Browse > Article
http://dx.doi.org/10.1016/j.net.2021.10.045

Verification of SARAX code system in the reactor core transient calculation based on the simplified EBR-II benchmark  

Jia, Xiaoqian (School of Nuclear Science and Technology, Xi'an Jiaotong University)
Zheng, Youqi (School of Nuclear Science and Technology, Xi'an Jiaotong University)
Du, Xianna (School of Nuclear Science and Technology, Xi'an Jiaotong University)
Wang, Yongping (School of Nuclear Science and Technology, Xi'an Jiaotong University)
Chen, Jianda (School of Nuclear Science and Technology, Xi'an Jiaotong University)
Publication Information
Nuclear Engineering and Technology / v.54, no.5, 2022 , pp. 1813-1824 More about this Journal
Abstract
This paper shows the verification work of SARAX code system in the reactor core transient calculation based on the simplified EBR-II Benchmark. The SARAX code system is an analysis package developed by Xi'an Jiaotong University and aims at the advanced reactor R&D. In this work, a neutron-photon coupled power calculation model and a spatial-dependent reactivity feedback model were introduced. To verify the models used in SARAX, the EBR-II SHRT-45R test was simplified to an ULOF transient with an input flowrate change curve by fitting from reference. With the neutron-photon coupled power calculation model, SARAX gave close results in both power fraction and peak power prediction to the reference results. The location of the hottest assembly from SARAX and reference are the same and the relative power deviation of the hottest assembly is 2.6%. As for transient analysis, compared with experimental results and other calculated results, SARAX presents coincident results both in trend and absolute value. The minimum value of core net reactivity during the transient agreed well with the reported results, which ranged from -0.3$ to -0.35$. The results verify the models in SARAX, which are correct and able to simulate the in-core transient with reliable accuracy.
Keywords
Sodium-cooled fast reactor; Transient analysis; ULOF transient;
Citations & Related Records
Times Cited By KSCI : 1  (Citation Analysis)
연도 인용수 순위
1 X. Jia, Y. Zheng, X. Du, M. He, Z. Zhai, Application of SARAX code system in transient analysis of sodium-cooled fast reactor, Atomic Energy Sci. Technol. 53 (7) (2019) 1195-1201.
2 T. Fei, B. Vezzoni, M. Marchetti, W. van Rooijen, Y. Zhang, Neutronics Benchmark for EBR-II Shutdown Heat Removal Test SHRT-45R, Argonne National Lab.(ANL), Argonne, IL (United States), 2016.
3 H. Lu, H. Wu, A nodal SN transport method for three-dimensional triangular-z geometry, Nucl. Eng. Des. 237 (2007) 830-839.   DOI
4 Y. Zheng, H. Wu, X. Du, et al., Improvement and verification of the fast reactor neutronics code NECP-SARAX, Qiangjiguang Yu Lizishu/High Power Laser Part Beams 29 (5) (2017).
5 A.E. Waltar, D.R. Todd, P.V. Tsvetkov, Fast Spectrum Reactors, Springer, 2011.
6 Y. Zheng, L. Qiao, Z. Zhai, X. Du, Z. Xu, SARAX: a new code for fast reactor analysis part II: verification, validation and uncertainty quantification, Nucl. Eng. Des. 331 (2018) 41-53, https://doi.org/10.1016/j.nucengdes.2018.02.033.   DOI
7 L. Wei, Y. Zheng, H. Wu, Improvement of few-group cross-section generation in fast reactor analysis system SARAX, Ann. Nucl. Energy 132 (2019) 149-160.   DOI
8 S. Zhou, H. Wu, L. Cao, Y. Zheng, K. Huang, M. He, X. Li, LAVENDER: a steady-state core analysis code for design studies of accelerator driven subcritical reactors, Nucl. Eng. Des. 278 (2014) 434-444.   DOI
9 Jianda Chen, Research on Fine Power Calculation Method and Code Development and Application for the Core with Hexagonal Assembly[D], Xi'an Jiaotong University, 2021.
10 T. Fanning, A. Brunett, T. Sumner, The SAS4A/SASSYS-1 Safety Analysis Code System, version 5, Argonne National Lab.(ANL), Argonne, IL (United States), 2017.
11 IAEA, Benchmark Analysis of EBR-II Shutdown Heat Removal Tests, IAEA, Vienna, 2017.
12 S. Sumner, Wei Tyler, Y. Thomas, Benchmark Specifications and Data Requirements for EBR-II Shutdown Heat Removal Tests SHRT-17 and SHRT-45R, 2012.
13 P. Chad, Experimental Breeder Reactor II Benchmark Evaluation, Idaho State University, 2017.
14 R. Macfarlane, D.W. Muir, R. Boicourt, A.C. Kahler III, J.L. Conlin, The NJOY Nuclear Data Processing System, Version 2016, Los Alamos National Lab.(LANL), Los Alamos, NM (United States), 2017.
15 B. Faure, P. Archier, J.-F. Vidal, J.M. Palau, L. Buiron, Neutronic calculation of an axially heterogeneous ASTRID fuel assembly with APOLLO3®: analysis of biases and foreseen improvements, Ann. Nucl. Energy 115 (2018) 88-104, https://doi.org/10.1016/j.anucene.2017.12.035.   DOI
16 H. Yu, X.Y. Yang, K.Y. Zhou, X.L. Chen, D.S. Hu, Research of criticality test for China experimental fast reactor, At. Energy Sci. Technol. 47 (2013) 62-65.   DOI
17 X.N. Du, L.Z. Cao, Y.Q. Zheng, H. Wu, in: Korea Jeju (Ed.), Developments of the Sodium Fast Reactor Analysis Code SARAX: Methods and Verification, 2017, pp. 16-20.