水冷包层模块第一壁流动传热特性初步分析

本文基于我国聚变工程实验堆水冷包层优化设计与安全分析的要求,针对水冷包层模块第一壁的流动传热特性进行三维数值模拟研究。采用计算流体力学方法,建立了水冷包层模块第一壁的三维数值模型,研究流量分配的特点以及温度分布情况,分析与评估在稳态工况、瞬态工况及失流事故下的水冷包层模块第一壁传热能力。研究结果表明,不同冷却管间存在流量分配不均匀的现象;在稳态工况下,水冷包层模块第一壁具有较好的传热能力,瞬态工况下水冷包层模块能够有效地导出反应堆热量;失流事故下冷却管内温度短时间上升至系统压力下的饱和温度,有待进一步研究。相关研究为优化包层第一壁传热设计提供参考,并为今后聚变堆的安全分析提供依据。     

低温堆上空腔失水事故模拟实验研究

叙述了位于低温堆上空腔位置的中小尺寸管道破裂引起的小破口失水事故研究。在核供热堆热工水力学实验系统ＨＲＴＬ－５上，对停堆后堆内有剩余功率的上空腔小破口失水事故进行了模拟实验，分析了小破口失水事故发生后，系统运行重要参数的变化，给出了上空腔小破口失水事故对低温安全性的影响。     

非能动安全系统应用于聚变-裂变混合能源堆的可行性研究

将非能动安全系统应用于聚变-裂变混合能源堆,使用RELAP5对混合能源堆包层、一回路系统、部分二回路系统和非能动安全系统进行建模,对主冷却剂泵卡转子事故和冷管段小破口失水事故进行瞬态计算和分析研究。计算结果显示,该非能动安全系统能够满足选取的3种反应堆事故的安全要求,验证非能动安全系统应用于聚变-裂变混合能源堆的可行性。非能动余热排出系统热交换器的面积增大后,能有效地导出包层衰变热;增大堆芯补水箱的容积对增大反应堆安全裕度有明显效果。     

聚变增殖堆液态金属Li自冷包层的热工水力分析

完成了托卡马克商用混合堆 TCB(Tokamak Commercial Breeder)Li 自冷包层设计的热工水力分析,讨论了热工水力设计中的一些关键问题。用两维有限元热传导程序 AYER 计算了 TCB 包层的温度分布,用液态金属 MHD(Magnetohydraudynamic)压降公式计算了包层的压降。同时,还分析了包层冷却剂丧失事故 LOCA 的瞬态热工过程。分析表明,正常工况下,包层结构材料最高温度,结构材料与冷却剂界面最高温度,以及包层总压降都满足堆设计要求。在 LOCA 工况下,如果停堆后1小时内包层中的燃料球能够借助重力卸出包层,第一壁和包层是安全的,并且不会受到损伤。     

中国双功能铅锂实验包层系统In-box LOCA事故瞬态压力传播特征分析

中国双功能铅锂实验包层系统(CN DFLL TBS)发生氦气—铅锂流道间破口(In-box LOCA)事故时,8 MPa高压氦气喷向低压铅锂增殖区,高压以压力波形式从包层模块(TBM)的铅锂增殖区传播到铅锂辅助系统(LLAS),造成系统超压,威胁包层安全。本文采用RELAP5/MOD4.0软件对DFLL包层系统进行建模,开展了破口事故下的系统瞬态压力传播分析,对破口位置、面积、爆破阀起爆压力等重要参数进行敏感性分析。分析表明:不同位置破口事故下,包层压力入口最高可达16.68 MPa,包层出口处最高可达13.92 MPa;单根与10根传热管破裂事故,包层出入口压力分别增加0.97 MPa、1.68 MPa;为降低包层内部的压力峰值,可在包层模块进出口管道设置体积不小于1.2×10~(-2) m~3稳压装置。通过将铅锂辅助系统的关键部件布置在稳压装置附近,可有效保护其不超出其压力限值。     

聚变堆氚的环境安全评估

对国家863项目聚变实验增殖堆工程概要设计(FEB-E)进行了氚环境安全问题评估。FEB-E是采用液态锂作为包层氚增殖剂，每个包层模块各区之间用隔板隔开．中间通高压氦气冷却、包层第一壁和偏滤器也用氦气冷却。运用自行研制的SWITRIM程序和Sieverts’定律研究了正常工作状态下和事故状态下可能造成氚的环境污染水平。研究表明．正常工作状态下包层液态锂中的氚分压在10^-6～10^-8pa。造成氚环境污染的主要危险来自氚循环回路中的偏滤器子系统的抽出气体泄漏。因此，提高堆芯等离子体燃耗和真空系统设计性能是重要的。     

聚变实验堆ITER实验包层瞬态热工安全分析

在中国向ITER(International Thermonuclear Experiment Reactor)实验包层工作组提交的双功能锂铅实验包层模块(DFLL-TBM)设计分析的基础上,通过对DFLL-TBM系统相关的瞬态事故如真空室内部冷却剂泄漏、TBM(实验包层模块)内部冷却剂泄漏以及真空室外部冷却剂泄漏事故进行计算分析,评价DFLL-TBM对ITER在热工方面对安全的影响.结果表明:当发生瞬态事故时,DFLL-TBM有能力通过热辐射将余热排出,且包层结构不会熔化.DFLL-TBM可满足ITER在热工方面对安全的要求.     

CAP1400 MSLB事故破口谱分析及极限工况瞬态分析

CAP1400核电厂与传统的"二代"核电厂区别较大。CAP1400反应堆在AP1000的基础上进行了一系列改进。采用RELAP5/MOD3.3程序建立CAP1400核电厂模型,对主蒸汽管道破裂事故的破口谱进行分析,结果表明,直到0.058m~2的蒸汽管道破口都不会触发反应堆停堆。对于0.059～0.105 m~2的蒸汽管道破口,反应堆由超功率△T信号触发停堆。对于0.106～0.15 m~2的蒸汽管道破口,反应堆由蒸汽管道低压力安注信号触发停堆。从DNB和燃料中心熔化保护角度考虑,极限工况是破口尺寸为超功率触发停堆的最大破口尺寸0.105 m~2。对极限工况的热工水力瞬态进行研究,分析堆芯流量、热流密度、温度、压力等关键参数随时间变化的趋势。采用VIPRE程序对DNBR进行计算,得到事故对应的最小DNBR为1.914,大于验收准则1. 45,表明CAP1400反应堆在主蒸汽管道破裂事故下安全可靠。     

聚变驱动次临界堆双冷嬗变包层材料活化计算与分析

对聚变驱动次临界堆 (FDS Ⅰ )包层进行了材料活化计算与分析。利用多功能中子学程序系统VisualBUS1 .0及多群数据库HENDL1 .0 /MG进行中子输运计算 ,以获得包层各个功能区的中子注量率能谱 ;在此基础上 ,使用欧洲活化计算程序FISPACT及IAEA聚变活化数据库FENDL/A 2 .0分别对停堆初期包层不同功能区的剂量率水平和衰变余热水平、停堆后期结构材料与氚增殖剂 /冷却剂的活化性能及其杂质的控制要求进行了计算及分析。     

非能动堆芯冷却系统LOCA下冷却能力分析

本文基于机理性分析程序建立了包括反应堆一回路冷却剂系统、专设安全设施及相关二次侧管道系统的先进压水堆分析模型,对典型的小破口失水事故和大破口失水事故开展了全面分析。针对不同破口尺寸、破口位置的失水事故,分析了非能动堆芯冷却系统（PXS）中非能动余热排出系统（PRHRS）、堆芯补水箱（CMT）、安注箱（ACC）、自动卸压系统（ADS）和安全壳内置换料水箱（IRWST）等关键系统的堆芯注水能力和冷却效果。研究表明,虽然破口尺寸、破口位置会影响事故进程发展,但所有事故过程中燃料包壳表面峰值温度不超过1 477 K,且反应堆堆芯处于有效淹没状态。PXS能有效排出堆芯衰变热,将反应堆引导到安全停堆状态,防止事故向严重事故发展。     

Thermal–hydraulic and safety analysis for Chinese helium-cooled solid breeder TBM cooling system

The thermal–hydraulic behavior and safety performance of the Chinese helium-cooled solid breeder (CH HCSB) test blanket module (TBM) with helium cooling system (HCS) has been studied using RELAP5/Mod3.4 code. According to accident analysis specification for TBM, two design basis accidents including loss of off-site power and TBM first wall (FW) ex-vessel coolant pipe break are investigated. The influences of different break locations and plasma termination behaviors are analyzed comprehensively. The results show that natural circulation is established in helium cooling circuit and the TBM can be cooled effectively after loss of off-site power. It is much more critical when the pipe break occurs at the downstream side of the circulator compared with that of upstream side of the circulator. In case of a more serious accident that the ex-vessel break extends to the TBM FW, the results reveal that TBM could be cooled down by natural circulation and radiation. In addition, at the beginning of ex-vessel loss of coolant accident (LOCA), large temperature difference between break and intact TBM FW pipes is found. The accidental results finally show that the integrity of the FW can be guaranteed if the plasma is terminated with a 3 s delay time by fusion power shutdown system (FPSS) in the case of ex-vessel LOCA.     

Preliminary safety analysis of ex-vessel LOCA for the European HCPB TBM system

Ex-vessel loss of coolant accident caused by a double-ended pipe break of the helium coolant system inside port cell is considered as one of the most critical accident for the European Helium Cooled Pebble Beds Test Blanket Module (HCPB TBM) system. The resulting rapid helium blow-down causes an immediate block of the TBM cooling, which requires a prompt plasma shutdown. Even after the plasma shutdown the temperature can increase over the design limit and the accident sequence can lead up to a break of the TBM box protection after the failure of different protection systems. Thus air ingresses in the vacuum vessel from the damaged TBM system and steam from the surrounding ITER blanket and divertor structures. The evaluation of this sequence is very important for the definition of the correct protection strategy of the system. To consider all these different events a methodology has been developed in KIT combining different codes for a complete analysis of the accident. In particular, this paper shows an application of MELCOR code to model beryllium–steam reaction in a particular accidental sequence for the long term cooling.     

Analysis of loss-of-coolant and loss-of-flow accidents in the first wall cooling system of NET/ITER

This paper presents the thermal-hydraulic analysis of potential accidents in the first wall cooling system of the Next European Torus or the International Thermonuclear Experimental Reactor. Three ex-vessel loss-of-coolant accidents, two in-vessel loss-of-coolant accidents, and three loss-of-flow accidents have been analyzed using the thermal-hydraulic system analysis code RELAP5/MOD3. The analyses deal with the transient thermal-hydraulic behavior inside the cooling systems and the temperature development inside the nuclear components during these accidents. The analysis of the different accident scenarios has been performed without operation of emergency cooling systems. The results of the analyses indicate that a loss of forced coolant flow through the first wall rapidly causes dryout in the first wall cooling pipes. Following dryout, melting in the first wall starts within about 130 s in case of ongoing plasma burning. In case of large break LOCAs and ongoing plasma burning, melting in the first wall starts about 90 s after accident initiation.     

Analysis of the reference accidental sequence for safety assessment of LLCB TBM system

Safety analysis of the reference accidental sequence has been carried out for Lead Lithium cooled Ceramic Breeder (LLCB) Test Blanket Module (TBM) system; India's prototype of DEMO blanket concept for testing in International Thermonuclear Experimental Reactor (ITER). The accidental event analyzed starts with a Postulated Initiating Event (PIE) of ex-vessel loss of first wall helium coolant due to guillotine rupture of coolant pipe with simultaneous assumed failure of plasma shutdown system. Three different variants of the sequences analyzed include simultaneous additional failures of TBM and ITER first wall, failure of TBM box resulting in to spilling of lead lithium liquid metal in to vacuum vessel and reactor trip on Loss of Coolant Accident (LOCA) signal from TBM system. The analysis address specific reactor safety concerns, such as pressurization of confinement buildings, vacuum vessel pressurization, release of activated products and tritium during these accidental events and hydrogen production from chemical reactions between lead–lithium liquid metal and beryllium with water. An in-house customized computer code is developed and through these deterministic safety analyses the prescribed safety limits are shown to be well within limits for Indian LLCB-TBM design and it also meets overall safety goal for ITER. This paper reports transient analysis results of the safety assessment.     

LOCA analysis for Korean helium cooled solid breeder TBM

One of the major ITER goals is test blanket module (TBM) program which is for the demonstration of the breeding capability that would lead to tritium self-sufficiency in a reactor and the extraction of high-grade heat suitable for electricity generation under the ITER fusion environment. While the engineering design of Korean helium cooled solid breeder (HCSB) TBM and its ancillary systems has been performed, a safety assessment on different possible accident scenarios should be carried out for the purpose of licensing. In this paper, accident analyses for several loss of coolant accident (LOCA) cases were performed in order to assess safety aspects of the TBM design using RELAP5/MOD3.2. Since the TBM forms a loop with helium cooling system (HCS) which is one of ancillary systems required for removing heat deposited in the TBM by neutron wall loading and surface heat flux from plasma, it is necessary to model the complete loop for accident analysis. In this study, the helium passage including the TBM and HCS was nodalized for each accident scenario. The TBM and HCS components were modeled as the associated heat structures provided by RELAP5 to include heat transfer across solid boundaries. Based on computational results it was found that current design of the TBM is robust from the safety point of view.     

Development of TSACO and application to Chinese HCCB TBM cooling system

In this study, a thermal-hydraulic and safety analysis code (TSACO) for helium cooling system has been developed using Fortran 90 language, and the simulation has been performed for the cooling system of the Chinese helium cooled ceramic breeder test blanket module (CH HCCB TBM). The semi-implicit finite difference technique was adopted for the solution of the dynamic behavior of helium cooling system. Furthermore, a detailed illustration of the numerical solution for heat structures and critical model was presented. The code was verified by the comparison of RELAP5 code with the same initial condition, boundary condition, heat transfer and flow friction models. The TBM inlet/outlet temperatures and pressure drop were obtained and the results simulated by TSACO were shown in good agreement with those by RELAP5. Thereafter, the design basis accident in-vessel loss of coolant accident (LOCA), was investigated for the CH HCCB TBM cooling system. The critical flow model was also verified by comparing with RELAP5 code. The results indicated that the TBM can be cooled down effectively. The vacuum vessel (VV) pressure and the mass of helium spilled into the VV maintained below the design limits with a large margin.     

Analyses for Passive Safety of Fusion Reactor during Ex-Vessel Loss of Coolant Accident

Passive safety of nuclear fusion reactors during ex-vessel Loss-of-Coolant Accidents (LOCAs) in the divertor cooling system has been investigated using a hybrid code, which can treat the interaction of the plasma and plasma facing components (PFCs). The code has been modified to include the impurity emission from PFCs with a diffusion model at the edge plasma. We assumed an ex-vessel LOCA of the divertor cooling system during the ignited operation in International Thermonuclear Experimental Reactor (ITER), in which a carbon-copper brazed divertor plate was employed in the Conceptual Design Activity (CDA). When a double-ended break occurs at the cold leg of the divertor cooling system, the impurity density in the main plasma becomes about twice within 2s after the LOCA due to radiation enhanced sublimation of graphite PFCs. The copper cooling tube of the divertor begins to melt at about 3s after the LOCA, even though the plasma is passively shut down a t about 4s due to the impurity accumulation. It is necessary to apply other PFC materials, which can shorten the time Period for passive shutdown, or an active shutdown system to keep the reactor structures intact for such rapid transient accident.     

Current status of accident analysis for Korean HCCR TBS

Korea has decided to test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in ITER and design of the TBM with its ancillary systems, i.e. Test Blanket System (TBS), is under progress. Since the TBM is operated at elevated temperature with high heat load, safety consideration is essential in design procedure. In this paper, preliminary accident analysis results for the current HCCR TBS design on selected scenarios are presented as an important part of safety assessments. To simulate transient thermo-hydraulic behavior, GAMMA-FR code which has been developed in Korea for fusion applications was used. The main cooling and tritium extraction circuit systems, as well as the TBM, were simulated and the main components in the TBS were modeled as the associated heat structures. The important accident scenarios were produced and summarized in the paper considering the HCCR TBS design and ITER conditions, which cover in-vessel Loss Of Coolant Accident (LOCA), in-box LOCA, ex-vessel LOCA, Loss Of Flow Accident (LOFA), Loss Of Heat Sink Accident (LOHSA) and purge pipe rupture case. The accident analysis based on the selected scenarios was performed and it was found that the current design of the HCCR TBS meets the thermo-hydraulic safety requirements.     

AP1000主给水管道断裂事故中PRHR系统冷却能力分析

使用机理性分析程序建立包括主冷却剂系统、专设安全设施及相关二回路管道的AP1000核电厂模型,对AP1000核电厂主给水管道断裂事故进程进行计算分析。着重分析了非能动余热排出（PRHR）系统在主给水管道断裂事故工况中的瞬态响应、热工水力行为及其冷却能力,并针对PRHR系统流道阻力特性的不确定性对冷却能力的影响进行分析。分析结果表明,在主给水管道断裂事故中,PRHR系统的热移出功率最终能够与堆芯的衰变功率相匹配,有能力带走衰变热,保证一回路系统最终处于安全停堆状态,不发生堆芯损伤,当PRHR系统阻力系数增加时,PRHR系统的流量和换热功率会降低,对PRHR系统冷却能力造成影响。     

Preliminary thermal-hydraulic and safety analysis of China DFLL-TBM system

China has proposed the dual-functional lithium-lead (DFLL) tritium breeding blanket concept for testing in ITER as a test blanket module (TBM), to demonstrate the technologies of tritium self-sufficiency, high-grade heat extraction and efficient electricity production which are needed for DEMO and fusion power plant. Safety assessment of the TBM and its auxiliary system should be conducted to deal with ITER safety issues directly caused by the TBM system failure during the design process. In this work, three potential initial events (PIEs) – in-vessel loss of helium (He) coolant and ex-vessel loss of He coolant and loss of flow without scram (LOFWS) – were analyzed for the TBM system with a modified version of the RELAP5/MOD3 code containing liquid lithium-lead eutectic (LiPb). The code also comprised an empirical expression for MHD pressure drop relevant to three-dimensional (3D) effect, the Lubarsky–Kaufman convective heat transfer correlation for LiPb flow and the Gnielinski convective heat transfer correlation for He flow. Since both LiPb and He serve as TBM coolants, the LiPb and He ancillary cooling systems were modeled to investigate the thermal-hydraulic characteristic of the TBM system and its influence on ITER safety under those accident conditions. The TBM components and the coolants flow within the TBM were simulated with one-dimensional heat structures and their associated hydrodynamic components. ITER enclosures including vacuum vessel (VV), port cell and TCWS vault were also covered in the model for accident analyses. Through this best estimate approach, the calculation indicated that the current design of DFLL-TBM and its auxiliary system meets the thermal-hydraulic and safety requirements from ITER.