In-situ observation of the effect of the precipitate/matrix interface on the evolution of dislocation structures in CLAM steel during irradiation

The key factor that affects the irradiation resistance of a material is its structure such as grain size and precipitates. Two types of China Low Activation Martensitic (CLAM) steels with a different number density of MC phase were pre-ion implanted and subsequently irradiated by electrons using ultra-high voltage electron microscope (HVEM). The effect of MC phase on the growth behavior of dislocation loops and the stability of pre-existing precipitates were investigated in situ and this may give some hints on the way to increase the ability against irradiation damage. The results show that a high number density of the fine MC phase improves the strength of the material and also helps to inhibit the fast growth of dislocation loops. The interface between the precipitate and the matrix acts as an effective sink to trap radiation induced point defects, which can possibly result in an improvement of irradiation resistance to some extent. However, the coarsening of precipitates because of radiation enhanced diffusion is another issue that needs to be seriously considered when developing a nuclear material.     

中国低活化马氏体钢CLAM热等静压扩散焊接初步研究

热等静压扩散焊接由于其独特的优点而被认为是聚变堆包层第一壁及冷却板等复杂结构件的首选制造技术之一.本文对中国低活化马氏体钢CLAM的热等静压焊接进行了初步研究.实验中CLAM钢自身焊接的拉伸性能已经达到母材的水平,但冲击断口为原接合面,且强度较低,X射线能谱(EDS)分析发现冲击断面上有强氧化物形成元素Al、Si、V等富集,初步认为焊接前期处理过程中待焊接面形成的稳定氧化膜导致冲击性能较低.     

CLAM钢异径管冷成形数值模拟及试验研究

研究高质量CLAM钢管件的成形性能对聚变堆管路系统制造具有重要意义.利用有限元方法对CLAM钢异径管冷成形过程进行了数值模拟.研究了成形过程中速度矢量的变化情况,并分析了异径管成形过程中的应力、应变分布及冲头推力变化;使用摩擦系数较小的表面涂层对管坯润滑处理后进行了冷成形.研究表明,在几何尺寸及厚度分布方面,实验结果和...     

Experimental investigation on the possible techniques of pebbles packing for the HCPB Test Blanket Module

The Helium Cooled Pebble Bed Test Blanket Module (TBM) features a structural box that consists of the first wall, two caps and a stiffening grid. Inside the stiffening grid the breeding units (BUs), consisting of the beryllium and lithium ceramic pebble beds and cooling plates, are accommodated. The BUs are closed by the BU back plates and several structural plates of the manifold system as well as the TBM back plate consequently the BUs may not be accessed directly after the assembly of the TBM box; however, access is possible through dedicated penetrations in the TBM caps. According to the current manufacturing strategy, the assembly of the TBM structural sub-components is based on several welding processes which require post-welding heat treatments (PWHT) at temperatures which exceed the temperature limit of the beryllium pebbles. For that reason the beryllium pebble beds will be packed after the TBM box is assembled and heat treated. The packing of the BUs will be performed using a small-diameter (5 mm) tube that will be inserted into some penetrations in the TBM caps. It is expected that the lithium ceramic pebbles can withstand the high temperatures of the PWHT (this assumption needs to be verified) therefore the current strategy is to pack the ceramic pebble beds during the TBM box assembly. This study experimentally demonstrates the packing procedures for the beryllium beds using a full-scale Plexiglas mock-up as well as the optimization of the packing process by dedicated measures such as vibrating and tilting of the mock-up. In addition the impacts of the experimental results on the TBM design are summarized and the paper is concluded by proposing a packing strategy that can be used to achieve a packing factor of 63.6%.     

Overview of Helium Cooled Ceramic Reflector Test Blanket Module development in Korea

Korea plans to install and test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in the ITER, because the HCCR blanket concept is one of options of the DEMO blanket. Currently, many design and R&D activities have been performed to develop the Korean HCCR TBM. An integrated design tool for a fusion breeder blanket has been developed based on nuclear technologies including a safety analysis for obtaining a license for testing in the ITER. A half-scale sub-module mockup of the first wall with the manifold was fabricated, and the manufacturability and thermo-hydraulic performances were evaluated. High heat load and helium cooling test facilities have been constructed. Next, the recent status of TBM material development in Korea was introduced including Reduced Activation Ferritic Martensitic (RAFM) steel, lithium ceramic pebbles and silicon carbide (SiC) coated graphite pebbles. Several fabrication methods of RAFM steel, lithium ceramic pebbles, and silicon carbide coating on graphite pebbles were investigated. Recent design and R&D progress on these areas are introduced here.     

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.     

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.     

Fabrication and high heat flux test of the first wall mock-ups for the Korean He Cooled Test Blanket (KO HCML TBM)

Korea has proposed and designed a Helium Cooled Molten Lithium (HCML) Test Blanket Module (TBM) to be tested in the International Thermonuclear Experimental Reactor (ITER). Ferrite Martensitic (FM) steel is designed to be used as a structural material in this design. Three mock-ups, especially for the first wall channels, were fabricated with a Hot Isostatic Pressing (HIP, 1050 °C, 100 MPa, 2 h) in order to develop the fabrication technology. One of them was used to observe the microstructure and it was found that there are no pores and cracks. Other mock-ups were used with high heat flux (HHF) tests performed with 20 cycles under 0.5 and 1.0 MW/m² heat fluxes in order to evaluate the integrity of the fabricated mock-ups. Here, HHF test conditions were determined with an ANSYS-CFX analysis. And then, the mock-ups were tested and broken under a 1.5 MW/m² heat flux, which is about the Critical Heat Flux (CHF) at the wall.     

聚变堆实验包层模块第一壁温度场及热结构分析

铍相对于众多聚变反应堆的第一壁护甲材料,有着许多优点,这些优点使它和钨及碳基材料一起被选作国际热核聚变实验堆（ITER）第一壁的候选防护材料。对中国氦冷固态增殖剂实验包层模块（CHHCSBTBM）第一壁进行多场耦合模拟分析结果表明,使用表面热负荷模拟分析时,未考虑中子负载情况下,模拟分析结果与其它结果有较大出入,故使用表面热负荷模拟分析时必须考虑中子负载情况。而对第一壁热结构分析表明,铍保护板的应力超过了其许用应力,可以寻找其它铍合金或第一壁护甲材料以满足第一壁护甲材料热结构应力要求。     

中国HCCB-TBM氚增殖单元三维多物理场耦合氚输运模拟分析

中国氦冷固态实验包层模块(HCCB-TBM)将在国际热核聚变实验堆(ITER)上安装测试,以验证其氚增殖能力与核热移出能力。HCCB-TBM中的氚输运与流体的传热和传质、氢同位素交换、结构材料的SORET效应密切相关。考虑以上物理因素,基于商业软件COMSOL完成了HCCB-TBM氚增殖单元多物理场耦合的氢同位素输运模拟分析。分析结果表明:球床吹洗气体中含氢有助于抑制氚渗透损失;当吹洗气体含氢浓度为4.66×10^-2 mol/m³时,产生约13.2倍的氚渗透阻止效应。     

Development of discrete ordinates code supporting unstructured tetrahedral mesh and applied in neutronics analysis for the Korea Helium Cooled Ceramic Reflector Test Blanket Module

The discrete ordinates code under development by KAERI uses an unstructured tetrahedral mesh, and thus it can be applied to solve the radiation transport in a complicated geometry. In addition, the geometry modeling process has become much easier because computational tetrahedral meshes are generated based on the CAD file by Gmsh.As our first phase of applying the code to a TBM neutronics analysis, the neutron flux distribution in the Korea HCCR TBM is compared with that of MCNPX, and visualized in a three-dimensional system domain. Visualization of the fluxes and associated reaction rates in the whole system with a single run is one of the merits of a deterministic method and is very useful for checking hot spots.     

Long-Term Integrity of Waste Package Final Closure for HLW Geological Disposal, (III)

Overpack, a high-level radioactive waste package for Japan's geological disposal program, is required to isolate the sealed vitrified waste from contact with groundwater for 1,000 years. In this study, EBW (Electron Beam Welding), a typical application of high energy density beams widely used in various industries, was examined for its applicability to sealing a carbon steel overpack lid with a thickness of 190 mm. Welding conditions and welding parameters were examined for single-pass welding in a narrow area for three different penetration depths. Weld joint tests including macro- and microstructure, and mechanical properties were conducted and assessed within tentatively applied criteria for weld joints. Measurement and numerical calculation for residual stress were also conducted and the tendency of residual stress distribution was discussed. These test results were compared with the basic requirements of the welding method for overpack, which were pointed out in our first report. The induced void and cold shut inside the weld joint and surface roughness were also discussed for their improvement and evaluation, which need to be established to assure the long-term integrity of overpack lid closure.     

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.