Validation of Monte Carlo based burnup codes against LWR-PROTEUS Phase-II experimental data

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• 作者：Marco Pecchia^a ; ^{marco.pecchia@psi.ch" class="auth_mail" title="E-mail the corresponding author} ; Damar Wicaksono^a ; ^{damar.wicaksono@psi.ch" class="auth_mail" title="E-mail the corresponding author} ; Peter Grimm^a ; ^{peter.grimm@psi.ch" class="auth_mail" title="E-mail the corresponding author} ; Alexander Vasiliev^a ; ^{alexander.vasiliev@psi.ch" class="auth_mail" title="E-mail the corresponding author} ; Gregory Perret^a ; ^{gregory.perret@psi.ch" class="auth_mail" title="E-mail the corresponding author} ; Hakim Ferroukhi^a ; ^{hakim.ferroukhi@psi.ch" class="auth_mail" title="E-mail the corresponding author} ; Andreas Pautz^a ; ^b ; ^{andreas.pautz@psi.ch" class="auth_mail" title="E-mail the corresponding author}
• 关键词：Monte Carlo codes ; Burnup ; Validation ; MCNP ; SERPENT ; VESTA ; Code comparison
• 刊名：Annals of Nuclear Energy
• 出版年：2016
• 出版时间：November 2016
• 年：2016
• 卷：97
• 期：Complete
• 页码：153-164
• 全文大小：4184 K

文摘

The increasing computer performance has extended the applicability of Monte Carlo codes to isotopic inventory prediction. Validation activities of Monte Carlo based burnup codes were started at PSI about a decade ago. The use of such codes is not foreseen as main “workhorse” for burnup calculations but rather as verification and validation tool, providing reference solutions.

During the LWR PROTEUS Phase II program precise radiochemical post irradiation experiments have been carried out. In the experimental program, 13 fuel samples were analyzed, and, 2 of these were selected for codes verification and validation comparison due to their relatively simple irradiation histories. Previous analyses were focused in investigating simple fuel pin models but representative of the irradiation history and boundary condition of the experimental sample.

The primary target of this work is to improve the level of validation of Monte Carlo based burnup codes using more complex fuel assembly model. The rationale for such improvement is to give the correct surrounding boundary condition for the experimental sample. The selected codes for the analysis are MCNPX2.7.0 and SERPENT2.1.21 with intrinsic fuel depletion modeling capabilities and VESTA2.14 in conjunction with MCNPX2.7.0 as a neutron transport solver. In particular, emphasis is given to the comparison of the codes, highlighting the performance in term of accuracy of results and computational cost. The same state-of-the-art cross section data library was used to ensure an adequate comparison. Then, fuel assembly model results are also compared with the equivalent pin cell cases results, to highlight the benefits in using more complex models in terms of gain in accuracy compared to experimental values.