聚变反应堆超临界水冷包层内对流换热及应力分析研究
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摘要
能源问题是当今人类社会面临的重要问题之一,目前主要使用的化石能源热值低、资源匮乏并且对环境造成很大的污染,不适合长期使用。核能是新能源家族中的重要成员,包括裂变能和聚变能两种主要形式。裂变能是重金属元素的原子通过裂变所释放出的巨大能量,目前已经实现商用化。核能的另一种形式就是目前尚未实现商用化的聚变能。聚变反应的燃料在自然界中大量存在,几乎“取之不尽,用之不竭”,并且其反应产物放射毒性较低,也不产生污染环境的硫、氮氧化物,不释放温室气体,所以聚变能被看成是清洁、安全和可再生的新型能源。
在产生聚变能的聚变反应堆中,包层是实现高环境适应性和低发电成本的聚变能应用的关键能量转换部件,作为堆内构建它包裹着等离子体,由第一壁和增殖区构成。聚变堆包层的主要功能是:产生和输运聚变反应所需的氚;将聚变中子的能量转换成热能并通过其内部的冷却通道将包层内的热量载出;第一壁材料还要抵御高温等离子体的热流;同时还要起到部分屏蔽作用。在众多包层设计和冷却方案中,超临界水冷包层是其中的一种,采用25MPa下的超临界水作为冷却剂。本文主要研究了超临界水冷包层第一壁结构材料中的温度和应力分布,以及增殖区冷却剂通道内的混合对流换热。
本文针对聚变反应堆超临界水冷包层第一壁结构,通过商用计算流体力学软件CFX和有限元分析软件ANSYS Workbench中的SIMULATION模块进行单向流固耦合,对现有设计的超临界水冷包层第一壁温度和应力进行数值模拟,验证了流固耦合分析手段的可行性。在此基础上,进一步研究了不同湍流模型对计算结果的影响,以及比较了不同冷却方案下第一壁结构材料中的最高温度和最大应力。同时,研究了流道截面对结构材料中最高温度和最大应力的影响,综合这些影响因素对第一壁结构作了优化设计,有效降低了结构材料中的最高温度和最大应力。
本文针对超临界水冷包层增殖区,利用商用计算流体力学软件CFX研究了冷却剂质量流量对通道内对流换热效果的影响。数值模拟结果表明超临界水在过拟临界点时剧烈的物性变化给冷却通道内的对流换热带来了一定的影响;更大的冷却剂流量能够带来的更好的对流换热效果,这一现象随着沿程高度的增加越来越明显。同时,还分析了不同流量下浮升力对混合对流换热的影响。数值模拟结果表明,随着流量大小和流动方向(顺浮升力方向和逆浮升力方向)的变化,浮升力作用下的对流换热出现“换热加强”和“换热弱化”的现象。
Nowadays energy issue is one of the most serious issues. The mostly used chemical energy is not recommended for long-term use due to its low caloric value and pollution potential to the environment. Nuclear energy is a vital member of new energy family, including fission and fusion. Fission energy comes from the fission of heavy metal atoms releasing enormous energy. Nuclear fission has been currently achieved commercial operation. Fusion energy is not yet commercially utilized. The fuel for fusion reaction is almost inexhaustible in nature, and its reaction products are less radioactive. Therefore, fusion energy is considered to be clean, safe and renewable.
In fusion reactor, blanket is the key component for energy transformation. It enwraps components in plasma reactor and made up by the first wall and breeder zone. The primary functions of fusion reactor blanket are as follows: producing and transporting tritium; transferring neutron energy into heat energy and removing heat by coolant flowing through the cooling channel. The first wall has to withstand the heat flux from high-temperature plasma and also partially achieve shielding effect. Supercritical water (SCW) cooled blanket is one kind of blanket design proposals. It is cooled by supercritical water at pressure of 25MPa.
This paper investigates the thermo-structural performance of the first wall and convective heat transfer in breeder zone cooling channel of SCW cooled blanket. In this study, a coupled code ANSYS CFX /ANSYS Workbench is used to analysis the temperature and stress distribution of structural material by one-way coupling approach. Firstly the temperature and stress in the first wall structural material of the existing SCW cooled blanket design is simulated to verify the fluid-structure interaction analysis method. Then different turbulence models are used in numerical simulation to find out their influence on temperature and stress distribution in the first wall. Different cooling schemes and different geometrical configurations of flow channel of the first wall are taken into consideration. Based on the results achieved so far, an optimized design solution is suggested on the modification of the design structure and geometric configuration of flow channels, which can effectively reduced the maximum temperature and stress of the structural material.
Furthermore, convective heat transfer in the cooling channel of breeder zone is investigated by CFX simulation. Numerical results show heat transfer reduction near the pseudo-critical temperature due to the dramatic change of physical properties of SCW. The results also indicate that larger coolant flow rate leads to improved heat transfer, which becomes more apparent along the flow pass. At the same time, the influence of buoyancy force on the heat transfer reduction at mixed convection conditions is studied. Simulation results show that phenomenon of“heat strengthened" and "heat weakened" occur alternatively according to the changes of flow rate and flow direction (along or again buoyancy direction).
在产生聚变能的聚变反应堆中,包层是实现高环境适应性和低发电成本的聚变能应用的关键能量转换部件,作为堆内构建它包裹着等离子体,由第一壁和增殖区构成。聚变堆包层的主要功能是:产生和输运聚变反应所需的氚;将聚变中子的能量转换成热能并通过其内部的冷却通道将包层内的热量载出;第一壁材料还要抵御高温等离子体的热流;同时还要起到部分屏蔽作用。在众多包层设计和冷却方案中,超临界水冷包层是其中的一种,采用25MPa下的超临界水作为冷却剂。本文主要研究了超临界水冷包层第一壁结构材料中的温度和应力分布,以及增殖区冷却剂通道内的混合对流换热。
本文针对聚变反应堆超临界水冷包层第一壁结构,通过商用计算流体力学软件CFX和有限元分析软件ANSYS Workbench中的SIMULATION模块进行单向流固耦合,对现有设计的超临界水冷包层第一壁温度和应力进行数值模拟,验证了流固耦合分析手段的可行性。在此基础上,进一步研究了不同湍流模型对计算结果的影响,以及比较了不同冷却方案下第一壁结构材料中的最高温度和最大应力。同时,研究了流道截面对结构材料中最高温度和最大应力的影响,综合这些影响因素对第一壁结构作了优化设计,有效降低了结构材料中的最高温度和最大应力。
本文针对超临界水冷包层增殖区,利用商用计算流体力学软件CFX研究了冷却剂质量流量对通道内对流换热效果的影响。数值模拟结果表明超临界水在过拟临界点时剧烈的物性变化给冷却通道内的对流换热带来了一定的影响;更大的冷却剂流量能够带来的更好的对流换热效果,这一现象随着沿程高度的增加越来越明显。同时,还分析了不同流量下浮升力对混合对流换热的影响。数值模拟结果表明,随着流量大小和流动方向(顺浮升力方向和逆浮升力方向)的变化,浮升力作用下的对流换热出现“换热加强”和“换热弱化”的现象。
Nowadays energy issue is one of the most serious issues. The mostly used chemical energy is not recommended for long-term use due to its low caloric value and pollution potential to the environment. Nuclear energy is a vital member of new energy family, including fission and fusion. Fission energy comes from the fission of heavy metal atoms releasing enormous energy. Nuclear fission has been currently achieved commercial operation. Fusion energy is not yet commercially utilized. The fuel for fusion reaction is almost inexhaustible in nature, and its reaction products are less radioactive. Therefore, fusion energy is considered to be clean, safe and renewable.
In fusion reactor, blanket is the key component for energy transformation. It enwraps components in plasma reactor and made up by the first wall and breeder zone. The primary functions of fusion reactor blanket are as follows: producing and transporting tritium; transferring neutron energy into heat energy and removing heat by coolant flowing through the cooling channel. The first wall has to withstand the heat flux from high-temperature plasma and also partially achieve shielding effect. Supercritical water (SCW) cooled blanket is one kind of blanket design proposals. It is cooled by supercritical water at pressure of 25MPa.
This paper investigates the thermo-structural performance of the first wall and convective heat transfer in breeder zone cooling channel of SCW cooled blanket. In this study, a coupled code ANSYS CFX /ANSYS Workbench is used to analysis the temperature and stress distribution of structural material by one-way coupling approach. Firstly the temperature and stress in the first wall structural material of the existing SCW cooled blanket design is simulated to verify the fluid-structure interaction analysis method. Then different turbulence models are used in numerical simulation to find out their influence on temperature and stress distribution in the first wall. Different cooling schemes and different geometrical configurations of flow channel of the first wall are taken into consideration. Based on the results achieved so far, an optimized design solution is suggested on the modification of the design structure and geometric configuration of flow channels, which can effectively reduced the maximum temperature and stress of the structural material.
Furthermore, convective heat transfer in the cooling channel of breeder zone is investigated by CFX simulation. Numerical results show heat transfer reduction near the pseudo-critical temperature due to the dramatic change of physical properties of SCW. The results also indicate that larger coolant flow rate leads to improved heat transfer, which becomes more apparent along the flow pass. At the same time, the influence of buoyancy force on the heat transfer reduction at mixed convection conditions is studied. Simulation results show that phenomenon of“heat strengthened" and "heat weakened" occur alternatively according to the changes of flow rate and flow direction (along or again buoyancy direction).
引文
[1]石星军,陈湘仁.聚变裂变混合堆展望[J].青岛大学学报(自然科学版),2009,22(1):91-95.
[2]霍小兰.核聚变研究与ITER计划[J].政策计划,2006,7:40-42.
[3]赵伟,朱远东.人类热核聚变项目[J].国外科技动态,2006,4:12-17.
[4] Feng K M, et al. Overview of design and R&D of solid breeder TBM in China [J]. Fusion Engineering and Design, 2008, 83:1149-1156.
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[8] Leshukov A Yu, et al. The RF concept of experimental breeding module for testing in ITER [J]. Fusion Engineering and Design, 2006, 81: 533-539.
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[10] Kovalenko V, et al. RF test blanket sub-module with ceramic breeder and heliumcooling for test in ITER [J]. Fusion Engineering and Design, 2006, 81:199-203.
[11] Kirillov I R, et al. RF TBMs for ITER tests [J]. Fusion Engineering and Design, 2006, 81:425-432.
[12] Enoeda M, Kosaku Y, et al. Design and technology development of solid breeder blanket cooled by supercritical water in Japan [J]. Nuclear Fusion, 2003, 43:1837-1844.
[13] Tobit K, Nishio S, et al. Design study of fusion DEMO plant at JAERI [J]. Fusion Engineering and Design, 2006, 81:1151-1158.
[14] Akiba M, Ishitsuka E, et al. Development of supercritical pressure water cooled solid breeder blanket in JAERI [J]. Plasma Fusion Research, 2003, 79(9):929-934.
[15] Cheng X, Schulenberg T. Heat transfer at Supercritical Pressures-LiteratureReview and Application to an HPLWR[R]. Scientific report FZKA 6609, 2001.
[16] Shatalov G. DEMO blanket testing in ITER. Influence on reaching DEMO [J]. Fusion Engineering and Design, 2001, 56-57:39-46.
[17]黄群英,郁金南,等.聚变堆低活化马氏体钢的发展[J].核科学与工程,2004,24(1):56-64.
[18] Kohyama A, Hishinuma A, et al. The developments of ferritic steels for DEMO blanket [J]. Fusion Engineering and Design, 1998, 41:1-6.
[19] Konishi S, Nishio S, et al. DEMO plant design beyond ITER [J]. Fusion Engineering and Design, 2002, 63-64:11-17.
[20] Tavassoli A-A F, et al. Materials design data for reduced activation martensitic steel type F82H [J]. Fusion Engineering and Design, 2002, 61-62:617-628.
[21] Reimann J, et al. Thermomechanics of solid breeder and Be pebble bed materials [J]. Fusion Engineering and Design, 2002, 61-62:319-331.
[22]贾小波,等. ITER中国固态TBM中子倍增剂和氚增殖剂布置优化[J].清华大学学报(自然科学版), 2006,46:1629-1632.
[23] Akiba M, Ishitsuka E, et al. Development of supercritical pressure water cooled solid breeder blanket in JAERI [J]. Plasma Fusion Research, 2003, 79(9):929-934.
[24]陶文铨.数值传热学(第2版)[M].西安:西安交通大学出版社,2001.
[25]梁在潮.工程湍流[M].武汉:华中理工大学出版社, 1999.
[26]翟建华.计算流体力学(CFD)的通用软件[J].河北科技大学学报,2005, 26(2):160-165.
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[28] ANSYS CFX, Release 11.0
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[35] Raffray A R, et al. Breeding blanket concepts for fusion and materials requirements [J]. Journal of Nuclear Materials, 2002, 307-311:21-30.
[36] Nishio S, et al. Consideration on blanket structure for fusion DEMO plant at JAERI [J]. Fusion Engineering and Design, 2006, 81:1271-1276.
[37] Ehrlich K, et al. International strategy for fusion materials development [J]. Journal of Nuclear Materials, 2000, 283-287:79-88.
[38] Yanagi Y, et al. Nuclear and thermal analyses of supercritical-water-cooled solid breeder blanket for Fusion DEMO reactor [J]. Journal of Nuclear Science and Technology, 2001, 38(11): 1014-1018.
[39] Tsuru D, et al. Recent progress in safety assessments of Japanese water-cooled solid breeder test blanket module [J]. Fusion Engineering and Design, 2008, 83:1747-1752.
[40] Klueh R L, et al. Ferritic/martensitic steels-overview of recent results, Section 4. Ferritic/martensitic steels [J]. Journal of Nuclear Materials, 2002, 307-311:455-465.
[41]谢正瑞,等.倾斜矩形通道中湍流混合对流换热的数值分析[J].核动力工程,2009,30(1):50-55.
[42]吴宜灿,等.磁约束聚变堆及ITER实验包层模块设计研究进展[J].原子核物理评论,2006,23(2):89-95.
[43]冯开明.ITER实验包层计划综述[J].核聚变与等离子体物理,2006,26(3):161-169.
[44]吴宜灿,等.聚变发电反应堆概念设计研究[J].核科学与工程,2005,25(1):76-85.
[45]许增裕.国际热核实验堆的建造与聚变堆材料研究[J].原子能科学技术,2005,39:46-52.
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[2]霍小兰.核聚变研究与ITER计划[J].政策计划,2006,7:40-42.
[3]赵伟,朱远东.人类热核聚变项目[J].国外科技动态,2006,4:12-17.
[4] Feng K M, et al. Overview of design and R&D of solid breeder TBM in China [J]. Fusion Engineering and Design, 2008, 83:1149-1156.
[5] Ying A,et al. Design Description Document for the U.S. Helium-Cooled Solid Breeder Test Blanket Module [Report]. Report to the ITER Test Blanket Working Group (TBWG), 2005, University of California Los Angeles.
[6] Boccaccini L V, et al. Design Description Document for the European Helium Cooled Pebble Bed (HCPB) Test Blanket Modules.2001, Forschungszentrum Karlsruhe.
[7] Boccaccini L V et al. EU Designs and Efforts on ITER HCPB TBM.ITERTBM Project Meeting UCLA, USA, February 23-25, 2004.
[8] Leshukov A Yu, et al. The RF concept of experimental breeding module for testing in ITER [J]. Fusion Engineering and Design, 2006, 81: 533-539.
[9] Leshukov A, et al. Ceramic helium-cooled blanket test module [J]. Fusion Engineering and Design, 2000, 49-50: 591-598.
[10] Kovalenko V, et al. RF test blanket sub-module with ceramic breeder and heliumcooling for test in ITER [J]. Fusion Engineering and Design, 2006, 81:199-203.
[11] Kirillov I R, et al. RF TBMs for ITER tests [J]. Fusion Engineering and Design, 2006, 81:425-432.
[12] Enoeda M, Kosaku Y, et al. Design and technology development of solid breeder blanket cooled by supercritical water in Japan [J]. Nuclear Fusion, 2003, 43:1837-1844.
[13] Tobit K, Nishio S, et al. Design study of fusion DEMO plant at JAERI [J]. Fusion Engineering and Design, 2006, 81:1151-1158.
[14] Akiba M, Ishitsuka E, et al. Development of supercritical pressure water cooled solid breeder blanket in JAERI [J]. Plasma Fusion Research, 2003, 79(9):929-934.
[15] Cheng X, Schulenberg T. Heat transfer at Supercritical Pressures-LiteratureReview and Application to an HPLWR[R]. Scientific report FZKA 6609, 2001.
[16] Shatalov G. DEMO blanket testing in ITER. Influence on reaching DEMO [J]. Fusion Engineering and Design, 2001, 56-57:39-46.
[17]黄群英,郁金南,等.聚变堆低活化马氏体钢的发展[J].核科学与工程,2004,24(1):56-64.
[18] Kohyama A, Hishinuma A, et al. The developments of ferritic steels for DEMO blanket [J]. Fusion Engineering and Design, 1998, 41:1-6.
[19] Konishi S, Nishio S, et al. DEMO plant design beyond ITER [J]. Fusion Engineering and Design, 2002, 63-64:11-17.
[20] Tavassoli A-A F, et al. Materials design data for reduced activation martensitic steel type F82H [J]. Fusion Engineering and Design, 2002, 61-62:617-628.
[21] Reimann J, et al. Thermomechanics of solid breeder and Be pebble bed materials [J]. Fusion Engineering and Design, 2002, 61-62:319-331.
[22]贾小波,等. ITER中国固态TBM中子倍增剂和氚增殖剂布置优化[J].清华大学学报(自然科学版), 2006,46:1629-1632.
[23] Akiba M, Ishitsuka E, et al. Development of supercritical pressure water cooled solid breeder blanket in JAERI [J]. Plasma Fusion Research, 2003, 79(9):929-934.
[24]陶文铨.数值传热学(第2版)[M].西安:西安交通大学出版社,2001.
[25]梁在潮.工程湍流[M].武汉:华中理工大学出版社, 1999.
[26]翟建华.计算流体力学(CFD)的通用软件[J].河北科技大学学报,2005, 26(2):160-165.
[27]方坤.计算流体力学的几种常用软件[J].煤炭技术, 2006, 25(12).
[28] ANSYS CFX, Release 11.0
[29]谢祚水.计算结构力学[M].武昌:华中科技大学出版社, 2004.
[30]张洪信.有限元基础理论与ANSYS应用[M].北京:机械工业出版社, 2006.
[31]周昌玉,贺小虎.有限元分析的基本方法及工程应用[M].北京:化学工业出版社, 2006.
[32]倪栋,et al.通用有限元分析ANSYS7.0实例精解[M].北京:电子工业出版社,2003.
[33] Release 11.0 Documentation for ANSYS Workbench
[34]刘志远,郑源,张文佳,司佳钧. ANSYS-CFX单向流固耦合分析的方法[J].水利水电工程设计,2009,28(2):29-31.
[35] Raffray A R, et al. Breeding blanket concepts for fusion and materials requirements [J]. Journal of Nuclear Materials, 2002, 307-311:21-30.
[36] Nishio S, et al. Consideration on blanket structure for fusion DEMO plant at JAERI [J]. Fusion Engineering and Design, 2006, 81:1271-1276.
[37] Ehrlich K, et al. International strategy for fusion materials development [J]. Journal of Nuclear Materials, 2000, 283-287:79-88.
[38] Yanagi Y, et al. Nuclear and thermal analyses of supercritical-water-cooled solid breeder blanket for Fusion DEMO reactor [J]. Journal of Nuclear Science and Technology, 2001, 38(11): 1014-1018.
[39] Tsuru D, et al. Recent progress in safety assessments of Japanese water-cooled solid breeder test blanket module [J]. Fusion Engineering and Design, 2008, 83:1747-1752.
[40] Klueh R L, et al. Ferritic/martensitic steels-overview of recent results, Section 4. Ferritic/martensitic steels [J]. Journal of Nuclear Materials, 2002, 307-311:455-465.
[41]谢正瑞,等.倾斜矩形通道中湍流混合对流换热的数值分析[J].核动力工程,2009,30(1):50-55.
[42]吴宜灿,等.磁约束聚变堆及ITER实验包层模块设计研究进展[J].原子核物理评论,2006,23(2):89-95.
[43]冯开明.ITER实验包层计划综述[J].核聚变与等离子体物理,2006,26(3):161-169.
[44]吴宜灿,等.聚变发电反应堆概念设计研究[J].核科学与工程,2005,25(1):76-85.
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