Architecture Design of Two-Fluid Two-Pressure Thermal-Hydraulic System Analysis Code LOCUST 2.0

Xu Caihong; Yuan Hongsheng; Liu Huannan; Ju Zhongyun; Li Changying; Wang Ting; Li Jinggang

doi:10.13832/j.jnpe.2023.06.0086

Volume 44 Issue 6

Dec. 2023

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Article Navigation > Nuclear Power Engineering > 2023 > 44(6): 86-94

Xu Caihong, Yuan Hongsheng, Liu Huannan, Ju Zhongyun, Li Changying, Wang Ting, Li Jinggang. Architecture Design of Two-Fluid Two-Pressure Thermal-Hydraulic System Analysis Code LOCUST 2.0[J]. Nuclear Power Engineering, 2023, 44(6): 86-94. doi: 10.13832/j.jnpe.2023.06.0086

Citation:

Xu Caihong, Yuan Hongsheng, Liu Huannan, Ju Zhongyun, Li Changying, Wang Ting, Li Jinggang. Architecture Design of Two-Fluid Two-Pressure Thermal-Hydraulic System Analysis Code LOCUST 2.0[J]. Nuclear Power Engineering, 2023, 44(6): 86-94. doi: 10.13832/j.jnpe.2023.06.0086

Citation:

Xu Caihong, Yuan Hongsheng, Liu Huannan, Ju Zhongyun, Li Changying, Wang Ting, Li Jinggang. Architecture Design of Two-Fluid Two-Pressure Thermal-Hydraulic System Analysis Code LOCUST 2.0[J]. Nuclear Power Engineering, 2023, 44(6): 86-94. doi: 10.13832/j.jnpe.2023.06.0086

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Architecture Design of Two-Fluid Two-Pressure Thermal-Hydraulic System Analysis Code LOCUST 2.0

doi: 10.13832/j.jnpe.2023.06.0086

China Nuclear Power Technology Research Institute Co., Ltd., Shengzhen, Guangdong, 518000, China

Received Date: 2022-12-20
Rev Recd Date: 2023-06-26

Available Online: 2023-12-11

Publish Date: 2023-12-15

Abstract

Abstract

China General Nuclear Power Corporation (CGN) is developing a new thermal hydraulic system analysis code LOCUST 2.0, which uses two-fluid two-field two-pressure seven-equation models, theoretically ensuring the strict well-posedness of conservation equations. The design application of LOCUST 2.0 is the safety analysis of loss-of-coolant accidents (LOCAs) for HPR1000, and its source codes have been preliminarily completed at present. In this paper, the conservation equations, numerical methods, software functions, and code architecture design are briefed introduced. Meanwhile, the calculation results of six typical test cases are given, and the results show that the code reasonably predicts the thermal-hydraulic processes and has good performance.
- LOCUST 2.0,
- Two-pressure models,
- System analysis code,
- Code architecture design

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References(25)

References

[1]	RETTIG W H, JAYNE G A, MOORE K V, et al. RELAP3: a computer program for reactor blowdown analysis: IN-1321[R]. Idaho Falls: Idaho Nuclear Corp. , 1970.
[2]	GRS. ATLET models and methods: GRS-P-1/Vol. 4, Rev. 5[Z]. 2019.
[3]	U.S. Nuclear Regulatory Commission. RELAP5/MOD3 code manual volume 1: code structure, system models, and solution methods: NUREG/CR-5535[R]. Washington: U.S. Nuclear Regulatory Commission, 1995.
[4]	Nuclear Regulatory Commission. TRACE V5.0 Theory manual: field equations, solution methods, and physical models: ML120060218[R]. Washington: U. S. Nuclear Regulatory Commission, 2008.
[5]	BESTION D. The physical closure laws in the CATHARE code[J]. Nuclear Engineering and Design, 1990, 124(3): 229-245. doi: 10.1016/0029-5493(90)90294-8
[6]	BERRY R A, PETERSON J W, ZHANG H, et al. RELAP-7 theory manual: INL/EXT-14-31366-Rev003[R]. Idaho Falls: Idaho National Lab, 2018.
[7]	EMONOT P, SOUYRI A, GANDRILLE J L, et al. CATHARE-3: a new system code for thermal-hydraulics in the context of the NEPTUNE project[J]. Nuclear Engineering and Design, 2011, 241(11): 4476-4481.
[8]	HA S J, PARK C E, KIM K D, et al. Development of the SPACE code for nuclear power plants[J]. Nuclear Engineering and Technology, 2011, 43(1): 45-62. doi: 10.5516/NET.2011.43.1.045
[9]	徐财红. 两相流热工水力系统分析软件LOCUST-1.2开发概述[C]. 重庆: 中国核学会核反应堆热工流体力学分会第一届学术年会, 2021.
[10]	单建强, 廖承奎, 苟军利, 等. 压水堆核电厂瞬态安全数值分析方法[M]. 西安: 西安交通大学出版社, 2016: 140-156.
[11]	BERRY R A, SAUREL R, LEMETAYER O, et al. The discrete equation method (DEM) for fully compressible, two-phase flows in ducts of spatially varying cross-section[J]. Nuclear Engineering and Design, 2010, 240(11): 3797-3818. doi: 10.1016/j.nucengdes.2010.08.003
[12]	CARISON K E, RIEMKE R A, ROUHANL S Z, et al. RELAP5/MOD3 Code manual, volume 3: developmental assessment problems (DRAFT): NUREG/CR-5535[R]. Washington: U. S. Nuclear Regulatory Commission, 1990.
[13]	U. S. Nuclear Regulatory Commission. TRACE V5.0 assessment manual: appendix a: fundamental validation cases: ML120060187[R]. Washington: U. S. Nuclear Regulatory Commission, 2008.
[14]	KATAOKA I, ISHII M. Drift flux model for large diameter pipe and new correlation for pool void fraction[J]. International Journal of Heat and Mass Transfer, 1987, 30(9): 1927-1939. doi: 10.1016/0017-9310(87)90251-1
[15]	HIBIKI T, ISHII M. One-dimensional drift–flux model for two-phase flow in a large diameter pipe[J]. International Journal of Heat and Mass Transfer, 2003, 46(10): 1773-1790. doi: 10.1016/S0017-9310(02)00473-8
[16]	NYLUND O, BECKER K M, EKLUND R, et al. Hydrodynamic and heat transfer measurements on a full-scale simulated 36-rod marviken fuel element with uniform heat flux distribution: FRIGG-2[R]. Stockholm: Aktiebolaget Atomenergi, 1968.
[17]	HENRY R E, FAUSKE H K. The two-phase critical flow of one-component mixtures in nozzles, orifices, and short tubes[J]. Journal of Heat Transfer, 1971, 93(2): 179-187. doi: 10.1115/1.3449782
[18]	MOODY F J. Maximum flow rate of a single component, two-phase mixture[J]. Journal of Heat Transfer, 1965, 87(1): 134-141. doi: 10.1115/1.3689029
[19]	U.S. Nuclear Regulatory Commission. The Marviken full scale critical flow tests: summary report: NUREG/CR-2671[R]. Washington: U.S. Nuclear Regulatory Commission, 1982.
[20]	YODER G L, MORRIS D G, MULLINS C B, et al. Dispersed-flow film boiling in rod-bundle geometry: steady-state heat-transfer data and correlation comparisons: NUREG/CR-2435[R]. Oak Ridge: Oak Ridge National Lab. , 1982.
[21]	SHAH M M. Unified correlation for heat transfer during boiling in plain mini/micro and conventional channels[J]. International Journal of Refrigeration, 2017, 74: 606-626. doi: 10.1016/j.ijrefrig.2016.11.023
[22]	GROENEVELD D C, SHAN J Q, VASIĆ A Z, et al. The 2006 CHF look-up table[J]. Nuclear Engineering and Design, 2007, 237(15-17): 1909-1922. doi: 10.1016/j.nucengdes.2007.02.014
[23]	BJORNARD T A, GRIFFITH P. PWR blowdown heat transfer[C]. U.S.: Proceedings of the Winter Annual Meeting of the American Society of Mechanical Engineers. New York: American Society of Mechanical Engineers, 1977.
[24]	GROENEVELD D C, LEUNG L K H, VASIC’ A Z, et al. A look-up table for fully developed film-boiling heat transfer[J]. Nuclear Engineering and Design, 2003, 225(1): 83-97. doi: 10.1016/S0029-5493(03)00149-3
[25]	LOFTUS M J, HOCHREITER L E, CONWAY C E, et al. PWR FLECHT SEASET unblocked bundle, forced and gravity reflood task data report. Volume 1: NUREG/CR-1532[R]. Pittsburgh: Westinghouse Electric Corp. , 1981.