Shintarou Ishiyama

Mechanical properties of HIP bonded W and Cu-alloys joint for plasma facing components

Abstract Hot isostatic pressing (HIP) bonding technology of W (tungsten) and Cu-alloys have been developed to fabricate plasma facing components of the fusion reactor. As regards W and oxygen free high conductivity copper (OFHC-Cu), the highest bonding strength was achieved at the HIP condition of 1273 K ×2 h ×147 MPa. On the other hand, W and dispersion strengthened copper (DS-Cu) were not bonded directly because of tungsten oxide production at the bonding interface. In this study, HIP bonding tests on W and DS-Cu with OFHC-Cu disk and/or Au-foil were performed. Bonding tests with OFHC-Cu disk were successfully bonded and it is shown that thickness of OFHC-Cu disk over 1.0 mm may be needed and the tensile strength are a little higher than that of HIP treated OFHC-Cu. Bonding tests with Au-foil were also performed and successfully bonded. Au-foil lead to an improvement in bonding strength and a lowering of bonding temperature.

Mechanical properties and damage behavior of non-magnetic high manganese austenitic steels

Abstract Fe–Cr–Mn steels have been considered as materials of structural components for fusion reactor because of their low induced-radio-activity compared with SUS316 stainless steels. It has been expected to develop a non-magnetic steel with a high stability of the austenitic phase and a strong resistance to irradiation environments. For these objectives, a series of the Fe–Cr–Mn steels have been examined by tensile tests and simulation irradiation by electrons. The main alloying compositions of the steels developed are; C:0.02–0.2 wt%, Mn: 15 wt%, Cr: 15–16 wt%, N: 0.2 wt%. These steels were heat-treated at 1323 K for 1 h. The structure of the steels after the heat-treatment was austenite single phase. The yield stress of the steels was 350–450 MPa and the elongation were 55–60%. When the steels of high C and N was electron-irradiated at below 673 K, no voids were nucleated and only small dislocation loops were formed with high density. The austenite phase was also stable during irradiation below 673 K. Thus, newly developed high manganese steels have excellent mechanical proprieties and high irradiation resistance at relatively low temperature.

Effect of stress history on cumulative fatigue damage of fine-grained isotropic GCR graphite

Abstract To investigate the effect of stress history on cumulative fatigue damage of fine-grained isotropic nuclear graphite, two step multi-loading fatigue tests were performed by changing peak load from high to low loading (H-L mode) or low to high loading (L-H mode) using the push-pull loading pattern. A remarkable reduction of fatigue life due to fatigue damage given in the first fatigue step was found in the secondary fatigue step in the H-L and L-H mode. These results suggested that there was fatigue interaction between the stress history of the first and secondary step as observed in the fatigue test of metals. This study shows that these phenomena can be explained by the fatigue damage mechanism of graphite with the concept of the intermediate fatigue process between first and secondary fatigue step. The cumulative fatigue damage rule of the graphite applicable for lifetime evaluation in H-L and L-H fatigue processes are proposed in equations (3)–(7), where a , b , c and d are constants, and a and b are positive and negative, and c and d are negative and positive, respectively, in the H-L mode and vice versa in the L-H mode.

Performance of V-4Cr-4Ti alloy exposed to the JFT-2M tokamak environment

A long-term test has been conducted in the JFT-2M tokamak fusion device to determine the effects of environmental exposure on the mechanical and chemical behavior of a V-4Cr-4Ti alloy. Test specimens of the alloy were exposed in the outward lower divertor chamber of JFT-2M in a region away from direct contact with the plasma and were preheated to 300 C just prior to and during selected plasma discharges. During their nine-month residence time in JFT-2M, the specimens experienced approximately 200 lower single-null divertor shots at 300 C, during which high energy particle fluxes to the preheated test specimens were significant, and approximately 2,010 upper single-null divertor shots and non-diverter shots at room temperature, for which high energy particle fluxes to and expected particle retention in the test specimens were very low. Data from post-exposure tests have indicated that the performance of the V-4Cr-4Ti alloy would not be significantly affected by environmental exposure to gaseous species at partial pressures typical for tokamak operation. Deuterium retention in the exposed alloy was also low (<2 ppm). Absorption of interstitial by the alloy was limited to the very near surface, and neither the strength nor the Charpy impact properties of the alloy appeared to be significantly changed from the exposure to the JFT-2M tokamak environment.

Irradiation damage analysis on the flat plate type target plate of the divertor for fusion experimental reactors

Abstract To design the relevant plasma facing components of fusion experimental reactors such as ITER, irradiation damage analysis, especially on divertor structures exposed to high heat flux and heavy neutron irradiation, is one of the most important problems. This paper presents finite element analytical results of the thermal and irradiation induced stresses which occurred in the divertor structures which are exposed to neutron irradiation at 0–1 dpa with a high heat flux up to 15 MW/m 2 . A type of target plate model of the divertor structure studied in present study e.g. flat plate model has bonded structure of one-dimensional high thermal conductivity carbon-carbon composite (C/C) and oxygen-free high conductivity copper (OFHC), as armor and substrate/heat sink materials, respectively. These results show that irradiation induced stresses at edges of bonded interface between an armor and a substrate/heat sink, become higher with increase of dpa and reach up to the critical values of the materials at 0 and 1 dpa. This indicates that drop-off of armor tiles from substrate structure is one of very serious problems for the safety design of target plate; thus the reduction of service conditions and change of divertor materials are important to extend lifetime of the model.

Characterization of non-magnetic Mn–Cr steel as a low induced activation material for vacuum vessels

Abstract The JT-60SU (Super Upgrade) program is under discussion at Japan Atomic Energy Research Institute (JAERI). Its design optimization activity requires the vacuum vessel material to be non-magnetic, very strong and with low induced activation. However, there is no suitable material available to fulfill all the requirements. JAERI started to develop a new material for the vacuum vessel together with the Japan Steel Works (JSW). Chemical composition and metallurgical processes were optimized and a new steel named VC9, which has the composition of Cr – 16 wt%, Mn – 15.5 wt%, C – 0.2 wt%, and N – 0.2 wt% with non-magnetic single γ phase, was selected as a candidate material. Here, mechanical properties and weldability of VC9 were examined and the results were compared with those of type 316 or 316L stainless steel. It was shown that VC9 has good mechanical properties and weldability.

The characterization of Ti- and CaMH systems in the high temperature chemical heat pump for gas-cooled reactor applications

Ca-Ni, Ca-Mg, Ti-Cu and Ti-Cr alloy systems, which have high melting points and hopefully high hydrogen absorption capacities with a high heat of reactor, are prepared in the compositional range 5-79 wt.% secondary additives. Absorption of hydrogen by these systems has been studied in the testing temperature range from 300 to 1200°C. Ca-Ni compounds (10-78.5 wt.% Ni), which encompass the intermetallic compounds CaNi z and CaNi 3 in this range, are found to show a high hydrogen absorption, much more than 1.5 wt.% H 2 up to 600 °C. On the contrary, Ca-Mg compounds (37.8-64.5 wt.% Mg) reveal good absorption and dissociation behaviors at temperatures from 320 to 400 °C under a hydrogen pressure of 0.5-2 Mpa and clear dependence of the absorption capacity on the composition ratio. Ti-52.4wt.% CU, with Cud ternary addition (5 wt.%) to enhance absorption rate, shows a very interesting behavior at 530°C, i.e. secondary plateau at 0.02-0.1 MPa. Ti-5 wt.% Cr shows 3.5 wt.% H 2 absorption capacity at 300 and 500 °C at 0.1 MPa, but not such good dissociation is found at 300 °C. These data suggest that in Ca-Mg or Ca-Ni systems the hydrogen absorption capacity is determined by the amount of the secondary element; the crystal structure appears to be of importance, and Ti-Cr systems are found to be possible for high temperature use as the MH material of the heat pump mentioned above. Keywords: Hydride reaction; Ti- and Ca-MH systems; Chemical heat pump; Intermetallic compound; Gas cooled reactor; Ternary additives 1. Introduction to nuclear thermal energy systems, as shown in Fig. 1, and the research and development of these systems The Japan Atomic Energy Research Institute have continued in JAERI since 1992. (JAERI) has been constructing the High Temperature The present work reports the recent activities of the Engineering Test Reactor (HTTR) since 1991. The HTTR (30 MW) has bee, n designed to be operated at very high temperatures (maximum temperature,