Application Research on Whole-Core Three-Dimensional Space-Time Kinetics Neutron Transport Code SAAFCGSN

Jiang Duoyu; Xu Peng; Jiang Xinbiao; Hu Tianliang; Wang Lipeng; Cao Lu; Li Da

doi:10.13832/j.jnpe.2025.01.0013

Volume 46 Issue 1

Feb. 2025

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Jiang Duoyu, Xu Peng, Jiang Xinbiao, Hu Tianliang, Wang Lipeng, Cao Lu, Li Da. Application Research on Whole-Core Three-Dimensional Space-Time Kinetics Neutron Transport Code SAAFCGSN[J]. Nuclear Power Engineering, 2025, 46(1): 13-23. doi: 10.13832/j.jnpe.2025.01.0013

Citation:

Jiang Duoyu, Xu Peng, Jiang Xinbiao, Hu Tianliang, Wang Lipeng, Cao Lu, Li Da. Application Research on Whole-Core Three-Dimensional Space-Time Kinetics Neutron Transport Code SAAFCGSN[J]. Nuclear Power Engineering, 2025, 46(1): 13-23. doi: 10.13832/j.jnpe.2025.01.0013

Citation:

Jiang Duoyu, Xu Peng, Jiang Xinbiao, Hu Tianliang, Wang Lipeng, Cao Lu, Li Da. Application Research on Whole-Core Three-Dimensional Space-Time Kinetics Neutron Transport Code SAAFCGSN[J]. Nuclear Power Engineering, 2025, 46(1): 13-23. doi: 10.13832/j.jnpe.2025.01.0013

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Application Research on Whole-Core Three-Dimensional Space-Time Kinetics Neutron Transport Code SAAFCGSN

doi: 10.13832/j.jnpe.2025.01.0013

1.
Rocket Force University of Engineering, Xi’an, 710025, China
2.
State Key Laboratory of Intense Pulsed Radiation Simulation and Effect, Northwest Institute of Nuclear Technology, Xi’an, 710024, China

Received Date: 2024-05-13
Rev Recd Date: 2024-06-10
Publish Date: 2025-02-15

Abstract

Abstract

In response to the evolving demands of advanced small reactor technologies, there is an increasing requirement for three-dimensional whole core transport calculations in reactor physics numerical simulation codes. This paper presents the development of a three-dimensional space-time kinetics neutron transport code, SAAFCGSN, based on the MOOSE platform. The code implements spatial variable discretization using the finite element method, and solves the steady-state and transient neutron transport equations as well as the delayed neutron precursor equations using a residual form approach. The Jacobian-Free Newton-Krylov (JFNK) method is employed to avoid the direct computation of the Jacobian matrix, thereby enhancing computational speed and reducing memory usage. To evaluate the transient computational capability of the code, we validated its reliability using the OECD/NEA C5G7-TD benchmark series and conducted a comparative analysis with high-fidelity deterministic neutron transport codes and Monte Carlo simulations. The study demonstrates that the SAAFCGSN code achieves high computational accuracy, effectively manages the cusping effect of control rods, and obtains detailed neutron flux distributions within the three-dimensional core. It meets the steady-state and transient neutronic calculation requirements for advanced small reactors.
- SAAFCGSN,
- MOOSE platform,
- Space-time kinetics,
- C5G7-TD benchmark question,
- Cusping effect

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References(26)

References

[1]	吴宏春,杨红义,曹良志,等. 金属冷却快堆关键分析软件的现状与展望[J]. 现代应用物理,2021, 12(1): 010201.
[2]	卢琳龙,李兵,高辉. 含氢反射层对金属活性区快中子脉冲反应堆特性参数的影响[J]. 现代应用物理,2022, 13(2): 020205.
[3]	CARLSON B G. Solution of the transport equation by S_n approximations: LA-1599[R]. Los Alamos: Los Alamos Scientific Laboratory, 1955.
[4]	STACEY W M. Nuclear reactor physics[M]. 2nd ed. Weinheim: WILEY-VCH Verlag GmbH & Co. KGaA, 2007: 338-353.
[5]	JIANG D Y, XU P, HU T L, et al. Coupled Monte Carlo and thermal-hydraulics modeling for the three-dimensional steady-state analysis of the Xi’an Pulsed Reactor[J]. Energies, 2023, 16(16): 6046. doi: 10.3390/en16166046
[6]	MOREL J E, MCGHEE J M. A self-adjoint angular flux equation[J]. Nuclear Science and Engineering, 1999, 132(3): 312-325. doi: 10.13182/NSE132-312
[7]	POMRANING G C, CLARK JR M. The variational method applied to the monoenergetic Boltzmann equation. Part II[J]. Nuclear Science and Engineering, 1963, 16(2): 155-164. doi: 10.13182/NSE63-A26495
[8]	ACKROYD R T. Least-squares derivation of extremum and weighted-residual methods for equations of reactor physics—I. The first-order Boltzmann equation and a first-order initial-value equation[J]. Annals of Nuclear Energy, 1983, 10(2): 65-99. doi: 10.1016/0306-4549(83)90011-7
[9]	LISCUM-POWELL J L. Finite element numerical solution of a self-adjoint transport equation for coupled electron-photon problems[D]. New Mexico: The University of New Mexico, 2000.
[10]	曹良志,吴宏春,周永强. 非结构几何下二阶自共轭中子输运方程中的简化球谐函数方法研究[J]. 核动力工程,2006, 27(3): 6-10. doi: 10.3969/j.issn.0258-0926.2006.03.002
[11]	叶青,吴宏春. R-Z几何中子输运方程的球谐函数方法[J]. 核动力工程,2008, 29(4): 19-23.
[12]	SCHUNERT S, WANG Y Q, MARTINEAU R, et al. A new mathematical adjoint for the modified SAAF-S_N equations[J]. Annals of Nuclear Energy, 2015, 75: 340-352. doi: 10.1016/j.anucene.2014.08.028
[13]	WANG Y Q, DEHART M D, GASTON D R, et al. Convergence study of Rattlesnake solutions for the two-dimensional C5G7 MOX benchmark[C]//Proceedings of Joint International Conference on Mathematics and Computation. Idaho Falls: Idaho National Lab. , 2015.
[14]	WANG Y Q, SCHUNERT S, LABOURÉ V. Rattlesnake theory manual: INL/EXT-17-42103[R]. Idaho Falls: Idaho National Laboratory, 2018.
[15]	LATIMER C, KÓPHÁZI J, EATON M D, et al. A geometry conforming isogeometric method for the self-adjoint angular flux (SAAF) form of the neutron transport equation with a discrete ordinate (S_N) angular discretisation[J]. Annals of Nuclear Energy, 2020, 136: 107049. doi: 10.1016/j.anucene.2019.107049
[16]	GIUDICELLI G, LINDSAY A, HARBOUR L, et al. 3.0 - MOOSE: enabling massively parallel multiphysics simulations[J]. SoftwareX, 2024, 26: 101690. doi: 10.1016/j.softx.2024.101690
[17]	KNOLL D A. KEYES D E. Jacobian-free Newton-Krylov methods: a survey of approaches and applications[J]. Journal of Computational Physics, 2004, 193(2): 357-397. doi: 10.1016/j.jcp.2003.08.010
[18]	JOO H S. Resolution of the control rod cusping problem for nodal methods[D]. Cambridge: Massachusetts Institute of Technology, 1984.
[19]	HOU J, IVANOV K N, BOYARINOV V F, et al. OECD/NEA benchmark for time-dependent neutron transport calculations without spatial homogenization[J]. Nuclear Engineering and Design, 2017, 317: 177-189. doi: 10.1016/j.nucengdes.2017.02.008
[20]	JIANG D Y, XU P, HU T L, et al. Transient multi-physics coupling analysis of the Xi'an Pulsed Reactor under pulsed condition[J]. Annals of Nuclear Energy, 2024, 200: 110379. doi: 10.1016/j.anucene.2024.110379
[21]	TAO S J, XU Y L. Neutron transport analysis of C5G7-TD benchmark with PANDAS-MOC[J]. Annals of Nuclear Energy, 2022, 169: 108966. doi: 10.1016/j.anucene.2022.108966
[22]	王博,刘宙宇,陈军,等. 基于NECP-X程序的C5G7-TD系列基准题的计算与分析[J]. 核动力工程,2020, 41(3): 24-30.
[23]	张旻婉,刘宙宇,温兴坚,等. 用NECP-X程序计算与分析VERA 9^#基准题[J]. 现代应用物理,2021, 12(1): 010212.
[24]	DEHART M D, MAUSOLFF Z, WEEMS Z, et al. Preliminary results for the OECD/NEA time dependent benchmark using Rattlesnake, Rattlesnake-IQS and TDKENO: INL/EXT-16-39723[R]. Idaho Falls: Idaho National Laboratory, 2016.
[25]	GUO X Y, SHANG X T, SONG J, et al. Kinetic methods in Monte Carlo code RMC and its implementation to C5G7-TD benchmark[J]. Annals of Nuclear Energy, 2021, 151: 107864. doi: 10.1016/j.anucene.2020.107864
[26]	SHEN Q C, WANG Y R, JABAAY D, et al. Transient analysis of C5G7-TD benchmark with MPACT[J]. Annals of Nuclear Energy, 2019, 125: 107-120. doi: 10.1016/j.anucene.2018.10.049