S. Sato

Compact DEMO, SlimCS: design progress and issues

The design progress in a compact low aspect ratio (low A) DEMO reactor, SlimCS, and its design issues are reported. The design study focused mainly on the torus configuration including the blanket, divertor, materials and maintenance scheme. For continuity with the Japanese ITER-TBM, the blanket is based on a water-cooled solid breeder blanket. For vertical stability of the elongated plasma and high beta access, the blanket is segmented into replaceable and permanent blankets and a sector-wide conducting shell is arranged inbetween these blankets. A numerical calculation indicates that fuel self-sufficiency can be satisfied when the blanket interior is ideally fabricated. An allowable heat load to the divertor plate should be 8 MW m−2 or lower, which can be a critical constraint for determining a handling power of DEMO.

Design and technology development of solid breeder blanket cooled by supercritical water in Japan

This paper presents results of conceptual design activities and associated RD neutronics experiments and analyses; and development of the blanket module fabrication technology. Upon developing the fabrication technology for the first wall and box structure, a hot isostatic pressing bonded F82H first wall mock-up with embedded rectangular cooling channels was successfully fabricated. It withstood the high heat flux test at 2.7 MW m−2. Also, a correlation parameter of the Li2TiO3 pebble bed made by the sol–gel method was verified by measurement of the thermal conductivity of the breeder pebble bed, which is one of the most important design data.

SlimCS—compact low aspect ratio DEMO reactor with reduced-size central solenoid

The concept for a compact DEMO reactor named SlimCS is presented. Distinctive features of the concept are low aspect ratio (A = 2.6) and use of a reduced-size centre solenoid (CS) which has the function of plasma shaping rather than poloidal flux supply. The reduced-size CS enables us to introduce a thin toroidal field coil system which contributes to reducing the weight and perhaps lessening the construction cost. Low-A has merits of vertical stability for high elongation (κ) and high normalized beta (βN), which leads to a high power density with reasonable physics requirements. This is because high κ facilitates high nGW (because of an increase in Ip), which allows efficient use of the capacity of high βN. From an engineering aspect, low-A may ensure ease in designing blanket modules robust to electromagnetic forces acting on disruptions. Thus, a superconducting low-A tokamak reactor such as SlimCS can be a promising DEMO concept with physics and engineering advantages.

Optimization of HIP bonding conditions for ITER shielding blanket/first wall made from austenitic stainless steel and dispersion strengthened copper alloy

Abstract Optimum bonding conditions were studied on the Hot Isostatic Pressing (HIP) bonded joints of type 316L austenitic stainless steel and Dispersion Strengthened Copper alloy (DSCu) for application to the ITER shielding blanket / first wall. HIP bonded joints were fabricated at temperatures in a 980–1050°C range, and a series of mechanical tests and metallurgical observations were conducted on the joints. Also, bondability of two grades of DSCu (Glidcop Al-25 ® and Al-15 ® ) with SS316L was examined in terms of mechanical properties of the HIP bonded joints. From those studies it was concluded that the HIP temperature of 1050°C was an optimal condition for obtaining higher ductility, impact values and fatigue strength. Also, SS316L/Al-15 joints showed better results in terms of ductility and impact values compared with SS316L/Al-25 joints.

Evaluation of Shutdown Gamma-ray Dose Rates around the Duct Penetration by Three-Dimensional Monte Carlo Decay Gamma-ray Transport Calculation with Variance Reduction Method

For the evaluation of gamma-ray dose rates around the duct penetrations after shutdown of nuclear fusion reactor, the calculation method is proposed with an application of the Monte Carlo neutron and decay gamma-ray transport calculation. For the radioisotope production rates during operation, the Monte Carlo calculation is conducted by the modification of the nuclear data library replacing a prompt gamma-ray spectrum with a decay gamma-ray spectrum. By multiplying each correction factor, which is ratio of the actual activation level after shutdown to the production rate during operation, with each decay gamma-ray flux due to each radioisotope, the decay gamma-ray dose rate is evaluated. In order to improve the statistical error, a variance reduction method is proposed by the application of the weight window importance technique and the specification of the decay gamma-ray generation location. We identify the cell producing the decay gamma-ray which can contribute the decay gamma-ray flux in evaluation locations, and forcibly terminate the gamma-ray transport calculation in the cells except for the identified cells. In order to validate the effectiveness of the method, shielding calculation for actual ITER (International Thermonuclear Experimental Reactor) configuration is performed, and small statistical errors below criteria are obtained. The effectiveness of the proposed method for ITER design analysis is demonstrated.

Mechanical properties of HIP bonded joints of austenitic stainless steel and Cu-alloy for fusion experimental reactor blanket

Abstract Tensile, fatigue and impact properties have been measured for hot isostatic pressing (HIP) bonded joints of type 316 austenitic stainless steel (SS316)/SS316, and of SS316/Al2O3 dispersion strengthened copper (DSCu). The HIP bonded joints of SS316/SS316 had almost the same tensile and fatigue properties as those of the base metal. The HIP bonded joints of SS316/DSCu had also almost the same tensile properties as those of the base metal of the DSCu, though total elongation and fatigue strength were slightly lower than those of the DSCu base metal. Further data accumulation, even with further optimization of fabrication conditions, is required, especially for HIP bonded SS316/DSCu joints, to confirm above data and reflect to blanket/first wall design.

Nuclear and Thermal Analyses of Supercritical-water-cooled Solid Breeder Blanket for Fusion DEMO Reactor

Within a design study of a fusion DEMO reactor aiming at demonstrating technologies of fusion power plant, supercritical water is applied as a coolant of solid breeder blanket to attain high thermal efficiency. The blanket has multi-layer composed of solid breeder pebbles (Li2O) and neutron multiplier pebbles (Be) which are radially separated by cooling panels. The first wall and the breeding region are cooled by supercritical water below and above the pseudo- critical temperature, respectively. Temperature distribution and tritium breeding ratio (TBR) have been estimated by one-dimensional nuclear and thermal calculations. The local TBR as high as 1.47 has been obtained after optimization of temperature distribution in the breeder region under the following conditions: neutron wall loading of 5 MW/m2, 6Li enrichment of 30% and coolant temperature at inlet of breeder region of 380°C. In the case of the higher coolant temperature 430°C of the breeder region the local TBR was reduced to be 1.40. This means that the net TBR higher than 1.0 could be expected with the supercritical-water-cooled blanket, whose temperature distribution in the breeder region would be optimized by following the coolant temperature, and where a coverage of the breeder region is assumed to be 70%.

Development of joining technology for Be/Cu-alloy and Be/SS by HIP

Joining of Be/DSCu and Be/SS by using HIP technique with and without various interlayers were investigated as a screening test for selecting optimum joining method and conditions. Metallurgical observation and shearing tests were performed for basic characterization of the bonded joints. For Be/DSCu, the use of Ag interlayer with 700°C HIP temperature would be a prime candidate if Cd formation under neutron irradiation would not seriously affect plasma operation and joint performance. Other than the Ag interlayer, a Cr/Cu interlayer gave relatively high joint strength in the present screening test. The lower HIP temperature, 550°C, for this joint contributes to prevent sensitization of stainless steel (SS) structural material. As for Be/SS, the highest joint strength was obtained with a Ti interlayer. The HIP temperature of 800°C or a little higher would be applied for this joint to avoid SS sensitization.

Neutronics experiments for DEMO blanket at JAERI/FNS

In order to verify the accuracy of the tritium production rate (TPR), neutron irradiation experiments have been performed with a mockup relevant to the fusion DEMO blanket consisting of F82H blocks, Li2TiO3 blocks with a 6Li enrichment of 40% and 95%, and beryllium blocks. Sample pellets of Li2TiO3 were irradiated and the TPR was measured by a liquid scintillation counter. The TPR was also calculated using the Monte Carlo code MCNP-4B with the nuclear data library JENDL-3.2 and ENDF-B/VI. The results agreed with experimental values within the statistical error (10%) of the experiment. Accordingly, it was clarified that the TPR could be evaluated within 10% uncertainty by the calculation code and the nuclear data. In order to estimate the induced activity caused by sequential reactions in cooling water pipes in the DEMO blanket, neutron irradiation experiments have been performed using test specimens simulating the pipes. Sample metals of Fe, W, Ti, Pb, Cu, V and reduced activation ferritic steel F82H were irradiated as typical fusion materials. The effective cross-sections needed to calculate the formation of the radioactive nuclei (56Co, 184Re, 48V, 206Bi, 65Zn and 51Cr) due to sequential reactions were measured. From the experimental results, it was found that the effective cross-sections increased remarkably while coming closer to polyethylene board, which was a substitute for water. As a result of this present study, it has become clear that the sequential reaction rates are important factors in the accurate evaluation of induced activity in fusion reactor design.

Evaluation of hot isostatic pressing for joining of fusion reactor structural components

Hot isostatic pressing (HIP) is a promising technology to fabricate the blanket structure of fusion reactors. HIP joining of solid materials has been selected as a reference fabrication method for the shielding blanket/first wall of the international thermonuclear experimental reactor (ITER). On the basis of experimental results obtained in Europe, Japan and Russia, an evaluation of HIP joining for fusion reactor structural components has been carried out. The parameters of HIP fabrication for copper alloys and stainless steels are given. The results of microscopic observations, X-ray microanalysis, tensile, impact toughness, fracture toughness and fatigue tests are presented. Material science criteria for an estimation of quality for joints fabricated by HIP are discussed.