Tatsuzo Tone

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.

Lifetime analysis for fusion reactor first walls and divertor plates

Lifetime analysis of fusion reactor first walls and divertor plates is performed by (1) a one-dimensional analytical plate model, and (2) a two-dimensional elastic-plastic finite element method. Life-limiting mechanisms and the limits of applicability for these analysis methods are examined. Structural design criteria are also discussed.

Fatigue strength of tungsten-copper duplex structures for divertor plates

Abstract A tungsten-copper duplex structure is specified in a conceptual design of the Japan Fusion Experimental Reactor (FER). The evaluation of the fatigue and creep life of the interface region between tungsten and copper is essential for design of the divertor plate. Fatigue crack initiation life and crack propagation behavior at room temperature and 200°C were measured for fully-annealed OFHC copper and for tungsten-OFHC copper joints brazed with amorphous nickel-base filler metal. The debonding fatigue strength for the brazed joints was relatively high, but less than that of the copper. Fatigue crack growth rates in the braze layer was approximately similar to that of the copper. Fatigue lives were estimated for the divertor plate with small defects, and a method for analyzing the apparent K- values of interface cracks was presented.

A thermal cycling durability test of tungsten copper duplex structures for use as a divertor plate

A tungsten-copper (W—Cu) duplex structure is the reference divertor plate design concept for the Fusion Experimental Reactor (FER). In the present study, a durability test of the W—Cu duplex structure against thermal cycling was performed. A tungsten disk was bonded to a copper disk by means of brazing or direct casting. The tungsten surface of the test piece was periodically heated by a high temperature argon plasma jet. Before and after the tests, the test pieces were examined with the aid of a scanning electron microscope and the Knoop hardness was measured. Grain boundary microcracks were observed after 200 and 1100 thermal cycles in brazed tungsten samples which contained a small amount of nickel and phosphorus. A cast tungsten specimen subjected to 2200 thermal cycles also contained microcracks. However, microcracks were not observed in a brazed tungsten sample containing an extremely small amount of impurities for thermal cycles up to 3700 times. Microcracks were observed in the brazing material of this test piece. None of the test specimens were broken. It is found that brazing is a valid bonding method and that W—Cu duplex structure, especially with high purity tungsten, is able to endure a practical number of thermal cycles.

Technical considerations on tritium breeding blanket systems for a fusion power reactor

Design studies of five candidate tritium breeding blankets for a fusion power reactor have been performed. The concept of a breeder outside of the coolant tubes (BOCT) was adopted for Li2O blankets in which the breeder is indirectly cooled by water or helium. For LiAlO2, a breeder inside the coolant tubes (BICT) concept was adopted on the basis of the chemical stability of LiAlO2. In this blanket the breeder is directly cooled by helium coolant. These solid breeders are fabricated in the form of small spheres to prevent thermal cracking and to ease in filling them in the blanket. A liquid lithium breeder employs a self-cooling concept and also a BOCT concept with helium indirect cooling of stagnant liquid lithium. In addition to evaluations of each blanket performance, technological problems and R&D issues for power reactor blanket development have been also identified.

On Elastic Removal Cross Sections of Light and Medium Elements near Sodium Resonance

A number of authors have discussed various aspects of the elastic removal cross section ssr for a fast reactor(1)~(6). None have however examined in detail the accuracy of the approximation methods adopted in typical standard sets such as the YOM(7) or the Russian ABN set(8). In this note we discuss the elastic removal cross sections of light elements in energies around 3 keV (sodium resonance energy) in order to examine the influence on cr„ of group boundaries and of different calculation methods. For the i-th group, where the energy interval DEi extends from Eui, to Eli (Eui>=Eli/a),

Analysis and experiments on lifetime predictions for first wall and divertor plate structures in JAERI

Analysis and experiments on lifetime predictions of the first wall and divertor plate of fusion reactors have been performed. The analysis is based on both a one-dimensional plate model and a two-dimensional elastic-plastic finite element method. The experiments consist of mechanical fatigue tests, thermal fatigue tests, and tests of vaporized loss of materials. From these results, discussions of lifetime prediction procedures for fusion reactor components are made.

Design studies of Tokamak power reactor in JAERI

Recent design studies of tokamak power reactor and related activities conducted in JAERI are presented. A design study of the SPTR (Swimming-Pool Type Reactor) concept was carried out in FY81 and FY82. The reactor design studies in the last two years focus on nuclear components, heat transport and energy conversion systems. In parallel of design studies, tokamak systems analysis code is under development to evaluate reactor performances, cost and net energy balance.

Analysis of magnetomechanical behavior of ferromagnetic first wall component

Ferritic stainless steel(HT-9) is a prospective candidate for a first wall material of a fusion reactor. It experiences magnetic stress due to magnetization in magnetic field. A ferromagnetic cantilever of mild steel was provided to carry out a test on magnetomechanical behavior and to compare with theoretical prediction. The theoretical prediction was made for an infinitely wide beam plate and magnetic stiffness was taken into account. Field distribution of the finite specimen is very different from one of an infinite specimen. It is made clear that deformation is proportional to the squared field for smaller applied field while linear with the field for larger one.

Effects of Fission Products on Physical Quantities Accompanying Burn-Up of Large Pu-U Fast Reactors

The effects brought by the presence of fission products (F.P.) on the effective multiplication factor k eff, the Na-void reactivity, the breeding ratio, the fuel composition and kinetics parameters have been calculated as functions of burn-up for Pu-U fast reactor with 3,000l core volume. The F.P. sharply reduce k eff and increase the positive values of the Na-void reactivity. Moreover, at a given burn-up, this effect of F.P. on k eff and Na-void reactivity is governed largely by the total amount of the F.P. found accumulated at the time of observation, and is independent of the history of the material. The F.P. hardly influence the transformation accompanying burn-up undergone by the ratio of Pu to 238U atoms and by the isotopic composition of Pu. Similarly, the effect on the internal breeding ratio also is very small. The total breeding ratio increases gradually with accumulation of the F.P. The effect on the effective delayed neutron fraction βeff is only slight, while that on the prompt-neutron lifeti...