# 15th International Conference on Nuclear Data for Science and Technology (ND2022)

US/Pacific
Gather.Town

#### Gather.Town

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Description

ND2022 is the latest in a series of conferences held every three years since 1978, most recently in Bruges, Belgium (ND2016) and Beijing, China (ND2019). This conference brings together international experts involved in generating and using nuclear data for a week of presentations and in-depth discussion.

ND2022 is being organized by Lawrence Livermore National Laboratory. The conference was planned for July 24-29, 2022 in Sacramento, California, but due to ongoing uncertainty related to the Covid19 pandemic it has been converted to a virtual event.

Thanks all who submitted abstracts!  Abstract submission has closed, but registration for ND2022 is still open! We invite everyone interested in attending to complete the registration. Presenters are also invited to review the Timetable and Contribution List using links to the left to check the time of their contribution(s).

For participants currently attending the meeting, we have put together a list of frequently asked questions that may be helpful in case you are running into problems.

Conference topics include:

• Nuclear reaction measurements,
• Nuclear mass, structure and decay measurements,
• Theoretical nuclear physics,
• Special focus on fission including theory, measurements and modeling,
• Development of new experimental capabilities and facilities,
• Generating, improving and validating evaluated nuclear data libraries,
• Processing nuclear data to prepare evaluated libraries for use in application codes,
• Improving data formats, tools and quality assurance,
• Uncertainty quantification and covariances,
• Applying machine learning to the nuclear data pipeline, and
• Applications of nuclear data to fields ranging from reactor design to medical isotope production to nuclear non-proliferation to space exploration and stellar nucleosynthesis.
ND2022 Organizers
• Thursday, 21 July
• 06:00 07:00
Chair / host training Sacramento ()

### Sacramento

• 06:00
Session chair training 1h
• 07:00 10:00
Speaker ready room American River ()

### American River

Time reserved for presenters to test out their audio / visual connections, screen sharing, etc.

• Friday, 22 July
• 09:00 16:00
Special Session on Nuclear Science Essentials: Introduction to Nuclear Data Sacramento ()

### Sacramento

• 09:00
Nuclear Data Pipeline 1h

Chair: Ali Dreyfuss

Speaker: Jo Ressler (LLNL)
• 10:00
Break 5m
• 10:05
Accelerator-based experiments for nuclear data and applications 1h

Chair: Erin Good

Speaker: Richard Hughes (LLNL)
• 11:05
Break 15m
• 11:20
Nuclear Reactions & Applications 1h

Chair: Amy Lovell

Speaker: Brett Carlson
• 12:20
Lunch 1h
• 13:20
Basics of Nuclear Fission 1h

Chair: Marc Verriere

Speaker: Ramona Vogt (LLNL, UC Davis)
• 14:20
Break 10m
• 14:30
Machine Learning for Nuclear Data 1h

Chair: Robert Casperson

Speaker: Denise Neudecker
• 15:30
Open discussion 30m
• Monday, 25 July
• 06:00 07:15
Plenary Session: I Sacramento ()

### Sacramento

Welcome and plenary talks

Convener: Caleb Mattoon (LLNL)
• 06:00
Speaker: Kim Budil (LLNL)
• 06:15
Nuclear Science: Meeting the Challenges of the Future 30m
Speaker: William D. Magwood (NEA Director General)
• 07:15 07:30
Break 15m
• 07:30 09:00
Plenary Session: II Sacramento ()

### Sacramento

Welcome and plenary talks

Convener: Ramona Vogt
• 07:30
Overview of Recent Advances in Experimental Capabilities 30m
Speaker: Alexandra Gade (FRIB)
• 08:00
Recent Advances in Nuclear Theory and Astrophysics 30m
Speaker: Karl-Heinz Langanke (GSI)
• 08:30
Facilitating Rapid Provenance Assessment using Intentional Nuclear Forensics Approaches 30m
Speaker: Naomi Marks (LLNL)
• 09:00 09:30
Break 30m
• 09:30 11:00
Fission: I American River ()

### American River

Convener: Amy Lovell (LANL)
• 09:30
New developments in fission studies within the time-dependent density functional theory framework 24m

LA-UR-21-28456

We have extended significantly our description of fission by examining a larger set of observables and more nuclei. We extract neutron and proton numbers of fission fragments, their spins and fission fragments relative orbital angular momentum and their correlations, investigate neutrons emitted at or shortly after scission, excitation energy sharing mechanism, total kinetic energy of fission fragments. So far, we have performed simulations with two independent nuclear energy density functionals on a number of uranium, plutonium, fermium, mercury isotopes, and spontaneous fission of 252Cf and we will extend these studies to a number of radon, polonium isotopes in the coming year (2022) and have results ready to present by the time of the meeting.

Speaker: Aurel Bulgac
• 09:54
Influence of fission fragment distributions on prompt emission quantities 12m

The influence of fragment distributions (experimental or theoretical) on the prompt emission results deserves a special investigation, as a part of methods and procedures used for the validation of fission fragment distributions. Such investigations are based on the primary model results (which consist of matrices of different prompt emission quantities as a function of fragment A, Z and its kinetic energy) already validated by their good description of existing experimental data. These validated matrices are then averaged over the studied fission fragment distributions. This paper includes the investigation of the influence of 4 fragment distributions Y(A,TKE) (three experimental and one theoretical) of 235U(nth,f) and 3 distributions (2 experimental and one theoretical) of 252Cf(SF) on prompt emission model results (as single distributions, e.g. prompt neutron multiplicity ν(A), ν(TKE) and centre-of-mass energy <ε>(A), <ε>(TKE) etc. and total average values, e.g. <ν>, <ε>, prompt neutron spectra, prompt γ-ray energy etc.).
The anti-correlation between the single fragment distribution TKE(A) and the distribution of total excitation energy of fully accelerated fragments TXE(A) reflected in the distribution of prompt neutron multiplicity of fragment pair νpair(A) is envisaged. The important role played on the distributions of different prompt emission quantities as a function of A by both the fission fragment distribution and the energy partition in fission is highlighted. Other correlations related to Y(TKE) and prompt emission quantities as a function of TKE as well as the role played by Y(A,TKE) on different total average prompt mission quantities are discussed, too.

Speaker: Anabella Tudora
• 10:06
Fission yield model for astrophysical use based on the four-dimensional Langevin model 12m

Nuclear fission is expected to play an essential role in the r-process nucleosynthesis via the fission recycling process. However, the impact of the fission recycling process on the r-process is still unclear due to the ambiguity of the fission yields for nuclei far from the beta-stability line and superheavy nuclei.

Our four-dimensional Langevin model can reproduce experimental nuclear fission properties, such as fission fragment mass distributions and total kinetic energies in a systematical manner including their sudden change due to shell structure. Then we developed a phenomenological yield model for nuclei Z=92 to 122, ranging from neutron deficient to very neutron-rich ones, by fitting the fission fragment mass yield obtained by the Langevin model with five Gaussians. We use the charge distributions for each isobar based on a parametrization which we have evaluated based on thousands of experimental data of actinides, to obtain the fission yields in the form of (Z, N) distribution.

What is unique here is that sudden change of the mass distribution from 2-peak to 1-peak structure for the region of Fm-256, and appearance of 5-peak structure for superheavy nuclei are predicted by the 5-D Langevin calculation, and these features are taken into account.

Speaker: Chikako Ishizuka
• 10:18
Langevin approach to fission dynamics with Cassini shape parameterization 12m

Fission plays an important role in the study of nuclear data in the actinide region. The dynamical approach to fission using the multi-dimensional Langevin equation has been widely accepted as a practical method. In this approach, nuclear shapes are expressed with a few shape parameters and the fission process is described as their time evolution by the Langevin equation. The shape parameters determine the model space and the calculation results depend critically on them.

The two-center shell model (TSCM) is a widely used model, which has five deformation degrees of freedom: elongation, asymmetry of fragment mass, neck radius, and deformation of right and left fragments [1]. In many cases, the neck degree of freedom is not treated as a dynamical variable. TSCM has been successfully applied in the four-dimensional Langevin calculation and provides good agreement with the experimental data such as fragment mass and kinetic energy distributions [2]. One drawback of TSCM is that only quadrupole deformations of the fragments are taken into account. On the other hand, Cassini shape parameterization can describe the various shapes of deformed nuclei by modifying the original Cassini oval with the Legendre expansion [3]. The coefficients $\alpha_i$ of the Legendre polynomial are the shape parameters. It has been applied in the static approach using the microscopic-macroscopic method to calculate the potential energy of deformation and provided the fragment mass and kinetic energy distributions in the actinide and super-heavy nuclear regions [4, 5]. These studies show that we can select five parameters $\alpha$, $\alpha_1$, $\alpha_3$, $\alpha_4$, and $\alpha_6$ among many Cassini parameters, corresponding to elongation, asymmetry of fragment mass, asymmetry of fragment shape, quadrupole deformation of fragments, and octupole deformation of fragments, respectively. However, its application to the dynamical treatment of fission was not realized.

In this study, we apply the Cassini shape parameterization to the dynamical calculation of fission using the multi-dimensional Langevin equation. We use the three- and the four-dimensional model space to calculate the fragment mass and kinetic energy distributions. As an example, we show the results for 14 MeV neutron-induced fission $^{235}$U($n$, $f$) with three- and four-dimensional calculation. As it is well known, the fragment mass distribution is asymmetric with the mass number of the heavy fragment at the peak position being 136 [6]. It is considered to reflect the magic numbers ($Z$, $N$) = (50, 82). Figure 1 shows the calculated fragment mass distributions for two cases: three-dimensional case with $\{\alpha,\, \alpha_1,\, \alpha_4\}$ and four-dimensional case with $\{\alpha,\, \alpha_1,\, \alpha_3,\, \alpha_4\}$. By including the shape asymmetric degree of freedom $\alpha_3$, the peak position moves from 95+141 to 100+136, which gives a better reproduction to the experimental data. The corresponding shapes at scission are shown at the top of Fig. 1. In the three-dimensional calculation, the shape is given by $\{\alpha,\, \alpha_1,\, \alpha_4\} =\{1.0,\,0.076,\,-0.036\}$, and in the four-dimensional calculation, it is given by $\{\alpha,\, \alpha_1,\, \alpha_3,\, \alpha_4\} =\{1.0,\,0.015,\,-0.091,\,-0.013\}$. It should be noted that the mass asymmetry is determined by the combination of $\alpha_1$ and $\alpha_3$. In the former case, two fragments have the same deformation with a slightly prolate shape. In the latter case, the heavy fragment is almost a sphere and the light one is elongated. It is important to consider not only the mass asymmetry but also the shape asymmetry to study the asymmetric fission as $^{236}$U. We examined the fission of Fm isotope as another example of actinide nuclei. For this element, it is known that the mass distribution in spontaneous fission changes drastically when we change the number of neutrons, i.e., it is asymmetric for $N$ = 156 and is symmetric for $N$ = 158 [7]. In the Langevin approach, we chose $\alpha_6$ as the fourth degree of freedom, by which we can describe the super-long and the super-short configurations in the symmetric fission. We performed the Langevin calculation at very low excitation energy to simulate the spontaneous fission. It has been confirmed that the transition from the asymmetric to the symmetric fission occurs at the correct neutron number when we include $\alpha_6$.

As a next step, we plan to extend the Langevin calculation to include five Cassini parameters $\{\alpha,\, \alpha_1,\, \alpha_3,\, \alpha_4,\, \alpha_6\}$. We will apply the five-dimensional calculation in the actinide region and compare the result with that of four-dimensional cases. Moreover, we will extend the system to super-heavy nuclei. It is suggested from the static approach that five Cassini parameters are necessary to describe the strongly deformed fragment that may appear in the fission of super-heavy nuclei [5]. It is important to confirm the mass and the shape of the fragment by the dynamical calculation using the Langevin equation with five Cassini parameters.

Reference
[1] J. Maruhn and W. Greiner, Z. Phys. 251, 431 (1972).
[2] M. D. Usang et al., Phys. Rev C 96, 064617 (2017).
[3] V. V. Pashkevich, Nucl. Phys. A 169, 275 (1971).
[4] N. Carjan et al., Nucl. Phys. A 942, 97 (2015).
[5] N. Carjan, F. A. Ivanyuk and Yu. Ts. Oganessian, Phys. Rev. C 99, 064606 (2019).
[6] K. Shibata et al., J. Nucl. Sci. Technol. 48(1), 1-30 (2011).
[7] D. C. Hoffman et al., Phys. Rev. C 21, 972 (1980).

• 10:30
Microscopic simulation of symmetric boost fission with antisymmetrized molecular dynamics 12m

We present our first results for a microscopic simulation of symmetric boost fission in terms of the antisymmetrized molecular dynamics (AMD), which incorporates the mean-field effects (as the TDDFT) via SLy4 effective interaction as well as stochastic 2-nucleon collisions similar to the cascade model. Due to these features, AMD is capable of obtaining not only the “expectation value” or “mean trajectory”, but “distributions” or “fluctuation” of observables can be obtained. We prepare the ground states of fissioning nuclei by the frictional cooling, which yielded the nuclear mass to accuracy of 50 keV/u. Then, each system is boosted symmetrically to split into 2 fragments. We calculated the total kinetic energy (TKE) and spins of the fission fragments. Dependence of the TKE on Z2/A1/3 of the fissioning system was found to be almost linearly increasing. We have also obtained orbital angular momenta of each fragment, their mutual orientation and their orientation with respect to the linear momenta. From this analysis, we can elucidate origin of the spins of fission fragments, either it is by the bending, wriggling, twisting or tilting motion. Furthermore, we have observed several ternary fission events, emitting tritons or 4He from the neck region, and average energy and angles of these ternary particles with respect to the fission axis were found to be in accord with experimental data.

Speaker: Jingde Chen
• 10:42
Modeling the angular momentum removal from fission fragments 12m

Fission fragment properties are set, to a large extent, during the scission process. It is generally assumed that the fully accelerated fission fragments can be treated as compound nuclei, and their de-excitation towards a long-lived state via prompt neutron and gamma emissions can be treated within a statistical model like the Hauser-Feshbach approach to nuclear reactions. Accurate modeling of prompt fission neutrons, and to a lesser extent prompt fission gamma-rays, is important, as measurements cannot detect fission fragments before neutron emission, and have to correct for the emitted particles to extract information about the scission process. One often makes the assumption that because most of the prompt fission neutrons are emitted with energies of about 1 MeV in the center-of-mass frame of the fragment, they do not carry much angular momentum when emitted from fission fragments. Using our state-of-the-art code CGMF, which treats the fission fragments as compound nuclei, we show that a significant number of neutrons are emitted with angular momentum of 3/2 and higher. As a consequence of quantum effects, the p-wave strength function peaks in the 90 – 100 mass region (the Ramsauer effect), emphasizing the importance of higher partial waves for the light fission fragments even at low neutron energies. When all of the contributions to the angular momentum removal from neutrons and statistical gamma rays (emitted before electromagnetic transitions between discrete levels) are counted, we find that on average the statistical decays remove about 4-5 units of angular momentum, with comparable standard deviation. This finding can be at odds with some other statistical models and assumptions made in the analysis of experimental data, and blurs the connection with the scission process, making difficult a direct connection between observables before and after neutron emission. We show that we can reproduce the shape of the average fission fragment spins as a function of their mass, as recently measured by Wilson et al., with somewhat different assumptions regarding the initial spin of the fission fragment, as illustrated in Fig. 1. We also show the impact of the known nuclear structure information, in particular isomeric states, illustrated in Fig. 1 by the difference between the CGMF results marked by green circles and purple pentagons. Finally, because different models disagree on details of the angular momenta of neutrons and statistical gamma rays, one should be aware of the possible model dependence when interpretation for some measurements is given.

LA-UR-21-28347

Speaker: Ionel Stetcu
• 09:30 11:00
Libraries: I Delta King ()

### Delta King

Convener: Osamu Iwamoto (JAEA)
• 09:30
The TENDL library: progress, success, and lessons learned 24m

The TALYS Evaluated Nuclear Data Library (TENDL) had 11 releases since 2008. A
number of achievements has been realized, translated in a growing user group, as well
as a high interaction with other library projects. The backbone of this achievement
is simple and robust: completeness, quality and reproducibility. In this presentation,
the strong points of TENDL will be outlined, as well as the existing weaknesses
with regards to the current nuclear data landscape. Finally, future improvements
will be discussed, based on the user needs, as well as realities from the nuclear data
communities.

Speaker: Dr Arjan Koning (IAEA)
• 09:54
ACEMAKER: A processing system for producing ACE-formatted files for Monte Carlo calculations 12m

Since 2016 the Nuclear Data Section of the International Atomic Energy Agency (NDS/IAEA) has encouraged the development of new codes for processing evaluated nuclear data for applications. The main motivation has been the need for improving processing methods and avoiding common failure modes. In the frame of this effort, the NDS/IAEA has supported the development of the processing system ACEMAKER, which takes advantage of the system of codes PREPRO and complements it with new modules for producing ACE-formatted files for Monte Carlo calculations. The new modules can process any correlated energy-angle distributions, gamma-production data, gamma-uncorrelated energy-angle distributions, thermal scattering laws and dosimetry files for producing the appropriated ACE-formatted file for Monte Carlo simulations. This paper describes the main features of the processing system ACEMAKER, its verification and validation process, and discuss planned developments.

1. List item
Speaker: Mr Daniel Lopez Aldama (IAEA/Consultant)
• 10:06
GIDI+: a GNDS 2.0 suite of C++ APIs to access nuclear and atomic data for use in neutronic transport codes. 12m

Transport codes used to simulate reactors, critical assemblies, medical diagnostics, etc. need access to nuclear and atomic data (e.g., data from ENDF/B-VIII.0 and EPICS2017). The Generalized Nuclear Data Structure (GNDS) supports storing evaluated and processed nuclear and atomic data - processed data are data in a form more suitable for transport codes (e.g., cross sections reconstructed from resonance parameters, heated cross sections, multi-grouped data). The specification for version 2.0 of the GNDS standard has recently been finalized. In this talk we will describe GIDI+ which is a suite of C++ APIs for reading GNDS 2.0 data, performing multi-group operations and sampling the data for Monte Carlo transport.

GIDI+ is comprised of the following C++ APIs:

• PoPI: Properties of Particles Interface
• Supports reading and accessing GNDS PoPs data. PoPs data describe properties of particles including a particle’s mass, spin, halflife and decay particles.

• GIDI: General Interaction Data Interface

• Supports reading and accessing data in a GNDS reactionSuite. A reactionSuite describes the nuclear or atomic reactions for a projectile hitting a target (e.g., “n + O16”).
• Supports reading GNDS map files. A map file links many reactionSuites together to form a nuclear and atomic data library.
• Provides functions for finding and reading in a reactionSuite for a specified projectile and target (e.g., “n + Th232”) found in a map file.
• Provides functions for accessing multi-group data in a reactionSuite which includes performing operations on the multi-group data (e.g., summing, collapsing and transport correcting with the Pendlebury/Underhill formalism).
• Allows one to mark some reactions as disabled in a reactionSuite instance. Disables reactions are skipped by multi-group routines and MCGIDI. For example, one can disable the (n,3n) reaction. With this reaction disabled, a subsequent call to the total multi-group cross section will not include the (n,3n) cross section.
• Support specifying modifications to reaction cross sections to perform sensitivity studies.

• MCGIDI: Monte Carlo GIDI

• Loads data from GIDI classes into classes that are better suited for Monte Carlo lookup, sampling and require less memory.
• Provides functions to evaluate the total cross section or the cross section for a reaction.
• Provides functions for sampling a reaction from the reactions in a reactionSuite.
• Provides functions for sampling outgoing particle data for a reaction.
• Supports continuous energy and multi-group cross section data.
• Supports unresolved region probability tables.
• Provides serializing a reactionSuite’s data which can then be broadcasted to MPI processes and GPUs.
• Supports shared memory to reduce footprint.

Currently, GIDI+ supports transporting n, p, d, t, h, $\alpha$ and $\gamma$. For $\gamma$ both photo-atomic and photo-nuclear data are supported. For neutrons, Thermal Neutron Scattering Law data are supported. We are currently adding support for electron transport and will describe its status at the meeting.

GIDI+ is freely available under the open-source MIT license and can be downloaded from https://github.com/LLNL/gidiplus.

Prepared by LLNL under Contract DE-AC52-07NA27344.

LLNL-ABS-827368

Speaker: Bret Beck (LLNL)
• 10:18
Current status of the GALILEE-1 processing code 12m

GALILEE-1 is the new verification and processing system for evaluated data, developed at CEA.
We have already presented in the past the reconstruction and Doppler broadening of cross-sections implemented in the GTREND module of GALILEE-1. The results obtained are of very good quality and we can explain the discrepancies observed with respect to other processing codes such as NJOY and PREPRO. This intensive comparison phase has increased the reliability of the GTREND module. GTREND is already able to handle the R-Matrix Limited format (LRF=7) with a large number of channels in the Resolved Resonance Region (RRR).
The treatment of the Unsolved Resonance Range is a more complex problem because it is difficult to obtain an experimental or theoretical reference to evaluate the different results of the various processing codes. We have chosen to implement two approaches in order to produce usable data for the deterministic and Monte Carlo transport codes, APOLLO3® and TRIPOLI-4® respectively, developed at CEA. These transport codes use URR cross-section data in the form of multi-group probability tables on an energy mesh chosen by the user. This job was previously done by the CALENDF code. The description of these probability tables is different from that of the NJOY PURR module, which produces pointwise probability tables on a coarse energy mesh using an interpolation method.
We will thus present a method that allows us to obtain a random pointwise representation over the whole energy range of URR, and a method similar to that of NJOY with a calculation on some energy points. Both approaches yield multi-group probability tables. Simulations of mono-atomic configurations and criticality benchmark configurations will be studied.

A new feature of the GTREND processing module is the possibility of processing thermal scattering data in a similar way as the THERMR module of NJOY. Some comparisons between GTREND and NJOY will be performed concerning the incoherent inelastic cross-section, the transfer probability density function and the angular distribution. The new capabilities of GTREND will be shown using an optimal processing for H1 bound in ZrH.

Speaker: Dr Cédric Jouanne (CEA Saclay)
• 10:30
Getting NJOY ready for ENDF/B-VIII.1 12m

Each release of the ENDF/B nuclear data library [1] pushes the limits of the processing codes through the introduction of new data and new data formats. Processing codes must be adapted to ensure that they are capable of using these new formats, or at the very least understand them. NJOY [2], the nuclear data processing code developed at Los Alamos National Laboratory is no exception to this. Since NJOY2016 is still the main production code for nuclear data libraries at LANL, we will continue to make the necessary changes in this version of NJOY to ensure full compatibility with the ENDF/B-VIII.1 nuclear data library. Where relevant, the modern NJOY21 components will also be updated.

The addition of the mixed elastic scattering option (LTHR = 2) for elastic thermal scattering (found in MF7 MT2) is one of the most anticipated changes for the ENDF/B-VIII.1 nuclear data library since it lifts the limitation of using either coherent or incoherent elastic scattering for moderators. For NJOY2016, this new format has required changes in various modules, the most important ones being THERMR (which processes the thermal scattering data) and ACER (which formats the data into an ACE file for use in application codes like the MCNP® code [3]). Figure 1 gives an example of what this mixed mode data looks like for D in 7LiD after processing through ACER in NJOY2016.65 which provides this new capability.

Because the coherent and incoherent data needs to be treated separately by MCNP6, the mixed mode scattering also necessitated an update to the ACE format itself by adding an additional elastic scattering data block [4], and a subsequent modification to MCNP so that the mixed mode elastic data can be used during a Monte Carlo simulation. Experimental support for this new format is available in MCNP, version 6.3.0 [5].

A second ENDF format change that will impact processing is the addition of R-matrix background elements for resonance channels in the R-matrix limited format (LRF = 7). Contrary to the mixed mode elastic scattering, the required changes to NJOY will be mainly limited to RECONR and NJOY21’s resonance reconstruction component.

In addition to these format changes, some processing limitations that were identified in the past will also have to be lifted to process ENDF/B-VIII.1 evaluations. Examples here would be MF34 covariance processing with multiple subsections in ERRORR or photonuclear data processing (in this case allowing for ACELAW=61 in the secondary photons).

In this paper we will give an overview of the most important changes that were made to NJOY2016 for processing ENDF/B-VIII.1 nuclear data, and where possible we will provide validation results that illustrate the impact that these changes will have for nuclear data users.

References
[1] D. Brown et al., “ENDF/B-VIII.0: The 8th Major Release of the Nuclear Reaction Data Library with CIELO-project Cross Sections, New Standards and Thermal Scattering Data,” Nuclear Data Sheets, 148, 2018, p. 1-142
[2] R. E. MacFarlane et al., “The NJOY Nuclear Data Processing System, Version 2016,” Los Alamos National Laboratory, LANL Technical Report LA-UR-20093
[3] J. A. Kulesza et al., “MCNP Code Version 6.3.0 Theory & User Manual,” Los Alamos National Laboratory, LANL Technical Report, to be published (2021)
[4] J. L. Conlin, P. Romano, W. Haeck, “A Compact ENDF (ACE) Format Specification,” Los Alamos National Laboratory, LANL Technical Report LA-UR-19-29016
[5] J. A. Kulesza, C. J. Josey, S. R. Bolding, J. C. Armstrong, J. E. Sweezy, J. S. Bull, M. E. Rising, E. J. Pearson, “MCNP® Code Version 6.3.0 Release Notes,” Los Alamos National Laboratory, LANL Technical Report, to be published (2021)

Speaker: Wim Haeck (LANL)
• 10:42
The current status of nuclear data processing code NECP-Atlas 12m

In the previous version of nuclear data processing code NECP-Atlas reported in the last ND-2019, the code could process the evaluated nuclear data into application libraries of typical formats with the conventional processing methods. NECP-Atlas is under continuous development and verification during the past years. Several new capabilities have been developed in the code.
(1) The code is extended to process the GNDS formatted evaluated nuclear data library. A generalized derived class is designed to be compatible with ENDF-6 and GNDS in nuclear data processing code NECP-Atlas, and the GNDS toolkit is developed in NECP-Atlas to parse and convert GNDS
(2) The module to calculate the thermal scattering law is updated to improve the accuracy. A method named anisotropic displacement parameters (ADP) is implemented to improve the coherent elastic scattering data. The 1-phonon correction method is implemented to improve the inelastic elastic scattering data.
(3) A new module named RESK is developed to generate Doppler-broadened resonance elastic scattering kernel in PENDF format. And then the data is processed into S(α, β) table used by Monte Carlo codes, and multi-group cross sections and scattering matrix used by deterministic based codes.
(4) The capability to generate photon data is newly developed. The code not only can provide the prompt parts of neutron-induced gamma production, but it can also generate the decayed gamma which is produced by decay of fission products.
(5) The capability to calculate neutron, electron, photon displacement cross section is newly developed. For the displacement damage model, apart from traditional NRT model, state-of-the-art ARC model is applied. Accuracy MOTT cross section is used as the electron different scattering cross section to improve the accuracy.

Speaker: Dr Tiejun Zu (Xi'an Jiaotong University)
• 09:30 11:00
Machine Learning: I Sutter's Fort ()

### Sutter's Fort

Convener: David Regnier (CEA)
• 09:30
Deep Learning applied to Capture Cross Section Data Analyisis 24m

The data analysis of neutron cross section measurements is a critical step for generating high quality and reliable nuclear databases [1-4] to be used in the fields of nuclear technologies, radioprotection, nuclear medicine or astrophysics, among others. Artificial intelligence techniques, and in particular deep learning, have proven to be a very useful tool for data recognition and analysis. We have investigated the suitability of deep learning and neural network techniques as a complementary tool for the analysis of neutron capture cross section measurements.

Some of the neutron capture cross section measurements performed at the n_TOF facility at CERN [5] use the so-called Total Absorption Calorimeter (TAC) [6]. This device is made of 40 BaF$_2$ crystals capable of detecting in coincidence the γ-rays emitted after the capture reactions. In a standard analysis, conditions (cuts) on the total energy deposition in the TAC and on the event multiplicity (number of crystals with signals in coincidence) serve to discriminate capture from background events.

We have trained a neural network with the list of energies deposited in the different crystals with Monte Carlo simulated capture and background events (due to scattered neutrons and competing fission reactions) and evaluated its capability of selecting the γ-ray cascades in a virtual capture experiment. We have compared different strategies and combinations of the neural network topologies and standard analysis methods in order to obtain at the same time large signal-to-background ratios and large capture detection efficiencies. The use of a neural network with 5 hidden fully connected layers and some conditions on the total energy and multiplicity of the crystals has achieved up to 5 times better signal-to-background ratio and ~15% more remaining capture events in the resulting dataset than with the standard analysis method.

In this conference, we will present these results and describe the deep learning methods used.

## References

[1] OTUKA, N., et al., Towards a more complete and accurate experimental nuclear reaction data library (EXFOR): international collaboration between nuclear reaction data centres (NRDC), Nucl. Data Sheets 120, 272 (2014).
[2] BROWN, D.A., The 8th major release of the nuclear reaction data library with CIELO-project cross sections, new standards and thermal scattering data, Nucl. Data Sheets, 148 (2028) 1-142.
[3] CABELLOS, O., et al., Benchmarking and validation activities within JEFF project, EPJ Web Conf., 146 (2017), Article 06004.
[4] SHIBATA, K., et al., JENDL-4.0: A new library for nuclear science and engineering, J. Nucl. Sci. Technol., 48 1 (2011) 1.
[5] GUERRERO, C., et al. Performance of the neutron time-of-flight facility n_TOF at CERN. Eur. Phys. J. A 49, 27 (2013).
[6] GUERRERO, C., et al., The n_TOF total absorption calorimeter for neutron capture measurements at CERN, Nucl. Instrum. Methods A, 608 3 (2009) 424-433.
[7] GARSONS, G.D., Interpreting neural networks connection weights, AI Expert 6 (1991) 47-51.

Speaker: Adrian Sanchez Caballero
• 09:54
Quantum computing calculations for nuclear structure and nuclear data 12m

Quantum computing opens up new possibilities for the simulation of many-body nuclear systems.

As the number of particles in a many-body system increases, the size of the associated Hamiltonian increases exponentially. This presents a challenge when performing calculations on large systems when using classical computing methods. By using a quantum computer, one may be able to overcome this difficulty thanks to the exponential way the Hilbert space of a quantum computer grows with the number of quantum bits (qubits).

Our aim is to develop quantum computing algorithms which can reproduce and predict nuclear structure such as level schemes and level densities. As a sample Hamiltonian, we use the Lipkin-Meshkov-Glick model [1] and have further tested our methods against recent quantum computing calculations of the deuteron binding energy [2].

We use an efficient encoding of the Hamiltonian onto many-qubit systems, and have developed an algorithm allowing the full excitation spectrum of a nucleus to be determined with a variational algorithm capable of implementation on today's quantum computers with a limited number of qubits. Our algorithm uses the variance of the Hamiltonian, $\langle H^2\rangle-\langle H\rangle^2$, as a cost function for the widely-used variational quantum eigensolver (VQE).

In our presentation, we will show the motivations for the work, the simplified encoding method, and example results obtained using variational quantum algorithms to produce nuclear excited states spectra, as well as commenting on prospects for future applications.

[1] H. J. Lipkin, N. Meshkov, and A. J. Glick, Nulcear Physics 62, 188 (1965)

[2] E. F. Dumitrscu \textit{et al.}, Physical Review Letters 120, 210501 (2018)

Speaker: Isaac Hobday
• 10:06
Improved calculations of Maxwellian-averaged \textit{n}-capture cross sections using machine learning for intentional forensics and astrophysics 12m

The National Nuclear Data Center (NNDC) maintains a repository of physical and theoretical quantities in the Evaluated Nuclear Data Format/B library (ENDF/B), currently on the 8th release (ENDF/B-VIII.0). Many fields rely on the accuracy of these quantities and their uncertainties, from real world applications (such as nuclear power) to more theoretical ones. This work is part of a larger effort by the NNDC to supply nuclear data for Intentional Forensics (IF) applications. The aim of IF is to develop a system for tagging commercially produced reactor fuel to alleviate a great security concern to the United States and other nation states and organizations. Apparently unrelated questions concerning the nature of the \textit{r}-process, including abundances, reaction rates, and astrophysical sites, remain enigmatic and an extremely active field of study. Both of these pursuits rely heavily on neutron capture in stable (but understudied) and unstable nuclei. The NNDC is analyzing and expanding the available data for these efforts, based on the ENDF/B-VIII.0 neutron sublibrary, containing 557 species of long lived isotopes and isomers relevant to neutron applications.

For both IF and many astrophysical applications the most relevant nuclear quantities are unmeasured (or under measured!), and will remain so for some time. Theoretical estimates are employed with varying levels of success due to assumptions that, while necessary, might not be supported by the physics. Current methods attempt to connect the cross section to physical quantities such as mass, level density, and neutron separation energy in order to identify patterns. These often focus on a single nuclear quantity or bulk property (such as mass model), which heavily weights that quantity's importance by default. As such, new methods are always of interest. This work is to first produce a catalogue of Maxwellian Averaged Cross Sections (MACS) for neutron-capture for the isotopes in the neutron sublibrary and spanning from thermal energies to the neutron activation range. From this library we will attempt to improve estimates of neutron capture Maxwellian Averaged Cross Sections, especially off-stability where few or no measurements exist. As the neutron fluence in many applications can be treated as a weighted sum of Maxwellian spectra, inverse problem theory allows us develop a regression model for the temperature dependence of the MACS. By applying the predictive capability of machine learning (ML) we can couple these estimates to additional physical quantities where known, to highlight or emphasize previously unknown relationships. These results will be provided to the IF group to inform their efforts. The results will also be implemented in new \textit{r}-process simulations, implementing the most up-to-date astrophysical models and nuclear network methods.

Speaker: Amber Lauer-Coles
• 10:18
Automated Nuclear Data Evaluation Tool; Nonconvex Decomposition Method for Resonance Regression and Uncertainty Quantification 12m

Nuclear data evaluation is a process that takes a long time partly due to resonance identification having to be performed manually by expert nuclear data evaluators. This procedure is laborious, time-consuming, and irreproducible from evaluator to evaluator. Additionally, it is well-documented that a systematic evaluation of the uncertainty on the evaluated cross section is unreliably low. Having these uncertainties that are “too small” results in additional work for the resonance evaluator to search for a non-systematic, artisanal way to make the estimated uncertainty be an appropriate size.
We offer an automated tool for the evaluation process to serve as a starting point for the evaluators by posing the resonance characterization problem as a mixed integer nonlinear program (MINLP). Since the number of resonances present is unknown, the model must be able to be determine the number of parameters to properly characterize the cross section curve as well as calculate the appropriate values for those parameters. Due to the size of the problem and nonconvexity of the parameterization, the optimization formulation is too difficult to solve as whole. We propose a novel method of decomposing the problem into small, overlapping windows that are now solvable, as well as method to stich the windows back together via parameter cardinality and value agreement routines to achieve the global solution. Lastly, a version of quantile regression is run on the windows to provide an uncertainty estimate on the parameter values. The results demonstrate the model being robust to finding the proper number of resonances, the appropriate values of the parameter values, and that the uncertainty estimation is directly reflective of the experimental conditions. Hence, this tool serves as a fast, accurate, and reproducible tool for nuclear data evaluators to have a starting place in their work.

Speaker: Noah Walton (University of Tennessee)
• 10:30
(WITHDRAWN) Automated grouping of spatially distributed detectors in neutron time-of-flight experiments based on multivariate similarity 12m

Nowadays, in neutron time of flight measurements, there are experimental setups in which many detectors record data during a single experiment. It is usually desirable to be able to sum several spectra in order to increase counting statistics, and therefore decrease uncertainties, for further analysis. A problem arises in time-of-flight experiments when the available spectra are acquired with a set of spatially distributed detectors, each forming a different source-sample-detector angle and at different sample-detector distances. Since these spectra record the neutron’s time of flight after scattering, and the neutron scattering depends on the Q vector, then these spectra are not arbitrarily summable. In this work, we propose an automated methodology for wisely adding spectra based on their multivariate similarity by means of machine learning techniques, such as k nearest neighbors combined with T-distributed Stochastic Neighbor Embedding (t-SNE). We exemplify it in the effective temperature determination of hydrogen in ethane and triphenylmethane samples by means of Deep Inelastic Neutron Scattering, measured at the VESUVIO spectrometer (ISIS facility, UK). The proposed methodology can be applied in other time-of-flight experiments, in which detectors located at different angles record complete spectra, and with this method their degree of compatibility can be determined.

Speaker: Jose I. Robledo
• 10:42
Scalable Risk-Informed Predictive Maintenance Strategy for Operating Nuclear Power Plants 12m

Over the years, the nuclear fleet has relied on labor-intensive, time-consuming preventive maintenance (PM) programs, driving up operation and maintenance (O&M) costs to achieve high capacity factors. The primary objective of the research presented in this paper is to develop scalable technologies deployable across plant assets and the nuclear fleet in order to achieve risk-informed predictive maintenance (PdM) strategies at commercial nuclear power plants (NPPs). A well-constructed risk-informed PdM approach for an identified plant asset was developed in this research, taking advantage of advancements in data analytics, machine learning (ML), artificial intelligence (AI), risk model, and visualization. These technologies would allow commercial NPPs to reliably transition from current labor-intensive PM programs to a technology-driven PdM program, eliminating unnecessary O&M costs.

The research and development approach (see Figure 1) presented in the paper is being developed as part of a collaborative research effort between Idaho National Laboratory and Public Service Enterprise Group (PSEG) Nuclear LLC. This paper presents the results of analyzing the heterogeneous data associated with the circulating water system (CWS) from both the Salem and Hope Creek NPP sites. Fault modes present in the data were identified based on logs and correlated with data to develop salient fault signatures associated with each fault mode. The fault signatures are used to develop diagnostic models using scalable predictive analytics and integrated with plant-level risk and economic models. The paper also outlines the development of a user-centric visualization application to ensure the right information is available to the right person, in the right format, and at the right time.

The research outcomes presented in this paper lay the foundation and provide a much-needed technical basis to start focusing on additional needs such as explainability and trustworthiness of ML- and AI-based technologies as part of future research.

Speaker: Vivek Agarwal
• 09:30 11:00
Measurements: I Folsom ()

### Folsom

Convener: Mark Chadwick (LANL)
• 09:30
Experimental activities and plans at the neutron time-of-flight facility n_TOF at CERN 24m

Based on an idea by Carlo Rubbia, the n_TOF facility has been operating during the last 20 years at CERN. n_TOF is a spallation neutron source, driven by the 20 GeV/c proton beam from the CERN PS accelerator. A massive Lead spallation target is feeding two experimental areas set at 185 meters (EAR1, horizonal with respect to the proton beam direction) and at 20 meters (EAR2, in the vertical direction) from the spallation source. Neutrons in a very wide energy range - from GeV down to sub-eV kinetic energy - are generated and selected by the time-of-flight technique, with the long flight paths ensuring the possibility of performing very high-resolution measurements.

Over the course of two decades, more than one hundred experiments have been performed by the n_TOF Collaboration, in the domain of nuclear data for advanced technologies (neutron capture, neutron induced fission and (n,cp) reactions for accelerator driven systems, Gen-IV and Th/Ufuel cycle), in nuclear astrophysics (synthesis of the heavy elements in stars, big bang nucleosynthesis, nuclear cosmo-chronology), and for basic nuclear science (nuclear structure and decay of highly excited compound states). Overall, measurements at n_TOF generated ~100 entries in the EXFOR database, covering over 90% of the data released by the n_TOF Collaboration.

During the planned shutdown of the CERN accelerator complex, between 2019 and 2021, the facility went through a substantial upgrade with a new target-moderator assembly, refurbishing of the neutron beam lines and experimental areas. An additional station (the NEAR Station) has been set up at approximately 2 m from the target-moderator assembly. The NEAR Station capabilities for performing material irradiation studies, neutron induced activation and new physics opportunities are presently explored.

An overview of the experimental activities performed at n_TOF will be presented, with a particular emphasis on the most recent results and planning for the future.

(*) Spokesperson of the n_TOF Collaboration

Speakers: Alberto Mengoni, The n_TOF Collaboration
• 09:54
Measurement of the $^{35}$Cl radiative neutron capture cross section at the n_TOF facility, CERN 12m

The $^{35}$Cl radiative capture rate is important in a number of applications. The long-lived radionuclide $^{36}$Cl is a by-product of the activation of $^{35}$Cl present in graphite moderated reactors both in the fuel cladding and as an impurity in nuclear graphite (< 2 ppm by mass). Reliable predictions of the amount of $^{36}$Cl present in the large volume of irradiated nuclear graphite waste relies on accurate reaction cross section data, essential for its safe disposal [1]. Moreover, in Boron Neutron Capture Therapy (BNCT), currently being considered more widely as a cancer therapy [2], accurate knowledge of the dose rate delivered both to tumours and the surrounding healthy tissue is imperative; chlorine has an important role in the dose rate, especially in brain tissue [3] where it represents 0.3% by mass. Simulations have indicated that around 11% of the total dose rate relevant to the neutron energy ranges used in BNCT (100 eV - 10 keV) comes from neutron capture on 35Cl. Finally, $^{35}$Cl is one of several ‘minor neutron poisons’ in the astrophysical s-process, reducing the efficiency of neutron recycling and is furthermore involved in the as yet unknown origin of $^{36}$S [4, 5]; accurate capture cross section data are important in determining the significance of $^{35}$Cl in the s-process and its impact on stellar reaction networks.
The reaction cross section has been measured twice via the time-of-flight method in the past [6, 7], for which the results of resonance analyses are discrepant by around 15%, and evaluations (ENDF/B-VIII.0, JEFF-3.3, JENDL-4.0) differ with respect to one another by around 10% in the resonance region. The recent AMS measurement of the 30 keV Maxwellian averaged cross section is inconsistent with the existing time-of-flight measurements [8].
Work has been performed to accurately measure the $^{35}$Cl$(n,\gamma$)$^{36}$Cl reaction cross section at the 185m beam-line at the neutron time-of-flight facility (n_TOF) at CERN, using a C6D6 total energy detection setup. We measured thirteen resonances up to a maximum energy of around 50 keV, limited to the strongest resonances by a prohibitive background. Preliminary results are in agreement with the results of Reference [6], and the Maxwellian averaged cross section extracted from the data is consistent with the recent AMS measurement [8]. Our results indicate that the ENDF/BVIII.0 and JEFF-3.3 evaluations (both based on the R-Matrix analysis of Sayer et al. [9]) underestimate the cross section by around 15%. The experimental procedure, analysis and results of this measurement shall be presented.
[1] R. Mills, Z. Riaz, A. Banford, Nuclear Data issues in the calculation of 14C and 36Cl in irradiated graphite, ENC 2012 Conference proceedings.
[2] International Atomic Energy Agency (IAEA), press release: https://www.iaea.org/newscenter/news/boron-neutron-capture-therapy-back-in-limelight-after-successful-trials
[3] R. F. Barth, M. G. H. Vicente, O. K. Harling et al., Current status of boron neutron capture therapy of high grade gliomas and recurrent head and neck cancer, Radiation Oncology 7, 146 (2012).
[4] H. Schatz, S. Jaag, G. Linker, R. Steininger, F. Käppeler, P. E. Koehler, S. M. Graff, and M. Wiescher, Phys. Rev. C 51, 379 (1995).  [5] R. Reifarth, K. Schwarz, and F. Käppeler, Astrophys. J. 528, 573 (2000).  [6] R. L. Macklin, Phys. Rev. C 29, 1996 (1984).  [7] K. H. Guber, R. O. Sayer, T. E. Valentine, L. C. Leal, R. R. Spencer, J. A. Harvey, P. E. Koehler, and T. Rauscher, Phys. Rev. C 65, 058801 (2002).  [8] S. Pavetich, A. Wallner, M. Martschini et al., Accelerator mass spectrometry measurement of the reaction $^{35}$Cl$(n,\gamma$)$^{36}$Cl at keV energies, Phys. Rev. C 99, 015801 (2019).
[9] R. O. Sayer, K. H. Guber, L. C. Leal et al., R-matrix analysis of Cl neutron cross sections up to 1.2 MeV, Physical Review C 73, 044603 (2006)

Speakers: Sam Bennett, The n_TOF Collaboration
• 10:06
New perspectives for neutron capture measurements in the upgraded CERN-n_TOF Facility 12m

Neutron capture cross-section measurements are of great interest for various nuclear data applications, such as the slow neutron capture (s-) process of nucleosynthesis in stars, innovative nuclear technology or medical applications. The neutron energy range of interest varies depending on the application and, hence, pulsed white neutron sources combined with the time-of-flight (TOF) technique are the best suited facilities for these measurements.

Since 2001, the high resolution neutron time-of-flight facility CERN-n_TOF-EAR1 [1] has provided neutron capture cross sections with an excellent energy resolution and broad energy range up to 1 MeV [2]. In 2014, the n_TOF Collaboration built a new vertical beam line, so-called n_TOF-EAR2 [3], with a flight path of only 20 m, approximately ten times shorter than the 185 m of n_TOF-EAR1. Given its high instantaneous flux [4], this new neutron beam line opened the door to challenging measurements of samples with high activity, available only in small quantities or with small cross sections [5, 6, 7].

The n_TOF facility has just undergone in 2021 a major upgrade with the installation of its third generation spallation target that has been designed to optimize the performance of the two n_TOF time-of-flight lines. This contribution will present the first results of reference capture measurements in the two beam lines of the upgraded n_TOF facility.

The performance and new possibilities for (n,Ɣ) measurements at n_TOF will be presented and compared with the currently most competitive time-of-flight facilities worldwide featuring white neutron beams. Several key aspects for capture measurements will be discussed, focusing on the maximum neutron energy limit, of relevance for astrophysics and fast reactor applications, the instantaneous neutron fluence, which determines the signal to background ratio in the case of radioactive samples, and the energy resolution. The latter is a key factor for both increasing the signal-to-background ratio and obtaining accurate Resonance Parameters [8]. In particular, the energy resolution has been clearly improved for the 20 m long vertical beam line with the new target design, according to our very preliminary results (see Fig. 1) while keeping the remarkably high resolution of the long beamline n_TOF-EAR1 [1].

FIGURE ATTACHED

Fig. 1.- Time-of-flight spectrum of 197Au(n,Ɣ) measured with C6D6 detectors at n_TOF-EAR2 (measuring position at 19.5 m) with the previous (2015) and the upgraded (2021) spallation target. The neutron energy between 200 and 300 eV is displayed.

Last, current experimental limitations for capture measurements at CERN n_TOF will be discussed together with some of the on-going detector R&D projects that will try to tackle them in the upcoming years [9].

[1] C. Guerrero et al., Performance of the neutron time-of-flight facility n_TOF at CERN, Eur. Phys. J. A 49, 27 (2013).
[2] G. Aerts et al., Neutron capture cross section of 232Th measured at the n_TOF facility at CERN in the unresolved resonance region up to 1 MeV, Phys. Rev. C 73, 054610 (2006).
[3] C. Weiss et al., The new vertical neutron beam line at the CERN n_TOF facility design and outlook on the performance, Nucl. Inst. and Methods A, 799, 90-98 (2015).
[4] M. Sabaté-Gilarte et al, High-accuracy determination of the neutron flux in the new
experimental area n_TOF-EAR2 at CERN, Eur. Phys. J. A 53, 10 (2017).
[5] M. Barbagallo et al., 7Be(n,α)4He Reaction and the Cosmological Lithium Problem: Measurement of the Cross Section in a Wide Energy Range at n_TOF at CERN, Phys. Rev. Lett. 117, 152701 (2016).
[6] V. Alcayne et al., Measurement of the 244Cm capture cross sections at both CERN n_TOF experimental areas, EPJ Web Conf., 239, 01034 (2020).
[7] Sabaté-Gilarte M. et al, The 33S(n,α)30Si cross section measurement at n_TOF-EAR2 (CERN): From 0.01 eV to the resonance region, EPJ Web Conf. 146, 08004 (2017).
[8] P. Koehler, Comparison of white neutron sources for nuclear astrophysics experiments using very small samples, Nucl. Inst. and Methods A 460, 352-361 (2001).
[9] V. Babiano-Suárez, J. Lerendegui-Marco, et al. Imaging neutron capture cross sections: i-TED proof-of-concept and future prospects based on Machine-Learning techniques. Eur. Phys. J. A 57, 197 (2021).

Speakers: Jorge Lerendegui, The n_TOF Collaboration
• 09:30 11:00
Medical Applications Capitol ()

### Capitol

Convener: Toni Kogler
• 09:30
Investigating the production of medical imaging radioisotopes using laser-accelerated protons. 24m

Speaker: Mr Juan Peñas
• 09:54
Determination of positron emission probability in the decay of 86Y 12m

In nuclear medicine, the positron emitting $^{86}$Y (14.7 h) is an emerging imaging isotope for use in combination with the $\beta$$^- emitting therapeutic radionuclide ^{90}Y (2.7 d), the pair being commonly called as a matched theranostic pair [1]. To produce the radionuclide ^{86}Y via the ^{86}Sr(p,n) reaction in a pure form, 96.4% enriched thin ^{86}SrCO_3 targets were irradiated with protons of energies 7 and 8 MeV at BC1710 cyclotron, Forschungszentrum Jülich (FZJ), Germany. We have determined the positron emission probability P_\beta$$^+$ in the $\epsilon$+$\beta$$^+ decay of ^{86}Y (14.7 h) by measuring the 511 keV annihilation \gamma-ray using an ORTEC HPGe detector. During counting, each irradiated ^{86}SrCO_3 sample was placed inside a circular groove (diameter 1 cm, depth 0.01 cm) of a Cu disk (diameter 3 cm, height 0.5 cm) and covered with another one of the same type and size. The Cu enclosure served to annihilate almost all \beta$$^+$ particles in $^{86}$Y decay, Q$^+$($^{86}$Y)=5.24$\pm$0.01 MeV, within its geometric volume. Several corrections were made including that for the counting geometry to deduce the final results. The electron capture probability P($\epsilon$+$\beta$$^+) has also been determined as an additional check, since %P(\epsilon+\beta$$^+$)=100, by measuring the K$_\alpha$ and K$_\beta$ X-rays of energies 14.1 and 15.8 keV, respectively, using a special ORTEC HPGe detector, with a 0.3 mm thick Be window, for low-energy $\gamma$-rays. The positron emission and electron capture probabilities are determined to be 27.6$\pm$1.4% and 72.2$\pm$4.0%, respectively. The normalization yields %P$_\beta$$^+=27.7\pm1.5 and %P\epsilon=72.3\pm1.5. The positron emission probability of this work is found to be in good agreement with the value 27.9\pm1.2% [2], deduced using the newly constructed \gamma-decay/level-scheme of ^{86}Sr and calculation of electron capture to positron decay ratio for each beta feeding levels, and is lower by 14% compared to the latest evaluated value of 32.5\pm2.0 [3], obtained using the prior \gamma-decay/level-scheme of ^{86}Sr and the same method of Ref. [2]. The deduced electron capture probability of the present work provided an additional confirmation for the measured positron emission probability of ^{86}Y. References: 1.F. Rösch, H. Herzog, and S. M. Qaim, Pharmaceuticals 10, UNSP article 56 (2017). 2.A. C. Gula, E. A. McCutchan, C. J. Lister, J. P. Greene, S. Zhu, P. A. Ellison, R. J. Nickles, M. P. Carpenter, S. V. Smith, and A. A. Sonzogni, Phys. Rev. C 102, 034316 (2020). 3.A. Negret and B. Singh, Nuclear Data Sheets 124, 1 (2015). Speakers: Dr M. S. Uddin (Institut für Neurowissenschaften und Medizin, INM-5:Nuklearchemie, Forschungszentrum Jülich, D-52425 Jülich, Germany; Tandem Accelerator Facilities, INST, Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh), Dr Shamsuzzoh Basunia (Nuclear Science Division, Lawrence Berkeley National Laboratory, Berkeley, California 94720, USA) • 10:06 Mesurement of the 160Gd(n, γ) cross section at n_TOF and its medical implications 12m 1. Istituto Nazionale di Fisica Nucleare, Sez. di Bari, Bari, Italy 2. Università degli studi di Bari, Dipartimento Interateneo di Fisica, Bari, Italy 3. Istituto Nazionale di Fisica Nucleare, Sez. di Perugia, Perugia, Italy 4. Istituto Nazionale di Astrofisica, Osservatorio Astronomico d’Abbruzzo, Teramo, Italy 5. Istituto Nazionale di Fisica Nucleare, Sez. di Bologna, Bologna, Italy 6. Università di Bologna, Dipartimento di Fisica e Astronomia, Bologna, Italy 7. European Organization for Nuclear Research, Geneva, Switzerland 8. ENEA Research Centre E. Clementel, Bologna, Italy The neutron capture reaction cross section of gadolinium isotopes play an important role in several fields of Physics, for instance in Nuclear Astrophysics for the understanding of the nucleosynthesis of heavy elements (beyond iron) in stars via the s- and r-processes [1] and in nuclear technology. Another important application of gadolinium is linked to the production of Terbium, that offers a set of clinically interesting isotopes for theranostics, characterized by complementary physical decay characteristics. In particular, the low-energy β- emitter Terbium-161 is very similar to Lutetium-177 in terms of half-life (6.89 d), β-energy and chemical properties. Being a significant emitter of conversion/Auger electrons, greater therapeutic effect can therefore be expected in comparison to Lu-177 [2, 3]. For this reason, in the last decade, the study of neutron capture reaction ^{160}Gd(n,γ)^{161}Gd and the subsequent β-decay in Terbium-161 is getting particular attention. As the nuclear data on the Gd-160 neutron capture reaction are quite scarce and inconsistent, a new measurement of the capture cross section of Gd-160 at the CERN neutron Time-Of-Flight facilty will provide high resolution, high-accuracy data on this important reaction of interest for Nuclear Astrophysics and Nuclear Medicine, in the energy range from thermal to tens of keV. In this talk, the preliminary results of the n_TOF measurement will be presented. References: [1] F. Käppeler, R. Gallino, S. Bisterzo, Wako Aoki, Rev. Mod. Phys. 83, 157 (2011) [2] C. Müller, K. Zhernosekov et al., Jour. of Nucl. Med. 53 (12) 1951 - 1959 [3] S. Lehenberger, C. Barkhausen et al., Nucl. Med. and Biol. 38 (2011) 917 - 924 Speakers: Dr Mario Mastromarco (INFN and UniBa, Bari, Italy), and the n_TOF Collaboration • 10:18 Measurement of the 176Yb(n,γ) cross-section and its application to nuclear medicine 12m Nuclear medicine has proven to be a much needed medical specialty in order to diagnose and treat several diseases, among them, cardiovascular diseases and cancer, the first and the second causes of mortality worldwide, respectively [1]. Several international agencies recommend to study of new routes and new facilities for producing radioisotopes with application to nuclear medicine as a complementary option to the conventional ones based on nuclear reactors or dedicated cyclotrons [2,3,4]. This has been specially pushed in the last years with the development and the availability of high-intensity accelerators and new installations because they allow the production of emergent or new radioisotopes. In addition, these new installations can provide quantities of various radioisotopes at regional level. CERN’s MEDICIS ISOLDE facility is an excellent example [5]. 177Lu is a versatile radioisotope used for therapy and diagnosis (theranostics) of cancer with good success in gastroenteropancreatic neuroendocrine tumours [6]. Currently, the use of 177Lu is under study for several other tumours with good results [7]. 177Lu is produced in a few nuclear reactors mainly by the neutron capture on 176Lu (direct route). However, it could be produced at high-intensity accelerator-based neutron facilities as IFMIF-DONES by means of the route 176Yb(n,γ) (indirect route) [8]. In fact, the produced 177Yb beta-decays to the 177Lu with a half-life of 1.9 h. Although the cross-section of the direct route is higher, several advantages in the indirect route have been pointed out: i) The specific activity is four times higher. About 100% of the theoretical specific activity can be achieved [9,10]. ii) The contaminants that remain in the final quantity of the material are much lower. iii) The undesirable 177mLu (t1/2≈160 d) is produced in the direct route, 0.05%, whereas in the indirect route is less than 10-5 % [11]. These properties have a direct impact on the quality of the diagnosis and the therapy. The higher specific activity allows a much better tumour uptake; thus, the dose delivered to the tumour for the same activity is much higher in case of the indirect route, and in addition, the quality of imaging of the tumour is much better [9,10]. The energy of the neutrons in accelerator-based neutron facilities is higher than in thermal reactors. Consequently, experimental data on the 176Yb(n,γ) cross-section in the eV and keV region are mandatory to accurately calculate the production of 177Lu. At present, there is no available experimental 176Yb(n,γ) cross-section data for thermal neutron energies up to 3 keV. In addition, the resonances have not been resolved in the range from 3 to 50 keV. An experimental campaign at the n_TOF facility provides data from the thermal to the resolved resonance region for the first time, resolving the largest resonances in the 176Yb(n,γ) cross-section. The γ-rays cascade, with a total energy of 5.24 MeV emitted after each capture reaction in 176Yb, have been detected using a set of four C6D6 detectors [12]. The Monte Carlo-based pulse height weighting technique (PWHT) have been in the analysis [13]. These data will be ready to be used in future evaluations of the 176Yb(n,γ) cross-section and for the calculation of 177Lu production in accelerator-based neutron sources as IFMIF-DONES. Results of the analysis will be shown with particular attention to the possible confirmation of the 1/v behaviour of between thermal resonance region, where there are not experimental data. Furthermore, this measurement will potentially clarify the existence of resonances which have been observed in transmission experiments but were not measured in previous capture measurements. References: [1] https://www.euro.who.int/en/health-topics/noncommunicable-diseases/cancer. [2] Nuclear Physics European Collaboration Committee: nuclear physics for medicine. ISBN: 978-2-36873-008-9. http://www.nupecc.org/pub/npmed2014.pdf. [3] European Nuclear Society, The Medical Isotope Crisis. http://www.euronuclear.org/1-information/news/medical-isotope-crisis.htm. [4] NuPEcc. Long Range Plan 2017 Perspectives in Nuclear Physics. https://www.esf.org/fileadmin/user_upload/esf/Nupecc-LRP2017.pdf. [5] https://medicis.cern/ [6] K. Kim et al., Lu-177-Based Peptide Receptor Radionuclide Therapy for Advanced Neuroendocrine Tumors, Nucl Med Mol Imag., 2018 Jun; 52(3): 208–215. https://doi.org/10.1007/s13139-017-0505-6. [7] J. Zhang et al., Clin. Nucl. Med., January 2020, Volume 45, Issue 1, p e48-e50. https://doi.org/10.1097/RLU.0000000000002655. [8] A. Dash et al., Production of 177Lu for targeted radionuclide therapy: available options. Nuc. Med. and mol. 1Ima., 49(2):85–107, 2015. https://pubmed.ncbi.nlm.nih.gov/26085854/. [9] R. Henkelmann, Thursday, 28 February 2013, Lu-177 production with a focus on radiation in the KBA at FRM II, SAAGAS 24, TUM Garching https://indico.frm2.tum.de/event/0/attachments/4/14/Henkelmann_177Lu_gek.pdf. [10] E. A. M. Ruigrok et al., Extensive preclinical evaluation of lutetium-177-labeledPSMA-specific tracers for prostate cancer radionuclide therapy, Eur. J. Nucl. Med. Mol. Imaging, 2020 Oct 23. 10.1007/s00259-020-05057-6. [11] K. M. Ferreira et al., Half-life measurement of the medical radioisotope 177Lu produced from the 176Yb(n,γ) reaction. EPJ Web of Conferences Vol 146, pag 08002, EDP Sciences, 2017. https://doi.org/10.1051/epjconf/201714608002. [12] P. F. Mastinu et al.; New C6D6 detectors: reduced neutron sensitivity and improved safety, (The n_TOF Collaboration), CERN-n TOF-PUB-2013-002 (2013). [13] U. Abbondanno, G. Aerts et al., New experimental validation of the pulse height weighting technique for capture cross-section measurements, https://doi.org/10.1016/j.nima.2003.09.066. Speaker: Mr Francisco García-Infantes (Universidad de Granada, Granada, Spain; European Organization for Nuclear Research (CERN), Geneva, Switzerland) • 09:30 11:00 Reactor Data: I Placerville () ### Placerville Convener: Dan Roubtsov • 09:30 From fission yield measurements to evaluation: correlations of multi-observable analysis 24m From fission yield measurements to evaluation: correlations of multi-observable analysis S.M. Cheikh 1, G. Kessedjian 1, A. Chebboubi 1, O. Serot 1 and C. Sage 2 1 CEA, DES, IRESNE, DER, SPRC, LEPh, Cadarache center, F-13108 Saint Paul lez Durance, France 2 LPSC, Université Grenoble-Alpes, CNRS/IN2P3, 38026 Grenoble, France Topics for submission: Fission yield evaluation Keywords: fission yields, evaluation, variance-covariance matrix Abstract: The study of fission yields has a major impact on the characterization and understanding of the fission process and its applications. The fission products have a direct influence on the amount of neutron poisons that limit the fuel burnup and on the evaluation of the decay heat of the reactor after shutdown. Fission yield evaluation represents the synthesis of experimental and theoretical knowledge in order to perform the best estimation of mass and independent fission product yields. Today, the lack of correlations between the different fission observables induces inconsistencies in the evaluations. For instance, the mass yield uncertainties are drastically overestimated while this observable is the best known. This last decade, different covariance matrices have been proposed but the experimental part of those are neglected. A consistent covariance matrix depends on the evaluation process. This consistency is deeply entangled to the statistical agreement between each data sets. Moreover, the experimental covariance data are crucial in the evaluation process and covariance of model parameters does not represent the only contribution to the covariance matrix associated to evaluation. A large range of data are present in the EXFOR data bank but most of them cover partial mass range, for different incident neutron energy, without necessarily an absolute mass or isotopic identification. Thus, the mix of data could lead to several solutions, which then has to be ranked. The LEPh Laboratory of CEA Cadarache develops a new methodology in the field of the fission products for the future version of the JEFF-library. Statistical test of datasets and data ranking are requested in order to define the confident region of the fission yield knowledge. A complete evaluation of the mass and independent fission product yields of the 235U(nth,f) reaction will be presented with the associated covariance matrix their impact on nuclear reactor calculations. Speaker: Grégoire Kessedjian (CEA Cadarache IRESNE/DER/SPRC/LEPh) • 09:54 Measurement of the fission yield of ^{136}Cs in the ^{239}Pu(n_{\text{th}},f) reaction and its impact on the total dose rate calculation 12m A. Chebboubi1, D. Bernard1, V. Vallet1, G. Kessedjian1, O. Méplan2, Y.H. Kim3, U. Köster3, Ch.E. Düllmann4,5,6, F. Géhin1, M. Houdouin-Quenault1,2, O. Litaize1, C. Mokry4,6, M. Ramdhane2, J. Runke4,5 ,C. Sage3, O. Serot1 1CEA, DES, IRESNE, DER, SPRC, Cadarache, Physics Studies Laboratory, 13108 Saint-Paul-lès-Durance, France 2 LPSC, Université Grenoble-Alpes, CNRS/IN2P3, 38026 Grenoble, France 3 Institut Laue-Langevin (ILL), 38042 Grenoble, France 4 Department of Chemistry - TRIGA site, Johannes Gutenberg University Mainz, 55128 Mainz, Germany 5 GSI Helmholtzzentrum für Schwerionenforschung GmbH, 64291 Darmstadt, Germany 6 Helmholtz Institute Mainz, 55128 Mainz, Germany A recent experimental campaign performed at the LOHENGRIN spectrometer at ILL aimed at measuring the independent fission yield of ^{136}Cs in the ^{239}Pu(n_{\text{th}},f) reaction. Indeed, recent works from CEA/LEPh have shown that this nuclide can have an important contribution to the total dose rate coming from the decay of radioactive nuclides. Moreover, its impact is of first-order on the uncertainty of the total dose rate calculated in specific areas of Nuclear Power Plants within accidental conditions. The production paths of this nuclide in a nuclear reactor are complex. A sensitivity analysis performed with the DARWIN package [1] has shown that one of the most important production paths was directly from the fission process. Therefore, a new measurement of its independent yield along with a rigorous uncertainty analysis would allow handling its impact on the total dose rate. The experimental campaign has been performed in June 2021 with a 129 µg/cm^{2} ^{239}Pu target (99.5% enrichment) produced by molecular plating [2] at the TRIGA Site of the Department of Chemistry at Johannes Gutenberg University Mainz. The experimental setup consisted of a vacuum chamber placed in the focal plane of the LOHENGRIN spectrometer and surrounded by two HPGe clover detectors. Due to its low independent yield, the signal from ^{136}Cs decays is very low in comparison to the γ–ray background at the measurement position. Therefore, the ions were collected by implantation of the mass-separated beam into Al foil placed inside the vacuum chamber. This foil was then removed and transfer to a low γ-ray background setup located at LPSC. The procedure is then repeated for different LOHENGRIN settings. The low γ-ray background setup features a considerably improved signal-to-background ratio compared to more conventional measurements in the on-line regime [3]. This talk will present the results of this experiment and the improvement on the total dose rate calculation uncertainty. [1] A. Tsilanizara, T. D. Huynh, Annals of Nuclear Energy 164, 108579 (2021) [2] J. Runke et al., J Radioanal Nucl Chem 299, 1081 (2014) [3] S. Julien-Laferrière et al., Phys. Rev. C 102, 034602 (2020) Speaker: Abdelaziz Chebboubi (CEA) • 10:06 Gamma ray spectrum measurement from capture reactions of Uranium-238 for thermal and resonance energy neutrons 12m For safe use of nuclear reactor core, slowing down of neutrons and reaction rates inside the core should be predicted accurately. They have made tremendous efforts to improve the calculation methods and the nuclear data for the prediction. However, the calculated neutron spectrum and the reaction rates have been validated only by post irradiation measurement of gamma rays. As an in-situ measurement technique, the authors have focused on spectrum measurement of prompt gamma rays. Nauchi et al. have succeeded in estimation of 238U(n,γ) reaction rates by detection of 4060 keV gamma rays [1]. Then they tried to determine energy of neutron which induces capture reactions of 238U. For that, neutron energy dependent gamma ray spectrum was measured for 238U at KURNS LINAC neutron source facility. Uranium metallic samples of the natural enrichment were irradiated by the pulsed neurons of white spectrum. The gamma rays were measured with a HP-Ge detector of 35% relative efficiency. The energy of neutrons which induces capture was measured by the time of flight technique. The gamma ray spectrum measured for the neutrons of thermal energy and resonances (En=6.67eV and 20.9eV) are shown in Fig. 1. Compared to the published literature [2], the spectrum was obtained with better resolution, and differences in peak gamma ray spectra for the thermal and resonances energy neutrons were observed. By measuring the gamma ray spectrum from a reactor core, the reaction rates ratio of the thermal and the resonance neutron energy might be unfolded taking advantage of the difference of the gamma ray spectrum. [1] Y. Nauchi, T. Sano, H. Unesaki, et al., Proceedings of 11th International conference on nuclear criticality safety, ICNC 2019, September 15-20, 2019 – Paris, France. [2] H. Harada S. Goko, A. Kimura et al., Journal of Korean Physical Society 59(2) 1547-1552, 2011. Speaker: Yasushi Nauchi (Central Research Institute of Electric Power Industry) • 10:18 Measurement of 233U-HU Substitution Reactivity Worth in KUCA for Validation of 233U Nuclear Data 12m As engineering discussions of the feasibility of new reactor systems, it is necessary to evaluate the impact of the fuels and materials for the neutronics characteristics such as criticality, conversion rate, and fuel balance. It is important to perform the validation of 233Th and 233U nuclear data for a thorium-based nuclear reactor, because those nuclear data contribute significantly to the neutronics characteristics of the reactor. In this study, in order to perform integral evaluation of 233U nuclear data, measurements of substitution reactivity worth between 233U sample and high enrichment uranium (HU) sample in Kyoto University Critical Assembly (KUCA) were carried out. In KUCA with solid moderator core, fuel and moderator plates in fuel element were set in a 1.5 mm thickness aluminum sheath and all material plates have nominal cross section of 2” square. In this study, the experimental core was consist of 24 fuel elements and 1 special fuel element. A fuel element was consisted of 31 unit cells and sandwiched by a upper and a lower polyethylene reflector. The unit cell has one enriched uranium (EU) plates of 1/16” thickness, three polyethylene plates of 1/8” thickness. In addition, a special fuel element was loaded in the center of the core. A 233U of HU sample was installed in the central cell of the fuel element (S). The 233U sample was about 9 g U3O8-Al and the HU sample was about 9 g U3O8-Al. The substitution reactivity worth was defined as difference of excess reactivates between 233U or HU inserted core. The excess reactivates were measured by positive period method. As the experimental results, the substitution reactivity was obtained 0.0146 ± 0.0006 %dk/k. On the other hand, a numerical results by MVP-3 with JENDL-4.0 was 0.0141±0.0006 and the C/E was 0.966 ± 0.041. Speaker: Prof. Tadafumi Sano (Kindai University) • 10:30 Fission Cross-section Measurement of U-233 with Time-of-Flight Method at the KURNS-LINAC 12m As engineering discussions of the feasibility of new reactor systems, it is necessary to evaluate the impact of the fuels and materials for the neutronics characteristics such as criticality, conversion rate, and fuel balance. It is important to perform the validation of 232Th and 233U nuclear data for a thorium-based nuclear reactor, because those nuclear data contribute significantly to the neutronics characteristics of the reactor. We have measured the fission cross section of 233U by the neutron time-of-flight method using a 12 m flight path and a 46-MeV electron linear accelerator at the Institute for Integrated Radiation and Nuclear Science, Kyoto University (KURNS-LINAC). Experiments were performed with two different linac operation modes. One was for low-energy measurement below about 1 eV with 50 Hz and an average current of about 18 μA. Another was for high-energy measurement above 1 eV with 200 Hz and an average current of about 72 μA. In the latter case, a Cd sheet with 0.5-mm thickness was inserted into the TOF beam line to prevent the overlap component from the previous pulse due to high frequency. Pulse width was 100 ns for both operation modes. A back-to-back type fission chamber was used for the detection of the fission fragments and to discriminate the decayed alpha-particles. A sample of uranium-233 oxide was electrodeposited on the stainless steel disk (28 mm in diameter and 0.2 mm thickness). A relative measurement to the 6Li(n,α) standard reaction was also made using a 6Li-glass detector, and normalized to the evaluated value from the JENDL-4.0 at thermal neutron energy. We have obtained the fission cross sections of 233U in the energy region from 0.002 eV to 1 keV. The present results were compared with the previous experimental data and the evaluated values. Moreover, the resonance parameters were obtained by using SAMMY code for the principal resonances below about 100 eV. Speaker: Jun-ichi Hori (Kyoto University) • 11:00 11:45 Break 45m • 11:45 13:15 Fission: II American River () ### American River Convener: Anton Tonchev • 11:45 Generation of Fragment Angular Momentum in Fission 24m The origin and character of the fission fragment angular momenta are currently topics of intense theoretical study. We describe how the angular-momentum bearing modes of the evolving dinuclear complex are being agitated at different rates by multiple transfers of individual nucleons, the mechanism long understood to be the primary cause of the transport phenomena displayed by damped nuclear reactions, including the generation of the angular momenta of the fragments [1,2]. The resulting fragment angular momenta are predominantly perpendicular to the fission axis and, although they are built up from collective rotational modes in which the two spins are highly correlated, they nevertheless emerge as largely independent [3,4], a feature supported by recent experimental data [5]. The fission simulation code FREYA [6], which treats angular momentum as suggested by the nucleon-exchange mechanism, presents a powerful tool for exploring observational consequences of the correlated fragment spin distributions and a number of examples will be discussed. For example, there is a marked correlation between the fragment spin magnitudes and the photon multiplicity [4]. [1] J. Randrup, Transport of angular momentum in damped nuclear reactions, Nucl. Phys. A 383, 468 (1982). [2] T. Dossing and J. Randrup, Dynamical evolution of angular momentum in damped nuclear reactions: (I) Accumulation of angular momentum by nucleon transfer, Nucl. Phys. A 433, 215 (1985). [3] J. Randrup and R. Vogt, Generation of fragment angular momentum in fission, Phys. Rev. Lett. 127, 062502 (2021). [4] R. Vogt and J. Randrup, Angular momentum effects in fission, Phys. Rev. C 103, 014610 (2021). [5] J. Wilson et al., Angular momentum generation in nuclear fission, Nature 590, 566 (2021). [6] J.M. Verbeke, J. Randrup, and R. Vogt, Fission Reaction Event Yield Algorithm FREYA 2.0.2, Comp. Phys. Comm. 222, 263 (2018). Speaker: Jorgen Randrup • 12:09 Novel Fully Microscopic Description of Fission: Odd-Even Staggering, Charge Polarization and Machine Learning 12m Nuclear fission is the fundamental mechanism that determines energy production in nuclear power plants. Predictive fission models are necessary for national security applications and are also key to explain the relative abundances of the elements in the universe. In the first part of this talk, we present a novel approach based on particle-number restoration techniques in the fission fragments to predict the preneutron fragment distributions. We show that we can obtain for the first time a qualitatively good description of the mass and charge yields Y(Z, A) using such microscopic approaches. In addition, our method can also describe fine-structural effects such as the odd-even staggering in the charge yields and the charge-polarization of the fission-fragments distributions. In the second part of this talk, we outline a new machine-learning framework to predict microscopic configurations at different nuclear deformations. Acknowledgments: This work was supported in part by the NUCLEI SciDAC-4 collaboration DE-SC001822 and was performed under the auspices of the U.S.\ Department of Energy by Lawrence Livermore National Laboratory (LLNL) under Contract DE-AC52-07NA27344. Computing support came from the LLNL Institutional Computing Grand Challenge program. Support was also provided by the LLNL Laboratory Directed Research \& Development (LDRD) 21-ERD-001. Release number: LLNL-ABS-827074 Speaker: Marc Verriere • 12:21 Correlations of Gamma-Ray Emission Between Fission Fragments: Experimental and Theoretical Challenges 12m Following the binary fission process, the two energetic fission fragments radiate their energy and angular momenta in the form of neutron and gamma-ray emission. These electrically neutral particles represent important signatures of fission in non-destructive assay of nuclear material and an analysis of the emitted radiation can be used to characterize the fissioning source. The emission of neutrons and gamma rays often represents the only probe we have of the nuclear fission process and the state of the fragments immediately following fission. Recent theoretical and experimental results have sparked renewed interest in the correlations between fission fragments. Of particular interest are the correlations between the fission fragment angular momenta. The most important observable related to the fission fragment angular momenta is the emission of gamma rays. It is thus of current interest to determine the correlations in the emission of gamma rays between fragments. While several conflicting models exist to model these correlations, experimental data is limited or inconclusive. In this work, we review the theoretical models showing how different models can give rise to differences in the gamma-ray emission correlations. We then review the experimental evidence, highlighting the experimental challenges involved and the expected systematic biases. Lastly, we present novel experimental techniques that allow us, for the first time, to directly determine the desired gamma-ray correlations. The technique generalizes the Maier-Leibniz Doppler-Shift method. In the original technique, the average yield of gamma rays is inferred from the small aberration of gamma-rays due to the speed of the emitting fragments. We have derived new sets of equations based on the same assumptions, showing that the aberration of gamma rays can be exploited to determine the second moment of the multiplicity distribution of gamma rays, i.e., the variances of the distributions and their covariance. We discuss this new technique, its applicability, and the experimental setups required to utilize it. Speaker: Stefano Marin • 12:33 Angular momentum of spherical fission products: experimental and theoretical aspects 12m Despite the numerous theoretical [1] and experimental [2] works published very recently, the way in which fission fragments acquire their angular momentum is still an open question. This angular momentum generation mechanism is important not only for improving our understanding of the fission process, but also for nuclear energy applications, since fission fragments angular momentum strongly impact the prompt gamma spectra and consequently decay heat in a reactor. In this context, within the framework of a collaboration between the ‘Laboratoire de Physique Subatomique et Corpusculaire’ (LPSC, France), the ‘Institut Laue Langevin’ (ILL, France) and the CEA-Cadarache (France), an experimental program was developed on the LOHENGRIN mass-spectrometer with the aim of measuring isomeric ratio of some fission products for different thermal neutron induced fission reactions. This presentation will be focused on results obtained for the spherical nucleus 132-Sn following thermal neutron induced fission of both 235-U and 241-Pu targets. To further constrain the angular momentum generation models, 132-Sn isomeric ratio (IR) is measured as a function of its kinetic energy (Ek). The angular momentum is determined by combining our experimental data with calculations performed with the FIFRELIN Monte-Carlo code [3]. A clear angular momentum decrease with Ek is observed for both reactions. A comparison of the present work with the predictions of the Madland-England model [4], the GEF [5] code and the FIFRELIN code will be also discussed. Lastly, we investigate the dependence of the 132-Sn angular momentum with the incident neutron energy, from thermal region up to the first chance fission. For that, the four free available parameters in FIFRELIN are selected in order to reproduce the average prompt neutron multiplicity. In this way, the angular momentum is deduced for each neutron energy. These results are discussed in terms of the impact of the fission fragment deformation at scission on the spin assignment. References [1] J. Randrup and R. Vogt, Phys. Rev. Lett. 127, 062502 (2021), and references therein [2] C.J. Sears et al., Nuclear Data Sheets 173 (2021) 118–143, and references therein [3] O. Litaize, O. Serot, and L. Berge, Eur. Phys. J. A 51, 177 (2015) [4] D. G. Madland and T. R. England, Nucl. Sci. Eng. 64, 859 (1977) [5] K. H. Schmidt et al., Nucl. Data Sheets 131, 107 (2016) Speaker: Olivier Serot • 12:45 Fission In R-process Elements (FIRE) 12m The goal of the FIRE topical collaboration in nuclear theory (2017-2021) was to determine the astrophysical conditions of the rapid neutron capture process (r-process), which is responsible for the formation of heavy elements. This was achieved by including in r-process simulations the most advanced models of fission (spontaneous, neutron-induced, β-delayed) that are beeing developed at LLNL and LANL. The collaboration was composed of LLNL (lead) and LANL for work on fission and nuclear models, BNL for nuclear data management, and the university of Notre Dame and North Carolina State University for r-process simulations. I will give a summary of the results obtained by the collaboration and their impact on nuclear data. Speaker: Nicolas Schunck • 11:45 13:15 Libraries: II Delta King () ### Delta King Convener: Bret Beck (LLNL) • 11:45 Towards implementing new isotopes for environmental research: The half-life of ^{32}Si 24m ^{32}Si is an extremely rare, naturally occurring radioactive isotope. With its half-life (\textit{T}$$_{1/2}$) of approximately 150 years $^{32}$Si would be one of the suitable candidates for radiometric dating in the range of 100–1000 years, where an appropriate dating nuclide is still missing (Fig. 1a).
The fact that the application of this nuclide for dating has been very limited so far is due to the imprecise and contradictory data for its half-life (Fig. 1b). With the current best estimate of 153(19) years, a more precise and more accurate determination is absolutely necessary.

The SNSF-funded project SINCHRON ($\textbf{Si}$ – a $\textbf{n}$ew $\textbf{chro}$nometer for $\textbf{n}$uclear dating) aims for an accurate half-life redetermination of $^{32}$Si with a relative standard uncertainty of less than 5% on the basis of several independent measurements. MBq quantities of $^{32}$Si have been successfully produced at the Paul Scherrer Institut (PSI, Switzerland) by exposing metallic vanadium discs to high-energy protons. In order to obtain radiochemically pure $^{32}$Si solutions, a robust chemical separation procedure has been developed (2). Several partners are involved in the SINCHRON-Project covering different tasks of the half-life determination.
Generally, two approaches are employed, while all measurements will be performed using aliquots of the same source material. The first approach is to follow the decay over a given time interval. For such measurements, the long-term stability of both the sample and the measurement device (e.g., ionization chamber (IC)) is essential. The second approach is the so-called direct method, where the $\textit{T}$$_{1/2} can be determined from the relationship \textit{T}$$_{1/2}$ = $\mathit{N}$ $\mathit{ln(2)}$/$\mathit{A}$. Here, inductively coupled plasma mass spectrometry (ICP-MS), and accelerator mass spectrometry (AMS) are utilized for the determination of the number of atoms ($\mathit{N}$). The activity ($\mathit{A}$) is measured using liquid scintillation counting (LSC) with two techniques that are well established in radionuclide metrology: the triple-to-double coincidence ratio (TDCR) method and CIEMAT/NIST efficiency tracing. In addition, a coincidence setup with a plastic scintillation detector and a gamma-ray detector is used to apply another independent efficiency tracing technique. Finally, enough sample material that meets the quality requirements has been produced and the individual measurements are currently ongoing - first, preliminary results of the half-life determination will be presented.

$\textbf{References}$:
(1) Ouellet, C., and Balraj, S. "Nuclear data sheets for A = 32" $\textit{Nuclear Data Sheets}$ 112.9 (2011): 2199-2355.

(2) Veicht, M., Mihalcea, I., Cvjetinovic, Đ., & Schumann, D. (2021). "Radiochemical separation and purification of non-carrier-added silicon-32". Radiochimica Acta.$\textit{Radiochimica Acta}$ (2021, pre-published online).

$\textbf{Acknowledgements}$:
The authors acknowledge the funding through the Swiss National Science Foundation (Grant No. 177229) and additionally from the Marie Skłodowska-Curie Grant (No. 701647).

Speaker: Mario Veicht (Paul Scherrer Institut (PSI); École polytechnique fédérale de Lausanne (EPFL))
• 12:09
Validating Gamma-rays Produced in Nuclear Reactions 12m

Authors:
Isabel Hernandez, Jennifer Jo Ressler, Bonnie Canion

Understanding gamma-ray production in a neutron environment is vital in many scientific concentrations. Emergency response, planetary exploration, safety and shielding designs – among many other applications – depend on analyzing neutron induced gamma-ray spectra from measurements. These research areas often require simulation capabilities to extend analysis and substitute as a low-cost and fast solution option to physical measurements. Due to the wide availability and applicability of neutron transport codes, many users depend on the status of current nuclear data libraries in order to produce accurate simulated results.

Discrepancies observed between simulations and measurements may be due to the availability of the nuclear data needed to model an interaction, the accuracy and precision of the nuclear data available, or the manner in which the simulation tool is incorporating the nuclear data to model an interaction. Therefore, it is of current interest to determine the condition of current nuclear data libraries and compare them to gamma-ray data. We focus on the ENDF/B VIII evaluation, due to the high volume of users. Proposed benchmark datasets, such as published IAEA compiled standards, will be used in its comparison. In this work, we developed a process for extracting neutron induced gamma-ray production data from the GNDS data libraries for discrete gamma-ray energies, as well as exploring best methods for including the continuum spectrum contribution.

We then compared these gamma-ray production cross sections in the GNDS library to experimental benchmarks. We saw that non-graphical library extraction methods provided a more accurate cross section compared to experimental results, though the continuum contribution that led to a comparable cross section was often non-trivial compared to its discrete counterpart, and the appropriate integration energy window size varied greatly. This is notable, since the importance of the non-discrete contribution to the total cross section is prevalent in all analyzed isotopes and may lead to interpretation issues in simulation codes. Lastly, we used popular neutron transport codes, MCNP, Mercury, and GEANT4 to extract the interpreted ENDF/B VIII data and provide comparisons to measurements. We found that isotopes with library storage issues found in the previous step had greatly misinterpreted cross sections across the board compared to the raw library data. These isotopes include 12C and 14N, two prominent isotopes relevant for applications such as non-destructive cargo screening techniques. Future work will include investigating why these Monte Carlo codes are accessing nuclear data differently and extending this process to other isotopes and interactions. Recommendations to the U.S. Nuclear Data Team have and will be made for further improvements where data are incorrect or missing.

Prepared by LLNL under Contract DE-AC52-07NA27344.
LLNL-ABS-827870

Speaker: Isabel Hernandez
• 12:21
Resolved Resonance Region Analysis of $^{206}$Pb, $^{207}$Pb, AND $^{208}$Pb in Support of Next Generation Lead-Cooled Fast Systems. 12m

{Evaluated nuclear data files are the basis for representing radiation interactions with nuclei in most modern transport and diffusion codes. It is therefore important to verify and update isotopic evaluations periodically to ensure they incorporate up-to-date experimental data and reaction theory in the hope to accurately model systems with low uncertainty. Recently, great interest has been generated in using lead as a coolant for fast neutron systems and as a result it is important to investigate the ENDF/B-VIII.0\cite{ENDF} isotopic evaluations that comprise stable lead. To this end, resonance parameters for $^{206}$Pb, $^{207}$Pb, and $^{208}$Pb were re-evaluated because their resolved resonance regions extend beyond 0.5 MeV meaning resonance parameters used in reconstructing cross sections and elastic scattering angular distributions impact fast systems. The impact of resonance parameters is demonstrated by the differences between the evaluations in predicting experimental results from the fully modeled RPI Quasi-Differential Scattering Experiment\cite{RPIscatter} in MCNP\cite{MCNP}, Figure \ref{fig:RPIScat2}. In addition, MCNP KCODE calculations of lead-sensitive fast spectra critical benchmarks showed variations of k$_{\mathit{eff}}$ on average of 200 pcm, caused solely from differences in elastic scattering angular distributions in $^{208}$Pb. Re-evaluation entailed fitting data with the program SAMMY\cite{SAMMY}. Resonance energies as well as neutron and radiation widths were fit sequentially to previously available high resolution transmission measurements and new capture yield experiments made available after the release of ENDF/B-VIII.0. Constraints for resonance parameters included SAMMY's reduced chi-square, resonance statistics (Wigner, Porter-Thomas, and summed strength plots), and resonance properties (scattering lengths, thermal quantities, and resonance integrals). Preliminary validation, done via a suite of critical benchmarks containing lead as well as MCNP simulations of the RPI scattering experiment, shows better agreement from the RPI evaluation for fast spectra criticality and scattering simulations than ENDF/B-VIII.0, JENDL-4.0 \cite{JENDL}, and JEFF-3.3\cite{JEFF}.

Speaker: Peter Brain (RPI)
• 12:33
Statistical Uncertainty Quantification of Probability Tables for Unresolved Resonance Cross Sections 12m

The self-shielding effect in the unresolved resonance region has a large impact on the fast- and intermediate-spectrum reactors. The probability table method is widely used for continuous-energy Monte Carlo calculation codes to treat the effect. In this method, a table provides the probability distribution of the cross-section for a nuclide in the given energy grid points. The table is generated by averaging with a lot of “ladders” which represent pseudo resonance structures. Though many nuclear data processing codes require the number of ladders as an input parameter to generate the probability table, an optimal number of ladders has not been investigated. Our previous study revealed that the suitable number of ladders depends on the nuclide and its resonance parameters. This result indicates that it is very difficult for users to find the optimal number of ladders.
We developed the calculation method of the statistical uncertainty for the probability table generation. The product of the probability table and total cross-section in each probability bin is considered as the target of statistical uncertainty of the probability table. The central limit theorem (CLT) method, the Bootstrap method, and the Jackknife method are used to calculate the statistical uncertainty and the statistical uncertainties of these methods are compared. The calculation results indicate that the statistical uncertainty of the CLT method is similar to that of the other methods. The CLT method is the best way to calculate the statistical uncertainty of the probability table with a short computational time.
This statistical uncertainty calculation method for the probability table generation will be implemented in the next version of the nuclear data processing code FRENDY, i.e., FRENDY version 2. Users can generate the probability table with the optimal number of ladders when users set the criterion of the statistical uncertainty of the probability table as the input parameter.

Speaker: Dr Kenichi Tada (Japan Atomic Energy Agency)
• 12:45
Updates and Validation for the n+63,65Cu Cross Sections 12m

The neutron induced total, elastic, and capture cross sections of 63,65Cu iso- topes were selected for evaluation in the resolved and unresolved resonance energy ranges by the National Criticality Safety Program to resolve discrepancies related to benchmark performance. This is especially evident for the series of ZEUS benchmarks in which copper is used as a reflector. Because copper is also used as structural material in both fission and fusion reactors, the need to address benchmark discrepancies linked to nuclear data deficiencies is a task of primary importance. The aim of this work is to describe the steps of evaluation work towards a consistent improvement of the benchmark performance.
The R-matrix analysis with the SAMMY code focused on the 63Cu(n,γ) reaction channel between 100-300 keV coupled to unresolved resonance region parameters up to 650 keV to fit average cross section data from a recent experiment. Due to the high sensitivity of many benchmarks to elastic scattering angular distribution data, especially for the 65Cu isotope, the impact of these data was tested by generating Legendre coefficients from both
resonance parameters and the Hauser-Feshbach model. Guided by the findings of Shaw et al., the performance of the current evaluation for 65Cu was compared to that of ENDF/B-VII.1 and ENDF/B-VIII.0 by testing the reactivity coefficients corresponding to the validation suite of experimental criticality benchmarks for thermal, intermediate, and fast systems taken from the International Criticality Safety Benchmark Experiments Project Handbook. The benchmark performance is especially sensitive to 63Cu(n,γ) and 65Cu elastic scattering for neutron energies in the 100–500 keV region, whereas 100 keV is the upper limit of the resolved resonance region in the ENDF/B-VIII.0 evaluations for 63,65Cu. The results highlight the need to handle the transition from the resolved resonance region to the high energy region carefully.

Speaker: Jordan McDonnell (UT-Battelle)
• 12:57
Results from the ARTIE experiment 12m

A measurement of the transmission coefficient for neutrons through a thick ($\sim 3$ atoms/b) liquid natural argon target in the energy range $30$-$70$ keV was performed by the Argon Resonance Transmission Interaction Experiment (ARTIE) using a time of flight neutron beam at Los Alamos National Laboratory.
In this energy range theory predicts an anti-resonance in the $^{40}$Ar cross section near $57$ keV, but the existing data, coming from an experiment performed in the 90s (Winters. et al.), does not support this.
This discrepancy gives rise to significant uncertainty in the penetration depth of neutrons through liquid argon, an important parameter for next generation neutrino and dark matter experiments.
In this talk, the final results from the ARTIE experiment will be presented.
The ARTIE measurement of the total cross section as a function of energy confirms the existence of the anti-resonance near $57$ keV, but not as deep as the theory predicts.

Speaker: Luca Pagani (UC Davis)
• 11:45 12:45
Machine Learning: II Sutter's Fort ()

### Sutter's Fort

Conveners: Amber Lauer-Coles (BNL), Cole Pruitt
• 11:45
Where Could Compensating Errors Hide in ENDF/B-VIII.0? 24m

LA-UR-21-29134

It is a well-known fact that our current nuclear data libraries, among them ENDF/B-VIII.0 [1] in the US, are home to compensating errors that lurk in unconstrained physics spaces [2]. Unconstrained physics spaces arise, on the one hand, due to differential information (from theory and experiments) that is too imprecise to fully describe nuclear data. One the other, this is combined with validating combinations of nuclear data with integral experiments cannot uniquely identify one nuclear data value as driving the difference between simulated and experimental values of these experiments. If we knew where these unconstrained physics spaces occur in nuclear data libraries, we could either build targeted experiments or propose theory developments that help to shine a light on them and, thus, reduce the possibility of compensating errors.

The EUCLID project (Experiments Underpinned by Computational Learning for Improvements in Nuclear Data) developed a work-flow [3] that identifies these unconstrained physics spaces by bringing together various experimental values of integral responses (criticality, $\beta_\mathrm{eff}$, reaction rates in critical assemblies, LLNL pulsed spheres neutron-leakage spectra, reactivity coefficients and sub-critical assembly observables) with their simulated counter-parts as well as differential information. This wealth of information is processed by machine learning tools supported by the random forest algorithms and analyzed by human experts [4]. Here, we will show how this work-flow is executed and a few representative examples of where compensating errors could potentially hide in ENDF/B-VIII.0.

Acknowledgments
Research reported in this publication was supported by the U.S. Department of Energy LDRD program at Los Alamos National Laboratory. This work was supported by the US Department of Energy through the Los Alamos National Laboratory. Los Alamos National Laboratory is operated by Triad National Security, LLC, for the National Nuclear Security Administration of the US Department of Energy under Contract No. 89233218CNA000001.

[1] D.A. Brown, M.B. Chadwick, R. Capote et al.,ENDF/B-VIII.0: The 8th Major Release of the Nuclear Reaction Data Library with CIELO-project Cross Sections, New Standards and Thermal Scattering Data,Nucl. Data Sheets 148, 1 (2018).
[2] E. Bauge, G. Belier, J. Cartier et al., Coherent Investigation of Nuclear Data at CEA DAM: Theoretical Models, Experiments and Evaluated Data, The European Physical Journal A 48, 113 (2012).
[3] D. Neudecker, O. Cabellos, A.R. Clark et al., Informing Nuclear Physics via Machine Learning Methods with Differential and Integral Experiments, Phys. Rev. C 104, 034611 (2021).
[4] D. Neudecker, M. Grosskopf, M. Herman et al., Enhancing Nuclear Data Validation Analysis by Using Machine Learning, Nucl. Data Sheet 167, 36 (2021).

Speaker: Denise Neudecker
• 12:09
Development of a New Fixed-source Sensitivity Tally Capability in the MCNP® Code 12m

The development of a new fixed-source sensitivity tally capability, targeted toward nuclear data parameters, is currently underway in the MCNP code. In recent research and development efforts that utilize machine learning to both seek problematic nuclear data as well as design experiments optimized to improve the nuclear data, the adjoint-weighted k-eigenvalue sensitivity tally capabilities have been essential. In this paper, the motivation to expand the sensitivity tally capabilities beyond k-eigenvalues toward diverse fixed-source problems along with preliminary results and verification will be discussed.

Speaker: Michael Rising
• 12:21
A Study of the Covariance Data in ENDF/B VIII.0 for Low Z Isotopes 12m
Speaker: Kent Parsons (Los Alamos National Laboratory)
• 12:33
Burnup Calculation Uncertainty Using Statistical Sampling Method 12m

The nuclear data are the basic data for reactor burnup calculation. It is significant to study the contribution of their uncertainty to the uncertainty of reactor burnup calculation for improving the safety and economy of reactor. we propagated the uncertainties coming from the nuclear data to the isotopic inventory of sample SF95-4. A Monte Carlo sampling code was developed and used to propagate the decay constant uncertainties, cross-section covariance information and uncertainties of the fission yields. The results showed that the impact of decay constant uncertainties was inappreciable and the contribution of cross-section covariance mostly affected the uncertainties of the actinides. Fission yields appeared to have the largest impact on the uncertainties of the fission products.

Speaker: Xiaofei Wu
• 11:45 13:15
Measurements: II Folsom ()

### Folsom

Convener: Keegan Kelly (LANL)
• 11:45
History of fission nuclear data from 1939-1945 and evolution to today's understanding 24m

Nuclear physics advances in the United States and Britain from 1939 to 1945 are described. The Manhattan Project’s work led to a rapid advance in our knowledge of nuclear science. A conference in April 1943 at Los Alamos provided a simple formula used to compute critical masses and laid out the research program needed to determine the key nuclear constants. In short order, four university accelerators were disassembled and reassembled at Los Alamos, and methods were established to make measurements on extremely small samples owing to the initial lack of availability of enriched 235U and plutonium. I trace the program that measured fission cross sections, fission-emitted neutron multiplicities and their energy spectra, and transport cross sections, comparing the measurements with our best under-standing today as embodied in the Evaluated Nuclear Data File ENDF/B-VIII.0. The large nuclear data uncertainties at the beginning of the project, which often exceeded 25% to 50%, were reduced by 1945 often to less than 5% to 10%. Uranium-235 and plutonium-239 fission cross-section assessments in the fast mega-electron-volt range were reduced following more accurate measurements, and the neutron multiplicity ν increased. By a lucky coincidence of canceling errors, the initial critical mass estimates were close to the final estimated masses. Some images from historical documents from our Los Alamos archives are shown. Many of the original measurements from these early years have not previously been widely available. Through this work, these data have now been archived in the international experimental nuclear reaction data library (EXFOR) in a collaboration with the International Atomic Energy Agency and Brookhaven National Laboratory.

• 12:09
Measurement of the $^{232}$Th/$^{235}$U fission cross section ratios at the Back-n white neutron source of CSNS 12m

Neutron-induced fission cross section of 232Th has important applications in the Th/U fuel cycle. A measurement of the 232Th/235U fission cross section ratios in 1-20 MeV energy region was carried out by the time-of-flight method and Fast Ionization Chamber Spectrometer for Fission Cross Section Measurement (FIXM) at the China Spallation Neutron Source (CSNS)- Back-n white neutron source (Back-n). The fission event-energy spectrum and amplitude spectrum of 232Th and 235U have been measured in the single-bunch and double-bunch operation modes. The time-of-flight (TOF) spectrum under double-bunch mode was unfolded by using Bayes method and the results of fission cross section ratios of 232Th/235U was obtained after various modifications. The results under single-bunch mode was also obtained and the experimental uncertainty of the two datasets was analyzed. The results of 232Th/235U fission cross section ratios under the two modes are in agreement in the range of experimental uncertainty. The results are compared with the main evaluation database and the previous experimental results. The current results can provide experimental data support for evaluation of relevant nuclear data and design of Th/U cycle nuclear system.

Speaker: Zhizhou Ren
• 12:21
Results on the $n+ ^{233}$U $\alpha$-ratio, prompt fission $\gamma$-ray spectra and isomeric fission states measured at n_TOF 12m

The 233U nucleus plays a key role as the fissile isotope in the Th-U fuel cycle. Consequently the accurate knowledge of its fission and capture cross sections is essential for a potential reactor design. As highlighted by the NEA High Priority Request List, the 233U(n,g) is of high importance for the production and destruction of 233U and is required to facilitate defining the reprocessing scheme for a molten salt reactor. Due to the 233U(n,f) reaction's inherent gamma-ray background the measurement of 233U(n,g) requires an efficient background discrimination. This can be achieved by employing a fission and a gamma-ray detector in anti-coincidence. For this purpose a pocket-sized fast fission chamber was designed and used together with the n_TOF Total Absorption Calorimeter 4$\pi$ gamma-ray detector at the n_TOF EAR1 facility at CERN. A brief description of the experimental setup is given and the fission detector and coupled detectors' performance will be presented. A description of the analysis methodology and procedure discussing relevant sources of background for the measurement will be given. The setup's capability to identify the fission gamma-rays for background estimation in turn allows us to extract information on the average energy and multiplicity of the prompt fission gamma-rays which are compared with other recent measurements and theoretical models. Furthermore, a careful analysis of the coincidence spectra enables to study the half-lives of 233U isomeric fission states. A good agreement with simulated gamma-ray cascades of the captured nucleus allows us to calculate the experimental detection efficiencies and finally the results on the 233U $\alpha$-ratio and 233U(n,g) will be shown and compared to existing data sets and evaluations.

Speaker: Michael Bacak
• 12:33
Towards the experimental validation of a small Time-Projection-Chamber for the quasi-absolute measurement of the fission cross section 12m

Improvement in neutronics codes jointly with the advent of high performance computing systems made the calculations more sensitive to the nuclear data. The latter are used both for solving the neutron transport equation together and for nuclear instrumentation validation and operation. Hence, it becomes relevant to improve the knowledge of the fission cross section of fertile secondary actinides as the 242Pu one, which is fissile in a fast neutron flux. This isotope has been chosen as a deposit for the fission chamber for the online monitoring of the fast flux in the experimental irradiation reactor Jules Horowitz (RJH) at CEA Cadarache. As any nuclear thermal or fast neutron reactor, it has a high neutron flux around 1 MeV. This motivates the improvement of the fission cross section for the fertile 242Pu isotope, for which the various experimental data show a dispersion of 10 to 15% around 1 MeV.
The standard measuring technique of a fission cross section is based on simultaneous comparison between the target nucleus and another one so-called reference nucleus. Usually secondary standards are used as reference reactions, known within a few percent, and calibrated to a primary standard of very high precision. The classical isotope chosen as reference for fission cross sections is the 235U whose fission cross section is known with an accuracy of 0.5 to 5%. The use of the same reaction as reference leads to correlations between the different measurements. Our approach aims to produce an independent measurement and to bypass the secondary standard by performing the measurement directly with the primary standard. Thereby, the obtained measurements are completely uncorrelated to any other. The reaction cross section 1H(n,n)p was chosen to achieve our goal since the latter is known from 0.2 to 0.5% over the energy range 0-20 MeV, allowing very accurate measurements. Quantifying the neutron flux with the 1H(n,n)p reaction requires a precise count of the number of recoil protons emitted by a hydrogenated sample of chosen thickness irradiated by this neutron flux. It is therefore essential to use a recoil proton detector having a perfectly known intrinsic efficiency in all operating regimes and a linear response with respect to the input signal. Above 1 MeV, the use of a silicon junction is fully adequate. However, this device is unsuitable at lower energies when a large number of gamma and electrons generate a crippling background noise. This presentation will therefore focus on the recent development and validation of the Gaseous Proton Recoil Telescope (GPRT), insensitive to gamma/electrons noise. This detector uses the Micromegas technology for the detection plane, contains a small time-projection chamber and will be used for the 242Pu fission cross section measurement from 200 keV to a few MeV. During this work, the optimal conditions and the intrinsic efficiency of this detector have been investigated and will be presented. The track reconstruction and the background rejection will also be shown.

Speaker: Carole Chatel
• 12:45
Measurement of neutron-induced fission cross sections of U-235 and U-238 relative to n-p scattering at CSNS Back-n facility 12m

Neutron-induced fission cross sections of U-235 and U-238 are ones of the most important nuclear data since they are fundamental to nuclear energy. Fission cross sections of U-235 and U-238 have been evaluated as standard data up to 200 MeV and they are always used as references for other cross section measurements. However, the experimental data in high neutron energy region are scarce. Especially above 30 MeV of neutron energy, there are only a few measurements with obvious discrepancies. Thus conducting a measurement in high energy region is quite necessary meaningful.

The back-streaming neutron facility (Back-n) at China Spallation Neutron Source (CSNS) is a newly built neutron beamline started commissioning since 2018. Back-n provides neutrons from 0.5 eV to 200 MeV with an achievable flux of 1.6×10^7 n/cm2/s at 55 meters away from the spallation target. It is therefore a good platform for nuclear data measurement. We performed an experiment at Back-n for measuring the fission cross sections of U-235 and U-238 relative to n-p elastic scattering. The U-235 and U-238 samples are sealed in an ionization chamber for measuring their fission reactions. A polythene (PE) foil and several recoiling proton telescopes (RPT), consisting of silicon detectors and cesium iodide scintillators, are setup in a vacuum chamber located at the downstream of the fission chamber. The proton events are selected by the ΔE-E identification. The proton events produced via 12C(n, p) reactions are obtained by measuring a graphite sample with equivalent thickness, which is used to correct the proton reaction rate measured with the PE sample.

We will firstly introduce the CSNS Back-n facility. Then we will go to details of the data analysis of fission chamber and RPT. Finally the preliminary results of fission cross section of U-235 and U-235 from 10 to 100 MeV will be shown.

Speaker: Yonghao Chen
• 12:57
Method to Compare Fission-To-Scattering Ratios using Uranium-238 12m

Neutron time-of-flight quasi-differential high energy scattering measurements of uranium-238 and natural iron were made at the Rensselaer Polytechnic Institute (RPI) Gaerttner Linear Accelerator Center. Data from the measurements were compared to evaluated nuclear data libraries using MCNP to identify energy-angle discrepancies from 0.5 to 20 MeV, and elastic-to-inelastic scattering ratios were generated using the detectors’ measured response functions from 1.4 to 2.0 MeV. A new analysis technique was applied to pre-existing uranium-238 data to derive uranium-238 fission response functions at several energies. Combining the scattering and fission response functions has allowed for the calculation of scattering-to-fission ratios for uranium-238. These response functions can also be used to isolate the scattering contribution and to constrain nuclear data physics models used in the development of evaluated nuclear data libraries.

• 11:45 13:21
Particle Therapy Capitol ()

### Capitol

Convener: Shamsu Basunia (LBL)
• 11:45
Nuclear fragmentation cross section measurements with the FOOT experiment 24m

FOOT (FragmentatiOn Of Target) is an applied nuclear physics experiment with the aim of performing high precision cross section measurements for fragmentation reactions of interest in Hadrontherapy and Radioprotection in Space. An in-depth knowledge of the physical and biological effects caused by nuclear fragments is in fact of great interest for both the improvement of Hadrontherapy treatment planning and the development of effective spacecraft shielding systems in long-term human missions in deep space (e.g. Mars explorations). However, the data needed to accurately model the behavior of nuclear fragments in these fields are currently scarce or totally unavailable in literature. To fill in the gaps in nuclear databases, the FOOT collaboration will perform a set of measurements with light ion beams, such as C and O, in the energy range of 100-800 MeV/n impinging on tissue-like and shielding material targets.

The FOOT experiment allows for a precise identification of the produced nuclear fragments through the measurement of their kinematic characteristics. Each detector has been studied to give the best possible resolution with the aim of performing measurements in inverse kinematics and with composite targets. To this purpose, the apparatus has been developed with the capability of accurately characterizing both the primary beam and all the fragments produced in nuclear reactions with the target. The system comprehends two complementary setups: an emulsion chamber spectrometer, specialized in the detection of lighter fragments (Z $\leq$ 3), and an electronic setup, dedicated to heavier particles (Z $\geq$ 3), which includes a magnetic spectrometer, a Time-Of-Flight system and a calorimeter.

In 2021, two data acquisition campaigns have been carried out with the electronic setup: one with $^{16}$O beams at the GSI laboratories (Darmstadt, Germany) and one with $^{12}$C ions at the CNAO facility (Pavia, Italy). On both occasions, fragmentation measurements were performed using graphite and polyethylene targets.
In this contribution, after an overview of the current status of the FOOT experiment, the first cross section measurements obtained from the GSI and CNAO campaigns are presented.

Speaker: Roberto Zarrella (University of Bologna - INFN Bologna)
• 12:09
MEASUREMENT OF THE REACTION CROSS SECTIONS OF THE SHORT- AND LONG-LIVED β+ EMITTERS FOR PET RANGE VERIFICATION IN PROTON THERAPY AT CNA, WPE AND HIT 12m

In proton therapy, Positron Emission Tomography (PET) range verification, which is based on the detection of the short-lived (online monitoring) or the long-lived (offline monitoring) $\beta^{+}$ emitters produced in the body of the patient, has been proved to be a well-suited technique to monitor the beam range $^{[1]}$. This technique requires the comparison of the observed activity distribution with a simulated one using a Monte Carlo code. As the reliability of the simulated activity distribution depends on the accuracy of the underlying cross sections for producing the $\beta^{+}$ emitters of interest $^{[2][3][4]}$, several studies confirm the need for more and better measurements and evaluations $^{[4][5][6]}$. New data related to the production of the short-lived nuclides involved in real-time verification $^{[7][8][9]}$ are especially needed, as there are no data available yet in the energy range of interest, up to 200 MeV.

In this context, we have developed two methods to measure the production cross sections of the mentioned long-lived ($^{11}$C with t$_{1/2}$ = 20.4 min, $^{13}$N with t$_{1/2}$ = 9.97 min and $^{15}$O with t$_{1/2}$ = 122 s) and short-lived ($^{12}$N with t$_{1/2}$ = 11 ms, $^{29}$P with t$_{1/2}$ = 4.14 s and $^{38mK}$ with t$_{1/2}$ = 924 ms) $\beta^{+}$ emitters. The methods are based on the multifoil activation technique combined with dynamic PET scanner imaging that is performed either in-beam (for short-lived $\beta^{+}$ emitters) or outside the irradiation room (for long-lived $\beta^{+}$ emitters). The technique has been first validated at the 18 MeV cyclotron of CNA in Spain $^{[10]}$, and then applied up to a nominal proton beam energy of 200 MeV at the WPE and HIT clinical facilities in Germany; in the first case using a commercial PET/CT scanner for $\beta^{+}$ long-lived emitters and in the latter using a state-of-the-art miniPET made of four i-TED $^{[11][12]}$ detectors for short-lived $\beta^{+}$ emitters. The results from both experimental campaigns will be presented and the relevance of the new data for PET range verification will be discussed.

$^{[1]}$ A. C. Kraan et al., Frontiers in Oncology 5 150 (2015)
$^{[2]}$ U. Oelfke, G. K. Lam and M. S. Atkins, Physics in Medicine and Biology 41 177 (1996)
$^{[3]}$ K. Parodi et al., Physics in Medicine and Biology 52 3369-3387 (2007)
$^{[4]}$ Tárkányi et al. J. Radioanalytical and Nuclear Chemistry. 319 533–666 (2019)
$^{[5]}$ S. España et al., Phys. Med. Biol. 56(9) 2687-2698 (2011)
$^{[6]}$ E. Seravalli et al., Phys. Med. Biol. 57 1659 (2012)
$^{[7]}$ P. Dendooven et al., Phys. Med. Biol. 64 129501 (2019)
$^{[8]}$ H. J. Buitenhuis et al., Phys. Med. Biol. 62 4654 (2017)
$^{[9]}$ I. Ozoemelam et al., Phys. Med. Biol. 65 245013 (2020)
$^{[10]}$ T. Rodríguez-González et al., Rad. Phys. and Chem. 190 109759 (2022)
$^{[11]}$ C. Domingo-Pardo. Nucl. Inst. and Methods A 825 78-86 (2016)
$^{[12]}$ V. Babiano et al., Nucl. Inst. and Methods A 953 163228 (2020)

Speaker: Maria Teresa Rodriguez Gonzalez (University of Sevilla)
• 12:21
Reaction cross sections of prompt-gamma and PET radioisotopes for range verification in proton therapy 12m

Although protontherapy is advantageous over more traditional radiotherapy from the point of view of dose delivery and sparing of organs at risk, its full potential has not been reached yet [1]. A lot of effort is focused on proton range verification techniques to improve dose localization. Several of these techniques profit from the secondary emission induced by protons to identify the proton range and to estimate the dose deposited in patients [2]. They include the generation and detection of PET radioisotopes, and the production of prompt gammas (PG) by proton-induced reactions. It is nonetheless crucial to have reliable cross section values of the more interesting reaction channels. The radiation induced on natural tissues is not always the most suitable to perform proton range verification. Thus the use of contrast agents that provide an increased induced radioactivity near the Bragg peak region has been suggested to improve the range verification capabilities [3,4]. Our studies show several promising candidates.
Water-18 (H218O) has a great potential as a contrast agent both in PET and in PG thanks to the oxygen 18O isotope. As a PET contrast agent the most interesting reaction channel is the (p,n) one, which produces 18F, which is a β⁺ emissor with a halflife of 109 min. The cross section values up to 10 MeV are well known because of its medical use as radiotracer. On the contrary, the relevant energies in protontherapy amount up to 220 MeV and therefore it is necessary to know the cross section values up to this energy. We will report on measurements performed using the 220 MeV cyclotron at the Quirónsalud protontherapy center in Madrid. Samples were irradiated in the beam at several energies and measured offline using a setup based on fast LaBr3(Ce) scintillator detectors coupled to a digital acquisition system. The results are illustrated in Fig. 1 (left), where measurements of an irradiated 100 μl water-18 sample was irradiated at 150 MeV. We will report on the cross section results and on the benchmarking measurements carried out at the 5-MV tandetron accelerator of the Center for Micro Analysis of Materials (CMAM) [5] in Madrid (Spain).
Prompt gamma emission from water-18 also is promising for proton range verification due to the presence of intense, discrete γ-rays. We have performed measurements of PG production at low energies at CMAM in the energy range 1–10 MeV using a set-up consisting of two pairs of collinear opposed LaBr3(Ce) detectors and a fully digital acquisition system with high-rate capabilities. We will report on results, shown in Fig. 1 (right), which highlight the presence of prompt γ-rays from 18O visible above background for the irradiation of an admixture of distilled water (93)% and water-18 (7%).

Figure 1: (Lef) PET activity as a function of time for a water-18 sample irradiated at Quirónsalud at 150 MeV. (Right) PG spectrum at 8.8 MeV measured at CMAM.

Another promising candidate for proton range verification using contrasts is Iodine. It is routinely used as a contrast agent in several other medical applications. Proton-induced reactions produce 127mXe with 69-s halflife, via the (p,n) channel, which can be used for online range verification [6]. Since the reaction cross sections were not known down to the low energies relevant for the Bragg peak we have performed irradiations in the 4–10 MeV range in the external microbeam line at CMAM. The measurements made use of the on-line setup based on LaBr3(Ce) detectors mentioned above. We will report on the results and the comparison to model calculations, which support the viability of 127I as a contrast agent for proton radiotherapy. Concerning the production of prompt gammas we will report on the measurements carried out at CMAM at low energies, and at the Quirónsalud cyclotron up to 220 MeV proton energies.

[1] Knopf, A., Lomax, A., Phys. Med. Biol. 58 (15), R131, 2013.
[2] H. Paganetti, Phys. Med. Biol. 57 (11), R99, 2012.
[3] L.M. Fraile et al., Nucl. Instrum. Methods A 814, 110–116, 2016.
[4] PRONTO-CM, 2020. Protontherapy and Nuclear Techniques for Oncology.
[5] A. Redondo-Cubero et al., Eur. Phys. J. Plus, 136:175, 2021.
[6] A. Espinosa Rodriguez et al., Radiation Physics and Chemistry 185, 109485, 2021.

• 12:33
Multi-Feature Treatment Verification in Particle Therapy 12m

Particle therapy is a promising and rapidly developing methodology of modern tumor treatment. In order to reach its full potential, however, it requires detailed verification that the clinical target volume receives the planned dose while sparing surrounding healthy tissues.

Although the applicability of in-beam positron emission tomography and prompt gamma rays has already been demonstrated in patients, range verification is not yet part of the clinical routine in particle therapy. This is due both to the methodological limitations of previous systems, but also to commercial, clinical and physical boundary conditions.

In pencil beam scanning, the most advanced particle therapy treatment method, the number of secondary particles available per spot ($\Delta t =$ 10 to 100 ms) is limited. In order to develop a clinically usable treatment verification system, as much information as possible must be extracted from the secondary radiation field. On the other hand, the instantaneous fluence rate of secondary particles is high ($5\cdot10^6$ to $1\cdot10^8$ cm$^{-1}$ s$^{-1}$), which challenges modern digital data acquisition systems connected to monolithic inorganic scintillators with typical sizes used in prompt gamma ray verification systems. In order to reduce the load on the detectors, and also with regard to the ever increasing current intensities of next generation medical accelerators, future systems have to be granular. Multi-Feature Treatment Verification (MFTV) combines and extends established methods (prompt gamma-ray imaging, spectroscopy, timing, etc.) in order to achieve higher reliability and performance.

This idea was taken up by the NOVO project and expanded by a multi-particle approach. The NOVO (A novel and holistic approach to real-time dose verification enabling a new era of radiotherapy for cancer treatment) consortium is a collaboration of medical, nuclear and detector physicists, nuclear engineers, and mathematicians, and aims at developing a holistic real-time treatment verification system in particle therapy. In addition to the development of the multi-particle detector, the NOVO consortium focuses on the development of crucial elements of a clinical system including image reconstruction methods, deterministic predictors, dose estimators, decision making schemes and the determination of neutron and photon production cross-sections of therapy-relevant isotopes in the proton energy range of 70 to 230 MeV.

Elements of a potential multi-channel MFTV system were characterized in a double time-of-flight experiment at the pulsed photo-neutron source nELBE. The essential properties (time resolution, light yield, detection efficiency and pulse shape discrimination) of EJ-276 plastic scintillators and novel organic glass scintillators from Sandia National Laboratories were determined. Organic glasses are reported to have higher light yield, improved pulse shape discrimination (PSD) properties and faster fall times compared to commercially available plastic scintillator materials such as the EJ-276. Therefore, organic glasses represent an interesting alternative to materials currently available in the market.

The first experimental results show that the time resolution of a $10\times10\times200$ mm$^3$ organic glass scintillator read-out at both ends will reach the high demands of such a proposed range verification system. Preliminary results show a time resolution of $\approx$390 ps, an energy resolution of $\approx$ 22 % / 17 % at 340 keV / 1061 keV and a pulse-shape discrimination figure-of-merit greater than 1.03 for separating neutrons and $\gamma$-rays over the entire usable kinetic energy range of nELBE. These properties make organic glasses a promising candidate for an MFTV system.

Speaker: Dr Toni Kögler (OncoRay – National Center for Radiation Research in Oncology)
• 12:45
Ambient neutron dosimetry in particle therapy facilities 12m

Neutrons are a highly penetrating radiation which can dominate the total adsorbed dose by the human body. Therefore, in different kind of facilities, it is essential to monitor neutron dose rates in order to ensure a minimal risk for workers, patients and public. This task is typically achieved by using commercial portable neutron detectors known as ambient neutron dosimeters. However, there are several concerns about the reliability of commercial ambient neutron dosimeters in modern facilities. In particular, those producing radiation fields with an important high energy contribution (E>20MeV) or a complex time structure [1, 2]. Reliability is a major issue indeed for the dosimetry and radiation protection in: 1) medical facilities, such as proton therapy, where high energy neutrons up to 250 MeV are produced as secondary stray radiation; 2) around synchrotron or cyclotron facilities, where either intended pulsed beams or beam losses produce short bursts of secondary neutron radiation; 3) pulsed facilities for fundamental research and applications, such as spallation, fusion neutron sources or high intensity lasers. Moreover, the International Commission on Radiation Units and Measurements (ICRU) has recently recommended the use of alternative definitions for the operational quantities that are currently in use for radiation protection. The new ICRU recommendation impacts directly on the expected performance of neutron dosimeters for energies lower than 100 eV and higher than 50 MeV. Accordingly, the LINrem project has been launched in 2018 in order to provide solutions to the new requirements of energy sensitivity and time resolution in neutron dosimetry with an special focus on medical applications.
In this work, the technical challenges for active and time-resolved neutron dosimetry in particle therapy are reviewed. It is discussed the impact of the new ICRU recommendation on the assessment, using commercial dosimeters, of neutron doses in hadron therapy. The nuclear data needs for novel high energy detector designs are also discussed. The conceptual design of LINrem dosimeters is presented and its performance is quantified by means of Monte Carlo simulations. Results are presented of the experimental validation of LINrem dosimeters in different facilities including proton therapy and pulsed neutron sources.
References
[1] J. Farah et al., Med. Phys. 42:5 (2015) 2572-2584.
[2] M. Caresana et al., NIMA 737 (2014) 203–213.

Speaker: Ariel Tarifeno-Saldivia
• 12:57
(WITHDRAWN) Nuclear data needs for the new era of Boron Neutron Capture Therapy 12m

Boron Neutron Capture Therapy (BNCT) is facing a renaissance with the development of accelerator-based neutron sources in different countries [1,2]. These facilities can be placed in hospitals, overcoming the main barrier in the expansion of this promising therapy, which up to now has been only performed at research reactors. BNCT is the only external radiotherapy option which is selective at the cellular level and is delivered in just one day, without any fractioning.

These new facilities are based on the production of neutrons by means of proton collisions on lithium or berillium targets, then, the produced neutron beam is moderated and collimated by a beam shaping assembly (BSA) providing an adequate neutron beam for the patient treatment. Each BSA must be specifically designed for each combination of proton energy and neutron production target (example in Ref. [3]). This allows to adjust the neutron spectrum in the optimal epithermal energy region. The spectrum is key in the dose delivered to the patient, both at the tumor and at the organs at risk.

There are different nuclear data needs for the improvement of the different elements involved in the therapy. First, there are different neutron interactions with the most common materials of the BSA (especially inelastic scattering processes) that are key for optimizing the final spectrum by means of Monte Carlo simulations. In addition, the cross sections of other neutron induced reactions of interest for the dosimetry and experimental spectrometry via multiactivation in the epithermal range are also not well known. Finally, the dose calculation at tumor and normal tissues requires accurate (with less than 5% uncertainty) data of the most important reactions leading to energy releases. In this sense, the n_TOF collaboration has recently measured the cross sections of the reactions 14N(n,p), 35Cl(n,p) and 35Cl(n,g) of interest for the determination of the normal tissue absorbed dose, especially at skin and brain.

In this contribution other cross sections corresponding to reactions of interest for BNCT will be mentioned and the current knowledge on them will be analysed.

[1] A.J. Kreiner et al., Present status of Accelerator-Based BNCT, Rep. Pract. Oncol. Radiother. 21, 95-101 (2016).
[2] I. Porras et al., Perspectives on Neutron Capture Therapy of Cancer, in Proceedings of the 15th International Conference on Nuclear Reaction Mechanisms, edited by F. Cerutti, A. Ferrari, T. Kawano, F. Salvat-Pujol, and P. Talou, CERN-Proceedings-2019-001 (CERN, Geneva, 2019), pp. 295-304.
[3] P. Torres-Sánchez, I. Porras, N. Ramos-Chernenko, F. Arias de Saavedra and J. Praena, Optimized beam shaping assembly for a 2.1-MeV proton-accelerator-based neutron source for boron neutron capture therapy. Sci Rep. 11:7576 (2021) https://doi.org/10.1038/s41598-021-87305-9

Speaker: Ignacio Porras (Universidad de Granada (ES))
• 11:45 13:15
Reactor Data: II Placerville ()

### Placerville

Convener: Robert Mills
• 11:45
Beta spectrum shape studies for the predictions of the antineutrino spectrum from reactors 24m

Nuclear reactors are copious sources of antineutrinos. This is the main reason why reactors have played a key role in the discovery of the neutrino and in the study of neutrino oscillation phenomena. The comparison of the neutrino flux measurements performed at short baselines with the improved conversion procedure by Huber [1] and Mueller [2] led to the discovery of the so-called reactor antineutrino anomaly [3], a surplus in the number of predicted antineutrinos compared with the measured, and more recently, to the finding of a spectrum distortion around 5 MeV [4]. There is another alternative for the prediction of the antineutrino spectrum, the summation calculation method, which has recently achieved a precision comparable to the conversion procedure [5].

Both methods of calculations of the antineutrino spectrum require assumptions on the shape of the beta transitions. In [6,7] it was argued that not taking into account the shape of first forbidden transitions could itself explain a large part of the discrepancy related to the anomaly.

With this motivation in mind, we have developed a new setup to measure the beta shape of the beta spectrum of relevant decays for the calculation of the antineutrino spectrum. In this presentation, the new setup, and some preliminary measurements performed at Jyväskylä with pure isotopic beams will be presented. The possible impact of the measurements in the calculations will be discussed.

[1] P. Huber, Phys. Rev. C 84, 024617 (2011).
[2] Th. A. Mueller, et al., Phys. Rev. C 83, 054615 (2011).
[3] G. Mention, et al., Phys. Rev. D 83, 073006 (2011).
[4] Double Chooz and Reno Collaborations in Proceedings of the Neutrino 2014 Conference, http://neutrino2014.bu.edu/; Daya Bay Collaboration in Proceedings of the ICHEP 2014 Conference, http://ichep2014.es/.
[5] M. Estienne, et al., Phys. Rev. Lett. 123, 022502 (2019).
[6] A. C. Hayes, et al., Phys. Rev. Lett. 112, 202501 (2014).
[7] L. Hayen, et al., Physical Review C 100, 054323 (2019).

Speaker: Gustavo Alcalá (Instituto de Física Corpuscular, CSIC-Univ. de Valencia, E-46071 Valencia, Spain.)
• 12:09
Total absorption beta decay studies for reactor decay heat calculations 12m

A. Algora1,2, V. Guadilla3, J. L. Tain1, M. Fallot4, M. Estienne4, A. Porta4, L. Giot4, L. Le Meur4, A. Beloeuvre et al. for the DTAS Jyvaskyla Collaboration

1)Instituto de Física Corpuscular, CSIC-Univ. de Valencia, E-46071 Valencia, Spain
2)Institute of Nuclear Research (ATOMKI), P.O. Box 51, H-4001 Debrecen, Hungary
3)Faculty of Physics, University of Warsaw, 02-093, Warsaw, Poland
4)Subatech, IMT-Atlantique, Univ. de Nantes, CNRS-IN2P3, F-44307, Nantes, France

Total absorption spectroscopy provides beta decay data free from the Pandemonium systematic error [1-3]. Avoiding this systematic error is of great relevance for nuclear structure, astrophysics and practical applications [3].

In this contribution we will present an overview of recent results from the research work performed by our collaboration employing this technique, which is relevant for reactor applications, in particular for decay heat summation calculations. The measurements, we are presenting here, have been performed at the University of Jyväskylä IGISOL IV Facility [4] using trap-assisted spectroscopy that provided radioactive beams of very high isotopic purity [5].

In this presentation we will emphasize mainly on highlights coming from our experimental campaign performed in 2014 [6-10] where special emphasis was devoted to cases that required isomeric separation [8]. This data has been shown to be of relevance in calculations of the decay heat and in calculations of the antineutrino spectrum from reactors [3,11]. The impact of the new results in nuclear structure and astrophysics will also be discussed. Some of the studied cases are beta delayed neutron emitters, for which gamma competition above the neutron separation energy has been determined [9]. In the framework of these analyses a new procedure to determine ground state to ground state feedings has also been introduced [10].

The impact of all the measurements performed until know by our collaboration for reactor decay heat calculations will be also discussed [12-16] and future perspectives presented.

[1] J. C. Hardy et. al., Phys. Lett. B 71, 307 (1977)
[2] B. Rubio et al., Journal of Physics G: Nuclear and Particle Physics 31, S1477 (2005)
[3] A. Algora et al., Eur. Phys. J. A 57, 85 (2021)
[4] I. D. Moore et al., Nucl. Instrum. and Methods B 317,208 (2013)
[5] T. Eronen et al., Eur. Phys. J. A 48, 46 (2012)
[6] V. Guadilla, PhD Thesis, Univ. of Valencia, Spain, (2017)
[7] L. Lemeur, PhD Thesis, Univ. of Nantes, France, (2018)
[8] V. Guadilla et al., Phys. Rev. Lett. 122, 042502 (2019)
[9] V. Guadilla et al., Phys. Rev. C 100, 044305 (2019)
[10] V. Guadilla et al, Phys. Rev. C 102, 064304 (2020)
[11] M. Estienne et al., Phys. Rev. Lett. 123, 022502 (2019)
[12] A. Algora et al., Phys. Rev. Lett. 105, 202501 (2010)
[13] D. Jordan et al., Phys. Rev. C 87, 044318 (2013)
[14] A. A. Zakari-Issoufou et al., Phys. Rev. Lett. 115, 102503 (2015)
[15] J. L. Tain et al., Phys. Rev. Lett. 115, 062502 (2015)
[16] E. Valencia et al., Phys. Rev. C 95, 024320 (2017)
[17] S. Rice et al., Phys. Rev. C 96, 014320 (2017)
[18] V. Guadilla et al., Phys. Rev. C C 100, 024311 (2019)

Speaker: A Algora (1) Instituto de Física Corpuscular, CSIC-Univ. de Valencia, E-46071 Valencia, Spain)
• 12:21
Neutron cross section measurements for burn-up credit approaches 12m

Criticality safety analysis is needed at various stages of the nuclear fuel cycle, including the back-end of the fuel cycle, i.e. reprocessing, transport, storage and final disposal of spent nuclear fuel (SNF). In the past, criticality safety assessments for SNF management for out-of-reactor applications assumed that the fuel is in its most reactive condition, which is usually prior to irradiation of the fuel in a reactor. Since the 80's, an effort is being made by the nuclear industry, research organisations and regulatory authorities to use more realistic and less conservative estimates of the nuclear reactivity of SNF by accounting for the reduction in reactivity due to fuel burnup in a (sub-)criticality analysis. This concept is referred to as Burn Up Credit (BUC).
Specific programmes to validate nuclear data of fission products that are important for criticality safety studies based on a BUC approach reveal shortcomings in these data. To improve the status of cross sections for fission products for BUC approaches a dedicated programme was defined as part of a collaboration between the CEA Cadarache (France) and the JRC Geel (Belgium). It includes experiments at the time-of-flight facility GELINA to characterise the samples that were used for the MINERVE experiments [1,2,3] and to produce accurate total and capture cross section data to improve the evaluated data in the resonance region. This collaborative effort triggered the interest of other institutes and organisation such as the IFIN-HH (Romania), INFN Bologna (Italy) and INRNE (Bulgaria) by participating in the experiments at GELINA and assisting in the data reduction and analysis.
In this report, the status of cross section data for $^{103}$Rh, $^{107,109}$Ag and $^{155,157}$Gd is discussed and results of a resonance shape analysis of new capture and transmission experiments at GELINA are presented. In addition, recommendations to improve the evaluated cross section data in the resonance region for these nuclides are given.

[1] Gruel et al., “Interpretation of fission product oscillations in the MINERVE reactor, from thermal to epithermal spectra”, Nucl. Sci. Eng. 169 (2011) 229.
[2] Šalamon et al., “Neutron resonance transmission analysis of cylindrical samples used for reactivity worth measurements”, J. Radioanal. Nucl.Chem. 321 (2019) 519.
[3] Fei Ma et al., “Non-destructive analysis of samples with a complex geometry by NRTA”, J. Anal. At. Spectrom. 35 (2020) 478.

Speaker: Carlos Paradela (European Commission, Joint Research Centre (JRC))
• 12:33
140,142Ce Neutron Cross Section Resolved Resonance Region Evaluation 12m

A resolved resonance region evaluation of 140,142Ce has been carried out by Oak Ridge National Laboratory. Requested by the US Nuclear Criticality Safety Program, this evaluation is based on recent high-resolution transmission and capture high-resolution measurements of natCe and 142Ce conducted at JRC-GEEL at the Geel Linear Accelerator facility, as well as recently measured thermal constants available from the EXFOR database [1]. Starting from the resonance parameters from the ENDF/B-VIII.0 library [2] and following a preliminary R-matrix analysis [3], an updated set of resonance parameters and corresponding covariance information was derived by the fit of these experimental datasets using the Reich-Moore approximation of the R-matrix theory as implemented in the SAMMY code system [4]. The resolved resonance region upper energy limit for 140Ce was kept at 200 keV while the 142Ce resonance region was extended from 13 to 26 keV. This new evaluation was found to be in good agreement not only with several integral quantities of interest to the reactor physics community, but also with the stellar Maxwellian-averaged cross section [5].

Acknowledgments

This work was supported by the U.S. Department of Energy (DOE) Nuclear Criticality Safety Program, which is funded and managed by the National Nuclear Security Administration for DOE.

• This manuscript has been authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 with the US Department of Energy (DOE). The US government retains and the publisher, by accepting the article for publication, acknowledges that the US government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for US government purposes. DOE will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan (http://energy.gov/downloads/doe-public-access-plan).}

References

[1] V. V. Zerkin, B. Pritychenko, “The experimental nuclear reaction data (EXFOR): Extended computer database and Web retrieval system,” Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 888, pp. 31-43 (2018).
[2] D. A. Brown et al, “ENDF/B-VIII.0: The 8th Major Release of the Nuclear Reaction Data Library with CIELO-project Cross Sections, New Standards and Thermal Scattering Data,” Nuclear Data Sheets, 148, pp. 1-142 (2018).
[3] C. W. Chapman, M. T. Pigni, K. Guber, “Progress on 140,142Ce Neutron Cross Section Resolved Resonance Region Evaluations,” ICNC2019-11th International Conference on Nuclear Criticality Safety, 15-20 September 2019 - Paris, France.
[4] N. M. Larson, “Updated user’s guide for Sammy: Multilevel R-matrix fits to neutron data using Bayes’ equations,” Tech. Rep. ORNL/TM-9179/R8, ENDF-364/R2, Oak Ridge National Laboratory (2008).
[5] B. Pritychenko, S. F. Mughabghab, “Neutron Thermal Cross Sections, Westcott Factors, Resonance Integrals, Maxwellian Averaged Cross Sections and Astrophysical Reaction Rates Calculated from the ENDF/B-VII.1, JEFF-3.1.2, JENDL-4.0, ROSFOND-2010, CENDL-3.1 and EAF-2010 Evaluated Data Libraries,” Nuclear Data Sheets, 113, pp. 3120-3144 (2012).

Speaker: Chris Chapman (Oak Ridge National Laboratory)
• 12:45 13:35
Education and Outreach Sutter's Fort ()

### Sutter's Fort

Convener: Lee Bernstein (UC Berkeley)
• 12:45
GRE@T-PIONEeR: teaching the nuclear data pipeline using innovative pedagogical methods 24m

GRE@T-PIONEeR - GRaduate Education Alliance for Teaching the PhysIcs and safety Of NuclEar Reactors - is a new project funded by the Euratom – Horizon 2020 Framework Programme. The project started on November 1st, 2020 for a duration of three years and gathers eight universities throughout Europe.

The project aims at developing and providing specialised and advanced courses in computational and experimental reactor physics at the graduate level (MSc and PhD levels) and post-graduate level, as well as the staff members working in the nuclear industry.

Although reactor physics has always been a core discipline in nuclear engineering, computational reactor physics relies on sophisticated models, databases and algorithms, which the engineer needs to understand, so that the tools are used most efficiently and in relevant applications. Moreover, these computational courses are often taught via advanced courses with fewer students.

One of the Work Package of GRE@T-PIONEeR is devoted to a specific course on the nuclear data pipeline processes and the role of nuclear data for calculations of nuclear reactor systems. It covers all steps starting from the measurements to their validation and final use in nuclear reactor calculations. The main topics covered in this course are: i) the process of generation and evaluation of nuclear data libraries, ii) the processing of nuclear data libraries for use in nuclear applications, iii) the assessment of nuclear data uncertainties, iv) the importance of nuclear data adjustments, v) and, finally, the presentation of activities, projects and international networks on nuclear data.

Beyond the technical contents of the courses being developed, the novelty of the project lies with the use of innovative pedagogical methods, such as flipped classes, aimed at promoting student learning. Before attending interactive sessions organised under the close supervision of the teachers, the students will have access to a handbook with the whole content of the course jointly with short videos summarising the key concepts, and online quizzes allowing testing one’s understanding of those concepts.

The interactive sessions are based on active learning, during which the students have to implement and use the techniques they learned in hands-on training exercises designed to promote learning. These exercises are computer-based assignments with existing open software and tools such as ENDF Utilities, NJOY, PREPRO, JANIS, DICE and NDaST. Finally, the students will be introduced to nuclear data adjustments algorithms and machine learning advanced techniques applied to the nuclear data field.

Furthermore, a set of experiments with direct application to cross-sections measurements will be carried out in research reactors: i) activation analysis experiments for a set of irradiated foils , ii) Doppler feedback measurement in the BME Training Reactor at the Budapest University of Technology and Economics, and iii) reactivity worth of samples using pile-oscillation techniques in the research reactors: AKR-2 at the Technical University Dresden and CROCUS at the École Polytechnique Fédérale de Lausanne.

In addition to the flipped classroom pedagogy, most of the interactive sessions are offered in a hybrid format: the students can decide to attend the interactive sessions either on-site or online. The sessions are also given in a condensed format. Combined with the hybrid set-up, the courses are very well suited for lifelong learning.

Acknowledgement. This project has received funding from the European Union’s Euratom research and training programme 2019-2020 under the Grant Agreement n°890675.

Speaker: Oscar Cabellos
• 13:09
Educational activity by using nuclear chart in Japan: constructing three-dimensional nuclear chart and distributing nuclear charts through crowdfunding 12m

(1) JAEA nuclear chart
JAEA has published a paper-based nuclear chart which is a sheet of 12 or 16 pages in A4 size, every four years since 1976. The chart is a two-dimensional chart that shows the number of protons on the vertical axis and the number of neutrons on the horizontal axis in order to represent the properties of various nuclei, and we distribute them to (mainly) researchers on nuclear science.1)

(2) Three-dimensional nuclear chart and educational activities
Nuclear chart is very useful, however, such a character-based chart is not rather friendly to be used. We constructed three-dimensional (3D) nuclear charts created using toy blocks, which represent the atomic masses per nucleon number and the total half-lives as heights of blocks for each nucleus in the entire region of the nuclear mass to visualize them intuitively2) (Fig.). The size is 96 (width) x 56 (depth) x 30 (maximum height) cm. By using the atomic-mass chart, for example, generation of atomic energy from nuclei, like nuclear fusion and fission, can be understood as a slope of the masses. Subsequently, these charts were used in outreach activities for the general public and high school students. As an example, an application for a lecture on nucleosynthesis in stars is introduced, and some explanations for the abundance of iron and the origin of uranium and heavy elements on the Earth are given with the 3D chart.
Using our charts, lectures entitled 'Alchemy of the Universe' were delivered to various places (science café, science lecture in high school, etc.) more than twenty times till now. We made some video movies3).

(3) Crowdfunding for distributing nuclear chart to high school students in Japan
In 2020, we started a crowdfunding project “one nuclear chart for one high school."4) The purpose of the project is to deliver nuclear chart to high schools in Japan. The projected budget is 1.5 million Yen (~15,000 USD as 1USD=100Yen) and the term of donation is set from Jan. to Mar. in 2020 (8 weeks). Finally, it was succeeded: the total amount of the donation reached to 1.731 million Yen (~17,310 USD) by 157 supporters. After that, we constructed 1,200 sheets of the chart written in Japanese, and distributed to 284 science-programmed high-schools, namely super science high school (SSH) approved by the ministry for education, overall Japan, 122 high schools in Fukushima prefecture, 98 high schools in Ibaraki prefecture, and 57 colleges of national institute of technology.

In our paper, we will show our educational activities with nuclear chart in Japan, including an introduction of the 3D charts and a crowdfunding project, and some examples of outreach lectures for heigh-schools.

Figure: Three-dimensional nuclear chart. The height of the block represents atomic mass excess per nucleon. The color of the block represents the length of half-lives of nuclei.

References
1) H. Koura, T. Tachibana, J. Katakura, F. Minato, Chart of the nuclides 2014, Japanese Nuclear Data Committee and Nuclear Data Center of JAEA (2015).
https://www.jaea.go.jp/02/press2014/p15031202/ (in Japanese)
2) H. Koura, Three-dimensional nuclear chart - understanding nuclear physics and nucleosynthesis in stars, Phys. Educ. 49 (2014) 215: Three-minute video abstract can also be seen:
https://iopscience.iop.org/article/10.1088/0031-9120/49/2/215
3) The promotion division of JAEA: (1) ‘Cosmic alchemy’ (2013), and (2) ‘Contributed to the Synthesis of New Element NIHONIUM - The World of Heavy Element Nuclear Science’ (2017), in the series of the JAEA official movie ‘JAEA Channel’
http://www.jaea.go.jp/english/jaea_channel/
4) A crowdfunding project “one nuclear chart for one high school” since 2020

Speaker: Hiroyuki Koura
• 13:21
The Nuclear Data Working Group in the United States and Workshop Outcomes 12m

The Nuclear Data Working Group (NDWG) was established in 2015 to identify cross-cutting nuclear data needs, and to facilitate communication and collaboration across programs on nuclear data activities. The NDWG currently consists of over forty members who represent the needs of program offices across the Department of Energy, the National Nuclear Security Administration, and other programs. National Laboratory representatives have been added to the NDWG to contribute knowledge of laboratory research needs and priorities. The NDWG meets annually to plan outreach activities and discuss collaboration on cross-cutting nuclear data priorities. The Nuclear Data Interagency Working Group (NDIAWG) is a federal program manager group led by the Department of Energy, Office of Science, Office for Nuclear Physics (NP). They meet regularly to coordinate funding efforts for nuclear data. Several of the programs collaborate through a multi-program Funding Opportunity Announcement managed by NP. The NDWG maintains a website that contains information about the NDWG activities and provides access to workshop reports, nuclear data needs documents and introductory presentations. The website link can be found on the National Nuclear Data Center webpage, and at www.nndc.bnl.gov/ndwg/.
The primary outreach mechanism of the NDWG is the annual Workshop for Applied Nuclear Data Activities (WANDA). Since 2015, the NDWG has organized six WANDA workshops which bring together nuclear data experts, the user community and program managers. The workshops include six to eight topical breakout sessions that encourage input from all attendees through facilitated discussion to identify and prioritize nuclear data needs, and to recommend tasks to improve the nuclear data. The recommendations are summarized in a workshop report. Many of the nuclear data tasks recommended in the WANDA workshops have received funding demonstrating the success of this type of collaborative interaction. This presentation will summarize the recommendations from the WANDA workshops and discuss recent outreach activities and future plans.

Speaker: Catherine Romano
• Tuesday, 26 July
• 06:00 07:30
Facilities: I Sutter's Fort ()

### Sutter's Fort

Convener: Katelyn Cook (RPI)
• 06:00
Back-n Facility Status and Recent Nuclear Data Measurements 24m

The Back-n white neutron source at CSNS (China Spallation Neutron Source) has been in operation since 2018. As a multidisciplinary research platform, Back-n has its main focus on nuclear data measurements. With its unique beamline using the back-streaming neutrons from a proton beam of 100 kW in beam power impinging on a thick spallation target in tungsten, Back-n owns the highest neutron flux for a given flight path among white neutron sources. The other excellent properties include a wide neutron energy spectrum covering from 0.3 eV to 200 MeV and a good time resolution of a few per mille in most of the energy range. Currently available detector systems or spectrometers are FIXM for fission cross-section measurements, C6D6 detectors and GTAF-II for neutron capture measurements, LPDA for light-particle emission measurements, NTOX for total cross-section measurements, and user-owned HPGe detectors for in-beam gamma spectrum measurements. All the above types of experiments have been carried out regularly at Back-n with an averaged beam time of 3000 hours per year. In addition, the experimental study on the time reversal violation with neutrons is also under way. This presentation will introduce the improvements of the Back-n facility in the last years, including proton beam power ramp-up, enhanced measurements on the neutron energy spectrum, detector systems, and also a summary of the nuclear data measurements.

Speaker: J. Tang (CSNS/CADS/IHEP)
• 06:24
Design, construction, commissioning and early operation of the third-generation n_TOF neutron spallation target at CERN 12m

The European Laboratory for Particle Physics (CERN) is equipped with a top-class, high-brightness, neutron spallation source dedicated to high-resolution neutron time-of-flight experiments: the n_TOF Facility.

During CERN’s Long Shutdown 2 (LS2, 2019-2021), the n_TOF neutron spallation target has been exchanged and the facility is now operating with its third-generation target [1]. The neutron production target is based on a segmented pure Pb (>99.9%) core, cooled by gaseous nitrogen at atmospheric pressure. The target has been designed for a single bunch proton beam impacting on the production target at 20 GeV/c with up to 1013 protons/bunch, and with a bunch length of 6 ns RMS. The produced neutrons span 11 orders of magnitude in kinetic energy, from sub-thermal to GeV and are serving two time-of-flight experimental stations plus an irradiation station. The initially fast neutrons are moderated by two independent moderator circuits, each one dedicated to an experimental station, that can be filled with either by demineralized water or by borated water in saturation.

The contribution will detail the physics optimization and engineering design processes that brought the facility from the second-generation target to the current evolutionary one. The work will include a description of the new target design features, its mechanical assembly, and its compliance with safety and physics performance requirements as well as the R&D which has been carried out, including beam testing at the CERN’s HiRadMat facility [2]. The contribution will also detail the first hardware results coming from the 2021 facility commissioning, first radiation protection assessments in the target zone and from early operation in 2022.

The paper will also discuss the plans for an autopsy of the second-generation target to investigate the status of the water-cooled monolithic Pb core after 10 years of operation and its impact on the radioactive waste characterization and in the operation of the third-generation target.

[1] R. Esposito et al., Design of the third-generation lead-based neutron spallation target for the neutron time-of-flight facility at CERN, Phys. Rev. Accel. Beams 24, 093001 (2021)

[2] R. Esposito and M. Calviani, Design of the third-generation neutron spallation target for the CERN’s n_TOF facility, J. Neutron Res. 22, 221 (2020)

Speaker: Michael Bacak
• 06:36
Status and perspectives of the n_TOF Facility at CERN following the upgrades and consolidation during CERN’s Long Shutdown 2 12m

The European Laboratory for Particle Physics (CERN) is equipped with a top-class, high-brightness, neutron spallation source dedicated to high-resolution neutron time of flight experiments: the n_TOF Facility.

The Facility has been constructed in 2000 [1] and has evolved significantly over the last 20 years, encompassing the creation of a new vertical experimental area on top of the spallation target [2] and the capability to handle unsealed radioactive samples in the two experimental stations. During CERN’s Long Shutdown 2 (2019-2021), a significant upgrade has been implemented to guarantee reliable operation of n_TOF for the years to come, maximizing the physics reach of the infrastructure.

The contributions will detail the upgrades and the status of the facility, including the construction of a third-generation spallation target [3], the consolidation of the neutron collimation systems, the complete overhaul of the target pit shielding as well as the realization of a new irradiation station (NEAR) for materials very close to the neutron spallation target.

Perspectives for operation and considerations for further upgrades of the facility will be provided as well.

[1] C. Borcea et al., Results from the commissioning of the n_TOF spallation neutron source at CERN, Nucl. Instr. Meth. A 513 (3), 524-537 (2003)

[2] M. Sabaté-Gilarte et al., High-accuracy determination of the neutron flux in the new experimental area n_TOF-EAR2 at CERN, Eur. Phys. J A 53, 210 (2017)

[3] R. Esposito et al., Design of the third-generation lead-based neutron spallation target for the neutron time-of-flight facility at CERN, Phys. Rev. Accel. Beams 24, 093001 (2021)

Speaker: Alberto Mengoni
• 06:48
Characterisation of the n_TOF/CERN 185m beam-line after the facility’s major upgrades 12m

The n_TOF facility at CERN is a pulsed neutron source for time-of-flight measurements, first conceptualised by Carlo Rubbia in 1998 [1]. It is designed to study neutron induced reactions bearing importance for fundamental research, nuclear astrophysics [2,3], nuclear technology applications [4] as well as nuclear medicine [5]. During its first two operation phases, n_TOF consisted of a single beam-line of 185 m length, while for its third phase in 2014, a vertical 20 m beam-line was constructed.

The facility’s 185 m flight path provides an excellent energy resolution, reaching as low as ΔE/E ~ 10-4 for the eV region [6], enabling high precision measurements that can significantly contribute to standard neutron cross section evaluations [7]. The energy range it covers stretches from a few meV up to the GeV region, allowing for significant extension of cross section data, especially for fission measurements. Furthermore, the high instantaneous neutron flux, provided by a high intensity 20 GeV/c proton beam impinging on a lead spallation target allows to measure highly radioactive samples.

Throughout CERN’s Long Shutdown 2, the n_TOF facility has been largely upgraded. The biggest change was the replacement of the water-cooled, single block lead spallation target, with a new nitrogen-cooled target consisting of several slabs of lead and optimised also for n_TOF’s vertical beam-line. After this major change, it must be ensured that the performance of the horizontal beam-line still fulfills its high standards and has not significantly changed. Additionally, the performance of the new target must be inspected thoroughly.

During the Commissioning in 2021, several detection systems, such as a fission and charged particles setup for the measurement of the neutron flux and beam profile. The energy resolution of the beam-line was investigated with neutron capture detectors complimented by extensive Monte Carlo simulations.

In this work, the characterisation of the horizontal beam-line after the facility’s spallation target replacement will be summarised. The preliminary results on the neutron flux will be presented along with a comparison with simulations as well as with the flux of the previous spallation target. Characteristics such as the energy resolution and the background conditions will also be discussed and compared to the beam-line’s previous performance.

References
[1] C. Rubbia et al., A high resolution spallation driven facility at the CERN-PS to measure neutron cross sections in the interval from 1 eV to 250 MeV, CERN/LHC/98-02 (EET), 1998
[2] U. Abbondanno et al., Neutron Capture Cross Section Measurement of 151Sm at the CERN Neutron Time of Flight Facility (n_TOF), Phys. Rev. Lett. 93, 161103 (2004)
[3] C. Lederer et al., Neutron Capture Cross Section of Unstable 63Ni: Implications for Stellar Nucleosynthesis, Phys. Rev. Lett. 110, 022501 (2013)
[4] N. Colonna, A. Tsinganis, R. Vlastou, et al. The fission experimental programme at the CERN n_TOF facility: status and perspectives. Eur. Phys. J. A 56, 48 (2020)
[5] J. Praena et al., Measurement and resonance analysis of the 33S(n,α)30Si cross section at the CERN n_TOF facility in the energy region from 10 to 300 keV, Phys. Rev. C 97, 064603 (2018)
[6] Guerrero, C., Tsinganis, A., Berthoumieux, E. et al. Performance of the neutron time-of-flight facility n_TOF at CERN. Eur. Phys. J. A 49, 27 (2013)
[7] S. Amaducci, L. Cosentino, M. Barbagallo, et al. Measurement of the 235U(n, f) cross section relative to the 6Li(n, t) and 10B(n,a) standards from thermal to 170 keV neutron energy range at n_TOF. Eur. Phys. J. A 55, 120 (2019)

Speaker: Michael Bacak
• 07:00
Characterisation of the n_TOF 20 m beam line at CERN with the new spallation target 12m

The n_TOF Collaboration operates the neutron time-of-flight facility at CERN [1], based on a 20 GeV/c pulsed proton beam impinging on a lead target employing water to moderate the spallation neutrons. The facility is characterized by a high-instantaneous neutron beam intensity, high energy resolution and a wide neutron energy spectrum, spanning from sub-thermal to GeV. The first experimental area, EAR1 [2], in operation since 2001, is located at 185 m from the spallation target nearly in the same direction as the incoming proton beam. In 2014, a new experimental area was commissioned, EAR2 [3], located at 20 m above the target in the perpendicular direction with respect to the proton beam. Thanks to the shorter flight path, this new beam line features a neutron flux around two orders of magnitudes bigger in the thermal region and about 30 times higher in the rest of the neutron energy spectrum compared to EAR1 [4]. Moreover, it presents a better signal to background ratio. This offers a unique opportunity of performing neutron induced cross section measurements for isotopes with very short half-life or small cross sections, paving the way for new challenging measurements in very diverse fields. Such is the case of the first measurement performed in EAR2 to determine the feasibility of neutron-induced fission, the 240Pu(n,f) cross section measurement [5], of special interest in nuclear waste management in the field of Nuclear Energy. In Astrophysics, the 7Be(n,α) and 7Be(n,p) reactions cross-section were measured for the first time in the energy regions of interest for the Big Bang nucleosynthesis [6-7], which are of importance to solve the Cosmological Lithium Problem or the 26Al(n,α) reaction which has a critical influence on the abundance of the 26Al cosmic γ-ray emitter [8]. Additionally, in the field of Medical Physics, the 33S(n,α) cross section was measured and data were provided for the first time from thermal to 10 keV [9].

In the course of CERN's second long shutdown (2019-2020), the facility has gone through a major upgrade comprised of the installation of a new spallation target design to fully optimise the features of the n_TOF experimental areas, unlike the previous one specifically designed for EAR1. These changes impact the characteristics of the neutron beam, i.e. the neutron flux, the energy resolution and the beam profile. The flux plays a key role in the determination of the energy dependence of the neutron induced cross-section. The energy resolution is the main feature in the characterization of the resonance region of the measured cross-sections. Compared to the previous target, the energy resolution and the characteristics of the flux are significantly improved, enhancing the capabilities of the facility EAR2.

During a commissioning phase in 2021 the changes in the characteristics of EAR2 were investigated. This work presents the features of the neutron beam with the new spallation target: neutron energy flux and neutron energy resolution. The characterisation of the neutron flux required of the combination of a series of measurements, making use of several detection systems such as micro-megas, parallel plate avalanche counters and silicon monitors, with diverse neutron-converting reactions considered standard in different energy regions. The preliminary results of the neutron flux measurement in EAR2 compared to extensive Monte Carlo simulations, as well as an overview of the improvement of the neutron energy resolution will be presented.

[1] C. Rubbia et al., A High Resolution Spallation Driven Facility ATTHE CERN-PS to Measure Neutron Cross Sections in the Interval from 1 eV to 250 MeV: a Relative Performance Assessment, CERN/LHC/98-002-EET, 1998
[2] Guerrero C. et al, Performance of the neutron time-of-ﬂight facility n_TOF at CERN, Eur. Phys. J. A 49, 2013
[3] Colonna N. et al, The Second Beam-Line and Experimental Area at n_TOF: A New Opportunity for Challenging Neutron Measurements at CERN, Nuclear Physics News, 25, 2015
[4] Sabaté-Gilarte M. et al, High-accuracy determination of the neutron flux in the new experimental area n_TOF-EAR2 at CERN, Eur. Phys. J. A 53, 2017
[5] Stamatopoulos A. et al. Investigation of the 240Pu(n,f) reaction at the n_TOF/EAR2 facility in the 9 meV–6 MeV range, Phys. Rev. C 102, 2020
[6] Barbagallo M. et al. Be(n,α)4He Reaction and the Cosmological Lithium Problem: Measurement of the Cross Section in a Wide Energy Range at n_TOF at CERN, Phys. Rev. Lett. 117, 2016
[7] Damone L. et al, 7Be(n,p)7Li Reaction and the Cosmological Lithium Problem: Measurement of the Cross Section in a Wide Energy Range at n_TOF at CERN. Phys. Rev. Lett. 121, 2018
[8] C. Lederer-Woods et al, Destruction of the cosmicγ-ray emitter 26Al in massive stars: Study of the key 26Al(n,α) reaction, Phys. Rev. Lett. C 104, 2021
[9] Sabaté-Gilarte M. et al, The 33S(n,α)30Si cross section measurement at n_TOF-EAR2 (CERN): From 0.01 eV to the resonance region, EPJ Web Conf. 146, 2017

Speaker: Jose Antonio Pavon Rodriguez (University of Seville and CERN)
• 07:12
The n_TOF NEAR Station: Physics case and commissioning 12m

A new experimental area, the NEAR Station, has recently been setup at the n_TOF facility at CERN. It is located at ~3 m distance from the spallation target, just outside the target-moderator shielding wall [1]. The main purpose of the new experimental area is to exploit the extremely high neutron flux in the vicinity of the spallation target to perform irradiation and activation measurements on short-lived radioactive isotopes or on small mass samples, of interest for Nuclear Astrophysics. The use of a suitable moderator and of filters of different thickness, is currently investigated, as is expected to produce Maxwellian-like neutron spectra of different temperatures, enabling challenging measurements of Maxwellian Averaged Cross Sections for nuclear astrophysics and other application purposes. Additionally, the extremely high intensity of the neutron beam is very convenient for measurements related to technological applications, such as material and fusion-technology related studies [2].
Following the completion of the NEAR Station [1], an extensive campaign for its commissioning has been undertaken, aiming at determining the spectral features of the neutron beam. The measurements rely on the use of the multi-foil activation method, as well as of a thermalization detector based on the activation of gold foils at different depths inside a moderator along the neutron beam direction (see contribution to this conference [3]).
In this presentation, the main characteristics of the newly built NEAR Station will be presented, together with Monte Carlo simulations of the moderator/filter assembly and of the foreseen physics program. Finally, the first results of the commissioning campaign will be reported.

References
[1] A.-P. Bernardes, et al., Design, construction, operation and first experimental results of the NEAR station at the n_TOF Facility, submitted to ND2022
[2] Mengoni, A. et al. (The n_TOF Collaboration), The new n_TOF NEAR station, Reports
No. CERN-INTC-2020-073 and No. INTC-I-222, CERN, Geneva, 2020,
https://cds.cern.ch/record/2737308
[3] P. Perez-Maroto et al., ANTILoPE: A NeuTron multi-foIL sPEctrometer for measuring neutron energy distributions up hundreds of MeV, submitted to ND2022

Speaker: Mrs Elisso Stamati
• 06:00 07:30
Fission: III American River ()

### American River

Convener: Marc Verriere
• 06:00
Dependence of TKE of fission fragments in neutron-induced fission on excitation energy by 4D Langevin mode 12m

The total kinetic energy (TKE) of fission fragments is accounting for about 80 to 90 % of the Q-value in nuclear fission. Therefore, it is important to understand the characteristics of TKE from the perspective of nuclear energy utilization and basic research, e.g., number of fission in reactor power or local heating of the r-process site including fission recycling. Naively thinking, the TKE is expected to increase, when the excitation energy of the fissioning system increases. In the reality, however, the TKE decreases as the excitation energy increases as is shown by many data on neutron-induced nuclear fission. Our 4-dimensional Langevin model has elucidated this mystery. TKE and the fission fragment deformation are related, the excitation energy dependence of TKE decreasing is discussed.

• 06:12
New Results in the Modeling of Fission and Radiative Neutron Capture with FIFRELIN 12m

The nuclear de-excitation process (through neutron, gamma and electron emission) simulated by the Monte Carlo code FIFRELIN [1] has been improved recently and compared with results from fission and radiative neutron capture experiments. Several examples will be presented during this conference.

Firstly, the initial goal of the code is to predict fission observables and associated correlations. That way a dedicated fission experiment has been performed at JRC Geel involving a fission chamber and several neutron and gamma scintillators (VESPA). FIFRELIN code has been used to calculate the relevant observables and especially the average neutron and gamma multiplicities as a function of fission fragment mass and kinetic energy [2]. Up to now it was very difficult to obtain a reasonable agreement for both neutron and gamma multiplicities within the same code using a unique set of model parameters. New calculations performed with HFB microscopic combinatorial level densities are in very good agreement with measured neutron and gamma multiplicities (as a function of mass or total kinetic energy).
Secondly, the code is used to estimate, among others, the gamma and electron cascades resulting from a neutron capture reaction. In a context of electron antineutrino spectrum emitted in a research reactor, the simulation of the STEREO compact detector was recently improved by applying the FIFRELIN 155,157Gd(nth,g) cascades [3]. As the detection efficiency has to be controlled to the %-level, a precise description of gamma cascades in Gd is necessary. The nuclear level scheme constructed from the RIPL-3 database and extended to initial capturing state at neutron separation energy has been updated with dedicated latest data from the EGAF database. Finally, a calculation of the angular distribution of gamma rays has been included in the code, with a perfect reproduction of all angular correlations [4].
In addition, FIFRELIN has been coupled with the depletion code DCHAIN to estimate delayed fission components and first results will be briefly presented [5].

[1] O. Litaize et al., Eur. Phys. J. A (2015) 51:177.
[2] M. Travar, V. Piau et al., Physics Letters B 817, 136293 (2021).
[3] H. Almazán et al., Eur. Phys. J. A (2019) 55:183.
[4] A. Chalil et al., submitted to EPJA.
[5] T. Ogawa, private communication.

Speaker: Olivier Litaize (CEA)
• 06:24
Determination of the Plasma Delay Time in PIPS detectors for fission fragments with the LOHENGRIN spectrometer 12m

The particle spectrometer VERDI (VElocity foR Direct particle Identification) allows high-precision measurements of fission fragment (FF) mass distributions by using the 2E-2v technique. VERDI consists of two arms with 16 Silicon PIPS (Passivated Implanted Planar Silicon) detectors and a Micro Channel Plate (MCP) each. The MCPs provide the pick-off signal used to trigger the time-of-flight (ToF) measurement, and the Si detectors are used both for energy detection and as a stop signal. In silicon detectors, the signal amplitude and shape get affected by the formation of a plasma from the interaction between the heavy ions and the detector material. In PIPS detectors this plasma causes a signal delay, the Plasma Delay Time (PDT), with typical delays up to 5 ns for FFs, resulting in a smearing of the mass distribution as well as increasing the systematic errors [1]. Moreover, it can render faulty fission-neutron correlations, if not properly corrected. Existing studies [2-4] propose different parameterizations to calculate the PDT contribution. These studies are not in agreement with each other and are limited to certain ion species and regions of energy. Moreover, the PDT effect is believed to be detector-specific, which necessitates a dedicated investigation.

A characterization of the PDT in the silicon detectors used in VERDI was performed at the LOHENGRIN recoil separator of the Institut Laue Langevin, for a wide range of energies and masses, $E$ $\sim$ 40-110 MeV and $A$ $\sim$ 80-160 u, respectively. With LOHENGRIN, characteristic FFs from $^{239}$Pu(n,f) were selected based on their A/q and E/q ratios. A unique collaboration allowed the utilization of a MCP detector from the STEFF spectrometer [5] and 6 silicon detectors from VERDI. The measured ToF, between the MCP and silicon, was compared to the true ToF derived from LOHENGRIN. The signals were recorded in a digital acquisition system to completely exploit the offline analysis capabilities. All the PIPS detectors were fully characterized to study their individual response to the PDT effect. The data will provide a calibration procedure in which the PDT contribution is calculated relative to alpha particles and protons. The achieved combined timing and energy resolutions of our experimental setup are around 160 ps and 0.1 MeV (FWHM), respectively.

During the presentation, we will discuss the PDT trend as a function of both mass and energy of the FF, and we will relate them to the protons and alpha particles. In addition, we will discuss the impact of the PDT correction on the fission fragment data measured with the 2E-2v method.

[1] Jansson, K., Al-Adili, A., Andersson Sundén, E. et al. The impact of neutron emission on correlated fission data from the 2E-2v method. Eur. Phys. J. A 54, 114 (2018). https://doi.org/10.1140/epja/i2018-12544-0

[2] J. Velkovska and R.L. McGrath. Fission fragment mass reconstruction from Si surface barrier detector measurement. Nucl. Instrum. Methods Phys. Res. A. 430, 2-3 507-511 (1999). https://doi.org/10.1016/S0168-9002(99)00225-9

[3] W. Seibt, K.E. Sundström, P.A. Tove. Charge collection in silicon detectors for strongly ionizing particles Nucl. Instrum. Methods Phys. Res. A. 113, 3 317-324 (1973). https://doi.org/10.1016/0029-554X(73)90496-5

[4] H-O. Neidel, H. Henschel, H. Geissel, Y. Laichter. Plasma delay of $^{238}$U ions in surface barrier detectors Nucl. Instrum. Methods Phys. 212, 1-3 299-300 (1983). https://doi.org/10.1016/0167-5087(83)90705-6

[5] I. Tsekhanovich, J.A. Dare, A. G. Smith, et. al. A novel 2v2e spectrometer in Manchester: new development in identification of Fission Fragments Seminar on Fission pp 189-196 (2008). https://doi.org/10.1142/9789812791061_0018

Speaker: Ana Maria Gomez
• 06:36
Simulation and design of a new ion guide for neutron-induced fission at the IGISOL facility 12m

Independent fission yields in neutron-induced fission at high neutron energy is important for the understanding of the fission process, and it is also in relevant for reactor physics applications. So far, measurements of independent fission yields in proton-induced fission has been performed at the IGISOL facility at the University of Jyväskylä, using the Penning trap as a high resolving-powermass-filter. In order to also facilitate neutron-induced measurements, a dedicated ion guide and a proton-to-neutron converter was developed. However, the first measurements [1] indicate that fewer fission products than expected reach the Penning trap. To explore potential reasons and possible improvements, a simulation model was developed [2, 3] and benchmarked [4]. The benchmark showed that the model is able to reproduce the performance of the ion guide remarkably well, but that the neutron flux from the converter has been over-estimated.

Based on the benchmark, the parameters of the setup, including the geometry of neutron converter, the distance from the converter to the uranium targets, and the size of the uranium targets and volume of the ion guide, have been optimized with regards to the production of fission products. However, the collection efficiency of the ion guide still limits the intensity of extracted fission products.

To significantly improve the collecting efficiency, as well as the extraction time, an electric field guidance system, inspired by the RF-system used in the CARIBU gas catcher [5], is considered to be deployed. In the gas catcher, fission products are prevented from hitting the walls of the ion guide by the RF-structure, and are guided by a DC electric field toward an extraction nozzle. With this poster we present the design of the guiding system for the IGISOL ion guide, as well as a simulation model for estimates of the extraction yields of fission products from the ion guide.

Rferences:
[1] A. Mattera et al. Production of sn and sb isotopes in high-energy neutron-induced fission of $^{nat}$U. Eur. Phys. J. A,54:33, 2018.
[2] A. Al-Adili et al. Simulations of the fission-product stopping efficiency in IGISOL.Eur. Phys. J. A,51:59, 2015.
[3] K. Jansson et al. Simulated production rates of exotic nuclei from the ion guide for neutron-induced fission at IGISOL. Eur. Phys. J. A,53:243, 2017.
[4] Zhihao et al.Benchmark of simulation of the ion guide for neutron-induced fission products. Submitted to Eur. Phys. J. A.
[5] G. Savard, A.F. Levand, and B.J. Zabransky. The caribu gas atcher. Nucl.Inst. Meth. B,376:246–250, 2016.

Speaker: Zhihao Gao
• 06:48
(WITHDRAWN) Simultaneous measurements of prompt fission γ-rays and neutrons with CLLBC detectors 24m

The knowledge about prompt neutrons and $\gamma$-rays spectral characteristics in nuclear fission is essential to understand the nuclear de-excitation process of fission fragments [1,2]. Preferably, these observables should be known in correlation with fragment properties like mass and kinetic energy.

Typical multi-parameter experiments require long measurement times and efficient detector systems. A particular problem for prompt-neutron spectral measurements is that usual detectors are not sensible to low neutron energies. Liquid-scintillator (LS) detectors, which allow neutron-$\gamma$ separation, exhibit low-energy thresholds for neutrons from $500~\text{keV}$ to $700~\text{keV}$. For stilbene-based detectors this threshold is about $100~\text{keV}$ lower. In order to overcome this low-energy limitation, usually combined measurements are performed. Lithium-glass ($^6\text{Li}$) detectors are employed for lower-energy neutrons. Afterwards spectra coming from LS and $\text{Li}$-glass need to be matched.

Recently, the so-called Caesium Lanthanum Lithium BromoChloride (CLLBC) detectors became available on the market. The detector material consists of a lanthanide-halide compound. Due to the presence of $^6\text{Li}$ and $\text{Cl}$, it is sensitive to $\gamma$ rays and neutrons. The pulse shape discrimination capabilities allow to distinguish between radiation types. We performed a full characterization of this type of detectors in terms of energy resolution, detection efficiency and coincidence-timing resolution using $\gamma$-rays between $80\text{keV}$ and $9\text{MeV}$. We determined the neutron response relative to the prompt-neutron spectrum from the spontaneous fission of $^{252}\text{Cf}$.

Applicability of CLLBC detectors for simultaneous measurements of prompt $\gamma$ rays and neutrons in fission will be discussed.

[1] O. Litaize et al., Eur. Phys. J. A (2015) 51:177.
[2] M. Travar, V. Piau et al., Physics Letters B 817, 136293 (2021)

Speaker: Cristiano Lino Fontana
• 06:00 07:30
Formats: I Delta King ()

### Delta King

Convener: Mike Herman (LANL)
• 06:00
GNDS-2.0 24m

The Generalised Nuclear Database Structure is a new standard for representing reaction and decay data in nuclear data libraries and is slated to replace the legacy ENDF-6 format. GNDS is designed to be both human readable and easy to work with so that downstream code developers can create software that supports GNDS. The specifications for GNDS-1.9 were published in 2020; this version describes the GNDS format used in the ENDF/B-VIII.0 library release. Work on GNDS-2.0 began soon after GNDS-1.9 was published with the focus on satisfying all of the requirements defined by WPEC Subgroup 38. In this contribution, we will detail the major changes in GNDS-2.0 including a new thermal neutron scattering markup that allows for mixed mode moderators, an expressive documentation format supporting rich meta data and improved support for covariance data, including thermal neutron scattering covariances. In addition, GNDS-2.0 improves and simplifies the resonance parameter descriptions, supports map-files (the GNDS equivalent of an ACE xsdir file), and provides a host of other smaller improvements. We will outline the timeline for finalizing GNDS-2.0 and the implementation status in processing codes.

This work was supported by the Nuclear Criticality Safety Program, funded and managed by the National Nuclear Security Administration for the Department of Energy. The work at Brookhaven National Laboratory was sponsored by the Office of Nuclear Physics, Office of Science of the U.S. Department of Energy under Contract No. DE-AC02-98CH10886 with Brookhaven Science Associates, LLC.

Speaker: David Brown (NNDC, Brookhaven National Laboratory)
• 06:24
WPEC SG50: Developing an Automatically Readable, Comprehensive and Curated Experimental Nuclear Reaction Database 12m

The EXFOR database is the major source to retrieve experimental data as input for nuclear data evaluations. It aggregates decades worth of international nuclear data measurements. EXFOR is specifically designed to store experimental data as it was reported by the experimentalists, with limited corrections made as needed. For this reason, the database does not store the subjective alterations to the data sets by users, such as evaluators. For example, evaluators may re-normalize data sets or augment uncertainties to increase consistency between data sets in their evaluation. These altered data sets, which are needed to reproduce the evaluations, are not readily available to others in the field. To improve the reproducibility of evaluations, the choices made by the evaluators need to be stored in a database that builds on the EXFOR database. In addition, some data in EXFOR are not easy to interpret automatically, including meta-data about experiments. If the new database improves the interpretability of this data with a different format, applying automatic processing and using machine learning techniques will be more tenable. To design this new database, an international collaboration has been established through the Organisation for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) Working Party on International Nuclear Data Evaluation Co-operation (WPEC) Subgroup 50 (SG50): Developing an Automatically Readable, Comprehensive and Curated Experimental Reaction Database. The goals of SG50 are to create a requirements document, a specifications document, and example files. The requirements document will contain the metadata to be stored in the new database, while the specifications document will outline the data structures used to store the information. This presentation will outline the current progress of SG50.

LA-UR-21-28324

Speaker: Dr Amanda Lewis (Naval Nuclear Laboratory)
• 06:36
Overview of experimental nuclear data from the n_TOF Collaboration 12m

The n_TOF neutron spallation source at CERN is used since 2001 for high quality nuclear data measurements from sub-thermal energy up to hundreds of MeV for the benefit of various communities in the fields of nuclear physics, nuclear astrophysics and nuclear technology. In the past twenty years, a considerable amount of valuable experimental results has been obtained and published, and measurements are still ongoing. In line with the CERN open data policy, the n_TOF Collaboration has taken actions [1] to preserve its unique data, to facilitate access to them, and to allow their re-use by expert users. For the vast majority of published results, reaction yields, cross sections and resonance parameters are now available in the international EXFOR database. However, these results have not been fully exploited yet for the benefit of the end-users, in particular for the improvement of evaluated nuclear data libraries. This contribution aims at updating the status and availability of n_TOF data, and to discuss ongoing efforts for better integration of the results in the evaluated library projects.

[1] E. Dupont et al. (The n_TOF Collaboration), Dissemination of data measured at the CERN n_TOF facility, EPJ Web Conf. 146 (2017) 07002

Speaker: Emmeric Dupont (CEA)
• 06:48
Dissemination of nuclear structure and decay data 12m

The availability of reliable, up-to-date and well-structured nuclear data libraries, with user-friendly visualization and retrieval interfaces, is a valuable tool for both nuclear specialists in the applications fields and nuclear physics researchers.

The International Atomic Energy Agency (IAEA) has been coordinating the international network of Nuclear Structure and Decay Data evaluators (NSDD) which maintains the Evaluated Nuclear Structure Data File (ENSDF) since 1974. In addition, it has been organising and leading a series of international projects with the purpose of collecting, evaluating and disseminating nuclear structure and decay data to address specific user needs. The output of the network activity and projects are disseminated via publications and online databases with user-friendly visualization and retrieval interfaces.

In this paper we present the status of the following three data development and dissemination efforts carried out at the IAEA:

a) Live Chart: interactive interface to the Evaluated Nuclear Structure Data File (ENSDF) produced by the NSDD network (https://www-nds.iaea.org/livechart/)
b) Reference database for beta-delayed neutrons: online database and interactive interface to the compiled and evaluated beta-delayed neutron data produced by an IAEA Coordinated Research Project (https://www-nds.iaea.org/beta-delayed-neutron/database.html)
c) Nuclear moments database: online database and interactive interface to the compiled and recommended magnetic dipole and electric quadrupole moments produced by N.J. Stone with contributions from experts within an IAEA project (https://www-nds.iaea.org/nuclearmoments/)

Speaker: Paraskevi Dimitriou (International Atomic Energy Agency)
• 07:00
A decay database of coincident $\gamma/\gamma$ and $\gamma/X$-ray branching ratios for in-field applications 12m

Current fieldable spectroscopy techniques often use single detector systems heavily impacted by interferences from intense background radiation fields. These effects result in low-confidence measurements that can lead to misinterpretation of the collected spectrum. To help improve interpretation of the fission products and short-lived radionuclides produced in a composite sample, a coincidence-$\gamma$ database is being developed in support of a robust portable $\gamma$/$X$-ray coincidence detector system concurrently under development at the Pacific Northwest National Laboratory for in-field deployment. Hitherto, no database exists containing coincident $\gamma/\gamma$ and $\gamma/X$-ray branching ratios on an absolute scale that will greatly enhance isotopic identification for in-field applications.

As part of this project, software has been developed to parse all radioactive-decay data sets from the Evaluated Nuclear Structure Data File (ENSDF) archive to enable translation into more useful eXtensible Markup Language (XML) and JavaScript Object Notation (JSON) formats that more readily support query-based data manipulation. The coincident database described in this work is the first of its kind and contains coincidence $\gamma/\gamma$, $\gamma/X$-ray, and $\gamma/< particle >$ branching ratios (where $< particle >$ = $\alpha$, $\beta^{-}$, $\beta^{+}$, $\epsilon$) and their corresponding uncertainties, together with auxiliary metadata associated with each decay data set. Both XML and JSON formats provide a convenient and portable means of data storage that can be imported into analysis frameworks with relatively low overhead allowing for meaningful comparison with measured data.

Speaker: Dr Aaron Hurst (University of California, Berkeley)
• 07:12
Overview of NNDC Web Services 2020-21 12m

The National Nuclear Data Center (NNDC) provides access to physics research data through a collection of 40 websites. Each is dedicated to a topic, a database, or to a previous conference. Many also have long histories, with the oldest websites having been deployed in 1998. Collectively, these websites receive millions of requests for data, with an estimated 5.2 million retrievals in 2020 alone. The popularity of the NNDC’s most prominent websites can attest to both its reach and its impact.

The Evaluated Nuclear Structure Data File (ENSDF) format has its own dedicated database and a website of the same name. The ENSDF website provides evaluated research data on excitation states for all known nuclides, in addition to data on nuclear reactions and decay. In 2020, the ENSDF website received its own chart interface to enable visual exploration of the ENSDF database.

ENSDF is the second most-visited website after NuDat, which maintains the original Chart of Nuclides. This interface visualizes patterns in half-lives, decay energies, fission yields, and more by mapping each nuclide onto a coordinate grid. In addition to the Chart of Nuclides, NuDat provides several plotting tools, all of which received updates in 2019. Of particular note is the Advanced Cross-Variable Plot, which graphs relationships between observable nuclear properties. In 2021, the NuDat 3.0 beta was released as a preview of a faster, smoother Chart of Nuclides.

To address the needs of specific users, the NNDC website also provides tools which query and visualize subsets of the main libraries for specific applications. The Medical Isotope Radiation Dose (MIRD) format is designed for isotopes used in nuclear medicine, and has its own dedicated website. The process of thermal neutron capture decay is also recorded in the CapGam website. In 2020, the MIRD website was updated to improve response times and add nuclear decay diagrams. Shortly afterwards, CapGam was redesigned with responsive user interfaces and a streamlined update process to keep up with new evaluations.

More recently, the NNDC has turned its attention towards modernizing its web services. Over time, the differences in each website’s creation and purpose have complicated updating code and correcting programming errors. The NNDC’s most recent web development project was started to create a standardized development process for present and future NNDC websites. It started with using Git version control for organizing and managing changes to code. It then progressed to using the Gradle build tool to make updates simpler and faster. The final (and still ongoing) step in this project is to re-deploy each website as a portable Web Archive (WAR) file. Each of these steps was taken with the intent of streamlining NNDC web development, which in turn will hopefully improve existing websites and guide the creation of new ones.

Work sponsored by the Office of Nuclear Physics, Office of Science of the U.S. Department of Energy, under Contract No. DE-AC02-98CH10886.

Speaker: Benjamin Shu
• 06:00 07:30
Measurements: III Folsom ()

### Folsom

Convener: Gilles Noguere (CEA)
• 06:00
First high-resolution $^{80}$Se(n,$\gamma$) cross section measurement between 1 eV and 100 keV and its astrophysical implications for the $s$-process 24m

The slow neutron capture ($s$-) process is responsible for the formation of half of the elements heavier than iron in the Universe. Despite the long time scale of this process, the long half-life of some isotopes throughout the $s$-process reaction flow creates branching points that lead to the division of the nucleosynthesis path. ${79}$Se ($t_{1/2} = 3.27 \times 10^5$ y [1]) represents one of the most relevant and debated s-branching nuclei [2] for two main reasons. On the one hand, the existence of quantum states in ${79}$Se, whose population varies with temperature, makes the s-process path sensitive to temperature. On the other hand, the observed abundances of the s-only isotopes of krypton (${80,82}$Kr) are very well-known from meteoric data. Thus, by comparing these abundances with those predicted by stellar models, information about the thermal conditions of the stellar media in which the $s$-process occurs can be obtained. To this aim hydrodynamic stellar models need experimental data on the neutron capture cross section of all isotopes involved in the branching.
In this context, we have measured the neutron capture cross section of ${80}$Se at CERN n_TOF, with very high energy resolution for the first time [3]. Although there is a previous measurement on of ${80}$Se(n,$\gamma$) [4], it suffers from a very limited energy resolution and a short neutron-energy range, as it can be appreciated in Fig. 1. These drawbacks have been remarkably improved in this time-of-flight measurement that covers the entire energy range of astrophysical interest between 1 eV and 100 keV. One hundred and thirteen resonances have been characterized, ninety-eight of them for the first time. The impact is sizable, being the MACS at kT = 8 keV 36\% smaller than the recommended value in KADoNiS [5]. In this work we present final results together with a discussion of their astrophysical implications.

[1] G. Jörg et al., Applied Radiation and Isotopes, 68(12):2339–2351, 2010.
[2] F. Kappeler et al., Reports on Progress in Physics, 52(8):945–1013, August 1989.
[3] V. Babiano-Suárez et al., CERN-INTC-2018-005, INTC-P-536, 2018.
[4] G. Walter et al., Astron. Astrophysic, 167:186-199, 1986.
[5] I. Dillmann et al., Nuclear Data Sheets, 120:171–174, 2014.

Speakers: Victor Babiano-Suarez, n_TOF collaboration
• 06:24
The Stellar $^{72}\mathrm{Ge}(n,\gamma)$ Cross Section for weak s-process: A First Measurement at n_TOF 12m

The slow neutron capture process (s-process) is responsible for producing about half of the elemental abundances heavier than iron in the universe. Neutron capture cross sections on stable isotopes are a key nuclear physics input for s-process studies. The $^{72}\mathrm{Ge}(n,\gamma)$ Maxwellian-Averaged Cross Section (MACS) has an important influence on production of isotopes between Ge and Zr in the weak s-process in massive stars [1] and so far only theoretical estimations are available [2].

An experiment was carried out at the neutron time-of-flight facility n$\_$TOF [3] at CERN to measure the $^{72}\mathrm{Ge}(n,\gamma)$ reaction for the first time at stellar neutron energies. At n$\_$TOF, the neutron beam covers a large energy range (few meV to several GeV). The capture measurement was performed using an enriched $^{72}\mathrm{GeO}_2$ sample at a flight path length of $184\,$m, which provided high neutron energy resolution. The prompt gamma rays produced after neutron capture were detected with a set of liquid scintillation detectors (C$_6$D$_6$). The neutron capture yield is derived from the counting spectra taking into account the neutron flux and the gamma-ray detection efficiency using the Pulse Height Weighting Technique [4].

Over $70$ new neutron resonances were identified, providing an improved resolved reaction cross section to calculate experimental MACS values for the first time. Furthermore, averaged resonance parameters such as $\langle \Gamma_\gamma \rangle$ and $D_0$ were derived from the resonance data. In summary, the experiment, data analysis and the new MACS results will be presented including their impact on stellar nucleosynthesis, which was investigated with $\mathsf{mppnp}$ [5] using a $25$ solar mass model.

References:
[1] M. Pignatari et al., The Astroph. J. 33, 1557-1577 (2010).; [2] I. Dillmann et al., Nuclear Data Sheets 120, 171-174 (2014); (http://www.kadonis.org).; [3] C. Guerrero et al., Eur. Phys. J. A 49, 27 (2013).; [4] U. Abbondanno et al., Nucl. Instr. Meth. Phys. Res. A 521, 454-467 (2004).; [5] M. Pignatari et al., The Astroph. J.,Suppl. Ser. 225, 24 (2016).

Speakers: Mirco Dietz, the n_TOF Collaboation
• 06:36
Direct measurements of $^2$H(α, γ)$^6$Li cross section at Big Bang energies and the primordial lithium problem. 12m

The correct prediction of the abundances of the light nuclides produced during of Big Bang Nucleosynthesis (BBN) is one of the main topics of modern cosmology. In order to precisely determine BBN 6 Li production the cross-section of the nuclear reaction 2H(α, γ)6Li must be directly measured within the astrophysical energy range of 30-400 keV. This measurement requires ultra low gamma-ray background, as obtained at LUNA, the deep underground accelerator laboratory installed in the INFN Gran Sasso National Laboratory (LNGS), Italy. On the basis of the new experimental data, the 2H(α, γ)6Li thermonuclear reaction rate has been derived. Our rate is even lower than previously reported, thus increasing the discrepancybetween predicted Big Bang 6Li abundance and the amount of primordial 6Li inferred from observations. The primordial 6Li/7Li isotopic abundance ratio has been consequently determined to be (1.5 ± 0.3) × 10^(−5) within standard BBN theory. The much higher 6Li/7Li values reported for halo stars will likely require a non-standard physics explanation, as discussed in the literature.

Speaker: Gianpiero Gervino
• 06:48
Stopping power of ions in matter: current status of experimental data and theoretical description 12m

Stopping powers are relevant to a wide range of applications, such as ion beam analysis, deposition ranges, ion implantation, and radiation damage, to name a few. Reliable values of stopping powers are also needed in isotope production for medical applications, in fusion technologies and detector development.

In this work we present the state of the art of the stopping power of ions in matter. We give an overview of our present knowledge, and discuss the areas of strength and weaknesses and where data is lacking, on the basis of the comprehensive experimental stopping power database of the International Atomic Energy Agency (IAEA) \cite{iaea,Montanari_SDB_2017}.

The field of stopping powers in matter is evolving with new trends in materials of interest, including oxides, polymers, and biological targets \cite{MATERNA20211,Provenzano_2015}. An example for different ions in Mylar is displayed in Figure \ref{fig:mylar}. Our goal is to identify areas of interest and emerging data needs to meet the requirements of a continuously developing user community.

\begin{figure}[h]
\centering
\includegraphics[width=0.75\textwidth]{Mylar_new.eps}
\caption{Stopping cross section of ions in Mylar compared with SRIM results \cite{srim} (lines). Experimental data (symbols) are from
\cite{iaea}; black triangles are recent data for Cs fragments reported in \cite{MATERNA20211} . }
\label{fig:mylar}
\end{figure}

\begin{thebibliography}{00}

\bibitem{iaea} IAEA Stopping Database. Electronic Stopping Power of Matter for Ions: Graphs, Data, Comments and Programs, Available at https://wwwnds.iaea.org/stopping/ (last update: June 2021).
\bibitem{Montanari_SDB_2017} Montanari, C. C.; Dimitriou, P. Nuclear Instruments and Methods in Physics Research B: 408,50–55 (2017).
\bibitem{MATERNA20211} Materna T. et al, Nuclear Instruments and Methods in Physics Research B: 505, 1–16 (2021).
\bibitem{Provenzano_2015} Provenzano L. et al, Journal of Physics: Conference Series 583, 012047 (2015).
\bibitem{srim} Ziegler, J. SRIM. The Stopping and Range of Ions in Matter, Available at http://www.srim.org/ (last update: 2013).

Speaker: Claudia Montanari
• 07:00
Hydrogen isotopes emissions from the nuclear muon capture reaction in silicon 12m

The main threat to electronics at the ground level is well-known as the secondary-cosmic-ray-induced soft error. The soft error is caused by an upset of the memory information due to an energy deposition by an energetic ionizing radiation. Among the cosmic-ray species, the muon has recently drawn attention as a new cause of the soft error due to the reduction of the critical charge of the static random-access memory. Our previous works [1] reported that the negative muon has much more serious effect on the occurrence of soft errors compared to the positive one because of the emission of light ions (hydrogen and helium) from the nuclear muon capture reaction in the silicon nuclei. In this work, we performed the experiment to accumulate the basic data of the fundamental physical process of muon-induced soft error, i.e., the light ion from the nuclear muon capture reaction, to improve the accuracy of the simulation of the muon-induced soft error. In addition to the experiment, a validation of the model calculation implemented in Particle and Heavy Ions Transport code System (PHITS) [2] was made by comparing the experimental data and the simulation.

The experiment was performed at the M1 muon beam line of Muon Science Innovative muon beam Channel [3] in Research Center for Nuclear Physics, Osaka University, Japan. The pions, which is a source of muons, were generated through the nuclear reaction between a 392-MeV proton beam and a graphite target. The produced pions almost decayed into muons in a superconducting solenoid magnet. The polarity and the momentum of the muon were selected by the solenoid magnet and the selected muons were transported to a vacuum chamber which was connected to the beam exit of the M1 beam line. The momentum of 36 MeV/c was chosen to maximize the number of stopping muons. A 100 µm thick silicon target was mounted at the center of the chamber at a tilt angle of 45° to the beam direction. The size of the target is 10 cm × 5 cm. Two forward plastic scintillators with the size of 5 cm × 7 cm × 0.5 cmt were set to count the number of incident muons and make the triggers for the data acquisition. Two telescopes were mounted parallelly to the target at both the upstream and downstream of the target to detect the secondary ions and measure their kinetic energies. The energies of the ions were determined by a ΔE-E method by using a 325 µm silicon detector (ΔE) and 25 mm thick CsI detector (E).

The energy spectra of proton, deuteron and triton were successfully measured in the energy range from 8 MeV to 35 MeV. The proton and deuteron spectra in the energy range of 8-15 MeV and the triton spectrum were first measured in this work. The measured proton and deuteron spectra were compared to those of the past work [4]. The comparison demonstrated the consistency between the present and past data. Next, a benchmark test of the PHITS calculation was performed. The nuclear muon capture reaction is described with the quantum molecular dynamics (QMD) [5] or modified QMD (MQMD) [6] for the dynamic process plus the generalized evaporation model (GEM) for statistical decay process [7]. The test indicated that the MQMD plus GEM has larger yields of deuteron and triton than the QMD plus GEM. However, both models still significantly underestimated the measured spectra in the high emission energy region.

[1] S. Manabe et al., IEEE Trans. Nucl. Sci., 65:8, 1742-1749 (2018).
[2] T. Sato et al., J. Nucl. Sci. and Technol., 55:5-6, 684-690 (2018).
[3] Y. Matsumoto et al., Proc. of 6th International Particle Accelerator Conference (IPAC), 2537-2540 (2015).
[4] S. E. Sobottka and E. L. Wills, Phys. Rev. Lett., 20:12, 596-598 (1968).
[5] K. Niita et al., Phys. Rev. C, 52:26, 2620-2635 (1995)
[6] Y. Watanabe and D. Kadrev, in Proc. of International Conference on Nuclear Data for Science and Technology 2007 (2008)
[7] S. Furihata et al., Japan Atomic Energy Research Institute; 2001. (JAERI-Data/Code 2001-015).

Speaker: Seiya Manabe
• 06:00 07:30
Reactor Data: III Placerville ()

### Placerville

Convener: Luiz Leal (IRSN)
• 06:00
Supplying Accurate Nuclear Data for energy and non-energy Applications – The SANDA Euratom project 24m

The SANDA H2020-EURATOM project, Supplying Accurate Nuclear Data for energy and non-energy Applications, brings together the majority of the European nuclear data community (35 partners from 19 countries), infrastructures and resources to produce accurate and reliable nuclear data tools including data, codes and methodologies that can be used to simulate, analyse, optimize, exploit and evaluate the safety of nuclear energy and non-energy applications.

The project covers all the aspects of the nuclear data cycle relevant to the needs for energy applications, including thermal and fast reactors and Accelerator Driven Systems (ADS), waste management, storage, reprocessing or disposal, decommissioning of nuclear facilities and other non-energy applications such as metrology, standards, medical imaging techniques, electron and proton cancer therapy and the production of isotopes for medical and industrial applications.

SANDA has been built taking into account the High Priority Nuclear Data needs list from OECD/NEA and IAEA to provide the final users with immediately usable data and tools for the cases where this is feasible during the project duration. In particular, it will bring at reach the target precision for some important isotopes/reactions. In addition, the R&D of SANDA will contribute to prepare the path (new detectors, facilities and methods) for future experiments addressing the remaining nuclear data needs in the years to come.

The structure of the project, outlook of the expected results and a selection of results achieved will be presented during the conference, including:

• The progress on new techniques, detectors and evaluation methods addressing high priority data needs: (n,f) and (n, $\gamma$) cross sections of highly radioactive isotopes, advanced detection setups (n,n), (n,n’$\gamma$) and (n,xn) reactions, $\beta$-decay experiments and fission yields.

• Advances in data evaluation and uncertainty assessment, verification and validation of libraries and codes with selected integral experiments.

• Nuclear data needs and applicability for nuclear reactor safety and fuel cycle analysis.

More information on SANDA can be found at http://www.sanda-nd.eu/.

Speaker: Dr Enrique M. González Romero (CIEMAT)
• 06:24
Testing the hypothesis of a universal behavior of the cross-section probability tables in the unresolved resonance region 12m

In the unresolved resonance region (URR), cross sections for neutrons interacting with nuclei have resonances that cannot be measured nor predicted, hence only statistical values are provided. Often, proper simulations by neutronic transport codes require special data and handling in the URR to account for the variation in the cross sections in this region. The current methodology used to describe this behavior is to construct the probability table of the cross section. This table is generated by extrapolating the average resonance widths and average resonance spacings from the resonance region, generating Monte Carlo realizations of resonance ladders and using these realizations to construct the probability tables. Although this is a standard and widely used technique, it is computationally very expensive. Therefore, an analytical fit would be preferable due to the considerable speed up of the computational time in real life applications. Since the only input for these tables is the average resonance parameters, which are functions of energy, orbital angular momentum, and total angular momentum, our hypothesis is that the probability tables might display a “universal” form for each reaction channel regardless of the nuclear species. If this is true, a single functional form per reaction channel could be used to parametrize the probability table for every nucleus. To test this hypothesis, our goal is to generate probability tables for different spin groups, different reaction channels, and different nuclei. As a first step, these tables will be analyzed searching for regularities and, if our hypothesis proves correct, as a second step we will use modern computational techniques to produce accurate analytical fits for each spin group.

Speaker: Matteo Vorabbi (Brookhaven National Laboratory)
• 06:36
The effect of nuclear data uncertainties and sensitivities on the potential use of uranium nitride fuels in current and future light water reactors. 12m

The Advanced Fuel Cycle Programme funded by the UK Government through the Department for Business, Energy & Industrial Strategy revolves around developing capability and capacity for nuclear while seeking to reduce costs across the nuclear lifecycle, to ensure that nuclear can play a part in delivering secure, low-carbon energy in the global market. This includes the development of Advanced Technology Fuels which are targeted at providing increased efficiency within current generation reactors. One novel fuel type that is being considered is uranium nitride to replace existing uranium dioxide fuels in current and future reactors. This fuel having improved thermal conductivity and higher actinide density. The use of such fuel will require the ability to accurately simulate its performance in a range of operational and safety scenarios across the fuel cycle.

This work considers the uncertainties introduced from nuclear data during calculations supporting manufacturing, irradiation, and waste management. The nuclear data sensitivities are studied from all sources but are principally concerned with the capture and scattering cross-sections of the nitrogen isotopes. As no experimental benchmarks exist to validate simulations of uranium nitride fuel, the original measurements and evaluations are studied in terms of their uncertainties. Additionally, the sensitivity of important parameters such as criticality, fuel life and increased ${^{14}}$C production are considered over a range of possible uranium enrichments in natural nitrogen and nitrogen reduced in ${^{14}}$N

Speaker: Dr Robert William Mills (UK National Nuclear Laboratory)
• 06:48
Update of the CIELO $^{238}$U resonance evaluation to improve LWR performance with burnup 12m

New U-235 and U-238 evaluations [1-3] were undertaken within the OECD/NEA Data Bank CIELO Project [4] and were adopted for the ENDF/B-VIII.0 library [5], which was released in 2018. Since then, several reports and publications were released that showed serious discrepancies with the light water reactor (LWR) performance of the previous ENDF/B-VII.1 library [6] in criticality studies as the function of the burnup, e.g., see Ref. [7]. A slight increase of the LWR reactivity was observed at the Beginning of Cycle (BOC) with a severe loss of reactivity at large burnup observed for the ENDF/B-VIII.0 library. Sensitivity studies showed some compensation effects at the BOC, but uniquely identified the U-238 evaluation as the responsible for the reactivity loss [7].

Recent changes in the U-235 resonance evaluation [8] slightly reduced the U-235 capture cross section below 20eV, but these changes are not expected to have an impact on reactivity trends with burnup. However, slight positive criticality bias may be expected due to U-235 changes.

In this work we focused on studying changes in resonance cross sections of U-238 that may explain the observed trend as a function of burnup. It was found that capture cross section from 0.1eV up to 10eV was reduced in ENDF/B-VIII.0 evaluation by about 2% [,4,5,9] compared to the ENDF/B-VII.1 evaluation [6,10] as shown in Figure 1.

(see attached PDF)
Figure 1. ENDF/B-VII.1 to ENDF/B-VIII.0 capture cross-section ratio in the resonance region.

The observed changes may explain the observed burnup trend as lower capture cross sections in U-238 around 1eV lead to increased criticality at the BOC, but results in lower criticality at higher burnup due to the reduced production of Pu-239 due to the lower capture cross section. This hypothesis was tested in deterministic calculations and confirmed in our preliminary studies.
.
Additional changes in ENDF/B-VIII.0 evaluation in the resonance region [9] are on-going to try to preserve the capture cross section below 10eV reproducing at the same time the existing transmission experiments.

[1] R. Capote, A. Trkov, M. Sin, M. Herman, A. Daskalakis, Y. Danon, Physics of neutron interactions with 238U: new developments and challenges, Nucl. Data Sheets 118, 26(2014)
[2] R. Capote, A. Trkov, M. Sin et al., IAEA CIELO evaluation of neutron-induced reactions on 235U and 238U targets, Nucl. Data Sheets 148, 254 (2018)
[3] R. Capote, A. Trkov (coordinators), IAEA CIELO Data Development Project within the International Pilot Project of the OECD/NEA [1], 235U and 238U files released December 1st, 2017, https://www-nds.iaea.org/CIELO/
[4] M.B. Chadwick et al.: “CIELO Collaboration Summary Results: International Evaluations of Neutron Reactions on Uranium, Plutonium, Iron, Oxygen and Hydrogen”, Nuclear Data Sheets 148 (2018) 189–213.
[5] D.A. Brown et al.: “ENDF/B-VIII.0: The 8th Major Release of the Nuclear Reaction Data Library with CIELO-project Cross Sections, New Standards and Thermal Scattering Data”, Nuclear Data Sheets 148 (2018) 1–142.
[6] M.B. Chadwick, M.W. Herman, P. Oblozinsky et al., ENDF/B-VII.1 nuclear data for science and technology: cross sections, covariances, fission product yields and decay data, Nucl. Data Sheets 112, 2887 (2012)
[7] Kang-Seog Kim and William A. Wieselquist, “Neutronic Characteristics of ENDF/B-VIII.0 Compared to ENDF/B-VII.1 for Light-Water Reactor Analysis”, J. Nucl. Eng. 2 (2021) 318–335. https://doi.org/10.3390/jne2040026
[8] M.T. Pigni, R. Capote, A. Trkov, Y. Danon. “Updates of the U-235 resonance parameters below 20 eV”, INDEN collaboration, IAEA 2020. See U-235 tab at https://www-nds.iaea.org/INDEN/.
[9] H.I. Kim, C. Paradela, I. Sirakov et al., Neutron capture cross section measurements for 238U in the resonance region at GELINA, Eur. Phys. J. A 52, 170 (2016)
[10] H. Derrien, L.C. Leal, N.M. Larson, and A. Courcelle, Report ORNL/TM-2005/241 (2005). “Neutron Resonance Parameters of 238U and the Calculated Cross Sections from the Reich-Moore Analysis of Experimental Data in the Neutron Energy Range from 0 keV to 20 keV”.

Speaker: Roberto Capote (IAEA NDS)
• 07:00
Target Accuracy Requirements for MYRRHA 12m

The Multi-purpose hYbrid Research Reactor for High-tech Applications (MYRRHA) is a flexible experimental facility being designed at SCK CEN, Mol, Belgium. It is conceived to operate both in sub-critical mode, as an Accelerator Driven System, and in critical mode, as a lead-bismuth cooled fast reactor.

In order to comply with MYRRHA reactor design requirements, uncertainties must be quantified. In nuclear reactor design the uncertainties mainly come from material properties, fabrication tolerances, operating conditions, simulation tools and nuclear data. Indeed, the uncertainty in nuclear data is one of the most important sources of uncertainty in reactor physics simulations, and significant gaps between the uncertainties and the target accuracies (i.e., maximum acceptable uncertainties for a certain parameter) have been systematically shown in the past. Meeting the target accuracy is required not only to achieve the requested level of safety for MYRRHA, but also to minimize the increase in the costs due to additional safety measures.

Target accuracy assessments allow identifying nuclear data needs and requirements (by nuclide, nuclear reaction and energy channel). Since the accuracy and priorities strongly depend on the assumed initial uncertainty data and since numerous nuclear data libraries with updated uncertainty evaluations have been released in the recent years (i.e., ENDF/B-VIII.0, JEFF-3.3, TENDL-2019) and new evaluations are currently being produced, such as JEFF-4 and JENDL-5, it is necessary to provide updated target accuracies for advanced reactor design parameters and nuclear data.

Therefore, in this work we present a Target Accuracy Requirement Assessments of the latest MYRRHA reactor design. The capability of calculating the uncertainty data requirements has been implemented in SANDY, a Python package that can read, write and perform a set of operations on nuclear data files written in the standard ENDF-6 format. Afterwards, SANDY has been used to perform Target Accuracy Requirement Assessments on a homogenised on fuel assembly level MYRRHA reactor model with JEFF-4T0 and other state-of-the-art nuclear data libraries. Nuclides and reactions in need of improvement have been identified to provide feedback to the NEA Nuclear Data High Priority Request List and the JEFF community.

Speaker: Pablo Romojaro (SCK CEN)
• 07:12
Methods for calculating self-shielded multigroup displacement damage cross section 12m

Neutron irradiation damage is a key factor that influences behaviors of nuclear materials. Folding the specific neutron flux to the pre-calculated displacement damage cross section is a widely used method for quantifying neutron irradiation damage in nuclear reactors. This method has been implemented in various neutron transport codes, such as ERANOS and OpenMC. However, because the displacement damage cross section is a generalization of nuclear cross section, current theories on the computation of self-shielded multigroup cross section are not directly applicable for the damage cross section.
Displacement damage cross section, briefly noted by Displacement per Atom (DPA) cross section hereinafter, is proportional to the product of nuclear reaction cross section and the average damage energy: σ_DPA (E)∝σ(E) E_a (E). There are three approaches to calculate the multigroup DPA cross section. Figure 1 shows the multigroup DPA cross sections computed with the three methods along with the quasi-point-wise (ECCO 1968 group) one for neutron elastic scattering on 56Fe. The first method is mathematically rigorous for computing multigroup cross section, but the self-shielding corrections on nuclear cross section cannot be directly applied. The latters can directly use the self-shielding corrections on nuclear cross section, whereas they are not mathematically well defined for multigroup reaction rate calculation.
The present work thoroughly studies the three approaches for computing multigroup DPA cross section and their influences on DPA rate based on Fe in the core and Mg in the reflector of a typical sodium-cooled fast breeder reactor. In addition, the uncertainty induced by the approximate treatments of self-shielded multigroup DPA cross sections is also quantified.

Speaker: Shengli Chen
• 06:00 07:30
Space Applications Capitol ()

### Capitol

Convener: Ramona Vogt (LLNL)
• 06:00
Nuclear Data Needs for Space Radiation Transport Calculations 24m
Speaker: Lawrence Heilbronn (University of Tennessee)
• 06:24
Theoretical Calculations and Evaluations for n+28,29,30Si Reactions 12m

Silicon is of great applied interest in the semiconductor industry and in detectors for physics experiments. An increased understanding of neutron induced charged particle producing reactions on silicon is particularly significant not only for basic physics but also for applications.
In the present work, an optimal set of optical model potential parameters is obtained for the n+28,29,30Si reactions up to 200 MeV, all the cross sections, the angle integrated spectra and the double differential cross sections of neutron, proton, deuteron, triton, helium and alpha-particle emission for n+28,29,30Si reactions are consistently calculated.
The optical model, the unifed Hauser-Feshbach and exciton model which includes the improved Iwamoto-Harada model, the distorted wave Born approximation, the intra-nuclear cascade model and recent experimental data are including in the theoretical calculations.
The results are analyzed and compared with existing experimental data and evaluations in ENDF/B-VIII.0 and JENDL-4.0.

Speakers: Mr Yinlu Han (中国原子能科学研究院), Yue Zhang (中国原子能科学研究院)
• 06:36
Total nuclear reaction cross-section database for radiation protection in space applications 12m

Monte Carlo simulation tools make use of models to describe the physics of everything that surrounds us. For such tools to be reliable, models as close to reality as possible need to be implemented in the simulation. Nuclear reaction cross-sections play an essential role in simulations for radiation protection in space. Therefore, a GSI-ESA-NASA collaboration project created a collection of total nucleus-nucleus reaction cross-sections, which have been experimentally measured over the last few decades*. The database has been made public and can be found at: https://gsi.de/fragmentation.
The data from the collection have been systematically compared to the cross-section models implemented in FLUKA, PHITS and Geant4, which are the most commonly used Monte Carlo codes for radiation protection in space. Such models are Tripathi, Kox, Shen, Kox–Shen, and Hybrid-Kurotama. An example of such a comparison for Fe-56 ions on C-12 target is shown in Figure 1.
Finally, literature gaps have been identified and considerations have been made about which models fit best the existing data for the most relevant systems to radiation protection in space and heavy-ion therapy.

*F Luoni et al 2021 New J. Phys. 23 101201, DOI: https://doi.org/10.1088/1367-2630/ac27e1

Speaker: Francesca Luoni (GSI, TUDa)
• 06:48
Prospects for Measurements of Production Cross Sections of Light Nuclei at RHIC 24m

The damage due to cosmic rays is a serious concern for astronauts, electronics, and spacecraft. Although the cosmic ray flux is comprised mostly of protons and helium nuclei, there is a non-negligible component comprised of heavy-ions up to and including iron nuclei. The importance of the heavy-ion component is enhanced because the damage due to ionization goes as $Z^2$. In addition to the damage due to primary ionization from a heavy-ion, the damage from secondary production of p, d, t, $^3$He, and $^4$He is also significant. Extensive double differential measurements for light fragment production have been made for projectile energies below 3 GeV/n. Many of these measurements have been made at the NASA Space Radiation Laboratory (NRSL), which uses beams extracted from the Booster synchrotron at Brookhaven National Laboratory (BNL). No light nucleus production data exist for heavy-ion projectile energies from 3-50 GeV/n. The Space Radiation Protection community has identified this high energy regime as an area of need. Both cross section data and models to describe those data are needed. Although several facilities that could produce heavy-ion beams in this high energy range are either under construction or planned, FAIR at GSI, NICA at Dubna, and an ion driver at JPARC, currently the only facilities that can address the needs in this high energy range are the Super Proton Synchrotron (SPS) at CERN (13-200 GeV/n) and the Relativistic Heavy Ion Collider (RHIC) at BNL (3-125 GeV/n). Although RHIC is a collider, the Solenoidal Tracker at RHIC (STAR) experiment has installed a fixed-target and the Collider Accelerator Division (CAD) has developed an efficient conduct of operations to deliver ion beams to the target. In the past few years, the RHIC/STAR fixed-target program has completed an energy scan with gold projectiles from 3 to 100 GeV/n on a gold foil target. Although gold ion beams are not relevant for space radiation concerns, these measurements demonstrate the capabilities of the STAR detector to make the light nucleus production cross section measurements using particle identification with both ionization density ($dE/dx$) and time-of-fight (TOF). RHIC is a flexible facility and can deliver the ion beam species (He, C, Si, Fe) and energies (3-125 GeV/n) of need to the Space Radiation Community. STAR can install the targets of interest (C, Al, Fe) and can make the necessary light nucleus production cross section measurements. This talk will discuss the prospects for making these measurements during the RHIC ion beam running periods from 2023-2025.

Speaker: Prof. Daniel Cebra (University of California, Davis)
• 07:12
Stopping Power and Neutron Scattering Measurements in Support of Space Exploration 12m

The galactic cosmic ray (GCR) background presents creates a hostile, high-dose environment for astronauts and the electronics they rely. While the GCR spectrum extends to 10’s of GeV/nucleon, much of the dose in electronics is attributable to the final slowing of heavy ion at the end of their path, e.g., the “Bragg peak”, and the location and magnitude of the Bragg peak is in turn highly sensitive to the stopping power. In contrast, long range dose attributable to the “neutron albedo” from backscattered neutrons arising from high-energy GCR cascades present a chronic health concern for long-term human habitation the surface of the moon and Mars. Researchers working to address these concerns rely on robust nuclear data to guide the modeling of both electronic stopping powers and neutron scattering. However, recent work by Sigmund and Baldez highlight the need for improved electronics stopping power data. In this talk I will review the charged-particle and neutron data needs relevant to space exploration and present a framework for how to address them through a program heavy-ion and neutron beams using the LBNL 88-Inch cyclotron. This work is supported .by the Lawrence Berkeley National Laboratory, USA under Contract No. DE-AC02-05CH11231 for the US Nuclear Data Program.

Speaker: Prof. Lee Bernstein (LBNL/UCB)
• 07:30 07:45
Break 15m
• 07:45 09:00
Poster session: I Gold Rush ()

### Gold Rush

• 07:45
Comparison of double-differential cross section between nuclear data library and experimental data for photoneutron production 3m

Radiation shielding and calculation are essential for the construction and operation of accelerator facilities. High energy gamma-rays in accelerators can produce secondary neutrons, which can penetrate through shielding materials and produce additional dose rate by activating accelerator components; thus, the double differential cross sections (DDXs) of photoneutron are very essential. Calculation of radiation and shielding in accelerators can be carried out by Monte Carlo simulation tools using nuclear data library (NDL). The DDXs of ($\gamma$,xn) reactions of a polarized photon beam with the mono energy of 16.6 MeV and different materials (including Pb, Au, Sn, Fe, Cu, and Ti) are reported in Reference [1]. In this study, we will show the comparisons in the DDXs between the experimental data measured in Reference [1] and the DDXs from NDL for medium-heavy atomic mass targets. Our python-based software extracted the DDXs from NDL to construct the neutron energy spectra, which were later analyzed considering the abundances of each target's isotopes, energy resolution of photon beam, and energy resolution of neutron detectors. For Pb, Au, Sn and Cu targets, the experimental DDX data at neutron energy higher 4 MeV are larger than the DDX values obtained from NDL. For Fe and Ti targets, the experimental DDX data are quite consistent with that of the NDL. The inconsistency between the neutron spectra of experimental data and NDL implies a need to improve physical models in generating the spectrum of photoneutron.

Keywords: differential double cross-section, photoneutron, polarized monoenergetic photon.
[1] T.K. Tuyet, T. Sanami, H. Yamazaki et al., “Energy and angular distribution of photo-neutrons for 16.6 MeV polarized photon on medium-heavy targets”, Nuclear Inst. and Methods in Physics, A 989 (2021)164965.

Speaker: Dr Kim Tuyet Tran (SOKENDAI)
• 07:48
Activation Cross Section Measurement of the (n,2n) Reaction on 203Tl 3m

Studies of neutron-induced reactions are of considerable significance, both for their importance to basic research in nuclear physics and for practical applications in nuclear technology, medicine and industry. Thallium is widely used in electronics, pharmaceuticals, fiber optics, infrared detectors and nuclear medicine. However, little information is available in literature for neutron induced reactions on Tl isotopes, with many discrepancies among the existing experimental data, especially in the energy region above ~14 MeV.
The aim of the present work was to study the cross-section of the (n,2n) reaction on 203Tl, by irradiating a natural TlCl pellet target with monoenergetic neutron beams at 16.9 and 18.9 MeV. The cross section of the 203Tl(n,2n)202Tl reaction was measured implementing the activation method, with respect to the 197Au(n,2n)196Au and 27Al(n,α)24Na reference reactions. The monoenergetic neutron beam was generated at the 5.5 MV Tandem accelerator of NCSR Demokritos, using the 3H(d, n)4He reaction. The target and reference foil assembly was placed at approximately 1.5 cm from the tritium target, thus limiting the angular acceptance to ± 23.5°, where the produced neutrons are practically isotropic and monoenergetic. The fluctuation of the neutron beam flux was monitored with a BF3 detector located 3 m from the neutron source. After the irradiation, the induced activity of the samples was measured with a HPGe detector of 80% efficiency, which was properly shielded with a lead block to reduce the contribution of natural radioactivity. Monte Carlo simulations using the MCNP code have been performed to take into account the gamma-ray self-absorption within the samples as well as the attenuation of the neutron flux through the target and reference foils.
Theoretical calculations with the EMPIRE code will also be presented, using the same parametrisation implemented in the theoretical study of Ir [1] and Au [2] nuclei, in an attempt to study the systematics and find a suitable statistical model for the description of all the experimental results in this mass region.

1. A. Kalamara et al., Phys. Rev. C 98, 034607 (2018).
2. A. Kalamara et al., Phys. Rev. C 97, 034615 (2018).
Speakers: Maria Diakaki (Department of Physics, National Technical University of Athens, Zografou campus, 15780 Athens, Greece), Maria Diakaki (Department of Physics, National Technical University of Athens, Zografou campus, 15780 Athens, Greece)
• 07:51
Fusion excitation function studies in the reactions forming medium heavy compound system 3m

Nuclear fusion is a complex process involving a complete re-arrangement of quantum systems with many degrees of freedom. Heavy ion fusion exhibits a strong entrance channel dependence at near fusion barrier energies and could be reasonably explained by coupled channels formalism [1,2] that explicitly includes the effects of internal degrees of freedom. Though tremendously successful in describing the data at near barrier energies, coupled channels theories fail to describe fusion at deep sub barrier and well above the fusion barrier energies [3]. Alternate theories [4,5] proposed recently also could not give a reasonable explanation to the experimental observations till date. Scarcity of experimental data spanning deep sub barrier to well above the barrier also adversely affect the theoretical developments to a great extent. It is pointed that instead of a coherent coupled channels approach, a model with a gradual onset of decoherence [6] between the superposed states may be required to describe fusion.

Evaporation residue (ER) excitation function measurements have been performed using the heavy ion reaction analyzer (HIRA) [7] at the Inter University Accelerator Center, New Delhi. The measurements were performed for the $^{16}$O+$^{142,150}$Nd reactions in the beam energy range 12% below the barrier to 50% above the barrier. Pulsed beams of $^{16}$O with a pulse separation of 4 μs, was used to bombard the isotopically enriched $^{142}$Nd and $^{150}$Nd targets. The measured ER cross section represents the total fusion cross section at near barrier energies, as fission cross section is negligibly small.

The fusion cross section at different energies is estimated by using the standard expression [8]. Coupled channels code CCFULL has been used to analyse the measured fusion cross sections for the two reactions. Nuclear potential parameters V$_0$, $r_0$ and $a$ were first fixed using the Akyüz-Winther parameterisation [9]. The measured excitation function is found to be significantly enhanced relative to the one-dimensional barrier penetration model (1-DBPM). Vibrational coupling of $2^+$ and $3^-$ states of $^{142}$Nd is found to explain the sub barrier fusion enhancement in $^{16}$O+$^{142}$Nd reaction. Vibrational effects of $^{16}$O seem to play no role in sub-barrier fusion in this reaction. The degree of fusion enhancement is larger for the $^{16}$O+$^{150}$Nd reaction, compared to $^{16}$O+$^{142}$Nd. Rotational couplings of ($2^+$ and $4^+$states) of the deformed target reproduced the fusion cross sections reasonably well in $^{16}$O+$^{150}$Nd reaction at sub- and near barrier energies.

Though CC calculations assuming AW potential parameters reasonably reproduce the fusion cross sections at below barrier energies, they overpredict fusion at energies well above the barrier. This difference is observed to increase with increasing beam energy. Careful analysis indicates that the observed difference is not due to fission or α particle emission. Diffuseness parameter in the range 1.0 - 1.1 fm is required to fit the cross sections. These values are significantly larger than the values obtained from elastic scattering measurements. The inadequacy of AW potential parameters hints the role of dynamical effects in fusion at higher energies.

Speaker: Visakh A C (Central University of Kerala)
• 07:54
Using Monte-Carlo method to analyze experimental data and produce uncertainties and covariances. 3m

The production of useful and high-quality nuclear data requires experimental measurements with high precision and extensive information on uncertainties and possible correlations. When performing experimental measurement data analysis, the physicist uses many external parameters (detector efficiencies, distance of flight, …) in addition to the raw data. All the steps of the data processing (event selection, calibration, …) may introduce uncertainties and correlations.

The usual method for combining and computing uncertainties is to use analytical developments based on the perturbation theory (e.g. 〖u_(f(x))〗^2=(∂f/∂x)^2×u_x^2). This method works well for simple cases, but with multiple parameters and sources of uncertainty, deriving the final total combined uncertainty can be long and complex. Reporting the formula into the analysis code becomes a tedious process where mistakes can appear and the final uncertainty value will be wrong.

Furthermore, it strictly applies only to small deviations from the central values, which is not always the case. Finally, this method makes calculation of covariances hard and the inclusion of some unusual form of uncertainty (assymetric, non Gaussian) even more difficult.

To overcome this issue, one may rely on random sampling methods (a.k.a Monte Carlo). With modern computing infrastructures, large storage space and many computing units are available to process data. This allows, within a reasonable time and coding budget, the repeated processing of data with variation of input parameters within their nominal distribution probability. From the collection of results, a central value, uncertainties and covariances can be extracted easily.

The poster will present an example of how the method is used in production of (n, n’ gamma) cross-sections with uncertainties. The advantages and drawbacks will be discussed, and the method will be compared to the analytical one.

Speaker: Dr Greg Henning (CNRS - IPHC)
• 07:57
Experimental measurement of 183W(n, n’γ) and (n, 2nγ) cross sections (preliminary) 3m

Most nuclear reactor developments rely on evaluated databases for numerical simulations to optimize and predict performance and reactor control parameters. However, these databases still present large uncertainties, preventing calculations from reaching the required precision. An improvement of evaluated databases requires new measurements and better theoretical descriptions of involved reactions. Among the reactions of interest, inelastic neutron scattering (or (n, xn)) are of great importance for the operation of a reactor as they modify the neutron spectrum, the neutron population, and produce radioactive species.

The CNRS-IPHC group is running an experimental program with the GRAPhEME setup installed at the neutron beam facility EC/JRC-GELINA (Geel, Belgium) to measure (n, xn gamma) reaction cross-sections using prompt gamma-ray spectroscopy and neutron energy determination by time-of-flight [1-3]. The obtained exclusive experimental data provide strong constraints on nuclear reaction mechanisms models. A measurement of (n, xn gamma) cross-section have been performed for 183W, which will be discussed in this contribution, in addition to numerous isotopes such as 232Th, 235,238U, natZr and 182,184,186W.

Tungsten is not an active element in nuclear reactors, but, because of its chemical and mechanical properties, it is used in many alloys. The interaction of neutrons with tungsten is therefore of importance for reactor physics, in particular for fusion reactors, in which tungsten is one of the most exposed material to high energy neutrons. From a theoretical point of view, a better description of (n, xn) reactions on tungsten nuclei allows an improvement of models for other key nuclei in reactors fuel. Indeed, tungsten isotopes are deformed like actinides, but also easier to describe as they do not present a neutron-induced fission channel. Still, there are very few measurements available today to test evaluations [4]. Our new experimental data will provide an extensive and constraining test to the predictability of models.

The experimental setup and the data analysis method will be presented. The preliminary experimental results for the 183W isotope will be compared to predictions from the usual nuclear reaction codes.

• [1] “What can we learn from (n, x n γ) cross sections about reaction mechanism and nuclear structure ?”, by Kerveno, Maëlle and Dupuis, Marc and Borcea, Catalin and Boromiza, Marian and Capote, Roberto and Dessagne, Philippe and Henning, Greg andHilaire, Stéphane and Kawano, Toshihiko and Negret, Alexandra and Nyman, Markus and Olacel, Adina and Party, Eliot and Plompen, Arjan and Romain, Pascal and Sin, Mihaela. ND 2019 : International Conference on Nuclear Data for Science and Technology (2019). 10.1051/epjconf/202023901023 https://hal.archives-ouvertes.fr/hal-02957494
• [2] “How to produce accurate inelastic cross sections from an indirect measurement method ?”, by Kerveno, Maëlle and Henning, Greg and Borcea, Catalin and Dessagne, Philippe and Dupuis, Marc and Hilaire, Stéphane and Negret, Alexandru and Nyman, Markus and Olacel, Adina and Party, Eliot and Plompen, Arjan in EPJ N - Nuclear Sciences & Technologies 4, (2018). 10.1051/epjn/2018020 https://hal.archives-ouvertes.fr/hal-02109918
• [3] “From γ emissions to (n,xn) cross sections of interest : The role of GAINS and GRAPhEME in nuclear reaction modeling”, by Kerveno, M. and Bacquias, A. and Borcea, C. and Dessagne, Ph. and Henning, G. and Mihailescu, C. and Negret, A. and Nyman, M. and Olacel, A. and Plompen, M. and Rouki, C. and Rudolf, G. and Thiry, C. in European Physical Journal A 51, 12 (2015). 10.1140/epja/i2015-15167-y https://hal.archives-ouvertes.fr/hal-02154831
• [4] Experimental Nuclear Reaction Data (EXFOR) https://www-nds.iaea.org/exfor/
Speaker: Greg Henning
• 08:00
Experiment-based determination of the excitation function for the production of $^{44}$Ti in proton-irradiated vanadium samples 3m

The artificial production of radionuclides has considerable significance for many different applications. Recently, we observe increasing interest in the field of nuclear medicine, with $^{44g}$Sc being one of the promising nuclides for positron emission tomography. Besides its direct production routes by irradiating suitable targets like Sc or Ti with charged particles, $^{44}$Ti is in the spotlight of radiopharmaceutical research because of the possibility to realize a $^{44}$Ti/$^{44g}$Sc radionuclide generator. Therefore, exploring the possible production routes for the mother nuclide is of great interest, and spallation reactions on V induced by high-energetic protons seem to be one of the most promising ones.
We determined the excitation function for the production of $^{44}$Ti in the nuclear reaction $^{nat}$(p,X) $^{44}$Ti (1). Seven metallic vanadium disks were proton irradiated within an energy range of 111 to 954 MeV between 1995 to 1996. The experiments’ cross sections were determined using two independent measurements by combining $\gamma$ spectrometry with both a low-energy germanium detector and a high-purity germanium detector. The maximum cross section was observed for energies of 145 $\pm$ 1.2 and 150.2 $\pm$ 1.2 MeV with values of 803 ± 28 and 805 $\pm$ 27 $\mu$b, respectively, and are in good agreement with former unpublished results (Fig. 1a). In combination with these earlier measurements, a consistent data set for the $^{nat}$V(p,X) $^{44}$Ti excitation function from 111 to 1350 MeV was obtained. Model calculations using Liège Intranuclear Cascade (INCL)/ABLA reproduce the shape of the excitation function correctly but over-predict the absolute values by factors of 2 to 3 (Fig. 1b). Some considerations regarding this already earlier observed systematic overestimation of neutron-poor residues are discussed and the experimental results will trigger new approaches in code development in order to improve the predictions.

Reference:
(1) Veicht, M., Kajan, I., David, J. C., Chen, S., Strub, E., Mihalcea, I., & Schumann, D. (2021). "Experiment-based determination of the excitation function for the production of Ti44 in proton-irradiated vanadium samples". $\textit{Physical Review C}$, 104(1), 014615.

Acknowledgements:
The authors acknowledge the funding from Swissnuclear (LRC_20_02), from the Marie Skłodowska-Curie Grant (No. 701647), and further, from the Swiss National Science Foundation (Grant No. 177229).

Speaker: Mario Veicht (Paul Scherrer Institut (PSI); École polytechnique fédérale de Lausanne (EPFL))
• 08:03
Measurement of differential and double differential cross-section for natural Gallium at 8.0 MeV neutrons 3m

The metal gallium is a very interesting element on earth. It is a liquid at temperatures greater than 29.76 °C and has a very high boiling point (2229 ◦C). Gallium has widely been used to make alloys with low melting points and has become a candidate element in Chinese Initiative Accelerator Driven Systems (CIADS) project for liquid–metallic coolant [ ]. The neutron cross section of gallium is useful for nuclear device design. The experimental differential cross section data is still rare in fast neutron energy region. The secondary neutron emission differential and double-differential cross sections (DX and DDXs) of n + natGa have been measured at the neutron energy of 8MeV using the multi-detector fast neutron time-of-flight (NTOF) spectrometer in China Institute of Atomic Energy (CIAE). The data was derived by comparing the measured NTOF spectra with Monte Carlo simulation, and corrected with n-p scattering cross section in small angles. Measured differential cross sections are compared with evaluated data base. It was found that the experimental results of angular distribution were in agreement with CENDL-3.2 data base better than ENDF/B-VIII.0 data base.

Speaker: Hanxiong Huang (CIAE)
• 08:06
Target mass dependence of photoneutron spectrum for 16.6MeV photons on medium-heavy mass targets 3m

Experimental data on photoneutron spectrum for reaction of 16.6MeV mono-energetic polalized photons on medium-heavy mass targets were obtained at BL-01, NewSUBARU, Hyogo, Japan. The data consists of low energy neutrons having energy distribution similar to tail of Maxwellain and relatively high energy ones originated from decay of pre-equilibrium status. In this presentation, quantities characterized the spectrum shape are presented as a function target mass. The discussions on the parameters expressing ratio of neutrons from pre equilibrium to equilibrium states are presented in comparison with past experimental data.

Speaker: Toshiya Sanami (KEK/SOKENDAI)
• 08:09
First on-beam measurements with the new detection system for the measurements of light-ion products of nuclear reactions induced by fast neutrons with energies from 18 to 33 MeV 3m

The new detection device for measurements of light ions (p, d, t, α) emitted as products of the
nuclear reactions induced by fast neutrons was recently developed at the Nuclear Physics Institute of
the Czech Academy of Sciences. The device consists of the vacuum chamber and rotational table with
the detection telescopes composed of two Si detectors for dE–E charged particle recognition. It is
coupled with the collimated beams of fast neutrons produced by an isochronous cyclotron U-120M
driven generator. The device can measure the double differential cross-sections with respect to the
angle of emitted light ions and to the incident neutron energy. The experimental data measured on
beams produced with the p(35 MeV)+Be(2.5mm) will be presented. The neutron field
characterizations by Proton-recoil-Method measured with the chamber will be shown and the
measured count rates will be compared with the estimations calculated by Geant4 simulations. In the
future, this new Chamber-for-Light-Ion-Detection (CLID) will produce new differential nuclear data of
high interest for the material applications related to the fusion and aerospace technologies and
potentially test and validate models of nuclear reactions themselves.

Speaker: Mr Martin Ansorge (Nuclear Physics Institute of the Czech Academy of Sciences)
• 08:12
Time dependence of prompt fission $\gamma$-ray emission 3m

Prompt fission $\gamma$ rays (PFG) account for 40% of the total $\gamma$ heating in the core of a typical fast reactor, which in turn constitutes about 10% of the total energy release in fission [1]. Therefore, precise knowledge about multiplicities and $\gamma$-ray energies is important. Prompt fission $\gamma$ rays are commonly defined as those emitted in coincidence with fission fragments, i.e. prior to $\beta$ decay. To assess all those $\gamma$ rays in an experiment is difficult, for two reasons: $\textit{i)}$ nuclear level lifetimes may reach up to microseconds, while a chosen prompt timing window usually is of the order of tens of nanoseconds at most, $\textit{ii)}$ extending the prompt window will include contributions from delayed $\gamma$ rays. Fortunately, the latter effect is negligible.

Not very long ago, the time dependence of PFG spectral characteristics, in particular the multiplicity, was already studied experimentally, however only during the first 10 ns after the spontaneous fission of $^{252}$Cf [2]. In a recent experiment with VESPA (VErsatile $\gamma$ SPectrometer Array) at JRC Geel, we managed to extend this time region considerably, including late $\gamma$-ray emission up to 100 $\mu$s after the instance of fission. Multiplicities were determined for different time bins, covering the entire time range. The observed time dependence is compared to results from calculations [3] with the Monte Carlo Hauser-Feshbach code CGMF [4, 5]. We will give details on the recent experiment, present the obtained results, and discuss them in terms of nuclear lifetimes.

References
[1] G. Rimpault, D. Bernard, D. Blanchet, C. Vaglio-Gaudard, S. Ravaux, A. Santamarina, Phys. Procedia $\textbf{31}$, 3 (2012).
[2] A. Oberstedt, A. Gatera, A. Göök, and S. Oberstedt, Eur. Phys. J. A $\textbf{56}$ 196 (2020).
[3] P. Talou, T. Kawano, I. Stetcu, J. P. Lestone, E. McKigney, and M. B. Chadwick, Phys. Rev. C $\textbf{94}$, 064613 (2016).
[4] T. Kawano, P. Talou, M. B. Chadwick and T. Watanabe, J. Nucl. Sci. Tech. $\textbf{47}$, 462 (2010).
[5] P. Talou, B. Becker, T. Kawano, M. B. Chadwick, and Y. Danon, Phys. Rev. C $\textbf{83}$, 064612 (2011).

Speaker: Andreas Oberstedt (Extreme Light Infrastructure - Nuclear Physics)
• 08:15
Prediction of prompt neutron spectra of the photon induced reactions on the 238U and 232Th targets at incident energies from 4 and 22 MeV 3m

The prediction of prompt neutron spectra of the photon-induced reactions on the 238U and 232Th targets are needed in the treatment of experimental data of the reactions induced by quasi monochromatic γ-ray beams produced in laser Compton-scattering at the NewSUBARU synchrotron radiation facility.
Unfortunately any experimental information concerning both the fission fragments and the prompt emission quantities for the monochromatic photon induced fission of 238U and 232Th is completely missing. Consequently the only way to predict the prompt neutron spectra of γ + 238U and γ + 232Th at incident energies Eγ ranging from 4 to 22 MeV, remains the use of the most probable fragmentation approach (known as the Los Alamos (LA) model) with input parameters provided by systematics. The LA version of Madland and Kahler [1] which consider different residual temperature distributions of the light and heavy fragment and the recent systematic of its input parameters [2] are employed.
At Eγ up to about 12 MeV, the spectra of prompt neutrons emitted only from the fission fragments of the main nucleus undergoing fission (i.e. 238U and 232Th, respectively) are considered. At higher Eγ multiple fission chances occur, so that the contributions of both, the prompt neutrons emitted from the fission fragments (FF) of each fissioning nucleus formed at the respective Eγ and of the so-called pre-fission neutrons must be taken into account.
At Eγ up to 22 MeV, only the fissioning nuclei of the main chain are involved (i.e. 238-236U, 232-230Th), their average excitation energies (at which the prompt emission calculations are performed) being given by iterative equations according to Refs.[3 - 6] and references therein.
The individual spectra of prompt neutrons emitted from the fission fragments Ni(E) of each fission chance (indexed i) and of each evaporated pre-fission neutron φev i(E) are then calculated without to need of the fission probabilities.
The total prompt neutron spectrum at a given Eγ is obtained as N(E) = NFF(E) + Nprefiss(E) in which both components NFF(E) and Nprefiss(E) are calculated as a superposition of individual spectra Ni(E) and φev i(E) respectively, weighted with the fission probabilities of each chance PFi and the average numbers of prompt neutrons. The fission probabilities can be taken as photo-fission cross-section ratios resulting from nuclear reaction calculations, i.e. RFi = σγ,xnf / σγ,F (in which x = i-1).
The present prediction of prompt neutron spectra is based on RFi obtained from the photon induced reaction calculations performed with the EMPIRE code, according to Ref.[7]

[1] D.G.Madland and A.C.Kahler, Nucl.Phys.A 957 (2017) 289-311 and corrigeum Nucl.Phys.A 961 (2017) 216-217.
[2] A.Tudora, Eur.Phys.J. A, 56 (9), (2020) 225.
[3] A.Tudora, G.Vladuca, B.Morillon, Nucl.Phys.A 740 (2004) 33-58.
[4] R.Capote et al. Nucl.Data Sheets 131 (2016) 1-106.
[5] A.Tudora, F.-J.Hambsch, V.Tobosaru, Phys.Rev.C 94 (2016) 044601.
[6] A.Tudora, F.-J.Hambsch, V.Tobosaru, Nucl.Sci.Eng. 192 (2018) 52-69.
[7] M.Sin, R.Capote, M.W.Herman, A.Trkov, B.V.Carlson, Phys.Rev.C 103 (2021) 054605.

Speaker: Anabella Tudora (University of Bucharest)
• 08:18
Application of Multipole humped fission barriers in the Unified Hauser-Feshbach and Exciton Model for Fission Nuclei 3m

We have added the multipole humped fission barriers model into the Unified Hauser-Feshbach and Exciton Model for Fission Nuclei (FUNF). With the help of the MINUIT, we optimized the parameters of FUNF to reproduce the cross-section of neutron-induced $^{238}U$. The double-humped fission barrier can be used to explain the low energy resonance structure of $n+^{238}U$ fission cross-section at low energy.

Speaker: Yuan Tian (China Institute of Atomic Energy)
• 08:21
Role of transport coefficients in fission dynamics 2m

The Langevin approach has been widely used in the study of fission. In this approach, the fission process is described as the time evolution of the shape deformation of the nucleus, and the fragment mass and kinetic energy distributions can be calculated. The physical inputs for the Langevin equation are the potential energy of deformation and two transport coefficients: the inertia and the friction tensor. The inertia tensor represents the geometric structure of the coordinates, while the friction tensor represents the coupling between collective and thermal nucleonic degrees of freedom. The fission process is mainly governed by the potential energy, such as the saddle points and the fission valleys. In the range from the ground state to the fission barrier, nuclear deformation can be roughly understood as a thermal equilibrium distribution due to the potential. On the other hand, the process from the saddle to scission is a dynamical motion governed by both the potential and the transport coefficients.

To elucidate the role of the transport coefficients, we have adopted two alternative methods: the Smoluchowski limit and the Metropolis walk [1, 2]. The Smoluchowski limit corresponds to the strong friction limit of the Langevin equation, which does not include the inertia tensor but depends on the friction tensor and the potential. The Metropolis walk is a method to simulate a random walk on the potential. This method chooses to stay or to move to the next point depending on the thermal probability based on the Boltzmann factor. With this method, the fission process can be calculated using only the potential. It should be noted that the Metropolis walk provides the evolution of the shape, but it is not the time evolution.

In this study, we compare the results calculated with the three methods in the actinide region. First, we focus on $^{264}$Fm which can be divided into a pair of stable double magic nuclei $(Z,\, N) = (50,\,82)$ and hence strongly induces 132+132 symmetric fission. To simplify the comparison, the temperature is fixed at 1 MeV and shell damping is neglected. As collective coordinates, we adopt three deformation parameters $\{\alpha$, $\alpha_1$, $\alpha_4\}$ corresponding to elongation, mass asymmetry, and quadrupole deformation of fragments, respectively, in Cassini shape parameterization [3]. Figure 1 shows the fragment mass distribution calculated with the three methods. In the Metropolis walk, the yield of symmetric fission is much large than that of asymmetric fission. In contrast, in the Langevin equation and the Smoluchowski limit, the yields of symmetric and asymmetric fission are almost the same.

To understand the origin of this different behavior, we analyze the fission paths along the fission valleys. The deformation energy is calculated with the microscopic-macroscopic method [4]. In the three-dimensional deformation space, we obtain the symmetric and the asymmetric saddle points which have almost the same height for $^{264}$Fm. These points are connected to the symmetric and the asymmetric fission valleys, respectively. Near scission, the symmetric fission valley is deeper than the asymmetric one. In the Metropolis walk, it is found that the trajectories move from the asymmetric valley to the symmetric one. On the other hand, in the Langevin equation and the Smoluchowski limit, the change in the mass-asymmetric degree of freedom is suppressed by the friction tensor when the neck becomes thin. It should be noted that the friction tensor is calculated using the completed wall-and-window formula which is a sum of the one-body wall-and-window friction and the friction regarding the mass transfer between the fragments [5]. From this analysis, we conclude that the Metropolis walk needs to be used with caution, in particular, when the distribution near the saddle region is different from that near scission.

The comparison between the Langevin equation and the Smoluchowski limit shows only a small difference. This indicates that the driving and friction force strongly contribute to the fission dynamics. On the other hand, around the saddle point where those forces are weak, it is found that the distribution of the trajectories is wider for the Langevin equation than for the Smoluchowski limit. This can be due to the effect of the inertia tensor.

As another example, we examined $^{236}$U and we found the asymmetric fragment mass distributions in the three models. When the potential is examined, the saddle point of the elongated symmetric fission is higher than that of the compact asymmetric fission and the asymmetric fission valley is more pronounced. The peak position of the mass distribution essentially coincides with the bottom of the valley. In such a case, Metropolis walk may give a similar result with that of the Langevin equation and the Smoluchowski limit.

The yield of the symmetric fission in $^{264}$Fm was smaller than expected. We have performed the four-dimensional Langevin calculation by adding a new parameter $\alpha_6$ corresponding to the octupole deformation of the fragments. In our study of the Fm isotope, it was found that the four-dimensional calculation better describes the saddle point and the valley of the symmetric fission. We found that the symmetric fission is dominant for $^{264}$Fm.

[1] J. Randrup and P. Moller, Phys. Rev. Lett. 106, 132503 (2011).
[2] J. Randrup and P. Moller, Phys. Rev C 88, 064606 (2013).
[3] V. V. Pashkevich, Nucl. Phys. A 169, 275 (1971).
[4] V. M. Strutinsky, Nucl. Phys. A 95 , 420 (1967).
[5] J. Randrup and W. J. Swiatecki, Nucl. Phys. A 429, 105 (1984).

Speaker: Mr Kazuki Okada (Department of Pure and Applied Physics, Kansai University)
• 08:23
Towards a Langevin Model for the Stochastic Dynamics of Nuclear Fission 2m

A robust description of the process of nuclear fission is essential to many research domains ranging from nuclear energy, national security, and nuclear data. However, owing to the nuclear many-body problem, a description of fission based on nucleon-nucleon interactions is unfeasible given current computational limitations, which has led to a number of alternative methods that greatly reduce the overall complexity of this difficult problem. In this work, we present results of recent efforts to model the process of nuclear fission from the perspective of a microscopic-macroscopic model of the atomic nucleus, where fission proceeds from an initially excited state to scission as a stochastic process according to a progression of increasingly sophisticated treatments of the stochastic dynamics. In contrast with our past work, which has treated fission in the strongly-damped limit described by a random-walk process, this approach furnishes the kinetic energies associated with the nascent fragments, which is subsequently used to model the de-excitation properties of fission fragments. LA-UR-21-29350

Speaker: Trevor Sprouse (Los Alamos National Laboratory)
• 08:25
Theoretical calculation and evaluation for n+232,234,236,237U reactions 2m

Abstract
In order to reduce the uncertainties in the design and operation of accelerator-driven systems (ADSs), high-precision nuclear data for neutron- and proton-induced reactions on a variety of isotopes in the energy range below 200 MeV are necessary. As the important component of the spent fuel of current nuclear power plants, U isotopes will be loaded into ADSs to produce energy and neutrons. Therefore, accurate nuclear data for neutron-induced reactions on U isotopes are needed in the calculation of neutron and energy balance and the prediction of transmutation rates of the various radioactive species. To meet this requirement, all cross sections, angular distributions, energy spectra, double differential cross sections of neutron, proton, deuteron, triton, helium-3 and alpha emissions and the number of neutron per fission for n+232,234,236,237U reactions are consistently calculated and analyzed by theoretical nuclear models in the energy range of En≤200 MeV. The optical model, the unified Hauser-Feshbach theory and the exciton model, the evaporation models, the linear angular momentum dependent exciton state density model, the fission model, the intranuclear cascade model, the distorted-wave Born approximation and the coupled channel theory. The calculated results reproduce the experimental data well, and the variation tendency of reaction cross sections related to the target mass numbers is obtained.

Speaker: Yinlu Han (China Institute of Atomic Energy)
• 08:27
(WITHDRAWN) Improving the prediction of TSL on-the-fly based on deep neural networks using experimentally measured double differential data and dynamic structure factor 2m

Improving the prediction of TSL on-the-fly based on deep neural networks using experimentally measured double differential data and dynamic structure factor

Vaibhav Jaiswal$^{1,*}$, Arthur Pignet$^2$, and Luiz Leal$^1$

$^1$Insititut de Radioprotection et de Sûreté Nucléaire (IRSN), Fontenay-aux-Roses, France
$^2$MINES ParisTech, PSL - Research University, Centre for Material Forming (CEMEF), CNRS UMR 7635, CS 10207, Sophia Antipolis Cedex, 06922, France

$^*$Corresponding author: vaibhav.jaiswal@irsn.fr

The slowing down of neutrons generated in a fission reaction, from the MeV range to the meV range, is governed by the scattering process. In the low energy region, when the energy of the neutron is comparable to the energy of the scattering medium, the probability of this interaction is described by Thermal Scattering Law (TSL). TSLs for moderator materials takes into account the dynamical properties of the system and governs the energy and momentum exchange between the neutron and the moderator. The TSLs libraries available to users in the ENDF evaluations are for fixed temperatures, and one makes various approximations and mechanisms to use for non-available temperatures in the evaluations. It is well known that direct interpolation of the TSL to obtain data at intermediate temperature leads to misleading results [1]. A recent approach to overcome this limitation is to interpolate the LEAPR parameters and re-run NJOY to obtain TSL at the required temperatures [2]. This is not a practical solution when multi-physics calculations demand TSL for a very fine grid of temperatures and running NJOY on-the-fly will not be a feasible solution. On the other hand, storing TSLs for a fine grid of temperatures is memory and data intensive to use with Monte Carlo codes.

This paper demonstrates the capabilities of deep neural networks to overcome the above challenges and that TSLs can be reconstructed for any temperatures preserving the dynamics of the system. This approach provides reliable and accurate results within the required memory and size of the data. Several recent works have been carried out in this domain that involves both machine learning and deep learning to carry out nuclear cross-section data processing on-the-fly. In particular, a recent work for TSL, focuses on training the neural network model using the alpha, beta, and temperature as parameters [3]. This paper proposes a new methodology to incorporate not only the alpha, beta and temperatures, but also the experimentally measured double differential data and the dynamic structure factor to improve the deep learning network's prediction capabilities and accuracy.

Presently, the most common thermal scattering cross-section evaluation tool is the LEAPR module of the NJOY code. In the LEAPR framework, the entire physics related to the dynamic structure factor of the material is reduced as a function of the phonon frequency spectrum. This methodology seems to perform well but can be improved involving the experimentally measured dynamic structure factor and the double differential data. Double differential data and the derived dynamic structure factor for light water data measured at the SNS facility at the Oak Ridge national laboratory [4-5] have been used in this work to train the deep learning model. This paper demonstrates an example of TSL for light water, but this method can be applied on a more generalized approach to proceed towards on-the-fly sampling of the TSL in Monte Carlo codes.

[1] W. Haeck and N. Leclaire. “Thermal scattering data and criticality safety”, International Conference on the Physics of Reactors “Nuclear Power: A Sustainable Resource”, Casino-Kursaal Conference Center, Interlaken, Switzerland, September 14-19, 2008.
[2] Vaibhav Jaiswal, “Theoretical and experimental approach towards generation of thermal scattering law for light water”, Ph.D. thesis, University of Lille, 2018.
[3] C. A. Manring, A. I. Hawari, “Development of neural thermal scattering (NeTS) modules for reactor multi-physics simulations”, EPJ Web of Conferences 247, 20004 (2021).
[4] Luiz Leal, Vaibhav Jaiswal, and Alexander I. Kolesnikov, “High-resolution neutron time-of-flight measurements for light water at the Spallation Neutron Source (SNS), Oak Ridge National Laboratory”, EPJ Web of Conferences 239, 14005 (2020).
[5] Vaibhav Jaiswal, Luiz Leal, and Alexander I. Kolesnikov, “Analysis of the time-of-flight neutron scattering cross-section data for light water measured at the SEQUOIA spectrometer, Spallation Neutron Source (SNS)”, EPJ Web of Conferences 239, 14007 (2020).

Speaker: Vaibhav Jaiswal (Institut de Radioprotection et de Sûreté Nucléaire (IRSN))
• 08:29
Calculation of fission product mass distribution by using a semi-empirical model and neutron multiplicity data 2m

Fission fragments are highly excited after scission, and they de-excite by the emission of neutrons and the subsequent gamma emission. The products before and after the emission of neutrons and gammas are called pre- and post-neutron fission fragments, respectively. Pre- and post-neutron emission fission product yields (FPY) can be related through the emitted neutrons, i.e. neutron multiplicity. To have comprehensive understanding of the fission process, one need to describe both pre- and post-neutron FPY data by taking into account the neutron multiplicity.
In this work, we attempt to describe post-neutron emission FPY by using pre-neutron FPY and neutron multiplicity data. We propose a simple empirical model for the calculation of pre-neutron FPY. Then, the experimental data for the neutron multiplicity of each fission fragment mass are used to calculate the post-neutron FPY. The probability of the number of emitted neutrons from each fragment is not known, and thus a few different assumptions for the neutron multiplicity are made. Despite naive approximations, calculated post-neutron fission product mass distributions reproduce the experimental data well. In this manner, we may describe the pre- and post-neutron emission FPY and the neutron multiplicity in a consistent way.

Speaker: Jounghwa Lee
• 08:31
Thermal scattering libraries for cold and ultra-cold neutron reflector materials 2m

Reflectors play an important role at reactor and spallation neutron sources, providing a means for otherwise lost neutrons to be potentially re-directed towards the neutron science instruments. Such materials typically surround the moderating volume, but can also be placed in the neutron beam extraction area. We present developments of improved modelling methods such as including crystallite effects in traditional beryllium reflectors [1] in addition to developments using the NJOY+NCrystal code for new proposed reflector materials such as Magnesium hydride and deuteride [2]. Of particular interest to very-cold and ultra-cold neutron sources is the possibility of using nanodiamonds in the neutron beam extraction area [2,3]. For the simulations of such a system, we present two alternative approaches. The first being the implementation of a small-angle neutron scattering model in OpenMC, to be combined with the standard bulk diamond ACE file and the second being the implementation of the small-angle scattering model inside NCrystal, which is then called directly during the Monte-Carlo simulation. In this work we present examples of cross-section computed for the above-mentioned materials and examples of Monte-Carlo calculations using the new developments.

This work was funded by the HighNESS project at the European Spallation Source. HighNESS is funded by the European Framework for Research and Innovation Horizon 2020, under grant agreement 951782.

[1] D.D. DiJulio et al., “Impact of crystallite size on the performance of a beryllium reflector”, Journal of Neutron Research, vol. 22, no. 2-3, pp. 275-279, 2020

[2] V. Santoro et al., “Development of high intensity neutron source at the European Spallation Source” Journal of Neutron Research, vol. 22, no. 2-3, pp. 209-219, 2020

[3] V. Nesvizhevsky, et al. "Fluorinated nanodiamonds as unique neutron reflector." Carbon, 130, 799-805, (2018).

Speaker: Douglas Di Julio (European Spallation Source ERIC)
• 08:33
The validation of S(α,𝛽) thermal neutron scattering libraries using pulsed-neutron die-away experiments 2m

Validation of the accuracy of S(α,𝛽) thermal scattering law (TSL) evaluations for moderator materials is an important task for the development of high-performance nuclear engineering systems. Many recent thermal neutron scattering evaluations have had limited experimental validation. Like other nuclear data, validation of TSL libraries has historically been by integral criticality benchmarks. While sufficient for general study, these benchmarks often have limited sensitivity to the tested TSLs, and compounding uncertainties from other nuclear data can make validation ambiguous. In some cases, no criticality benchmarks exist that are sensitive to the TSLs of interest. With the development of high-performance next-generation thermal nuclear reactors, alternative validation of applicable TSLs is of high importance.
By performing thermal neutron measurements via pulsed-neutron die-away (PNDA) experiments, along with parallel simulations, the integral performance of various TSL evaluations can be compared to measured experimental data. An experimental testbed using a D-T neutron generator, moderator sample, and thermal neutron detector was assembled at Rensselaer Polytechnic Institute. A Thermo Scientific D211 Deuterium-Tritium Neutron generator is used to generate 10 𝜇s neutron pulses. Various targets of different sizes and geometries are used to moderate the neutrons. Multiple detector types and configurations were tested to optimize the experiment. The room-temperature polyethylene TSL evaluation is well vetted and is similar across different evaluations. This makes it an ideal evaluation to compare with the experimental results. Measurements with different sample sizes and comparison to simulations will be discussed.

Speaker: Mr Benjamin Wang (RPI)
• 08:35
Evaluation of Thermal Neutron Scattering Law and Cross Sections for Calcium Hydride 1m

Metal hydrides have a variety of mechanical, thermal, and neutronic properties that are ideal for applications to nuclear technologies. As a result, zirconium hydride has been used in the fuel of TRIGA reactors for over half a century. More recently, calcium hydride (CaH${}_{2}$), a saline hydride, has been investigated and shows promise for use as a moderator in microreactors. As of yet, there is not a Thermal Scattering Law (TSL) evaluation for CaH${}_{2}$ in the ENDF database. In addition, the evaluation that exists in the European counterpart, JEFF, contains physical inaccuracies in the Ca portion of the cross section due to limitations in the thermal neutron scattering code used for the evaluation. The purpose of this work is to evaluate the thermal neutron scattering cross sections of CaH${}_{2\ }$from first principles, using accurate physical models, for publication and use in reactor design.

The probability of a thermal neutron to scatter from incident energy E into energy E${}^{'}$ through solid angle $\mathrm{\Omega }$ is described by the double differential scattering cross section and, therefore, the TSL. In the case of CaH${}_{2}$, the incoherent, harmonic, and cubic approximations are implemented in the calculation of the TSL, which is performed using a phonon expansion where all second- and higher-order terms represent inelastic scattering via the exchange of phonons. The phonon density of states (DOS), also known as the vibrational DOS, is a key input in this calculation under the cubic approximation. Using ab initio lattice dynamics (AILD), density functional theory (DFT) calculations were performed using the Vienna ab initio simulation package (VASP) with a GGE-PBE pseudopotential. The structure optimization led to lattice constant and atom positions that differ by less than 1% from experiment and determined that a plane wave cutoff energy of 675 eV and a 9x9x9 Monkhorst-Pack k-point mesh were sufficient. The obtained Hellman-Feynman forces were used in the PHONON code to calculate the phonon dispersion curves and DOS using the dynamical matrix method. Finally, the cross sections were calculated, under the incoherent approximation, using the Full Law Analysis Scattering System Hub, $FLASSH$, a code developed by the Low Energy Interaction Physics (LEIP) group at North Carolina State University.

Consequently, this work produced a phonon DOS that closely matches experiment and thermal neutron scattering cross sections for the three CaH${}_{2}$ nonequivalent sites (Ca, H${}_{1}$, H${}_{2}$) that demonstrate the expected physical behavior. In addition to the incoherent inelastic cross sections, the Ca evaluation contains a coherent elastic component that accounts for interference effects from the entire CaH${}_{2}$ lattice, while the H${}_{1}$ and H${}_{2\ }$evaluations contain incoherent elastic contributions. Additionally, the hydrogen inelastic cross sections demonstrate oscillatory behavior that was predicted by Fermi in 1936.${}^{\ }$The cross sections calculated in this work are being finalized for entry into the ENDF database.

Speaker: Mr Benjamin K. Laramee (North Carolina State University)
• 08:36
Validation of calculational determination of 18O(p,n) secondary neutron field 1m

The medical cyclotrons intended to produce medical isotopes are relatively widespread. Nowadays, it is popular to place small and compact accelerators directly to hospitals. This approach simplifies handling with produced radiopharmaceuticals, but it imposes radiation safety measures during production. Radiation protection issues are gaining on importance especially as production increases on current cyclotrons which leads to higher radiation loads than originally designed. For optimal utilization of isotope production cyclotron, the exact knowledge of leakage neutron field is essential due to the deep penetration ability of the high energy neutrons and accompanied secondary radiation production. Our paper presents measurement of neutron leakage spectra in various angles from open target assembly located in research laboratory of Czech academy of sciences. These spectra are compared with data obtained from compact medical cyclotron IBA Cyclone 18/9 accelerator in the UJV Rez cyclotron laboratory. The neutron spectra were measured by the organic scintillator coupled with fast two-parameter spectrometric system NGA-01 equipped with an active voltage divider. The spectra measurement was accompanied by reaction rates measurement of reactions with various threshold for validation of the shape of neutron spectra. The preliminary results show significant disagreement between experiments and theoretical predictions in both cases, open target assembly and production cyclotron IBA 18/9. These findings could have implications not only to the nuclear data community but also to the production accelerators’ operators at the licensing stage.

Speaker: Marek Zmeškal (Research Centre Řež s.r.o.)
• 08:37
(WITHDRAWN) Short-Lived Fission Product Yield Results At Oregon State University 1m

Fission product yields play an important role for nuclear reactor fuel cycle, nuclear reactor
decay heat, and nuclear reactor waste inventory. Multiple techniques have been used to determine
fission product yields for many decades to support reactor design. For our purposes, we will be
performing $\gamma$-ray spectroscopy on $\beta$-decaying fission products following a prompt fission neutron
spectrum irradiation. Previous work has been performed using the Godiva-IV critical assembly to
determine fission product yields starting 1 hour after irradiation and data was recorded for 7 days.
In order to observe shorter-lived fission products, a new rabbit system and $\gamma$-ray counting setup is
being developed by Lawrence Livermore National Laboratory (LLNL) and Oregon State University
(OSU) for use in a campaign of measurements on $^{238}$U at the OSU TRIGA reactor. The counting
setup will consist of six BGO Compton-suppressed HPGe clover detectors connected to a new
VME-based digital data acquisition system. Data will be recorded in list-mode to analyze time-
dependant behavior of the $\gamma$-spectra. The clover detectors have relative efficiency of 150% (to 3×3
in. NaI(Tl) crystal at 1.33 MeV) and the solid angle coverage of four detectors is approximately
9% for the array. The $^{238}$U samples will be delivered in front of the detectors directly from the
reactor core via the rabbit facility. One complication with rabbit systems is that there may exist trace elements that will undergo neutron capture reactions and introduce extraneous $\gamma$-rays in the
time scale of interest. This prompted a study of different pure polyethylene sample materials from
various suppliers at the OSU TRIGA reactor and the development of a new rabbit. Results from this measurement will be discussed.

Speaker: Aaron Tamashiro (Oregon State University)
• 09:00 09:30
Break 30m
• 09:30 11:00
Facilities: II Sutter's Fort ()

### Sutter's Fort

Convener: Emilio Andrea Maugeri (PSI)
• 09:30
(WITHDRAWN) Combining DICER and DANCE to Obtain Improved Neutron Cross Sections and Resonance Parameters for Very Small Samples 24m

P. E. Koehler1, A. Stamatopoulos1, E. Bond2, T. A. Bredeweg2, A. Couture1, B. DiGiovine1, M. E. Fassbender2, A. C. Hayes-Sterbenz3, G. Keksis2, A. Matyskin2, K. Parsons4, G. Rusev2, J. Ullmann1, C. Vermeulen2

1. Physics division, Los Alamos National Laboratory, 87545, NM, USA
2. Chemistry division, Los Alamos National Laboratory, 87545, NM, USA
3. Theory Division, Los Alamos National Laboratory, 87545, NM, USA
4. X-Computational Physics division, Los Alamos National Laboratory, 87545, NM, USA

Combined measurement and analysis of neutron total and capture cross sections results in more accurate cross sections and resonance parameters for applications and for testing and improving theory. The results are even more valuable when the apparatus is capable of enhanced resonance spin and parity determination. Furthermore, the ability to make these measurements on very small samples extends the capability to a wider range of nuclides. The new Device for Indirect Capture Experiments on Radionuclides (DICER) [1] at the Los Alamos Neutron Science Center (LANSCE) is being developed to make neutron transmission measurements on very samples. The current 1-mm-diameter collimation system typically requires samples in the mg range although samples as small as about 1 g are possible in exceptional circumstances. Typical neutron transmission measurements requires samples in the grams to tens of grams or more range. DICER will soon be upgraded to a 0.1-mm-diameter collimation system, allowing routine measurements on samples in the 100 g range. DICER features a unique, innovative binocular collimator system which enables simultaneous sample-in and sample-out measurements, thereby minimizing both the sample size and measurement time. The Detector for Advanced Neutron Capture Experiments (DANCE) at LANSCE has for years measured neutron capture cross sections for samples in the mg range. DANCE also has been shown to be an excellent neutron resonance “spin meter” [3, 4]. In addition to spin and parity, each neutron resonance has both a neutron and a gamma width; hence, both neutron capture and total cross sections are needed to fully determine all the parameters for as many resonances as possible. Because they have a wider range of resonance spins as well as typically large cross sections, odd-A nuclides in particular represent a fertile testing ground [3, 5] for theory as well as a challenge for applications [6, 7]. The combined power of DICER and DANCE will be demonstrated via new data and combined resonance analysis on 147,149Sm and 191,193Ir.
[1] A. Stamatopoulos et al., submitted to Nucl. Instrum. Meth. A (2021).
[2[ A. Stamatopoulos et al., submitted to this conference.
[3] P. E. Koehler et al., Phys. Rev. Lett. 108, 142502 (2012).
[4] S. A. Sheets et al., Phys. Rev. C 76, 064317 (2007).
[5] P. E. Koehler et al., in Compound-Nuclear Reactions, CNR*18, editedby J. Escher, Y. Alhassid, L. A. Bernstein, D. Brown, C. Frohlich, P. Talou, and W. Younes (Springer 2021) p. 187.
[5] L. C. Leal, H. Derrien, M. E. Dunn, and D. E. Mueller, “Assessment of Fission Product Cross-Section Data for Burnup Credit Applications”, ORNL/TM-2005/65.
[6] N. Leclaire, I. Duhamel, and M. Monestier, in Criticality safety – Pushing Boundaries by modernizing and Integration Data, Methods, and Regulations, edited by C. Associates (American Nuclear Society, 2017), p. 15.

Speaker: Paul Koehler (LANL)
• 09:54
New Capabilities of the RPI Gamma-Multiplicity Detector to Measure Gamma Production 12m

Accurate gamma production in capture reactions is critical for simulation of nuclear reactor applications. This includes modeling of strength functions and level densities and calculating cross sections from resonance parameters. To improve this work, the Rensselaer Polytechnic Institute (RPI) 16-segment gamma-multiplicity NaI(Tl) detector at the Gaerttner Linear Accelerator Center (LINAC) has been upgraded by implementing a digitizing data acquisition system. The new digitized system can measure the gamma energy distribution in each individual detector, and gamma-multiplicity values as a function of neutron time-of-flight. With the new capabilities, high precision capture (and fission) yield measurements can be made, and the accuracy of simulation tools used to predict the capture gamma cascades can be tested. To validate the yield measurements completed with the updated system, an experiment was performed using a natural Ta sample to measure $^{181}$Ta and $^{180m}$Ta resonance capture yield as a function of neutron energy by detecting prompt gammas emitted from neutron capture interactions using the time-of-flight method. Capture yield was also calculated as a function of the measured gamma multiplicity of each capture event. The Ta results confirm earlier measurements and agree with theoretical yield in the low energy resonance region from 1 to 20-eV. This measurement also includes the 0.01-eV thermal region, where there is currently little experimental data. In addition to the yield validation, a measurement of $^{238}$U was performed to generate gamma emission spectra for observed multiplicities. For capture gamma cascades where the total gamma energy deposition is close to the neutron binding energy, gamma emission spectra were measured for individual resonance energies for the $^{238}$U(n, γ) reaction. The results were compared to a recent measurement done using the Detector for Advanced Neutron Capture Experiments (DANCE) array at Los Alamos Neutron Science Center (LANSCE), and Monte-Carlo n-particle simulations. The measured gamma emission spectra for observed two step cascades had general agreement with the LANSCE shape; however, there were noticeable differences between the current measurement (RPI) and previous work (LANSCE).

Speaker: Katelyn Cook (Rensselaer Polytechnic Institute)
• 10:06
Combining neutron/gamma radiography and tomography at CNA 12m

Centro Nacional de Aceleradores (CNA) is a joint interdisciplinary research center from Universidad de Sevilla, Junta de Andalucía and CSIC (Spanish National Research Council) open to external users. CNA has six different facilities, namely: a 3 MV Pelletron tandem accelerator, an 18/9 MeV Cyclotron accelerator, a 1 MV mass spectrometer accelerator, a PET/CT scanner, a 200 kV radiocarbon dating system, and a Co-60 irradiator. In addition, a Mo x-ray tube is available for cultural heritage studies, among other uses.

Radiography is a non-destructive imaging technique that uses the attenuation of penetrating radiation passing through a sample to study its internal structure. Although the term usually refers to x-rays, different kinds of radiation can be used as a probe. Each type of radiation features different mechanisms of interaction with matter, revealing different structures and properties of an object. Therefore, the combined information from different types of radiographies of a given object provides a more complete knowledge about said sample.

This work presents the commissioning of a mobile imaging setup at CNA with the capability of combining four different radiography types taking advantage of the different facilities available at the center: fast neutron radiography, thermal neutron radiography, gammagraphy, and conventional x-ray imaging. Among these, fast neutron imaging is the only technique that is not yet commercially available despite having great interest in industry due to the high penetration range and the possibility of contrast under strongly shielded objects.

At CNA, acceptable quality images in exposures of several minutes can be obtained, distinguishing between different materials and different thicknesses of the same material and resolving millimetric structures. Furthermore, since 2021, neutron and gamma tomography are also available, complementing the flexibility of this setup with volumetric information of a sample. This new feature consolidates the imaging techniques as a new tool at CNA, with widespread use in industry and research, particularly interesting for cultural heritage applications.

Speaker: Ms Maria de los Ángeles Millán Callado (Centro Nacional de Aceleradores (CNA, CSIC-US-Junta de Andalucía) - Dpto. Física Atómica, Molecular y Nuclear (FAMN), Universidad de Sevilla (US))
• 10:18
GENESIS: A $\gamma-n$ coincident spectrometer for nuclear data inquiry 12m

The Gamma Energy Neutron Energy Spectrometer for Inelastic Scattering (GENESIS) was commissioned at the 88-Inch Cyclotron at Lawrence Berkeley National Laboratory (LBNL). The array consists of up to 27 EJ-309 organic scintillators capable of $n-\gamma$ particle discrimination. The $\gamma$-ray detectors include two Compton-suppressed 4-fold segmented Eurosys CLOVER HPGe detectors, several ORTEC PopTop single-crystal HPGe detectors, and LaBr inorganic scintillators. GENESIS includes a custom software analysis suite and modeling framework. The modeling framework for the system is a Geant4-based modular simulation, which can be easily adapted to match changes in the experimental configuration as needed. In order to benchmark the model and provide system characterization, several auxiliary measurements were conducted. These include a long-dwell-time measurement using a 10~mCi $^{252}$Cf source for validating the neutron detection array model, coincident timing resolution measurements, and coincident time calibration measurements. Initial experiments have focused on $(n,n'\gamma)$ reactions, but the use of GENESIS for providing measurements of correlated fission observables is also being explored. Preliminary results from a measurement using a 42~g natural uranium target will be presented. The GENESIS array will enable a variety of nuclear physics measurements providing differential measurements of correlated observables for neutrons and $\gamma$ rays.

Speaker: Josh Brown (UC Berkeley)
• 10:30
The new Device for Indirect Capture Experiments on Radionuclides at LANSCE: Efforts on measuring the resonance(s) responsible for the extremely large 88Zr(n,g) cross section 12m

The new Device for Indirect Capture Experiments on Radionuclides at LANSCE: Efforts on measuring the resonance(s) responsible for the extremely large 88Zr(n,g) cross section

A. Stamatopoulos1, P. Koehler1, E. Bond2, T. A. Bredeweg2, A. Couture1, B. DiGiovine1, M. E. Fassbender2, A. C. Hayes-Sterbenz3, G. Keksis2, A. Matyskin2, K. Parsons4, G. Rusev2, J. Ullmann1, C. Vermeulen2

1. Physics division, Los Alamos National Laboratory, 87545, NM, USA
2. Chemistry division, Los Alamos National Laboratory, 87545, NM, USA
3. Theory Division, Los Alamos National Laboratory, 87545, NM, USA
4. X-Computational Physics division, Los Alamos National Laboratory, 87545, NM, USA

The thermal neutron capture cross section of 88Zr was recently reported1 to be the second largest in nature with the largest resonance integral2 measured. Presumably, these very large values are caused by a resonance or resonances very near thermal energy, and determining their energies and widths, and hence the shape of the cross section away from thermal energy, would be very useful for applications. The short half-life (83.4 days) and associated large background renders direct measurements of the neutron capture cross section impossible using current techniques. However, it is possible to measure the total neutron cross section, and hence the resonance properties, using the newly commissioned Device for Indirect Capture Experiments on Radionuclides (DICER)3 at the Los Alamos Neutron Science Center (LANSCE). Transmission measurements are utilized as a surrogate method to perform capture measurements. The 88Zr needed for a DICER measurement was produced at the Isotope Production Facility (IPF) and cleanly separated from the production target material. The final steps of loading the ~1 g 88Zr sample into DICER and performing the measurement are expected to be completed very soon. A description of the new instrument and efforts on 88Zr will be presented.

1. J. Shusterman et al., Nature volume 565, pages 328–330 (2019)
2. J. Shusterman et al., Phys. Rev. C 103, 024614 (2021)
3. A. Stamatopoulos et al., submitted to Nucl. Instrum. Meth. A (2021)

LA-UR-21-29581

Speaker: Thanos Stamatopoulos (Los Alamos National Laboratory)
• 09:30 11:00
Fission: IV American River ()

### American River

Convener: Maria Anastasiou (LLNL)
• 09:30
The fast neutron induced fission of actinide nuclei-TKE and Mass Distributions 12m

The study of the fast neutron (En = 3-100 MeV) induced fission of actinide nuclei allows us to study the relative role of Coulomb and dissipative forces in a large scale nuclear collective motion. One is aware of the practical applications of knowing the energy release and its dependence on neutron energy and fragment mass. We have been engaged in studies of the total kinetic energy release in the fast neutron induced fission of 232Th, 233U, 235U, 237Np and 239Pu, which we will summarize. We have prepared the targets for these studies by vapor deposition which leads to very uniform, high purity targets, an essential component of such studies.

Our experiments were done at the LANSCE facility at LANL using the WNR “white” neutron spectrum beams, with typical intensities of 105 n/s. The neutron energies were measured by time of flight with typical uncertainties of 5%. Fission fragments were detected by 9 pairs of Si PIN diode detectors. Corrections were applied for the pulse height defect of the detectors and the fragment energy loss in the target and backing. Our apparatus, targets, etc. were benchmarked by measuring the TKE release in the thermal neutron induced fission of 233U, 235U and 239Pu. No normalizations of the data were needed and thus our measurements are absolute measurements.

The TKE distributions for each system studied were Gaussian in shape. In Figure 1, we show the measured TKE values (as a function of neutron energy) for several of the systems we have studied. The data are in good agreement with previous work with some deviations at higher energies. The data can be parameterized in terms of Viola scaling. The variances of the TKE distributions show jumps at neutron energies corresponding to 2nd, 3rd, etc, chance fission.
The fission mass distributions show the expected evolution from asymmetric fission at low neutron energies to symmetric fission at higher neutron energies. The data clearly show the importance of shell closures near A=134 reflecting the extra stability of Z=56, due to octupole deformation. The data are in general agreement with the predictions of the GEF model but with smaller jumps in the TKE release at nth chance fission energies.

Speaker: Walter Loveland
• 09:42
Experimental Prompt Fission Neutron Spectra for the $^{235,238}$U(n,f), $^{239}$Pu(n,f), and $^{240,242}$Pu(sf) Reactions 12m

The Chi-Nu project at the Los Alamos Neutron Science Center (LANSCE) continues to measure Prompt Fission Neutron Spectra (PFNS) as accurately and completely as possible, with fully understood uncertainties including uncertainty correlations. These measurements use the LANSCE/WNR white-spectrum neutron beam and cover incident neutron energies from below 1 MeV to 20 MeV, and were originally focussed on the major actinides. However, the program has grown to include PFNS measurements for neutron-induced fission of $^{240}$Pu, as well as some spontaneous fissioning nuclei. Two neutron detector arrays are used, combined with multi-cell Parallel Plate Avalanche Counters[3] or a single-cell fission chamber (for $^{242}$Pu) for fission detection. Detailed Monte Carlo simulations of the experimental setup were used to model the neutron response and determine PFNS for each case, as a function of incident neutron energy for the neutron-induced cases. The experimental details are discussed in detail in Ref. [1].

To date, Chi-Nu PFNS data have been reported on neutron-induced fission of $^{239}$Pu[1] and $^{235}$U[2], and data on neutron-induced fission of $^{238}$U and spontaneous fission of $^{242}$Pu and $^{240}$Pu have been taken. These data will be presented and compared with prior data and current evaluations. All of these data sets were taken using similar detector arrangements and analysis methods, so comparisons of the PFNS of the various isotopes can be made with improved accuracy, as some of the uncertainties are similar in all cases. We will discuss a few such comparisons. Such comparisons will also be discussed and compared to evaluations and integral experiment results.

This work is supported by the U.S. Department of Energy under Contracts No. 89233218CNA000001 (LANL) and DE-AC52-07NA27344 (LLNL).

[1] K.J. Kelly, et al., Phys. Rev. C 102, 034615 (2020).
[2] M. Devlin, et al., Nucl. Data Sheets 148, 322 (2018).
[3] C.-Y. Wu, et al., Nucl. Instr. And Meth. A794, 76 (2015).

LA-UR-21-29659

Speaker: Matthew Devlin
• 09:54
Fission Product Yields from Neutron-Induced Fission of Major Actinides at 6.5 MeV 12m

Fission product yields (FPY) are an important observable for fundamental and applied nuclear physics. We are conducting a comprehensive study of the cumulative FPY from neutron-induced fission of $^{235}$U, $^{238}$U, and $^{239}$Pu between 0.5 and 15 MeV at the Triangle Universities Nuclear Laboratory. Actinide foils are irradiated using monoenergetic neutron beams produced by a tandem accelerator. The activation foils are placed in a dual fission chamber with thin reference foils of the same isotope to determine the total number of induced fission events in the target. Fission products in the sample are identified following gamma-ray spectra measurements with high-purity germanium detectors in a low-background counting facility. Gamma rays from decay of about 40 fission products with half-lives between 10 minutes and a few days have been measured after 1 hour of irradiation. Details of the experiment setup and preliminary results for FPY at a neutron energy of 6.5 MeV will be presented.

This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

Speaker: Ronald Malone
• 10:06
Measurement of Independent Fission Product Yields with SPIDER 12m

Independent fission product yields (FPY), i.e., yields of fission products right after the prompt neutron emission, but before beta decay, are an important piece of data for nuclear fission modeling and fission applications. The Spectrometer for Ion Determination in fission Research (SPIDER), was developed at Los Alamos Neutron Science Center (LANSCE) for measuring FPYs from neutron-induced fission using the 2E-2v method, eventually spanning from thermal up to 20 MeV in incident neutron energy. SPIDER has recently undergone various improvements for increasing the fidelity and resolution of the extracted data. In particular, a gamma-ray tagging system has been implemented for improving the accuracy of the mass calibration by measuring strong gamma-ray transitions from isotopes of known mass. In this presentation, an overview of the upgraded SPIDER system and preliminary results from FPY measurements on $^{252}$Cf(sf), $^{235}$U(n$_{th}$,f), and $^{239}$Pu(n$_{th}$,f) will be discussed.

*This work is supported by the US Department of Energy through the Los Alamos National Laboratory, Contract No. 89233218CNA000001 (LA-UR-21-29666)

Speaker: Panagiotis Gastis
• 10:18
The Extension of the Deterministic Hauser-Feshbach Fission Fragment Decay Model to Multi-chance Fission and its Extension to $^{239}$Pu 12m

The Hauser-Feshbach fission fragment decay model HF$^3$D calculates the statistical decay of fission fragments through both prompt and delayed neutron and $\gamma$-ray emissions in a deterministic manner. While previously limited to the calculation of only first-chance fission, the model has recently been extended to include multi-chance fission, up to neutron incident energies of 20 MeV. The deterministic decay takes as input pre-scission quantities—fission probabilities, pre-fission neutron energies, and the average energy causing fission—and post-scission quantities—yields in mass, charge, total kinetic energy, spin, and parity. From those fission fragment initial conditions, the full decay is followed through both prompt and delayed particle emissions. The evaporation of the prompt neutrons and $\gamma$ rays is calculated through the Hauser-Feshbach statistical theory, taking into account the competition between neutron and $\gamma$-ray emission, conserving energy, spin, and parity. The delayed emission is taken into account using time-independent calculation using decay data. This whole formulation allows for the calculation of prompt neutron and $\gamma$-ray properties, such as multiplicities and energy distributions, both independent and cumulative fission yields, and delayed neutron observables, in a consistent framework. Here, we describe the implementation of multi-chance fission into the HF$^3$D model, and show an example of prompt and delayed quantities beyond first-chance fission, using the example of neutron-induced fission on $^{239}$Pu. This expansion represents significant progress in consistently modeling the emission of prompt and delayed particles from fissile systems.
LA-UR-21-28700

Speaker: Amy Lovell
• 09:30 11:00
Formats: II + Evaluation Methodology I Delta King ()

### Delta King

Convener: Amanda Lewis
• 09:30
Nuclear data activities for medium mass and heavy nuclei at Los Alamos 24m

Nuclear data is critical for many modern applications from stockpile stewardship to cutting edge scientific research. Central to these pursuits is a robust pipeline for nuclear modeling as well as data assimilation and dissemination. We summarize the ongoing nuclear data efforts at Los Alamos for medium mass to heavy nuclei. We begin with a discussion of a novel toolkit for model parameter optimization that is based on a Bayesian technique called hyperparameter optimization. This mathematical framework affords the combination of different measured data in determining model parameters and their associated correlations. It also has the advantage of being able to quantify outliers in data. We exemplify the power of this procedure by highlighting evaluated cross sections along the Pu isotopic chain and emphasize the importance of model consistency in such evaluations. Finally, we highlight the success of our tools and pipeline by covering the insight gained from incorporating the latest nuclear modeling and data in astrophysical simulations as part of the Fission In R-process Elements (FIRE) collaboration.

Speaker: Matthew Mumpower (Los Alamos National Laboratory)
• 09:54
The National Nuclear Data Center is a Public Reusable Research (PuRe) Data Resource 12m

In May 2021, the National Nuclear Data Center (NNDC) was designated by the US Department of Energy Office of Science (SC) as a Public Reusable Research (PuRe) Data Resource. The DOE defines "PuRe Data Resources are data repositories, knowledge bases, analysis platforms, and other activities that make data publicly available to enable better communication, better stewardship, and better science." The NNDC is the only PuRe resource in SC's Nuclear Physics program. Such a designation is a source of great pride for the NNDC, but comes with numerous conditions, obligations and opportunities. In this talk, we will outline the NNDC's progress towards implementing the requirements of being a PuRe resource: the generation of Document Object Identifiers for all NNDC managed libraries, modernization of the NNDC website, modernization of the underlying library infrastructure, and development of a robust data preservation strategy.

The work at Brookhaven National Laboratory was sponsored by the Office of Nuclear Physics, Office of Science of the U.S. Department of Energy under Contract No. DE-AC02-98CH10886 with Brookhaven Science Associates, LLC.

Speaker: David Brown
• 10:06
Modernization efforts for the R-Matrix code SAMMY 12m

The R-Matrix code SAMMY \cite{SAMMY} is a widely used nuclear data evaluation code focused on the resolved range,
which includes corections for experimental effects. The code is still mostly written in Fortran 77, and uses
a memory managment system suitable for the time of its initial writing (1984). A modernization effort is under way
to bring the code in-line with modern software development practices. A continuous-integration testing
framework was added, automating the large exisiting set of test cases. It is run on every commit.
The memory management was updated to current standard practices suitable for modern software analysis tools.
The code can be optained from \href{https://code.ornl.gov/RNSD/SAMMY}{https://code.ornl.gov/RNSD/SAMMY}.

The resonance parameters and covariance information are now stored in C++ objects shared by SAMMY and AMPX \cite{AMPX},
the processing code that generates nuclear data libraries for SCALE \cite{SCALE}. This
allows for easier maintainance and access to the resonance parameters inside and outside of SAMMY.
This feature is already used by accessing and changing parameters in memory in the
Bayesian Monte Carlo Evaluation Framework for Cross Sections Nuclear Data and Integral Benchmark Experiments project \cite{ND10},

Further plans include the switch to the ENDF reading and writing routines in AMPX, as these routines
are more robust, easier to maintain, and support more features. Of note here is support for the new GNDS format \cite{GNDS}.
Previously it wasn't easy to share the full covariance matrix for evaluations containing more than one isotope due to
limitations on the ENDF format; this is now supported in GNDS. The data are currently available in a binary SAMMY format and can be exported
to GNDS to make them more widely available and sharable.

The next step will be to use the same resonance processing code at 0K in AMPX and SAMMY as one of the available
Reich-Moore R-Matrix formalism.
The first step toward this goal is to isolate the reconstruction into a module that takes resonance parameters as its input and
does not depend on SAMMY gobal parameters. This goal has been achieved and it should now be possible to more easily
change the resonance formalism
and add enhancements as the Phenomenological $R$-Matrix parameterization of direct, doorway, and compound nuclear reactions discussed
elsewhere on this conference.
This concerted modernization and enhancement effort provides multiple advantages to the nuclear data community.
It will allow parameter optimization using enhanced formalisms, including experimental effects, that better match complex experimental data.
Then those evaluated parameters can immediately be passed off to AMPX to be reconstructed with the exact same cross section model and
be put into a data library for subsequent testing using SCALE and the Valid Benchmark suite \cite{VALID} or other suitable benchmark suites.

{\bf Acknowledgments:} This work was supported by the U.S. Department of Energy (DOE) Nuclear Criticality Safety Program, which is funded and managed by the National Nuclear Security Administration for DOE.

\thispagestyle{empty}
\begin{thebibliography}{1}
\bibitem{SAMMY} N.M. Larson, {\em Updated Users' Guide for SAMMY: Multilevel R-matrix Fits to Neutron Data Using Bayes' Equations"}, ORNL/TM-9179/R8 (2008)

\bibitem{AMPX} D. Wiarda, M. L. Williams, C. Celik, and M. E. Dunn,
{\em AMPX-2000: A Cross-Section Processing System for Generating Nuclear Data for Criticality Safety Applications},
International Conference on Nuclear Criticality Safety (ICNC 2015).

\bibitem{SCALE} W. A. Wieselquist, R. A. Lefebvre, and M. A. Jessee, Eds., {\em SCALE Code System},
ORNL/TM-2005/39, Version 6.2.4, Oak Ridge National Laboratory, Oak Ridge, TN (2020).

\bibitem{ND10} G. Arbanas, J.M. Brown, A.M. Holcomb, and D. Wiarda,
{\em Bayesian Monte Carlo Evaluation Framework for Cross Sections Nuclear Data and
Integral Benchmark Experiments}, in the proceedings of the ANS Winter Meeting, NCSP Session (2020)

\bibitem{GNDS} Specifications for the Generalised Nuclear Database Structure (GNDS)
\href{https://www.oecd-nea.org/science/wpec/documents/7519-GNDS.pdf}{https://www.oecd-nea.org/science/wpec/documents/7519-GNDS.pdf},
(accessed September 15, 2021).

\bibitem{VALID} W. J. Marshall and B. T. Rearden,
{\em The SCALE Verified, Archived Library of Inputs and Data – VALID},
ANS NCSD 2013 - Criticality Safety in the Modern Era: Raising the Bar, Wilmington, NC, September 29 –October 3, 2013, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2013).

\end{thebibliography}

Speaker: Dorothea Wiarda (ORNL)
• 10:18
239Pu R-matrix Analysis and Neutron Multiplicities in the Neutron Energy Region up to a few keVs 12m

The evaluation of $^{239}$Pu neutron resonance parameters coupled to neutron multiplicities $\bar{\nu}_{p}$ is of particular importance to investigate the $(n,\gamma f)$ reaction in which a $\gamma$-ray emission occurs before the scission of the compound nuclear. This reaction has offered one explanation for the fluctuations in the measured values of $\bar{\nu}_{p}$. In this regard, the competition between $(n,\gamma f)$ reaction and the direct fission process can be also included in the $R$-matrix analysis of fission and capture measured data.

The goal of this work is the coupled evaluation of the $n$+$^{239}$Pu resonance parameters and related neutron multiplicities by ensuring the adoption of thermal neutron constants recently evaluated at the International Atomic Nuclear Energy as well as the recommended (thermal-neutron) induced prompt neutron fission spectrum. Moreover, this new set of physical evaluated quantities should also guarantee the agreement for high-leakage solution benchmarks while keeping the good performance of large thermal solution assemblies.

Speaker: Dr Marco Pigni (ORNL)
• 09:30 11:00
Measurements: IV Folsom ()

### Folsom

• 09:30
GRAPhEME: performances, achievements (@EC-JRC/GELINA facility) and future (@ GANIL/SPIRAL2/NFS facility) 24m

GRAPhEME is a $\gamma$ spectrometer developed by CNRS/IPHC Strasbourg (France), in collaboration with EC-JRC/Geel (Belgium) and IFIN-HH Bucharest (Romania). With its 6 High Purity Planar Germanium detectors, GRAPhEME, installed at the EC-JRC GELINA facility, was optimized for measurements of accurate (n,xn $\gamma$) cross sections on actinides. The experimental methodology is based on prompt $\gamma$-ray spectroscopy coupled to time of flight measurements. In a first configuration, involving 4 HPGe, several measurement campaigns have produced numerous sets of data for $^{235}$U [1], $^{238}$U [2] , $^{232}$Th [3] and $^{183,182,184,186,nat}$W isotopes [4]. An update of the setup in 2016, with a segmented (6x6 pixels) HPGe has opened the way for measurements with very active targets. A first campaign on $^{233}$U has been performed and a second one, on $^{239}$Pu is planned in the coming months. Beyond the experimental work, a strong collaboration with theoreticians has emerged allowing the use of the data produced with GRAPhEME to test and constraint nuclear reaction codes [5] like TALYS, CoH and EMPIRE.

In this paper, we would like to present an overview of fifteen years of experiments with GRAPhEME and to highlight the performances and achievements done at EC-JRC GELINA facility. The limitations of our apparatus and methodology will be also described with the solutions we implement continuously to improve GRAPhEME and the associated data analysis methodology. The next challenge tackled with GRAPhEME is to perform (n,2n) and (n,3n) cross sections measurements at the new GANIL/SPIRAL2/NFS facility in Caen, France. A status of this new project will be discussed. This overview paper, if accepted, will be completed by five other submitted papers, which will present more focused items related to GRAPhEME (new inelastic cross section data on $^{233}$U and $^{183}$W, improvement of data analysis methodology, inferring total cross section from measured partial ones, problematic of nuclear structure uncertainties, evaluation).
- On the need for precise nuclear structure data for high quality (n, n'$\gamma$) cross section measurements. G.Henning et al.
- (poster) Using Monte-Carlo method to analyze experimental data and produce uncertainties and covariances. G.Henning et al.
- Experimental measurement of $^{183}$W(n, n'$\gamma$) and (n, 2n$\gamma$) cross sections (preliminary). G.Henning et al.
- Measurement and evaluation of partial and total (n,xn) reaction cross-sections on highly radioactive nuclei of interest for energy production application. F. Claeys et al.
- (poster) From $^{232}$Th(n,n' $\gamma$) cross sections to level production and total inelastic scattering cross section. N. Dari Bako et al.
[1] M.Kerveno et al., Physical Review C 87, 24609 (2013)
[2] M.Kerveno et al., Physical Review C, accepted for publication (2021)
[3] E. Party, Thesis University of Strasbourg, paper in preparation
[4] G. Henning et al., Proceedings of PHYSOR 2020 conference: https://hal.archives-ouvertes.fr/hal-02956052
[5] M.Kerveno et al., Eur. Phys. J Web of Conferences 239, 01023 (2020)

Speaker: Maelle Kerveno
• 09:54
Activation Cross-Sections for Short-Lived Reaction Products on Hafnium Isotopes Induced by 1 – 20 MeV Neutrons 12m

Neutron-induced reaction cross-sections on hafnium isotopes are important for research and nuclear applications. Hafnium is considered as a constituent of the control elements and structural materials of nuclear reactors due to its high absorption cross-section for thermal neutrons, good mechanical properties and extremely high resistance to corrosion. Hafnium is an alloying element of low activation materials. The majority of the neutron-induced reactions on hafnium isotopes populate metastable states. Experimental cross-sections provide a database for investigation of the sensitivity of nuclear models to level properties and decay schemes. The present study is an integral part of previous measurements on Hf, W and Ta isotopes and contributes to the database for consistent nuclear modelling in this mass region. Activation cross-section data for short-lived reaction products on hafnium isotopes are very scarce. Results of cross-section measurements will be presented for the following reactions. 178Hf(n,n’)178m1Hf, 179Hf(n,2n)178m1Hf, 179Hf(n,n’g)179m1Hf, 180Hf(n,2n)179m1Hf. The irradiations were carried out at the 7-MV Van de Graaff accelerator at EC-JRC, Geel. Neutrons in the 1-3 MeV energy range were produced via the 3H(p,n)3He reaction. Deuterium beam and a deuterium gas target were used to produce 5.6 and 6.5 MeV neutrons. For the production of quasi-monoenergetic neutrons between 15 and 20 MeV the 3H(d,n)4He reactions was employed. Both samples with natural composition and isotopic enrichment were employed to differentiate reactions leading to the same product. The half-lives of 178m1Hf and 179m1Hf are 4.0 s and 18.67 s, respectively. An automated pneumatic transport system was used for sample transport from the irradiation to the measurement position. Cycles of irradiations and measurements were repeated to enhance counting statistics. Corrections were applied for the gamma-ray self-absorption, secondary background neutrons, coincidence summing correction, interference between reactions leading to the same reaction product. The studied cross-sections were determined relative to 27Al(n,alpha)24Na reaction cross-section. Stacks with monitor foils were irradiated in separate irradiations in order determine flux density distributions. The neutron flux for the short irradiations was determined by normalization to the count rates registered by a Long Counter. Cross-section data for the 178Hf(n,n’)178m1Hf reaction were obtained for the first time. For the other reactions the new result extended the range of the experimental data to higher energy.

Speaker: Valentina Semkova
• 10:06
Cross Section Measurements and Theoretical Study of the $^{174,176}$Hf$(n,2n) ^{173,175}$Hf Reactions 12m

Neutron induced reactions can provide significant information in the field of nuclear physics and applications. Hafnium (Hf) is one of the rare-earth isotopes with a relative large total neutron cross-section in the thermal neutron energy region and neutron induced reactions in reactor materials could lead to the production of long-lived isomeric states of Hf isotopes. Thus, the knowledge of neutron cross-sections on Hf isotopes is of great importance both for applications concerning the interaction of neutrons with matter as well as for testing nuclear models.

In this study, experimental cross section measurements for the $^{176}$Hf$(n,2n)^{175}$Hf and $^{174}$Hf$(n,2n)^{173}$Hf reactions were carried out, using the activation technique. The neutron beam energy at 15.3 and 17.1 MeV was produced via the $^{3}$H(d,$n)^{4}$He reaction at the 5.5 MeV Tandem Van de Graaf accelerator laboratory of NCSR “Demokritos”. A thin metallic foil of natural Hf was used, while for the determination of the neutron flux at the target position, reference foils of Al were placed at the front and back of the Hf target. The irradiation was continuous for 28 hours leading to a total neutron fluence of $10^{10}-10^{11}$ n/cm$^{2}$ and a BF3 detector was used for monitoring the neutron flux during the irradiation. After the end of the irradiation, the activity of the Hf target and the Al foils were measured off-line by two HPGe detectors. The $^{176}$Hf$(n,2n)^{175}$Hf reaction has been corrected from the contribution of the $^{177}$Hf$(n,3n)^{175}$Hf and $^{174}$Hf$(n,γ)^{175}$Hf reactions.
Statistical model calculations based on the Hauser-Feshbach theory have also been performed using the code EMPIRE 3.3.2. The predictions have been compared with the data of the present work as well as with data from literature.

Speaker: Rosa Vlastou
• 10:18
Transmission measurements of thick $^{\rm nat}$Fe targets at nELBE 12m

Significant shortcomings in the resolved resonance region of all existing evaluations of iron isotopes have been identified from leakage neutron measurements to be related to inaccurate elastic cross sections minima between 50 and 700 keV [1,2]. Those regions are only sensitive to highly accurate transmission measurements through very thick samples. The overestimation of the measured yield in leakage neutron measurements reach 30-40% around 300 keV. Similar disagreements have been identified from 50 up to 700 keV. Therefore, new transmission measurements are needed to improve the description of the measured minima of the total cross section. The total cross section in this energy range is dominated by elastic scattering, as the capture cross section above 70 keV is only a correction of the order 10$^{-3}$. We have measured the neutron transmission of 30, 50 and 90 mm thick iron samples with high statistical accuracy at the nELBE neutron time-of-flight setup [3] at Helmholtz-Zentrum Dresden-Rossendorf in 16.5 days of beam time. The neutron transmission was measured using a plastic scintillator in the energy range above 100 keV with a short flight path of 868 cm. The new transmission and total cross section data of iron can be used to improve the elastic scattering cross section minima in the INDEN-III (structural materials) evaluation of IAEA [4]. This project has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 847594 (ARIEL).

1 B.Jansky et al., JEFFDOC-1918, NEA (2018)
[2] M. Schulc et al., App. Rad. and Isotopes 130 (2017) 224
[3] A.R. Junghans et al., EPJ Web of Conferences 239 (2020) 01006
[4] International Nuclear Data Evaluation Network on the Evaluated Nuclear Data of the Structural Materials, IAEA Vienna, 2018 INDEN-III
INDEN-III webpage

Speaker: Roland Beyer
• 10:30
Measurement of Photoneutron Yields Using the RPI LINAC and Assessment of Evaluated Photoneutron Data for Tantalum and Beryllium 12m

A new experiment configuration was designed, developed and implemented to measure photoneutron yields using the Rensselaer Polytechnic Institute (RPI) electron linear accelerator (LINAC) at the RPI Gaerttner LINAC Center. The new experiment configuration includes a new target assembly that converts the LINAC electron beam into a high energy bremsstrahlung photon flux incident upon a sample material of interest. The photons excite nuclei in the sample of interest, which can subsequently emit neutrons (photoneutrons). The photoneutrons emitted in the direction of the detector system (90 degree orientation from the initial electron beam) travel through a series of collimated vacuum pipes before reaching a pair of proton-recoil (EJ-301) high energy neutron detectors. The signals generated by the neutron detectors are processed using a digital data acquisition system and subsequently analyzed to determine the energy-dependent photoneutron yield from the sample of interest.

The new experiment configuration was used to perform proof-of-concept experiments to measure the photoneutron yields from samples of tantalum and beryllium. The measured results were then compared against the results from Monte Carlo simulations of the detailed experiment configurations to perform preliminary assessments of evaluated photoneutron data libraries. For example, comparing the measured photoneutron spectrum from a tantalum sample with a Monte Carlo simulation using the ENDF/B-8.0 evaluation showed overall good agreement in the energy range from 0.5-20 MeV; however, differences observed in the range of 5-9 MeV warrant further investigation. Lessons learned and potential upgrades were identified and documented to support improving the photoneutron measurement capability for future experiments and allow more rigorous validation assessments of evaluated photoneutron data libraries.

Speaker: Brian Epping
• 10:42
Thermal cross section measurements at the RPI LINAC 12m

Recently, a cold moderator was designed and developed for use at the RPI LINAC. This cold moderator proved to easily and safely couple to an existing neutron producing target, while enhancing neutron flux below 0.02 eV by cooling polyethylene down to 29 K. This cold moderator capability allowed for significantly improved counting statistics below 0.02 eV not previously possible due to a poor signal to background ratio. Additionally, testing was performed to characterize the energy resolution of the new cold moderator system and found the system easily capable of resolving resonances in Ta-181 at 4 and 10 eV, while also clearly resolving the Bragg edges found in Be metal below 0.01 eV. Following the design and development of a cold polyethylene moderator, a series of thermal total cross section measurements were performed for polyethylene, polystyrene, Plexiglas and yttrium hydride in the thermal region. These measurements serve to help validate thermal scattering law (TSL) evaluations in the 0.0006 – 20 eV energy range. For polyethylene and polystyrene, two sets of experiments were performed – one with the Enhanced Thermal Target (ETT) and another with the ETT plus the new cold moderator capability (ETTC). The yttrium hydride and Plexiglas measurements were only performed with the ETTC. The measurements for polyethylene help to validate the data processing methodology when using the ETTC, while extending the measured range of polyethylene down to 0.0007 eV. Two different Plexiglas, Plexiglas G and Plexiglas G-UVT, and two different concentrations of yttrium hydride, H/Y = 1.85 and 1.68, were measured. Overall, all materials had generally good agreement with their ENDF/B-VIII.0 TSL evaluations, though some discrepancies were noticed. In the case of the yttrium hydride, the high energy oscillations in the hydrogen cross section and the low energy Bragg edges in the yttrium cross section were clearly seen. These measurements represent the first total cross section measurements that encompass the entire thermal region from 0.0006 – 3 eV for polystyrene and yttrium hydride.

Speaker: Dominik Fritz
• 09:30 11:06
Nuclear Astrophysics Capitol ()

### Capitol

Convener: Lawrence Heilbronn (UTK)
• 09:30
High precision re-determination of the half-lives of $^{146}$Sm, $^{148}$Gd, and $^{154}$Dy. 24m

Radioisotopes with half-lives (t$_{1/2}$) in the order of millions of years play an essential role in the description of galactic events, as well as in the nuclear dating of extraterrestrial samples. Among the different benchmark radionuclides for astrophysics and geophysics studies, the radio-lanthanides $^{146}$Sm, $^{148}$Gd and $^{154}$Dy are of especial interest. In fact, through the pure α-decay sequence $^{154}$Dy → $^{150}$Gd → $^{146}$Sm → $^{142}$Nd, these radionuclides directly influence both the abundance of $^{146}$Sm, as well as the natural isotopic composition of stable Nd, thus contributing to the Sm/Nd chronometer. The latter is one of the most powerful tools to date silicate differentiation and primitive crust formation events occurred during the first hundreds of million years (My) of the Solar System [1, 2]. Furthermore, photodisintegration processes on $^{148}$Gd, the rate of the $^{144}$Sm(α,γ)$^{148}$Gd reaction, and competing nuclear reactions like $^{154}$Dy(α, γ)$^{158}$Er, $^{154}$Dy(γ, α)$^{150}$Gd, and $^{154}$Dy(γ, n)$^{152}$Dy [3-5], just to mention a few, are essential for a better understanding of the nucleosynthetic processes that led to the present isotopic abundances in the lanthanide region. It is apparent that an accurate knowledge on the nuclear properties of the above-mentioned radionuclides is crucial. However, the currently available data for the half-life of $^{146}$Sm, $^{148}$Gd, and $^{154}$Dy are inconsistent, or affected by uncertainties even up to 50% [6-8]. Recently, the re-evaluated half-life data of a considerable number of radionuclides showed a significant disagreement with previous values, as in the case of $^{146}$Sm [9]. The shorter measured half-life for this radio-lanthanide implies a higher abundance of $^{146}$Sm in the early Solar System, and thus, planetary events dated with the $^{146}$Sm-$^{142}$Nd chronometer converge now to a shorter time span than previously estimated. Reasons for these imprecise nuclear data lie in the difficulty of obtaining samples of the isotopes of interest in sufficient amounts and purity, together with inherent problematics in performing such demanding measurements. In this work, which belongs to the initiative “ERAWAST - Exotic Radionuclides from Accelerator Waste for Science and Technology” [10], we could obtain sufficient amounts of $^{146}$Sm, $^{148}$Gd, and $^{154}$Dy by reprocessing irradiated Ta materials available at the PSI accelerator-facilities. The re-determination of the half-lives of the above-mentioned radio-lanthanides proceeded by successively applying the “direct” or the “decay” method, depending on the order of magnitude of the expected t$_{1/2}$. A scheme for both methods is presented in the Figure.
Here, in the framework of the ERAWAST project, preliminary results on the measured half-lives of $^{148}$Gd and $^{154}$Dy, together with the first steps towards determining the decay constant of $^{146}$Sm, will be presented.

Figure: Process followed for the determination of half-lives below 150 years (a) and the determination of half-lives above 150 years (b).

Acknowledgement
This project is funded by the Swiss National Science Foundation (SNF grant no 200021-159738).

References
1 Bouvier, A., and M. Boyet. Nature 537, no. 7620 (2016): 399-402 [2] Harper, C. L., and Stein B. Jacobsen. Nature 360, no. 6406 (1992): 728-732 [3] Scholz, P., H. Wilsenach, H. W. Becker, A. Blazhev, F. Heim, V. Foteinou, U. Giesen et al. Physical Review C 102, no. 4 (2020): 045811. [4] Rauscher, T. Physical Review C 73, no. 1 (2006): 015804. [5] Woosley, S. E., and W. M. Howard. The Astrophysical Journal 354 (1990): L21-L24. [6] Khazov, Yu, A. Rodionov, and G. Shulyak. Nuclear Data Sheets 136 (2016): 163-452. [7] Nica, N. Nuclear Data Sheets 117 (2014): 1-229. [8] Reich, C. W. Nuclear Data Sheets 110, no. 10 (2009): 2257-2532. [9] Kinoshita, N., M. Paul, Y. Kashiv, P. Collon, C. M. Deibel, B. DiGiovine, J. P. Greene et al. Science 335, no. 6076 (2012): 1614-1617. [10] Schumann, D., and J. Neuhausen. Journal of Physics G: Nuclear and Particle Physics 35, no. 1 (2007): 014046.

Speaker: Nadine Mariel Chiera
• 09:54
Neutron capture cross-section measurements of Mn-53 12m

Short-lived radionuclides, i.e., radioactive isotopes with half-lives less than 100 Ma, were present in the proto-solar cloud and during the early phases of the formation of our Solar system. The origin of individual short-lived radionuclides is still under debate. Due to the comparatively short half-lives, these isotopes are nowadays not present in cosmic samples, but are recognizable as enhancements of their decay products e.g. in samples of meteorites.
A remarkable case is $^{53}$Mn, which is expected to be one of the most abundant short-lived radioisotopes present in our Galaxy. Sahijpal modelled the general galactic chemical evolution of the stellar cluster surrounding our Solar system within galactic timescales [1]. It can be efficiently produced and released into the interstellar medium during supernovae explosions and thus be able to reach our Solar system. The analysis of deep sea manganese crust samples reveal the presence of layers with enhanced $^{53}$Mn concentrations pointing to the precipitation after close-by supernovae explosions [2]. In addition, it was shown that $^{53}$Mn is continuously deposited on Earth by analyzing 500 kg snow sample from Antarctica [3]
Different to other short lived isotopes, $^{53}$Mn can also be formed in dust that originates from asteroid collisions and comets via spallogenic reactions. The estimation of the amounts arriving on Earth and its relation to the originally produced quantity in the supernovae event are still the subject of intensive discussions. Secondary particle reactions are one of the essential components in this debate. However, the dominating nuclear reactions in the dust leading to $^{53}$Mn are proton and neutron induced reactions on iron. In such an environment, one has to consider also the follow-up reactions of $^{53}$Mn with these particles, which could be one of the sources for the observed reduced $^{53}$Mn content. In any case, the synthesized $^{53}$Mn must pass through regions of high neutron densities and therefore undergoes further nuclear reactions that modulate the total content of expelled $^{53}$Mn. One of the possible reactions causing such an effect could be neutron capture.
Due to the rarity of $^{53}$Mn on Earth – it only occurs in usable quantities in meteorites – the measurement of nuclear properties is challenging. Therefore, only the thermal neutron capture cross-section was determined so far using samples containing about 10$^{13}$ atoms of $^{53}$Mn. In the course of the ERAWAST (Exotic Radionuclides from Accelerator Waste for Science and Technology) initiative it was possible to gain a stock of about 10$^{19}$ atoms $^{53}$Mn from proton activated materials at the ring cyclotron at PSI. Parts of this stock were used to fabricate samples to measure neutron capture cross-sections of $^{53}$Mn at different neutron facilities. These samples were used to measure the neutron capture cross-section in a wide range of neutron energies starting from very cold neutron till stellar neutrons utilizing different neutron facilities [4]. Figure 1 shows an overall plot of all obtained results together with an adopted TALYS calculation and the normalized used neutron spectra.
In the case of using cold and thermal neutrons, the results are in good agreement with each other as well as with reported values of the thermal capture cross-section obtained 50 years ago, but with an order of magnitude reduced uncertainties. In addition, due to the direct determination of the number of atoms in the samples, these values do not depend of the half-life of $^{53}$Mn. The resonance integrals and the capture cross-section at very cold and stellar neutron energies were measured for the first time.

Acknowledgment:
We would like to thank T. Jenke, St. Roccia and U. Köster at the Institut Laue-Langevin (France) for the support conducting experiments at the very cold neutron beam line PF2; N. Kneip, D. Studer, T. Kieck, and K. Wendt of the Johannes Gutenberg-Universität Mainz (Germany) for preparing samples using the RISIKO off-line laser mass separation facility for the experiments at the PF2 beam line; L. Viererbl ,M. Vinš, H. Assmann-Vratislavská of the Research Centre Řež (Czech Republic) for conduction the thermal and epithermal neutron activations at the research reactor LRV-15; O. Aviv, A. Barak, Y. Buzaglo, H. Dafna, B. Kaizer, D. Kijel, A. Kreisel, M. Tessler, L. Weissman, Z. Yungrais of the Soreq National Research Center (Israel) for providing the proton beam at the LiLiT Neutron Source at SARAF and support and participating in the experiment at stellar neutron energy; M. Paul and E. Peretz of Racah Institute of Physics at the Hebrew University Jerusalem (Israel) for participating and co-analysing the experiment at stellar neutron energy; as well as the following colleagues of the Paul Scherrer Institut: M. Ayranov and T. Wieseler for elaborating the chemical extraction methods and performing the purification to create the stock of $^{53}$Mn; P. Sprung for determining the total $^{53}$Mn quantities of the used samples via ICP-MS measurements and A. Kaestner for support to perform experiments at the cold neutron beam line ICON at the neutron source SINQ at PSI.

[1] S. Sahijpal, J. Astrophys. Astron. 35 (2014) 121
[2] G.Korschinek, et al.: Phys. Rev. Lett.125 (2020) 031101
[3] D. Koll, et al.: Phys. Rev. Lett. 123 (2019) 072701
[4] J. Ulrich, PhD thesis 2020, High precision nuclear data of $^{53}$Mn for astrophysics and geosciences, University of Berne, Switzerland

Speaker: Dr Rugard Dressler (Paul Scherrer Institute)
• 10:06
$^{77,78}$Se(n,$\gamma$) and $^{68}$Zn(n,$\gamma$) Cross-section Measurements at n\_TOF Relevant for the Astrophysical $s$-process 12m

Around half of the elements heavier than iron are produced via the slow neutron capture process ($s$-process), whereby a seed nucleus undergoes a series of neutron captures and beta decays, with the timescales of the captures being slower than those of beta decays. The $s$-process is further subdivided into several different components which occur in different types of stars. Of particular interest to this presentation is the weak $s$-process component, which occurs in massive ($>$10M$_\odot$) stars during He-core and C-shell burning, and is responsible for producing elements between mass numbers 60 and 90.

A recent study by Nishimura $et$ $al$. [1] showed that $^{77}$Se(n,$\gamma$), $^{78}$Se(n,$\gamma$) and $^{68}$Zn(n,$\gamma$) reactions have a key impact on the prediction of the $^{77}$Se and $^{78}$Se abundances, respectively, considering the current large uncertainties ($>$10% [2]) in their stellar cross sections.

Recently, $^{77}$Se, $^{78}$Se and $^{68}$Zn neutron capture cross sections have been measured at the Neutron Time-of-Flight (n_TOF) Facility at CERN, chosen for its wide spectrum of neutron energies relevant to stellar environments and high neutron energy resolution. The campaign was conducted using C$_6$D$_6$ gamma-ray detectors at the experimental area at the end of the 185-meter measurement station. The details of the experiments and subsequent analyses and results will be presented.

[1] N. Nishimura et al., Uncertainties in s-process nucleosynthesis in massive stars de-
termined by Monte Carlo variations, MNRAS 469, Issue 2, 2017
[2] I. Dillman et al., The Karlsruhe Astrophysical Database of Nucleosynthesis in Stars
Project – Status and Prospects, Nuclear Data Sheets 120, pp. 171-174, 2014

Speaker: Nikolay Sosnin (University of Edinburgh)
• 10:18
Measurement of the 140Ce(n,γ) cross section at astrophysical energies at n_TOF 12m

It is well ascertained since the late '50, that the vast majority of the elements above the iron peak are synthesized in stars, via sequences of neutron captures and β-decays. Among the nucleosynthesis mechanisms, the slow (s-)process represents one of the better known and is very effective into modelling the stellar evolution. Thanks to the high accuracy of the nuclear data available in the major libraries, the s-process models can predict the final element abundances from iron to lead with high precision.
An accurate measurement of the 140Ce(n,γ) energy dependent cross section has been performed at the n_TOF facility at CERN. This measurement was motivated by the significant discrepancy in the cerium abundance observed in the globular cluster M22 and the value predicted by theoretical stellar models [1].

The experimental apparatus was based on a highly enriched 140Ce sample and on C6D6 liquid scintillators to detect the γ produced following the neutron capture events. The experimental data up to 65 keV allowed us to resolve 81 neutron resonances, which provides the largest contribution to the MACS at the temperature of interest for the s-process. The n_TOF data show large discrepancies compared to the cross section reported in the major nuclear data libraries, as shown in Fig.1 (Resonance Shape Analysis of a p-wave resonance: n_TOF data presents large disceprancies compared to both ENDF/B-VIII and JENDL-4.0 libraries.).

The robustness of the n_TOF results is ensured by a good agreement with the MACS measured by activation by Käppeler et al. [2] at kT = 30 keV. At lower temperatures, high discrepancy with respect to the values provided by KADoNiS has been observed. The experimental results of the 140Ce(n,γ) cross section and the astrophysical impact of the new MACS on the s-process modelling in the Ba-Ce-Nd region will be presented.

[1] O. Straniero, S. Cristallo, L. Piersanti. Heavy elements in globular clusters: the role of asymptotic giant branch stars, ApJ 2014, 785, 77.
[2] F. Käppeler et al. Neutron capture cross sections of the cerium isotopes for s- and p-process studies, Phys. Rev. C 1996, 53, 1397.

Speaker: Simone Amaducci (Laboratori Nazionali del Sud (IT))
• 10:30
(WITHDRAWN) New experimental decay data for nuclei around $^{78}$Ni and its astrophysical impact 12m

How elements are made in the Universe is an open long-standing question. Several processes are invoked to explain the observed elemental abundances in our Solar System [1] and in our Galaxy [2].
Complex simulations of astrophysical events are used to study the origin of the heavy elements and quantify the contribution of the r-process to the observed elemental abundances [3]. This process proceeds through very neutron rich nuclei, most of them experimentally inaccessible so far and theoretical estimates supply the lack of experimental data, introducing large uncertainties in the calculated abundances [4]. The main decay mode of such nuclei is beta-decay accompanied by the emission of one or more neutrons. Accurate experimental data of these nuclei, particularly their half-lives $\mathrm{T}_\mathrm{1/2}$ and neutron emission probabilities $\mathrm{P}_\mathrm{xn}$, will contribute to reduce the uncertainties of the calculated elemental abundances and allow a critical comparison with several nuclear structure models.
The BRIKEN (Beta delayed neutron measurements at RIKEN) collaboration was launched to provide precise new measurements of $\mathrm{T}_\mathrm{1/2}$ and $\mathrm{P}_\mathrm{xn}$ of very neutron rich nuclei which are important for r-process nucleosynthesis [5]. We will report new experimental values from RIBF127 experimental campaign for 39 nuclei between $^\mathrm{75}$Co and $^\mathrm{94}$Br, for which 12 $\mathrm{P}_\mathrm{xn}$ and 7 $\mathrm{T}_\mathrm{1/2}$ are measured for first time and the remaining ones improved. The analysis methods [6] were refined and systematic errors carefully considered in order to improve the accuracy of results. The astrophysical impact in several scenarios was explored and the most relevant cases will be presented.

Speakers: Dr Alvaro Tolosa-Delgado (University of Jyvaskyla, Finland), BRIKEN collaboration BRIKEN collaboration (BRIKEN collaboration)
• 09:30 11:06
Reactor Data: IV Placerville ()

### Placerville

Convener: Grégoire Kessedjian (CEA)
• 09:30
Photo-Disintegration of Deuterium: from Evaluations to Applications 24m

Among the photo-nuclear reactions of light nuclides, the photo-disintegration of deuterium, $^2{\mathrm H}(\gamma,{\mathrm n})^1{\mathrm H}$, is arguably the most important one for nuclear systems with heavy/light water. In multiplying/critical systems, this reaction is also a source of delayed neutrons due to the delayed gamma emission of certain fission products. Therefore, it is important to review the modern evaluations of photo-nuclear reaction data of deuterium, their processability from the ENDF format into application specific ones (e.g., ACE), and from the standpoint of Monte Carlo neutron and gamma transport applications, the sampling of out-going neutron energy and direction by modern Monte Carlo codes. The necessary application of relativistic kinematics vs. non-relativistic limits in sampling of neutrons and protons will be discussed in detail. In addition, a comparison with the $^2{\mathrm H}({\mathrm n},2{\mathrm n})^1{\mathrm H}$ reaction will be made for completeness of the physics background of deuterium disintegration in the coupled neutron and gamma flux.

In studying the time-dependent phenomena in critical systems, a standard tool in nuclear engineering is the application of the point kinetic equations. Some Monte Carlo codes can be used to analyze time-dependent phenomena in neutronics without the assumption that the system of interest is a point'' in space. While feasible in theory, coupled neutron-gamma transport with the rigorous inclusion of photo-nuclear reactions is still practically beyond the reach of time-dependent Monte Carlo methods. However, similar to the point kinetics treatment of systems with heavy water/beryllium, extra delayed neutron groups can be added to the existing delayed group data of actinides (6 or 8 delayed groups), but the extra data blocks have to be prepared in the ENDF format. As a result, the treatment of time-dependent phenomena in heavy water systems with a time-dependent Monte Carlo method (and using nuclear data libraries in ACE format) will be possible, and time scale constrained by the delayed photo-neutrons can be reached. The problem of adding extra delayed neutron groups to the nuclear data files of actinides (e.g., $^{235}$U and $^{238}$U in ENDF/B-VIII.0 and JEFF-3.3) and post-processing/testing of the results will be discussed in detail. In addition, some interesting examples of modelling steady-state and time-dependent phenomena in the ZED-2 (heavy water) reactor at the Chalk River Laboratories (Canada) will be discussed from the standpoint of coupled neutron gamma transport and photo-nuclear data applications.

Speaker: Dr Danila Roubtsov (Canadian Nuclear Laboratories (CNL))
• 09:54
Resonance Region Evaluation of O16 for Criticality Safety and Reactor Applications 12m

Recent experimental data measurements, namely transmission and (n, alpha), have motivated revising the O16 resonance region evaluation. Issues with the normalization of the (n, alpha) cross sections have been investigated and a very dependable measurement has been carried out at the GELINA facility at the Joint Research Center Geel from the energy threshold (2.354 MeV) to 9 MeV. Additionally, transmission measurements were done at the nELBE time of flight facility at Helmholtz-Zentrum Dresden - Rossendorf (HZDR). Experimental data used in previous evaluation were also considered in the evaluation.
The resonance evaluation was performed in the energy range from 0 eV to 6 MeV using the computer code SAMMY resulting in a set of resonance parameters (RPs) that describes well the experimental data used in the evaluation. The recent transmission measurements and the (n, alpha) cross section data are well reproduced in the RPs. The RPs were converted to the evaluated nuclear data file (ENDF) format using the R-Matrix Limited format option LRF=7. The intent of the full paper is to describe the procedures used in the evaluation of the RPs and the use of the RPs in calculations of critical benchmark experiments. Preliminary results for Pu-SOL-THERM-041 (PST-041) configurations listed in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) handbook indicates an improvement on the average keff values with calculated keff being consistent with the benchmark keff, discrepancies on keff not exceeding the combined effect of experimental and Monte Carlo uncertainties. This series of 40 experiments performed at Valduc research center involves solutions of low 240Pu content (3 %) plutonium nitrate with a concentration in plutonium varying between 20 g/L and 190 g/L, in a 500 mm/200 mm annular tank. In particular, for the PST-041 series the changes in keff values indicate that these benchmarks are sensitive to the (n, alpha) cross section of 16O and that the new measurements were essential to improve the benchmark results.

Speaker: Luiz Leal (IRSN)
• 10:06
From $^{232}Th$ $(n,n’\gamma)$ cross sections to level production and total neutron inelastic scattering cross sections. 12m

For the improvement of Generation IV nuclear reactors simulation, accurate evaluated data are required. Among others, a reaction of interest is the neutron inelastic scattering as it modifies the energy distribution of neutrons. In the context of the innovated Th/U cycle and for a better knowledge of the neutron inelastic scattering on thorium 232, experiments have been performed at EC-JRC/GELINA.
Using the prompt gamma ray spectroscopy and the GRAPhEME device to detect the emitted photons, 70 $(n,n'\gamma)$ transition cross sections have been measured . From these cross sections and nuclear structure data, such as gamma intensities or internal conversion coefficients (ICC), one can determine inelastic level production cross sections and the total neutron inelastic scattering cross section. Assumptions had to be made regarding the multipolarity of several gamma transitions as they were not experimentally known. They enabled to exploit more data and obtain more level production cross sections.
The obtained cross sections, presented in this poster, have shown that conversion electrons are a relevant parameter. The misreading of this parameter and its associated uncertainty have a large impact on the total neutron inelastic scattering cross section uncertainty. There is also a need to probe deeper the thorium 232 nuclear structure in order to get more accurate inelastic level production cross sections.
Comparisons made with the nuclear reaction code TALYS, using default parameters and improved input data (TENDL files), have revealed that inelastic level production cross sections are not well predicted by the code. Studies are ongoing to improve the TALYS predictions.

Speaker: Nicolas Dari Bako (IPHC/CNRS)
• 10:18
Neutrons propagation in Lead: a feasibility study for experiments in the RSV TAPIRO fast research reactor 12m

The increasing interest worldwide in Lead-cooled Fast Reactors (LFRs) substantiates the need to validate the analytical codes and methods used to support their design. For neutronic analyses, this is chiefly reflected in assessing the impact of nuclear data uncertainties on the integral and local parameters resulting from such analyses. Besides this driving interest, the aim of refining nuclear data moves continuous efforts for more accurate measurements, be them differential or integral, for which adequate facilities are required.
A preliminary attempt at adjusting the ENDF/B-VIII.0 neutron data libraries for application to ALFRED – the demonstration reactor of future European LFRs – provided clear indications on the need to refine the cross-sections of Lead isotopes, and particularly of the elastic and inelastic ones at high energies. However, experiments dealing specifically with Lead's neutronics are not frequent in literature and even scarce in international databases establishing benchmark cases for the validation of neutronic codes or the adjustment of nuclear data libraries. Therefore, the availability at the ENEA's Casaccia research center of a fast source reactor – RSV TAPIRO – provides a unique opportunity to perform new integral experiments in support of fast reactors, including LFRs, owing to the well-characterized neutron spectrum of the thermal column, a spacious experimental slot within the biological shield of the reactor in the proximity of the RSV TAPIRO's external copper reflector.
Accordingly, a series of experiments has been envisaged, dealing with the use of Lead in a reactor, with the possibility for some of those experiments to become benchmark cases of international relevance. The experiments concern the propagation of neutrons through blocks of materials representing relevant elements of a reactor core, and ranging from pure Lead to mixtures reproducing portions of the reflector and shield in LFRs.The paper will be focused on the feasibility study of some of these experiments, where Leadand mixture blockswill be inserted in the so-called thermal column of the RSV TAPIRO reactor and irradiated by the neutron flux emerging from the copper reflector surrounding the reactor core. Despite the name of the experimental opening, the spectrum of the incident flux is relative hard, owing to the almost-pure core fission spectrum.
The scope of the feasibility study will include the conceptual design of the blocks, moving from representative studies with target LFRs. An in-depth investigation will be also performed concerning the required instrumentation: appropriate Neutron detectors and threshold activation foilswill be selectedaiming at characterizing the assembly's neutron flux intensity and energy spectrum, so thatthe measured reaction rates and spectral indexes could be used to assess the impact of nuclear data on these parameters, retrieving information for ad-hoc adjustments applied to LFRs.

Speaker: Mrs Valentina Fabrizio
• 10:30
Simultaneous evaluation of uranium and plutonium fast neutron fission cross sections up to 200 MeV for JENDL-5 12m

The fast neutron fission cross sections of 233,235,238U and 239,240,241Pu were evaluated for the JENDL-5 library up to 200 MeV. The experimental fission cross sections and their ratios in the EXFOR library were reviewed with the source articles. Additionally, Poenitz’s data compiled in his GMA database were reviewed. We found about 160 datasets are archived with the uncertainty information sufficient for covariance matrix construction and converted them from EXFOR to an experimental database with their covariance matrices. When the uncertainty information in the source article is missing in the EXFOR entry, we updated the EXFOR entry. We minimized corrections to the experimental database to make our evaluation becomes traceable.

The least-squares fitting was performed to the logarithms of the cross sections and their rations in the experimental database by using the simultaneous least-squares fitting code SOK. The best precisions of the group-wise cross sections were achieved around 2 to 3 MeV, where the external uncertainties are 1% (235,238U), 1.5% (233U, 239Pu) or 2.5% (241Pu). Well documented experimental fission cross sections are desired for 235U between 100 and 300 keV and 241Pu in the whole energy range since the numbers of the usable experimental datasets are limited and it makes the uncertainties in our evaluation relatively high.

The evaluated cross sections were also validated against the spectrum averaged fission cross sections measured under the 252Cf and Σ-Σ neutron fields, and we found the newly evaluated cross sections are consistent with the experimental spectrum averaged cross sections well except for 238U fission cross sections measured in the 252Cf neutron fields, which is underestimated by the newly evaluated cross sections as well as those in many other data libraries.

Speakers: Dr Naohiko Otsuka (International Atomic Energy Agency), Dr Osamu Iwamoto (Japan Atomic Energy Agency)
• 10:42
Theoretical calculation and evaluation for n+239,240,241,242,242m,243,244Am reactions 12m

In order to reduce the uncertainties in the design and operation of accelerator-driven systems (ADSs), high-precision nuclear data for neutron- and proton-induced reactions on a variety of isotopes in the energy range below 200 MeV are necessary. Knowledge of accurate neutron-induced fission cross sections is crucially important for the design of various reactor systems. Therefore, accurate nuclear data for neutron-induced reactions on Am isotopes are needed in the calculation of neutron and energy balance and the prediction of transmutation rates of the various radioactive species. To meet this requirement, all cross sections, angular distributions, energy spectra, double differential cross sections of neutron, proton, deuteron, triton, helium-3 and alpha emissions and the number of neutron per fission for n+239,240,241,242,242m,243,244Am reactions are consistently calculated and analyzed by theoretical nuclear models in the energy range of En≤200 MeV. The corresponding models include the optical model, the unified Hauser-Feshbach theory and the exciton model, the improved Iwamoto-Harada model, the evaporation models, the linear angular momentum dependent exciton state density model, the fission model, the intranuclear cascade model, the distorted-wave Born approximation and the coupled channel theory. The calculated results reproduce the experimental data well, and the variation tendency of reaction cross sections related to the target mass numbers is obtained.

Speaker: Yongli Xu
• 10:54
Measurement of the delayed-neutron yield and group parameters in the thermal neutron induced fission of 239Pu 12m

Delayed neutron data is essential in inherent reactor safety and in reactor control since it is used to estimate the reactivity. The quantities of interest related to delayed neutrons are the average delayed neutron yield, and the group constants that define the kinetics of delayed neutron emission.
Discrepancies among the evaluated data as well as missing covariance in most of international databases result in excessive conservatism in the safety margins. In order to improve this status, a collaboration among CEA and CNRS was settled to produce high-quality delayed neutron data for several fissioning systems of interest in GEN-II & III reactors.
Following successful campaigns at ILL (Institut Laue-Langevin) in 2018 and 2019, devoted to 235U, a new campaign was realized in march 2021, focusing on the thermal fission of 239Pu. The experiment consists of a cold neutron beam hitting a fissile target, placed at the center of a long-counter made of a polyethylene matrix in which 16 helium-3 tubes are introduced. After a defined irradiation length, during which fissions occur and precursors buid up, the beam is quickly interrupted and the delayed neutrons emitted by the precursors' is detected. First estimations of the delayed neutron yield and group constants are proposed and discussed in this paper, in comparison with data from the literature.

Speaker: Pierre Leconte
• 11:06 11:15
Break 9m
• 11:15 12:45
Facilities: III Sutter's Fort ()

### Sutter's Fort

Convener: Yaron Danon (RPI)
• 11:15
Improvement of targetry methods for nuclear data measurements 24m

According to the Nuclear Data High Priority Request List published by the Nuclear Energy Agency (http://www.oecd-nea.org/dbdata/hprl/), values of cross sections used in nuclear technology should be known with uncertainties between 1% and 5%, while cross sections of key isotopes involved in s- and r-processes must be known with uncertainties of about 1%.
The uncertainty associated with the measured cross sections has two sources, data evaluation and experimental procedure. Experimental uncertainties are mainly generated by time resolution of experimental configurations (duration and speed of neutron burst, etc.), detectors and electronics [1], and by the targetry (defined as discipline of design, manufacture and characterization of targets) [2]. On the other end several measurements of important cross sections have been hindered by the lack of adequate amount of the relative target nuclide with the requested high chemically and isotopically purity.
The first part of this contribution focuses on the production and purification of some exotic radionuclides at the Paul Scherrer Institut, Switzerland. Some examples, to illustrate the importance of the target chemical purity for obtaining the envisaged measurement, are presented.
The main part of the talk centers on recent developments of novel methods for the production of homogeneous and uniform solid targets allowing obtaining cross sections with lower uncertainty. In fact, only if the target material is homogeneous and has uniform thickness, the incident particles will have the same probability of interaction with the target nuclei and the energy loss of all the detected particles, especially if charged, will be similar. This will allow measurements that are more precise.
New findings on the role of the solvent on Molecular Plating, one of the most used techniques for solid target production [3], are presented. These results show that less volatile solvent allow obtaining more uniform and thinner targets. Figure 1 illustrates three scanning electron microscopy imagines (x 600 magnification) of three different targets of holmium obtained with identical conditions in term of temperature and concentration, but using solvent with different vapor pressure, i.e. isopropanol (5.78 kPa), Isobutanol (1.53 kPa) and N,N-Dimethylformamide (0.44 kPa). It is possible to see that the material deposited from solvents with higher vapor pressure are characterized by a less homogeneous and island-like structure, with presence of cracks and peeling off the deposited material.
A still unsolved problem related with the Molecular Plating is the inevitable co-precipitation of organic compounds, which causes thickening and introduction of impurities into the deposited layer, enhancing the background during the cross section measurement.
The development of a new method, based on electrodeposition of exotic radionuclides from ionic liquid, is under development in our laboratories to deposit chemically pure elements avoiding any co-precipitation. Some preliminary results of this project are presented too.
Reference
[1] H. Postma, P. Schillebeeckx, Neutron Resonance Capture and Transmission Analysis, Encyclopedia of Analytical Chemistry, John Wiley & Sons, Ltd, 2006.
[2] P. Schillebeeckx, A. Borella, J.C. Drohe, R. Eykens, S. Kopecky, C. Massimi, L.C. Mihailescu, A. Moens, M. Moxon, R. Wynants, Target requirements for neutron-induced cross-section measurements in the resonance region, Nucl. Instr. Meth. Phys. Res. A, 613 (2010) 378.
[3] W. Parker, R. Falk, Molecular plating: A method for the electrolytic formation of thin inorganic films, Nucl. Instr. Meth., 16 (1962) 355.

Speaker: Emilio Andrea Maugeri (Paul Scherrer Institut)
• 11:39
A next-generation neutron detector with interaction localization capabilities and its application to beta-delayed neutron spectroscopy 12m

The new generation of radioactive ion beam facilities will allow the production of unstable nuclei far away from the valley of stability. In the very neutron-rich region, the beta-decay energy window is much larger than the neutron separation energy of the daughter nucleus. Thus, delayed neutron emission after beta decay becomes a prevalent decay mode. To obtain information about the nuclear structure of these nuclei, precise spectroscopic measurements of beta-delayed neutrons are essential. The Neutron dEtector with Xn Tracking (NEXT) array has been designed to measure neutron energies via the time-of-flight technique with improved precision [1]. A single NEXT module is composed of a highly segmented plastic scintillator bar and position-sensitive photomultipliers. This design enables the localization of the neutron interaction point, reducing the uncertainties associated with neutron time-of-flight measurement without sacrificing the neutron detection efficiency.
The NEXT prototyping phase has already been completed, and the construction of 40 NEXT modules is ongoing. The array will be fully implemented by fall 2022. We will present NEXT construction status along with key results from characterization measurements. Details on the readout electronics will also be covered.
[1] J. Heideman et al., Nucl. Instrum. Methods Phys. Res., Sect. A 946, 162528 (2019).

Speaker: Noritaka Kitamura (University of Tennessee, Knoxville)
• 11:51
Response of nuclear physics detectors to laser-driven neutrons at the PW DRACO laser facility 12m

Pulsed neutron beams are a valuable tool in nuclear physics, but their applications are strongly restricted by the limited number of neutron sources available worldwide [1]. These neutron beams, suitable for "time of flight" (TOF) experiments, can be characterized by their time resolution, the intensity per pulse, and the frequency or repetition rate.

In terms of neutron production, laser-driven ion sources are garnering the interest of the nuclear physics community due to the fast development of ultra-short (fs) and ultra-high power (> $10^{19}$ W/cm$^2$) lasers and their use as compact particle accelerators [2]. Neutrons produced using this alternative method can reach higher instantaneous neutron fluxes than conventional neutron sources [3-4]. Nevertheless, the potential use of these neutrons in nuclear physics applications relies on the behavior of nuclear physics detectors, which is still not characterized for the peculiarities of the harsh environment produced in a laser facility.

This work presents the results of a multiple detectors’ experimental campaign performed at the PW-class DRACO laser system at HZDR [5] during fall 2021. During this campaign, different detectors employed in conventional nuclear physics experiments were tested under a laser-driven neutron source. This set included lithium detectors for thermal and epithermal neutron detection, organic liquid and solid detectors for fast neutrons, diamond and silicon detectors for charged particles, and a neutron camera to study the possibilities of neutron imaging.

The experiment was performed under different neutron production configurations and also with and without moderation of the neutron flux. This systematic analysis based on the response of the different diagnostics allowed us to set an optimized setup to perform a dedicated time-of-flight measurement. The purpose of this experiment is to establish the viability of carrying out nuclear physics experiments in this type of neutron sources and to identify the advantages and disadvantages of this production method as opposed to conventional systems.

[1] INTERNATIONAL ATOMIC ENERGY AGENCY, Development Opportunities for Small and Medium Scale Accelerator Driven Neutron Sources – Report of a technical meeting held in Vienna, 18–21 May 2004, IAEA-TECDOC-1439, IAEA, Vienna (2005).
[2] Alejo A., Ahmed H., Green A. et al. Recent advances in laser-driven neutron sources. Il nuovo cimento 38 C, 188 (2015). 10.1393/ncc/i2015-15188-8
[3] Roth M., Jung D., Falk K. et al. Bright Laser-Driven Neutron Source Based on the Relativistic Transparency of Solids. Phys. Rev. Lett. 110, 044802 (2013). 10.1103/PhysRevLett.110.044802
[4] Guerrero C., Domingo-Pardo C., Käppeler F. et al. Prospects for direct neutron capture measurements on s-process branching point isotopes. Eur. Phys. J. A 53, 87 (2017). 10.1140/epja/i2017-12261-2
[5] Schramm U., Bussmann M., Irman A. et al. First results with the novel petawatt laser acceleration facility in Dresden. J. Phys.: Conf. Ser. 874, 012028 (2017). 10.1088/1742-6596/874/1/012028

Speaker: Ms María de los Ángeles Millán Callado (Centro Nacional de Aceleradores (CNA, US - Junta de Andalucía - CSIC) - Universidad de Sevilla (US), Dpt. Física Atómica, Molecular y Nuclear (FAMN))
• 12:03
Experimental evaluation of energy resolutions for pulsed neutron beam in the KURNS-LINAC 12m

Energy resolutions of pulsed neutrons are one of the important parameters for nuclear data measurement by TOF method. The energy resolutions are a facility-specific parameters, and has been evaluated experimentally and numerically at various TOF facilities using pulsed neutrons.
KURNS-LINAC is an L-band electron linear accelerator with maximum acceleration voltage of 46 MeV established at Institute for Integrated Radiation and Nuclear Science, Kyoto University. By using a target according to a research purpose, we can use various types of particle beam sources, i.e. neutron, electron, and photon. In addition, the pulsed width of the electron beam can be easily changed. For a pulsed neutron source, a water-cooled tantalum (Ta) target as a photo-neutron target and various moderator are used. A typical neutron flight tube used for nuclear data measurement at KURRI-LINAC is installed in the direction of 135 degree to the LINAC electron beam line, and the neutron flight path from the Ta target to a sample is about 12.7 m.
In this study, experimentally evaluations for the energy resolution of pulsed neutron flux in the neutron path were carried out. The capture gamma-rays from a Ta-181 sample were measured by a BGO detector with a TOF method and the TOF spectra for well-known resonances were obtained. The energy resolution was evaluated by comparing the full width at half maximum of the Ta-181 resonances in the JENDL-4.0. In order to obtain relationships between the energy resolution and the pulsed neutron beam width, the measurements were carried out with the pulsed neutron beam width of 100 nsec, 1 μsec and 4 μsec, respectively. As the experimental results, the energy resolution of neutron energy range from 4 eV to 100 eV corresponding to each pulse width were evaluated. For example, the energy resolution at 4.28 eV (Ta-181 first resonance peak) was about 1.0 % for a pulse width of 4 μsec.

Speaker: Mr Yasunori Matsuo (Kindai University)
• 12:15
miniBELEN: a modular neutron long counter for (α,n) reactions 12m

Abstract
Neutron production through α-induced nuclear reactions is a key issue in several fields. Specifically, (α,n) reactions are interesting in nuclear astrophysics as a source of neutrons for the slow neutron capture nucleosynthesis (the s-process) [TAI16] and in the α-particles capture process (the α-process) [WOO92, BLI17]. Other fields of interest include the neutron-induced background in underground laboratories [BET10], which is a crucial issue in low counting rate experiments, and in nuclear facilities such as nuclear reactors and particle accelerators [MUR02]. Currently, evaluated data is available only for a limited number of isotopes and the databases present large discrepancies in some cases.

In this work we present the miniBELEN detector, a new modular and transportable neutron counter for (α,n) reactions. This detector has been developed in the context of a Spanish effort for the Measurement of Alpha Neutron Yields and spectra (MANY collaboration). The miniBELEN detector is based on the use of several long 3He-filled proportional neutron counters embedded in a modular high density polyethylene moderator. In order to provide the detector with a response independent of the neutron energy, namely a flat response, an innovative design methodology has been applied. The method is based on the optimization of the contribution of each counter to the total efficiency by using thermal neutron absorbers. This allows to obtain flat responses up to 10 MeV. The experimental characterization of miniBELEN using 252Cf neutron sources and results of the commissioning with the measurement of the 27Al(α,n)30P reaction at Spanish research facilities will be presented.

Acknowledgments
This work has been supported by the Spanish Ministerio de Economía y Competitividad under grants FPA2017-83946-C2-1 & C2-2 and PID2019-104714GB-C21 & C22.

Bibliography
[BET10] A. Bettini, Nucl. Instrum. Methods A 626 - 627 (2010) S64 - S68.
[BLI17] J. Bliss et al. J. Phys. G.: Nucl. Part. Phys. 44 (2017) 054003
[MUR02] T. Murata and K. Shibata, J. Nucl. Sci. Technol. 39 (2002) 76 - 79
[TAI16] J.L. Tain et al., J. Phys.: Conf. Ser. 665 (2016) 012031.
[WOO92] S.E. Woosley and R.D. Hoffmann, Astrophys. J. 395 (1992) 202 - 239

Speaker: Nil Mont i Geli (Universitat Politècnica de Catalunya)
• 11:15 12:45
Fission: V American River ()

### American River

Convener: Lucas Snyder (LLNL)
• 11:15
Measurements of fission product masses and isotopic yields from $^{252}$Cf spontaneous fission at the FRS Ion Catcher 24m

Measurements of independent isotopic fission yields (IIFYs) provide access to the underlying probability distribution of products resulting from fission. The knowledge of IIFY distributions contributes to our understanding of the nuclear fission process in more depth with respect to mass yield distributions. A better understating of the fission process has wide implications, including the abundance of elements through nucleosynthesis, nuclear structure and reactions, and nuclear waste management and safety [1].

We will present the first results of a novel method for measuring IIFYs of 252Cf spontaneous fission (SF) via direct mass measurements [2], at the FRS Ion Catcher (FRS-IC) at GSI, Germany [3]. Fission products were generated from a 252Cf source that was installed inside the cryogenic stopping cell (CSC) [4], and were identified and counted with the multiple-reflection time-of-flight mass spectrometer (MR-TOF-MS) [5] of the FRS-IC, utilizing well-established measurement and data analysis methods [6]. The high performance of the MR-TOF-MS can resolve isobars unambiguously, even with limited statistics. The non-scanning nature of the MR-TOF-MS ensures minimal relative systematic uncertainties in IIFY amongst fission products.

Our high-accuracy mass results constitute by themselves new important data, as they include first (or first direct) mass measurements at the N≈90 and Z=56-62 region. We compare our results to existing data of indirect measurements and to the Atomic Mass Evaluation 2020 (AME2020) [7].
The analysis for extracting IIFYs includes isotope-dependent efficiency corrections for all components of the FRS-IC. In particular, we applied a self-consistent technique that takes into account the element- dependent survival efficiencies in the CSC, due to chemical reactions with the buffer gas.

Our IFY results, which cover several tens of fission products in the less-accessible high-mass peak down to fission yields at the level of 10-5, are generally similar to those of the NuDat 2 database [8]. Nevertheless, they reveal some structures that are not observed in the database smooth trends.
These are the first results of a planned campaign to investigate IIFY distributions of spontaneous fission at the FRS-IC. Upcoming experiments will extend our results to wider Z and N ranges, lower fission yields, and other spontaneously-fissioning actinides.

References:
[1] Dimitriou, P., Hambsch, F.-J., & Pomp, S. (2016). Fission Product Yields Data: Current status and perspectives Summary Report of an IAEA Technical Meeting (INDC(NDS)--0713). International Atomic Energy Agency (IAEA)
[2] Mardor et al., EPJ Web of Conferences 239, 02004 (2020)
[3] W. R. Plass et al., Nucl. Inst. Meth. B 317, 457 (2013)
[4] M. Ranjan et al., Nucl. Inst. Meth. A 770, 87 (2015)
[5] T. Dickel et al., Nucl. Inst. Meth. A 777, 172 (2015)
[6] S. Ayet San Andres et al., Phys. Rev. C 99, 064313 (2019)
[7] M. Wang et al., Chinese Phys. C 45 030003 (2021)
[8] National Nuclear Data Center, information extracted from the NuDat 2 database, https://www.nndc.bnl.gov/nudat2/

Speaker: Yuval Waschitz
• 11:39
Energy Dependent Fission Product Yields from Neutron Induced Fission 12m

Fission product yields (FPY) are essential ingredients for addressing questions relevant to a range of basic and applied physics. Examples include the cosmic nucleosynthesis processes that created the elements from iron to uranium, decay heat release in nuclear reactors, reactor neutrino studies, radioisotope production, development of advanced reactor and transmutation systems, and many national security applications. While new applications will require accurate energy-dependent FPY data over a broad set of incident neutron energies, the current evaluated FPY data files contain only three energy points: thermal, fast, and 14-MeV incident energies. The goal of this study is to provide high-precision and energy dependent FPY data using monoenergetic neutron beams with energies between 0.5 and 15 MeV.

Absolute cumulative fission product yields have been determined for about 100 fission products representing 40 mass chains during neutron-induced fission of 235U, 238U, and 239Pu. Using rapid belt-driven irradiated target transfer system (RABITTS) and irradiations with varying duration, gamma-ray decay history of fission products between 1 second to a few days have been measured. The number of fissions during the irradiation times was determined via a dual fission ionization chamber loaded with thin electroplated foils with the same actinide material. The obtained new FPY data provides a complete picture of the fission product yield landscape; from the initial distribution produced directly by fission, through the complex, time-dependent evolution of the yields from beta-decay and neutron emission. This work also provides a unique capability to bridge short-lived fission product yields to our measured long-lived chain fission yields [1]. An overview of the recent experimental results will be presented.

This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under contract DE-AC52-07NA27344.

[1] M. Gooden et al., Nucl. Data Sheets 131, 319 (2016).

Speaker: Anton Tonchev
• 11:51
Measuring the $^{235}$U(n,f)/$^6$Li(n,t) cross section ratio in the NIFFTE fissionTPC 12m

While nuclear data play an important role in nuclear physics applications, it has become important to have a better understanding and try to minimize their uncertainties. In particular, there is a need for precision neutron-induced fission cross section measurements on fissile nuclei. Neutron-induced fission cross sections are typically measured as ratios, with a well-known standard in the denominator. While the $^{235}$U(n,f) standard is well measured, some light particle reactions are also well-known and their use as reference can provide information to remove shared systematic uncertainties that are present in an actinide-only ratio. The NIFFTE collaboration’s fission time projection chamber (fissionTPC) is a 2×2π charged particle tracker designed for measuring neutron-induced fission. Detailed 3D track reconstruction of the reaction products enables evaluation of systematic effects and corresponding uncertainties which are less directly accessible by other measurement techniques. This work focuses on the recent measurement of the $^{235}$U(n,f) using as a reference the standard $^6$Li(n,t) reaction. Preliminary data of the $^{235}$U(n,f)/$^6$Li(n,t) measurement deployed at the Los Alamos Neutron Science Center will be presented.

LLNL-ABS-800573. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

Speakers: Maria Anastasiou, the NIFFTE Collaboration
• 12:03
Spectral averaged cross sections as a probe to a high energy tail of $^{235}$U PFNS 12m

The systematic evaluations of spectrum averaged cross sections of dosimetric reactions over a broad range of energies were performed in 252Cf (spontaneous fission) and 235U(nth,f) neutron fields. The neutron sources used in this study were LR-0, VR-1 zero power research light water reactors, LVR-15 10 MW research light water reactor, and 252Cf high emission isotopic source with emission specified precisely by the manganese sulphate bath. All spectral averaged cross sections were inferred from measured reaction rates. Reaction rates were derived from gamma spectrometry using a high purity germanium detector with a validated computational efficiency determination. These gamma spectrometry measurements were performed using single detector in all cases. The ratios of 252Cf and 235U spectral averaged cross sections can be used to specify the high energy tail of the 235U prompt fission neutron spectrum as 252Cf spontaneous fission spectrum is considered as a standard. Dosimetric cross sections were validated in 252Cf neutron field. Furthermore, ratios are independent on cross section uncertainties since uncertainties in the cross sections are eliminated. Theoretical models of fission can be tested based on our ratios measurements. The calculations were performed in MCNP6.2 transport code using different prompt fission neutron spectra and IRDFF-II cross sections for threshold reactions. The ratios are in good agreement using only ENDF/B-VIII.0 235U prompt fission neutron spectrum suggesting harder 235U(nth,f) prompt fission neutron spectrum than in other evaluations.

Speaker: Martin Schulc
• 11:15 12:45
Formats: Evaluation Methodology II Delta King ()

### Delta King

Convener: Dorothea Wiarda (ORNL)
• 11:15
Updates to the evaluation of neutron induced reactions on $^{56}$Fe target 24m

Iron is an important structural and shielding material that appears in many applications. Nuclear reaction data of iron have been addressed within the Subgroup-40 (CIELO) of the OECD/NEA Data Bank [1]. The resulting evaluated data files from the BNL/IAEA collaboration for the iron isotopes [2] were included in the ENDF/B-VIII.0 library [3] and performed well in criticality benchmarks. Unfortunately, inadequate performance of the $^{56}$Fe evaluation was discovered for shielding benchmarks just before the release of the ENDF/B-VIII.0 library: the leakage spectra from very thick iron shells with a $^{252}$Cf source in the centre were significantly under-predicted in the energy range 1-8 MeV.
Additional analysis was performed which showed that the root cause of the problem was the non-elastic cross section, which seemed to be too high. This hypothesis was in contradiction with the new measurement of the inelastic cross section by Beyer et al [4] but was partially confirmed by a new measurement of the elastic cross section [5]. A compromise between the measured elastic and the inelastic cross section was sought such that unitarity with the total cross section was preserved. In addition, the elastic cross section was increased in the resonance interference minima (being the largest increase around 300 keV). The increase was guided by the measured leakage spectra from thick iron shells with a $^{252}$Cf source in the centre to eliminate the observed overestimation by 30% of the measured neutron leakage around 300 keV. Some of the improvements were already presented orally at various meetings, but no formal report or publication has been released.
There was also a question of the thermal capture cross section. R. Firestone [6] published the results from prompt-gamma activation analysis, which suggested lowering of the thermal capture cross section by 7%. Validation measurements at CEA within the MAESTRO project [7] did not support such a decrease [8]. The same validation showed that JEFF evaluation was essentially correct, so the recommended value of the thermal capture cross section for $^{56}$Fe of 2.577 barns was adopted. This value is also in agreement with latest Firestone evaluation [8].
The resulting evaluated data files were extensively tested. The previously observed deficiency of under-predicting the fast-neutron leakage spectra from thick iron shells with a $^{252}$Cf source was avoided. The 30% overestimation of the neutron leakage around 300 keV was also eliminated. Similar benchmark experiments with a D-T source also showed an improvement. In a benchmark involving 40 MeV neutrons the new evaluation performed much better than the ENDF/B-VII.1 library. At the same time, performance in criticality benchmarks was not compromised.
The benchmarking exercise illustrates how integral benchmarks can be used to discriminate between discrepant differential data to produce evaluated data files that perform well in situations that are critical for the safe operation of nuclear installations. Details of the changes to the differential data will be given and selected representative benchmark results will be presented.

References
[1] M.B. Chadwick et al.: “CIELO Collaboration Summary Results: International Evaluations of Neutron Reactions on Uranium, Plutonium, Iron, Oxygen and Hydrogen”, Nuclear Data Sheets 148 (2018) 189–213.
[2] M.W. Herman et al: “Evaluation of Neutron Reactions on Iron Isotopes for CIELO and ENDF/B-VIII.0”, Nuclear Data Sheets 148 (2018) 214–253.
[3] D.A. Brown et al.: “ENDF/B-VIII.0: The 8th Major Release of the Nuclear Reaction Data Library with CIELO-project Cross Sections, New Standards and Thermal Scattering Data”, Nuclear Data Sheets 148 (2018) 1–142.
[4] R. Beyer et al.: “Inelastic scattering of fast neutrons from 56Fe”, EPJ Web of Conferences 146, 02017 (2017).
[5] E. Pirovano et al.: “Cross section and neutron angular distribution measurements of neutron scattering on natural iron”, Physical Review C 99, 024601 (2019)
[6] R.B. Firestone et al.: “Thermal neutron capture cross section for 56Fe(n,γ )”, Physical Review C 95, 014328 (2017).
[7] P. Leconte, “Nuclear data feedback on structural, moderating and absorbing materials through the MAESTRO experimental programme in MINERVE,” Tech. Rep. JEF/DOC-1849, Nuclear Energy Agency, OECD, Paris, France (2017).
[8] R.B. Firestone, private communication, to be published (2020).

Speaker: Prof. Andrej Trkov (Jozef Stefan Institute)
• 11:39
Bayesian network evaluation of neutron-induced reactions of Fe-56 12m

We present the progress made in the evaluation of neutron-induced reactions of Fe-56 using a Bayesian network. The Bayesian network framework facilitates the incorporation of a diverse set of measurements, such as residual production cross sections and angular distributions. A nuclear physics model code in combination with Gaussian processes to take into account model imperfections is used in the fast energy range. In the energy range between about one and five MeV, we use a flexible Gaussian process construction instead of a physics model. The Bayesian network framework enables us to evaluate these two energy ranges simultaneously with a smooth transition between them and sum rule constraints between channels being respected.

Speaker: Georg Schnabel (IAEA)
• 11:51
Limited confidence and overall consistency concepts in nuclear data evaluation 12m

LA-UR-21-29369

We use an example of Ta181 evaluation to illustrate modern evaluation methodology which, while based on the classic three pillars of the evaluation (differential experimental data, nuclear reaction modeling, and validation with integral experiments), makes predominant use of the "limited confidence" and "overall consistency" concepts. The limited confidence applies to all three pillars. Differential data have limited precision and in some cases might be misleading. Reaction modeling might use inappropriate models, be affected by model defects, and leaves pretty wide leeway to parameter adjustment. The integral experiments, because of the fact of being "integral", may lead to the compensation of errors and, similarly to the differential measurements, might be misleading. We will show how in this confusing realm requiring the overall consistency, and obeying physics laws leads to ignoring a number of differential experiments and results in a clean, model based evaluation reproducing wide range of differential observables and improving performance in integral testing.

Tantalum is nearly mono-isotopic metal with Ta181 being 99.988% of the
natural mixture. Therefore, there is an exceptional wealth of experimental data, including cross sections (also to isomeric states), angular distributions, as well as spectra and double-differential cross sections. This abundance allows, in several cases, to eliminate datasets that are not consistent with the remaining bulk of experimental data. We will show that this elimination goes beyond simple rejection of outliers.

The modeling was performed with EMPIRE-3.2.3 reaction model code. In view of multiple choices for selecting models available in EMPIRE various combinations of those were first explored in attempt to reproduce vast set of existing experimental data, while refraining from adjusting individual model parameters. This way the following models were adopted:

• Coupled-Channels with dispersive Optical Model potential.
• Quantum-mechanical TUL Multistep Direct (MSD) model for pre-equilibrium emission of neutrons.
• Heidelberg formulation of the Multistep Compound (MSC) model for pre-equilibrium neutron, proton and gamma emission.
• Exciton model with Iwamoto-Harada extension for pre-equilibrium cluster emission.
• Hauser-Feshbach with width-fluctuations correction, full gamma-cascade, and Gilbert-Cameron level densities.

Combination of MSD and MSC models provides very good description of the neutron spectra, which is also essential for correct description of channels such as (n,2n) and (n,p). Gilbert-Cameron level densities were chosen over EGSM and microscopic Hartree-Fock-Bogoliubov because the former produce better capture cross sections between 1-3 MeV, and slightly better gamma spectra.

Speaker: Mike Herman (LANL)
• 12:03
Bayesian Monte Carlo Evaluation Framework for Imperfect Data and Models 12m
Speaker: Jesse Brown (ORNL)
• 11:15 12:45
Measurements: V Folsom ()

### Folsom

Convener: Hye Young Lee (LANL)
• 11:15
Indirect measurements of neutron-induced reaction cross-sections at storage rings 24m

Obtaining reliable cross sections for neutron-induced reactions on unstable nuclei is a highly important task and a major challenge. These data are essential for understanding the synthesis of heavy elements in stars and for applications in nuclear technology. However, their measurement is very complicated as both projectile and target are radioactive. The NECTAR (NuclEar reaCTions At storage Rings) project aims to circumvent these problems by using the surrogate reaction method in inverse kinematics, where the nucleus formed in the neutron-induced reaction of interest is produced by a reaction (typically a transfer or an inelastic-scattering reaction) involving a radioactive heavy-ion beam and a stable, light target nucleus. The probabilities as a function of the compound-nucleus excitation energy for gamma-ray emission, neutron emission and fission, which can be measured with the surrogate reaction, are particularly useful to constrain model parameters and to inform more accurate predictions of neutron-induced reaction cross sections [1].

Yet, the full development of the surrogate method is hampered by the numerous long-standing target issues. The objective of the NECTAR project is to solve these issues by combining surrogate reactions with the unique and largely unexplored possibilities at heavy-ion storage rings. In a storage ring heavy radioactive ions revolve at high frequency passing repeatedly through an electron cooler, which will greatly improve the beam quality and restore it after each passage of the beam through the internal gas-jet serving as ultra-thin, windowless target. This way, excitation energy and decay probabilities can be measured with unrivaled accuracy.

In this contribution, we will present the conceptual idea of the setup, which will be developed within NECTAR to measure for the first time simultaneously the fission, neutron and gamma-ray emission probabilities at the storage rings of the GSI/FAIR facility. We will also discuss the developments that are being carried out towards these measurements. In particular, we will present the first results of a proof of principle experiment, which will be conducted in June 2022 at the ESR storage ring of GSI/FAIR.

[1] R. Pérez Sánchez, B. Jurado et al., Phys. Rev. Lett. 125 (2020) 122502

Acknowledgement: This work has received funding from the European Research Council (ERC) under the European Union’s Horizon 2020 research and innovation programme (ERC-Advanced grant NECTAR, grant agreement No 884715).

• 11:39
Time-of-flight measurements of MINERVE samples containing fission products and neutron absorbing isotopes 12m

The zero-power reactor MINERVE (CEA Cadarache) was designed to perform reactivity worth measurements by the oscillation technique. The various experimental programs, undertaken for the last thirty years, involved cylindrical samples with a diameter of about 1 cm and a height ranging from a few cm to 10 cm. Most of the samples are composed of UO2 pellets with a given fission product, actinide or neutron absorbing isotope in a double-sealed Zry-4 container.

An experimental program started in 2015 in collaboration with the Joint Research Centre of Geel to study the MINERVE samples at the time-of-flight facility GELINA by the neutron transmission technique. The two main objectives consist of checking both the composition of the MINERVE samples provided by the manufacturer and the quality of the resonance parameters recommended in the evaluated neutron data library JEFF-3.3 [1]. The pioneer experiments on MINERVE samples containing 107Ag and 109Ag revealed a substantial Tungsten contamination that was not reported by the manufacturer [2]. Such a Tungsten contamination is related to the manufacturing process of the sample pellets by powder compacting. The observed Tungsten contaminations lead to non-negligible increases of the C/E ratios up to a few percent. A second experimental campaign on MINERVE samples containing 99Tc provided useful insight on the quality of the 99Tc resonance parameters measured at the GELINA facility at the end of the 90s [3].

The ongoing program continuing through 2022 will deliver data for samarium (natural, 147, 149, 152), neodymium (natural,143,145), gadolinium (natural, 155), europium (151, 153), rhodium (103), Cesium (133), hafnium (180), dysprosium (160, 161, 162, 163, 164) and erbium (168, 170). The present work will focus on the data analysis technique developed for long cylindrical samples with a diameter smaller than the neutron beam, and on the grain size distribution model implemented in the resonance shape analysis REFIT and CONRAD.

References
[1] A. J. M. Plompen et al, The joint evaluated fission and fusion nuclear data library JEFF-3.3, Eur. Phys. J. A (2020) 56:181
[2] L. Salamon et al., Neutron resonance transmission analysis of cylindrical samples used for reactivity worth measurements, J. Radioanal Nucl. Chem. https://doi.org/10.1007/s10967-019-06611-9 (2019)
[3] G. Noguere et al., Average neutron cross sections of Tc, Phys. Rev. C 102, 015807 (2020)

Speaker: Gilles Noguere
• 11:51
Measurement of the $^{235}$U(n,f) cross section relative to the $^{10}$B(n,α) reaction with Micromegas detectors at the CERN n_TOF facility: first results 12m

Neutron induced reaction cross section measurements are often performed relative to a neutron cross-section standard. Thus, the accuracy of the neutron standards determines the accuracy of the neutron cross section measurements. The $^{235}$U(n,f) cross section is widely used as reference reaction, while it is considered as standard at 0.0253 eV (thermal energy), between 7.8 and 11 eV and from 0.15 to 200 MeV [1].

Moreover, it is a very important reaction for neutronic calculations of nuclear reactors and has been the subject of many experimental and theoretical works. Nevertheless, certain issues with the cross section of this reaction have been pointed out, especially in the energy region below 100 keV (e.g. [2] - [4] etc).

For this reason, high accuracy and high resolution cross section data for the $^{235}$U(n,f) reaction are needed to improve the accuracy of this reaction cross section and to extend the energy range of the Resolved Resonances Region above 2.5 keV.
In this context, the $^{235}$U(n,f) reaction cross has been measured relative to the standard $^{10}$B(n,α) cross section within the n_TOF collaboration, with an independent experimental setup from the previous measurement of Amaducci et al. [4] in n_TOF. The measurement was carried out in experimental area EAR-1 of the n_TOF facility at CERN, with the aim to cover the widest energy range possible. The high purity targets were produced at JRC-Geel in Belgium, while the experimental setup was based on Micromegas detectors.
The preliminary results from this work will be presented and discussed.

[1] A.D.Carlson et al., Nucl. Data Sheets 148 (2018) 143
[2] M. Barbagallo et al., Eur. Phys. J. A 49 (2013) 156
[3] M. Jandel et al., Phys. Rev. Lett. 109 (2012) 202506
[4] S. Amaducci et al., Eur. Phys. J. A 55 (2019) 120

Speakers: Veatriki Michalopoulou, the n_TOF Collaboration
• 12:03
Design and Simulation of a Next-Generation Dual n-gamma Detector Array at Los Alamos National Laboratory 12m

Measurements of neutron elastic scattering and angular distributions remain one of the largest uncertainties in simulations of fission-driven nuclear systems. Current experimental techniques are limited by issues arising from measuring only neutrons or gammas or, in the case where both particles are measured, the relatively small number of angles current dual n-gamma detectors are capable of covering. Next-generation elpasolite detectors offer near-perfect n-gamma pulse shape discrimination across a wide array of energies and the availability of these detectors in sufficient quantities make the construction of a large, highly-segmented detector array a possibility for the first time. Preliminary studies have isolated CLYC detectors as the ideal candidate for use in measuring neutron elastic scattering. We describe the mechanical design and simulation of the Correlated gamma-neutron Array for sCattering (CoGNAC) utilizing these detectors currently under development at Los Alamos National Laboratory. Additionally, we discuss the results of an initial experimental campaign on $^9$Be, $^{27}$Al, and $^{56}$Fe utilizing a partially completed CoGNAC in conjunction with Chi-Nu’s existing array of 54 liquid scintillator detectors.

Speaker: Eames Bennett
• 12:15
Measurement of the $^{233}$U(n,$\gamma$) cross section at LANSCE 12m

The $^{233}$U plays an important role in the Th-U fuel cycle, that has been proposed as an alternative to the U-Pu fuel cycle due to its reduced amount of transuranium elements. The available experimental $^{233}$U(n,$\gamma$) cross section data in the literature are scarce and were measured decades ago [1-3]. In 2008, the $^{233}$U(n,$\gamma$) cross section was investigated at LANL using the DANCE detector combined with a PPAC, however the statistics in the keV regime were inadequate for a reliable extraction of the cross section at 100 keV. An accurate measurement of the $^{233}$U(n,$\gamma$) cross section is required by the NCSP (National Critically Safety Program) to complete the neutron-induced cross section data, and also, as reported by the ORNL, a new evaluation with revised (renormalized) fission cross section is needed on $^{233}$U.

Speaker: Esther Leal Cidoncha
• 11:15 12:45
Measurements: XIII (capture) Capitol ()

### Capitol

• 11:15
First precise measurement of the U-235 neutron capture cross-section at thermal and sub-thermal neutron energies 24m

The recommended cross-section value for $^{235}$U neutron-capture at thermal energies is largely based on the difference from total and competing cross-sections of $^{235}$U. Despite its importance as a thermal neutron constant and its high value (100 barn), direct capture measurements are rare (only two exist for thermal energies) and exhibit large uncertainties. The reason is the difficulty to measure the characteristic radiation of the reaction product $^{236}$U within a dominant fission background and with $^{236}$U having a long half-life of 23.4 Myr. Importantly, capture of $^{235}$U may exhibit a significant deviation from a pure 1/v-behaviour around and below thermal energies.

Here, we present highly accurate capture data by applying a multi-isotope spike method to minimize systematic effects: we used the combination of neutron activation and subsequent accelerator-mass-spectrometry (AMS) for direct atom counting of the reaction product $^{236}$U. By utilizing different neutron fields we mapped the energy dependence of the capture cross section in six activations from thermal to ultra-cold energies: with an almost pure Maxwellian spectrum at room temperature at BR1 (Mol, Belgium), with cold neutrons at MLZ (Munich, Germany) and ILL (Grenoble, France) as well as with very cold neutrons also at ILL. AMS was performed at the VERA laboratory in Vienna. Using AMS, the capture cross-section is deduced simply by the measured isotope ratio $^{236}$/$^{235}$U divided by the neutron fluence.

By irradiating Uranium samples of natural U composition, we measured also $^{239}$Pu, the decay product of $^{239}$U produced from $^{238}$U(n,γ) in the same samples. I.e. we deduced simultaneously the $^{236}$U/$^{239}$Pu ratio, giving directly the cross-section ratio of $^{235}$U(n,γ) relative to $^{238}$U(n,γ). The latter ratio is therefore completely independent of the neutron fluence.

Activation with AMS allowed for an accurate determination of the $^{235}$U(n,γ) and $^{238}$U(n,γ) cross sections with an uncertainty of ~2% across the whole energy range. We will present our new experimental results and will also demonstrate the potential of this method for independent and accurate nuclear reaction cross section measurements.

Speaker: Anton Wallner
• 11:39
Neutron capture and total cross-section measurements on $^{94,95,96}$Mo at n_TOF and GELINA 12m

Cross-sections for neutron-induced interactions with molybdenum, in particular for the neutron capture reaction, play an important role in various fields ranging from nuclear astrophysics to safety assessment of conventional nuclear power plants and the development of innovative technologies. It is found as a pollutant in pre-solar silicon carbide grains and it has a crucial role in stellar nucleosynthesis in Asymptotic Giant Branch (AGB) stars[1]. During reactor operation molybdenum is produced as a fission product. Moreover, the use of molybdenum for the production of Accident Tolerant Fuel (ATF) is under study [2]. It is a promising candidate for new generation research reactors using UMo alloys with Low Enriched Uranium (LEU)[3]. This shows the importance of an accurate knowledge of the total and capture cross-section for molybdenum isotopes.

Experimental data in the literature for the capture cross-section of Mo isotopes suffer from large uncertainties. This is also reflected in the large uncertainties of the cross-sections recommended in the ENDF/B-VIII.0 library[4]. Below 1 eV the relative uncertainty of the capture cross-section is above 18% for $^{94}$Mo and around 40% for $^{96}$Mo, while above 2 keV the uncertainties are in the order of 10-20% for $^{94,95,96}$Mo. One of the reasons for these large uncertainties is related to the absence of transmission data for enriched samples.

In this contribution results of accurate transmission and capture cross-section measurements using natural Mo samples and samples enriched in $^{94}$Mo, $^{95}$Mo and $^{96}$Mo are presented. The data cover an energy region from thermal energy (0.025 eV) up to hundreds of keV. The capture measurements are performed at the n_TOF facility at CERN (CH) and the GELINA facility at JRC-Geel (BE). The transmission measurements are performed at GELINA. The experimental data, i.e. transmissions and capture yields, will be delivered to the EXFOR data library and will be used to improve the evaluated cross-section data for neutron interactions with $^{94,95,96}$Mo in the resolved and unresolved resonance region.

REFERENCES
[1] N.Liu, T.Stephan, S.Cristallo et al., Astrophysical Journal, 881, 28 (2019).
[2] B. Cheng, Y.-J. Kim, P. Chou, Nuclear engineering and Technology, 48, 16-25 (2016).
[3] P. Herve et al., EPJ Nuclear Sciences & Technologies, 4, 49 (2018).
[4] D.A. Brownet et al., Nuclear Data Sheets, 148, 1 (2018).

Speakers: Riccardo Mucciola, the n_TOF Collaboration
• 11:51
Radiative capture cross section measurements of $^{54}$Fe at the RPI LINAC 12m

$^{54}$Fe capture cross section measurements were conducted at the Rensselaer Polytechnic Institute (RPI) Gaerttner Linear Accelerator (LINAC) Center using an enriched $^{54}$Fe sample in the keV energy region. $^{54}$Fe is a constituent of natural iron, which is present in a large variety of nuclear grade materials.Therefore, it is important to have an accurate understanding of the cross section of $^{54}$Fe, which can be measured experimentally. In the time-of-flight measurements previously conducted at the LINAC, an array of four C$_{6}$D$_{6}$ detectors surrounded the sample and data were collected using a digital data acquisition system. In this work, the experimental apparatus was upgraded with additional C$_{6}$D$_{6}$ detectors, allowing for more accurate measurements from 1 keV – 2 MeV while preserving the low neutron sensitivity of the system. The capture yield of the $^{54}$Fe measurements was normalized using the saturated resonances in Au and Ta. The capture yield was subsequently compared to evaluations and prior experimental data. Results were found to have good agreement in certain energy regions. The improved experimental apparatus used in these measurements will greatly enhance RPI’s ability to measure neutron capture cross sections in the keV region.

Speaker: Sukhjinder Singh
• 12:03
Measurement of neutron capture cross section and capture gamma-ray spectrum of Te-128 in keV-neutron energy region 12m

Tellurium isotopes are produced in nuclear reactors as fission products. Evaluation of neutron nuclear data of Tellurium isotopes was made recently [1]. On the other hand, $^{128}$Te is one of the candidate nuclides for neutrino-less double beta decay searches. Neutron-induced reactions of $^{128}$Te are important to evaluate background in measurement [2]. However, experimental data of the neutron capture cross section of $^{128}$Te at neutron energies below 1 MeV are old and have large uncertainties. Thus, in this work, new measurements were carried out to determine the capture cross section of $^{128}$Te using the time-of-flight (TOF) method in the keV energy region. Experiments were performed at the Laboratory for Zero Carbon Energy at the Tokyo Institute of Technology. Incident neutrons were generated through the $^7$Li(p,n)$^7$Be reaction by a pulsed proton beam from a Pelletron accelerator bombarding a lithium target. Experiments were carried out in the two different neutron energy regions: 15 – 100 keV and around 550 keV. The incident neutron energy spectrum was measured by the TOF method. Capture gamma-rays from the sample were detected with an anti-Compton NaI(Tl) spectrometer in the TOF experiments. The capture cross sections were obtained from the pulse height spectra by the pulse-height weighting technique. A comparison of the present results with past experimental data and evaluated cross sections was made. Gamma-ray spectra by unfolding the pulse height spectra with the detector response function were also derived.

References
1. K. Shibata, Journal of Nuclear Science and Technology, 52, 490-502 (2015).
2. W. Tornow et al., EPJ web of Conferences, 146, 09013 (2017).

Speaker: Tatsuya Katabuchi
• 12:15
Results of the $^{244}$Cm, $^{246}$Cm and $^{248}$Cm neutron-induced capture cross sections measurements at EAR1 and EAR2 of the n_TOF facility 12m

V. Alcayne a), A. Kimura b), E. Mendoza a), D. Cano-Ott a), T. Martinez a), E. González-Romero a), A. Pérez de Rada a), and the n_TOF collaboration c)

a) Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), E-28040 Madrid, Spain
b) Japan Atomic Energy Agency (JAEA), Tokai-mura, Japan
c) Conseil Européen pour la Recherche Nucléaire, CERN, Switzerland

The safe and efficient management of the high level waste produced in the operation of nuclear reactors requires more accurate nuclear data. In particular, inventory calculations of the spent nuclear fuel (SNF) and the derived magnitudes such as the decay heat, radio toxicity or neutron and gamma dose, among others, rely on the accuracy of neutron induced reaction cross sections ruling the burn-up in the reactor. The Cm isotopes require a special attention due to their various implications along the fuel cycle. For instance, $^{244}$Cm is responsible of ∼10% of the radio-toxicity and the decay heat in the spent nuclear fuel in a Light Water Reactor (LWR) during the first fifteen years after unloading the SNF from the reactor. Furthermore, the neutron emission in the spent fuel is dominated by the $^{244}$Cm and $^{246}$Cm spontaneous fission during the first 10$^4$ years of disposal. From the point of view of reactor (LWR) and fuel cycle parameters, the sensitivity analyses performed in [1] indicate that uncertainties in the $^{244}$Cm capture cross section need to be reduced to 4.1% between 4 and 22.6 eV and to 14.4% between 22.6 and 454 eV. Last, but not least, a more accurate knowledge of the capture cross sections of $^{244}$Cm, $^{246}$Cm and $^{248}$Cm is required for improving the calculations on the formation of heavier isotopes such as Bk, Cf and other Cm isotopes.

In this work, we will present the final results of the capture cross section measurement on $^{244}$Cm, $^{246}$Cm and $^{248}$Cm performed at the CERN n_TOF facility [2]. It is important to notice that, the Cm samples used in the experiment at n_TOF have been used previously in an experiment at J-PARC [3]. At n_TOF, the capture cross section measurements of $^{244}$Cm, $^{246}$Cm and $^{248}$Cm were performed at the 20 m vertical flight path (EAR2) with three C$_6$D$_6$ total energy detectors [4]. In addition, the cross section of $^{244}$Cm was measured at the 185 m flight path (EAR1) with a Total Absorption Calorimeter (TAC) [5]. The combination of measurements in EAR1 and EAR2 and the use of two complementary experimental techniques, the total absorption calorimetry and the pulse height weighting technique, has contributed to control and reduce the systematic uncertainties in the results.
We will present the radiative kernels of the resonances of $^{244}$Cm, $^{246}$Cm and $^{248}$Cm obtained in the energy ranges from 7 to 300 eV, 4 to 400 eV, and 7 to 100 eV, respectively, and compare the results of the measurement with previous capture and transmission data and evaluated cross sections.

References
[1] G. Aliberti et. al., , Ann. Nucl. Ener. 33, 700 (2006)
[2] C. Guerrero et al. Eur. Phys. J. A 49, 27 (2013)
[3] A. Kimura et. al., Jour. Nucl. Sc. Tech. 49, 708 (2012)
[4] U. Abbondannoet al., Nuc. Inst. Meth. A. 521, 454-467, (2004)
[5] C.Guerrero et al., Nucl. Instrum. Meth. A 608, 424 (2009)

Speakers: Victor Alcayne, the n_TOF Collaboration
• 12:27
Neutron Capture Cross Section Measurement and Resonance Analysis of Pd-107 Using ANNRI at MLF/J-PARC 12m

The long-term accumulation of Long-Lived Fission Products (LLFPs) in nuclear waste has been a significant issue in nuclear industry due to their long half-life. The nuclear transmutation of LLFPs into short-lived or stable nuclides is expected to contribute reducing the current amount of high-level radioactive waste. Highly accurate nuclear data for the neutron-induced nuclear reactions are necessary in order to design LLFPs nuclear transmutation systems.

Palladium-107 (half life: 6.5×10$^{6}$ y) is one of the important LLFPs, and accurate data for the neutron capture cross section are needed for the study on LLFPs transmutation systems. Nevertheless, only a few experiments to measure the neutron capture cross section of Pd-107 have been performed. The neutron energy regions of most of the measurements are limited. A new measurements with a wide neutron energy range from the thermal to keV energies is needed to improve the neutron capture cross section of $^{107}$Pd.

In the present work, the neutron capture cross section measurements were carried out using the Accurate Neutron Nucleus Reaction Measurement Instrument (ANNRI) at the Materials and Life Science Facility (MLF) of the Japan Proton Accelerator Research Complex (J-PARC). A high intensity pulsed neutron beam from Japan Spallation Neutron Source at the MLF using the 3 GeV proton beam was utilized. NaI(Tl) detectors of ANNRI were used for capture measurements. The time-of-flight (TOF) method was employed to determine the incident neutron energy. Two-dimensional data, TOF and pulse-height (PH), were acquired and the data were analyzed based on a PH weighting technique. Resonance parameters were derived from resonance analysis using the REFIT code.

Speaker: Hideto Nakano
• 11:15 12:45
Theory/Codes: I Placerville ()

### Placerville

Convener: Ionel Stetcu (LANL)
• 11:15
Evaluation of charged-particle reactions in the resolved resonance region 24m

Charged-particle-induced reactions at low energies in the resolved resonance region are important for a range of applications in basic and applied sciences. Ion Beam Analysis of materials for cultural heritage, environment and climate control, and forensics, depends on the knowledge of proton-, deuteron- and alpha-induced reactions at energies of a few MeV. Management of fuel in nuclear reactors involves the control of neutrons produced after the reactor operation is shutdown. For the most widely used fuel materials, UO2, UF6, PuF4 and PuO2, the dominant neutron producing reactions are (α,xn) reactions on isotopes of O and F, at incident energies above the neutron emission threshold. In nuclear astrophysics, main stellar processes producing the energy of the stars and leading to the synthesis of the light and medium-mass elements up to the iron nuclei, are fuelled by thermonuclear reactions at temperatures of tens of millions of degrees Kelvin. These conditions simulated in the earth laboratories translate into charged-particle induced reactions on light and medium-mass nuclei at energies of a few hundreds of keV. Alpha-nucleus reactions up to 10 MeV are also a potential source of low-energy neutron background for rare event research in astroparticle physics (dark matter search, neutrino, etc.).
Significant effort has been made over the past decades to measure some of these cross sections. The Ion Beam Analysis Data Library (IBANDL) maintained by the IAEA contains over 6000 datasets of differential and total experimental cross sections for charged-particle-induced reactions in the energy region below several MeV. In nuclear astrophysics, several dedicated compilations (NACRE-I and II) and theoretical reaction rate databases (REACLIB, BRUSLIB, KADONIS, NUCASTRODATA.ORG) have been made available to meet the needs of the stellar nucleosynthesis calculations.
On the other hand, the evaluated nuclear data libraries maintained by national or international coordinated efforts (ENDF, JEFF, JENDL, CENDL) are to date, incomplete as far as charged-particle- induced reactions in the resolved resonance region are concerned.
In response to the above-mentioned gap in the evaluated libraries, the IAEA Nuclear Data Section is coordinating an international effort to (i) verify that the existing R-matrix codes are consistent, (ii) evaluate charged-particle cross sections in the resolved resonance region, (iii) produce evaluated nuclear data files for further processing and finally (iv) disseminate the evaluated data through general purpose evaluated nuclear data libraries.
In this paper we present the results of the effort made thus far on 1) verification of the available R-matrix codes, minimization methods and calculation of covariances, 2) the evaluation of the compound system 7Be*, and 3) improving (alpha,n) reaction data for the applications.
We also discuss the open issues in R-matrix calculations as we extend to higher energies, such as dealing with the rapidly growing number of open channels and merging with the regime of smooth cross sections described by the statistical model. Finally, we summarize the main conclusions of the IAEA Technical meeting on ‘(alpha,n) reaction evaluation and data needs’.

The work of I.J. Thompson was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.
The work of G. Hale and M. Paris was supported by the U.S. Department of Energy through the Los Alamos National Laboratory. Los Alamos National Laboratory is operated by Triad National Security, LLC, for the National Nuclear Security Administration of U.S. Department of Energy (Contract No. 89233218CNA000001). [LA-UR-21-29650]

Speaker: Ian Thompson
• 11:39
Interchange, Extension, and Validation of R-matrix fits for gamma production 12m

The R-matrix method of Lane and Thomas is the standard procedure for modelling resonances at low energies, to determine widths and angular distributions needed for nuclear evaluations. Many different codes have been written, all with different input and output file formats. To ensure that results can be replicated, I have written a python code FERDINAND using FUDGE to read & write ENDF, GNDS standard formats, read AMUR, AZURE, EDA, FRESCO and RAC formats, and write AZURE, EDA, FRESCO, HYRMA and latex formats.

Because of the increasing need for gamma production information, R-matrix codes are being extended to include primary and secondary gammas and predict the angular distributions for both of kinds. More excited states of residual nuclei need to be included in the fits, since these give directly many of the secondary gamma decays. These place computational demands on the codes, and using Google’s Tensorflow framework enables the development of GPU codes such as LLNL’s RFLOW for fitting.

There is also a need to go higher-energy states in the compound nucleus. This will require theory developments to ensure smooth transitions from unresolved and/or overlapping resonances in an R-matrix code to Hauser-Feshbach predictions that are based on transmission coefficients derived from optical potentials. A good connection between these two methods should be possible because the Hauser-Feshbach method may be derived as an approximation to R-matrix resonance descriptions.

Reactions of n+14N and other reactions producing 15N* have been evaluated as part of the INDEN project. The thresholds for n+14N, p+14C and +11B are close to each other, and so all three channels must be fitted simultaneously. ENDF has a 1992 evaluation with these three channels up to 2 MeV, and Notre Dame has recently re-evaluated them up to higher energies. Here I report on assessments that include the higher excited states of 14N and 11B that are responsible for high-energy gamma production.

This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

Speaker: Ian Thompson (LLNL)
• 11:51
Spin assignments of neutron resonances with Machine Learning 12m

The different processes known in astrophysics to govern the successions of nuclear formation and decay, r-process and s-process, depend strongly on nuclear properties such as the intrinsic density of levels of a given nucleus and how they behave when emitting or absorbing particles such as neutrons, protons and/or photons. There are few experimental constraints that we can use to narrow down such properties and most of the information we know comes from experimental measurements of resonance states seen in compound nuclei formed by neutron-induced reactions. These resonances are seen as sharp increases in measured neutron transmission and capture due to the proximity of the neutron energy with excited levels in the compound nucleus.
Therefore, a proper and reliable account of all resonances is crucial for the description of nuclear reactions. We have developed a Machine-Learning method, the Bayesian Resonance Reclassifier, to make use of the statistical properties of resonances to train an algorithm capable to identify missing and misassigned resonances. We are able to train the model on synthetic data and use transfer learning to assess resonance sequences from the Atlas of Neutron Resonances, evaluated files or experimental data. We performed an extensive optimization through multiple grid search of hyper-parameter values for each of the classifiers studied. We were also able to use experimental resonance data obtained through measurements with polarized neutron beams as a validation tool to assess the quality of the method, with encouraging results.

Speaker: Gustavo Nobre (BNL)
• 12:03
A novel R-matrix formalism for three-body channels 12m

Three-body break-up channels play an important role in light nuclear systems even at low incident energies. In standard two-body R-matrix theory these processes can only be treated approximatively for instance by using sequential decay models. We will present a novel three-body R-matrix formalism that is based on a proposal by Glöckle [1]. In this framework the Faddeev equations are solved by dividing the space of Jacobi-coordinates into interior regions with strong local interactions and an asymptotic exterior region. Using the asymptotic form of the three-body wave functions one obtains a system of linear equations which includes the boundary conditions at the division of the Jacobi-coordinates. Its solution directly yields the elastic and the break-up transition amplitudes and thus the elastic and the breakup cross sections. The form of these equations resembles a similar one in the standard R-matrix theory and justifies its denomination.

In a first step we generalized the proposal of Glöckle to three spinless particles with different masses. A first numerical implementation indicated severe shortcomings as well as ill-posedness of the algorithm. Significant changes especially with regard to the conditions and the choice of basis functions as well as the normalization of bound states were required leading to a novel formalism of Glöckle type. The ill-posedness of the equations has been cured by regularization. The novel formalism was first applied to the neutron-deuteron system. Systematic studies of different sets of basis functions and regularization parameters were performed and stable results were achieved. The calculated break-up cross section and the elastic cross section are in fair agreement to the experimental data up to 10 MeV since we include s-waves only. Furthermore, we present an outlook and first results regarding the n+9Be system.

Acknowledgement: The work has been carried out within the framework of the EUROfusion Consortium and has received funding from the European research and training programme 2014-2018 and 2019-2020 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

[1] W. Glöckle, Z. Phys. 271, 31 (1974)

Speaker: Mr Benedikt Raab (TU WIEN)
• Wednesday, 27 July
• 06:00 07:30
Covariances: I Delta King ()

### Delta King

Convener: Oscar Cabellos
• 06:00
(WITHDRAWN) Reactor Antineutrino Energy Spectra and Associated Uncertainties Using the GEF Model 24m

Reactor antineutrino energy spectra are the subject of active experimental researches nowadays, either through large reactor neutrino experiments, short baseline neutrino experiments at research reactors or through dedicated nuclear physics measurements of the properties of the fission products. The summation method relies on the use of nuclear data to build the reactor antineutrino energy spectrum emitted by a reactor, or after the fission of each actinide at the origin of the thermal power in-core[1,2]. It is a unique alternative to converted spectra[3], which rely on measurements of integral beta spectra and a subsequent conversion approach. The summation method presents in addition the advantage of being predictive ; it can be used for the computation of reactor antineutrino spectra associated to innovative fuels, of interest for non-proliferation purposes, or to predict the antineutrino spectrum in wider energy ranges (lower/larger energies) than accessible with the conversion method, of interest for the fundamental study of neutrino properties.
During the last decade, the precision achieved by the summation method to predict reactor antineutrino spectra has been greatly improved by nuclear data measurements using the Total Absorption Gamma-ray Spectroscopy technique (TAGS)[4], reaching a level comparable to the conversion method[1,3]. The main drawback of the summation predictions is the difficulty to propagate the uncertainties of nuclear data on the antineutrino observable, which require covariance matrices for fission yields and for beta decay data. Covariance matrices for fission yields could be obtained consistently with predictions for fission yields from a model such as the GEF model[5].
Recently, we have used the reactor antineutrino observable to improve the GEF model [6], making it a useful tool to predict fission yields for antineutrino spectra calculations, along with their uncertainties. The level of prediction achieved by the antineutrino flux obtained with the GEF fission yields is now close to the one obtained in[4], it is thus possible to use GEF to compute the associated uncertainties.
At this conference, we will present new summation method antineutrino predictions based on fission yields computed with the new version of the GEF code, updated with the latest Total Absorption Gamma-ray Spectroscopy (TAGS) results [6] and, for the first time, with their associated computed uncertainties. Spectra will be displayed for the main uranium and plutonium isotopes contributing to the fissions in a Pressurized Water Reactor, but also for various fuels to be used in innovative reactor designs. The results will be compared with recent measurements of reactor antineutrino flux and energy spectra.

[1] Th. A. Mueller et al., Phys.Rev. C 83 , 054615 (2011)
[2] M. Fallot et al., Phys. Rev. Lett. 109 , 202504 (2012)
[3] P. Huber, Phys. Rev. C 84, 024617 (2011). Also presented in ND2013.
[4] M. Estienne et al., Phys. Rev. Lett. 123, 022502 (2019).
[5] K.-H. Schmidt, B. Jurado, C. Amouroux, C. Schmitt, Nucl. Data Sheets 131 (2016) 107
[6] K.-H. Schmidt, M. Estienne, M. Fallot et al., Nuclear Data Sheets Volume 173, (2021)

Speaker: Magali Estienne (Subatech)
• 06:24
Toward an exact Bayesian inference of fission nuclear data 12m

Evaluating nuclear data reduces the information of a set of experimental data and theoretical models to provide the best estimate of a physical quantity. In the last decade safety applications pushed the evaluation community toward better quantification of the uncertainties associated to evaluated nuclear data. The Bayesian inference appears as a standard and versatile mathematical framework to estimate such uncertainties. As a matter of fact, it is now a widespread tool to evaluate neutron cross sections, especially in the resolved resonance region [1]. On the other hand, the current JEFF evaluations related to the fission cross sections at higher energies as well as the neutron multiplicities and prompt neutron spectra do not often rely on a rigorous application of a Bayesian inference and if so only within some approximated treatments [1,2]. In a long term objective to better ground the determination of uncertainties in evaluations, we explore the feasibility of an exact Bayesian inference of the nuclear data associated to fission.

In this presentation, I will emphasize our preliminary attempts to perform an exact Bayesian inference to evaluate the total neutron induced cross section of an actinide in the continuum energy range as well as on the prompt neutron multiplicity and spectrum. I will emphasize our current results and conclude on the difficulties and challenges of such developments.

[1] C. DE SAINT JEAN, P. ARCHIER, E. PRIVAS, G. NOGUÈRE, B. HABERT, P. TAMAGNO, Nuclear Data Sheets 148, 383-419 (2018)
[2] D. ROCHMAN, E. BAUGE, A. VASILIEV, H. FERROUKHI, S. PELLONI, A.J. KONING, J.Ch. SUBLET, European Physical Journal Plus 133, 537 (2018)

Speaker: David Regnier
• 06:36
Surrogate Modeling for Fission Cross Sections, Criticality Studies, and Uncertainty Quantification 12m

Christian Brazell
10/15/21
TAMU-LLNL
LLNL-ABS-827556
ND2022

Surrogate Modeling for Fission Cross Sections, Criticality Studies, and Uncertainty Quantification
ABSTRACT

Computational methods have advanced to the point where, in many cases, the main source of uncertainty in neutronics calculations comes from the underlying Nuclear Data (ND). Assessing the impact of the underlying uncertainties can be a challenging and computationally expensive task. Yet, it is critical to understand how ND affect the predicted behavior of nuclear systems. This is the goal of Uncertainty Quantification (UQ) and Sensitivity Analysis (SA), research areas of great interest to nuclear engineering researchers and practitioners alike. The present study considers the sensitivities of the effective neutron multiplication factor (keff) to multi-group fission cross sections (XS). Machine Learning (ML) techniques are employed to enable and accelerate UQ research beyond the capabilities of current methods.
Adjoint methods are used to examine the sensitivity of a Quantity of Interest (QoI) to perturbations in a model parameter, such as ND. As in the case of this study, computing the sensitivity of keff to uncertainty in multigroup XS using the adjoint requires only two neutronics calculations. While the adjoint methods are computationally inexpensive, assumptions are made along the way which do not necessarily hold outside of small XS perturbations. Such classical adjoint sensitivities are based on first-order approximations, and cannot capture the true non-linearity of keff’s response to uncertain XS values. This leads to inaccuracies when evaluating the uncertainty of keff over the entire uncertain region of the XS.
This study explores a new approach to UQ and SA using data-driven methods. Advances in ML have shown that regression models are able to capture complex non-linear behavior in systems, exhibiting high accuracy and precision in their predictions. Here, we use ML to train surrogate models that map a realization of a XS to the corresponding keff for a given criticality problem. Once trained, these models can accurately compute keff across the full uncertain distribution of the XS.
Realizations of the uncertainty in the XS are constructed by sampling from the cross section’s multi-group covariance matrix. These samples form a set of inputs which are representative of the uncertainty in the evaluated data. The keff corresponding to each sample is then computed using a neutron transport code. This set of (XS, keff) pairs is then used to train, validate, and test the ML models. Performing this analysis using various critical benchmark experiments, multiple nuclides’ fission cross sections, and several ML models, the present study then compares the performance of the ML models to each other and to the adjoint solution as a baseline.
Initial results showing the performance of the ML methods on a test set can be seen in Figure 1. Here, Gaussian Process (GP) regression, Support Vector Machine (SVM) regression, Multivariate Adaptive Regression Splines (MARS), and Artificial Neural Networks (ANNs) are used as surrogates to map the U-235 fission XS to keff in the Godiva benchmark. The distribution of predicted keff values is compared to the true distribution in the test set, with the difference (residual) between the two being plotted below. Each method shows strong performance in capturing the full relationship between the multigroup XS and the system criticality, with some showing a mean absolute error in keff as low as 2 pcm.

Figure 1. The performance of each regression method for the Godiva problem is shown in the histogram/residual plots above. The models map realizations of the U-235 fission XS to keff.

This study demonstrates the usefulness of ML in nuclear data applications. Computational costs are often prohibitive in UQ/SA tasks, and perturbation methods designed to avoid these costs involve assumptions which do not always hold true. By building surrogate models which capture the nonlinear relationship between nuclear data and applications such as criticality, UQ and other such studies can be enabled with negligible cost. This study also opens the door to a broad range of future work. Directions for further research include exploring the data requirements for good model performance and implementation of the UQ for more complex QoIs, such as reaction rates, multiphysics effects, and transient system behaviors.

Speaker: Christian Brazell (Texas A&M University)
• 06:48
Stochastically Estimated Fission Yield Covariance Matrices 12m

A Monte-Carlo method for the generation of correlation and covariance matrices for independent and cumulative fission yields has been developed. The method uses a constrained Monte-Carlo resampling structure in order to vary evaluated fission yield libraries in a way that meets basic conservation principles. This results in the generation of correlation/covariance matrices with limited model bias and uncertainty; the matrices are primarily reflective of the evaluated fission yield uncertainties and correlations that arise from the evaluation process. This method has been applied to generate correlation and covariance matrices for all of the fissioning systems of the ENDF/B-VIII.0 and JEFF-3.3 evaluations, marking the first time such matrices have been generated for all of these systems. These covariance matrices have been published online for immediate public use. Presented is the method by which these matrices were generated, a discussion of the results, and examples of the use of these matrices in applications.

Speaker: Eric Matthews
• 07:00
Uncertainty Quantification for Phenomenological Optical Potentials 12m

Phenomenological optical model potentials (OMPs) have played an essential role in scattering calculations for several decades. Despite their successes, none have had their parametric uncertainty fully quantified, making reliable extrapolation difficult. To address this gap, we have developed a generic OMP uncertainty quantification framework that leverages Markov-Chain Monte Carlo for OMP parameter inference. Using this framework, we have revisited the original analyses for the Koning-Delaroche and Chapel Hill '89 potentials, assigned well-calibrated uncertainties, and validated their effectiveness against a broad test corpus of recent experimental scattering data. Finally, we propagate these new uncertainties into two illustrative studies: predictions of neutron cross sections along well-studied isotopic chains, and examination of selected proton-induced reactions relevant for astrophysical nucleosynthesis. We conclude that, contrary to common practice, the substantial inherent uncertainty of standard OMPs can no longer be ignored.

This work was performed in part under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

Speaker: Cole Pruitt (Lawrence Livermore National Laboratory)
• 06:00 07:30
Facilities: IV Sutter's Fort ()

### Sutter's Fort

Convener: Dr Richard Hughes (LLNL)
• 06:00
Nuclear Data Production System of RAON 24m

The Nuclear Data Production System (NDPS) is an experimental system for measuring nuclear data by use of neutron Time-of-Flight detection systems built at Daejeon, Korea. The RAON (Rare isotope Accelerator complex for ON-line experiment) provides deuterons and protons up to 98 MeV and 83 MeV, respectively. They are accelerated by a superconducting driver LINAC and are delivered to the neutron production target to produce neutrons. Pulsed beams with intensities up to ~12 μA can be used to do experiments for measuring neutron-induced cross sections. The range of the beam repetition rate is to be 1 kHz ~ 500 kHz while the RF frequency of the LINAC is 81.25 MHz. The pulse width of the beams is an important factor in determining the accuracy of nuclear data and is aimed to be as small as 1~2 ns. Both white neutrons and monoenergetic neutrons will be produced. White neutrons will be generated by thick C (graphite) targets, while thin Li targets will produce monoenergetic neutrons. Beam lines for charged particles and neutrons together with a neutron collimator and beam dumps are constructed. Nuclear data such as (n, fission), (n, xn), and (n,γ) cross sections can be measured with various detectors. Neutron detectors based on MICROMEGAS and PPAC with converters are installed for monitoring neutron flux. The current status of the construction of the NDPS will be presented.

Speaker: Prof. Seung-Woo Hong (IBS/SKKU)
• 06:24
Overview of Nuclear Data Production System at RAON 12m

Neutron beams have been utilized not only in the basic science, but also in the various industry sectors such as nuclear power, aerospace, defense industry, and semiconductor industry, over the past decades. Although the needs for the nuclear data using the neutron beam extensively have been increased worldwidely, however, it is still insufficient, in particular for the high energy neutron-induced cross sections due to the lack of the high energy neutron beam facility. A neutron beam facility, so-called Nuclear Data Production System (NDPS), is being constructed, so as to measure the nuclear data by employing Time-of-Flight (ToF) technique at Rare isotope Accelerator complex for ON-line experiment (RAON) in Korea. NDPS is planning to complete the installation in 2023. which will provide both white and mono-energetic neutrons using 49 MeV/nucleon deuteron and 83 MeV/nucleon proton beams with graphite and lithium targets, respectively. Since the ToF technique is employed in NDPS, pulsed beams at least less than 200 kHz repetitions with 1-2 nsec width of micro bunch are required in order to obtain sufficient accuracy of the nuclear data. The current status of NDPS will be presented along with the brief overview of RAON.

Speaker: Dr Cheolmin Ham (Rare Isotope Science Project, Institute for Basic Science, Republic of Korea)
• 06:36
Neutron monitoring detector system for Nuclear Data Production System of RAON 12m

The Nuclear Data Production System (NDPS) which is a neutron time-of-flight (TOF) experimental facility of RAON is under construction. The proton and deuteron beams are accelerated up to 83 and 98 MeV, respectively. The proton beams bombarded with a thin Li target will produce mono-energetic neutrons, and continuous-energy neutron spectra will be generated by colliding deuteron beams to a thick C target. To monitor the neutron flux we developed neutron monitoring detectors based MICROMEGAS and PPAC. In order to convert neutrons to charged particles, Thorium converters of 6 cm diameter are installed. The electrodeposition method was used with isopropyl alcohol as a solvent to dissolve thorium nitrate powder. The amount of deposited Th was measured by using alpha spectroscopy. These neutron monitoring detectors were tested with a $^{252}$Cf neutron source. The test results of these two monitoring detectors will be presented.

Speaker: Mr Dal-Ho Moon (Sungkyunkwan University)
• 06:48
Dead-Time correction method with pulse pile-up rejection for neutron TOF experiments 12m

In the field of radiation detection using Ge detectors, pulse pile-up rejection processing (PUR) is widely used. PUR reduces the distortion of the spectrum shape but it causes further increase of dead time. Many papers have been written on the dead time losses with PUR and methods to correct the influence [1],[2]. The proposed correction methods presuppose that the timing of pulses is random. However, in neutron time-of-flight (TOF) experiments, event rates dramatically change depending on TOF and the presupposed randomness is not satisfied. Then a new dead time correction method applicable for neutron TOF experiments is required.

The dead-time influence in TOF is simulated by using a CAEN v1724 digitizer, a BNC DB-2 random pulse generator, a BNC PB-5 pulse generator and a clock generator. 1 kHz triggers were produced by the clock generator and fed into the digitizer as a reset signal of a clock counter (dummy of triggers from an accelerator). With the rate of 70% of the reset signals, trigger signals delayed by 12 $\mu$s were fed into the external trigger input of PB-5 to produce input signals. The random pulse generator was used to produce dummy constant background signals. The produced input and background signals were fed together into the digitizer. The time interval from the reset signal and the pulse height of the signal for each event were recorded as a list-formatted data.

An example of the obtained time dependence spectrum is shown in Figure. The constant background rate decreases not only after the event, but also before it as well. In the time range from 11.5 $\mu$s to 12.25 $\mu$s, the constant background rate is recovered. This is because the digitizer is incapable of discriminating two events in this range and reports as a single event. The reported signal height indicates the sum of the deposited energy in the detector of the two events. In the present work, we refer the time range from 9.5 $\mu$s to 11.5 $\mu$s and 12.25 $\mu$s to 15.2 $\mu$s as “dead state” and that from 11.5 $\mu$s to 12.25 $\mu$s as “accidental sum state”.

The dead time of the digitizer can be calculated in offline data analysis using the list format data. In a normal gamma-ray spectroscopy, the accidental sum state is not acceptable, because single gamma-ray energy is needed. In this case, the rate of the dead time, $P_{dt}(i)$, is calculated at each TOF channel as follows:
$P_{dt} (i)=(T_{ds} (i)+T_{as} (i))/(N_{shot} \delta_t )$
where $T_{ds}(i)$ is the time length of the digitizer being in the dead state, $T_{as}(i)$ is that being in the accidental sum state, $N_{shot}$ is the number of the reset signals and $\delta_t$ is the TOF bin width. On the other hand, in neutron TOF experiments with the pulse height weighting technique, the accidental sum state is acceptable if a weighting function of a detector is close to be a linear. (Of course, its uncertainty should be estimated.) In this case, the rate of the dead time is calculated by following equation.
$P_{dt}(i)=(T_{ds}(i))/(N_{shot} \delta_t )$

This dead time correction has been applied to the analysis of the data measured with CAEN V1724 digitizers in Beam Line #04 (ANNRI) at J-PARC to deduce neutron capture cross section.

Speaker: Atsushi Kimura (Japan Atomic Energy Agency (JP))
• 07:00
JRC MONNET - the intense fast-neutron source for fundamental and application-driven research 12m

MONNET is a fast neutron source based on a tandem-accelerator, located at the Geel (BE) site of the Joint Research Centre (JRC). It became operational in 2020. MONNET may deliver intense neutron beams in the energy range from $30~\text{keV}$ to $10.1~\text{MeV}$ and from $12.8~\text{MeV}$ to $24~\text{MeV}$. Neutrons are produced with protons or deuterons on lithium, tritium or deuterium targets. MONNET delivers a neutron flux of up to $10^9~\text{n/sr/s}$, depending on the producing reaction and the neutron energy. Neutron beams are essentially mono-energetic ($\Delta E_n/E_n < 6%$ with $E_n > 300~\text{keV}$). The accelerator may also be used with proton, deuteron and alpha beams. The production of photon beams is possible and presently under investigation.

The MONNET neutron source is a user facility within the JRC EUFRAT Open Access program. Proposal evaluation by an independent panel is taking place twice per year.

The research program ranges from cross section measurements, e.g. $(n, f)$, $(n, p)$, $(n, \gamma)$ as well as $(p, p')$, $(p, n)$ and $(p, \gamma)$, nuclear fission studies, material studies (e.g. radiation-induced damage), to the investigation of advanced methods in nuclear technologies, safety and security. The activities may have exploratory character for the development of new scientific concepts or testing of new equipment.

We will give a detailed overview of the MONNET infrastructure, the possibilities offered to scientists and will report about first experiment campaigns.

Speaker: Dr Cristiano Lino Fontana (European Commission, Joint Research Centre (JRC))
• 06:00 07:30
Fission: VI American River ()

### American River

Convener: Andrea Mattera (BNL)
• 06:00
New measurement of the $^{235}$U($n_{\rm th}$,f) fission yields and development of a Time of Flight line at the LOHENGRIN spectrometer 24m

The study of nuclear fission yields has a major impact on the characterization and understanding of the fission process and is mandatory for reactor applications. In the framework of a collaboration between the CEA, the LPSC and the ILL, a program of actinide fission yields measurements has been initiated for several years at the LOHENGRIN spectrometer. However, the measurement of very low fission yields in the symmetry region and the heavy wing of the distributions are difficult to achieve due to the strong contamination by other masses with much higher yields and require the development of a new experimental setup.

This talk will first present the results of an absolute measurement of the $^{235}$U(n$_{th}$,f) mass yields using an ionization chamber placed at the exit of the spectrometer. Although very well documented in the literature, these yields show uncertainties lying from 3% to 10% with large discrepancies between libraries and a lack of correlation matrices. New experimental data obtained at the LOHENGRIN spectrometer will be detailed, along with the measurement method and the production of experimental covariance matrices.

The second part will show the development of a Time of Flight (ToF) line in order to improve the background rejection in the mass yield measurements. In the symmetry region, the precision of the measurement is limited by the background estimation due to the charge exchanges with the residual gas of the separator. We foresee to analyze the events using a triple coincidence (ΔE x E) x ToF, whereas today only ΔE x E selection is available. The new ToF line is built using Si$_3$N$_4$ foils and electron detectors for the start and stop detectors. The talk will present the choices made for the electron detectors technology along with the progress achieved on the ToF line characterization.

Speaker: Christophe Sage
• 06:24
Measurement of $^6$Li(n,t) and $^{235}$(n,f) with cold neutrons 12m

The Alpha-Gamma device at the National Institute of Standards and Technology (NIST) utilizes neutron capture on a totally absorbing $^{10}$B deposit to measure the absolute flux of a monochromatic cold neutron beam to better than 0.1% precision. Gammas produced by the boron capture are counted using high purity germanium detectors, which are calibrated using a precisely characterized alpha source and the alpha-to-gamma ratio from neutron capture on a thin $^{10}$B target. This device has been successfully operated and used to calibrate the neutron flux monitor for a beam-based neutron lifetime experiment at NIST. Here we will discuss work underway to provide systemically-distinct measurements of the $^6$Li(n,t)$^4$He and $^{235}$U(n,f) cross sections with competitive precision.

Speaker: Pieter Mumm
• 06:36
Charge polarization calculated with a microscopic model for the fission fragments of U-236 12m

The charge distribution of fission fragments is a significant quantity to estimate and evaluate the neutron emission yields from fission fragments. The distribution has been evaluated as the deviation from the unchanged charge distribution (UCD) assumption in which the fragments keep the neutron-proton ratio of the fissile parent nucleus. The deviation is called the charge polarization (CP).

For the major fission reactions in the nuclear reactor, their CP have been compiled in a library called Wahl's systematics. Although the library is designed for the nuclear engineering field, this is not suitable to predict the fission fragments from unmeasured fissile nuclei.

In order to provide the CP of fragments generated from unknown fissile nuclei without empirical ways, we suggest a new method based on the microscopic nuclear theory. We employ the mean-field theories, which are the constrained Skyrme Hartree-Fock+BCS theory and the canonical-basis time-dependent Hartree-Fock-Bogoliubov theory, represented in three-dimensional space.

In the presentation, we will report the calculated CP of the fission products on the $^{236}$U, assuming the reaction of $^{235}$U absorbed a thermal neutron. We will discuss the effectiveness of our method and will mention the dynamical effects on the CP through the comparison among results calculated by static, dynamic mean-field models and the data in Wahl's systematics.

Speaker: Shuichiro Ebata
• 06:48
Experimental determination of fission barrier parameters 12m

The multi-humped fission barrier is the result of superposing microscopic shell corrections to the macroscopic liquid drop model [1]. A manifestation of that is the existence of shape isomers [2, 3], as observed in several actinide nuclides. However, the - most often - double-humped barrier is much more than just a concept, since it is related to several fission observables. These data may have an impact on both fundamental nuclear physics and nuclear applications. Sometimes, model calculations of the barrier are performed and observables like various cross sections are predicted [4], which then have to withstand comparison with experimental results. In a recent work [5] it was demonstrated how different measured data from the decay of the shape isomer in $^{235}$U together with fission fragment properties could be used to estimate barrier parameters and to construct the shape of the fission barrier. While that work is based on neutron-induced fission measurements, a recent experiment at GSI was performed to study fission isomers, produced in fragmentation reactions, with the FRS. Data analysis from this experiment is still ongoing, but we will give an overview of the results obtained so far and present the current status of the data treatment.

References:
[1] V. M. Strutinsky, Nucl. Phys. $\textbf{A95}$, 420 (1967).
[2] S. Bjørnholm and J. E. Lynn, Rev. Mod. Phys. $\textbf{52}$, 725 (1980).
[3] P. Thirolf and D. Habs, Prog. Part. Nucl. Phys. $\textbf{49}$, 325 (2002).
[4] M. Sin, R. Capote, M.W. Herman, and A.Trkov, Nucl. Data Sheets $\textbf{139}$, 138 (2017).
[5] A. Oberstedt and S. Oberstedt, Phys. Rev. C $\textbf{104}$, 024611 (2021).

Speaker: Andreas Oberstedt
• 06:00 07:30
Integral Measurements: I Capitol ()

### Capitol

Convener: Marie-Anne Descalle (LLNL)
• 06:00
Outcomes of WPEC SG47 on »Use of Shielding Integral Benchmark Archive and Database for Nuclear Data Validation« 24m

Working Party on International Nuclear Data Evaluation Co-operation Subgroup 47 (WPEC SG47) entitled "Use of Shielding Integral Benchmark Archive and Database for Nuclear Data Validation" was started in June 2019 with the objectives to promote a more systematic and wider use of shielding benchmark experiments in nuclear data and transport code validation and development, to provide feedback on the Shielding Integral Benchmark Archive and Database (SINBAD), and to promote its further development in coordination with Expert Group on Physics of Reactor Systems (EGPRS). Altogether 6 meetings, the large majority (5) held remotely by videoconference, were organised during the past 3 years to discuss the experience on the use of SINBAD, new benchmark experiment evaluations and improvements to be contributed to the database which is severely malnourished and lacking maintenance since the past ~10 years or more. Several proposals for new or updated benchmark evaluation were presented and discussed, such as FNG copper, LLNL pulsed spheres, CIAE iron sphere, KFK 1977 gamma measurements, Rez Fe sphere, ASPIS, Oxygen ORNL broomstick, TIARA and other benchmark experiments. Complementing the database with new features was also discussed. For example, providing nuclear data sensitivity profiles more systematically would facilitate and better guide the use of data. Also, the information on the geometry, (radiation source) and materials in CAD format is expected to allow an easier and less error prone computational model preparation for different transport codes. Inputs for various transport codes and other benchmark data from participants have been shared via NEA GitLab which could hopefully in future evolve and form a bases for critically checked and validated set for computer code analysis tools.
Examples of the use and some views on future development of the SINBAD benchmark database will be presented in the paper.

Speaker: Ivan Alexander Kodeli (CCFE/UKAEA Abingdon)
• 06:24
EUCLID: A New Approach to Improve Nuclear Data Coupling Optimized Experiments with Validation using Machine Learning 12m

\documentclass{style/nseJournal}

\usepackage{enumitem}
\usepackage{xcolor}
\usepackage{hyperref}
\usepackage{cleveref}
\usepackage{graphicx}
\usepackage{rotating}
\usepackage{ulem}

\newcommand{\beff}{$\beta_\mathrm{eff}$}
\newcommand{\keff}{$k_\mathrm{eff}$}

\begin{document}

\title{EUCLID: A New Approach to Improve Nuclear Data Coupling Optimized Experiments with Validation using Machine Learning}

\correspondingEmail{jesson@lanl.gov}

\addAffiliation{a}{Los Alamos National Laboratory\ P.O. Box 1663, MS B228, Los Alamos, New Mexico 87545}

\titlePage

\section{Abstract}
\label{sec:Intro}
Unconstrained physics spaces arise between nuclear data due to imprecise knowledge or missing differential experimental data and theory combined with the fact that validation with integral experiments cannot uniquely identify one nuclear-data value as driving the difference between simulated and experimental values of these experiments. Compensating errors of nuclear data can easily hide within these unconstrained physics spaces. This can potentially have large impacts on application calculations that do not closely resemble validation experiments we validate our nuclear data against.

The EUCLID project (Experiments Underpinned by Computational Learning for Improvements in Nuclear Data) aims to resolve part of these compensating errors by performing an integral experiment optimized to better constrain nuclear data. In addition to that, the project will adjust ENDF/B-VIII.0 nuclear data with respect to this new experiment to better constrain nuclear data.

EUCLID includes four major components: (1) simulations, (2) experiments, (3) machine learning, and (4) nuclear data validation/adjustment. One major component of EUCLID is the evaluation of several integral measurement methods (criticality, reaction rate ratios, neutron noise, etc.). Regarding simulations, new capabilities are being developed to estimate sensitivity coefficients for all of these methods. Regarding experiments, the National Criticality Experiments Research Center (NCERC) will be utilized to perform measurements. These experiments will be optimized using Bayesian optimization. The measured responses (along with the simulated responses and sensitivities) will be utilized via Random Forest to perform nuclear data adjustment. This will include requirements to obey physics constraints and will result in adjusted nuclear data.

\begin{figure}[htb!]
%\vspace{-20pt}
\centering\includegraphics[width=6in]{figures/EUCLIDimage.pdf}
%\vspace{-5pt}
\caption{Unconstrained physics spaces within nuclear data can adversely impact application simulations, but cannot be resolved using current techniques. EUCLID aims to provide a capability to help understand such unconstrained spaces. In order to achieve this, we utilize machine learning with simulated sensitivities in MCNP of experiments and measured results to augment expert identification of areas where compensating errors may be resolved. The design of new experiments will then be optimized using machine learning. After the new experiments are completed, the new data will be used to produce adjusted libraries.}
\label{Fig_EUCLID}
%\vspace{-25pt}
\end{figure}

\newpage

\section*{Acknowledgments}
\label{sec:acknowledgements}
\textcolor{black}{Research reported in this publication was supported by the U.S. Department of Energy LDRD program at Los Alamos National Laboratory. The National Criticality Experiments Research Center (NCERC) is supported by the DOE Nuclear Criticality Safety Program, funded and managed by the National Nuclear Security Administration for the Department of Energy. This work was supported by the US Department of Energy through the Los Alamos National Laboratory. Los Alamos National Laboratory is operated by Triad National Security, LLC, for the National Nuclear Security Administration of the US Department of Energy under Contract No. 89233218CNA000001.}

\end{document}

Speaker: Jesson Hutchinson (LANL)
• 06:36
Use of nickel sphere and copper cube with Cf-252 neutron source in the centre for test of nuclear data library ENDF/B-VII, ENDF/B-VIII, JEFF-3.3, BROND-3.1 and ROSFOND 12m

Abstract
The leakage neutron spectra measurements have been done on benchmark spherical assembly-nickel sphere with diameter of 50 cm and copper cube with an edge of 48.4 cm. The Cf-252 neutron source was placed into the centre of nickel sphere and copper cube. The proton recoil method was used for neutron spectra measurement using spherical hydrogen proportional detectors (HPD) with pressure of 400 and 1000 kPa (diameter of detectors is 4 cm) and using scintillation stilbene detector (diam. 1x1 cm). The neutron energy range of spectrometer is 0.06-1.3MeV for HPD (HPD region) and 1-10MeV for stilbene (stilbene region). The adequate MCNP neutron spectra calculations based on data libraries ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.3, BROND-3.1 and ROSFOND-2010 were done. The neutron energy structure used for calculations and measurements was 40 and 200 groups per decade (gpd) and step of 100 keV for stilbene. Structure 200 gpd represents lethargy step about of 1%.

Speaker: Dr Bohumil Jansky (Research Centre Rez)
• 06:48
Data Assimilation using Non-invasive Monte Carlo Sensitivity Analysis of Reactor Kinetics Parameters 12m

LA-UR-21-29561

Abstract

High-fidelity simulation of nuclear systems are paramount to the operational success and safety of many nuclear applications. Simulations rely on quality nuclear data. The nuclear data input for these simulations is continuously being reassessed by nuclear data evaluators. One way to support this ongoing evaluation is to present the evaluators with detailed analysis from data assimilation techniques. These techniques examine major sources of nuclear data-induced uncertainties in important nuclear parameters, such as measures of criticality. In this work, a data assimilation technique called the bias factor method will be utilized to improve the predicted accuracy and precision of the effective neutron multiplication factor documented for multiple International Criticality Safety Benchmark Experiment Project (ICSBEP) benchmarks [1]. This data assimilation technique has been shown to be successful in reducing nuclear data-induced uncertainty in the effective neutron multiplication factor by researchers at Nagoya University [2]. To perform calculations of the bias factor, one must calculate sensitivity coefficient matrices of a parameter of interest. The parameters to be examined in this work include the effective delayed neutron fraction, reactivity coefficient, and prompt neutron decay constant. Monte Carlo N-Particle® Code Version 6.21 (MCNP®6.2) will be used in a non-invasive manner (i.e., source code will not be modified) to calculate these sensitivity coefficients. The calculation of these sensitivity coefficients follows work performed by researchers at TerraPower LLC [3] and Jožef Stefan Institute [4]. The data will be subsequently used to assist with the optimization of an integral experiment for the Experiments Underpinned by Computational Learning for Improvements in Nuclear Data (EUCLID) project at Los Alamos National Laboratory. By examining the results of the bias factor method with the three aforementioned parameters, more certainty can be attributed to the optimal integral experiment providing results that can improve/adjust uncertain or inaccurate nuclear data.


1 MCNP® and Monte Carlo N-Particle® are registered trademarks owned by Triad National Security, LLC, manager and operator of Los Alamos National Laboratory. Any third party use of such registered marks should be properly attributed to Triad National Security, LLC, including the use of the designation as appropriate. For the purposes of visual clarity, the registered trademark symbol is assumed for all references to MCNP within the remainder of this paper.

Acknowledgements

Research reported in this publication was supported by the U.S. Department of Energy Laboratory Directed Research and Development (LDRD) program at Los Alamos National Laboratory. Los Alamos National Laboratory is operated by Triad National Security, LLC, for the National Nuclear Security Administration of the US Department of Energy under Contract No.89233218CNA000001.
`

References

[1] J. D. Bess, T. Ivanova, L. Scott, I. Hill, “The 2019 Edition of the ICSBEP Handbook,” Transactions of the American Nuclear Society, 121, 901-904 (2019).

[2] T. Endo, A. Yamamoto, “Data Assimilation Using Subcritical Measurement of Prompt Neutron Decay Constant,” Nuclear Science and Engineering, 194, 1089-1104 (2020).

[3] N. Touran, J. Yang, “Sensitivities and Uncertainties Due to Nuclear Data in a Traveling Wave Reactor,” PHYSOR, Sun Valley, Idaho, May 1 – 5 (2016).

[4] I. Kodeli, “Sensitivity and uncertainty in the effective delayed neutron fraction (β_eff),” Nuclear Instruments and Methods in Physics Research A, 715, 70-78 (2013)

Speaker: Noah Kleedtke (LANL)
• 07:00
(WITHDRAWN) Automation of Nuclear Data Validation through Integral Benchmarks 12m

Verification and validation are essential components of the nuclear data development process, since only those data that have been demonstrated to accurately simulate real-life applications can be relied upon for those applications. While most nuclear data work relies on so-called 'differential' measurements, the validation of this work requires data from full systems also called Integral Benchmarks. Thousands of integral benchmarks are available and their use for Nuclear Data validation requires an automated process to minimize work time and improve Quality Assurance. For this purpose, an automated tool is developed at LANL.

The automated Nuclear Data Validation Tool is built as a tool kit. It is designed to be open source and user friendly. The automated tool is composed of five modules. When using all the modules successively, it aims at processing one or several ENDF format data file to be used by the MCNP transport code to run a selection of integral benchmarks and provide simulation results into graphs or tables. Each module can be run separately as well.

The first module encapsulates the NJOY data processing code to create a full or a partial nuclear data file library in the ACE format and its associated xsdir file. Parameters available to users are temperatures and the NJOY version code. The second module constitutes a database of MCNP input files associated to a search functionality. About 1,200 of input files from the International Criticality Safety Benchmark Evaluation Project (ICSBEP) are available. Benchmarks and benchmark input files can be selected through metadata and isotope occurrence. User-defined libraries or the regular MCNP libraries are available. The library definition operates at the isotope level. The third module is an MCNP run launcher. For a list of cases, it submits runs either sequentially or simultaneously. Workstation as well as HPC platforms are supported; 6.1 and newer versions of MCNP are accessible. Parallelism through MPI, OMP or mixed MPI/OMP is user-defined as well as the architecture of the computing platform. The postprocessing of the MCNP output files is performed in the fourth module. Results are stored in json files. This module can also perform comparisons to experimental values.
The last module is dedicated to documentation and the presentation of results through graphs and/or tables. For each plot, all parameters are user-defined. Plot parameters can be saved and stored into a list of pre-defined graphs for further use.

Speaker: Cecile Toccoli
• 06:00 07:30
Measurements: VI Folsom ()

### Folsom

Convener: Frank Gunsing (CEA)
• 06:00
The past and the future of the GAINS spectrometer @ GELINA 24m

The GAINS spectrometer operating at the GELINA neutron source (EC-JRC-Geel) is one of the best known setups within the nuclear data community. It provides reliable, high resolution $\gamma$-production cross-section data of importance both for the fundamental nuclear physics research and its many applications. The story of this spectrometer started around 2000 with just 2 large volume HPGe detectors and a Fission Chamber but, during the following years, it extended to 12 detectors allowing our group to perform numerous measurements ($^{52}$Cr, $^{208}$Pb, $^{23}$Na, $^{28}$Si, $^{76}$Ge, $^{56}$Fe, $^{24}$Mg, $^{206}$Pb, $^{46-50}$Ti, $^{7}$Li, $^{57}$Fe, $^{54}$Fe, $^{16}$O, $^{58,60,61,62,64}$Ni, $^{40}$Ca) [1-17]. The extracted data were consistently used, among others, in several evaluations and also for establishing a new inelastic $\gamma$-production reference cross section. This talk presents an overview of the work done within the JRC – IPHC – IFIN-HH collaboration at GELINA (which recently extended to include also ESRIG and University of Helsinki), focusing on the achievements of GAINS. We will not only describe past work but we will also emphasize the future plans: upcoming experiments ($^{14}$N, $^{56}$Fe, $^{35-37}$Cl) and moving to a new digital acquisition system (both hardware and software upgrades) and others.

$^{1}$L. C. Mihailescu, PhD Thesis, University of Bucharest (2006).
$^{2}$L.C. Mihailescu, et al., Nucl. Phys. A786, 1 (2007).
$^{3}$L.C. Mihailescu, et al., Nucl. Phys. A811,1 (2008).
$^{4}$C. Rouki, et al., Nucl. Instrum. Meth. Phys. Res. A672, 82 (2012).
$^{5}$A. Negret, et al., Phys. Rev. C88, 034604 (2013).
$^{6}$C. Rouki, et al., Phys. Rev. C88, 054613 (2013).
$^{7}$A. Negret, et al., Phys. Rev. C90, 034602 (2014).
$^{8}$A. Olacel, et al., Phys. Rev. C90, 034603 (2014).
$^{9}$A. Negret, et al., Phys. Rev. C91, 064618 (2015).
$^{10}$A. Olacel, PhDThesis, University of Bucharest (2016).
$^{11}$A. Olacel, et al., Phys. Rev. C96, 014621 (2017).
$^{12}$M. Nyman, et al., Phys. Rev. C93, 024610 (2016).
$^{13}$A. Negret, et al., Phys. Rev. C96, 024620 (2017).
$^{14}$A. Olacel, et al., Eur. Phys. Journal A54 (2018).
$^{15}$M. Boromiza, PhDThesis, University of Bucharest (2018).
$^{16}$M. Boromiza, et al., Phys. Rev. C 101, 024604 (2020).
$^{17}$A. Olacel et al., Article in preparation (2021).

Speaker: Dr Adina Olacel (Horia Hulubei National Institute for R&D in Physics and Nuclear Engineering (IFIN-HH))
• 06:24
Measurements of $^{35}$Cl(n,x) reaction cross sections 12m

Measurements of $^{35}$Cl(n,x) reaction cross sections were conducted at Lawrence Berkeley National Laboratory’s (LBNL) 88-Inch Cyclotron using neutrons produced via thick target deuteron breakup from a 14 MeV deuteron beam. These cross sections are vital to the design of Molten Chloride Fast Reactors (MCFR), especially in the 0.1 MeV – 1.0 MeV region where little experimental data exist but almost half of the MCFR neutron spectrum lies. The nuclear data evaluation process that produces the cross sections used for the design of an MCFR assumes a fixed total $^{35}$Cl(n,x) cross section. The result is that an increase in one channel causes a corresponding decrease in one or more other evaluated channel(s). To address this aspect of nuclear data evaluation, we performed an experiment with the goal of measuring all energetically possible reaction channels. The experiment consisted of three independent parts. First, energy-angle differential (n,n’${\gamma}$) cross sections were obtained using the GENESIS array, a collection of high-purity germanium detectors and LaBr inorganic scintillators for ${\gamma}$-ray detection as well as EJ-309 organic scintillators for neutron detection. Second, (n,p) and (n,${\alpha}$) energy differential cross sections were obtained using a CLYC elpasolite scintillator as an active target. Finally, energy integral (n,p) and (n,${\alpha}$) cross sections were obtained from the production of the $^{35}$S and $^{32}$P activation products in a pressed NaCl tablet. The results were compared to reaction modeling using the TALYS reaction code package.

This work was performed under the auspices of the U.S. Department of Energy (DOE) by Lawrence Berkeley National Laboratory under Contract DE-AC02-05CH11231, the DOE Nuclear Energy University Program (NEUP), and Lawrence Livermore National Laboratory under Contract DE-AC52- 07NA27344.

Speaker: Tyler Nagel
• 06:36
Direct measurements of neutron-induced charged-particle reactions on radioactive $^{59}$Ni and $^{56}$Ni 12m

Nuclear reaction data for neutron induced reactions on unstable nuclei are critical for a wide range of applications spanning from studies of nuclear astrophysics, nuclear reactor designs, and for radiochemistry diagnostics. However, due to the difficulty in performing this class of measurements, and the resulting lack of experimental data, nuclear data evaluations of the reaction cross sections are typically unavailable or unreliable. In this work, we present partial and total $^{59}$Ni(n,p)$^{59}$Co and $^{59}$Ni(n,$\alpha$)$^{56}$Fe cross sections that were measured directly with a radioactive $^{59}$Ni target, using the fast neutrons available at the WNR facility at LANSCE. The results are compared to the available nuclear data evaluations and to a recent study of the $^{59}$Ni(n,xp) reaction cross section that was performed via an indirect surrogate ratio method. The current study of $^{59}$Ni was also used to inform some of the background contributions in our recent direct measurement of the $^{56}$Ni(n,p)$^{56}$Co reaction using a radioactive $^{56}$Ni (T$_{1/2}$ = 6 days) target. This $^{56}$Ni target was produced at the Isotope Production Facility at LANSCE and measured using the recently commission hotLENZ experimental setup at the WNR facility. Details on the target fabrication process, technical considerations for the experimental setup, and preliminary analysis of the $^{56}$Ni(n,p)$^{56}$Co measurement will also be presented.

Speaker: Hye Young Lee
• 06:48
A new measurement on $^{56}$Fe(n,inl) using GAINS at GELINA 12m

The extended dataset of 56Fe(n,ng) cross sections measured by our group more than a decade ago at GELINA [1] was used in many recent evaluations like ENDF, JEFF and CIELO. Despite the special measures we took to ensure reliability and accuracy, concerns were raised by various groups with regard to several features of our dataset (absolute normalization an/or shape) and therefore the 56Fe(n,inl) cross section is still under the evaluation by the International Nuclear Data Evaluation Network (INDEN) [2]. Consequently a new experiment is now under preparation aiming to take advantage of the numerous experimental improvements of the GAINS setup implemented over the years. While gamma spectroscopy combined with the time-of-flight method will remain the main techniques involved, several other experimental details will differ substantially:
- A new enriched target is already available,
- The number of HPGe detectors of GAINS increased from 4 to 12,
- The ToF flight path decreased from 200 m to 100 m,
- A new digitized data acquisition for GAINS is being implemented.
We will present all these aspects with the purpose of updating the community about our experimental plans while triggering the debate over any additional experimental features that could be considered.

[1] A. Negret, C. Borcea, Ph. Dessagne, M. Kerveno, A. Olacel, A.J.M. Plompen, and M. Stanoiu, Physical Review C90, 034602 (2014)
[2] https://www-nds.iaea.org/INDEN/

Speaker: Alexandru Negret
• 07:00
Beta-delayed neutron emission - a new opening 24m

Beta-delayed neutron emission is a dominant decay mode for the very neutron-rich nuclei. With the opportunities offered by new generation radioactive beam facilities, we can perform experiments with a larger and more exotic pool of isotopes. This enabled us to revisit the conventional views on the fundamentals of beta delayed neutron emission. However, new data and improved models are necessary because of the critical role of beta-delayed neutron emission for astrophysics (r-process) and reactor physics. Using experimental results obtained for very exotic isotopes with neutron arrays such as VANDLE or BRIKEN, I will show how these recent measurements enabled us to improve the understanding of this process. One of the central questions is if and how the nu