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http://dx.doi.org/10.1016/j.net.2020.06.016

Validation of UNIST Monte Carlo code MCS for criticality safety calculations with burnup credit through MOX criticality benchmark problems  

Ta, Duy Long (Department of Nuclear Engineering, Hanyang University)
Hong, Ser Gi (Department of Nuclear Engineering, Hanyang University)
Lee, Deokjung (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology)
Publication Information
Nuclear Engineering and Technology / v.53, no.1, 2021 , pp. 19-29 More about this Journal
Abstract
This paper presents the validation of the MCS code for critical safety analysis with burnup credit for the spent fuel casks. The validation process in this work considers five critical benchmark problem sets, which consist of total 80 critical experiments having MOX fuels from the International Criticality Safety Benchmark Evaluation Project (ICSBEP). The similarity analysis with the use of sensitivity and uncertainty tool TSUNAMI in SCALE was used to determine the applicable benchmark experiments corresponding to each spent fuel cask model and then the Upper Safety Limits (USLs) except for the isotopic validation were evaluated following the guidance from NUREG/CR-6698. The validation process in this work was also performed with the MCNP6 for comparison with the results using MCS calculations. The results of this work showed the consistence between MCS and MCNP6 for the MOX fueled criticality benchmarks, thus proving the reliability of the MCS calculations.
Keywords
Spent fuel cask; Critical experiment; MCS; Validation; MCNP6;
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1 Y. Gao, N.J. McFerran, A. Enqvist, J.E. Tulenko, J.E. Baciak, Dry cask radiation shielding validation and estimation of cask surface dose rate with MAVRIC during long-term storage, Ann. Nucl. Energy 140 (2020) 107145.   DOI
2 A. Rimpler, M. Borst, D. Seifarth, Neutron measurement around a TN-85 type storage cask with high-active waste, Radiat. Meas. 45 (2010) 1290-1292.   DOI
3 H. Lee, W. Kim, P. Zhang, M. Lemaire, A. Khassenov, J. Yu, Y. Jo, J. Park, D. Lee, MCS - a Monte Carlo particle transport code for large-scale power reactor analysis, Ann. Nucl. Energy 139 (2020) 107276.   DOI
4 International Handbook of Evaluated Criticality Safety Benchmark Experiments, 2011. NEA/NSC/DOC(95)03.
5 Los Alamos National Laboratory, MCNP6 User's Manual, LA-CP-13-00634, 2013.
6 Transnuclear, Inc, TN-32 dry storage cask system safety evaluation report. https://www.nrc.gov/docs/ML0036/ML003696918.pdf.
7 Ju Chan Lee, et al., Usage Inspection of KN-12 Spent Fuel Transport Cask, KAERI/TR-3402/2007, Korea Atomic Energy Research Institute, 2007.
8 S.S. Shapiro, R.S. Francia, An approximate analysis of variance test for normality, J. Am. Stat. Assoc. 67 (1972) 215-216.   DOI
9 P. Royston, A pocket-calculator algorithm for the shapiro-francia test for non-normality: an application to medicine, Stat. Med. 12 (1993) 181-194.   DOI
10 R. Poskas, V. Simonis, H. Jouhara, P. Poskas, Modeling of decay heat removal from CONSTOR RMBK-1500 casks during long-term storage of spent nuclear fuel, Energy 170 (2019) 978-985.   DOI
11 H.S. Yoo, S.H. Yoo, E.S. Kim, Heat transfer enhancement in dry cask storage for nuclear spent fuel using additive high density inert gas, Ann. Nucl. Energy 132 (2019) 108-118.   DOI
12 D. Price, M.I. Radaideh, D. O'Grady, T. Kozlowski, Advanced BWR criticality safety part II: cask criticality, burnup credit, sensitivity, and uncertainty analyses, Prog. Nucl. Energy 115 (2019) 126-139.   DOI
13 A. Mohammadi, M. Hassanzadeh, M. Gharib, Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors, Appl. Radiat. Isot. 108 (2016) 129-132.   DOI
14 H. Spilker, M. Peehs, H.P. Dyck, G. Kaspar, K. Nissen, Spent LWR fuel dry storage in large transport and storage casks after extended burnup, J. Nucl. Mater. 250 (1997) 63-74.   DOI
15 H.J. Yun, D.Y. Kim, K.H. Park, S.G. Hong, A criticality analysis of the GBC-32 dry storage cask with hanbit nuclear power plant unit 3 fuel assemblies from the viewpoint of burnup credit, Nuclear Engineering and Technology 48 (2016) 624-634.   DOI
16 H. Yun, K. Park, W. Choi, S.G. Hong, An effective evaluation of depletion uncertainty for a GBC-32 dry storage cask with PLUS7 fuel assemblies using the Monte Carlo uncertainty sampling method, Ann. Nucl. Energy 110 (2017) 679-691.   DOI
17 K.C. Chen, K. Ting, Y.C. Li, Y.Y. Chen, W.K. Cheng, W.C. Chen, C.T. Liu, A study of the probabilistic risk assessment to the dry storage system of spent nuclear fuel, Int. J. Pres. Ves. Pip. 87 (2010) 17-25.   DOI
18 G. Pugliese, R.L. Frano, G. Forasassi, Spent fuel transport cask thermal evaluation under normal and accident conditions, Nucl. Eng. Des. 240 (2010) 1699-1706.   DOI
19 S.E. Smith, X. Sun, C.A. Ford, A.W. Fentiman, MCNP simulation of neutron energy spectra for a TN-32 dry shielded container, Ann. Nucl. Energy 35 (2008) 1296-1300.   DOI
20 C. Greulich, C. Hughes, Y. Gao, A. Enqvist, J. Baciak, High energy neutron transmission analysis of dry cask storage, Nucl. Instrum. Methods Phys. Res. A 874 (2017) 5.   DOI
21 Oak Ridge National Laboratory, SCALE: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design, June 2011. ORNL/TN-2005/39, Version 6.1.
22 J. Jang, W. Kim, S. Jeong, E. Jeong, J. Park, M. Lemaire, H. Lee, Y. Jo, P. Zhang, D. Lee, Validation of UNIST Monte Carlo code MCS for criticality safety analysis of PWR spent fuel pool and storage cask, Ann. Nucl. Energy 114 (2018) 495-509.   DOI
23 D.E. Mueller, W.J. Marshall, D.G. Bowen, J.C. Wagner, Bias Estimates Used in Lieu of Validation of Fission Products and Minor Actinides in MCNP Keff Calculations for PWR Burnup Credit Casks, NUREG/CR-7205, U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, Oak Ridge (TN), 2015. ORNL/TM-2012/544.
24 M. Lemaire, H. Lee, B. Ebiwonjumi, C. Kong, W. Kim, Y. Jo, J. Park, D. Lee, Verification of photon transport capability of UNIST Monte Carlo code MCS, Comput. Phys. Commun. 231 (2018) 1-18.   DOI
25 J.M. Scaglione, D.E. Mueller, J.C. Wagner, W.J. Marshall, An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Criticality (Keff) Predictions, NUREG/CR-7109, ORNL/TM-2011/514, U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, Oak Ridge (TN), 2012.
26 M.I. Radaideh, D. Price, T. Kozlowski, Criticality and uncertainty assessment of assembly misloading in BWR transportation cask, Ann. Nucl. Energy 113 (2018) 1-14.   DOI
27 J.C. Dean, R.W. Tayloe Jr., Guide for Validation of Nuclear Criticality Safety Calculational Methodology, NUREG/CR-6698, U.S. Nuclear Regulatory Commission, 2001.
28 Standard Review Plan for Spent Fuel Dry Storage Systems at a General License Facility, NUREG-1536, Revision 1, U.S. Nuclear Regulatory Commission, 2010.
29 S.H. Chung, et al., Evaluation of the KN-12 spent fuel shipping cask, in: Proceedings of the Korean Nuclear Society Spring Meeting, May 2001. Cheju, Korea.
30 G. Radulescu, I.C. Gauld, G. Ilas, J.C. Wagner, An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses - Isotopic Composition Predictions, NUREG/CR-7108, US Nuclear Commission, 2012.