Shinzo Saito

Corrosion studies in liquid Pb–Bi alloy at JAERI: R & D program and first experimental results

Abstract The program of corrosion study in liquid Pb–Bi at Japan Atomic Energy Research Institute (JAERI) is described. It was planned to clarify effects of parameters such as temperature, temperature difference between high-temperature and low-temperature parts, oxygen concentration in liquid Pb–Bi, flow rate, irradiation, stress and chemical composition of materials. The program contains the basic corrosion study in static Pb–Bi for elucidation of corrosion mechanism and effects of oxygen concentration in liquid Pb–Bi and ion irradiation on corrosion behavior. It also contains the corrosion study in flowing Pb–Bi using a corrosion testing loop. From the initial result of the static corrosion tests in oxygen saturated Pb–Bi at 550 ° C for 500 h, it was shown that the thickness of the corrosion film decreases with increasing Cr content in steels.

Present status of the High Temperature Engineering Test Reactor (HTTR)

Abstract In Japan, the research and development on the High Temperature Gas-cooled Reactors (HTGRs) had been carried out for more than fifteen years since 1969 as the multi-purpose Very High Temperature gas-cooled Reactor (VHTR) program for direct utilization of nuclear process heat such as nuclear steel making. Recently, reflecting the change of the social and energy situation and with less incentives for industries to introduce such in the near future, the JAERI changed the program to a more basic ‘HTTR program’ to establish and upgrade the HTGR technology basis. The HTTR is a test reactor with a thermal output of 30 MW and reactor outlet coolant temperature of 950°C, employing a pin-in-block type fuel block, and has the capability to demonstrate nuclear process heat utilization using an intermediate heat exchanger. Since 1986 a detailed design has been made, in which major systems and components are determined in line with the HTTR concept, paying essential considerations into the design for achieving the reactor outlet coolant temperature of 950°C. The safety review of the Government started in February 1989. By request of the Science and Technology Agency the Reactor Safety Research Association reviewed the safety evaluation guideline, general design criteria, design code and design guide for the graphite and the high-temperature structure of the HTTR. The installation permit of the HTTR was issued by the Government in November 1990.

Corrosion–erosion test of SS316 in flowing Pb–Bi

Abstract Corrosion tests of austenitic stainless tube were done under flowing Pb–Bi conditions for 3000 h at 450 °C. Specimens were 316SS produced as a tubing form with 13.8 mm outer diameter, 2 mm thickness and 40 cm length. During operation, maximum temperature, temperature difference and flow velocity of Pb–Bi at the specimen were kept at 450, 50 °C, and 1 m/s, respectively. After the test, specimen and components of the loop were cut and examined by optical microscope, SEM, EDX, WDX and X-ray diffraction. Pb–Bi adhered on the surface of the specimen even after Pb–Bi was drained out to the storage tank from the circulating loop. Results differed from a stagnant corrosion test in that the specimen surface became rough and the corrosion rate was maximally 0.1 mm/3000 h. Mass transfer from the high temperature to the lower temperature area was observed: crystals of Fe–Cr were found on the tube surface in the low-temperature region. The sizes of crystals varied from 0.1 to 0.2 mm. The depositing crystals were ferrite grains and the chemical composition ratio (mass%) of Fe to Cr was 9:1.

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.

Analytical method and result of off-site exposure during normal operation of High Temperature engineering Test Reactor (HTTR)

In order to evaluate off-site radiation exposure during normal operation of the High Temperature Engineering Test Reactor (HTTR), some amounts of fission products are assumed to be released from the fuel particles with defect of coating layers. In the evaluation, 1% of coated fuel particles is conservatively assumed to be failed in the HTTR. The release paths of fission products from the core to the stack are divided into two patterns, namely continuous release and periodical release. The annual offsite radiation exposure from the HTTR is 0.77 μSv/y in effective dose equivalent. This value is sufficiently lower than the reference dose of 50 μSv/y in normal operation.

Depressorization accident analysis for the FTTR by the TAC-NC

The two-dimensional thermal analysis code TAC-NC is modified from the analytical code TAC-2D in order to calculate temperature transients in the case of loss of forced cooling accidents of the HTTR (High Temperature engineering Test Reactor) such as a depressurization accident. The TAC-NC code includes a special function to calculate heat transfer by natural convection between hotter and colder regions in the pressure vessel during the depressurization accident.

Design and safety consideration in the High-Temperature engineering Test Reactor (HTTR)

The High Temperature engineering Test Reactor (HTTR) is a test reactor in Japan with thermal output of 30MW and reactor outlet coolant temperature of 950 °C at a high temperature test operation. This report describes features of the HTTR, emphases being laid on the safety design.

Experimental program of JT-60 and its relevance with the research of divertors

Abstract JT-60 is a large hydrogen tokamak with a compact poloidal divertor. The maximum input energy into the torus is about 500 MJ, and divertor studies are among the most essential experiments. The divertor in JT-60 has been studied numerically from three different view points, i.e. MHD equilibrium control, divertor plasma control and surface temperature control of the device. These results suggest that a serious heat flux can be avoided and a plasma contacting the divertor plates can be cooled down to 10 eV or less by controlling the magnetic configuration and divertor plasma density. The magnetic configuration with the divertor can be controlled by changing various coil currents, and the divertor plasma density can be controlled by changing the particle recycling condition in the divertor. The particle number in the main plasma can be also controlled by combining the compact divertor and gas influx both in the main chamber and in the divertor chamber.

Present status of HTGR development program in Japan

The JAERI started the HTGR development program in Japan in 1969 so as mainly to construct an experimental reactor for direct heat application. The energy situation, however, has changed remarkably in Japan during the last 20 years, then, the construction of High Temperature Engineering Test Reactor (HTTR) of 30MW in thermal power was decided instead of an experimental HTGR. The HTTR aims at establishing and upgrading the technology basis necessary for an HTGR, serving at the same time as a potential tool for new and innovative basic researches.

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.