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

Qian Guanhua; Yu Tao; Yang Tao; Zhao Yanan; Zhao Pengcheng

doi:10.13832/j.jnpe.2022.06.0051

Volume 43 Issue 6

Dec. 2022

Turn off MathJax

Article Contents

Article Navigation > Nuclear Power Engineering > 2022 > 43(6): 51-60

Qian Guanhua, Yu Tao, Yang Tao, Zhao Yanan, Zhao Pengcheng. Research and Platform Development of Multi-physical Coupling Scheme Based on Unified Framework[J]. Nuclear Power Engineering, 2022, 43(6): 51-60. doi: 10.13832/j.jnpe.2022.06.0051

Citation:

Qian Guanhua, Yu Tao, Yang Tao, Zhao Yanan, Zhao Pengcheng. Research and Platform Development of Multi-physical Coupling Scheme Based on Unified Framework[J]. Nuclear Power Engineering, 2022, 43(6): 51-60. doi: 10.13832/j.jnpe.2022.06.0051

Citation:

PDF( 5914 KB)

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

doi: 10.13832/j.jnpe.2022.06.0051

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

Received Date: 2021-11-19
Rev Recd Date: 2022-01-04
Publish Date: 2022-12-14

Abstract

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

FullText(HTML)

References(35)

References

[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.