본문 바로가기
KERD
논문KCI 등재

A validation study of the SLTHEN code for hexagonal assemblies of wire-wrapped pins using liquid metal heating experiments

2

조회

0

다운로드

0

인용

0

참고문헌

키워드

주제분류원자력공학

초록

This paper presents a validation study of the subchannel analysis code SLTHEN used for the core thermalhydraulic design of the Prototype Gen-IV sodium-cooled fast reactor (PGSFR). To assess the performance of the ENERGY model of SLTHEN, four liquid metal heating experiments conducted by ORNL, WARD, and KIT with hexagonal assemblies of wire-wrapped rod bundles were analyzed. These experiments were performed with 19- and 61-pin bundles and varying power distributions of axial and radial peaking factors up to 1.4 and 3.0, respectively. The coolant subchannel temperatures measured at different axial locations were compared with the SLTHEN predictions with the Novendstern, Chiu-Rohsenow-Todreas (CRT), and Cheng-Todreas (CT) correlations for flow split and mixing in wire-wrapped pin bundles. The results showed that the SLTHEN predicts the measured subchannel temperatures reasonably well with root-mean-square errors of ~10 % and maximum errors of ~20 %. It was also observed that the CRT and CT correlations consistently outperform the Novendstern correlation.

Abstract

This paper presents a validation study of the subchannel analysis code SLTHEN used for the core thermalhydraulic design of the Prototype Gen-IV sodium-cooled fast reactor (PGSFR). To assess the performance of the ENERGY model of SLTHEN, four liquid metal heating experiments conducted by ORNL, WARD, and KIT with hexagonal assemblies of wire-wrapped rod bundles were analyzed. These experiments were performed with 19- and 61-pin bundles and varying power distributions of axial and radial peaking factors up to 1.4 and 3.0, respectively. The coolant subchannel temperatures measured at different axial locations were compared with the SLTHEN predictions with the Novendstern, Chiu-Rohsenow-Todreas (CRT), and Cheng-Todreas (CT) correlations for flow split and mixing in wire-wrapped pin bundles. The results showed that the SLTHEN predicts the measured subchannel temperatures reasonably well with root-mean-square errors of ~10 % and maximum errors of ~20 %. It was also observed that the CRT and CT correlations consistently outperform the Novendstern correlation.

참고문헌 (25)

  1. J. Yoo, "Overall system description and safety characteristics of prototype Gen IV sodium cooled fast reactor in Korea", Nucl. Eng. Technol, 48, pp.1059~1070 (2016)
  2. User Manual COBRA-WC: A Version of COBRA for Single-phase, Single and Multiassembly Thermal-Hydraulic Transient Analysis, "E. P. Coomes", Pacific Northwest Laboratory, ~, pp.PNL-4303 (1982)
  3. SABRE-3 - A Computer Program for the Calculation of Steady State Boiling in Rod-Clusters, "J. N. Lillington", Unit Kingdom Atomic Energy Authority, ~, pp.AEEW-M-1647 (1979)
  4. W. S. Kim, "A subchannel analysis code MATRA-LMR for wire wrapped LMR subassembly", Ann. Nucl. Energy, 29, pp.303~321 (2002)
  5. W. S. Yang, "An LMR core thermal-hydraulics code based on the ENERGY model", J. Korean Nucl. Soc, 29, pp.406~416 (1997)
  6. Research on the Development of a Computational Code for Steady-State Thermal-Hydraulics Analysis in an LMR Core, vols. 95–83, "W. S. Yang", Basic Electric Power Engineering Research Institute, ~ (1996)
  7. E. U. Khan, "A porous body model for predicting temperature distribution in wire-wrapped rod assemblies", Nucl. Eng. Des, 35, pp.1~12 (1975)
  8. E. H. Novendstern, "Turbulent flow pressure drop model for fuel rod assemblies utilizing a helical wire-wrap spacer system", Nucl. Eng. Des, 22, pp.19~27 (1972)
  9. C. Chiu, "Turbulent flow split model and supporting experiments for wire-wrapped core assemblies", Nucl. Technol, 50, pp.40~52 (1990)
  10. S. K. Cheng, "Hydrodynamic models and correlations for bare and wire-wrapped hexagonal rod bundles – bundle friction factors, subchannel friction factors and mixing parameters", Nucl. Eng. Des, 92, pp.227~251 (1986)
  11. S. K. Chang, "Flow distribution and pressure loss in subchannels of a wirewrapped 37-pin rod bundle for a sodium-cooled fast reactor", Nucl. Eng. Technol, 48, pp.376~385 (2016)
  12. S. K. Chang, "Experimental study of the flow characteristics in an SFR type 61-pin rod bundle using iso-kinetic sampling method", Ann. Nucl. Energy, 106, pp.160~169 (2017)
  13. H. Kim, "Flow Mixing Characteristics in Subchannels of a Wire-Wrapped 61-pin Rod Assembly for a Sodium-Cooled Fast Reactor", 16th International Meeting on Nuclear Reactor Thermal Hydraulics (2015)
  14. H. Kim, "Investigations of single-phase flow mixing characteristics in a wirewrapped 37-pin bundle for a sodium-cooled fast reactor", Ann. Nucl. Energy, 87, pp.541~546 (2016)
  15. S. R. Choi, "Assessment of subchannel flow mixing coefficients for wirewrapped hexagonal fuel rod bundles", Ann. Nucl. Energy, 166, pp.108810~ (2022)
  16. B. Chen, "Prediction of the coolant temperature field in a breeder reactor including interassembly heat transfer", Nucl. Eng. Des, 35, pp.423~440 (1975)
  17. SUPERENERGY-2: A Multiassembly, Steady-State Computer Code for LMFBR Core Thermal-Hydraulic Analysis, "K. L. Basehore", Pacific Northwest Laboratory, ~, pp.PNL-3379 (1982)
  18. W. S. Yang, "Assessment of the SE2-ANL Code Using EBR-II Temperatures Measurement", 7th International Meeting on Nuclear Reactor Thermal Hydraulics (1995)
  19. R. E. Fontana, "Temperature distribution in the duct wall and at the exit of a 19-rod simulated LMFBR fuel assembly(FFM bundle 2A)", Nucl. Technol, 24, pp.176~200 (1974)
  20. Single-phase Sodium Tests in 61-pin Full-Length Simulated LMFBR Assembly- Record of Phase 1 Experimental Data for THOR Bundle 9, "R. H. Morris", ~, pp.ORNL (1980), TM-7313
  21. F. C. Engel, "Characterization of heat transfer and temperature distributions in an electrically heated model of an LMFBR blanket assembly", Nucl. Eng. Des, 62, pp.335~347 (1980)
  22. J. Pacio, "Experimental study of heavy-liquid metal(LBE)flow and heat transfer along a hexagonal 19-rod bundle with wire spacers", Nucl. Eng. Des, 301, pp.111~127 (2016)
  23. H. A. Abderrahim, "Myrrha – a multi-purpose fast spectrum research reactor", Energy Convers. Manage, 63, pp.4~10 (2012)
  24. J. Pacio, "Thermal-hydraulic study of the LBE-cooled fuel assembly in the MYRRHA reactor : experiments and simulations", Nucl. Eng. Des, 312, pp.327~337 (2017)
  25. M. D. Carelli, "Hot channel factors for rod temperature calculations in LMFBR assemblies", Nucl. Eng. Des, 62, pp.155~180 (1980)

관련 논문AI추천

Development and application of sub-channel thermal-hydraulic analysis code for liquid metal cooled reactorsChen Ronghua, Gui Minyang, Liang Yu, Qiu Suizheng, Tian Wenxi, Wang Jinshun, Zhu Xinyang한국원자력학회 · Nuclear Engineering and Technology2025 · 제57권 제6호In-depth understanding of sodium heat transfer characteristics with low Prandtl number in wire-wrapped fuel assemblies using high-fidelity CFD technique송한섭, 정재호, MUGABI JOPHOUS한국원자력학회 · Nuclear Engineering and Technology2025 · 제57권 제10호Code development and preliminary validation for lead-cooled fast reactor thermal-hydraulic transient behaviorQiu Suizheng, Su Guanghui, Tian Wenxi, Wang Chen, Wang Chenglong한국원자력학회 · Nuclear Engineering and Technology2024 · 제56권 제6호Experimental study and sub-channel analysis on the transverse mixing of the helical cruciform fuel assemblyCai Mengke, Gu Hanyang, Hong Zhenhan, Liu Li, Liu Zhenhai, Pang Hua, Wang Haoyu, Xiao Yao, Yang Guangliang, Zhang Qi, Zhang Yile한국원자력학회 · Nuclear Engineering and Technology2025 · 제57권 제9호Mesh morphing-based pre-processing and numerical simulation of blockage accident in lead–bismuth fast reactor fuel assemblyLiang Huo, Lin Zhikang, Liu Zihua, Luo Hanyan, Ouyang Yong, Sun Hao한국원자력학회 · Nuclear Engineering and Technology2024 · 제56권 제12호Numerical analysis of temperature fluctuation characteristics associated with thermal striping phenomena in the PGSFR정요한, 최선락, 홍종간한국원자력학회 · Nuclear Engineering and Technology2022 · 제54권 제10호

이 간행물의 다른 논문 (39건)