Architecture Design of Two-Fluid Two-Pressure Thermal-Hydraulic System Analysis Code LOCUST 2.0
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摘要: 中国广核集团(CGN)研发了一款新的热工水力系统分析软件LOCUST 2.0,它采用两流体两场双压力七方程模型,从理论上保证了守恒方程的严格适定性。LOCUST 2.0的应用对象为华龙一号失水事故类的安全分析,目前已初步完成代码开发。本文简要描述了该软件的守恒方程、数值算法、软件功能、源代码架构设计;给出了6个典型案例的分析计算,结果表明该软件能够合理预测热工水力进程,性能良好。
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关键词:
- LOCUST 2.0 /
- 双压力模型 /
- 系统分析软件 /
- 软件架构设计
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.-
Key words:
- LOCUST 2.0 /
- Two-pressure models /
- System analysis code /
- Code architecture design
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表 1 不同系统程序结构的对比
Table 1. Structure Comparison of Different System Codes
序号 程序结构 优点 缺点 典型代表 1 两场五方程模型
(一个混合物动量方程+滑速比模型)无条件适定;计算量相对更少 两相动力学描述偏于简单,对于
非泡状流型的模拟能力不足RETRAN-02 2 两场六方程模型 两相动力学描述较精细 存在不适定区域(可通过虚拟质量力等进行改善) RELAP5、
TRACE3 两场七方程模型 无条件适定;两相动力学描述较精细 计算量相对六方程模型更大;成熟性尚有限 RELAP7、
LOCUST 2.0 -
[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.