Speaker
Description
The safe management of the spent nuclear waste (SNF) requires a special effort of characterization, since its isotopic composition is the key to its transport, handling and storage by means of the decay heat, gamma emission and neutron emission it produces due to radioactive decay.
The effort is being done in the frame of the EU EURAD project (European Joint Programme on Radioactive Waste Management) project, funded under H2020-Euratom-1.2 programme, where its Work Package 8 (Spent Fuel Characterization and Evolution Until Disposal) is in charge of producing procedures to determine the nuclide content of SNF, including uncertainties (in particular, in Task 2.1: Theoretical study of SNF source terms).
Among the work to be done in the mentioned task, sensitivity studies have been proposed to define confidence limits on the current nuclear data libraries and identify nuclear data requiring improvement. The goal of these sensitivity calculations is to estimate the uncertainties in the isotopic composition after the irradiation due to some parameters. Differences between the reference calculation and the sensitivity one are then evaluated to obtain which parameter provides a larger impact and has to be hence better quantified.
In this work, a sensitivity study involving the most recent versions of two nuclear databases, JEFF-3.3 and ENDF/B-VIII.0, has been done. A simulation of the Takahama-3 reactor, a Japanese light water reactor (LWR) operated by Kansai Electric Power Company (KEPCO) whose experimental information was available at the NEA SFCOMPO, has been done with these two libraries. In particular, the fifth pellet of the SF95 rod has been chosen for this task. This pellet had an initial $^{235}$U enrichment of 4.1% and was irradiated for a total of 30.4 GWd/tU.
The reference code used in this task has been EVOLCODE 2.0, successfully validated against LWR experimental data in previous studies. EVOLCODE 2.0 is a combined neutronics and depletion evolution simulation code developed at CIEMAT to describe the burnup evolution of nuclear reactors. The neutron transport is currently being solved by the general MC transport code MCNP/X while the depletion of the materials is carried out by the activation code ACAB.
Results of this sensitivity calculation have been calculated for a list of isotopes due to their relevance in different parameters of interest. This list includes actinides relevant for operation or transmutation causes, fission products for criticality safety reasons, for radioprotection, for post irradiation examination comparison and impurities in the fuel. For the actinides, a small impact can be found, because only $^{242M}$Am presents a significant discrepancy (of around 22%), while the other actinides show differences smaller than 6%. The discrepancy of $^{242M}$Am comes as a consequence of having different values of the $^{241}$Am capture cross sections in both libraries. In particular, since EVOLCODE requires the calculation of one-group cross sections to perform the irradiation, this difference could be quantified. A systematic ~37% larger value of the $^{241}$Am capture cross sections along burn-up was found for JEFF-3.3 regarding ENDF/B-VIII.0.
Most of the studied fission products show similar results for both libraries, with differences smaller than 5%. However, some of them show a significant discrepancy. This is the case of $^{14}$C, $^{109}$Ag, $^{110M}$Ag, $^{129}$I, $^{142}$Nd, $^{151}$Eu, $^{157}$Gd and $^{158}$Gd. An exhaustive analysis has been done for each of these fission products to understand the origin of the discrepancies.
As an example, $^{157}$Gd shows a discrepancy between libraries of around 50%. This isotope appears due to several processes. First, it can be generated as an independent fission yield. Second, it can be created by decay from other direct fission yields having A=157. Third, it can be generated by captures in $^{156}$Gd. Finally, it can appear from a successive capture and decay starting in $^{156}$Eu. $^{156}$Eu is a β$^{-}$ emitter with a half-life of 15.2 d that is created by its own chain of fission yields. However, having this long enough half-life together with a large enough capture cross section, makes that its capture rate competes with its decay, leading to $^{157}$Eu, parent of $^{157}$Gd. This is the cause of the discrepancy of 50% between libraries. This cross section has a long energy range up to around 1 eV where there is a discrepancy of two orders of magnitude. This creates a one-group cross section of 342 b in JEFF-3.3 and 52 b in ENDF/B-VIII.0. None of these libraries provides the uncertainty in the $^{156}$Eu capture cross section. Besides, the larger $^{156}$Eu capture in JEFF-3.3 also causes a smaller amount in $^{156}$Eu and hence a smaller $^{156}$Gd amount of ~11.3% at the end of the irradiation in this library, with respect to ENDF/B-VIII.0.
Recommendations to improve the JEFF and ENDF/B libraries will be provided as a result of the sensitivity analysis performed as part of this work.