DOI QR코드

DOI QR Code

Platform development for multi-physics coupling and uncertainty analysis based on a unified framework

  • Guan-Hua Qian (School of Nuclear Science and Technology, University of South China) ;
  • Ren Li (College of Nuclear Science and Technology, Harbin Engineering University) ;
  • Tao Yang (School of Nuclear Science and Technology, University of South China) ;
  • Xu Wang (School of Nuclear Science and Technology, University of South China) ;
  • Peng-Cheng Zhao (School of Nuclear Science and Technology, University of South China) ;
  • Ya-Nan Zhao (School of Nuclear Science and Technology, University of South China) ;
  • Tao Yu (School of Nuclear Science and Technology, University of South China)
  • Received : 2022.12.08
  • Accepted : 2023.02.04
  • Published : 2023.05.25

Abstract

The multi-physics coupled methodologies that have been widely used to analyze the complex process occurring in nuclear reactors have also been used to the R&D of numerical reactors. The advancement in the field of computer technology has helped in the development of these methodologies. Herein, we report the integration of ADPRES code and RELAP5 code into the SALOME-ICoCo framework to form a multi-physics coupling platform. The platform exploits the supervisor architecture, serial mode, mesh one-to-one correspondence and explicit coupling methods during analysis, and the uncertainty analysis tool URANIE was used. The correctness of the platform was verified through the NEACRP-L-335 benchmark. The results obtained were in accordance with the reference values. The platform could be used to accurately determine the power peak. In addition, design margins could be gained post uncertainty analysis. The initial power, inlet coolant temperature and the mass flow of assembly property significantly influence reactor safety during the rod ejections accident (REA).

Keywords

Acknowledgement

The authors would like to express their deepest gratitude to NEAL (Nuclear Engineering and Application Laboratory) Team for its help during this research.

References

  1. S. Stimpson, J. Powers, K. Clarno, R. Pawlowski, R. Gardner, S. Novascone, K. Gamble, R. Williamson, Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1-3, Nucl. Eng. Des. 327 (2018) 172-186. https://doi.org/10.1016/j.nucengdes.2017.12.015
  2. N. Capps, S. Stimpson, K. Clarno, B.D. Wirth, J. Rashid, Assessment of the analysis capability for core-wide PWR pellet-clad interaction screening of Watts Bar Unit 1, Nucl. Eng. Des. 333 (2018) 131-144. https://doi.org/10.1016/j.nucengdes.2018.04.018
  3. A.R. Larzelere, Nuclear Energy Advanced Modeling and Simulation, NEAMS), 2010. https://www.energy.gov/sites/default/files/NEAC042910-NEAM.pdf.
  4. T. Sofu, J.W. Thomas, US DOE NEAMS Program and SHARP Multi-Physics Toolkit for High-Fidelity SFR Core Design and Analysis, Argonne Natl. Lab., 2017, pp. 1-10. IAEA-CN245-054.
  5. Y. Yu, E.R. Shemon, T.K. Kim, E. Merzari, Evaluation of hot channel factor for sodium-cooled fast reactors with multi-physics toolkit, Nucl. Eng. Des. 365 (2020), 110704.
  6. E.R. Shemon, Y. Yu, Y.S. Jung, B. Feng, T.K. Kim, Extension and Demonstration of NEAMS Multiphysics Tools to Lead- Cooled, Sodium-Cooled, and Molten Salt Fast Reactor Applications, Argonne Natl. Lab., 2019, pp. 1-70. ANL/NSE19/30.
  7. C. Chauliac, J.M. Aragones, D. Bestion, D.G. Cacuci, N. Crouzet, F.-P. Weiss, M.A. Zimmermann, NURESIM - a European simulation platform for nuclear reactor safety: multi-scale and multi-physics calculations, sensitivity and uncertainty analysis, Nucl. Eng. Des. 241 (2011) 3416-3426. https://doi.org/10.1016/j.nucengdes.2010.09.040
  8. B. Chanaron, C. Ahnert, N. Crouzet, Advanced multi-physics simulation for reactor safety in the framework of the NURESAFE project, Ann. Nucl. Energy 84 (2015) 166-177. https://doi.org/10.1016/j.anucene.2014.12.013
  9. B. Chanaron, Overview of the NURESAFE European project, Nucl. Eng. Des. 321 (2017) 1-7. https://doi.org/10.1016/j.nucengdes.2017.09.001
  10. I. Spasov, S. Mitkov, N.P. Kolev, S. Sanchez-Cervera, N. Garcia-Herranz, A. Sabater, D. Cuervo, J. Jimenez, V.H. Sanchez, L. Vyskocil, Best-estimate simulation of a VVER MSLB core transient using the NURESIM platform codes, Nucl. Eng. Des. 321 (2017) 26-37. https://doi.org/10.1016/j.nucengdes.2017.03.032
  11. Y. Perin, K. Velkov, CTF/DYN3D multi-scale coupled simulation of a rod ejection transient on the NURESIM platform, Nucl. Eng. Technol. 49 (6) (2017) 1339-1345. https://doi.org/10.1016/j.net.2017.07.010
  12. N.U. Aydemir, A. Trottier, T. Xu, M. Echlin, T. Chin, Coupling of reactor transient simulations via the SALOME platform, Ann. Nucl. Energy 126 (2019) 434-442. https://doi.org/10.1016/j.anucene.2018.11.049
  13. K. Zhang, A.C. Munoz, V.H. Sanchez-Espinoza, Development and verification of the coupled thermal- hydraulic code - TRACE/SCF based on the ICoCo interface and the SALOME platform, Ann. Nucl. Energy 155 (2021), 108169.
  14. P. An, D. Liu, J. Pan, W. Zhao, W. Liu, Development of steady reactor core multi-physics coupling system CSSS V1.0, At. Energy Sci. Technol. 53 (5) (2019) 863-868.
  15. W. Yang, C. Hu, T. Liu, A. Wang, M. Wu, Research progress of China virtual reactor (CVR1.0), At. Energy Sci. Technol. 53 (10) (2019) 1821-1832.
  16. X. Li, W. Xiao, T. Zhang, J. Li, X. Liu, Preliminary Research on a multi-physics integration platform for heat pipe reactors, Nucl. Power Eng. 42 (2021) 208-212, 02.
  17. Z. Liu, J. Chen, L. Cao, C. Zhao, Q. He, H. Wu, Development and verification of the high-fidelity neutronics and thermal-hydraulic coupling code system NECP-X/SUBSC, Prog. Nucl. Energy 103 (2018) 114-125. https://doi.org/10.1016/j.pnucene.2017.11.010
  18. Z. Liu, B. Wang, M. Zhang, X. Zhou, L. Cao, H. Wu, 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, Ann. Nucl. Energy 146 (2020), 107645.
  19. R. Jiang, W. Feng, H. Chen, S. Qiang, Z. Li, J. Pan, X. Zhang, X. Luo, Development and verification of core physical-thermal coupling code based on unified coupling framework, J. Sichuan Univ. (Nat. Sci. Ed. 58 (2021), 044006.
  20. X. Luo, X. Zhang, H. Chen, S. Wang, C. Guo, C. Wang, Analysis of transient thermal-hydraulic and safety of lead-cooled fast reactor based on unified coupling framework, Nucl. Power Eng. 42 (S1) (2021) 11-16.
  21. SALOME, The open source integration platform for numerical simulation, 2022. https://www.salome-platform.org.
  22. K. Zhang, V.H. Sanchez-Espinoza, R. Stieglitz, Implementation of the system thermal-hydraulic code TRACE into SALOME platform for multi-scale coupling, in: 50th Annual Meeting on Nuclear Technology (AMNT 2019), 2019. Berlin.
  23. SALOME, Salome platform documentation, 2016. https://docs.salome-platform.org/7/dev/MEDCoupling /library.html.
  24. E. Deville, F. Perdu, Documentation of the Interface for Code Coupling, ICOCO, CEA, 2012, pp. 9-10. STMF/LMES/RT/12-029/A.
  25. ADPRES, An open nuclear reactor simulator and reactor core analysis tool, 2021. https://imronuke.github.io/ADPRES/method.
  26. M. Imron, Development and verification of open reactor simulator ADPRES, Ann. Nucl. Energy 133 (2019) 580-588. https://doi.org/10.1016/j.anucene.2019.06.049
  27. J.-B. Blanchard, G. Damblin, J.-M. Martinez, G. Arnaud, F. Gaudier, The Uranie platform: an open-source software for optimisation, meta-modelling and uncertainty analysis, EPJ Nucl. Sci. Technol. 5 (2019) 4.
  28. S.S. Wilks, Statistical prediction with special reference to the problem of tolerance limits, Ann. Mathemat. Statist. 13 (4) (1942) 400-409. https://doi.org/10.1214/aoms/1177731537
  29. K. Zhang, The multiscale thermal-hydraulic simulation for nuclear reactors: a classification of the coupling approaches and a review of the coupled codes, Int. J. Energy Res. 44 (5) (2020) 3295-3315. https://doi.org/10.1002/er.5111
  30. X. Zhang, K. Zhang, V.H. Sanchez-Espinoza, H. Chen, Multi-scale coupling of CFD code and sub-channel code based on a generic coupling architecture, Ann. Nucl. Energy 141 (2020), 107353.
  31. K. Zhang, X. Zhang, V.H. Sanchez-Espinoza, R. Stieglitz, 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, Nucl. Eng. Des. 362 (2020), 110602.
  32. K. Zhang, Multi-Scale Thermal-hydraulic Developments for the Detailed Analysis of the Flow Conditions within the Reactor Pressure Vessel of Pressurized Water Reactors, Karlsruher Instituts fur Technologie (KIT), 2020.
  33. H. Finnemann, A. Galati, NEACRP-L-335: 3-D LWR Core Transient Benchmark Specification, OECD/NEA, 1992, pp. 5-17. NEACRP-L-335 (Revision 1).
  34. S. Pinem, T.M. Sembiring, P.H. Liem, NODAL3 sensitivity analysis for NEACRP 3D LWR core transient benchmark (PWR), Sci. Technol. Nucl. Ins. 2016 (2016), 7538681.
  35. H. Finnemann, H. Bauer, A. Galati, R. Martinelli, Results of LWR Core Transient Benchmarks, NEA, 1993, pp. 15-33. NEA/NSC/DOC(93)25.
  36. C.R. Hernandez, J. Wallenius, J. Luxat, Dynamic sensitivity and uncertainty analysis of a small lead cooled reactor, Ann. Nucl. Energy 144 (2020), 107512.
  37. C. Hao, P. Li, D. She, X. Zhou, R. Yang, Sensitivity and Uncertainty analysis of the maximum fuel temperature under accident condition of HTR-PM, Sci. Technol. Nucl. Ins. 2020 (2020), 9235783.
  38. B. Batki, B. Kvizda, A. Kereszturi, I. Panka, Uncertainty analyses of transients on the allegro reactor, in: ANS Best Estimate Plus Uncertainty International Conference (BEPU 2018), Real Collegio, May 13-19, 2018. Lucca, Italy.
  39. B. Batki, A. Kereszturi, I. Panka, Uncertainty analyses of unprotected transients in fast reactors from reactor physics point of view, in: 26th International Conference Nuclear Energy for New Europe (NENE2017), 2017, pp. 11-14. Bled, Slovenia, September.
  40. B.T. Hallee, Feed-and-Bleed Transient Analysis of OSU APEX Facility Using the Modern Code Scaling, Applicability, and Uncertainty Method, Oregon State University (OSU), 2013.