超临界水堆核热耦合及系统安全特性研究
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摘要
超临界水堆具有堆芯进出口温差大、冷却剂流量低和燃料棒间隙窄等特征,由此带来强烈的物理-热工反馈,以及比普通压水堆更高的冷却剂流量供应要求,从而影响其系统安全特性。在本文中,选用日本Super LWR为研究对象,开发了超临界水堆核热耦合分析程序与安全特性分析程序,展开了耦合特性分析及安全特性分析。在耦合特性分析中率先提出了超临界窄缝效应的概念,并且将核热耦合引入超临界水堆安全分析。
首先,进行了超临界水堆稳态核热耦合特性分析,并结合超临界窄间隙和跨临界的设计特征进行窄缝效应研究：(1)对比了5%富集度的耦合计算结果与余弦曲线拟合轴向功率分布的非耦合计算结果,发现物理热工耦合导致内、外组件堆芯功率峰值因子沿轴向发生明显偏移,并且使得最高包壳温度降低。(2)进行了耦合条件下不同燃料棒间隙下的流动换热特性分析,发现燃料棒间隙减小,冷却剂换热系数明显增加,而最高包壳温度明显降低,但是变化幅度均较非耦合计算结果小。研究了不同流量时燃料棒间隙最大允许值,为设计优化提供理论参考。
其次,进行了超临界水堆瞬态耦合特性分析及滑压启动特性研究：(1)通过冷却剂温度升高瞬态及慢化剂温度升高瞬态的特性分析,发现物理与热工之间的耦合作用将导致堆芯功率随时间明显降低,其中冷却剂通道入口温度升高引起功率降低幅度最明显。(2)通过堆芯功率升高瞬态的特性分析,发现冷却剂温度随功率升高呈现升高趋势,而物理热工的反馈作用机制抑制了最高包壳温度的升高幅度。(3)以滑压启动的功率上升过程为例,进行平均流量条件下不同组件的启动特性分析,提出堆芯组件冷却剂流量分配方案和滑压启动曲线优化方案。
最后,进行了超临界水堆安全特性分析：(1)以主给水流量降低、温度降低和压力升高三种扰动为例,对比分析了不同扰动时控制参数及安全特性变化,发现主给水温度降低与流量降低导致更加显著的特性变化；(2)以5%流量丧失事件、单台冷却剂泵故障事件及丧失厂外电源事件为例,进行单通道安全特性分析及其敏感性分析,发现基于时空动力学耦合求解的最高包壳温度始终低于点堆方程求解的最高包壳温度；(3)以给水加热丧失事件和单台冷却剂泵故障事件为例,进行多通道安全特性分析。结果表明具有最大功率因子燃料组件的最高包壳温度峰值远远高于其它燃料组件,但是仍满足安全准则要求。
For supercritical light water-cooled reactor, it is characterized as lower flow rate design, narrower fuel rod gaps and larger temperature increase from inlet to outlet of the core than that of traditional light water reactors, which bring new problems for its safety analysis especially with the influence of neutron-thermal coupling and coolant supply. According to the coupling and safety codes, the research on coupling between thermal-hydraulic and neutronic is carried out, and the safety analysis is put forward focusing on the transients or events of supercritical water-cooled reactor associated with coolant flow rate changing. For the coupling characteristics analysis, supercritical narrow effect was firstly brought forward. And for the first time in China, the coupling method was introduced into safety characteristics analysis of supercritical water-cooled reactor.
Firstly, the research on steady coupling characteristics analysis as well as supercritical narrow effects was given.(ⅰ) By comparing the coupling results with that results given by uncoupling method with cosine approximated, it shows that neutronic and thermal-hydraulic coupling will lead to the axial power peak factor shifting along axial direction for both the internal and external components. It makes part of the cladding temperature rising but the maximum cladding temperature decreasing.(ⅱ) For given design flow rate of coolant, thermal hydraulic analysis of supercritical pressure light water reactor in different gap sizes is given by changing the fuel rod pitch only, and the characteristics of narrow channels effect with enhancing heat transfer are proposed. As the gap size decreases, maximum cladding temperature decreases while the convective heat transfer coefficient between fuel rod and coolant will be risen. Based on thermal analysis in different coolant flow rate, the reasonable value range of gaps between fuel rods is given, in which the maximum cladding temperature safety limits and installation technology are comprehensively considered.
Secondly, studies on the transient coupling characteristics analysis as well as the thermal characteristics during start up are put forward.(ⅰ) Transient physical characteristics analysis is given by changing coolant temperature as well as moderator temperature. It shows that, as the coolant temperature increases, the core power decreases with the coupling influence. While the moderator temperature increases, it gives the same results but with smaller decreased degree.(ⅱ) Transient thermal characteristics analysis is given by changing core power. It shows that the coolant temperature increases while maximum cladding temperature decreases but with the inhibition by coupling influence.(iii) According to the given slidding pressure mode of SCWR, the transient characteristics for power increasing is detailed analyzed. Considering with radial heterogeneity of power distribution, thermal characteristics for different assemblies are also put forward. Then on the principle of startup safety and stability, optimization design of slipping startup for supercritical pressure light water reactor is given by adjusting flow rate distribution in different fuel assemblies or changing power setting during startup.
Finally, safety characteristics analysis is given.(i) The comparative analysis of transient safety characteristics with main parameters disturbance is made by taking the feed-water flow rate decreasing, temperature decreasing and pressure increasing for example. It shows that the feed-water flow rate decreasing and temperature decreasing gives serious variation of system characteristics (ii) Considering the importance of coolant supply for supercritical water-cooled reactor, transient analysis of single coolant pump failure rate is given by coupled neutronics and thermal hydraulics calculation method, as well as that of5%partial loss of coolant flow rate. On this basis, sensitivity analysis is performed. The results shows that maximum cladding temperature calculated by coupling method is always lower than that calculated by uncoupling method (iii) By the example of Loss of Feed-water Heating Transient and Partial Loss of Reactor Coolant Flow Transient, multi-channel safety analysis is given by introducing the radial power distribution factors and flow distribution calculation. In which, the axial power distribution factors is obtained by curve fitting based on the coupled calculation results. During the above two transient process, there is a peak value of maximum cladding temperature exists in each fuel assembly, in which the maximum cladding temperature peak of fuel assembly with the highest power factor is much higher than that of other fuel assembly. But this situation still meets the safety standards.
首先,进行了超临界水堆稳态核热耦合特性分析,并结合超临界窄间隙和跨临界的设计特征进行窄缝效应研究：(1)对比了5%富集度的耦合计算结果与余弦曲线拟合轴向功率分布的非耦合计算结果,发现物理热工耦合导致内、外组件堆芯功率峰值因子沿轴向发生明显偏移,并且使得最高包壳温度降低。(2)进行了耦合条件下不同燃料棒间隙下的流动换热特性分析,发现燃料棒间隙减小,冷却剂换热系数明显增加,而最高包壳温度明显降低,但是变化幅度均较非耦合计算结果小。研究了不同流量时燃料棒间隙最大允许值,为设计优化提供理论参考。
其次,进行了超临界水堆瞬态耦合特性分析及滑压启动特性研究：(1)通过冷却剂温度升高瞬态及慢化剂温度升高瞬态的特性分析,发现物理与热工之间的耦合作用将导致堆芯功率随时间明显降低,其中冷却剂通道入口温度升高引起功率降低幅度最明显。(2)通过堆芯功率升高瞬态的特性分析,发现冷却剂温度随功率升高呈现升高趋势,而物理热工的反馈作用机制抑制了最高包壳温度的升高幅度。(3)以滑压启动的功率上升过程为例,进行平均流量条件下不同组件的启动特性分析,提出堆芯组件冷却剂流量分配方案和滑压启动曲线优化方案。
最后,进行了超临界水堆安全特性分析：(1)以主给水流量降低、温度降低和压力升高三种扰动为例,对比分析了不同扰动时控制参数及安全特性变化,发现主给水温度降低与流量降低导致更加显著的特性变化；(2)以5%流量丧失事件、单台冷却剂泵故障事件及丧失厂外电源事件为例,进行单通道安全特性分析及其敏感性分析,发现基于时空动力学耦合求解的最高包壳温度始终低于点堆方程求解的最高包壳温度；(3)以给水加热丧失事件和单台冷却剂泵故障事件为例,进行多通道安全特性分析。结果表明具有最大功率因子燃料组件的最高包壳温度峰值远远高于其它燃料组件,但是仍满足安全准则要求。
For supercritical light water-cooled reactor, it is characterized as lower flow rate design, narrower fuel rod gaps and larger temperature increase from inlet to outlet of the core than that of traditional light water reactors, which bring new problems for its safety analysis especially with the influence of neutron-thermal coupling and coolant supply. According to the coupling and safety codes, the research on coupling between thermal-hydraulic and neutronic is carried out, and the safety analysis is put forward focusing on the transients or events of supercritical water-cooled reactor associated with coolant flow rate changing. For the coupling characteristics analysis, supercritical narrow effect was firstly brought forward. And for the first time in China, the coupling method was introduced into safety characteristics analysis of supercritical water-cooled reactor.
Firstly, the research on steady coupling characteristics analysis as well as supercritical narrow effects was given.(ⅰ) By comparing the coupling results with that results given by uncoupling method with cosine approximated, it shows that neutronic and thermal-hydraulic coupling will lead to the axial power peak factor shifting along axial direction for both the internal and external components. It makes part of the cladding temperature rising but the maximum cladding temperature decreasing.(ⅱ) For given design flow rate of coolant, thermal hydraulic analysis of supercritical pressure light water reactor in different gap sizes is given by changing the fuel rod pitch only, and the characteristics of narrow channels effect with enhancing heat transfer are proposed. As the gap size decreases, maximum cladding temperature decreases while the convective heat transfer coefficient between fuel rod and coolant will be risen. Based on thermal analysis in different coolant flow rate, the reasonable value range of gaps between fuel rods is given, in which the maximum cladding temperature safety limits and installation technology are comprehensively considered.
Secondly, studies on the transient coupling characteristics analysis as well as the thermal characteristics during start up are put forward.(ⅰ) Transient physical characteristics analysis is given by changing coolant temperature as well as moderator temperature. It shows that, as the coolant temperature increases, the core power decreases with the coupling influence. While the moderator temperature increases, it gives the same results but with smaller decreased degree.(ⅱ) Transient thermal characteristics analysis is given by changing core power. It shows that the coolant temperature increases while maximum cladding temperature decreases but with the inhibition by coupling influence.(iii) According to the given slidding pressure mode of SCWR, the transient characteristics for power increasing is detailed analyzed. Considering with radial heterogeneity of power distribution, thermal characteristics for different assemblies are also put forward. Then on the principle of startup safety and stability, optimization design of slipping startup for supercritical pressure light water reactor is given by adjusting flow rate distribution in different fuel assemblies or changing power setting during startup.
Finally, safety characteristics analysis is given.(i) The comparative analysis of transient safety characteristics with main parameters disturbance is made by taking the feed-water flow rate decreasing, temperature decreasing and pressure increasing for example. It shows that the feed-water flow rate decreasing and temperature decreasing gives serious variation of system characteristics (ii) Considering the importance of coolant supply for supercritical water-cooled reactor, transient analysis of single coolant pump failure rate is given by coupled neutronics and thermal hydraulics calculation method, as well as that of5%partial loss of coolant flow rate. On this basis, sensitivity analysis is performed. The results shows that maximum cladding temperature calculated by coupling method is always lower than that calculated by uncoupling method (iii) By the example of Loss of Feed-water Heating Transient and Partial Loss of Reactor Coolant Flow Transient, multi-channel safety analysis is given by introducing the radial power distribution factors and flow distribution calculation. In which, the axial power distribution factors is obtained by curve fitting based on the coupled calculation results. During the above two transient process, there is a peak value of maximum cladding temperature exists in each fuel assembly, in which the maximum cladding temperature peak of fuel assembly with the highest power factor is much higher than that of other fuel assembly. But this situation still meets the safety standards.
引文
[1]程旭,刘晓晶.超临界水冷堆国内外研究现状与趋势[J].原子能科学技术,2008,42(2)：167-172
[2]陆道纲,彭常宏.超临界水冷堆评述[J].原子能科学技术,2009,43(8)：743-749
[3]Oka Y., Koshizuka S., Ishiwatari Y., et al. Super light water reactors and super fast reactors[M]. New York:Springer,2010
[4]Yamaji A., Kamei K., Oka Y, et al. Improved core design of the high temperature supercritical-pressure light water reactor[J]. Annals of Nuclear Energy,2005(32): 651-670
[5]黄禹,沈飚,张鹏.超临界流体传热综述[J].制冷技术,2008,36(10)：44-50
[6]Cheng X. Supercritical water reactor-fuel qualification test (SCWR-FQT). http://www.ifrt.kit.edu/english/101_290.php
[7]刘华,周世荣,巢哲雄.我国应对福岛核事故的措施及启示[J].中国核工业,2011,10：18-19
[8]龚曙光.ANSYS工程应用实例解析[M].北京：机械工业出版社,2003
[9]程旭.超临界水冷堆是我国水冷堆技术路线的自然发展[J].核动态,2007,4：26-28
[10]Department of Energy. A Technology Roadmap for Generation Ⅳ Nuclear Energy Systems-Technical Roadmap Report[R]. GIF-002-00,2002
[11]Dobashi K., Kimura A., Oka Y., et al. Conceptual design of a high temperature power reactor cooled and moderated by Supercritical Light Water[J]. Annals of Nuclear Energy,1998,25(8):487-505
[12]Yamada K., Ookawa M., Asanuma Y, et al. Recent activities and future plan of thermal-spectrum SCWR development in Japan[C]. The 3rd Int. Symposium on SCWR-Design and Technology,2007, Shanghai, China
[13]Mukohara T., Koshizuka S., Oka Y. Core design of a high-temperature fast reactor cooled by supercritical light water[J]. Annals of Nuclear Energy,1999,26: 1423-1436
[14]Oka Y, Koshizuka S.. Supercritical-pressure, once-through cycle light water cooled reactor [J]. Journal of Nuclear Science and Technology,2001,38(12): 1081-1089
[15]Yoo J., Ishiwatari Y, Oka Y, et al. Conceptual design of compact supercritical water-cooled fast reactor with thermal hydraulic coupling [J]. Annals of Nuclear Energy,2006,33:945-956
[16]Kamei K., Yamaji A., Ishiwatari Y., et al. Fuel and core design of super light water reactor with low leakage fuel loading pattern[J]. Journal of Nuclear Science and Technology,2006,43(2):129-139
[17]Yoo J., Oka Y., Ishiwatari Y., et al. Thermo-mechanical analysis of supercritical pressure light water-cooled fast reactor fuel rod by FEMAXI-6 code[J]. Annals of Nuclear Energy,2006,33:1379-1390
[18]Hofmeister J., Waata C., Starflinger J., et al. Fuel assembly design study for a reactor with supercritical water[J]. Nuclear Engineering and Design,2007,237: 1513-1521
[19]Fischer K., Schulengberg T., Laurien E.. Design of a supercritical water-cooled reactor with a three-pass core arrangement[J]. Nuclear Engineering and Design, 2009,239:800-812
[20]Schulengberg T., Starflinger J., Marsault P., et al. European supercritical water cooled reactor[J]. Nuclear Engineering and Design,2011,241:3505-3513
[21]Torgerson D.F., Shalaby B.A., Pang S.. CANDU technology for generation III+ and IV reactors[J]. Nuclear Engineering and Design,2006,236:1565-1572
[22]Yang P., Cao L.Z., Wu H.C., et al. Core design study on CANDU-SCWR with 3D neutronics/thermal-hydraulics coupling[J]. Nuclear Engineering and Design,2011, 241(12):4714-4719
[23]Buongiorno J., MacDonald P.E.. Supercritical water reactor (SCWR) progress report for the FY-03 generation-Ⅳ R&D activities for the development of the SCWR in the U.S.[R]. INEEL/EXT-03-01210,2003
[24]程旭,刘晓晶.混合能谱超临界水堆堆芯设计分析[J].核科学与工程,2009,29(1)：43-49
[25]Yang T., Liu X.J., Cheng X.. Optimization of multilayer fuel assemblies for supercritical water-cooled reactors with mixed neutron spectrum[J]. Nuclear engineering and design,2012,249:159-165.
[26]超临界水堆特定条件下堆芯物理特性及其分析方法的基础研究报告[R].2007BC209807,2011
[27]孙灿辉.超临界水堆MOX燃料物理热工特性研究[D].北京：华北电力大学,2012
[28]严家騄,余晓福,王永青.水和水蒸气热力性质图表[M].北京：高等教育出版社,2008
[29]马庆芳,方荣生,项立成等.实用热物理性质手册[M].北京：中国农业机械出版社,1986
[30]Yang X.B., Su G.H., Tian W.X., et al. Numerical study on flow and heat transfer characteristics in the rod bundle channels under super critical pressure condition[J]. Annals of nuclear energy,2010,37:1723-1734.
[31]Cheng X., Liu X.J., Gu H.Y.. Fluid-to-fluid scaling of heat transfer in circular tubes cooled with supercritical fluids[J]. Nuclear Engineering and Design,2011, 241:498-508
[32]张曙明,陈玉宙,赵民富.竖直圆管内跨临界压力区水对流传热数值模拟研究[J].原子能科学技术,2009,43(6)：491-495.
[33]张亚奇.超临界压力下竖直上升管传热分析与回归评价[D].上海：上海交通大学,2008.
[34]Shang Z.. CFD investigation of vertical rod bundles of supercritical water-cooled nuclear reactor[J]. Nuclear Engineering and Design,2009, 239:2562-2572.
[35]文彦,高超,秋穗正,等.矩形窄缝通道内水稳态和瞬态流动换热特性实验[J].核动力工程,2010,31(1)：28-32
[36]刘平,周涛,张明,等.自然循环条件下窄通道ONB点影响因素灰色关联度研究[J].核动力工程,2011,32(4)：29-32
[37]张明,周涛,盛程,等.窄通道潜热沸腾起始点计算模型的分析[J].核动力工程,2011,32(3)：73-76
[38]Yamaji A., Oka Y., Koshizuka S.. Three-dimensional core design of high temperature supercritical-pressure light water reactor with neutronic and thermal-hydraulic coupling[J]. Journal of Nuclear Science and Technology,2005, 42(1):8-19
[39]Reiss T., Feher S., Czifrus S.Z.. Coupled neutronics and thermohydraulics calculations with burn-up for HPLWRs[J]. Progress in Nuclear Energy,2008,50: 52-61
[40]Monti L., Starflinger J., Schulenberg T.. Development of a coupled neutronic/ thermal-hydraulic tool with multi-scale capabilities and applications to HPLWR core analysis[J]. Nuclear Engineering and Design,2011,241:1579-1591
[41]胡珀,杨燕华.超临界水堆系统分析程序的改进[J].原子能科学技术,2009,43(6)：548-551
[42]Shan J.Q., Zhang B., Li C.Y., et al. SCWR subchannel code ATHAS development and CANDU-SCWR analysis[J]. Nuclear Engineering and Design,2009,239: 1979-1987
[43]刘晓晶,程旭.混合能谱超临界水堆堆芯热工-物理性能分析[J].原子能科学技术,2009,43(6)：538-542
[44]张鹏,王侃,李满仓.均匀化与多群蒙卡输运研究[C].核反应堆系统设计技术重点实验室2010年会,2010
[45]安萍,姚栋.超临界水堆反应堆物理-热工水力耦合程序系统MCATHAS的开发[J].核动力工程,2010,31(6)：52-55
[46]刘占权,蒋朱敏,将校丰,等.超临界水堆堆芯轴向一维物理热工耦合稳态分析[J].核科学与工程,2009,29(1)：16-21
[47]刘晓壮,周涛,李臻洋,等.超临界水堆燃料组件核热耦合计算分析[C].核反应堆系统设计技术国家级重点实验室2009年报,2009
[48]李臻洋.超临界水堆物理热工程序研究[D].华北电力大学,2011
[49]Sun C.H., Zhou T., Hou Z.S., et al. Improving on assembly of SCWR using MOX fuel[J]. Advanced materials research,2012,347-353:1633-1636
[50]Ishiwatari Y., Oka Y., Koshizuka S., et al. Safety of Super LWR(I):Safety System Design[J]. Journal of Nuclear Science and Technology,2005,42(11):927-934
[51]Ishiwatari Y, Oka Y, Koshizuka S., et al. Safety of Super LWR(II):Safety Analysis at Supercritical Pressure[J]. Journal of Nuclear Science and Technology, 2005,42(11):935-948
[52]Lee J.H., Koshizuka S., Oka Y. Development of a LOCA analysis code for the supercritical-pressure light water cooled reactors[J]. Annual of Nuclear Energy, 1998,25(16):1341-1361
[53]Ishiwatari Y, Oka Y., Koshizuka S., et al. LOCA analysis of Super LWR[J]. Journal of Nuclear Science and Technology,2006,43(3):231-241
[54]Ishiwatari Y., Oka Y, Koshizuka S., et al. ATWS characteristics of Super LWR with/without alternative action[J]. Journal of Nuclear Science and Technology, 2007,44(4):572-589
[55]Ikejiri S., Ishiwatari Y., Oka Y. Safety analysis of a supercritical-pressure water-cooled fast reactor under supercritical pressure[J]. Nuclear Engineering and Design,2010,240:1218-1228
[56]Maraczy Cs., Kereszturi A., Trosztel I., et al. Safety analysis of reactivity initiated accidents in a HPLWR reactor by the coupled ATHLET-KIKO3D code[J]. Progress in Nuclear Energy,2010,52:190-196
[57]Xu Z.H., Hou D., Fu S.W., et al. Loss of flow accident and its mitigation measures for nuclear system with SCWR-M[J]. Annals of Nuclear Energy,2011, 38:2634-2644
[58]周翀,刘晓晶,杨燕华,等.ATHLET-SC程序的开发及适用性分析[J].原子能科学技术,2009,43(6)：556-560
[59]Zhu D.H., Zhao H., Tian W.X., et al. Development of TACOS code for loss of flow accident analysis of SCWR with mixed spectrum core[J]. Progress in Nuclear Energy,2012,54:150-161
[60]Zhu D.H., Tian W.X., Zhao H., et al. Comparatice study of transient thermal-hydraulic characteristics of SCWRs with different core design[J]. Annals of Nuclear Energy,2013,51:135-145
[61]胡雨.控制系统对超临界水堆事故影响分析[D].北京：华北电力大学,2010
[62]周涛,李臻洋,王晗丁,等.超临界水堆核热耦合及其瞬态相关控制方式研究[C].核电站新技术交流研讨会论文集,2010
[63]王晗丁,周涛,陈娟,等.超临界快堆给水控制失效瞬态控制分析[J].核动力工程,2011,32(5)：18-22
[64]王晗丁,周涛,陈娟,等.超临界水堆控制条件下扰动特性分析[J].原子能科学技术,2012,46(5)：555-560
[65]李磊,张志俭.并联通道瞬态流量分配方法研究[J].核动力工程,2010,31(5)：97-101
[66]蔡章生,桂学文,于雷.反应堆时空动力学方程的解法研究[J].海军工程大学学报,2006,18(3)：28-29
[67]宋英明,马远乐,单文志,等.高温气冷堆堆芯中子时空动力学模拟计算[J].计算物理,2009,26(6)：911-916
[68]廖承奎,求解中子扩散方程的半解析节块方法[J].原子能科学技术,2009,43(6)：496-500
[69]夏榜样,谢仲生.通量展开节块法求解六角形几何三维多群中子扩散方程[J].西安交通大学学报,2006,40(1)：84-87
[70]赵文博.瞬态节块格林函数方法及其与热工-水力耦合研究[D].清华大学,2012
[71]于平安,朱瑞安,喻真烷,等.核反应堆热工分析[M].上海：上海交通大学出版社,2002
[72]赵冬建,路璐,史国宝.超临界水堆子通道分析[J].原子能科学技术,2009,43(6)：543-547
[73]Gou J.L., Shang Z., Yuki I. et al. CFD analysis of heat transfer in subchannel of a super fast reactor[J]. nuclear engineering and design,2010,240:1819-1829
[74]Ivanov K., Avramova M.. Challenges in coupled thermal-hydraulics and neutronics simulations for LWR safety analysis[J]. Annals of Nuclear Energy, 2007,34:501-513
[75]孔衍,于雷,鄢炳火.自然循环下堆芯热工水力与时空中子动力学的耦合计算[J].船海工程,2010,39(5)：201-204
[76]宋英明,周志伟,马远乐.高温气冷堆温度反馈的集总参数法模拟[J].清华大学学报,2009,49(9)：92-95
[77]Volkanovski A., Cepin M.. Implication of PSA uncertainties on risk-informed decision making[J]. Nuclear Engineering and Design,2011,241(4):1108-1113
[78]Volkanovshi A., Cepin M., Mavko B.. Application of the fault tree analysis for assessment of power system reliability[J]. Reliability engineering and system safety,2009,94:1116-1127
[79]朱继洲.核反应堆安全分析[M].北京：原子能出版社,西安交大出版社,2002
[80]Durga K., Gopika V., Sanyasi V.V.S, et al. Dynamic fault tree analysis using Monte Carlo Simulation in probabilistic safety assessment[J]. Reliability engineering and system safety,2009,94:872-883
[81]Sajith T., John A., Parthasarathy U., etc. Passive system reliability analysis using Response Conditioning Method with an application to failure frequency estimation of Decay Heat Removal of PFBR[J]. Nuclear Engineering and Design, 2011,241(6):2257-2270
[82]Nayak A.K., Jain V., Gartia M.R., et al. Reliability assessment of passive isolation condenser system of AHWR using APSRA methodology[J]. Reliability engineering and system safety,2009,94:1064-1075
[83]Zhou T., Chen J., Luo F., et al. Fuzzy PSA evaluation method for passive residual heat removal system[J]. Nuclear Engineering and Design,2012,24:230-235
[84]刘晓晶.混合能谱超临界水冷堆堆芯热工与物理性能的研究[D].上海：上海交通大学,2010
[85]广东核电培训中心.900MW压水堆核电站系统与设备[M].北京：原子能出版社,2004
[86]Marleau G., Hebert A., Roy R.. A user guide for dragon version4[R]. IGE-294, 2007
[87]Marleau G.. Dragon theory manual Part 1:Collision probability calculations[R]. IGE-236,2001
[88]吴宏春,巨海涛,姚栋.复杂几何燃料组件的参数计算[J].原子能科学技术,2006,40(4)：433-438
[89]刘晓壮,周涛,王增辉,等.DRAGON程序在超临界水冷堆应用的可行性研究[C].973项目《超临界水冷堆关键科学问题的基础研究》中期学术讨论会,2008
[90]弗兰克P.英克鲁佩勤,大卫P.德维特,狄奥多尔L.伯格曼,等.传热和传质基本原理[M].北京：化学工业出版社,2007
[91]马建隆,宋之平,吴民强,等.实用热工手册,北京：水利电力出版社,1988
[92]谢仲生.核反应堆物理分析[M].西安：西安交通大学出版社,北京：原子能 出版社,2004
[93]谢仲生,核反应堆物理数值计算[M].北京：原子能出版社,1997
[94]谢仲生,尹邦华,罗经宇.核反应堆物理分析[M].北京：原子能出版社,1996
[95]俞冀阳,贾宝山.反应堆热工水力学[J].北京：清华大学出版社,2003
[96]Hu P.. Coupled neutronics/thermal-hydraulics analysis of supercritical water reactor[D]. University of Wisconsin- Madison,2008
[97]Sekki D., Hebert A.. A user guide for donjon version4[R]. IGE-300,2007
[98]汤晓斌,谢芹,耿长冉,等.基于MCNP的超临界水堆堆芯建模及中子通量计算[J].科技导报,2012,30(20)：39-43
[99]徐琪,王侃,李世悦,等.反应堆时空动力学蒙卡改进准静态方法研究[C].第十四届反应堆数值计算和粒子输运学术会议,中国银川,2012
[100]李世悦.基于蒙特卡罗方法的时空动力学改进准静态方法研究[D].清华大学,2011
[101]徐李,含热工反馈的快堆中子时空动力学计算分析研究[D].清华大学,2012
[102]廖承奎,谢仲生.耦合的PWR三维物理与热工-水力堆芯瞬态分析程序系统NLS ANMT/COBRA-I V[J].核动力工程,2003,24(5)：412-416
[103]Ishiwatari Y., Oka Y., Koshizuka S., et al. Control of high temperature supercritical pressure light water cooled and moderated reactor with water rods[J]. Journal of Nuclear Science and Technology,2003,40(5):298-306.
[104]洪德训.不同间隙矩形窄通道自然循环换热特性研究[D].北京：华北电力大学,2013
[105]Nakatsuka T., Oka Y., Koshizuka S.. Startup thermal considerations for supercritical-pressure light water-cooled reactors[J]. Nuclear Technology,2001, 134,221-230
[2]陆道纲,彭常宏.超临界水冷堆评述[J].原子能科学技术,2009,43(8)：743-749
[3]Oka Y., Koshizuka S., Ishiwatari Y., et al. Super light water reactors and super fast reactors[M]. New York:Springer,2010
[4]Yamaji A., Kamei K., Oka Y, et al. Improved core design of the high temperature supercritical-pressure light water reactor[J]. Annals of Nuclear Energy,2005(32): 651-670
[5]黄禹,沈飚,张鹏.超临界流体传热综述[J].制冷技术,2008,36(10)：44-50
[6]Cheng X. Supercritical water reactor-fuel qualification test (SCWR-FQT). http://www.ifrt.kit.edu/english/101_290.php
[7]刘华,周世荣,巢哲雄.我国应对福岛核事故的措施及启示[J].中国核工业,2011,10：18-19
[8]龚曙光.ANSYS工程应用实例解析[M].北京：机械工业出版社,2003
[9]程旭.超临界水冷堆是我国水冷堆技术路线的自然发展[J].核动态,2007,4：26-28
[10]Department of Energy. A Technology Roadmap for Generation Ⅳ Nuclear Energy Systems-Technical Roadmap Report[R]. GIF-002-00,2002
[11]Dobashi K., Kimura A., Oka Y., et al. Conceptual design of a high temperature power reactor cooled and moderated by Supercritical Light Water[J]. Annals of Nuclear Energy,1998,25(8):487-505
[12]Yamada K., Ookawa M., Asanuma Y, et al. Recent activities and future plan of thermal-spectrum SCWR development in Japan[C]. The 3rd Int. Symposium on SCWR-Design and Technology,2007, Shanghai, China
[13]Mukohara T., Koshizuka S., Oka Y. Core design of a high-temperature fast reactor cooled by supercritical light water[J]. Annals of Nuclear Energy,1999,26: 1423-1436
[14]Oka Y, Koshizuka S.. Supercritical-pressure, once-through cycle light water cooled reactor [J]. Journal of Nuclear Science and Technology,2001,38(12): 1081-1089
[15]Yoo J., Ishiwatari Y, Oka Y, et al. Conceptual design of compact supercritical water-cooled fast reactor with thermal hydraulic coupling [J]. Annals of Nuclear Energy,2006,33:945-956
[16]Kamei K., Yamaji A., Ishiwatari Y., et al. Fuel and core design of super light water reactor with low leakage fuel loading pattern[J]. Journal of Nuclear Science and Technology,2006,43(2):129-139
[17]Yoo J., Oka Y., Ishiwatari Y., et al. Thermo-mechanical analysis of supercritical pressure light water-cooled fast reactor fuel rod by FEMAXI-6 code[J]. Annals of Nuclear Energy,2006,33:1379-1390
[18]Hofmeister J., Waata C., Starflinger J., et al. Fuel assembly design study for a reactor with supercritical water[J]. Nuclear Engineering and Design,2007,237: 1513-1521
[19]Fischer K., Schulengberg T., Laurien E.. Design of a supercritical water-cooled reactor with a three-pass core arrangement[J]. Nuclear Engineering and Design, 2009,239:800-812
[20]Schulengberg T., Starflinger J., Marsault P., et al. European supercritical water cooled reactor[J]. Nuclear Engineering and Design,2011,241:3505-3513
[21]Torgerson D.F., Shalaby B.A., Pang S.. CANDU technology for generation III+ and IV reactors[J]. Nuclear Engineering and Design,2006,236:1565-1572
[22]Yang P., Cao L.Z., Wu H.C., et al. Core design study on CANDU-SCWR with 3D neutronics/thermal-hydraulics coupling[J]. Nuclear Engineering and Design,2011, 241(12):4714-4719
[23]Buongiorno J., MacDonald P.E.. Supercritical water reactor (SCWR) progress report for the FY-03 generation-Ⅳ R&D activities for the development of the SCWR in the U.S.[R]. INEEL/EXT-03-01210,2003
[24]程旭,刘晓晶.混合能谱超临界水堆堆芯设计分析[J].核科学与工程,2009,29(1)：43-49
[25]Yang T., Liu X.J., Cheng X.. Optimization of multilayer fuel assemblies for supercritical water-cooled reactors with mixed neutron spectrum[J]. Nuclear engineering and design,2012,249:159-165.
[26]超临界水堆特定条件下堆芯物理特性及其分析方法的基础研究报告[R].2007BC209807,2011
[27]孙灿辉.超临界水堆MOX燃料物理热工特性研究[D].北京：华北电力大学,2012
[28]严家騄,余晓福,王永青.水和水蒸气热力性质图表[M].北京：高等教育出版社,2008
[29]马庆芳,方荣生,项立成等.实用热物理性质手册[M].北京：中国农业机械出版社,1986
[30]Yang X.B., Su G.H., Tian W.X., et al. Numerical study on flow and heat transfer characteristics in the rod bundle channels under super critical pressure condition[J]. Annals of nuclear energy,2010,37:1723-1734.
[31]Cheng X., Liu X.J., Gu H.Y.. Fluid-to-fluid scaling of heat transfer in circular tubes cooled with supercritical fluids[J]. Nuclear Engineering and Design,2011, 241:498-508
[32]张曙明,陈玉宙,赵民富.竖直圆管内跨临界压力区水对流传热数值模拟研究[J].原子能科学技术,2009,43(6)：491-495.
[33]张亚奇.超临界压力下竖直上升管传热分析与回归评价[D].上海：上海交通大学,2008.
[34]Shang Z.. CFD investigation of vertical rod bundles of supercritical water-cooled nuclear reactor[J]. Nuclear Engineering and Design,2009, 239:2562-2572.
[35]文彦,高超,秋穗正,等.矩形窄缝通道内水稳态和瞬态流动换热特性实验[J].核动力工程,2010,31(1)：28-32
[36]刘平,周涛,张明,等.自然循环条件下窄通道ONB点影响因素灰色关联度研究[J].核动力工程,2011,32(4)：29-32
[37]张明,周涛,盛程,等.窄通道潜热沸腾起始点计算模型的分析[J].核动力工程,2011,32(3)：73-76
[38]Yamaji A., Oka Y., Koshizuka S.. Three-dimensional core design of high temperature supercritical-pressure light water reactor with neutronic and thermal-hydraulic coupling[J]. Journal of Nuclear Science and Technology,2005, 42(1):8-19
[39]Reiss T., Feher S., Czifrus S.Z.. Coupled neutronics and thermohydraulics calculations with burn-up for HPLWRs[J]. Progress in Nuclear Energy,2008,50: 52-61
[40]Monti L., Starflinger J., Schulenberg T.. Development of a coupled neutronic/ thermal-hydraulic tool with multi-scale capabilities and applications to HPLWR core analysis[J]. Nuclear Engineering and Design,2011,241:1579-1591
[41]胡珀,杨燕华.超临界水堆系统分析程序的改进[J].原子能科学技术,2009,43(6)：548-551
[42]Shan J.Q., Zhang B., Li C.Y., et al. SCWR subchannel code ATHAS development and CANDU-SCWR analysis[J]. Nuclear Engineering and Design,2009,239: 1979-1987
[43]刘晓晶,程旭.混合能谱超临界水堆堆芯热工-物理性能分析[J].原子能科学技术,2009,43(6)：538-542
[44]张鹏,王侃,李满仓.均匀化与多群蒙卡输运研究[C].核反应堆系统设计技术重点实验室2010年会,2010
[45]安萍,姚栋.超临界水堆反应堆物理-热工水力耦合程序系统MCATHAS的开发[J].核动力工程,2010,31(6)：52-55
[46]刘占权,蒋朱敏,将校丰,等.超临界水堆堆芯轴向一维物理热工耦合稳态分析[J].核科学与工程,2009,29(1)：16-21
[47]刘晓壮,周涛,李臻洋,等.超临界水堆燃料组件核热耦合计算分析[C].核反应堆系统设计技术国家级重点实验室2009年报,2009
[48]李臻洋.超临界水堆物理热工程序研究[D].华北电力大学,2011
[49]Sun C.H., Zhou T., Hou Z.S., et al. Improving on assembly of SCWR using MOX fuel[J]. Advanced materials research,2012,347-353:1633-1636
[50]Ishiwatari Y., Oka Y., Koshizuka S., et al. Safety of Super LWR(I):Safety System Design[J]. Journal of Nuclear Science and Technology,2005,42(11):927-934
[51]Ishiwatari Y, Oka Y, Koshizuka S., et al. Safety of Super LWR(II):Safety Analysis at Supercritical Pressure[J]. Journal of Nuclear Science and Technology, 2005,42(11):935-948
[52]Lee J.H., Koshizuka S., Oka Y. Development of a LOCA analysis code for the supercritical-pressure light water cooled reactors[J]. Annual of Nuclear Energy, 1998,25(16):1341-1361
[53]Ishiwatari Y, Oka Y., Koshizuka S., et al. LOCA analysis of Super LWR[J]. Journal of Nuclear Science and Technology,2006,43(3):231-241
[54]Ishiwatari Y., Oka Y, Koshizuka S., et al. ATWS characteristics of Super LWR with/without alternative action[J]. Journal of Nuclear Science and Technology, 2007,44(4):572-589
[55]Ikejiri S., Ishiwatari Y., Oka Y. Safety analysis of a supercritical-pressure water-cooled fast reactor under supercritical pressure[J]. Nuclear Engineering and Design,2010,240:1218-1228
[56]Maraczy Cs., Kereszturi A., Trosztel I., et al. Safety analysis of reactivity initiated accidents in a HPLWR reactor by the coupled ATHLET-KIKO3D code[J]. Progress in Nuclear Energy,2010,52:190-196
[57]Xu Z.H., Hou D., Fu S.W., et al. Loss of flow accident and its mitigation measures for nuclear system with SCWR-M[J]. Annals of Nuclear Energy,2011, 38:2634-2644
[58]周翀,刘晓晶,杨燕华,等.ATHLET-SC程序的开发及适用性分析[J].原子能科学技术,2009,43(6)：556-560
[59]Zhu D.H., Zhao H., Tian W.X., et al. Development of TACOS code for loss of flow accident analysis of SCWR with mixed spectrum core[J]. Progress in Nuclear Energy,2012,54:150-161
[60]Zhu D.H., Tian W.X., Zhao H., et al. Comparatice study of transient thermal-hydraulic characteristics of SCWRs with different core design[J]. Annals of Nuclear Energy,2013,51:135-145
[61]胡雨.控制系统对超临界水堆事故影响分析[D].北京：华北电力大学,2010
[62]周涛,李臻洋,王晗丁,等.超临界水堆核热耦合及其瞬态相关控制方式研究[C].核电站新技术交流研讨会论文集,2010
[63]王晗丁,周涛,陈娟,等.超临界快堆给水控制失效瞬态控制分析[J].核动力工程,2011,32(5)：18-22
[64]王晗丁,周涛,陈娟,等.超临界水堆控制条件下扰动特性分析[J].原子能科学技术,2012,46(5)：555-560
[65]李磊,张志俭.并联通道瞬态流量分配方法研究[J].核动力工程,2010,31(5)：97-101
[66]蔡章生,桂学文,于雷.反应堆时空动力学方程的解法研究[J].海军工程大学学报,2006,18(3)：28-29
[67]宋英明,马远乐,单文志,等.高温气冷堆堆芯中子时空动力学模拟计算[J].计算物理,2009,26(6)：911-916
[68]廖承奎,求解中子扩散方程的半解析节块方法[J].原子能科学技术,2009,43(6)：496-500
[69]夏榜样,谢仲生.通量展开节块法求解六角形几何三维多群中子扩散方程[J].西安交通大学学报,2006,40(1)：84-87
[70]赵文博.瞬态节块格林函数方法及其与热工-水力耦合研究[D].清华大学,2012
[71]于平安,朱瑞安,喻真烷,等.核反应堆热工分析[M].上海：上海交通大学出版社,2002
[72]赵冬建,路璐,史国宝.超临界水堆子通道分析[J].原子能科学技术,2009,43(6)：543-547
[73]Gou J.L., Shang Z., Yuki I. et al. CFD analysis of heat transfer in subchannel of a super fast reactor[J]. nuclear engineering and design,2010,240:1819-1829
[74]Ivanov K., Avramova M.. Challenges in coupled thermal-hydraulics and neutronics simulations for LWR safety analysis[J]. Annals of Nuclear Energy, 2007,34:501-513
[75]孔衍,于雷,鄢炳火.自然循环下堆芯热工水力与时空中子动力学的耦合计算[J].船海工程,2010,39(5)：201-204
[76]宋英明,周志伟,马远乐.高温气冷堆温度反馈的集总参数法模拟[J].清华大学学报,2009,49(9)：92-95
[77]Volkanovski A., Cepin M.. Implication of PSA uncertainties on risk-informed decision making[J]. Nuclear Engineering and Design,2011,241(4):1108-1113
[78]Volkanovshi A., Cepin M., Mavko B.. Application of the fault tree analysis for assessment of power system reliability[J]. Reliability engineering and system safety,2009,94:1116-1127
[79]朱继洲.核反应堆安全分析[M].北京：原子能出版社,西安交大出版社,2002
[80]Durga K., Gopika V., Sanyasi V.V.S, et al. Dynamic fault tree analysis using Monte Carlo Simulation in probabilistic safety assessment[J]. Reliability engineering and system safety,2009,94:872-883
[81]Sajith T., John A., Parthasarathy U., etc. Passive system reliability analysis using Response Conditioning Method with an application to failure frequency estimation of Decay Heat Removal of PFBR[J]. Nuclear Engineering and Design, 2011,241(6):2257-2270
[82]Nayak A.K., Jain V., Gartia M.R., et al. Reliability assessment of passive isolation condenser system of AHWR using APSRA methodology[J]. Reliability engineering and system safety,2009,94:1064-1075
[83]Zhou T., Chen J., Luo F., et al. Fuzzy PSA evaluation method for passive residual heat removal system[J]. Nuclear Engineering and Design,2012,24:230-235
[84]刘晓晶.混合能谱超临界水冷堆堆芯热工与物理性能的研究[D].上海：上海交通大学,2010
[85]广东核电培训中心.900MW压水堆核电站系统与设备[M].北京：原子能出版社,2004
[86]Marleau G., Hebert A., Roy R.. A user guide for dragon version4[R]. IGE-294, 2007
[87]Marleau G.. Dragon theory manual Part 1:Collision probability calculations[R]. IGE-236,2001
[88]吴宏春,巨海涛,姚栋.复杂几何燃料组件的参数计算[J].原子能科学技术,2006,40(4)：433-438
[89]刘晓壮,周涛,王增辉,等.DRAGON程序在超临界水冷堆应用的可行性研究[C].973项目《超临界水冷堆关键科学问题的基础研究》中期学术讨论会,2008
[90]弗兰克P.英克鲁佩勤,大卫P.德维特,狄奥多尔L.伯格曼,等.传热和传质基本原理[M].北京：化学工业出版社,2007
[91]马建隆,宋之平,吴民强,等.实用热工手册,北京：水利电力出版社,1988
[92]谢仲生.核反应堆物理分析[M].西安：西安交通大学出版社,北京：原子能 出版社,2004
[93]谢仲生,核反应堆物理数值计算[M].北京：原子能出版社,1997
[94]谢仲生,尹邦华,罗经宇.核反应堆物理分析[M].北京：原子能出版社,1996
[95]俞冀阳,贾宝山.反应堆热工水力学[J].北京：清华大学出版社,2003
[96]Hu P.. Coupled neutronics/thermal-hydraulics analysis of supercritical water reactor[D]. University of Wisconsin- Madison,2008
[97]Sekki D., Hebert A.. A user guide for donjon version4[R]. IGE-300,2007
[98]汤晓斌,谢芹,耿长冉,等.基于MCNP的超临界水堆堆芯建模及中子通量计算[J].科技导报,2012,30(20)：39-43
[99]徐琪,王侃,李世悦,等.反应堆时空动力学蒙卡改进准静态方法研究[C].第十四届反应堆数值计算和粒子输运学术会议,中国银川,2012
[100]李世悦.基于蒙特卡罗方法的时空动力学改进准静态方法研究[D].清华大学,2011
[101]徐李,含热工反馈的快堆中子时空动力学计算分析研究[D].清华大学,2012
[102]廖承奎,谢仲生.耦合的PWR三维物理与热工-水力堆芯瞬态分析程序系统NLS ANMT/COBRA-I V[J].核动力工程,2003,24(5)：412-416
[103]Ishiwatari Y., Oka Y., Koshizuka S., et al. Control of high temperature supercritical pressure light water cooled and moderated reactor with water rods[J]. Journal of Nuclear Science and Technology,2003,40(5):298-306.
[104]洪德训.不同间隙矩形窄通道自然循环换热特性研究[D].北京：华北电力大学,2013
[105]Nakatsuka T., Oka Y., Koshizuka S.. Startup thermal considerations for supercritical-pressure light water-cooled reactors[J]. Nuclear Technology,2001, 134,221-230