A Two-dimensional Coupled Neutron Transport Method for MOC and S<sub>N</sub> via Boundary Fluence Rate Coupling

Zhang Sifan; Yuan Yuan; Liu Zhouyu; Zhou Xinyu; Cao Liangzhi

doi:10.13832/j.jnpe.2022.02.0009

Volume 43 Issue 2

Apr. 2022

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Article Navigation > Nuclear Power Engineering > 2022 > 43(2): 9-16

Zhang Sifan, Yuan Yuan, Liu Zhouyu, Zhou Xinyu, Cao Liangzhi. A Two-dimensional Coupled Neutron Transport Method for MOC and SN via Boundary Fluence Rate Coupling[J]. Nuclear Power Engineering, 2022, 43(2): 9-16. doi: 10.13832/j.jnpe.2022.02.0009

Citation:

Zhang Sifan, Yuan Yuan, Liu Zhouyu, Zhou Xinyu, Cao Liangzhi. A Two-dimensional Coupled Neutron Transport Method for MOC and S_N via Boundary Fluence Rate Coupling[J]. Nuclear Power Engineering, 2022, 43(2): 9-16. doi: 10.13832/j.jnpe.2022.02.0009

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A Two-dimensional Coupled Neutron Transport Method for MOC and S_N via Boundary Fluence Rate Coupling

doi: 10.13832/j.jnpe.2022.02.0009

1.
School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049, China
2.
China Nuclear Power Engineering Co., Ltd., Beijing, 100840, China

Received Date: 2021-01-11
Rev Recd Date: 2021-07-29
Publish Date: 2022-04-02

Abstract

Abstract

When the method of characteristics (MOC) is used to calculate the out-of-pile detector or some special heavy water moderated light water cooled experimental reactor, the dense characteristics will lead to a large waste of computing resources because the external structural material or moderator area of its active region is too large. To solve this problem, a new coupled transport method based on MOC and discrete ordinate (S_N) nodal method is proposed and implemented in the numerical reactor physics calculation program NECP-X. In this method, the calculation region is divided into MOC domain (including complex structure regions such as active region) and S_N domain (including simple structure regions such as moderator and reflector), and then the grids of the two regions are subject to hybrid sweeping and coupled through the angular fluence rate of the region interface; At the same time, some feasible methods are proposed to reduce the error caused by the fluence rate of the coupling boundary angle. Finally, the calculation effect of the coupling method is verified by the test of two-dimensional C5G7 benchmark task and whole core problem. The numerical results show that the coupling method achieves superior efficiency and accuracy.
- Method of characteristics (MOC),
- Discrete ordinates (S_N) nodal method,
- Hybrid sweeping scheme,
- NECP-X

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References(19)

References

[1]	LIU Z Y, WU H C, CAO L Z, et al. A new three-dimensional method of characteristics for the neutron transport calculation[J]. Annals of Nuclear Energy, 2011, 38(2-3): 447-454. doi: 10.1016/j.anucene.2010.09.021
[2]	刘宙宇. 中子输运方程的三维模块化特征线计算方法研究[D]. 西安: 西安交通大学, 2014.
[3]	COLLINS B, STIMPSON S, KELLEY B W, et al. Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT[J]. Journal of Computational Physics, 2016, 326: 612-628. doi: 10.1016/j.jcp.2016.08.022
[4]	LEE G S, CHO N Z. 2D/1D fusion method solutions of the three-dimensional transport OECD benchmark problem C5G7 MOX[J]. Progress in Nuclear Energy, 2006, 48(5): 410-423. doi: 10.1016/j.pnucene.2006.01.010
[5]	LEE G S. Development of 2D/1D fusion method for three-dimensional whole-core heterogeneous neutron transport calculations[D]. Seoul: Korea Advanced Institute of Science & Technology, 2006.
[6]	CHO J Y, KIM K S, LEE C C, et al. Axial SP_N and radial MOC coupled whole core transport calculation[J]. Journal of Nuclear Science and Technology, 2007, 44(9): 1156-1171. doi: 10.1080/18811248.2007.9711359
[7]	CHO N Z, LEE G S, PARK C J. Fusion of method of characteristics and nodal method for 3-D whole-core transport calculation[J]. Transactions of the American Nuclear Society, 2002, 86: 322-324.
[8]	JOO H G, CHO J Y, KIM K S, et al. Methods and performance of a three-dimensional whole-core transport code DeCART[C]//Proceeding of PHYSOR 2004-The Physics of Fuel Cycles and Advanced Nuclear Systems: Global Developments. Chicago, Illinois: American Nuclear Society, 2004.
[9]	JUNG Y S, SHIM C B, LIM C H, et al. Practical numerical reactor employing direct whole core neutron transport and subchannel thermal/hydraulic solvers[J]. Annals of Nuclear Energy, 2013, 62: 357-374. doi: 10.1016/j.anucene.2013.06.031
[10]	CHEN J, LIU Z Y, ZHAO C, et al. A new high-fidelity neutronics code NECP-X[J]. Annals of Nuclear Energy, 2018, 116: 417-428. doi: 10.1016/j.anucene.2018.02.049
[11]	LIU Z Y, LIANG L, CHEN J, et al. Accuracy and analysis of the iteratively coupling 2D/3D method for the three-dimensional transport calculations[J]. Annals of Nuclear Energy, 2018, 113: 130-138. doi: 10.1016/j.anucene.2017.10.020
[12]	LIU S C, WANG G B, WU G C, et al. Neutronics comparative analysis of plate-type research reactor using deterministic and stochastic methods[J]. Annals of Nuclear Energy, 2015, 79: 133-142. doi: 10.1016/j.anucene.2015.01.027
[13]	WALTERS W F, O'DELL R D. Nodal methods for discrete-ordinates transport problems in (X, Y) geometry[C]//Proceeding of the ANS/ENS Joint Topical Meeting, Mathematical Methods in Nuclear Engineering, Munich, FRG, 1981.
[14]	WALTERS W F. Use of the Chebyshev-Legendre quadrature set in discrete-ordinate codes: LAUR-87-3621[R]. Los Alamos, USA: Los Alamos Scientific Laboratory, 1987.
[15]	YAMAMOTO A, TABUCHI M, SUGIMURA N, et al. Derivation of optimum polar angle quadrature set for the method of characteristics based on approximation error for the bickley function[J]. Journal of Nuclear Science & Technology, 2007, 44(2): 129-136.
[16]	SANCHEZ R, MAO L, SANTANDREA S. Treatment of boundary conditions in trajectory based deterministic transport methods[J]. Nuclear Science and Engineering, 2002, 140(1): 23-50. doi: 10.13182/NSE140-23
[17]	LU H L, WU H C. A nodal SN transport method for three-dimensional triangular-z geometry[J]. Nuclear Engineering and Design, 2007, 237(8): 830-839. doi: 10.1016/j.nucengdes.2006.10.025
[18]	BADRUZZAMAN A. An efficient algorithm for nodal-transport solutions in multidimensional geometry[J]. Nuclear Science and Engineering, 1985, 89(3): 281-290. doi: 10.13182/NSE85-A17550
[19]	SMITH M A, LEWIS E E, NA B C. Benchmark on deterministic transport calculations without spatial homogenization: a 2-D/3-D MOX fuel assembly benchmark: ISBN 92-64-02139-6[R]. Paris: OECD/NEA, 2003