Speaker
Description
In recent years, molten salt reactor (MSR) concepts have seen renewed interest both within academia and emerging industry. Several widespread reactor concepts are developed, sharing the feature that the fissile fuel is dissolved in a molten fluoride or chloride salt which remains liquid over a wide temperature range at atmospheric pressure. One of the attractive features of MSRs is precisely that high pressure is avoided, thereby eliminating a range of possible accident scenarios. Combining this with the fuel salt fission product retention properties and the ability to do online fission product processing, paves the way for a completely different safety case for MSR concepts with respect to conventional reactors, allowing significant savings on construction costs.
The MSR concept is not new. Two reactors were built and operated at ORNL in the 50’ies and 60’ies, but the technology never matured into commercial reactors. Though generally successful according to assessment reports, the operation of these reactors did identify several challenges 1. Among the more concerning were (irradiation assisted) corrosion issues of the structural material, while another relates to the mechanical and moderation properties of the graphite moderator under irradiation [2].
Although significant efforts have been made, the issues relating to the graphite behaviour under irradiation have not been resolved during the half century that has passed. At present, some MSR developers are focusing their research on solving this issue [3], while others plan frequent replacements of the core, to limit irradiation damage [4].
At Seaborg Technologies, a Danish based MSR developer, a different path is chosen, namely to replace the graphite moderator with sodium hydroxide (NaOH) [5].
While this introduces several complications relating to corrosion and chemistry control, it does also promise several benefits. During operation of the reactor, the NaOH would be liquid, thus eliminating issues relating to radiation damage. Additionally, due to the presence of hydrogen, the moderation power is significant allowing for a compact core design, which is essential for the assembly line based deployment and central to Seaborg’s business case.
Another challenge following the use of sodium hydroxide relates to the fact that only limited past experience is available in the literature [6]. Although the interactions of neutrons with materials can be reliably predicted using Monte-Carlo techniques for energetic neutrons using nuclear data per nuclide, in the thermalization regime, where the neutron kinetic energy approaches the energy levels of the molecules of the moderator, molecular levels need to be considered.
To facilitate the usage of NaOH as a moderator, the modelling must be supplemented with thermal scattering kernels describing interactions between neutrons and NaOH in the thermal regime. The development of such scattering kernels relies on experiments as well as modeling and will be described in detail in this conference contribution. The outline is as follows:
Using the pulsed neutron beam available at the ISIS Pulsed Neutron and Muon Source, at the Rutherford Appleton Laboratory (UK), it is possible to determine total cross sections as a function of the energy using the time-of-flight (TOF) technique and over a wide range of energies. For the present work, we carried out experiments in May 2020 using the VESUVIO beamline - see figure below, a neutron spectrometer optimized for the routine measurement of the total cross sections over a broad energy range [7,8].
A sample of 8.47g of NaOH at room temperature was prepared [9], and placed centrally in the neutron beam. To limit multiple scattering and undesired scattering on the sample container, the sample was pressed to 1.5 mm thickness inside a polyethylene bag.
VESUVIO instrument at ISIS [8].
The incident as well as transmitted beams were monitored, allowing the TOF spectrum to be converted into a neutron energy spectrum, ranging over several orders of magnitude and covering the energy region of interest for thermal scattering kernel development (roughly: 1meV-0.1eV).
In the post-processing, the measured data was normalized to the well-known epithermal cross-section. The measured results have been compared with calculations using NJOY [10] and NCrystal [11], before making them available for use in modern Monte-Carlo codes, such as Serpent, MCNP and OpenMC.
The results for the total cross-section of neutrons on solid NaOH at ambient temperatures will be presented along with examples of usage in Monte-Carlo reactor modeling.
In addition, the process forward for measurements on liquid NaOH at elevated temperatures, more relevant to reactor operation, will be described.
References:
1 P. N. Haubenreich J. R. Engel, ”Experience with the Molten-Salt Reactor Experiment”, Nucl. App. and Tech. Vol 8, 1970 - Issue 2, pp 118-136
[2] P. R.Kasten et al, “Graphite behavior and its effects on MSBR performance”, Nucl. Eng. and Design, Vol 9, Issue 2, (1969), pp157-195
[3] G., Zhu et al, “Neutronic effect of graphite dimensional change in a small modular molten salt reactor”,Int.J of Energy Research — 2020, Vol 45, Issue 8, pp. 11976-11991
[4] https://thorconpower.com/ and https://www.terrestrialenergy.com/ are examples.
[5] Patent WO 2018/229265 Al - “Molten Salt Reactors”
[6] C.E. Teeter, et al.. Catalog Of Nuclear Reactor Concepts. Part I. Homogeneous and Quasi-Homogenous Reactors. Section III. Reactors Fueled With Molten-Salt Solutions. No. ANL-7092. Argonne National Lab., Ill., 1965.
[7] J.I. Robledo, et al, “Measurement of neutron total cross section at the VESUVIO spectrometer”, Nucl. Inst. Meth. A, A 971 (2020) 164096
[8] https://www.isis.stfc.ac.uk/Pages/Vesuvio.aspx
[9] Sodium hydroxide white powder 757527 – 25g ultra dry from Sigma Aldrich. https://www.sigmaaldrich.com/GB/en/product/aldrich/757527
[10] R. Macfarlane, et al. “The NJOY nuclear data processing system”, version 2016. No. LA-UR-17-20093. Los Alamos National Lab.(LANL), Los Alamos, NM (United States), 2017.
[11] XX. Cai, and T. Kittelmann. "NCrystal: A library for thermal neutron transport." Computer Physics Communications 246 (2020): 106851.