铅基快堆关键热工水力问题研究综述
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摘要
铅基快堆是一种极具发展潜力的第4代核能系统,在燃料增殖和嬗变方面具有独特优势,具有良好的非能动安全特性和经济性,且有利于实现小型化,是目前国际核能领域研究的热点。本文总结了国内外主要铅基堆型,指出了小型化是铅基快堆的发展方向,同时也指出了当前铅基快堆发展所面临的主要问题。针对热工水力关键问题的5个方面,即液态铅/铅铋流动换热特性研究、堆芯/组件热工水力分析、铅池内流动换热现象研究、系统热工水力安全分析以及特殊现象的热工水力分析,对国内外研究现状展开了分析,总结了当前研究成果,并分析了研究的发展趋势以及遇到的技术瓶颈。本文可为铅基快堆的设计和热工水力分析提供一定的建议和指导。
Lead-based fast reactor(LFR) is one of the most promising types among the Gen Ⅳ nuclear reactors. It has better transmutation and breeding capability, as well as higher passive safety and economical efficiency, and it is easy to be miniaturized. As a result, LFR becomes a major concern in the nuclear industry. In this paper, the main concepts of LFR were concluded. A research tendency on small modular LFR was indicated and the existing challenges were pointed out. There are mainly 5 key thermal-hydraulic issues of LFR, namely the flow dynamics and heat transfer of the coolant, thermal-hydraulic phenomena in reactor core and lead pool, system safety characteristics and some specific thermal-hydraulic phenomena of LFR. The research progress in these 5 issues was analyzed and the current research findings were concluded. Besides, the research tendency and difficulties in each research direction were suggested. This paper can provide useful information for LFR design and thermal-hydraulic analysis.
Lead-based fast reactor(LFR) is one of the most promising types among the Gen Ⅳ nuclear reactors. It has better transmutation and breeding capability, as well as higher passive safety and economical efficiency, and it is easy to be miniaturized. As a result, LFR becomes a major concern in the nuclear industry. In this paper, the main concepts of LFR were concluded. A research tendency on small modular LFR was indicated and the existing challenges were pointed out. There are mainly 5 key thermal-hydraulic issues of LFR, namely the flow dynamics and heat transfer of the coolant, thermal-hydraulic phenomena in reactor core and lead pool, system safety characteristics and some specific thermal-hydraulic phenomena of LFR. The research progress in these 5 issues was analyzed and the current research findings were concluded. Besides, the research tendency and difficulties in each research direction were suggested. This paper can provide useful information for LFR design and thermal-hydraulic analysis.
引文
[1] GROMOV B F, BELOMITCEV Y S, YEFIMOV E I. Use of lead-bismuth coolant in nuclear reactors and accelerator-driven systems[J]. Nuclear Engineering and Design, 1997, 173: 207-217.
[2] ORLOV V V, FILIN A I, LOPATKIN A V, et al. The closed on-site fuel cycle of the Brest reactors[J]. Progress in Nuclear Energy, 2005, 47(1-4): 171-177.
[3] ZRODNIKOV A V, TOSHINSKY G I, KOMLEV O G, et al. Innovative nuclear technology based on modular multi-purpose lead-bismuth cooled fast reactors[J]. Progress in Nuclear Energy, 2008, 50(2): 170-178.
[4] DAVIS C, HERRING S, MACDONALD P, et al. Design of an actinide burning, lead-bismuth cooled reactor that produces low cost electricity[R]. Idaho: Office of Scientific and Technical Information Technical Reports, 2002.
[5] BORTOT S, MOISSEYTSEV A, SIENICKI J J, et al. Core design investigation for a SUPERSTAR small modular lead-cooled fast reactor demonstrator[J]. Nuclear Engineering and Design, 2011, 241(8): 3 021-3 031.
[6] MAES D. Mechanical design of the small-scale experimental ADS: MYRRHA[J]. Energy Conversion and Management, 2006, 47(17): 2 710-2 723.
[7] ALEMBERTI A, CARLSSON J, MALAMBU E, et al. European lead fast reactor: ELSY[J]. Nuclear Engineering and Design, 2011, 241(9): 3 470-3 480.
[8] GRASSO G, PETROVICH C, MATTIOLI D, et al. The core design of ALFRED, a demonstrator for the European lead-cooled reactors[J]. Nuclear Engineering and Design, 2014, 278: 287-301.
[9] WU Y. CLEAR-S: An integrated non-nuclear test facility for China lead-based research reactor[J]. International Journal of Energy Research, 2016, 40(14): 1 951-1 956.
[10] CHEN H, CHEN Z, CHEN C, et al. Conceptual design of a small modular natural circulation lead cooled fast reactor SNCLFR-100[J]. International Journal of Hydrogen Energy, 2016, 41(17): 7 158-7 168.
[11] TAKAHASHI M, UCHIDA S, KASAHARA Y. Design study on reactor structure of Pb-Bi-cooled direct contact boiling water fast reactor (PBWFR)[J]. Progress in Nuclear Energy, 2008, 50: 197-205
[12] SEKIMOTO H, YAN M. Design study on small CANDLE reactor[J]. Energy Conversion and Management, 2008, 49(7): 1 868-1 872.
[13] CHANG J E, SUH K Y, HWANG S. Natural circulation capability of Pb-Bi cooled fast reactor: PEACER[J]. Progress in Nuclear Energy, 2000, 37(1-4): 211-216
[14] CHOI S, CHO J H, BAE M H, et al. PASCAR: Long burning small modular reactor based on natural circulation[J]. Nuclear Engineering and Design, 2011, 241(5): 1 486-1 499.
[15] PACIO J, MAROCCO L, WETZEL T. Review of data and correlations for turbulent forced convective heat transfer of liquid metals in pipes[J]. Heat and Mass Transfer, 2015, 51(2): 153-164.
[16] JAEGER W, HERING W, LUX M, et al. Liquid metal thermal hydraulics in rectangular ducts: Review, proposal and validation of empirical models[C]//International Conference on Nuclear Engineering. Japan: JSME, 2015.
[17] JAEGER W. Heat transfer to liquid metals with empirical models for turbulent forced convection in various geometries[J]. Nuclear Engineering and Design, 2017, 319: 12-27.
[18] JAEGER W. Empirical models for liquid metal heat transfer in the entrance region of tubes and rod bundles[J]. Heat and Mass Transfer, 2017, 53(5): 1-18.
[19] MARTELLI D, FORGIONE N, PIAZZA I D, et al. HLM fuel pin bundle experiments in the CIRCE pool facility[J]. Nuclear Engineering and Design, 2015, 292: 76-86.
[20] PACIO J, WETZEL T, DOOLAARD H, et al. Thermal-hydraulic study of the LBE-cooled fuel assembly in the MYRRHA reactor: Experiments and simulations[J]. Nuclear Engineering and Design, 2017, 312: 327-337.
[21] 田永红. 新概念铅铋合金冷却快堆PBWFR热工水力特性分析研究[D]. 西安:西安交通大学,2014.
[22] 吕科锋. 液态铅铋合金在带绕丝棒束组件内热工水力行为研究[D]. 合肥:中国科学技术大学,2016.
[23] PACIO J, DAUBNER M, FELLMOSER F, et al. Heavy-liquid metal heat transfer experiment in a 19-rod bundle with grid spacers[J]. Nuclear Engineering and Design, 2014, 273: 33-46.
[24] POINTER W D, FISCHER P, SMITH J, et al. RANS simulations of turbulent diffusion in wire-wrapped sodium fast reactor fuel assemblies[C]//International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities. Vienna: IAEA, 2009.
[25] NATESAN K, SUNDARARAJAN T, NARASIMHAN A, et al. Turbulent flow simulation in a wire-wrap rod bundle of an LMFBR[J]. Nuclear Engineering and Design, 2010, 240(5): 1 063-1 072.
[26] SHAMS A, ROELOFS F, BAGLIETTO E, et al. Assessment and calibration of an algebraic turbulent heat flux model for low-Prandtl fluids[J]. International Journal of Heat and Mass Transfer, 2014, 79(4): 589-601.
[27] VIRGILIO M, FARCHIONI R, GROSSO G. Status and perspective of turbulence heat transfer modelling for the industrial application of liquid metal flows[J]. Nuclear Engineering and Design, 2015, 290: 99-106.
[28] GE Z, LIU J, ZHAO P, et al. Investigation on the applicability of turbulent-Prandtl-number models in bare rod bundles for heavy liquid metals[J]. Nuclear Engineering and Design, 2017, 314: 198-206.
[29] GAJAPATHY R, VELUSAMY K, SELVARAJ P, et al. CFD investigation of helical wire-wrapped 7-pin fuel bundle and the challenges in modeling full scale 217 pin bundle[J]. Nuclear Engineering and Design, 2007, 237(24): 2 332-2 342.
[30] LIU P, CHEN X, RINEISKI A, et al. Transient analyses of the 400 MWth-class EFIT accelerator driven transmuter with the multi-physics code: SIMMER-Ⅲ[J]. Nuclear Engineering and Design, 2010, 240(10): 3 481-3 494.
[31] KRIVENTSEV V, RINEISKI A, MASCHEK W. Application of safety analysis code SIMMER-Ⅳ to blockage accidents in FASTEF subcritical core[J]. Annals of Nuclear Energy, 2014, 64: 114-121.
[32] TENCHINE D. Some thermal hydraulic challenges in sodium cooled fast reactors[J]. Nuclear Engineering and Design, 2010, 240(5): 1 195-1 217.
[33] VELUSAMY K, CHELLAPANDI P, CHETAL S C, et al. Overview of pool hydraulic design of Indian prototype fast breeder reactor[J]. Sadhana, 2010, 35(2): 97-128.
[34] PATWARDHAN A W, MALI R G, JADHAO S B, et al. Argon entrainment into liquid sodium in fast breeder reactor[J]. Nuclear Engineering and Design, 2012, 249(10): 204-211.
[35] BANERJEE I, CHANDRA L, LAXMAN D, et al. Gas entrainment in scaled model of pool type LMFBR[C]// ICAPP’07. Nice, France: ANS, 2007.
[36] TAK N I, SONG T Y, PARK W S, et al. Preliminary thermal hydraulic analysis of HYPER fuel assembly using MATRA[C]∥Seventh Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation. Jeju, Korea: Korea Hydro and Nuclear Power-Central Research Institute, 2002.
[37] BANDINI G, POLIDORI M, GERSCHENFELD A, et al. Assessment of systems codes and their coupling with CFD codes in thermal-hydraulic applications to innovative reactors[J]. Nuclear Engineering and Design, 2015, 281(3): 22-38.
[38] MA W M, BUBELIS E, KARBOJIAN A, et al. Transient experiments from the thermal-hydraulic ADS lead bismuth loop (TALL) and comparative TRAC/AAA analysis[J]. Nuclear Engineering and Design, 2006, 236 (13): 1 422-1 444.
[39] BORGOHAIN A, JAISWAL B K, MAHESHWARI N K, et al. Natural circulation studies in a lead bismuth eutectic loop[J]. Progress in Nuclear Energy, 2011, 53(4): 308-319.
[40] WU Q, SIENICKI J J. Stability analysis on single-phase natural circulation in Argonne lead loop facility[J]. Nuclear Engineering and Design, 2003, 224(1): 23-32.
[41] BENAMATI G, FOLETTI C, FORGIONE N, et al. Experimental study on gas-injection enhanced circulation performed with the CIRCE facility[J]. Nuclear Engineering and Design, 2007, 237(7): 768-777.
[42] NISHI Y, KINOSHITA I, NISHIMURA S. Experimental study on gas lift pump performance in lead-bismuth eutectic[C]//Proceedings of International Congress on Advanced Nuclear Power Plant, ICAPP’03. Cordoba, Spain: ANS, 2003.
[43] SUZUKI T, TOBITA Y, KONDO S, et al. Analysis of gas-liquid metal two-phase flows using a reactor safety analysis code SIMMER-Ⅲ[J]. Nuclear Engineering and Design, 2003, 220(3): 207-223.
[44] MIKITYUK K, CODDINGTON P, CHAWLA R. Development of a drift-flux model for heavy liquid metal/gas flow[J]. Journal of Nuclear Science and Technology, 2005, 42(7): 600-607.
[45] BEZNOSOV A V, PINAEV S S, DAVYDOV D V, et al. Experimental studies of the characteristics of contact heat exchange between lead coolant and the working body[J]. Atomic Energy, 2005, 98(3): 170-176.
[46] SIBAMOTO Y, KUKITA Y, NAKAMURA H. Small-scale experiment on subcooled water jet injection into molten alloy by using fluid temperature-phase coupled measurement and visualization[J]. Journal of Nuclear Science and Technology, 2007, 44(8): 1 059-1 069.
[47] FURUYA M, KINOSHITA I. Effects of polymer, surfactant, and salt additives to a coolant on the mitigation and the severity of vapor explosions[J]. Experimental Thermal and Fluid Science, 2002, 26(2): 213-219.
[48] FURUYA M, ARAI T. Effect of surface property of molten metal pools on triggering of vapor explosions in water droplet impingement[J]. International Journal of Heat and Mass Transfer, 2008, 51(17): 4 439-4 446.
[49] FLAD M, WANG S, MASCHEK W. Simulation of a steam generator tube rupture accident in a lead-cooled accelerator driven system[C]//International Conference on Nuclear Engineering. Beijing: CNS, 2010.
[50] 黄望哩. 铅基堆SGTR事故下铅铋与水接触碎化行为研究[D]. 合肥:中国科学技术大学,2015.
[2] ORLOV V V, FILIN A I, LOPATKIN A V, et al. The closed on-site fuel cycle of the Brest reactors[J]. Progress in Nuclear Energy, 2005, 47(1-4): 171-177.
[3] ZRODNIKOV A V, TOSHINSKY G I, KOMLEV O G, et al. Innovative nuclear technology based on modular multi-purpose lead-bismuth cooled fast reactors[J]. Progress in Nuclear Energy, 2008, 50(2): 170-178.
[4] DAVIS C, HERRING S, MACDONALD P, et al. Design of an actinide burning, lead-bismuth cooled reactor that produces low cost electricity[R]. Idaho: Office of Scientific and Technical Information Technical Reports, 2002.
[5] BORTOT S, MOISSEYTSEV A, SIENICKI J J, et al. Core design investigation for a SUPERSTAR small modular lead-cooled fast reactor demonstrator[J]. Nuclear Engineering and Design, 2011, 241(8): 3 021-3 031.
[6] MAES D. Mechanical design of the small-scale experimental ADS: MYRRHA[J]. Energy Conversion and Management, 2006, 47(17): 2 710-2 723.
[7] ALEMBERTI A, CARLSSON J, MALAMBU E, et al. European lead fast reactor: ELSY[J]. Nuclear Engineering and Design, 2011, 241(9): 3 470-3 480.
[8] GRASSO G, PETROVICH C, MATTIOLI D, et al. The core design of ALFRED, a demonstrator for the European lead-cooled reactors[J]. Nuclear Engineering and Design, 2014, 278: 287-301.
[9] WU Y. CLEAR-S: An integrated non-nuclear test facility for China lead-based research reactor[J]. International Journal of Energy Research, 2016, 40(14): 1 951-1 956.
[10] CHEN H, CHEN Z, CHEN C, et al. Conceptual design of a small modular natural circulation lead cooled fast reactor SNCLFR-100[J]. International Journal of Hydrogen Energy, 2016, 41(17): 7 158-7 168.
[11] TAKAHASHI M, UCHIDA S, KASAHARA Y. Design study on reactor structure of Pb-Bi-cooled direct contact boiling water fast reactor (PBWFR)[J]. Progress in Nuclear Energy, 2008, 50: 197-205
[12] SEKIMOTO H, YAN M. Design study on small CANDLE reactor[J]. Energy Conversion and Management, 2008, 49(7): 1 868-1 872.
[13] CHANG J E, SUH K Y, HWANG S. Natural circulation capability of Pb-Bi cooled fast reactor: PEACER[J]. Progress in Nuclear Energy, 2000, 37(1-4): 211-216
[14] CHOI S, CHO J H, BAE M H, et al. PASCAR: Long burning small modular reactor based on natural circulation[J]. Nuclear Engineering and Design, 2011, 241(5): 1 486-1 499.
[15] PACIO J, MAROCCO L, WETZEL T. Review of data and correlations for turbulent forced convective heat transfer of liquid metals in pipes[J]. Heat and Mass Transfer, 2015, 51(2): 153-164.
[16] JAEGER W, HERING W, LUX M, et al. Liquid metal thermal hydraulics in rectangular ducts: Review, proposal and validation of empirical models[C]//International Conference on Nuclear Engineering. Japan: JSME, 2015.
[17] JAEGER W. Heat transfer to liquid metals with empirical models for turbulent forced convection in various geometries[J]. Nuclear Engineering and Design, 2017, 319: 12-27.
[18] JAEGER W. Empirical models for liquid metal heat transfer in the entrance region of tubes and rod bundles[J]. Heat and Mass Transfer, 2017, 53(5): 1-18.
[19] MARTELLI D, FORGIONE N, PIAZZA I D, et al. HLM fuel pin bundle experiments in the CIRCE pool facility[J]. Nuclear Engineering and Design, 2015, 292: 76-86.
[20] PACIO J, WETZEL T, DOOLAARD H, et al. Thermal-hydraulic study of the LBE-cooled fuel assembly in the MYRRHA reactor: Experiments and simulations[J]. Nuclear Engineering and Design, 2017, 312: 327-337.
[21] 田永红. 新概念铅铋合金冷却快堆PBWFR热工水力特性分析研究[D]. 西安:西安交通大学,2014.
[22] 吕科锋. 液态铅铋合金在带绕丝棒束组件内热工水力行为研究[D]. 合肥:中国科学技术大学,2016.
[23] PACIO J, DAUBNER M, FELLMOSER F, et al. Heavy-liquid metal heat transfer experiment in a 19-rod bundle with grid spacers[J]. Nuclear Engineering and Design, 2014, 273: 33-46.
[24] POINTER W D, FISCHER P, SMITH J, et al. RANS simulations of turbulent diffusion in wire-wrapped sodium fast reactor fuel assemblies[C]//International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities. Vienna: IAEA, 2009.
[25] NATESAN K, SUNDARARAJAN T, NARASIMHAN A, et al. Turbulent flow simulation in a wire-wrap rod bundle of an LMFBR[J]. Nuclear Engineering and Design, 2010, 240(5): 1 063-1 072.
[26] SHAMS A, ROELOFS F, BAGLIETTO E, et al. Assessment and calibration of an algebraic turbulent heat flux model for low-Prandtl fluids[J]. International Journal of Heat and Mass Transfer, 2014, 79(4): 589-601.
[27] VIRGILIO M, FARCHIONI R, GROSSO G. Status and perspective of turbulence heat transfer modelling for the industrial application of liquid metal flows[J]. Nuclear Engineering and Design, 2015, 290: 99-106.
[28] GE Z, LIU J, ZHAO P, et al. Investigation on the applicability of turbulent-Prandtl-number models in bare rod bundles for heavy liquid metals[J]. Nuclear Engineering and Design, 2017, 314: 198-206.
[29] GAJAPATHY R, VELUSAMY K, SELVARAJ P, et al. CFD investigation of helical wire-wrapped 7-pin fuel bundle and the challenges in modeling full scale 217 pin bundle[J]. Nuclear Engineering and Design, 2007, 237(24): 2 332-2 342.
[30] LIU P, CHEN X, RINEISKI A, et al. Transient analyses of the 400 MWth-class EFIT accelerator driven transmuter with the multi-physics code: SIMMER-Ⅲ[J]. Nuclear Engineering and Design, 2010, 240(10): 3 481-3 494.
[31] KRIVENTSEV V, RINEISKI A, MASCHEK W. Application of safety analysis code SIMMER-Ⅳ to blockage accidents in FASTEF subcritical core[J]. Annals of Nuclear Energy, 2014, 64: 114-121.
[32] TENCHINE D. Some thermal hydraulic challenges in sodium cooled fast reactors[J]. Nuclear Engineering and Design, 2010, 240(5): 1 195-1 217.
[33] VELUSAMY K, CHELLAPANDI P, CHETAL S C, et al. Overview of pool hydraulic design of Indian prototype fast breeder reactor[J]. Sadhana, 2010, 35(2): 97-128.
[34] PATWARDHAN A W, MALI R G, JADHAO S B, et al. Argon entrainment into liquid sodium in fast breeder reactor[J]. Nuclear Engineering and Design, 2012, 249(10): 204-211.
[35] BANERJEE I, CHANDRA L, LAXMAN D, et al. Gas entrainment in scaled model of pool type LMFBR[C]// ICAPP’07. Nice, France: ANS, 2007.
[36] TAK N I, SONG T Y, PARK W S, et al. Preliminary thermal hydraulic analysis of HYPER fuel assembly using MATRA[C]∥Seventh Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation. Jeju, Korea: Korea Hydro and Nuclear Power-Central Research Institute, 2002.
[37] BANDINI G, POLIDORI M, GERSCHENFELD A, et al. Assessment of systems codes and their coupling with CFD codes in thermal-hydraulic applications to innovative reactors[J]. Nuclear Engineering and Design, 2015, 281(3): 22-38.
[38] MA W M, BUBELIS E, KARBOJIAN A, et al. Transient experiments from the thermal-hydraulic ADS lead bismuth loop (TALL) and comparative TRAC/AAA analysis[J]. Nuclear Engineering and Design, 2006, 236 (13): 1 422-1 444.
[39] BORGOHAIN A, JAISWAL B K, MAHESHWARI N K, et al. Natural circulation studies in a lead bismuth eutectic loop[J]. Progress in Nuclear Energy, 2011, 53(4): 308-319.
[40] WU Q, SIENICKI J J. Stability analysis on single-phase natural circulation in Argonne lead loop facility[J]. Nuclear Engineering and Design, 2003, 224(1): 23-32.
[41] BENAMATI G, FOLETTI C, FORGIONE N, et al. Experimental study on gas-injection enhanced circulation performed with the CIRCE facility[J]. Nuclear Engineering and Design, 2007, 237(7): 768-777.
[42] NISHI Y, KINOSHITA I, NISHIMURA S. Experimental study on gas lift pump performance in lead-bismuth eutectic[C]//Proceedings of International Congress on Advanced Nuclear Power Plant, ICAPP’03. Cordoba, Spain: ANS, 2003.
[43] SUZUKI T, TOBITA Y, KONDO S, et al. Analysis of gas-liquid metal two-phase flows using a reactor safety analysis code SIMMER-Ⅲ[J]. Nuclear Engineering and Design, 2003, 220(3): 207-223.
[44] MIKITYUK K, CODDINGTON P, CHAWLA R. Development of a drift-flux model for heavy liquid metal/gas flow[J]. Journal of Nuclear Science and Technology, 2005, 42(7): 600-607.
[45] BEZNOSOV A V, PINAEV S S, DAVYDOV D V, et al. Experimental studies of the characteristics of contact heat exchange between lead coolant and the working body[J]. Atomic Energy, 2005, 98(3): 170-176.
[46] SIBAMOTO Y, KUKITA Y, NAKAMURA H. Small-scale experiment on subcooled water jet injection into molten alloy by using fluid temperature-phase coupled measurement and visualization[J]. Journal of Nuclear Science and Technology, 2007, 44(8): 1 059-1 069.
[47] FURUYA M, KINOSHITA I. Effects of polymer, surfactant, and salt additives to a coolant on the mitigation and the severity of vapor explosions[J]. Experimental Thermal and Fluid Science, 2002, 26(2): 213-219.
[48] FURUYA M, ARAI T. Effect of surface property of molten metal pools on triggering of vapor explosions in water droplet impingement[J]. International Journal of Heat and Mass Transfer, 2008, 51(17): 4 439-4 446.
[49] FLAD M, WANG S, MASCHEK W. Simulation of a steam generator tube rupture accident in a lead-cooled accelerator driven system[C]//International Conference on Nuclear Engineering. Beijing: CNS, 2010.
[50] 黄望哩. 铅基堆SGTR事故下铅铋与水接触碎化行为研究[D]. 合肥:中国科学技术大学,2015.