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.     

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.     

基于RELAP5的中国氦冷固态包层真空室外破口瞬态特性分析

利用RELAP5/MOD3.4对中国氦冷固态包层、氦气冷却剂回路和二次侧水冷系统进行建模和系统热工水力安全评价。依据ITER事故分析制定的事故序列,对设计基准真空室外破口进行了瞬态分析,并对比了不同破口位置、面积和停堆方式对第一壁的影响。结果表明：真空室外破口发生在风机的下游较上游危险,且小破口较大破口更危险;若真空室外破口同时包层第一壁破口,也可通过自然循环和辐射换热带走衰变热冷却包层;真空室外破口事故中采用聚变停堆系统的3s停堆方式,可避免第一壁熔化。     

基于RELAP5的CFETR氦冷陶瓷增殖包层模块热工安全分析

本文对中国聚变工程实验堆(CFETR)氦冷陶瓷增殖(HCCB)包层进行热工安全分析。采用大型反应堆瞬态分析程序RELAP5对HCCB包层建模,并进行稳态分析和假设事故的模拟。计算结果表明,CFETR HCCB包层在真空室内氦气泄漏和增殖区盒内氦气泄漏事故中均未出现结构材料熔化,同时各部分的压强变化情况均未超出设计阈值,包层系统在事故发生后均能有效快速地排出余热。CFETR HCCB包层的设计满足热工安全方面的要求。     

基于RELAP5的双功能液态锂铅实验包层模块安全分析

利用嵌入了液态锂铅（LiPb）的热工水力子模块的系统程序RELAP5/MOD3,对双功能液态锂铅（DFLL）实验包层模块（TBM）的安全特性进行评价。对DFLL-TBM及其辅助冷却系统的稳态运行工况、预期运行事件和相关事故工况进行了建模、计算和分析。计算结果表明,稳态运行时第一壁（FW）结构材料表面最高温度低于允许值550 ℃。事故工况下氦气泄漏引起的ITER真空室（VV）、窗口设备室（port cell）以及托卡马克冷却水系统大厅拱顶（TCWS vault）的增压均低于ITER要求的限值0.2 MPa。实验包层钢结构不会熔化且可通过辐射换热有效地导出衰变余热。DFLL-TBM的设计可满足ITER对其热工水力安全方面的要求。     

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.     

Neutronic Calculation Analysis for CN HCCB TBM-Set

Using the Monte Carlo transport code MCNP.neutronic calculation analysis for China helium cooled ceramic breeder test blanket module(CN HCCB TBM) and the associated shield block(together called TBM-set) has been carried out based on the latest design of HCCB TBM-set and C-lite model.Key nuclear responses of HCCB TBM-set.such as the neutron flux,tritium production rate,nuclear heating and radiation damage,have been obtained and discussed.These nuclear performance data can be used as the basic input data for other analyses of HCCB TBM-set,such as thermal-hydraulics,thermal-mechanics and safety analysis.     

Steady State and Transient Analysis of Novel Design Helium Cooled Ceramic Blanket (HCCB) System of China Fusion Engineering Test Reactor (CFETR)

An upgraded form of China fusion engineering test reactor (CFETR) was investigated for the safety performance. In the current study, modification of the designs were presented with relative tolerance. The steady state were calculated for the new design using Relap5 code. Two accidents were simulated i.e., in-vessel and In-box loss of coolant accident. These accidents were simulated in helium cooled ceramic blanket (HCCB) system for the purpose to investigate the safety measures of the CFETR. It is utmost important to ensure the safety performance of the reactor. In this research, sudden break at blanket system was assumed and calculated different parameters including temperature, pressure and coolant fluxes to observe the differences in pattern during the accident under limited time domain. The research is very important because the design of HCCB is new and there is a need to conduct steady state and transient state of the reactor in order to make sure and authenticate the design and to safer the reactor.     

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 and optimization of the manifolds of the HCCB blanket for CFETR

Thermal-hydraulic performance is a challenging issue in helium-cooled ceramic breeder (HCCB) blanket design due to the extremely complicated working environment and the strict limits of materials temperature. The heat loads deposited on the HCCB blanket comprises of severe surface heat flux from plasma and the volumetric nuclear heat from neutron irradiation, which can be exhausted by the built-in cooling channels of the components. High pressure helium with 8 MPa, distributed from the coolant manifolds is employed as coolant in the blanket. The design and optimization of the manifolds configuration was performed to guarantee the accurate flow control of helium coolant. The flow distribution in the coolant manifolds was investigated based on the structural improvement of manifolds aiming at overall uniform mass flow rates and better flow streamline distribution without obvious vortexes. The peak temperature of different functional materials in the blanket under normal operating condition is below allowable material limits. It is found that the components in the current blanket module could be cooled effectively under the intense thermal loads due to the updated design and optimization analysis of manifolds.     

基于SCDAP/RELAP5耦合堆腔注水的非能动压水堆熔融池冷却分析

采取堆腔注水策略冷却熔融池对缓解严重事故后果、降低安全壳的失效概率具有十分重要的作用。本文采用SCDAP/RELAP5程序,首先以韩国APR1400相关实验结果对堆腔外部注水自然对流冷却能力进行比对分析,然后建立了耦合堆腔注水措施的融熔池冷却的核电厂模型,以非能动压水堆为研究对象,针对冷段大破口失水事故(LBLOCA)始发严重事故序列,分析堆芯熔融进展过程中实施堆腔注水策略后融熔池的冷却特性及堆腔外部注水的自然循环能力。分析结果表明,LBLOCA下,当堆芯出口温度达到923K时,实施堆腔注水后能有效冷却下封头内的熔融池,从而保持压力容器的完整性。     

基于RELAP5的氦氙流动换热计算模块开发与验证

反应堆热工系统分析程序是开展热工水力计算与安全评价的重要工具。为开发适用于氦氙气冷空间堆的热工系统分析程序,本文在RELAP5/MOD40程序中拓展了氦氙混合气体（He Xe）物性计算模块,添加了适用于He Xe的传热关系式,将拓展后程序计算值与实验值进行对比。结果表明：程序默认的Sutherlands定律用于He Xe物性计算时将引入较大误差;Dittus Bolter公式对He Xe对流换热时的Nu预测偏高,将导致不保守的壁温计算结果。拓展后的程序对He Xe压降和换热计算结果均与实验值吻合较好,验证了程序开发的正确性以及程序用于He Xe流动换热计算的功能。本研究可为系统层面程序开发奠定基础。     

Safety analysis of multiple-failure of passive safety systems in SBWR-1200 SBLOCA

The design of the simplified boiling water reactor (SBWR-1200) is characterized by utilizing fully passive safety systems. The emergency core cooling is realized by the gravity driven core cooling system, and the decay heat removal is done by the passive containment cooling system and isolation condenser system. All of the systems have multiple units and could be partially failed. The objective of this paper is to analyze the system response under the multiple malfunctions of passive safety systems in the SBWR-1200.

The chosen accident scenario is a small break loss of coolant accident with one of three gravity driven core cooling system drain lines blocked and one of three passive containment cooling system condensers disabled. An integral test has been carried out in the PUMA facility for 16 h. The facility is designed for low pressure, long term cooling operation with the multiple safety related components; therefore, it has the flexibility to demonstrate the asymmetric or multiple-failure effects with the combination of disability of safety systems. The test initial conditions at 1 MPa (150 psi) are obtained from RELAP5/MOD3.2 code simulation for the SBWR-1200 with appropriate scaling considerations.

Comparisons have been first made between the multiple-failure test and a single-failure test preformed previously. It shows that the core has been covered with liquid coolant during all of accident transient even though there is an apparent coolant inventory reduction in the multiple-failure test. The decay heat removal has no significant difference because the remaining two passive containment cooling condensers increase their cooling capacities, and even the drywell pressure is slightly lower due to the cold water injection from the suppression pool. Comparisons have also been made between the scaled-up test data and the code simulation at the prototypic level. The prototypic simulation is done by RELAP5/MOD3.2. Agreements between the code simulation and the scaled-up test data confirm the code applicability and the facility scalability for this accident scenario.     

针对REPEC加热实验的RELAP5程序模拟与分析

为研究压力容器外部流道的冷却能力及流动传热过程,在反应堆压力容器外部冷却(REPEC, Reactor Pressure vessel External Cooling)实验台架前期加热实验的基础上,采用RELAP5程序对实验工况进行模拟和对比。模拟结果与实验数据一致性较好。随加热热流、进出口面积的增加,系统内自然循环流量也增加;入口欠热度对自然循环流量的影响不是很明显;近饱和沸腾条件下,系统出现明显的两相不稳定流动。     

Thermal-Hydraulic System Study of the Helium Cooled Pebble Bed （HCPB） Test Blanket Module （TBM） for ITER Using System Code RELAP5

The HCPB concept has been a European DEMO reference concept for nearly one decade. Detailed thermal-hydraulic study on the control behavior of the whole system is one of the important parts of this development. The thermal-hydraulic effect of the TBM-combined cooling circuit during a cyclic operation in ITER has been studied using the system code RELAP5. The RELAP5 is based on an one-dimensional, transient two-fluid model for the flow of a two-phase steam-water mixture that can contain noncondensable components like Helium. The RELAP5-models are modified to take the cyclic operation of the circulator, heat exchanger, bypass, valves etc in to account. A sequence of operational phases is investigated, starting from the cold state through the heating phase that brings the system to a stand-by condition, followed by typical power cycles applied in ITER. The results show that the implemented control mechanisms keep the inlet temperature to the TBM and the total mass flow rate at the required values through all phases.     

小型压水堆严重事故下一回路承压管道蠕变预测分析模型开发

一回路承压管道蠕变是压水堆核电厂严重事故重要现象之一。针对小型压水堆,本文基于SCDAP/RELAP5程序开发了严重事故分析模型,利用实验拟合方法得到了一回路主管道(SA321)、自然循环式蒸汽发生器传热管(00Cr25Ni35Al Ti)两种材料蠕变预测分析模型,改进了SCDAP/RELAP5程序蠕变预测分析功能模块,并通过假想事故序列验证了SA321、00Cr25Ni35Al Ti蠕变预测分析模型的合理性。为后续开展小型压水堆严重事故下一回路承压管道蠕变规律研究提供基础参考。     

基于RELAP5与MELCOR联合分析方法的压水堆严重事故研究

针对严重事故的模拟研究,本文提出结合热工水力系统程序和严重事故一体化程序的分析方法,以典型三环路传统压水堆为对象,分别采用RELAP5和MELCOR程序建立模型,分析在全厂断电叠加汽动辅助给水泵失效事故下系统的瞬态响应。为了尽可能地利用RELAP5计算早期热工水力响应,同时保证严重事故计算结果的准确性,以MELCOR锆合金氧化模型开始工作温度的下限,即包壳温度达到1 100 K作为程序衔接准则并利用RELAP5的大编辑功能,提取所需计算结果导入MELCOR输入卡作为初始参数继续模拟。计算结果表明,数据连接过程整体保持了连续性,两种方法计算得出的主冷却剂系统压力、堆芯和稳压器水位、燃料包壳温度等参数的数值以及堆芯传热恶化和压力容器失效等现象的时序存在不同程度的差异,例如堆芯熔毁时间延后了约538 s。由于采用了RELAP5计算严重事故前的系统暂态响应,联合分析方法的计算结果比单独使用MELCOR分析的结果更加准确,该方法可以提高传统严重事故分析的可靠性。     

基于RELAP5与MELCOR联合分析方法的压水堆严重事故研究

针对严重事故的模拟研究,本文提出结合热工水力系统程序和严重事故一体化程序的分析方法,以典型三环路传统压水堆为对象,分别采用RELAP5和MELCOR程序建立模型,分析在全厂断电叠加汽动辅助给水泵失效事故下系统的瞬态响应。为了尽可能地利用RELAP5计算早期热工水力响应,同时保证严重事故计算结果的准确性,以MELCOR锆合金氧化模型开始工作温度的下限,即包壳温度达到1 100 K作为程序衔接准则并利用RELAP5的大编辑功能,提取所需计算结果导入MELCOR输入卡作为初始参数继续模拟。计算结果表明,数据连接过程整体保持了连续性,两种方法计算得出的主冷却剂系统压力、堆芯和稳压器水位、燃料包壳温度等参数的数值以及堆芯传热恶化和压力容器失效等现象的时序存在不同程度的差异,例如堆芯熔毁时间延后了约538 s。由于采用了RELAP5计算严重事故前的系统暂态响应,联合分析方法的计算结果比单独使用MELCOR分析的结果更加准确,该方法可以提高传统严重事故分析的可靠性。     

RELAP5/MOD3.3程序对非能动核电厂小破口失水事故的适用性研究

AP1000核电厂采用非能动堆芯冷却系统来缓解小破口失水事故(SBLOCA),缓解事故的理念是流动冷却。RELAP5/MOD3.3程序适用于传统核电厂SBLOCA研究,对于非能动电厂SBLOCA研究的适用性需重新研究与评估。本工作基于非能动电厂小破口失水事故的分析,结合RELAP5/MOD3.3的结构与模型,对其进行评估和改进。为验证改进后的RELAP5/MOD3.3的适用性,以AP1000小破口失水事故的验证试验台架APEX-1000为模拟对象,分析模拟DBA-02、NRC-05事故工况。分析结果表明,改进后的RELAP5/MOD3.3的计算结果与试验数据符合较好。     

Status and progress of Indian LLCB test blanket systems for ITER

The lead–lithium ceramic breeder (LLCB) TBM and its auxiliary systems are being developed by India for testing in ITER machine. The LLCB TBM consists of lithium titanate as ceramic breeder (CB) material in the form of packed pebble beds. The FW structural material is ferritic martensitic steel cooled by high-pressure helium gas and lead–lithium eutectic (Pb–Li) flowing separately around the ceramic breeder pebble bed to extract the nuclear heat from the CB zones. Low-pressure helium is purged inside the CB zone for in situ extraction of bred tritium. Currently the LLCB blanket design optimization is under progress. The performance of tritium breeding and high-grade heat extraction is being evaluated by neutronic analysis and thermal–hydraulic calculations for different LLCB cooling configurations and geometrical design variants. The LLCB TBM auxiliary systems such as, helium cooling system (HCS), lead–lithium cooling system (LLCS), tritium extraction system (TES) process design are under progress. Safety analysis of the LLCB test blanket system (TBS) is under progress for the contribution to preliminary safety report of ITER-TBMs. This paper will present the status of the LLCB TBM design, process integration design (PID) of the auxiliary systems and preliminary safety analysis results.