系统安全分析程序在超临界水冷堆和钠冷快堆上的适用性研发与应用
|
摘要
系统安全分析程序（以下简称系统程序）是对反应堆系统进行瞬态和事故安全分析的重要工具。在近十年来第四代先进核能系统研发热潮背景下，国际核能界针对第四代反应堆系统程序开展了大量研究工作。由于轻水堆系统程序发展成熟，经过了全面系统的验证和广泛的应用，其中一些被加以修改以适用于第四代核能系统。这些修改后的程序的适用性及可靠性还有待进一步研究。
本文针对第四代反应堆中的超临界水冷堆和钠冷快堆进行系统程序的模型开发、验证与应用研究。首先对轻水堆系统程序ATHLET在这两种第四代堆型上的适用性进行评价，在此基础上对ATHLET程序进行二次模型开发，即开发适用于超临界水和钠两种流体的相关计算模型，获得多流体系统程序ATHLET-MF（Multi-Fluid）；其次，对修改后的ATHLET-MF程序进行初步验证，并采用法国PHENIX钠冷快堆的自然对流实验对程序进行评估；最后，将ATHLET-MF程序应用于超临界水冷堆燃料性能验证实验回路的安全分析，并提出对该实验回路的安全系统设计的改进意见。
论文的主要工作包括：
1. ATHLET程序的适用性评价与多流体模型开发：（1）就ATHLET程序对超临界水冷堆和钠冷快堆的适用性进行分析；（2）增加超临界水的物性计算、传热、压降、临界流模型；（3）增加钠的物性计算、传热、压降模型；（4）引入多流体通用接口，将针对超临界水和钠两种流体的程序扩展整合为多流体模型，获得ATHLET-MF程序版本。
2. ATHLET-MF程序多流体模型的初步验证与评估：（1）超临界水冷堆扩展功能的初步验证，包括与理论计算对比、程序比对验证；（2）钠冷快堆扩展功能的评估，即PHENIX反应堆自然对流实验验证。结果表明：ATHLET-MF程序对超临界水冷堆和钠冷快堆系统的瞬态模拟具有良好的适用性和可靠性。
3. ATHLET-MF程序多流体模型的应用，即超临界水冷堆燃料性能验证实验回路的安全分析：（1）计算分析了5类设计基准事故，包括冷却剂丧失、主泵卡轴、实验回路断电、冷却剂流道堵塞、以及由于压力管内部构件破裂导致的冷却剂旁流事故；（2）针对初始安全系统设计的缺陷提出改进措施：增加了2个安全信号，提出长期余热排出方案并进行了可行性分析；（3）对安全系统的部分设计参数进行了敏感性分析，提出对安全系统控制信号及设备参数的建议。
本文旨在开发和验证适用于超临界水冷堆和钠冷快堆的系统安全分析工具，为超临界水冷堆燃料性能验证实验回路进行安全分析，探索超临界水冷系统的动态和安全特性，提出安全系统设计的改进方向。本文为超临界水冷堆燃料性能验证实验回路的设计和安全许可证申请提供了安全分析工具和分析结果，对该实验回路安全系统的设计具有现实指导意义。
System safety analysis code is an important tool to perform transientsand accidents safety analysis for reactor systems. Under the context ofglobal R&D upsurge about Generation IV innovative nuclear systems,plenty of work has been done for Gen-IV reactor system codedevelopment. Because light water reactor system codes are well developedwith systematic V&V and extensive application, some of them aremodified for Gen-IV nuclear systems. However, feasibility and realiabilityof these modified codes are still to be further confirmed.
This dissertation aims at development, validation and application ofthe system code for Supercritical water-cooled reactor (SCWR) andSodium-cooled fast reactor (SFR). Firstly, assess the applicability of thelight water reactor system code ATHLET for these two reactor types, anddevelop the multi-fluid model for supercritical water and sodium, to obtainthe system code ATHLET-MF (multi-fluid). Secondly, preliminaryverification of the multi-fluid model and validation with the French SFRPHENIX’s natural convection test. Thirdly, application of theATHLET-MF code to safety analysis and safety system designimprovement of the SCWR Fuel Qualification Test (SCWR-FQT) Loop.
The main contents of this dissertation include:
1. Applicability assessment of the ATHLET code and development ofthe multi-fluid model.(1) Applicability analysis of the ATHLET code forSCWR and SFR.(2) Implementation of property model, heat transfermodel, pressure drop model and critical flow model for supercritical water.(3) Implementation of property model, heat transfer model and pressuredrop model for sodium.(4) Introduction of the multi-fluid generalinterface to integrate code extensions for supercritical water and sodiuminto the multi-fluid model, and obtain the code version ATHLET-MF.
2. Preliminary verification and validation of the multi-fluid model ofthe ATHLET-MF code.(1) Validation of the supercritical watermodifications by comparson with analytical calculation and other codes.(2) Validation of the sodium modifications with the SFR PHENIX naturalconvection test. Results show good applicability and accuration of theATHLET-MF code in transient simulation of SCWR and SFR systems.
3. Application of the ATHLET-MF code to safety analysis of theSCWR-FQT loop.(1)5types of design basis accidents are simulated,including loss of coolant accident, primary pump seize, loss of power tothe test loop, blockage of any coolant supply line, and coolant bypassingthe test section due to crack in the internal structures of the pressure tube.(2) Improvement of the initial safety system design,2safety signals areadded, and the long term residual heat removal strategy is proposed and itsfeasibility is analyzed.(3) Sensitivity analysis of some design parametersof the safety system, and suggestions to control signal design as well as equipment selection are proposed.
This dissertation focuses on development and validation of thesystem safety analysis tool for Gen-IV nuclear systems, mainly for SCWRand SFR, and its engineering application to safety analysis and safetysystem design improvement of the SCWR-FQT loop. The outcomings ofthis dissertation provide safety analysis tools and analysis results fordesign and licensing the SCWR-FQT loop, which give realistic referenceand guidance for the safty system design of the test loop.
本文针对第四代反应堆中的超临界水冷堆和钠冷快堆进行系统程序的模型开发、验证与应用研究。首先对轻水堆系统程序ATHLET在这两种第四代堆型上的适用性进行评价，在此基础上对ATHLET程序进行二次模型开发，即开发适用于超临界水和钠两种流体的相关计算模型，获得多流体系统程序ATHLET-MF（Multi-Fluid）；其次，对修改后的ATHLET-MF程序进行初步验证，并采用法国PHENIX钠冷快堆的自然对流实验对程序进行评估；最后，将ATHLET-MF程序应用于超临界水冷堆燃料性能验证实验回路的安全分析，并提出对该实验回路的安全系统设计的改进意见。
论文的主要工作包括：
1. ATHLET程序的适用性评价与多流体模型开发：（1）就ATHLET程序对超临界水冷堆和钠冷快堆的适用性进行分析；（2）增加超临界水的物性计算、传热、压降、临界流模型；（3）增加钠的物性计算、传热、压降模型；（4）引入多流体通用接口，将针对超临界水和钠两种流体的程序扩展整合为多流体模型，获得ATHLET-MF程序版本。
2. ATHLET-MF程序多流体模型的初步验证与评估：（1）超临界水冷堆扩展功能的初步验证，包括与理论计算对比、程序比对验证；（2）钠冷快堆扩展功能的评估，即PHENIX反应堆自然对流实验验证。结果表明：ATHLET-MF程序对超临界水冷堆和钠冷快堆系统的瞬态模拟具有良好的适用性和可靠性。
3. ATHLET-MF程序多流体模型的应用，即超临界水冷堆燃料性能验证实验回路的安全分析：（1）计算分析了5类设计基准事故，包括冷却剂丧失、主泵卡轴、实验回路断电、冷却剂流道堵塞、以及由于压力管内部构件破裂导致的冷却剂旁流事故；（2）针对初始安全系统设计的缺陷提出改进措施：增加了2个安全信号，提出长期余热排出方案并进行了可行性分析；（3）对安全系统的部分设计参数进行了敏感性分析，提出对安全系统控制信号及设备参数的建议。
本文旨在开发和验证适用于超临界水冷堆和钠冷快堆的系统安全分析工具，为超临界水冷堆燃料性能验证实验回路进行安全分析，探索超临界水冷系统的动态和安全特性，提出安全系统设计的改进方向。本文为超临界水冷堆燃料性能验证实验回路的设计和安全许可证申请提供了安全分析工具和分析结果，对该实验回路安全系统的设计具有现实指导意义。
System safety analysis code is an important tool to perform transientsand accidents safety analysis for reactor systems. Under the context ofglobal R&D upsurge about Generation IV innovative nuclear systems,plenty of work has been done for Gen-IV reactor system codedevelopment. Because light water reactor system codes are well developedwith systematic V&V and extensive application, some of them aremodified for Gen-IV nuclear systems. However, feasibility and realiabilityof these modified codes are still to be further confirmed.
This dissertation aims at development, validation and application ofthe system code for Supercritical water-cooled reactor (SCWR) andSodium-cooled fast reactor (SFR). Firstly, assess the applicability of thelight water reactor system code ATHLET for these two reactor types, anddevelop the multi-fluid model for supercritical water and sodium, to obtainthe system code ATHLET-MF (multi-fluid). Secondly, preliminaryverification of the multi-fluid model and validation with the French SFRPHENIX’s natural convection test. Thirdly, application of theATHLET-MF code to safety analysis and safety system designimprovement of the SCWR Fuel Qualification Test (SCWR-FQT) Loop.
The main contents of this dissertation include:
1. Applicability assessment of the ATHLET code and development ofthe multi-fluid model.(1) Applicability analysis of the ATHLET code forSCWR and SFR.(2) Implementation of property model, heat transfermodel, pressure drop model and critical flow model for supercritical water.(3) Implementation of property model, heat transfer model and pressuredrop model for sodium.(4) Introduction of the multi-fluid generalinterface to integrate code extensions for supercritical water and sodiuminto the multi-fluid model, and obtain the code version ATHLET-MF.
2. Preliminary verification and validation of the multi-fluid model ofthe ATHLET-MF code.(1) Validation of the supercritical watermodifications by comparson with analytical calculation and other codes.(2) Validation of the sodium modifications with the SFR PHENIX naturalconvection test. Results show good applicability and accuration of theATHLET-MF code in transient simulation of SCWR and SFR systems.
3. Application of the ATHLET-MF code to safety analysis of theSCWR-FQT loop.(1)5types of design basis accidents are simulated,including loss of coolant accident, primary pump seize, loss of power tothe test loop, blockage of any coolant supply line, and coolant bypassingthe test section due to crack in the internal structures of the pressure tube.(2) Improvement of the initial safety system design,2safety signals areadded, and the long term residual heat removal strategy is proposed and itsfeasibility is analyzed.(3) Sensitivity analysis of some design parametersof the safety system, and suggestions to control signal design as well as equipment selection are proposed.
This dissertation focuses on development and validation of thesystem safety analysis tool for Gen-IV nuclear systems, mainly for SCWRand SFR, and its engineering application to safety analysis and safetysystem design improvement of the SCWR-FQT loop. The outcomings ofthis dissertation provide safety analysis tools and analysis results fordesign and licensing the SCWR-FQT loop, which give realistic referenceand guidance for the safty system design of the test loop.
引文
[1] U.S. DOE Nuclear Energy Research Advisory Committee and the GenerationIV International Forum. A Technology Roadmap for Generation IV NuclearEnergy Systems [R]. U.S.,2002.
[2] U.S. DOE Nuclear Energy Research Advisory Committee and the GenerationIV International Forum. GIF R&D Outlook for Generation IV Nuclear EnergySystems [R]. U.S.,2009.
[3] Oka Y., Koshizuka S. Conceptual Design of a Supercritical-pressure, DirectCycle Light Water Reactor [J]. Nuclear Technology,1993,103:295-302.
[4] Jevremovic T., Oka Y., Koshizuka S. Core Design of a Direct-cycle,Supercritical-water-cooled Fast Breeder Reactor [J]. Nuclear Technology,1994,108:24-32.
[5] Oka Y., Jevremovic T., Koshizuka S. A Direct-Cycle, Supercritical-Water-Cooled Fast Breeder Reactor [J]. Journal of Nuclear Science andTechnology,1994,108:83-85.
[6] Oka Y., Koshizuka S., Jevremovic T. et al. Systems Design of Direct-cycleSupercritical-water-cooled Fast Reactors [J]. Nuclear Technology,1995,109:1-10.
[7] Mukohara T., Koshizuka S., Oka Y. Core Design of a High-temperature FastReactor Cooled by Supercritical Light Water [J]. Annals of Nuclear Energy,1999,26:1423-1436.
[8] Oka Y. Review of High Temperature Water and Steam Cooled ReactorConcepts, Proceedings of the1st International Symposium on SupercriticalWater-cooled Reactors, Design and Technology,2000[C]. Tokyo, Japan,2000.
[9] Oka Y., Koshizuka S. Design Concept of Once-through Cycle SupercriticalPressure Light Water Cooled Reactors, Proceedings of the1st InternationalSymposium on Supercritical Water-cooled Reactors, Design and Technology,2000[C]. Tokyo, Japan,2000.
[10] Oka Y., Koshizuka S., Ishiwatari Y. Elements of Design Consideration ofOnce-through Cycle, Supercritical Pressure Light Water Cooled Reactor,Proceedings of the2002International Congress on Advanced Nuclear PowerPlants (ICAPP’02),2002[C].Hollywood, USA,2002.
[11] Ishiwatari Y. Japanese R&D Projects on Pressure-vessel Type SCWR,Proceedings of the4th International Symposium on Supercritical Water-cooledReactors,2009[C]. Heidelberg, Germany,2009.
[12] Buongiorno J., MacDonald P.E. Progress Report for the FY-03Generation-IVR&D Activities for the Development of the SCWR in the U.S.,INEEL/EXT-03-01210[R]. Idaho National Engineering and EnvironmentalLaboratory,2003.
[13] MacDonald P.E., Buongiorno J., Sterbentz J.W. et al. Feasibility Study ofSupercritical Light Water Cooled Reactors for Electric Power Production, FinalReport,12th Quarterly Report, INEEL/EXT-04-02530[R]. Idaho NationalEngineering and Environmental Laboratory,2005.
[14] Modro S.M. The Supercritical Water Cooled Reactor Research andDevelopment in the U.S., Proceedings of the2005International Congress onAdvances in Nuclear Power Plants (ICAPP’05),2005[C]. Seoul, Korea,2005.
[15] Rimpault G. Core Design Feature Studies and Research Needs for HighPerformance Light Water Reactors, Proceedings of the2003InternationalCongress on Advanced Nuclear Power Plants (ICAPP’03),2003[C].Cordoba,Spain,2003.
[16] Schulenberg T, Starflinger J, Heinecke J. Three Pass Core Design Proposal fora High Performance Light Water Reactor [J]. Progress in Nuclear Energy,2008,50:526-531.
[17]李满昌,王明利.超临界水冷堆开发现状与前景展望[J].核动力工程,2006,27(2):1-4.
[18]程旭,刘晓晶.超临界水冷堆国内外研发现状与趋势[J].原子能科学技术,2008,42(2):167-172.
[19]欧阳予.世界核电国家的发展战略历程与我国核电发展[J].中国核电,2008,1(2):118-125.
[20]欧阳予.世界核电国家的发展战略历程与我国核电发展[J].中国核电,2008,1(3):194-201.
[21]欧阳予.先进核能技术研究新进展[J].中国核电,2009,2:98-105.
[22] Cheng X. R&D Activities on SCWR in China, Proceedings of the4thInternational Symposium on Supercritical Water-Cooled Reactors,2009[C].Heidelberg, Germany,2009.
[23] IAEA. Status of Liquid Metal Cooled Fast Reactor Technology,IAEA-TECDOC-1083[R]. IAEA,1999.
[24] IAEA. LMFR LMFR Core Thermohydraulics: Status and Prospects,IAEA-TECDOC-1157[R]. IAEA,2000.
[25] IAEA. Fast Reactor Database2006Update, IAEA-TECDOC-1531[R]. IAEA,2006.
[26] IAEA. Fast Reactors and Related Fuel Cycles: Challenges and Opportunities(FR09), Proceedings of an International Conference, Kyoto, Japan,7-11December2009[M]. Austria: IAEA,2012.
[27] Niwa H., Fiorini G.L., Sim Y.S. et al. Status of the Design and Safety Projectfor the Sodium-cooled Fast Reactor as a Generation IV Nuclear Energy System,Proceedings of GLOBAL2005,2005[C]. Tsukuba, Japan,2005.
[28] Ichimiya M. The Status of Generation IV Sodium-cooled Fast ReactorTechnology Development and its Future Project [J], Energy Procedia,2011,7:79-87.
[29]徐銤.中国实验快堆的安全特性[J].核科学与工程,2011,31(2):116-126.
[30]徐銤.我国快堆和第4代先进核能系统[J].中国原子能科学研究院年报,2006:3-4.
[31]徐銤.我国快堆发展战略目标研究[J].核科学与工程,2008,28(1):20-25.
[32] Xu M. Fast Reactor Development for a Sustainable Nuclear Energy Supply inChina, Proceedings of the International Conference of Fast Reactors andRelated Fuel Cycles: Challenges and Opportunities,2009[C]. Kyoto, Japan,2009.
[33]徐銤.快堆和我国核能的可持续发展[J].中国核电,2009,2(2):106-110.
[34]严陈昌.钠冷快堆标准现状及其体系预研工作的初步设想[J].核标准计量与质量,2012,1:2-7.
[35] Jones W. P. Air Conditioning Engineering: Fourth Edition [M]. Great Britain:Edward Arnold,1994.
[36] Pioro I.L., Duffey R.B. Heat Transfer and Hydraulic Resistance at SupercriticalPressure in Power-engineering Applications [M]. US: ASME Press,2007.
[37] Lee J.H., Koshizuka S., Oka Y. Development of a LOCA Analysis Code for theSupercritical Pressure Light Water Cooled Reactors [J]. Annals of NuclearEnergy,1998,25:1341-1361.
[38] Koshizuka S., Shimamura K., Oka, Y. Large-break Loss-of-coolant AccidentAnalysis of a Direct-cycle Supercritical-pressure Light Water Reactor [J].Annals of Nuclear Energy,1994,21:177-187.
[39] Okano Y., Koshizuka S., Oka Y. Core Design of a Direct-Cycle,Supercritical-Pressure, Light Water Reactor with Double Tube Water Rods [J].Journal of Nuclear Science and Technology,1996,33:365-373.
[40] Ishiwatari Y., Oka Y., Koshizuka S. Control of a High TemperatureSupercritical Pressure Light Water Cooled and Moderated Reactor with WaterRods [J]. Journal of Nuclear Science and Technology,2003,40:298-306.
[41] Ishiwatari Y., Oka Y., Koshizuka S. et al. Safety of Super LWR,(I) SafetySystem Design [J]. Journal of Nuclear Science and Technology,2005,42:927-934.
[42] Ishiwatari Y., Oka Y., Koshizuka S. et al. Safety of Super LWR,(II) SafetyAnalysis at Supercritical Pressure [J]. Journal of Nuclear Science andTechnology,2005,42:935-948.
[43] Ishiwatari Y., Oka Y., Koshizuka S. et al. LOCA Analysis of Super LWR [J].Journal of Nuclear Science and Technology,2006,43:231-241.
[44] Ishiwatari Y., Oka Y., Koshizuka S. et al. ATWS Characteristics of Super LWRwith/without Alternative Action [J]. Journal of Nuclear Science and Technology,2007,44:2335-2341.
[45] Ikejiri S., Ishiwatari Y., Oka Y. Safety Analysis of a Supercritical-pressureWater-cooled Fast Reactor under Supercritical Pressure [J]. NuclearEngineering and Design,2010,240:1218-1228.
[46] The RELAP5-3D Code Development Team. RELAP5-3D Code ManualVolume I: Code Structure, System Models, and Solution Methods, Revision2.4,INEEL-EXT-98-00834[R]. Idaho National Engineering and EnvironmentalLaboratory,2005.
[47] The RELAP5-3D Code Development Team. RELAP5-3D Code ManualVolume IV: Models and Correlations, Revision2.3, INEEL-EXT-98-00834[R].Idaho National Engineering and Environmental Laboratory,2005.
[48] Riemke R., Davis C., Schultz R. RELAP5-3D Code for Supercritical PressureLight Water Cooled Reactors, Proceedings of the11th International Conferenceon Nuclear Engineering (ICONE-11),2003[C]. Tokyo, Japan,2003.
[49] Buongiorno J., Davis C. Validation of Existing Thermal-Hydraulic Codes forUse in Supercritical-Water Cooled Reactor Studies, NE-INERI-2001001[R].Idaho National Engineering and Environmental Laboratory,2001.
[50] Davis C., Boungiorno J., MacDonald P. A Parametric Study of theThermal-Hydraulic Response of Supercritical Light Water Reactors duringLoss-of-feedwater and Turbine-trip Events, Proceedings of GENES4/ANP2003,2003[C]. Kyoto, Japan,2003.
[51] MacDonald P., Buongiorno J., Davis C. et al. Feasibility Study of SupercriticalLight Water Cooled Fast Reactors for Actinide Burning and Electric PowerProduction, Progress Report for Work through September2002,3rd QuarterlyReport, INEEL/EXT-02-00925[R]. Idaho National Engineering andEnvironmental Laboratory,2002.
[52] MacDonald P., Buongiorno J., Davis C. et al. Feasibility Study of SupercriticalLight Water Cooled Fast Reactors for Actinide Burning and Electric PowerProduction, Progress Report for Work through September2002,4th QuarterlyReport, INEEL/EXT-02-01330[R]. Idaho National Engineering andEnvironmental Laboratory,2002.
[53] Buongiorno J., MacDonald P. E. Study of Solid Moderators for theThermal-Spectrum Supercritical Water-Cooled Reactor (Neutronics),Proceedings of the11th International Conference on Nuclear Engineering(ICONE-11),2003[C]. Tokyo, Japan,2003.
[54] Buongiorno J., MacDonald P. E. Study of Solid Moderators for theThermal-Spectrum Supercritical Water-Cooled Reactor (Thermo-mechanicsand Cost), Proceedings of the11th International Conference on NuclearEngineering (ICONE-11),2003[C]. Tokyo, Japan,2003.
[55] MacDonald P.E., Buongiorno J., Davis C. et al. Feasibility Study ofSupercritical Light Water Cooled Reactors for Electric Power Production,Progress Report for Work through September2003,2nd Annual Report,8thQuarterly Report, INEEL/EXT-03-01277[R]. Idaho National Engineering andEnvironmental Laboratory,2003.
[56] Bishop A. A., Sandberg R. O., Tong L. S. Forced Convection Heat Transfer toWater near Critical Temperature and Supercritical Pressures, WCAP-2056-P,Part-III-B [R]. U.S.,1964.
[57] Koshizuka S., Oka Y. Computational Analysis of Deterioration Phenomena andThermal-hydraulic Design of SCR, Proceedings of the1st InternationalSymposium on Supercritical Water-cooled Reactors, Design and Technology,2000[C]. Tokyo, Japan,2000.
[58] Jackson J. D. Consideration of the Heat Transfer Properties of SupercriticalPressure Water in Connection with the Cooling of Advanced Nuclear Reactors,Proceedings of the13th Pacific Basin Nuclear Conference (PBNC-13),2002
[C]. Shenzhen, China,2002.
[59] Jackson J. D. Mixed Forced and Free Convection–The Influence of Buoyancyon Turbulent Forced Flow in Vertical Pipes, Proceedings of the EuromechMeeting on Boundary Layers and Turbulence in Internal Flows,1979[C].Salford, England,1979.
[60] Petrov N. E., Popov V. N. Heat Transfer and Hydraulic Resistance withTurbulent Flow in a Tube of Water at Supercritical Parameters of State [J].Thermal Engineering,1988,35:577-580.
[61] Division of Risk Assessment and Special Projects, Office of NuclearRegulatory Research, U.S. Nuclear Regulatory Commission. TRACE V5.0Theory Manuel: Field Equations, Solution Methods, and Physical Models [R].U.S.,2007.
[62] Monti L., Schulenberg T., Jaeger W. et al. Application and Improvements of theSystem Code TRACE for HPLWR Core Analyses, Proceedings of the7thInternational Topical Meeting on Nuclear Reactor Thermal Hydraulics,Operation and Safety (NUTHOS-7),2008[C]. Seoul, Korea,2008.
[63] J ger W., Sánchez Espinoza V.H., Hurtado A. Investigations of Experimentswith Supercitical Water with the System Code TRACE, Proceedings of the7thInternational Topical Meeting on Nuclear Reactor Thermal Hydraulics,Operation and Safety (NUTHOS-7),2008[C]. Seoul, Korea,2008.
[64] Monti L. Multi-scale, Coupled Reactor Physics/Thermal-Hydraulics Systemand Applications to the HPLWR3Pass Core [D]. Karlsruhe Institute ofTechnology, Germany,2009.
[65] J ger W., Sánchez Espinoza V.H., Hurtado A. Review and Proposal for HeatTransfer Predictions at Supercritical Water Conditions Using ExistingCorrelations and Experiments [J]. Nuclear Engineering and Design,2011,241:2184-2203.
[66] Squarer D., Schulenberg T., Struwe D. et al. High Performance Light WaterReactor [J]. Nuclear Engineering and Design,2003,221:167-180.
[67] Schulenberg T., Starflinger J. Core Design Concepts for High PerformanceLight Water Reactors [J]. Nuclear Engineering and Technology,2007,39:249-256.
[68] Schulenberg T., Starflinger J., Heinecke J. Three Pass Core Design Proposal fora High Performance Light Water Reactor [J]. Progress in Nuclear Engineering,2008,50:526-531.
[69] Barre F., Bernard M. The CATHARE Code Strategy and Assessment [J].Nuclear Engineering and Design,1990,124:257-284
[70] Geffraye G., Antoni O., Farvacque M. et al. CATHARE2V2.52: A SingleVersion for Various Applications [J]. Nuclear Engineering and Design,2011,241:4456-4463.
[71] Dumaz P., Antoni O. The Extension of the CATHARE Computer Code Abovethe Critical Point, Applications to a Supercritical Light Water Reactor,Proceedings of the10th International Topical Meeting on Nuclear ReactorThermal Hydraulics (NURETH-10),2003[C]. Seoul, Korea,2003.
[72] Robert M., Farvacque M., Parent M. et al. CATHARE2V2.5: a Fully ValidatedCATHARE Version for Various Applications Proceedings of the10thInternational Topical Meeting on Nuclear Reactor Thermal Hydraulics(NURETH-10),2003[C]. Seoul, Korea,2003.
[73] Antoni O., Dumaz P. Preliminary Calculations of a Supercritical Light WaterReactor Concept Using the CATHARE Code, Proceedings of the2003International Congress on Advanced Nuclear Power Plants (ICAPP’03),2003
[C]. Cordoba, Spain,2003.
[74] Marsault Ph., Renault C., Rimpault G. et al. Pre-design Studies of SCWR inFast Neutron Spectrum: Evaluation of Operating Conditions and Analysis ofthe Behavior in Accidental Situations, Proceedings of the2004InternationalCongress on Advanced Nuclear Power Plants (ICAPP’04),2004[C]. Pittsburgh,U.S.,2004.
[75] Manera A., Antoni O. Code-to-Code Comparison for Blow-down Transients atSupercritical Conditions, Proceedings of the Annual Meeting on NuclearTechnology,2008[C]. Hamburg, Germany,2008.
[76] Alobaid F., Strohle J., Epple B. et al. Dynamic Simulation of a SupercriticalOnce-through Heat Recovery Steam Generator During Load Changes and StartUp Procedures [J]. Applied Energy,2009,86:1274-1282.
[77]蔡宝玲,王哲,魏湘等.基于引进仿真支撑软件APROS的电站仿真培训系统[J].热力发电,2004,11:7-9.
[78]蔡宝玲,王哲,魏湘等.超临界600MW机组仿真系统动态数学模型的开发及其分析[J].热力发电,2006,2:34-36.
[79]杨冬,马彦花,潘杰等.600MW超临界循环流化床锅炉水冷壁动态特性的研究[J].动力工程,2009,29(8):722-727.
[80]周东阳,魏湘,蔡宝玲等.超超临界1000MW机组仿真系统开发与应用[J].热力发电,2011,11:76-77.
[81] H nninen M., Kurki J. Simulation of Flows at Supercritical Pressures with aTwo-fluid Code, Proceedings of the7th International Topical Meeting onNuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-7),2008[C]. Seoul, Korea,2008.
[82] Kurki J. Simulation of Thermal Hydraulics at Supercritical Pressures withAPROS [D]. Helsinki University of Technology, Finland,2008.
[83] H nninen M. Phenomenological extensions to APROS six-equation modelNon-condensable gas, supercritical pressure, improved CCFL and reducednumerical diffusion for scalar transport calculation [D]. LappeenrantaUniversity of Technology, Finland,2009.
[84] Kurki J., Sepp l M. Thermal Hydraulic Transient Analysis of the HighPerformance Light Water Reactor Using APROS and SMABRE, Proceedingsof the20th International Conference on Structural Mechanics in ReactorTechnology,2009[C]. Espoo, Finland,2009.
[85] Kurki J., H nninen M. On the Applicability of a One-dimensional Two-fluidFlow Model to Simulation of Flows and Heat Transfer at Near and SupercriticalPressures, Proceedings of the8th International Topical Meeting on NuclearReactor Thermal Hydraulics, Operation and Safety (NUTHOS-8),2010[C].Shanghai, China,2010.
[86] Kurki J., H nninen M. Simulation of Large-break Loss of Coolant Accident inthe High Performance Light Water Reactor, Proceedings of the IAEA TechnicalMeeting on Heat Transfer, Thermal-Hydraulics and System Design forSupercritical Water Cooled Reactors,2010[C]. Pisa, Italy,2010.
[87] Andreani M., Bittermann D., Marsault Ph. et al. Evaluation of a PreliminarySafety Concept for the HPLWR [J]. Progress in Nuclear Energy,2012,55:68-77.
[88] Johnsen G., Davis C., Bayless P. ATHENA-3D for Generation IV ReactorAnalysis, BE-2004, Proceedings of the ANS Winter Meeting,2004[C].Washington D.C., U.S.,2004.
[89] Riemke R., Davis C., Schultz R. RELAP5-3D Code Includes Athena Featuresand Models, Proceedings of the14th International Conference on NuclearEngineering (ICONE-14),2006[C]. Miami, U.S.,2006.
[90] Sumner T., Ghiaasiaan S.M. Effects of fuel type on the safety characteristics ofa sodium-cooled fast reactor: Part I: Background, modeling tools andpre-transient calculations [J]. Annals of Nuclear Energy,2011,38:1559-1568.
[91] Sumner T., Ghiaasiaan S.M. Effects of fuel type on the safety characteristics ofa sodium cooled fast reactor. Part II: Simulation results [J]. Annals of NuclearEnergy,2011,38:1760-1768.
[92] Spore J.W. TRAC-M/FORTRAN90(V3.0) Theory Manual LAUR-00-910[R].Los Alamos National Laboratory/Penn State University, U.S.,2000.
[93] Mikityuk K., Pelloni S., Coddington P. et al. FAST: An advanced code systemfor fast reactor transient analysis [J]. Annals of Nuclear Energy,2005,32:1613-1631.
[94] Chenu A., Mikityuka K., Chawla R. TRACE simulation of sodium boiling inpin bundle experiments under loss-of-flow conditions [J]. Nuclear Engineeringand Design,2009,239:2417-2429.
[95] Chenu A., Adams R., Mikityuk K., et al. Analysis of selected Phenix EOL testswith the FAST code system, Part I: control-rod-shift experiments [J]. Annals ofNuclear Energy,2012,49:182-190.
[96] Chenua A., Mikityuka K., Chawlaa R. Analysis of selected Phenix EOL testswith the FAST code system, Part II: Unprotected phase of the NaturalConvection Test [J]. Annals of Nuclear Energy,2012,49:191-199.
[97] Tenchine D., Baviere R., Bazin P., et al. Status of CATHARE code for sodiumcooled fast reactors [J]. Nuclear Engineering and Design,2012,245:140-152.
[98] Mochizuki H. Verification of NETFLOW code using plant data of sodiumcooled reactor and facility [J]. Nuclear Engineering and Design,2007,237:87-93.
[99] Mochizuki H. Inter-subassembly heat transfer of sodium cooled fast reactors:validation of the NETFLOW code [J]. Nuclear Engineering and Design,2007,237:2040-2053.
[100] Mochizuki H. Development of the plant dynamics analysis code NETFLOW++[J]. Nuclear Engineering and Design,2010,240:577-587.
[101] Mochizuki H. Plant behavior of a fast breeder reactor under loss of AC powerfor long period [J]. Nuclear Engineering and Design,2012,245:19-27.
[102] Tobita Y., S Kondo., Yamano H. et al. The Development of SIMMER-III, anAdvanced Computer Program for LMFR Safety Analysis, Proceedings of theJoint IAEA/NEA Technical Meeting on Use of Computational Fluid Dynamics(CFD) Codes for Safety Analysis Reactors Systems Including Containment,2002[C]. Pisa, Italy,2002.
[103] Tobita Y., Kondo S., Yamano H., et al. The development of SIMMER-III, anadvanced computer program for lmfr safety analysis, and its application tosodium [J]. Nuclear Technology,2006,153:245-255.
[104] Yamano H., Fujita S., Tobita Y., et al. Development of a three-dimensionalCDA analysis code: SIMMER-IV and its first application to reactor case [J].Nuclear Engineering and Design,2008,238:66-73.
[105] Maschek W., Rineiski A., Suzuki T., et al. SIMMER-III and SIMMER-IVSafety Code Development for Reactors with Transmutation Capability,Proceedings of the Mathematics and Computation, Supercomputing, ReactorPhysics and Biological Applications,2005[C]. Palais de Papes, France,2005.
[106] Suzuki T., Chen X.N., Rineiski A., et al. Transient analyses for acceleratordriven system PDS-XADS using the extended SIMMER-III code [J]. NuclearEngineering and Design,2005,235:2594-2611.
[107] Dunn F.E., Prohammer F.G. The SASSYS LMFBR systems analysis code [J].Mathematics and Computers in Simulation,1984,26:23-26.
[108] Ha K.S., Jeong H.Y., Cho C., et al. Simulation of the EBR-II Loss-of-FlowTests Using the MARS Code [J]. Nuclear Technology,2010,169:134-142.
[109] Jeong H.Y., Ha K.S., Lee K.L., et al. Pre-Test Analysis of Natural CirculationTest of PHENIX End-of-Life With the MARS-LMR Code, Proceedings of the18th International Conference on Nuclear Engineering (ICONE-18),2010[C].Xi'an, China,2010.
[110] Woo S.M., Chang S.H. Multi dimensional analysis of Design Basis Eventsusing MARS-LMR [J]. Nuclear Engineering and Design,2012,244:83-91.
[111] Guppy J.G. Super System Code (SSC, Rev.0) An Advanced ThermohydraulicSimulation Code for Transients in LMFBRs, NUREG/CR-3169(BNL-NUREG-51650)[R]. U.S.,1983.
[112]杨红义,徐銤. OASIS程序的开发与应用[J].核科学与工程,2001,21(4):322-340.
[113]许义军.中国实验快堆钠池三维热工水力分析[D].中国原子能科学研究院,2003.
[114]张春明. DINROS程序在中国实验快堆事故分析中的应用[J].核科学与工程,2006,26(2):142-148.
[115]任丽霞.钠冷快堆系统分析程序实用开发[D].中国原子能科学研究院,2007.
[116]李硕.钠冷堆瞬态安全分析方法与初步研究[D].华北电力大学（北京）,2011.
[117]陆道纲,隋丹婷,任丽霞等.池式快堆系统分析软件稳态功能开发[J].原子能科学技术,2012,46(4):422-428.
[118]隋丹婷,陆道纲,张盼.用于池式快堆系统分析的钠池三维模型开发[J].原子能科学技术,2012,46(4):429-436.
[119]陆道纲,隋丹婷.池式快堆系统瞬态分析软件开发[J].原子能科学技术,2012,46(5):542-548.
[120] Lerchel G., Austregesilo H., ATHLET Mod2.1Cycle A, User's Manual,GRS-p-1/Vol.1, Rev.4[R]. GRS,2006.
[121] Ulrych F. Darstellung der IFC-Wasser-Dampf-Zustandsgleichungen alsALGOL-Prozeduren bzw. FORTRAN-Funktionen (D2O und H2O), TS116Nr.1/68[R]. Deutschland, Siemens,1968.
[122] The IFC-Formulation for Industrial Use. A Formulation of the ThermodynamicProperties of Ordinary Water Substance [M]. Germany: VDI-Verlag,1967.
[123] Müller W.Ch. Fast and accurate water and steam properties programs fortwo-phase flow calculations [J]. Nuclear Engineering and Design,1994,149:449-458.
[124] Ball A., Trambauer K. Stoffwertapproximation der Transportgr en vonWasser für Temperaturen bis3000°C, GRS-A-1710[R]. Deutschland, GRS,1990.
[125] Goldammer H.D. H2OSTA-A General Purpose Software Package for theCalculation of Thermodynamic Properties of Water/Steam, Final Report forContract Nr.3048-86-10[R] Germany, EP/ISP D,1987.
[126] MATPRO-Version11(Rev.2), A Handbook of Material Properties for Use inthe Analysis of Light Water Reactor Fuel Behaviour, NUREG/CR-0497,TREE-1280, Rev.2[R]. Idaho National Engineering Laboratory,1981.
[127] Revised release on the IAPWS industrial formulation1997for thethermodynamic properties of water and steam [R]. International Association forthe Properties of Water and Steam, Switzerland,2007.
[128] Supplementary Release on Backward Equations for Specific Volume as aFunction of Pressure and Temperature v(p,T) for Region3of the IAPWSIndustrial Formulation1997for the Thermodynamic Properties of Water andSteam [R]. The International Association for the Properties of Water and Steam,Greece,2005.
[129] Fu S.W., Zhou C., Xu Z.H., et al. Modification and Application of theATHLET-SC Code to Trans-critical Simulations, Proceeding of the5thInternational Symposium on Supercritical Water-Cooled Reactors,2011[C].Canada,2011.
[130] Cheng X., Schulenberg T. Heat transfer at supercritical pressure--literaturereview and application to an HPLWR, FZKA6609[R]. Karlsruhe Institute ofTechnology, Germany,2001.
[131] Krasnoshchekov E.A., Protopopov V.S. Experimental study of heat exchange incarbon dioxide in the supercritical range at high temperature drops [J].Teplofizika Vysokikh Temperature,1966,4(3):389-398.
[132] Yamagata K., Nishikawa K., Hasegawa S., et al. Forced convective heattransfer to supercritical water flowing in tubes [J]. International Jounal of HeatMass Transfer,1972,15(12):2575-2593.
[133] Jackson J.D. Validation of an extended heat transfer equation for fluids atsupercritical pressure, Proceedings of the4th International Conference onSupercritical Water-Cooled Reactors,2009[C]. Heidelberg, Germany,2009.
[134] Cheng X., Yang Y.H., Huang S.F. A simplified method for heat transferprediction of supercritical fluids in circular tubes [J]. Annals of Nuclear Energy,2009,36:1120-1128.
[135] Bishop A.A., Sandberg R.O., Tong L.S. High temperature supercritical pressurewater Loop-IV-Forced convection heat transfer to water at near-criticaltemperatures and near supercritical pressures, WCAP-2056, Part IV [R]. U.S.,1964.
[136] Herkanrath H., Mork-Morkenstein P., Jung U., et al. Heat transfer in water withforced circulation in the pressure range from140-250bar, E.U.R.3658d, HTFS916[R]. Joint Nuclear Research Centre, Italy,1967.
[137] Griem H. A new procedure for the prediction of forced convection heat transferat near-and supercritical pressure [J]. International Jounal of Heat and MassTransfer,1996,31:301-305.
[138] Fewster J., Jackson J.D. Experiments on supercritical pressure convective heattransfer having relevance to SPWR, Proceedings of the2004InternationalCongress on Advanced Nuclear Power Plants (ICAPP’04),2004[C]. Pittsburgh,U.S.,2004.
[139] Xu F. Studies on heat transfer characteristics in tubes at supercritical pressures(in Chinese)[D]. Xi’an Jiao Tong University, China,2004.
[140] Cheng X., Gu H.Y. Test data of School of Nuclear Science and Engineering [R].Shanghai Jiao Tong University,2011, Personal communication.
[141] Domin G. W rmeübergang in kritischen und überkritischen Bereichen vonWasser in Rohren BWK15[R].1963.
[142] Chen Y.Z., Yang C.S., Zhao M.F., et al. Experimental Studies on Critical Flowand Heat Transfer of Water for Near-critical and Supercritical Pressures,Proceedings of the IAEA Technical Meeting on Heat Transfer,Thermal-Hydraulics and System Design for Supercritical Water-CooledReactors,2010[C]. Pisa, Italy,2010.
[143] Bystrov P.I., Kagan D.N., Krechetova G.A., et al. Liquid-Metal Collants forHeat Pipes and Power Plants [M]. US: Ed. V. A. Kirillin, Hemisphere Pub.Corp,1990.
[144] Fink J.K., Leibowitz L. Thermodynamic and Transport Properties of SodiumLiquid and vapor, ANL/RE-95/2[R]. Argonne National Laboratory,1995.
[145] Borishanski V.M., Firsova E.V. Heat Transfer in Liquid Metals, FluidMechanics and Heat Transfer of Heat Exchanger Design Handbook, vol.2,chapter2.5.13[M]. US: Hemisphere Publishing Corporation,1990.
[146] Mikityuk K. Heat transfer to liquid metal: Review of data and correlations fortube bundles [J]. Nuclear Engineering and Design,2009,239:680-687.
[147] Cheng X., Tak N. Investigation on turbulent heat transfer to lead-bismutheutectic flows in circular tubes for nuclear applications [J]. NuclearEngineering and Design,2006,236:385-393.
[148] Pfrang W., Struwe D. Assessment of Correlations for Heat Transfer to theCoolant for Heavy Liquid Metal Cooled Core Designs, FZKA7352[R].Karlsruhe Institute of Technology, Germany,2007.
[149] Lyon R.N. Liquid metal heat transfer coefficients [J]. Chemical EngineeringProgress,1951,47:75-79.
[150] Skupinski E., Tortel J., Vautrey L. Tetermination des coefficients de convectionD’un Alliage sodium-potassium Dans un Tube Circulaire [J]. InternationalJounal of Heat Mass Transfer,1965,8:937-951.
[151] Sleicher C.A., Awad A.S., Notter R.H. Temperature and Eddy diffusivityprofiles in NaK [J]. International Jounal of Heat Mass Transfer,1973,16:1565-1575.
[152] Gr ber V.H., Rieger M. Experimentelle Untersuchung des W rmeübergangs anFlüssigmetalle (NaK) in parallel durchstr mten Rohrbündeln bei konstanter undexponentieller W rmeflussdichteverteilung [J]. Atomkernenergie (ATKE) Bd.,1972,19:23-40.
[153] Ushakov P.A., Zhukov A.V., Matyukhin M.M. Heat transfer to liquid metals inregular arrays of fuel elements [J]. High Temperature,1977,15:868-873.
[154] Chenu A., Mikityuka K., Chawla R. Pressure drop modeling and comparisonswith experiments for single-and two-phase sodium flow [J]. NuclearEngineering and Design,2011,241:3898-3909.
[155] Bubelis E., Schikorr M. Review and proposal for best fit of wire-wrapped fuelbundle friction factor and pressure drop predictions using various existingcorrelations [J]. Nuclear Engineering and Design,2008,238:3299-3320.
[156] Edwards A.R., O’Brien T.P. Studies of phenomena connected with thedepressurization of water reactors [J]. Jounal of British Nuclear Energy Society,1970,9:125-135.
[157] Manera A., Antoni O. Code-to-code comparison for blowdown transients atsupercritical conditions, Proceedings of the Annual meeting on nucleartechnology2011[C]. Berlin, Germany,2011.
[158] Cheng X., Yang Y.H. A point-hydraulics model for flow stability analysis [J].Nuclear Engineering and Design,2008,238:188-199.
[159] Cheng X., Kuang B., Yang Y.H. Investigation on Flow Stability ofSupercritical Water Cooled Systems, Proceedings of the2006Internationalcongress on advances in nuclear power plants (ICAPP’06),2006[C]. Reno,U.S.,2006.
[160] Tenchine D. Some thermal hydraulic challenges in sodium cooled fast reactors[J]. Nuclear Engineering and Design,2010,240:1195-1217.
[161] GauthéP. THINS (WP2and WP5): Natural convection test in PHENIX, datalist, test description and initial state, CEA/DEN/CAD/DER/SESI/LSMR/NT,DO6du07/09/10[R]. CEA, France,2010.
[162] Pialla D., Tenchine D., Gauthé P., et al. Natural convection test in Phenixreactor and associated CATHARE calculation, Proceedings of the14thInternational Topical Meeting on Nuclear Reactor Thermal hydraulics [C].Toronto, Canada,2011.
[163] RaquéM., Vojecek A., Hajek P., et al. Design and1D analysis of the safetysystems for the SCWR fuel qualification test, Proceedings of the9thInternational Topical Meeting on Nuclear Reactor Thermal Hydraulics,Operation and Safety (NUTHOS-9),2012[C]. Kaohsiung, Taiwan,2012.
[164] Coffman F. LOCA Temperature criterion for stainless steal clad fuel [R].NUREG-0065,1976.
[2] U.S. DOE Nuclear Energy Research Advisory Committee and the GenerationIV International Forum. GIF R&D Outlook for Generation IV Nuclear EnergySystems [R]. U.S.,2009.
[3] Oka Y., Koshizuka S. Conceptual Design of a Supercritical-pressure, DirectCycle Light Water Reactor [J]. Nuclear Technology,1993,103:295-302.
[4] Jevremovic T., Oka Y., Koshizuka S. Core Design of a Direct-cycle,Supercritical-water-cooled Fast Breeder Reactor [J]. Nuclear Technology,1994,108:24-32.
[5] Oka Y., Jevremovic T., Koshizuka S. A Direct-Cycle, Supercritical-Water-Cooled Fast Breeder Reactor [J]. Journal of Nuclear Science andTechnology,1994,108:83-85.
[6] Oka Y., Koshizuka S., Jevremovic T. et al. Systems Design of Direct-cycleSupercritical-water-cooled Fast Reactors [J]. Nuclear Technology,1995,109:1-10.
[7] Mukohara T., Koshizuka S., Oka Y. Core Design of a High-temperature FastReactor Cooled by Supercritical Light Water [J]. Annals of Nuclear Energy,1999,26:1423-1436.
[8] Oka Y. Review of High Temperature Water and Steam Cooled ReactorConcepts, Proceedings of the1st International Symposium on SupercriticalWater-cooled Reactors, Design and Technology,2000[C]. Tokyo, Japan,2000.
[9] Oka Y., Koshizuka S. Design Concept of Once-through Cycle SupercriticalPressure Light Water Cooled Reactors, Proceedings of the1st InternationalSymposium on Supercritical Water-cooled Reactors, Design and Technology,2000[C]. Tokyo, Japan,2000.
[10] Oka Y., Koshizuka S., Ishiwatari Y. Elements of Design Consideration ofOnce-through Cycle, Supercritical Pressure Light Water Cooled Reactor,Proceedings of the2002International Congress on Advanced Nuclear PowerPlants (ICAPP’02),2002[C].Hollywood, USA,2002.
[11] Ishiwatari Y. Japanese R&D Projects on Pressure-vessel Type SCWR,Proceedings of the4th International Symposium on Supercritical Water-cooledReactors,2009[C]. Heidelberg, Germany,2009.
[12] Buongiorno J., MacDonald P.E. Progress Report for the FY-03Generation-IVR&D Activities for the Development of the SCWR in the U.S.,INEEL/EXT-03-01210[R]. Idaho National Engineering and EnvironmentalLaboratory,2003.
[13] MacDonald P.E., Buongiorno J., Sterbentz J.W. et al. Feasibility Study ofSupercritical Light Water Cooled Reactors for Electric Power Production, FinalReport,12th Quarterly Report, INEEL/EXT-04-02530[R]. Idaho NationalEngineering and Environmental Laboratory,2005.
[14] Modro S.M. The Supercritical Water Cooled Reactor Research andDevelopment in the U.S., Proceedings of the2005International Congress onAdvances in Nuclear Power Plants (ICAPP’05),2005[C]. Seoul, Korea,2005.
[15] Rimpault G. Core Design Feature Studies and Research Needs for HighPerformance Light Water Reactors, Proceedings of the2003InternationalCongress on Advanced Nuclear Power Plants (ICAPP’03),2003[C].Cordoba,Spain,2003.
[16] Schulenberg T, Starflinger J, Heinecke J. Three Pass Core Design Proposal fora High Performance Light Water Reactor [J]. Progress in Nuclear Energy,2008,50:526-531.
[17]李满昌,王明利.超临界水冷堆开发现状与前景展望[J].核动力工程,2006,27(2):1-4.
[18]程旭,刘晓晶.超临界水冷堆国内外研发现状与趋势[J].原子能科学技术,2008,42(2):167-172.
[19]欧阳予.世界核电国家的发展战略历程与我国核电发展[J].中国核电,2008,1(2):118-125.
[20]欧阳予.世界核电国家的发展战略历程与我国核电发展[J].中国核电,2008,1(3):194-201.
[21]欧阳予.先进核能技术研究新进展[J].中国核电,2009,2:98-105.
[22] Cheng X. R&D Activities on SCWR in China, Proceedings of the4thInternational Symposium on Supercritical Water-Cooled Reactors,2009[C].Heidelberg, Germany,2009.
[23] IAEA. Status of Liquid Metal Cooled Fast Reactor Technology,IAEA-TECDOC-1083[R]. IAEA,1999.
[24] IAEA. LMFR LMFR Core Thermohydraulics: Status and Prospects,IAEA-TECDOC-1157[R]. IAEA,2000.
[25] IAEA. Fast Reactor Database2006Update, IAEA-TECDOC-1531[R]. IAEA,2006.
[26] IAEA. Fast Reactors and Related Fuel Cycles: Challenges and Opportunities(FR09), Proceedings of an International Conference, Kyoto, Japan,7-11December2009[M]. Austria: IAEA,2012.
[27] Niwa H., Fiorini G.L., Sim Y.S. et al. Status of the Design and Safety Projectfor the Sodium-cooled Fast Reactor as a Generation IV Nuclear Energy System,Proceedings of GLOBAL2005,2005[C]. Tsukuba, Japan,2005.
[28] Ichimiya M. The Status of Generation IV Sodium-cooled Fast ReactorTechnology Development and its Future Project [J], Energy Procedia,2011,7:79-87.
[29]徐銤.中国实验快堆的安全特性[J].核科学与工程,2011,31(2):116-126.
[30]徐銤.我国快堆和第4代先进核能系统[J].中国原子能科学研究院年报,2006:3-4.
[31]徐銤.我国快堆发展战略目标研究[J].核科学与工程,2008,28(1):20-25.
[32] Xu M. Fast Reactor Development for a Sustainable Nuclear Energy Supply inChina, Proceedings of the International Conference of Fast Reactors andRelated Fuel Cycles: Challenges and Opportunities,2009[C]. Kyoto, Japan,2009.
[33]徐銤.快堆和我国核能的可持续发展[J].中国核电,2009,2(2):106-110.
[34]严陈昌.钠冷快堆标准现状及其体系预研工作的初步设想[J].核标准计量与质量,2012,1:2-7.
[35] Jones W. P. Air Conditioning Engineering: Fourth Edition [M]. Great Britain:Edward Arnold,1994.
[36] Pioro I.L., Duffey R.B. Heat Transfer and Hydraulic Resistance at SupercriticalPressure in Power-engineering Applications [M]. US: ASME Press,2007.
[37] Lee J.H., Koshizuka S., Oka Y. Development of a LOCA Analysis Code for theSupercritical Pressure Light Water Cooled Reactors [J]. Annals of NuclearEnergy,1998,25:1341-1361.
[38] Koshizuka S., Shimamura K., Oka, Y. Large-break Loss-of-coolant AccidentAnalysis of a Direct-cycle Supercritical-pressure Light Water Reactor [J].Annals of Nuclear Energy,1994,21:177-187.
[39] Okano Y., Koshizuka S., Oka Y. Core Design of a Direct-Cycle,Supercritical-Pressure, Light Water Reactor with Double Tube Water Rods [J].Journal of Nuclear Science and Technology,1996,33:365-373.
[40] Ishiwatari Y., Oka Y., Koshizuka S. Control of a High TemperatureSupercritical Pressure Light Water Cooled and Moderated Reactor with WaterRods [J]. Journal of Nuclear Science and Technology,2003,40:298-306.
[41] Ishiwatari Y., Oka Y., Koshizuka S. et al. Safety of Super LWR,(I) SafetySystem Design [J]. Journal of Nuclear Science and Technology,2005,42:927-934.
[42] Ishiwatari Y., Oka Y., Koshizuka S. et al. Safety of Super LWR,(II) SafetyAnalysis at Supercritical Pressure [J]. Journal of Nuclear Science andTechnology,2005,42:935-948.
[43] Ishiwatari Y., Oka Y., Koshizuka S. et al. LOCA Analysis of Super LWR [J].Journal of Nuclear Science and Technology,2006,43:231-241.
[44] Ishiwatari Y., Oka Y., Koshizuka S. et al. ATWS Characteristics of Super LWRwith/without Alternative Action [J]. Journal of Nuclear Science and Technology,2007,44:2335-2341.
[45] Ikejiri S., Ishiwatari Y., Oka Y. Safety Analysis of a Supercritical-pressureWater-cooled Fast Reactor under Supercritical Pressure [J]. NuclearEngineering and Design,2010,240:1218-1228.
[46] The RELAP5-3D Code Development Team. RELAP5-3D Code ManualVolume I: Code Structure, System Models, and Solution Methods, Revision2.4,INEEL-EXT-98-00834[R]. Idaho National Engineering and EnvironmentalLaboratory,2005.
[47] The RELAP5-3D Code Development Team. RELAP5-3D Code ManualVolume IV: Models and Correlations, Revision2.3, INEEL-EXT-98-00834[R].Idaho National Engineering and Environmental Laboratory,2005.
[48] Riemke R., Davis C., Schultz R. RELAP5-3D Code for Supercritical PressureLight Water Cooled Reactors, Proceedings of the11th International Conferenceon Nuclear Engineering (ICONE-11),2003[C]. Tokyo, Japan,2003.
[49] Buongiorno J., Davis C. Validation of Existing Thermal-Hydraulic Codes forUse in Supercritical-Water Cooled Reactor Studies, NE-INERI-2001001[R].Idaho National Engineering and Environmental Laboratory,2001.
[50] Davis C., Boungiorno J., MacDonald P. A Parametric Study of theThermal-Hydraulic Response of Supercritical Light Water Reactors duringLoss-of-feedwater and Turbine-trip Events, Proceedings of GENES4/ANP2003,2003[C]. Kyoto, Japan,2003.
[51] MacDonald P., Buongiorno J., Davis C. et al. Feasibility Study of SupercriticalLight Water Cooled Fast Reactors for Actinide Burning and Electric PowerProduction, Progress Report for Work through September2002,3rd QuarterlyReport, INEEL/EXT-02-00925[R]. Idaho National Engineering andEnvironmental Laboratory,2002.
[52] MacDonald P., Buongiorno J., Davis C. et al. Feasibility Study of SupercriticalLight Water Cooled Fast Reactors for Actinide Burning and Electric PowerProduction, Progress Report for Work through September2002,4th QuarterlyReport, INEEL/EXT-02-01330[R]. Idaho National Engineering andEnvironmental Laboratory,2002.
[53] Buongiorno J., MacDonald P. E. Study of Solid Moderators for theThermal-Spectrum Supercritical Water-Cooled Reactor (Neutronics),Proceedings of the11th International Conference on Nuclear Engineering(ICONE-11),2003[C]. Tokyo, Japan,2003.
[54] Buongiorno J., MacDonald P. E. Study of Solid Moderators for theThermal-Spectrum Supercritical Water-Cooled Reactor (Thermo-mechanicsand Cost), Proceedings of the11th International Conference on NuclearEngineering (ICONE-11),2003[C]. Tokyo, Japan,2003.
[55] MacDonald P.E., Buongiorno J., Davis C. et al. Feasibility Study ofSupercritical Light Water Cooled Reactors for Electric Power Production,Progress Report for Work through September2003,2nd Annual Report,8thQuarterly Report, INEEL/EXT-03-01277[R]. Idaho National Engineering andEnvironmental Laboratory,2003.
[56] Bishop A. A., Sandberg R. O., Tong L. S. Forced Convection Heat Transfer toWater near Critical Temperature and Supercritical Pressures, WCAP-2056-P,Part-III-B [R]. U.S.,1964.
[57] Koshizuka S., Oka Y. Computational Analysis of Deterioration Phenomena andThermal-hydraulic Design of SCR, Proceedings of the1st InternationalSymposium on Supercritical Water-cooled Reactors, Design and Technology,2000[C]. Tokyo, Japan,2000.
[58] Jackson J. D. Consideration of the Heat Transfer Properties of SupercriticalPressure Water in Connection with the Cooling of Advanced Nuclear Reactors,Proceedings of the13th Pacific Basin Nuclear Conference (PBNC-13),2002
[C]. Shenzhen, China,2002.
[59] Jackson J. D. Mixed Forced and Free Convection–The Influence of Buoyancyon Turbulent Forced Flow in Vertical Pipes, Proceedings of the EuromechMeeting on Boundary Layers and Turbulence in Internal Flows,1979[C].Salford, England,1979.
[60] Petrov N. E., Popov V. N. Heat Transfer and Hydraulic Resistance withTurbulent Flow in a Tube of Water at Supercritical Parameters of State [J].Thermal Engineering,1988,35:577-580.
[61] Division of Risk Assessment and Special Projects, Office of NuclearRegulatory Research, U.S. Nuclear Regulatory Commission. TRACE V5.0Theory Manuel: Field Equations, Solution Methods, and Physical Models [R].U.S.,2007.
[62] Monti L., Schulenberg T., Jaeger W. et al. Application and Improvements of theSystem Code TRACE for HPLWR Core Analyses, Proceedings of the7thInternational Topical Meeting on Nuclear Reactor Thermal Hydraulics,Operation and Safety (NUTHOS-7),2008[C]. Seoul, Korea,2008.
[63] J ger W., Sánchez Espinoza V.H., Hurtado A. Investigations of Experimentswith Supercitical Water with the System Code TRACE, Proceedings of the7thInternational Topical Meeting on Nuclear Reactor Thermal Hydraulics,Operation and Safety (NUTHOS-7),2008[C]. Seoul, Korea,2008.
[64] Monti L. Multi-scale, Coupled Reactor Physics/Thermal-Hydraulics Systemand Applications to the HPLWR3Pass Core [D]. Karlsruhe Institute ofTechnology, Germany,2009.
[65] J ger W., Sánchez Espinoza V.H., Hurtado A. Review and Proposal for HeatTransfer Predictions at Supercritical Water Conditions Using ExistingCorrelations and Experiments [J]. Nuclear Engineering and Design,2011,241:2184-2203.
[66] Squarer D., Schulenberg T., Struwe D. et al. High Performance Light WaterReactor [J]. Nuclear Engineering and Design,2003,221:167-180.
[67] Schulenberg T., Starflinger J. Core Design Concepts for High PerformanceLight Water Reactors [J]. Nuclear Engineering and Technology,2007,39:249-256.
[68] Schulenberg T., Starflinger J., Heinecke J. Three Pass Core Design Proposal fora High Performance Light Water Reactor [J]. Progress in Nuclear Engineering,2008,50:526-531.
[69] Barre F., Bernard M. The CATHARE Code Strategy and Assessment [J].Nuclear Engineering and Design,1990,124:257-284
[70] Geffraye G., Antoni O., Farvacque M. et al. CATHARE2V2.52: A SingleVersion for Various Applications [J]. Nuclear Engineering and Design,2011,241:4456-4463.
[71] Dumaz P., Antoni O. The Extension of the CATHARE Computer Code Abovethe Critical Point, Applications to a Supercritical Light Water Reactor,Proceedings of the10th International Topical Meeting on Nuclear ReactorThermal Hydraulics (NURETH-10),2003[C]. Seoul, Korea,2003.
[72] Robert M., Farvacque M., Parent M. et al. CATHARE2V2.5: a Fully ValidatedCATHARE Version for Various Applications Proceedings of the10thInternational Topical Meeting on Nuclear Reactor Thermal Hydraulics(NURETH-10),2003[C]. Seoul, Korea,2003.
[73] Antoni O., Dumaz P. Preliminary Calculations of a Supercritical Light WaterReactor Concept Using the CATHARE Code, Proceedings of the2003International Congress on Advanced Nuclear Power Plants (ICAPP’03),2003
[C]. Cordoba, Spain,2003.
[74] Marsault Ph., Renault C., Rimpault G. et al. Pre-design Studies of SCWR inFast Neutron Spectrum: Evaluation of Operating Conditions and Analysis ofthe Behavior in Accidental Situations, Proceedings of the2004InternationalCongress on Advanced Nuclear Power Plants (ICAPP’04),2004[C]. Pittsburgh,U.S.,2004.
[75] Manera A., Antoni O. Code-to-Code Comparison for Blow-down Transients atSupercritical Conditions, Proceedings of the Annual Meeting on NuclearTechnology,2008[C]. Hamburg, Germany,2008.
[76] Alobaid F., Strohle J., Epple B. et al. Dynamic Simulation of a SupercriticalOnce-through Heat Recovery Steam Generator During Load Changes and StartUp Procedures [J]. Applied Energy,2009,86:1274-1282.
[77]蔡宝玲,王哲,魏湘等.基于引进仿真支撑软件APROS的电站仿真培训系统[J].热力发电,2004,11:7-9.
[78]蔡宝玲,王哲,魏湘等.超临界600MW机组仿真系统动态数学模型的开发及其分析[J].热力发电,2006,2:34-36.
[79]杨冬,马彦花,潘杰等.600MW超临界循环流化床锅炉水冷壁动态特性的研究[J].动力工程,2009,29(8):722-727.
[80]周东阳,魏湘,蔡宝玲等.超超临界1000MW机组仿真系统开发与应用[J].热力发电,2011,11:76-77.
[81] H nninen M., Kurki J. Simulation of Flows at Supercritical Pressures with aTwo-fluid Code, Proceedings of the7th International Topical Meeting onNuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-7),2008[C]. Seoul, Korea,2008.
[82] Kurki J. Simulation of Thermal Hydraulics at Supercritical Pressures withAPROS [D]. Helsinki University of Technology, Finland,2008.
[83] H nninen M. Phenomenological extensions to APROS six-equation modelNon-condensable gas, supercritical pressure, improved CCFL and reducednumerical diffusion for scalar transport calculation [D]. LappeenrantaUniversity of Technology, Finland,2009.
[84] Kurki J., Sepp l M. Thermal Hydraulic Transient Analysis of the HighPerformance Light Water Reactor Using APROS and SMABRE, Proceedingsof the20th International Conference on Structural Mechanics in ReactorTechnology,2009[C]. Espoo, Finland,2009.
[85] Kurki J., H nninen M. On the Applicability of a One-dimensional Two-fluidFlow Model to Simulation of Flows and Heat Transfer at Near and SupercriticalPressures, Proceedings of the8th International Topical Meeting on NuclearReactor Thermal Hydraulics, Operation and Safety (NUTHOS-8),2010[C].Shanghai, China,2010.
[86] Kurki J., H nninen M. Simulation of Large-break Loss of Coolant Accident inthe High Performance Light Water Reactor, Proceedings of the IAEA TechnicalMeeting on Heat Transfer, Thermal-Hydraulics and System Design forSupercritical Water Cooled Reactors,2010[C]. Pisa, Italy,2010.
[87] Andreani M., Bittermann D., Marsault Ph. et al. Evaluation of a PreliminarySafety Concept for the HPLWR [J]. Progress in Nuclear Energy,2012,55:68-77.
[88] Johnsen G., Davis C., Bayless P. ATHENA-3D for Generation IV ReactorAnalysis, BE-2004, Proceedings of the ANS Winter Meeting,2004[C].Washington D.C., U.S.,2004.
[89] Riemke R., Davis C., Schultz R. RELAP5-3D Code Includes Athena Featuresand Models, Proceedings of the14th International Conference on NuclearEngineering (ICONE-14),2006[C]. Miami, U.S.,2006.
[90] Sumner T., Ghiaasiaan S.M. Effects of fuel type on the safety characteristics ofa sodium-cooled fast reactor: Part I: Background, modeling tools andpre-transient calculations [J]. Annals of Nuclear Energy,2011,38:1559-1568.
[91] Sumner T., Ghiaasiaan S.M. Effects of fuel type on the safety characteristics ofa sodium cooled fast reactor. Part II: Simulation results [J]. Annals of NuclearEnergy,2011,38:1760-1768.
[92] Spore J.W. TRAC-M/FORTRAN90(V3.0) Theory Manual LAUR-00-910[R].Los Alamos National Laboratory/Penn State University, U.S.,2000.
[93] Mikityuk K., Pelloni S., Coddington P. et al. FAST: An advanced code systemfor fast reactor transient analysis [J]. Annals of Nuclear Energy,2005,32:1613-1631.
[94] Chenu A., Mikityuka K., Chawla R. TRACE simulation of sodium boiling inpin bundle experiments under loss-of-flow conditions [J]. Nuclear Engineeringand Design,2009,239:2417-2429.
[95] Chenu A., Adams R., Mikityuk K., et al. Analysis of selected Phenix EOL testswith the FAST code system, Part I: control-rod-shift experiments [J]. Annals ofNuclear Energy,2012,49:182-190.
[96] Chenua A., Mikityuka K., Chawlaa R. Analysis of selected Phenix EOL testswith the FAST code system, Part II: Unprotected phase of the NaturalConvection Test [J]. Annals of Nuclear Energy,2012,49:191-199.
[97] Tenchine D., Baviere R., Bazin P., et al. Status of CATHARE code for sodiumcooled fast reactors [J]. Nuclear Engineering and Design,2012,245:140-152.
[98] Mochizuki H. Verification of NETFLOW code using plant data of sodiumcooled reactor and facility [J]. Nuclear Engineering and Design,2007,237:87-93.
[99] Mochizuki H. Inter-subassembly heat transfer of sodium cooled fast reactors:validation of the NETFLOW code [J]. Nuclear Engineering and Design,2007,237:2040-2053.
[100] Mochizuki H. Development of the plant dynamics analysis code NETFLOW++[J]. Nuclear Engineering and Design,2010,240:577-587.
[101] Mochizuki H. Plant behavior of a fast breeder reactor under loss of AC powerfor long period [J]. Nuclear Engineering and Design,2012,245:19-27.
[102] Tobita Y., S Kondo., Yamano H. et al. The Development of SIMMER-III, anAdvanced Computer Program for LMFR Safety Analysis, Proceedings of theJoint IAEA/NEA Technical Meeting on Use of Computational Fluid Dynamics(CFD) Codes for Safety Analysis Reactors Systems Including Containment,2002[C]. Pisa, Italy,2002.
[103] Tobita Y., Kondo S., Yamano H., et al. The development of SIMMER-III, anadvanced computer program for lmfr safety analysis, and its application tosodium [J]. Nuclear Technology,2006,153:245-255.
[104] Yamano H., Fujita S., Tobita Y., et al. Development of a three-dimensionalCDA analysis code: SIMMER-IV and its first application to reactor case [J].Nuclear Engineering and Design,2008,238:66-73.
[105] Maschek W., Rineiski A., Suzuki T., et al. SIMMER-III and SIMMER-IVSafety Code Development for Reactors with Transmutation Capability,Proceedings of the Mathematics and Computation, Supercomputing, ReactorPhysics and Biological Applications,2005[C]. Palais de Papes, France,2005.
[106] Suzuki T., Chen X.N., Rineiski A., et al. Transient analyses for acceleratordriven system PDS-XADS using the extended SIMMER-III code [J]. NuclearEngineering and Design,2005,235:2594-2611.
[107] Dunn F.E., Prohammer F.G. The SASSYS LMFBR systems analysis code [J].Mathematics and Computers in Simulation,1984,26:23-26.
[108] Ha K.S., Jeong H.Y., Cho C., et al. Simulation of the EBR-II Loss-of-FlowTests Using the MARS Code [J]. Nuclear Technology,2010,169:134-142.
[109] Jeong H.Y., Ha K.S., Lee K.L., et al. Pre-Test Analysis of Natural CirculationTest of PHENIX End-of-Life With the MARS-LMR Code, Proceedings of the18th International Conference on Nuclear Engineering (ICONE-18),2010[C].Xi'an, China,2010.
[110] Woo S.M., Chang S.H. Multi dimensional analysis of Design Basis Eventsusing MARS-LMR [J]. Nuclear Engineering and Design,2012,244:83-91.
[111] Guppy J.G. Super System Code (SSC, Rev.0) An Advanced ThermohydraulicSimulation Code for Transients in LMFBRs, NUREG/CR-3169(BNL-NUREG-51650)[R]. U.S.,1983.
[112]杨红义,徐銤. OASIS程序的开发与应用[J].核科学与工程,2001,21(4):322-340.
[113]许义军.中国实验快堆钠池三维热工水力分析[D].中国原子能科学研究院,2003.
[114]张春明. DINROS程序在中国实验快堆事故分析中的应用[J].核科学与工程,2006,26(2):142-148.
[115]任丽霞.钠冷快堆系统分析程序实用开发[D].中国原子能科学研究院,2007.
[116]李硕.钠冷堆瞬态安全分析方法与初步研究[D].华北电力大学（北京）,2011.
[117]陆道纲,隋丹婷,任丽霞等.池式快堆系统分析软件稳态功能开发[J].原子能科学技术,2012,46(4):422-428.
[118]隋丹婷,陆道纲,张盼.用于池式快堆系统分析的钠池三维模型开发[J].原子能科学技术,2012,46(4):429-436.
[119]陆道纲,隋丹婷.池式快堆系统瞬态分析软件开发[J].原子能科学技术,2012,46(5):542-548.
[120] Lerchel G., Austregesilo H., ATHLET Mod2.1Cycle A, User's Manual,GRS-p-1/Vol.1, Rev.4[R]. GRS,2006.
[121] Ulrych F. Darstellung der IFC-Wasser-Dampf-Zustandsgleichungen alsALGOL-Prozeduren bzw. FORTRAN-Funktionen (D2O und H2O), TS116Nr.1/68[R]. Deutschland, Siemens,1968.
[122] The IFC-Formulation for Industrial Use. A Formulation of the ThermodynamicProperties of Ordinary Water Substance [M]. Germany: VDI-Verlag,1967.
[123] Müller W.Ch. Fast and accurate water and steam properties programs fortwo-phase flow calculations [J]. Nuclear Engineering and Design,1994,149:449-458.
[124] Ball A., Trambauer K. Stoffwertapproximation der Transportgr en vonWasser für Temperaturen bis3000°C, GRS-A-1710[R]. Deutschland, GRS,1990.
[125] Goldammer H.D. H2OSTA-A General Purpose Software Package for theCalculation of Thermodynamic Properties of Water/Steam, Final Report forContract Nr.3048-86-10[R] Germany, EP/ISP D,1987.
[126] MATPRO-Version11(Rev.2), A Handbook of Material Properties for Use inthe Analysis of Light Water Reactor Fuel Behaviour, NUREG/CR-0497,TREE-1280, Rev.2[R]. Idaho National Engineering Laboratory,1981.
[127] Revised release on the IAPWS industrial formulation1997for thethermodynamic properties of water and steam [R]. International Association forthe Properties of Water and Steam, Switzerland,2007.
[128] Supplementary Release on Backward Equations for Specific Volume as aFunction of Pressure and Temperature v(p,T) for Region3of the IAPWSIndustrial Formulation1997for the Thermodynamic Properties of Water andSteam [R]. The International Association for the Properties of Water and Steam,Greece,2005.
[129] Fu S.W., Zhou C., Xu Z.H., et al. Modification and Application of theATHLET-SC Code to Trans-critical Simulations, Proceeding of the5thInternational Symposium on Supercritical Water-Cooled Reactors,2011[C].Canada,2011.
[130] Cheng X., Schulenberg T. Heat transfer at supercritical pressure--literaturereview and application to an HPLWR, FZKA6609[R]. Karlsruhe Institute ofTechnology, Germany,2001.
[131] Krasnoshchekov E.A., Protopopov V.S. Experimental study of heat exchange incarbon dioxide in the supercritical range at high temperature drops [J].Teplofizika Vysokikh Temperature,1966,4(3):389-398.
[132] Yamagata K., Nishikawa K., Hasegawa S., et al. Forced convective heattransfer to supercritical water flowing in tubes [J]. International Jounal of HeatMass Transfer,1972,15(12):2575-2593.
[133] Jackson J.D. Validation of an extended heat transfer equation for fluids atsupercritical pressure, Proceedings of the4th International Conference onSupercritical Water-Cooled Reactors,2009[C]. Heidelberg, Germany,2009.
[134] Cheng X., Yang Y.H., Huang S.F. A simplified method for heat transferprediction of supercritical fluids in circular tubes [J]. Annals of Nuclear Energy,2009,36:1120-1128.
[135] Bishop A.A., Sandberg R.O., Tong L.S. High temperature supercritical pressurewater Loop-IV-Forced convection heat transfer to water at near-criticaltemperatures and near supercritical pressures, WCAP-2056, Part IV [R]. U.S.,1964.
[136] Herkanrath H., Mork-Morkenstein P., Jung U., et al. Heat transfer in water withforced circulation in the pressure range from140-250bar, E.U.R.3658d, HTFS916[R]. Joint Nuclear Research Centre, Italy,1967.
[137] Griem H. A new procedure for the prediction of forced convection heat transferat near-and supercritical pressure [J]. International Jounal of Heat and MassTransfer,1996,31:301-305.
[138] Fewster J., Jackson J.D. Experiments on supercritical pressure convective heattransfer having relevance to SPWR, Proceedings of the2004InternationalCongress on Advanced Nuclear Power Plants (ICAPP’04),2004[C]. Pittsburgh,U.S.,2004.
[139] Xu F. Studies on heat transfer characteristics in tubes at supercritical pressures(in Chinese)[D]. Xi’an Jiao Tong University, China,2004.
[140] Cheng X., Gu H.Y. Test data of School of Nuclear Science and Engineering [R].Shanghai Jiao Tong University,2011, Personal communication.
[141] Domin G. W rmeübergang in kritischen und überkritischen Bereichen vonWasser in Rohren BWK15[R].1963.
[142] Chen Y.Z., Yang C.S., Zhao M.F., et al. Experimental Studies on Critical Flowand Heat Transfer of Water for Near-critical and Supercritical Pressures,Proceedings of the IAEA Technical Meeting on Heat Transfer,Thermal-Hydraulics and System Design for Supercritical Water-CooledReactors,2010[C]. Pisa, Italy,2010.
[143] Bystrov P.I., Kagan D.N., Krechetova G.A., et al. Liquid-Metal Collants forHeat Pipes and Power Plants [M]. US: Ed. V. A. Kirillin, Hemisphere Pub.Corp,1990.
[144] Fink J.K., Leibowitz L. Thermodynamic and Transport Properties of SodiumLiquid and vapor, ANL/RE-95/2[R]. Argonne National Laboratory,1995.
[145] Borishanski V.M., Firsova E.V. Heat Transfer in Liquid Metals, FluidMechanics and Heat Transfer of Heat Exchanger Design Handbook, vol.2,chapter2.5.13[M]. US: Hemisphere Publishing Corporation,1990.
[146] Mikityuk K. Heat transfer to liquid metal: Review of data and correlations fortube bundles [J]. Nuclear Engineering and Design,2009,239:680-687.
[147] Cheng X., Tak N. Investigation on turbulent heat transfer to lead-bismutheutectic flows in circular tubes for nuclear applications [J]. NuclearEngineering and Design,2006,236:385-393.
[148] Pfrang W., Struwe D. Assessment of Correlations for Heat Transfer to theCoolant for Heavy Liquid Metal Cooled Core Designs, FZKA7352[R].Karlsruhe Institute of Technology, Germany,2007.
[149] Lyon R.N. Liquid metal heat transfer coefficients [J]. Chemical EngineeringProgress,1951,47:75-79.
[150] Skupinski E., Tortel J., Vautrey L. Tetermination des coefficients de convectionD’un Alliage sodium-potassium Dans un Tube Circulaire [J]. InternationalJounal of Heat Mass Transfer,1965,8:937-951.
[151] Sleicher C.A., Awad A.S., Notter R.H. Temperature and Eddy diffusivityprofiles in NaK [J]. International Jounal of Heat Mass Transfer,1973,16:1565-1575.
[152] Gr ber V.H., Rieger M. Experimentelle Untersuchung des W rmeübergangs anFlüssigmetalle (NaK) in parallel durchstr mten Rohrbündeln bei konstanter undexponentieller W rmeflussdichteverteilung [J]. Atomkernenergie (ATKE) Bd.,1972,19:23-40.
[153] Ushakov P.A., Zhukov A.V., Matyukhin M.M. Heat transfer to liquid metals inregular arrays of fuel elements [J]. High Temperature,1977,15:868-873.
[154] Chenu A., Mikityuka K., Chawla R. Pressure drop modeling and comparisonswith experiments for single-and two-phase sodium flow [J]. NuclearEngineering and Design,2011,241:3898-3909.
[155] Bubelis E., Schikorr M. Review and proposal for best fit of wire-wrapped fuelbundle friction factor and pressure drop predictions using various existingcorrelations [J]. Nuclear Engineering and Design,2008,238:3299-3320.
[156] Edwards A.R., O’Brien T.P. Studies of phenomena connected with thedepressurization of water reactors [J]. Jounal of British Nuclear Energy Society,1970,9:125-135.
[157] Manera A., Antoni O. Code-to-code comparison for blowdown transients atsupercritical conditions, Proceedings of the Annual meeting on nucleartechnology2011[C]. Berlin, Germany,2011.
[158] Cheng X., Yang Y.H. A point-hydraulics model for flow stability analysis [J].Nuclear Engineering and Design,2008,238:188-199.
[159] Cheng X., Kuang B., Yang Y.H. Investigation on Flow Stability ofSupercritical Water Cooled Systems, Proceedings of the2006Internationalcongress on advances in nuclear power plants (ICAPP’06),2006[C]. Reno,U.S.,2006.
[160] Tenchine D. Some thermal hydraulic challenges in sodium cooled fast reactors[J]. Nuclear Engineering and Design,2010,240:1195-1217.
[161] GauthéP. THINS (WP2and WP5): Natural convection test in PHENIX, datalist, test description and initial state, CEA/DEN/CAD/DER/SESI/LSMR/NT,DO6du07/09/10[R]. CEA, France,2010.
[162] Pialla D., Tenchine D., Gauthé P., et al. Natural convection test in Phenixreactor and associated CATHARE calculation, Proceedings of the14thInternational Topical Meeting on Nuclear Reactor Thermal hydraulics [C].Toronto, Canada,2011.
[163] RaquéM., Vojecek A., Hajek P., et al. Design and1D analysis of the safetysystems for the SCWR fuel qualification test, Proceedings of the9thInternational Topical Meeting on Nuclear Reactor Thermal Hydraulics,Operation and Safety (NUTHOS-9),2012[C]. Kaohsiung, Taiwan,2012.
[164] Coffman F. LOCA Temperature criterion for stainless steal clad fuel [R].NUREG-0065,1976.