Tsuneo Kodaira

Evaluation of toughness degradation by small punch (SP) tests for neutron-irradiated 214Cr-1Mo steel

Mechanical properties correlations between the small punch (SP) test and conventional tensile, Charpy impact and fracture toughness tests were investigated on neutron irradiated 214Cr-1Mo ferritic steel. Estimation of radiation induced changes on tensile strength, elastic-plastic fracture toughness (JIC) and ductile-to-brittle transition temperature (DBTT) are thought to be basically possible by mechanical properties correlations when based on sufficient pre-irradiation data.

High resolution static imaging system using a cooled CCD camera

Abstract An intelligent filter, originally a median filter, was developed and introduced as noise correction in order to distinguish white spots from pixels which hold image information. The algorithm of the intelligent filter is able to correct dark spots as well as white spots using independent criteria and does not affect any other signals. The experimental results confirmed that the intelligent filter worked as a white spot correction effectively.

Weldability of neutron-irradiated type 316 stainless steel

Abstract The post-irradiation weldability of neutron-irradiated type 316 stainless steel has been examined. The tungsten inert-gas arc welded joints were made of type 316 stainless steel used as wrapper tube in the JOYO fast reactor at 668–683 K to a maximum dose and helium content of about 22 dpa and 9 appm respectively. Large loss of tensile ductility was observed in welded joints made of the irradiated steels, being accompanied with intergranular brittle fracture in the heat affected zone at ambient temperature (about 293 K) and at 773 K, although ductile fracture of unirradiated specimens was observed in the weld metal region. The brittle fracture is closely related to helium bubble formation on grain boundaries during the welding.

Dimensional Stability of Uranium Silicide Plate-Type Reactors at Transient Condition

This paper describes the result of transient experiments using low enriched uranium silicide plate-type fuel for research reactors. The pulse irradiation was carried out at Nuclear Safety Research Reactor (NSRR) in Japan Atomic Energy Research Institute. The results obtained were: 1. At fuel plate temperature of below 400°C, a good dimensional stability of the tested fuel was kept. No fuel failure occurred. 2. At a plate temperature of about 540°C, a local crack was initiated on the AI-3%Mg alloy cladding. Once the cladding temperature exceeded the melting point of 640°C, the fuel plate was degraded much by increased bowing and cracking of the denuded fuel meat occurred after relocation of molten Al cladding. Despite of these degradation, neither fragmentation of the fuel plate nor mechanical energy generation occurred up to the cladding temperature of 971 °C. 3. At the temperatures of around 925°C, the reaction of silicide particles with molten Al in the matrix and that of cladding occurred, forming Al r...

Swelling behavior of welded type 316 stainless steel and its improvement

Abstract Type 316 stainless steel was electron beam welded with titanium foil insertion. The concentration of introduced titanium in the weld metal was 0.1, 0.3 and 0.6 wt% corresponding to the inserted foil thickness of 10, 30 and 60 μm, respectively. All the weld joint showed good mechanical performance. The swelling resistance of the weld metal is effectively improved by the introduced titanium. Although inhomogeneous distribution of titanium makes it difficult to estimate the extent of the improvement quantitatively, the results suggest the applicability of this method to Ti-modified 316 stainless steel, where weld metal is already reported to show reduced swelling resistance.

Low cycle fatigue behavior of titanium carbide coated molybdenum

Low cycle fatigue tests of TiC coated and uncoated molybdenum materials were performed in a vacuum at room temperature, 773 and 1073 K in order to examine the influence of TiC coating on the fatigue strength. The relation between the fatique life, N f and the plastic strain range, Δe ap was expressed by the equation, Δe ap N f β = C , where β and C are constants. At 773 and 1073 K uncoated molybdenum fractured on the plane of the maximum shear stress, but on the plane of the maximum tensile stress at room temperature. The low cycle fatigue strength of TiC coated molybdenum decreased compared with that of the uncoated one at 773 K, where cracks were observed to be initiated at the TiC coating layer and to extend into the molybdenum matrix by a notch effect. At 1073 K, however, the TiC coated molybdenum exhibited higher fatigue strength than the uncoated one since the TiC coating layer seemed to restrict a slip deformation of molybdenum matrix.

Influence of microstructural change caused by cyclic strain on the low cycle fatigue strength of sintered molybdenum

Low cycle fatigue tests of sintered molybdenum with a 70% reduction by hot rolling were performed in a vacuum at room temperature, 573, 773, 1073 and 1223 K in order to examine the influence of microstructural change on the low cycle fatigue strength. The molybdenum exhibited cyclic softening at the temperatures between 573 and 1073 K, though it did not indicate the softening at 1223 K because of static recrystallization. The cyclic softening occurred progressively in some parts of the specimen due to the dynamic microstructural change caused by cyclic strain. Dynamic recrystallization occurred at 1073 K apparently during low cycle fatigue tests, and this recrystallized grain grew with increasing cycles. The plastic deformation was consequently concentrated in the softening region of the specimen. Therefore, the low cycle fatigue lives of molybdenum tested at temperatures below 1073 K were shorter than those at 1223 K, while the specimens were recrystallized and deformed uniformly at 1223 K.

Temper and neutron irradiation embrittlement in 2 1/4 Cr-1 Mo steels for pressure vessels of high-temperature gas-cooled reactors

A 2 1/4 Cr-1 Mo steel is a promising candidate material for structural components of the pressure vessel of the experimental very high temperature gascooled reactor (VHTR) in Japan. Since the service temperature of such components is expected to be about 400/sup 0/C, the behavior of the temper and neutron irradiation embrittlements in these chromium-molybdenum steels should be confirmed from the viewpoint of structural integrity. The experimental verification on the degree of the embrittlement due to thermal aging, including the effect of applied stress and neutron irradiation, is described. Steel containing substantial amounts (about 100 ppm) of phosphorus atoms, which are believed to cause the temper embrittlement, showed that applied stress enhanced the embrittlement due to thermal aging. Embrittlement caused by neutron irradiation appears to be minimal in the case of the material containing <1000 ppm of copper as impurity with neutrons irradiated at about 400/sup 0/C.