Minoru Narui

Radiation induced conductivity of ceramic insulators measured in a fission reactor

In-reactor measurements of the long-term change in the electrical conductivity of α-alumina were carried out in the JMTR fission reactor. Special attention was focussed on the effect of applied voltage on electrical degradation. Two experiments were carried out for 96 and 48 reactor full power days at 600–630 K and 770–800 K, with an applied electric field of 500 V/m AC and 500 kV/m DC, respectively. A long-term increase in electrical conductivity was observed, which is thought to be radiation induced electrical degradation, RIED.

Irradiation hardening of reduced activation martensitic steels

Abstract Irradiation response on the tensile properties of 9Cr2W steels has been investigated following FFTF/MOTA irradiations at temperatures between 646 and 873 K up to doses between 10 and 59 dpa. The largest irradiation hardening accompanied by the largest decrease in the elongation is observed for the specimens irradiated at 646 K at doses between 10 and 15 dpa. The irradiation hardening appears to saturate at a dose of around 10 dpa at the irradiation temperature. No hardening but softening was observed in the specimens irradiated at above 703 K to doses of 40 and 59 dpa. Microstructural observation by transmission electron microscope (TEM) revealed that the dislocation loops with the a〈100〉 type Burgers vector and small precipitates which were identified to be M6C type carbides existed after the irradiation at below 703 K. As for the void formation, the average size of voids increased with increasing irradiation temperature from 646 to 703 K. No voids were observed above 703 K. Irradiation softening was attributed to the enhanced recovery of martensitic structure under the irradiation. Post-irradiation annealing resulted in hardening by the annealing at 673 K and softening by the annealing at 873 K.

Charpy impact testing using miniature specimens and its application to the study of irradiation behavior of low-activation ferritic steels

Abstract The effectiveness of mini-size Charpy V-notch specimens with a 1.5 or 1.0 mm square cross section in measuring the ductile brittle transition temperature (DBTT) and upper shelf energy (USE) compared with full-size specimens is evaluated for a ferritic steel. It is shown that the data from the mini-size specimens can be used to estimate the DBTT and USE for full-size specimens when the measured absorbed energy-temperature curves are normalized by appropriate parameters. The result is applied to the study of neutron irradiation embrittlement of low-activation ferritic steels.

Mechanical property changes of low activation ferritic/martensitic steels after neutron irradiation

Mechanical property changes of Fe–XCr–2W–0.2V,Ta (X: 2.25–12) low activation ferritic/martensitic steels including Japanese Low Activation Ferritic/martensitic (JLF) steels and F82H after neutron irradiation were investigated with emphasis on Charpy impact property, tensile property and irradiation creep properties. Dose dependence of ductile-to-brittle transition temperature (DBTT) in JLF-1 (9Cr steel) irradiated at 646–700 K increased with irradiation up to 20 dpa and then decreased with further irradiation showing highest DBTT of 260 K at 20 dpa. F82H showed similar dose dependence in DBTT to JLF-1 with higher transition temperature than that of JLF-1 at the same displacement damage. Yield strength in JLF steels and F82H showed similar dose dependence to that of DBTT. Yield strength increased with irradiation up to 15–20 dpa and then decreased to saturate above about 40 dpa. Irradiation hardening in 7–9%Cr steels (JLF-1, JLF-3, F82H) were observed to be smaller than those in steels with 2.25%Cr (JLF-4) or 12%Cr (JLF-5). Dependences of creep strain on applied hoop stress and neutron fluence were measured to be 1.5 and 1, respectively. Temperature dependence of creep coefficient showed a maximum at about 700 K which was caused by irradiation induced void formation or irradiation enhanced creep deformation. Creep coefficient of F82H was larger than those of JLF steels above 750 K. This was considered to be caused by the differences in N and Ta concentration between F82H and JLF steels.

Irradiation test of diagnostic components for ITER application in the Japan Materials Testing Reactor

Radiation effects in components and materials will be one of the most serious technological issues in nuclear fusion systems realizing burning-plasmas. Especially, diagnostic components, which should play a crucial role in controlling plasmas and understanding the physics of burning-plasmas, will be exposed to high-flux neutrons and gamma rays. Dynamic radiation effects will affect the performance of components substantially from the beginning of exposure to radiation environments, and accumulated radiation effects will gradually degrade their functioning abilities in the course of their service. High-power-density fission reactors will be the only realistic tools to simulate the radiation environments expected to occur in burning-plasma fusion machines such as the International Thermonuclear Experimental Reactor (ITER), at present. Some key diagnostic components, namely magnetic coils, bolometers, and optical fibres, were irradiation-tested in a fission reactor, to evaluate their performances in heavy radiation environments. Results indicate that ITER-relevant radiation-resistant diagnostic components could be developed in time, although there are still some technological problems to be overcome.

Effect of specimen size on the ductile-brittle transition behavior and the fracture sequence of 9Cr-W steels

Abstract The effect of specimen size on the ductile-brittle transition behavior and the fracture sequence were investigated by means of Charpy absorbed energy measurement and fractography, using the full size, the half size and the one-third size V-notch specimens of 9Cr-W steels. The steels used are reduced-activation ferritic steels for fusion reactor structures. Attempts were made to correlate the impact data between the different specimen sizes by using normalizing parameters, such as nominal fracture area and nominal fracture volume for the upper shelf energy and ligament size for the ductile-brittle transition temperature. Fractography showed a similar fracture sequence for the three different sizes of the specimens.

Irradiation embrittlement of neutron-irradiated low activation ferritic steels

Abstract Effects of neutron irradiation and additions of small amounts of alloying elements on the ductile-brittle transition temperature (DBTT) of three different groups of ferritic steels were investigated by means of the Charpy impact test in order to gain an insight into the development of low-activation ferritic steels suitable for the nuclear fusion reactor. The groups of ferritic steels used in this study were (1) basic 0–5% Cr ferritic steels, (2) low-activation ferritic steels which are FeCrW steels with additions of small amounts of V, Mn, Ta, Ti, Zr, etc. and (3) FeCrMo, Nb or V ferritic steels for comparison. In Fe-0–15% Cr and FeCrMo steels, Fe-3–9% Cr steels showed minimum brittleness and provided good resistance against irradiation embrittlement. Investigations on the effects of additions of trace amounts of alloying elements on the fracture toughness of low-activation ferritic steels made clear the optimum amounts of each alloying element to obtain higher toughness and revealed that the 9Cr-2W-Ta-Ti-B ferritic steel showed the highest toughness. This may result from the refinement of crystal grains and improvement of quenching characteristics caused by the complex effect of Ti and B.

Enhancement of irradiation hardening by nickel addition in the reduced-activation 9Cr–2W martensitic steel

Reduced-activation martensitic (RAM) steels with and without an addition of 1% Ni were irradiated in a so called multisection-multidivision controlled irradiation capsule in the JMTR at 220°C up to 0.15 dpa. The 1/4 power dependence of the irradiation hardening on neutron dose was observed for the specimens irradiated in the controlled capsule. A part of the specimens were simultaneously irradiated in the capsule out of the reactor core where the irradiation temperature was considered to be lower than 170°C. The out of-reactor core irradiation induced a tremendous irradiation hardening as much as 350 MPa in the Ni added RAM steel but only 120 MPa of the hardening in the unadded RAM steel. The tremendous irradiation hardening was never observed following the irradiation at 220°C. As for the results of positron annihilation measurements, no significant effect of the Ni addition was observed in the life time spectrum. Post-irradiation annealing studies indicate that the irradiation hardening observed in the Ni added RAM steel begins to recover at 190°C and diminishes after the annealing at 250°C.

Dependence of impact properties on irradiation temperature in reduced-activation martensitic steels

Ductile–brittle transition (DBT) behavior of 9%Cr-2%W reduced-activation martensitic (RAM) steels has been investigated following neutron irradiation in the fast flux test facility, materials open test facility (FFTF/MOTA) at different temperatures. Both the irradiations at 663 and 733 K cause an increase in DBT temperature, while the irradiation at 663 K induces the hardening and the softening at 733 K. Microstructural observation by transmission electron microscope (TEM) revealed that small dislocation loops existed in the specimen irradiated at 663 K and no such a loop, but relatively large M6C carbides and Laves phase were formed by the irradiation at 733 K. There appears to be a linear dependence between ΔDBTT and ΔσY in neutron irradiated RAM steels when irradiation induces the hardening. Irradiation embrittlement accompanied by the softening is considered to be due to reduction of cleavage fracture stress caused by the irradiation-induced recovery of the martensitic structure, namely decrease in dislocation density and formation of large precipitates.

Radiation-induced electrical degradation experiments in the Japan materials testing reactor

Abstract An experiment to measure radiation-induced electrical degradation (RIED) in a sapphire sample and in three MgO-insulated cables was conducted at the JMTR light water reactor. The materials were irradiated at about 260°C to a fluence of 3 × 1024 n/m 2 ( E > 1 MeV) with an applied DC electric field between 100 kV/m and 500 kV/m. Even though the results for the sapphire sample are somewhat ambiguous because of an unexplained offset current of about 0.6 μA substantial degradation was not observed in the sapphire: instead, radiation-induced conductivity (RIC) seemed to decrease slightly during the experiment. Substantial increase in leakage current, that increased with applied electric field, occurred in the MgO-insulated cables. This increased conductivity disappeared when the reactor was shut down and sample temperature returned to ambient. However, the physical degradation apparently remained in the material while the reactor was off because restarting the irradiation brought the conductivity back to its previous, degraded, reactor-on value. This effect is different from the RIED effect reported by Hodgson but is similar to previous results reported by Shikama et al. Considerable data were taken to determine the sample temperature and leakage currents during the irradiation.