Seiichiro Yamazaki

A fusion power reactor concept using SiC/SiC composites

JAERI studied a concept of a commercial fusion power reactor (5.5 GW, electric output: 2.7 GW) having high environmental safety, high thermal efficiency and high availability. The reactor configuration was designed to achieve good maintainability, high performance breeding blanket, high efficiency power generation system and less radwastes. The design was based on the use of low activation structural material (SiC/SiC Composites) and helium as a coolant. (1) Easy maintenance is attained by sector replacement with the radiation environment less than 103 R/h in a reactor chamber. (2) The net thermal efficiency over 45% is attained by high temperature helium gas Brayton cycle. (3) Most of radwastes of DREAM reactor can be disposed in shallow land burial as a low level radwaste after cooling of several tens of years.

Blanket and divertor design for the Steady State Tokamak Reactor (SSTR)

Abstract The tritium breeding blanket and plasma facing components (first wall and divertor) have been designed for the Steady State Tokamak Reactor (SSTR). Low activation ferritic steel (F82H) was chosen for the structural material of the first wall and the blanket. The ceramic breeder pebble bed concept with beryllium neutron multiplier was adopted. The blanket is divided into two parts; a replaceable blanket and a permanent blanket. Electrical insulation made of functionally gradient material (FGM) was introduced in the blanket box structure to drastically reduce electromagnetic loads during a plasma disruption and to simplify the attaching mechanism of the blanket. Low-temperature and high-density divertor plasma conditions and a radiative cooling concept lower the heat flux to the divertor plate to below 3 MW/m2 and enable us to adopt the same cooling conditions as the blanket coolant. Puffing mechanisms of gas (deuterium) and iron are installed for the radiative cooling of the divertor plasma.

A simulated plasma disruption experiment using an electron beam as a heat source

An experimental study was made on the behavior of a solid surface subjected to an extremely high heat flux similar to that expected during a plasma disruption. An electron beam was used as the heat source to simulate the high heat flux. The beam was defocused in an attempt to give as much uniform heat flux as possible on the test surface. The 5-mm-diameter test pieces were made of 304 stainless steel, aluminum, and zinc. Heat fluxes from 10 to 110 MW/m2 were applied on the test pieces for durations of 90 to 180 msec. Special attention was paid to the measurement of the surface heat flux on the test surface. Comparison between experimental and analytical results on melt layer thickness and evaporation loss is made. An improved thermal analysis code (DAT-K) was developed for the analysis. Agreement between the experimental and analytical results on melt layer thickness is good. For evaporation loss, experimental and analytical results are in fair agreement. Features of the experiments and analysis that lead to the differences in the results are discussed.

Improvement of an electron beam facility as a heat source for disruption simulation experiments

To perform simulated plasma disruption experiments, a heat source which can provide fast rise and high heat flux on a target surface is required. An existing electron beam facility has been improved to provide higher heat fluxes uniformly over a wider area by installing a high speed beam rastering system. The beam rastering coils can provide a beam oscillation angle of 4.4° with a frequency of up to 400 kHz. The high frequency ensures temporal uniformity of the heat flux on a test surface. Triangular wave shaped current is supplied to excite the coils to provide spatially uniform heat flux over a test surface. Using this electron beam facility, we can provide uniform heat fluxes as high at 160 MW/m2 on an area of 13 mm × 13 mm, and 20 MW/m2 on 38 m × 38 mm.

Prototype tokamak fusion reactor based on SiC/SiC composite material focusing on easy maintenance

If the major part of the electric power demand is to be supplied by tokamak fusion power plants, the tokamak reactor must have an ultimate goal, i.e. must be excellent in construction cost, safety aspect and operational availability (maintainability and reliability), simultaneously. On way to the ultimate goal, the approach focusing on the safety and the availability (including reliability and maintainability) issues must be the more promising strategy. The tokamak reactor concept with the very high aspect ratio configuration and the structural material of SiC/SiC composite is compatible with this approach, which is called the DRastically Easy Maintenance (DREAM) approach. This is because SiC/SiC composite is a low activation material and an insulation material, and the high aspect ratio configuration leads to a good accessibility for the maintenance machines. As the intermediate steps along this strategy between the experimental reactor such as international thermonuclear experimental reactor (ITER) and the ultimate goal, a prototype reactor and an initial phase commercial reactor have been investigated. Especially for the prototype reactor, the material and technological immaturities are considered. The major features ofthe prototype and commercial type reactors are as follows. The fusion powers of the prototype and the commercial type are 1.5 and 5.5 GW, respectively. The major/minor radii for the prototype and the commercial type are of 12/1.5 m and 16/2 m, respectively. The plasma currents for the prototype and the commercial type are 6 and 9.2 MA, respectively. The coolant is helium gas, and the inlet/outlet temperatures of 500/800 and 600/900°C for the prototype and the commercial type, respectively. The thermal efficiencies of 42 and 50% are obtainable in the prototype and the commercial type, respectively. The maximum toroidal field strengths of 18 and 20 tesla are assumed in the prototype and the commercial type, respectively. The thermal conductivities of 15 and 60 W/m per K are assumed in the prototype and the commercial type, respectively.

Development of fabrication technologies for the ITER divertor

Abstract The ITER divertor cassette consists of the cassette body and the high heat flux components (HHFCs), which include the vertical target, wing, dome and liner. In order to develop and fabricate these HHFCs, an improvement to the existing brazing of the surface materials to the cooling structures is necessary. Carbon fiber composites (CFCs) are one of the candidate surface materials, which are brazed to the cooling structures made of copper or copper alloys such as dispersion strengthened copper (DSCu). Available brazing materials were researched and some candidates were tested to assess their capability in this application. The fabrication technology of the cooling structure of HHFCs, which will have a complicated geometry, is also of concern. The wing must withstand the maximum heat flux of 5 MW/m 2 . Also, it has to satisfy connectivity to the neighboring HHFCs and the cassette body. For a wing concept, a CFC mono-block with copper alloy cooling tubes was proposed and some mock-ups using this concept were fabricated. Through the trial fabrication of the full-sized mock-up, the feasibility of the fabrication procedure for the CFC mono-block wing has been confirmed.

High heat flux test on first wall materials

The plasma facing components in a tokamak type fusion reactor are subjected to intense heat load during a plasma disruption. The effects of the resulting high heat flux on the wall materials must be evaluated. The authors performed high heat flux tests with a high-frequency rastering beam as the heat source and thermal and stress analyses to investigate the thermo-mechanical behaviors of candidate wall materials. Based on the results of the studies, melting, evaporation and crack initiation, propagation behaviors of metallic and carbon materials are discussed. The heat transfer enhancement of melted metal motion of metallic materials, and the particle emission of carbon materials is important to the disagreement between experimental and analytical results on melting and evaporation. For stainless steel, a lot of micro-cracks with a depth of about 0.1 mm along dendrites initiated in the resolidified zone. They did not propagate in the non-melted zone. For tungsten, cracks with a depth of a few millimeters initiated from the periphery zone of the heated surface and propagated into the non-melted zone with repeated heating.

Thermal shock fracture of graphite armor plate under the heat load of plasma disruption

Experiments on the thermal shock brittle fracture of graphite plates were performed. Thermal loading which simulated a plasma disruption was produced by an electron beam facility. Pre-cracks produced on the surface propagated to the inside of the specimen even if the thermal stress on the surface was compressive. Two mechanisms are possible to produce tensile stress around the crack tip under thermal shock conditions. Temperature, thermal stress, and the stress intensity factor for the specimen were analyzed based on the finite element method for various heating conditions. The trend of experimental results under the asymmetric heating agrees qualitatively with the analytical results. This phenomenon is important for the design of plasma facing components made of graphite. Establishment of a lifetime prediction procedure including fatigue, fatigue crack growth, and brittle fracture is needed for graphite armors.

Maintenance and material aspects of DREAM reactor

A concept of a commercial fusion power reactors (Fusion Power: 5.5 GW, electric output: 2.7 GW) having high environmental safety, high thermal efficiency and high availability has been studied in JAERI. The gross reactor configuration was designed to achieve good maintainability, high performance breeding blanket, high efficient power generation system and little radwastes. Design was based on the use of low activation structural material (SiC/SiC composites) and helium as a coolant. In this paper, maintenance and material aspects of DREAM reactor design is discussed. The concluding remarks are as follows. (1) The difficulty of development of maintenance tool is alleviated by sector replacement and the radiation dose environment less than 10 Gy/h in a reactor chamber. (2) Design requirement and present status of SiC/SiC composites was investigated. (3) The SiC/SiC composite development program is planned to satisfy the requirements of DREAM reactor.

Proposal of Integrated Test Facility for in-Vessel Thermofluid Safety of Fusion Reactors

A vacuum vessel (VV) of a tokamak fusion reactor like the International Thermonuclear Experimental Reactor (ITER) consists the first confinement barrier that includes the largest amount of radioactive materials such as tritium and activation products. The ingress of coolant event (ICE) is a design basis event in the ITER where water is used as the coolant. The loss of vacuum event (LOVA) is also considered as an independent design basis event. Based on the results of ICE and LOVA preliminary experiments, an integrated in-vessel thermofluid test is being planned and conceptual design of the facility is in progress. The main objectives of the integrated test are to investigate the consequences of possible interaction of the ICE and the LOVA and to validate the analytical model of thermofluid events in the VV of the fusion reactor. This paper introduces a conceptual design of the integrated test facility and a testing plan.