Flow Distribution Analysis of Water-Cooled Solid Breeder Test Blanket Module

Blanket system is one of the most important systems in a fusion reactor, which plays an important role in heat removing, radiation shielding and tritium breeding. Water-cooled ceramic breeder (WCCB) blanket module (BM) is one of tritium breeding blanket module concepts for Chinese Fusion Engineering Testing Reactor. According to the preliminary design of WCCB BM, there are complicated cascade and parallel cooling channels, in which maybe exists the nonuniform distribution of flow rate, and resulting adverse effect on heat transfer and safety. In this paper, the whole model of one BM is built by thermal hydraulic analytical code named Relap 5 and the flow distribution issue of water-cooled solid breeder (WCSB) test blanket module (TBM) is analyzed. The systematic analysis results show that the flow rate differences of most parts of the WCSB TBM are less than 4 % in a steady state. Start-up, operational transient and loss of flow accident are also studied, and flow instability in these transient cases is found and needs for further analysis. Three dimensional local model of First Wall is also built by CFX, to investigate flow characteristics at partial WCSB TBM, which shows that flow distribution calculated by CFX is consistent with the results from the thermal hydraulic analytical code. Both of the results of the steady state and transient analysis show that the thermal hydraulic analytical code is appropriate in analyzing the flow distribution and transient issue from the systematic view.     

Thermo-mechanical analysis of a DEMO divertor cooling finger under the EFREMOV test conditions

A divertor component of the forthcoming DEMO fusion reactor should be able to withstand heat flux loads larger than 10 MW/m². Successful design should withstand high flux loads for a number of load cycles since initially the DEMO reactor is expected to operate in a non-steady-state mode. Computations for evaluating the structural response of the divertor published so far have, however, been based on the stationary approach. A combined computational fluid dynamics and structural model for evaluating the structural response of a divertor under the non-stationary load conditions is therefore developed in this work. Heat transfer coefficients between the helium and inner surface of the thimble are calculated first by solving the helium steady-state flow equations. Spatially distributed heat transfer coefficients are then used as a boundary condition in a non-stationary thermo-mechanical analysis of the divertor. This analysis is performed for a number of load cycles under different surface heat flux levels. The model is validated against the EFREMOV test experimental conditions, designed to be close to reactor operation conditions. Good agreement of the highest temperatures on the tile’s top surface with the experimental data is obtained. The results suggest that there are three critical regions in the design where damage could initialize: (a) the thimble’s inner surface with the highest thermal gradients, (b) the tile’s outer surface and (c) the filler layer of the brazed tile-thimble joint where the temperature is higher than permissible. Post-examination data of experimental specimen confirm these conclusions as cracks were observed in the above mentioned areas (a) and (b), while melting of the layer (c) was also observed.     

Neutronics Analysis of Water-Cooled Ceramic Breeder Blanket for CFETR

In order to investigate the nuclear response to the water-cooled ceramic breeder blanket models for CFETR, a detailed 3D neutronics model with 22.5otorus sector was developed based on the integrated geometry of CFETR, including heterogeneous WCCB blanket models,shield, divertor, vacuum vessel, toroidal and poloidal magnets, and ports. Using the Monte Carlo N-Particle Transport Code MCNP5 and IAEA Fusion Evaluated Nuclear Data Library FENDL2.1,the neutronics analyses were performed. The neutron wall loading, tritium breeding ratio, the nuclear heating, neutron-induced atomic displacement damage, and gas production were determined.The results indicate that the global TBR of no less than 1.2 will be a big challenge for the watercooled ceramic breeder blanket for CFETR.     

DEMO水冷包层第一壁结构优化设计研究

针对聚变示范堆（DEMO）水冷包层,通过计算流体力学程序CFX和计算结构力学程序ANSYSWorkbench中的SIMULATION模块进行单向流固耦合分析。在对现有设计的DEMO水冷包层第一壁温度和应力数值模拟分析的基础上,改变了第一壁流道结构,着重研究了不同流道结构下的温度和应力分布,分析了几何结构对最高温度和最大应力的影响,提出第一壁结构的优化设计方案。数值模拟结果表明,优化设计方案能有效降低第一壁结构中的最高温度和最大应力。     

Preliminary Analysis of Liquid Metal MHD Pressure Drop in the Blanket for the FDS

Preliminary analysis and calculation of liquid metal Li17Pb83 magnetogydrodynamic (MHD) pressure drop in the blanket for the FDS have been presented to evaluate the significance of MHD effects on the thermal-hydraulic design of the blanket.To decrease the liquid metal MHD pressure drop,Al2O3 is applied as an electronically insulated coating onto the inner surface of the ducts.The requirement for the insulated coating to reduce the additional leakage pressure drop caused by coating imperfections has been analyzed.Finally,the total liquid metal MHD pressure drop and magnetic pump power in the FDS blanket have been given.     

Estimation of TBR on the Gap Between Neighboring Blanket Modules in the DEMO Reactor

When one wants to simply estimate tritium breeding ratio (TBR), the TBR may be reduced from a local TBR for the breeding zones of a blanket module by multiplying the breeder coverage (= the surface area of effective breeding region / the surface area of the first wall around plasma). When blanket modules are arranged, the gap between neighboring modules and the frames of the modules are regarded as non-breeding zones. On the other hand, neutrons scattered in the non-breeding zones can enter breeding zones, contributing to tritium production. This means that the estimation method mentioned above tends to underestimate TBR. In order to assess the scattering effect quantitatively, we carried out a three-dimensional Monte Carlo N-particle transport MCNP-5 calculation. It was found from the calculation that there is little decrease in TBR for gaps less than 4 cm when the blanket thickness is 70 cm. The result indicates that such a wide allowance of the gap will facilitate access of remote handling equipment for the replacement of blanket modules and improve access of diagnostics.     

He冷却试验包层模块的热-力耦合分析

试验包层模块(TBM)是国际热核聚变实验堆(ITER)的关键核心组件,其设计涉及多学科综合优化分析.本文介绍了He冷却固态增殖试验包层的设计概念,并应用热功耦合模拟方法对所提出的包层概念模型的热力响应特性进行分析.结果表明,包层内部各区域的最大温度值和最大应力值均未超过材料容许的限值,所提出的包层设计概念在正常运行工况下是安全可靠的.     

Nuclear Analysis of an ITER Blanket Module

ITER blanket system is the reactor’s plasma-facing component, it is mainly devoted to provide the thermal and nuclear shielding of the Vacuum Vessel and external ITER components, being intended also to act as plasma limiter. It consists of 440 individual modules which are located in the inboard, upper and outboard regions of the reactor. In this paper attention has been focused on to a single outboard blanket module located in the equatorial zone, whose nuclear response under irradiation has been investigated following a numerical approach based on the Monte Carlo method and adopting the MCNP5 code. The main features of this blanket module nuclear behaviour have been determined, paying particular attention to energy and spatial distribution of the neutron flux and deposited nuclear power together with the spatial distribution of its volumetric density. Moreover, the neutronic damage of the structural material has also been investigated through the evaluation of displacement per atom and helium and hydrogen production rates. Finally, an activation analysis has been performed with FISPACT inventory code using, as input, the evaluated neutron spectrum to assess the module specific activity and contact dose rate after irradiation under a specific operating scenario.     

Thermal Hydraulic Design and Analysis of a Water-Cooled Ceramic Breeder Blanket with Superheated Steam for CFETR

The water-cooled ceramic breeder blanket (WCCB) is one of the blanket candidates for China fusion engineering test reactor (CFETR). In order to improve power generation efficiency and tritium breeding ratio, WCCB with superheated steam is under development. The thermal-hydraulic design is the key to achieve the purpose of safe heat removal and efficient power generation under normal and partial loading operation conditions. In this paper, the coolant flow scheme was designed and one self-developed analytical program was developed, based on a theoretical heat transfer model and empirical correlations. Employing this program, the design and analysis of related thermal-hydraulic parameters were performed under different fusion power conditions. The results indicated that the superheated steam water-cooled blanket is feasible.     

Non-linear failure analysis of HCPB Test Blanket Module

For the European Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) the reduced activation ferritic martensitic (RAFM) steel EUROFER 97 is selected as a structural material. During operation the TBM will be subjected to complex thermo-mechanical loadings which yield at certain positions of the structure to stresses beyond the design limits of the structural material. Preliminary structural analyses of the TBM have shown critical behavior in several key points of the structure. An improved design has been proposed and in order to identify and assess the problematic positions in the improved version of the TBM a non-linear failure analysis is performed, for which a coupled deformation damage model developed at KIT for RAFM steels and recently implemented in the finite element code ABAQUS is used. The thermal loads in the form of non-homogeneous temperature fields distributions are obtained from a thermal analysis performed using the finite element code ANSYS on the same structure. Importing the temperature fields into the finite element code ABAQUS and applying the remaining loads – coolant internal pressure and structural boundary conditions – non-linear simulations are conducted taking into account the ITER-typical cyclic nature of the loading. The simulation results are evaluated and discussed considering ratcheting and damage at most critical highly loaded areas of the structure.     

Assessment of DEMO challenges in technology and physics

The challenges that DEMO designs encounter in both technology and physics are reviewed. It is shown that it is very important to respect the interlinks between these fields when developing designs for DEMO. Examples for areas where such interlinks put very strict requirements are the development of a steady state tokamak operation scenario and the question of power exhaust taking into account the boundary conditions set by materials questions. Concerning steady state operation, we find that demands on the physics scenario are so high that pulsed operation of a tokamak DEMO should seriously be considered in conservative DEMO designs. Alternatively, the device could foresee a large fraction of externally driven current which calls for optimization of both plasma CD efficiency as well as wall plug efficiency of the CD system. In the exhaust area, a realistic estimate of the admissable time averaged peak heat flux at the target is of the order of 5 MW/m², leading to strict requirements for the operational scenario, which has to rely on an unprecedented high level of radiation loss by impurity seeding and the facilitation of partial detachment. Thus, exhaust scenarios along these lines have to be developed which are compatible with the confinement needs and the H-L back transition power for DEMO. In both areas, we discuss possible risk mitigation strategies based on conceptually different approaches.     

The Temperature Control Mechanism of a Breeding Blanket Module for Fusion Reactor

A temperature control mechanism is presented for the nuclear material design of a breeding blanket module. It aims at exploring the acceptable material fractions maintaining the thermal balance inside a blanket module, which support the temperature requirement for tritium release and inventory. One-dimensional neutronics model is utilized to discuss the calculations. It is found that there is a mutual influence in the blanket interior by using the multi-layers structure. Firstly, the pure Li₄SiO₄ zones in the front area are the relative independent zones and have little thermal influence to other zones. The Beryllium zones only affect the adjacent zones, but the effect can be ignored. The second beryllium zone and the first mixed pebble zones are mainly limited to the reduced activation ferritic/martensitic material due to the lower temperature limitation. It indicates that the structural casing for nuclear materials should be reckoned in detail because of no cooling with it. It is confirmed that there is a temperature control methodology for the design of blanket nuclear materials using the linear variation of the temperature profile.     

Safety Analysis of ITER Test Blanket Module for Water Cooled Blanket with Pebble Bed Breeder

Performance test of test blanket modules in the fusion environment using the International Thermonuclear Experimental Reactor (ITER) is one of the most important mile-stone for the development of the breeding blanket of the fusion power plant. In the design of test blanket modules in the ITER, it is very important to show that test modules do not cause additional safety concern to the ITER. This work has been performed for the evaluation of the preliminary safety of the test blanket module of a water cooled solid blanket, which is the primary candidate of the breeding blanket in Japan currently. Major issues of the evaluation were, establishment of post-accident cooling in the test blanket module, hydrogen gas generation by Be/steam reaction, and pressure increase and spilled water amount by the event of coolant leakage. The analyses results showed that, suppression tank system is necessary to accommodate the over-pressure by the coolant water after pipe break in the box of the test module. Coolant water pipe break of the first wall of the test blanket module will result relatively small impact to the ITER safety because of the small inventory of the coolant water of the test module and large volume of the vacuum vessel of the ITER. However, it was clarified that the water cooled blanket with beryllium pebble as the multiplier will have the potential hazard of the hydrogen formation. Further investigation to maintain the safety on this aspect is required.     

Preliminary Thermo-mechanical Analysis of Water Cooled Ceramic Breeder Blanket for China Fusion Engineering Test Reactor

Bridging from ITER to DEMO, China Fusion Engineering Test Reactor aims at tritium self-sufficiency which is one of the main functions of blanket. The structure and thermo-mechanical performance influences strongly the operation and tritium breeding of blanket. In this paper, Water Cooled Ceramic Breeder blanket was designed with multilayer mixed pebble beds. And preliminary thermo-mechanical analysis has been done by the coupling of ANSYS finite element (FE) model and self-developed finite difference (FD) code under normal steady state condition. The results showed that the temperature distribution of the FE model corresponds well to that of the FD code. The obtained equivalent stress of the blanket is presented and critically verified the compliance with the SDC-IC code as reference criteria. At last, possible improvements such as adding fillets and plug-in materials are proposed to ameliorate the structure.     

Analysis and Optimization of Cooling Channels in ITER Blanket Module

Thermal hydraulic analysis was performed for the 2004 ITER blanket design, which was based on the fabrication methods of forging, drilling and welding. Results show that, in the original design, the flow distribution in radial holes was far from uniform because different flow drivers were used in the radial holes. Improvement of the flow driver is proposed in this study. With the optimization of the flow drivers, it is clearly shown that the total pressure drop decreases and a better velocity distribution in radial holes is obtained. The more uniform heat transfer coefficient among the radial holes is beneficial for reducing thermal stress.     

Welding techniques development of CLAM steel for Test Blanket Module

Fabrication techniques for Test Blanket Module (TBM) with CLAM are being under development. Effect of surface preparation on the HIP diffusion bonding joints was studied and good joints with Charpy impact absorbed energy close to that of base metal have been obtained. The mechanical properties test showed that effect of HIP process on the mechanical properties of base metal was little. Uniaxial diffusion bonding experiments were carried out to study the effect of temperature on microstructure and mechanical properties. And preliminary experiments on Electron Beam Welding (EBW), Tungsten Inert Gas (TIG) Welding and Laser Beam Welding (LBW) were performed to find proper welding techniques to assemble the TBM. In addition, the thermal processes assessed with a Gleeble thermal-mechanical machine were carried out as well to assist the fusion welding research.     

Development of the Water Cooled Ceramic Breeder Test Blanket Module in Japan

The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and evaluation toward DEMO blanket, the module fabrication technology development by a candidate structural material, reduced activation martensitic/ferritic steel, F82H, is one of the most critical items from the viewpoint of realization of TBM testing in ITER. In Japan, fabrication of a real scale first wall, side walls, a breeder pebble bed box and assembling of the first wall and side walls have succeeded. Recently, the real scale partial mockup of the back wall was fabricated. The fabrication procedure of the back wall, whose thickness is up to 90 mm, was confirmed toward the fabrication of the real scale back wall by F82H. Important key technologies are almost clarified for the fabrication of the real scale TBM module mockup. From the view point of testing and evaluation, development of the technology of the blanket tritium recovery, development of advanced breeder and multiplier pebbles and the development of the blanket neutronics measurement technology are also performed. Also, tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been started as the verification test of tritium production performance. This paper overviews the recent achievements of the development of the WCCB TBM in Japan.     

Conceptual design of Tritium Extraction System for the European HCPB Test Blanket Module

The HCPB (Helium Cooled Pebble Bed) Test Blanket Module (TBM), developed in EU to be tested in ITER, adopts a ceramic containing lithium as breeder material, beryllium as neutron multiplier and helium at 80 bar as primary coolant.In HCPB-TBM the main function of Tritium Extraction System (TES) is to extract tritium from the breeder by gas purging, to remove it from the purge gas and to route it to the ITER Tritium Plant for the final tritium processing.In this paper, starting from a revision of the so far reference process considered for HCPB-TES and considering a new modeling activity aimed to evaluate tritium concentration in purge gas, an updated conceptual design of TES is reported.     

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.     

基于失效物理模型的聚变堆包层管道可靠性数据修正方法研究

由于聚变堆部件特殊的运行环境,部件可靠性数据极度匮乏,通常采用环境因子方法对现有可靠性数据进行修正,但现有可靠性数据修正模型未考虑解决高温模型不能适用的问题。本文提出了高温环境下的失效物理模型修正优化方法,提升了高温失效物理模型在极端环境下的适用范围和部件服役寿命修正精度,并基于失效物理模型修正方法开展了ITER中国氦冷固态包层氚提取系统(TES)管道可靠性数据修正研究,为TES系统可靠性分析提供了数据支持。