K. Yokoyama

Burnout experiments on the externally-finned swirl tube for steady-state and high-heat flux beam stops

An experimental study to develop beam stops for the next generation of neutral beam injectors was started, using an ion source developed for the JT-60 neutral beam injector. A swirl tube is one of the most promising candidates for a beam stop element which can handle steady-state and high-heat flux beams. In the present experiments, a modified swirl tube, namely an externally-finned swirl tube, was tested together with a simple smooth tube, an externally finned tube, and an internally finned tube. The major dimensions of the tubes are 10 mm in outer-diameter, 1.5 mm in wall thickness, 15 mm in external fin width, and 700 mm in length. The burnout heat flux (CHF) normal to the externally finned swirl tube was 4.1 ± 0.1 kW/cm2, where the Gaussian e-folding half-width of the beam intensity distribution was about 90 mm, the flow rate of the cooling water was 30 l/min, inlet and outlet gauge pressures were about 1 MPa and 0.2 MPa, respectively, and the temperature of the inlet water was kept to 20 °C during a pulse. A burnout heat flux ratio, which is defined by the ratio of the CHF value of the externally-finned swirl tube to that of the externally-finned tube, turned out to be about 1.5. Burnout heat fluxes of the tubes with a swirl tape or internal fins increase linearly with an increase of the flow rate. It was found that the tube with external fins has effects that not only reduce the thermal stress but also improve the characteristics of boiling heat transfer.

High heat flux test of a HIP-bonded first wall panel of reduced activation ferritic steel F-82H

Abstract Reduced activation ferritic steel F-82H is a primary candidate structural material of DEMO fusion reactors. In fabrication technology, development of the DEMO blanket in JAERI, a hot isostatic pressing (HIP) bonding method, especially for the first wall structure with built-in cooling tubes has been proposed. A HIP-bonded F-82H first wall panel was successfully fabricated with selected manufacturing parameters. A high heat flux test of the HIP-bonded F-82H first wall panel has been performed to examine the thermo-mechanical performance of the panel including the integrity of the HIP-bonded interfaces and the fatigue behavior. A maximum heat flux of 2.7 MW/m 2 was applied to accelerate the fatigue test up to 5000 cycles in test blanket inserted ITER. The maximum temperature of the panel was ∼450°C under this heat flux. Through this test campaign, no damage such as cracks was observed on the surface of the panel, and no degradation in heat removal performance was observed either from the temperature responses. The thermal fatigue lifetime of the panel was found to be longer than the fatigue data obtained by mechanical testing.

High heat flux testing of a HIP bonded first wall panel with built-in circular cooling tubes

Abstract A HIP bonded DS-Cu/SS first wall (FW) with built-in circular cooling tubes was fabricated under the optimized HIP conditions. High heat flux testing of the panel was carried out on electron beam facility, JEBIS , at JAERI. The objective of this test is to examine the thermomechanical performance of the panel, including the integrity of the HIP bonded interfaces and also to examine the relation between the design fatigue curve and experimental results. Test conditions applied during these tests were 5.0–7.0 MW/m 2 in average, much higher than the ITER normal operation condition of 0.5 MW/m 2 , to accelerate the fatigue test with a repetition cycle up to 2500 cycles in total. High heat flux tests consisted of two test campaigns. Throughout these tests, no damages such as cracks were observed and no degradation in heat removal performance was also observed from temperature responses measured with thermocouples embedded within the panel. Thermomechanical integrity of the panel was confirmed within the parameter tests and the fatigue lifetime of the panel was found to be much longer than the design fatigue curve of this material, or even beyond the raw fatigue data.

Thermal shock tests on various materials of plasma facing components for FER/ITER

Development of plasma facing components and materials is a key element in the R&D program for the Fusion Experimental Reactor (FER), which has been designed at JAERI, and the International Thermonuclear Experimental Reactor (ITER), which has been designed under international collaboration. In these next-step tokamak devices, the plasma facing components and materials will be exposed to severe heat load and incident particle flux. The concern is especially acute that the extremely high thermal shock due to plasma disruption could cause material fracture. Efforts on developing the first wall and divertor have been energetically undertaken at JAERI. The present paper describes recent experimental and analytical results on thermal shock characteristics of various materials.

Disruption erosions of various kinds of tungsten

Abstract Thermal shock experiments on CVD-W, powder sintered tungsten (P-W), monocrystal tungsten (M-W) and brush tungsten (B-W) have been performed under the heat flux of 1000–2400 MW m −2 with the pulse length of 2 ms in JAERI electron beam irradiation system (JEBIS). The samples were exposed to a single pulsed beam with the electron energy of 70 keV and preheated up to 1000°C. In these experiments, the following information was obtained: (1) CVD-W is the most promising from an erosion point of view; (2) the increased weight loss was mainly caused by severe particle emission; (3) weight loss for every kind of tungsten was greater at 1000°C than at room temperature; (4) no cracks were observed at 1000°C for every sample by SEM observation; (5) secondary cracks were observed along the rolled direction at room temperature and may cause the exfoliation of a piece of tungsten surface during normal operation.

High heat flux experiment on B4C-overlaid C/C composites for plasma facing materials of JT-60U

High heat flux experiments (5–40 MW/m 2 , 5 s and 550 MW/m 2 , 5–10 ms) in the JAERI electron beam irradiation stand (JEBIS) have been carried out on three kinds (conversion, CVD and LLPS) of B 4 C-overlaid C/C composites, on which B 4 C is overlaid with a thickness of 100–250 μm. Measurements were made with respect to the weight loss, changes of the surface morphology and of the surface atomic composition, and the surface temperature. As a result of these experiments, it is found that B 4 C layers of all samples have no damages except small weight losses up to 12 MW/m 2 heat loads, which are estimated at the divertor tiles of JT-60U in normal plasma operation, and that the conversion method is the best of the three methods applied in the present tests, since no exfoliation has occurred even under the disruption conditions.

Design and experimental results of a new electron gun using a magnetic multipole plasma generator

A new electron gun utilizing a magnetic multipole plasma generator was designed and fabricated as the heat source of the high heat flux test facility, called JEBIS (JAERI electron beam irradiation stand). By changing the acceleration grids, this electron gun is able to produce a pencil to a sheetlike electron beams up to 4 A at 100 keV for 1 ms to continuous mode. In this electron gun, magnetic lens system is not adopted to focus the electron beam, but the space charge neutralization effect by the beam plasma produced downstream of the electron gun is utilized to prevent the blow‐up of the electron beam. In addition, high permeability metal is embedded in the first and the second grids to magnetically shield the earth field and the stray field from the beam bending magnet. It was experimentally demonstrated that wide range of heat flux from 0.2 MW/m2 to over 2000 MW/m2 can be realized at the test sample position about 1.7 m downstream of the electron gun.

Disruption tests on repaired tungsten by CVD coating

Abstract The chemical vapor deposition (CVD) coating is considered as one of the possible methods for in situ repairing of the tungsten armor. In the present work, CVD coatings on eroded tungsten specimens were prepared to investigate the applicability of this method to repair the eroded tungsten surface. To simulate the damaged surface relevant to the disruption erosion, specimens were irradiated by an electron beam at a heat flux of 1250 MW/m2 before the CVD repairing procedure. From metallographic results, no pores or cracks were observed at the interface between the CVD layer and the eroded layer seemed to be successfully repaired by the CVD coating. However, the CVD layer was delaminated by thermal shock tests which simulate disruption heat loads. It was found that complete removal of the resolidified area before CVD coating is effective to repair the eroded surface.

High heat flux experiments of saddle type divertor module

Abstract JAERI has been extensively developing plasma facing components for next tokomak devices. The authors have developed a saddle type divertor module which consists of saddle-shaped armor tiles brazed on metal heat sink. This paper presents the experimental and analytical results of thermal cycling experiments of the saddle type divertor module. The divertor module has unidirectional CFC armor tiles brazed on OFHC copper heat sink. A twisted tape was inserted in the cooling tube to enhance the heat transfer. In the experiments, thermal response of the divertor module was monitored by an infrared camera and thermocouples. The maximum incident heat flux was 24.5 MW/m2 for a duration of 30 s. No degradation of thermal response was observed during the experiment. As a result, the saddle type divertor module successfully endured at an incident heat flux of over 20 MW/m2 under steady state conditions for 1000 cycles.

Experimental and analytical results of carbon based materials under thermal shock heat loads for fusion application

Abstract To evaluate the durability and the lifetime of carbon materials as plasma facing components against plasma disruption, thermal shock tests were performed using an electron beam test facility. Experimental results show that the weight loss increases with heat flux at a constant absorbed energy. It is found that the erosion depth measured in the experiment is about 2 to 2.5 times larger than that predicted by the numerical analysis. Since solid particle emission is observed during heating experiments, the difference between experimental and numerical results is mainly attributable to the solid particle losses which are not considered in the numerical analysis.