基于统一框架的多物理耦合方案研究与平台开发

钱冠华; 于涛; 杨涛; 赵亚楠; 赵鹏程

doi:10.13832/j.jnpe.2022.06.0051

基于统一框架的多物理耦合方案研究与平台开发

doi: 10.13832/j.jnpe.2022.06.0051

钱冠华^{1, 2,},
于涛^{1, 2, ,},
杨涛^{1, 2},
赵亚楠^{1, 2},
赵鹏程^{1, 2}

1.
南华大学核科学技术学院，湖南衡阳，421001
2.
数字化反应堆湖南省工程技术研究中心，湖南衡阳，421001

基金项目: 国家自然科学基金（11875162）；湖南省科技创新团队项目资助（2020RC4053）；湖南省教育厅优秀青年项目（20B490）

详细信息

作者简介:
钱冠华（1995—），男，博士研究生，现从事反应堆热工流体方向研究，E-mail: qianguanhua@stu.usc.edu.cn

通讯作者:
于　涛，E-mail: yutao29@sina.com

中图分类号: TL365
计量
- 文章访问数: 363
- HTML全文浏览量: 151
- PDF下载量: 77
- 被引次数: 0
出版历程
- 收稿日期: 2021-11-19
- 修回日期: 2022-01-04
- 刊出日期: 2022-12-14

Research and Platform Development of Multi-physical Coupling Scheme Based on Unified Framework

Qian Guanhua^{1, 2
,},
Yu Tao^{1, 2
, ,},
Yang Tao^{1, 2},
Zhao Yanan^{1, 2},
Zhao Pengcheng^{1, 2}

1.
School of Nuclear Science and Technology, University of South China, Hengyang, Hunan, 421001, China
2.
Research Center for Digital Nuclear Reactor Engineering and Technology of Hunan Province, University of South China, Hengyang, Hunan, 421001, China

摘要

摘要: 为实现反应堆多物理、多过程、高保真数值计算，捕捉堆芯内部更真实的物理学行为，本文深入研究了多物理程序耦合方案，并基于上层监控架构、串行计算模式、网格一一映射的显式耦合方案，依托开源集成平台SALOME、通用平台接口ICoCo、三维堆芯中子学程序ADPRES和系统热工水力程序RELAP5搭建了基于统一框架的多物理耦合平台。经NEACRP-L-335压水堆弹棒基准题验证表明，耦合平台计算结果与基准例题吻合良好，耦合平台在功率峰捕获上更加准确，可释放部分安全裕量；对瞬态末各参数的计算结果也有足够高的精度，证明了耦合平台可对反应堆多物理、多过程耦合工况进行更精细、更深入的数值计算与安全分析。
- 数值反应堆 /
- 多物理耦合 /
- 统一耦合框架 /
- 开源集成平台 /
- 通用平台接口
Abstract: In order to realize the multi-physics, multi-process and high-fidelity numerical calculation of the reactor and capture the more real physical behavior in the reactor core, in this paper, the coupling scheme of multiple physical programs is deeply studied, and based on the upper monitoring architecture, serial computing mode, and the explicit coupling scheme of grid one-to-one mapping, a unified framework-based multi-physical coupling platform is built by relying on the open source integration platform SALOME, the common platform interface ICoCo, the three-dimensional core neutronics program ADPRES, and the system thermal hydraulic program RELAP5. The verification of NEACRP-L-335PWR PWR rod ejection benchmark task shows that the calculation results of the coupling platform are in good agreement with the benchmark task, and the coupling platform is more accurate in power peak capture, which can release part of the safety margin; the calculation results of the parameters at the end of the transient are also accurate enough, which proves that the coupling platform can carry out more precise and in-depth numerical calculation and safety analysis of the reactor multi-physical and multi-process coupling conditions.
- Virtual nuclear reactor /
- Multi-physics coupling /
- Unified coupling framework /
- Open-source coupling platform /
- Common platform interface

HTML全文

图 1 两种架构的示意图

Figure 1. Schematic Diagram of Two Architectures

下载: 全尺寸图片幻灯片

图 2 两种运行模式的逻辑

Figure 2. Logic of Two Operation Modes

下载: 全尺寸图片幻灯片

图 3 网格映射的两种方式

Figure 3. Two Methods of Mesh Mapping

下载: 全尺寸图片幻灯片

图 4 时间耦合的三种方式

Figure 4. Three Methods of Temporal Coupling

下载: 全尺寸图片幻灯片

图 5 SALOME主要功能模块

Figure 5. Main Functional Modules of SALOME

下载: 全尺寸图片幻灯片

图 6 MED的结构

MPI—消息传递接口

Figure 6. MED Architecture

下载: 全尺寸图片幻灯片

图 7 ICoCo计算流程及组件功能

Figure 7. Calculation Process and Component Functions of ICoCo　　　　　　

下载: 全尺寸图片幻灯片

图 8 ADPRES-RELAP5耦合计算流程

Figure 8. Coupling Calculation Process of ADPRES-RELAP5

下载: 全尺寸图片幻灯片

图 9 堆芯径向几何

Figure 9. Radial Geometry of Core

下载: 全尺寸图片幻灯片

图 10 ADPRES堆芯径向建模

Figure 10. Core Radial Modeling in ADPRES

下载: 全尺寸图片幻灯片

图 11 RELAP5流道建模

Figure 11. Modeling of Flow Channel in RELAP5

下载: 全尺寸图片幻灯片

图 12 相对功率因子随时间的变化趋势

Figure 12. Change Trend of Relative Power Factor with Time

下载: 全尺寸图片幻灯片

图 13 平均芯块温度随时间的变化趋势

Figure 13. Change Trend of Average Pellet Temperature with Time

下载: 全尺寸图片幻灯片

图 14 最大芯块温度随时间的变化趋势

Figure 14. Change Trend of Maximum Pellet Temperature with Time

下载: 全尺寸图片幻灯片

图 15 A2工况堆芯径向归一化功率分布

Figure 15. Radial Normalized Power Distribution of Core in A2 Case

下载: 全尺寸图片幻灯片

表 1 PWR基准题额定参数

Table 1. Rated Parameters of PWR Benchmark Task

参数	参数值	参数	参数值
组件总数	157	包壳密度/(g·cm⁻³)	6.6
组件边长/cm	21.606	燃料释热率/%	98.1
活性区总长度/cm	367.3	冷却剂释热率/%	1.9
吸收区总长度/cm	362.159	入口冷却剂温度/℃	286
芯块材料	UO₂	入口冷却剂质量流量/(kg·s⁻¹)	12893
芯块密度/(g·cm⁻³)	10.412	系统压力/10⁵Pa	155
包壳材料	Zr-4	额定功率/MW	2775

下载: 导出CSV

表 2 A2工况稳态计算结果

Table 2. Results of Steady State Calculation in A2 Case

参数	ADPRES	耦合平台	相对误差/%
相对功率因子	1	1	0
平均冷却剂温度/℃	302.93	303.14	0.069
最冷却剂大温度/℃	329.91	330.87	0.291
平均芯块温度/℃	544.58	546.20	0.297
最大芯块温度/℃	1669.14	1668.15	−0.059

下载: 导出CSV

表 3 B2工况稳态计算结果

Table 3. Results of Steady State Calculation in B2 Case

参数	ADPRES	耦合平台	相对误差/%
相对功率因子	1	1	0
平均冷却剂温度/℃	302.60	303.02	0.139
最冷却剂大温度/℃	330.15	330.38	0.007
平均芯块温度/℃	542.40	543.11	0.131
最大芯块温度/℃	1576.18	1576.36	0.0114

下载: 导出CSV

表 4 A2工况瞬态计算结果

Table 4. Results of Transient State Calculation in A2 Case

参数	参考值	ADPRES	ADPRES相对误差/%	耦合平台	耦合平台相对误差/%
功率峰时间/s	0.1	0.1	0	0.1	0
功率峰因子	1.08	1.0812	0.11	1.0804	0.037
5 s时功率因子	1.035	1.0357	0.068	1.0351	0.0097
5 s时平均芯块温度/℃	554.6	553.26	−0.21	554.35	−0.045
5 s时最大芯块温度/℃	1691.8	1691.09	−0.04	1691.81	0.0005

下载: 导出CSV

表 5 B2工况瞬态计算结果

Table 5. Results of Transient State Calculation in B2 Case

参数	参考值	ADPRES	ADPRES相对误差/%	耦合平台	耦合平台相对误差/%
功率峰时间/s	0.12	0.12	0	0.12	0
功率峰因子	1.063	1.0633	0.0282	1.0631	0.0094
5 s时功率因子	1.038	1.0395	0.1445	1.039	0.0963
5 s时平均芯块温度/℃	552.0	550.62	0.25	551.33	0.0214
5 s时最大芯块温度/℃	1588.1	1587.97	−0.0082	1588.15	0.0031

下载: 导出CSV

参考文献(35)

[1]	STIMPSON S, POWERS J, CLARNO K, et al. Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1-3[J]. Nuclear Engineering and Design, 2018, 327: 172-186. doi: 10.1016/j.nucengdes.2017.12.015
[2]	CAPPS N, STIMPSON S, CLARNO K, et al. Assessment of the analysis capability for core-wide PWR pellet-clad interaction screening of Watts Bar Unit 1[J]. Nuclear Engineering and Design, 2018, 333: 131-144. doi: 10.1016/j.nucengdes.2018.04.018
[3]	LARZELERE A R. Nuclear energy advanced modeling and simulation (NEAMS)[EB/OL]. (2010-04-29). https://www.energy.gov/sites/prod/files/NEAC042910-NEAM.pdf.
[4]	SOFU T, THOMAS J W. US DOE NEAMS program and SHARP multi-physics toolkit for high-fidelity SFR core design and analysis[C]//FR17: International Conference on Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable Development. Yekaterinburg: IAEA, 2017.
[5]	YU Y Q, SHEMON E R, KIM T K, et al. Evaluation of hot channel factor for sodium-cooled fast reactors with multi-physics toolkit[J]. Nuclear Engineering and Design, 2020, 365: 110704. doi: 10.1016/j.nucengdes.2020.110704
[6]	SHEMON E R, YU Y Q, JUNG Y S, et al. Extension and demonstration of NEAMS multiphysics tools to lead-cooled, sodium-cooled, and molten salt fast reactor applications[Z]. Argonne: Argonne National Lab., 2019.
[7]	CHAULIAC C, ARAGONÉS J M, BESTION D, et al. NURESIM - a European simulation platform for nuclear reactor safety: multi-scale and multi-physics calculations, sensitivity and uncertainty analysis[J]. Nuclear Engineering and Design, 2011, 241(9): 3416-3426. doi: 10.1016/j.nucengdes.2010.09.040
[8]	CHANARON B, AHNERT C, CROUZET N, et al. Advanced multi-physics simulation for reactor safety in the framework of the NURESAFE project[J]. Annals of Nuclear Energy, 2015, 84: 166-177. doi: 10.1016/j.anucene.2014.12.013
[9]	CHANARON B. Overview of the NURESAFE European project[J]. Nuclear Engineering and Design, 2017, 321: 1-7. doi: 10.1016/j.nucengdes.2017.09.001
[10]	SPASOV I, MITKOV S, KOLEV N P, et al. Best-estimate simulation of a VVER MSLB core transient using the NURESIM platform codes[J]. Nuclear Engineering and Design, 2017, 321: 26-37. doi: 10.1016/j.nucengdes.2017.03.032
[11]	PÉRIN Y, VELKOV K. CTF/DYN3D multi-scale coupled simulation of a rod ejection transient on the NURESIM platform[J]. Nuclear Engineering and Technology, 2017, 49(6): 1339-1345. doi: 10.1016/j.net.2017.07.010
[12]	AYDEMIR N U, TROTTIER A, XU T, et al. Coupling of reactor transient simulations via the SALOME platform[J]. Annals of Nuclear Energy, 2019, 126: 434-442. doi: 10.1016/j.anucene.2018.11.049
[13]	ZHANG K L, MUÑOZ A C, SANCHEZ-ESPINOZA V H. Development and verification of the coupled thermal-hydraulic code - TRACE/SCF based on the ICoCo interface and the SALOME platform[J]. Annals of Nuclear Energy, 2021, 155: 108169. doi: 10.1016/j.anucene.2021.108169
[14]	安萍,刘东,潘俊杰,等. 稳态堆芯多物理耦合系统CSSS V1.0的研发[J]. 原子能科学技术,2019, 53(5): 863-868. doi: 10.7538/yzk.2018.youxian.0473
[15]	杨文,胡长军,刘天才,等. 数值反应堆及CVR1.0研究进展[J]. 原子能科学技术,2019, 53(10): 1821-1832.
[16]	李相越,肖维,张滕飞,等. 热管反应堆多物理耦合平台初步研究[J]. 核动力工程,2021, 42(2): 208-212. doi: 10.13832/j.jnpe.2021.02.0208
[17]	LIU Z Y, CHEN J, CAO L Z, et al. Development and verification of the high-fidelity neutronics and thermal-hydraulic coupling code system NECP-X/SUBSC[J]. Progress in Nuclear Energy, 2018, 103: 114-125. doi: 10.1016/j.pnucene.2017.11.010
[18]	LIU Z Y, WANG B, ZHANG M W, et al. An internal parallel coupling method based on NECP-X and CTF and analysis of the impact of thermal-hydraulic model to the high-fidelity calculations[J]. Annals of Nuclear Energy, 2020, 146: 107645. doi: 10.1016/j.anucene.2020.107645
[19]	姜荣,冯文培,陈红丽,等. 基于统一耦合框架的堆芯物理热工耦合程序的开发及验证[J]. 四川大学学报:自然科学版,2021, 58(4): 044006.
[20]	罗晓,张喜林,陈红丽,等. 铅冷快堆多物理耦合分析方法及三维瞬态特性研究[J]. 核动力工程,2021, 42(S1): 11-16. doi: 10.13832/j.jnpe.2021.S1.0011
[21]	ZHANG K L. The multiscale thermal-hydraulic simulation for nuclear reactors: a classification of the coupling approaches and a review of the coupled codes[J]. International Journal of Energy Research, 2020, 44(5): 3295-3315. doi: 10.1002/er.5111
[22]	ZHAO P C, LEI Z Y, YU T, et al. Uncertainty analysis in coupled neutronic/thermal-hydraulic calculations based on computational fluid dynamics[J]. Annals of Nuclear Energy, 2021, 156: 108215. doi: 10.1016/j.anucene.2021.108215
[23]	ZHANG X L, ZHANG K L, SANCHEZ-ESPINOZA V H, et al. Multi-scale coupling of CFD code and sub-channel code based on a generic coupling architecture[J]. Annals of Nuclear Energy, 2020, 141: 107353. doi: 10.1016/j.anucene.2020.107353
[24]	ZHANG K L, ZHANG X L, SANCHEZ-ESPINOZA V, et al. Development of the coupled code–TRACE/TrioCFD based on ICoCo for simulation of nuclear power systems and its validation against the VVER-1000 coolant-mixing benchmark[J]. Nuclear Engineering and Design, 2020, 362: 110602. doi: 10.1016/j.nucengdes.2020.110602
[25]	ZHANG K L. Multi-scale thermal-hydraulic developments for the detailed analysis of the flow conditions within the reactor pressure vessel of pressurized water reactors[D]. Karlsruhe: Karlsruher Institut für Technologie (KIT), 2020.
[26]	SALOME. The open source integration platform for numerical simulation[EB/OL]. [2021-11-10]. https://www.salome-platform.org.
[27]	ZHANG K L, SANCHEZ-ESPINOZA V, STIEGLITZ R. Implementation of the system thermal-hydraulic code TRACE into SALOME platform for multi-scale coupling[C]//50th Annual Meeting on Nuclear Technology (AMNT 2019). Berlin: INFORUM Verlags-und Verwaltungsgesellschaft, 2019.
[28]	SALOME. Salome platform documentation[EB/OL]. [2021-11-10]. https://docs.salome-platform.org/7/dev/MEDCoupling/library.html
[29]	DEVILLE E, PERDU F. Documentation of the interface for code coupling: ICOCO[Z]. CEA, STMF/LMES/RT/12-029/A, 2012.
[30]	ADPRES. An open nuclear reactor simulator and reactor core analysis tool[EB/OL]. [2021-01-09]. https://imronuke.github.io/ADPRES/method
[31]	IMRON M. Development and verification of open reactor simulator ADPRES[J]. Annals of Nuclear Energy, 2019, 133: 580-588. doi: 10.1016/j.anucene.2019.06.049
[32]	刘余,张虹. RELAP5程序耦合接口的开发[J]. 核动力工程,2009, 30(6): 38-40,45.
[33]	何帆. 基于RELAP5/FLUENT耦合程序的熔盐堆热工水力瞬态分析[D]. 上海: 中国科学院大学（中国科学院上海应用物理研究所）, 2021.
[34]	FINNEMANN H, GALATI A. NEACRP-L-335: 3-D LWR core transient benchmark specification: NEACRP-L-335 (Revision 1)[Z]. OECD/NEA, 1992.
[35]	FINNEMANN H, BAUER H, GALATI A, et al. Results of LWR core transient benchmarks[Z]. Paris: Nuclear Energy Agency, 1993.