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The authors previously conducted thermogravimetric analyses on zircaloy-2 in air. By using the thermogravimetric data, an oxidation model was constructed in this study so that it can be applied for the modeling of cladding degradation in spent fuel pool (SFP) severe accident condition. For its validation, oxidation tests of long cladding tube were conducted, and computational fluid dynamics analyses using the constructed oxidation model were proceeded to simulate the experiments. In the oxidation tests, high temperature thermal gradient along the cladding axis was applied and air flow rates in testing chamber were controlled to simulate hypothetical SFP accidents. The analytical outputs successfully reproduced the growth of oxide film and porous oxide layer on the claddings in oxidation tests, and validity of the oxidation model was proved. Influence of air flow rate for the oxidation behavior was thought negligible in the conditions investigated in this study., これまでにジルカロイ2被覆管の熱天秤による空気中酸化試験を行ったが、本研究ではそのデータに基づいて使用済み燃料プール(SFP)事故時の被覆管の酸化挙動の解析に適用可能な酸化モデルの構築を行った。その検証にあたり長尺被覆管の酸化試験に関して当該酸化モデルを適用した数値流体解析を行った。酸化試験はSFP事故を模擬した空気流量条件中で高温の温度勾配を付与して実施した。構築した酸化モデルを適用した解析は酸化試験での被覆管表面の酸化皮膜及び多孔質な酸化層の成長をよく再現し、酸化モデルの妥当性が確認できた。本研究の試験条件の範囲では空気流量条件の酸化挙動への影響は明らかには見られなかった。

After the Fukushima Daiichi nuclear power plant accident due to the pacific coast of Tohoku earthquake and Tsunami, we have been developing an analysis method considering creep damage mechanisms based on three-dimensional analysis in order to estimate condition of the fuel debris and failure behavior of the reactor due to severe accidents for the early completion of the decommissioning of nuclear power plants. We have also been obtaining material properties that are not provided in existing databases or literature for the analysis. In this study, we measure the tensile and creep properties of low alloy steel, Ni-based alloy, and stainless steel at high temperatures near the melting points. Using experimental data, some parameters for the creep constitutive law and creep failure evaluation method are determined. It is confirmed that the estimated rupture time by the failure evaluation method agree well with the experimental one.

In existing severe accident codes such as MELCOR and THALES2, rupture of reactor pressure vessel (RPV) by relocated molten core is judged using simple models such as temperature and/or stress criteria. However, it is difficult to assess rupture behavior of the lower head of RPV in boiling-water-type nuclear power plants due to severe accident like Fukushima Daiichi. One reason is that boiling water reactors (BWRs) have geometrically complicated structure with a lot of penetrations. Another one is that BWR lower head is composed of various types of materials of RPV, weld-overlay cladding, control rod guide tubes, stub tubes, welds, etc. Therefore, we have been developing an analysis method to predict time and location of RPV lower head rupture of BWRs considering creep damage mechanisms based on coupled analysis of three-dimensional thermal-hydraulics (TH) and thermal-elastic-plastic-creep analyses. The detailed three-dimensional model of RPV lower head with control rod guide tubes, stub tubes, and welds are constructed. TH analysis is performed to obtain three-dimensional temperature distribution in relocated debris. Using TH analysis results, structural analysis is carried out to evaluate creep damage distributions using four types of damage criterions of “considere”, strain, Kachanov, and Larson-Miller-parameter (LMP) criteria. Creep damage evaluation based on Kachanov and LMP models is made by using experimentally determined parameters. From comparison of damage criterions, it is shown that failure regions of BWR lower head are only penetrations under simulated conditions, although there is a large difference in failure time.

To examine the effects of minor elements on stress corrosion cracking (SCC) susceptibility of low carbon stainless steels with a work hardened layer, a high purity type 304 stainless steel was fabricated and minor elements, Si, S, P, C or Ti, were added. A work hardened layer was introduced by shaving on the surface of stainless steels. The specimens were exposed to a boiling 42% MgCl 2 solution for 20 h and the number and the length of initiated cracks were examined. SCC susceptibility of the specimen with P was the highest and that of the specimen with C was the lowest in all specimens. By magnetic force microscope examination, a magnetic phase expected to be a martensitic phase was detected near the surface. Since corrosion resistance of martensite is lower than that of austenite, the minor elements additions would affect SCC susceptibility through the amount of the transformed martensite.

The authors previously reported that magnetic flux leakage d ata, measured using a flux gate (FG) sensor, showed a correlation with the irradiation-assisted stress corros ion cracking (IASCC) susceptibility of neutron-irradiate d austenitic stainless alloys. This paper presents a study conducted to develop a diagnostic system that can detect IASCC precursors in stainless steels by measuring the magnetic properties of the material. The eddy current method and alternating current (AC) magnetization method were used, as these will be more practical for use in actual reactors. Probes were developed for these measurement methods, providing sufficient tolerance for en vironments in nuclear reactors. An attempt was also made to improve spatial resolution by manufacturing a smaller probe. A sensor system was designed for remote control, performance tests were conducted by measuring neutron-irradiated specimens and mock specimens, and magnetic data were evaluated by comparing the IASCC susceptibility of the specimens. It was proved that the sensor system developed in this study is capable of detecting IASCC precursors. Further developments necessary for application in actual nuclear reactors and the mechanism of correlation between magnetic properties and IASCC susceptibility were also discussed. Many of the problems experienced in existing light water reactors (LWR) are caused by damage to structural materials. Stress corrosion cracking (SCC) and irradiation-assisted SCC (IASCC) of structural materials have been especially serious problems, and it has not been possible to predict these in the reactor design process. These problems will become more significant when existing reactors start to age from now on, and when advanced reactors are operated and structural materials are used in more severe irradiation environments in future. It is therefore import ant to develop diagnostic techniques that can

Crack initiation and crack growth processes of irradiation assisted stress corrosion cracking of stainless steels were studied by slow strain rate testing (SSRT) in oxygenated high temperature 561 K water. In-situ observation was carried out during SSRT for type 304 stainless steel irradiated to a neutron fluence of 1.0 × 1025 n/m2 (E > 1 MeV) at 323 K in the Japan material testing reactor. The specimens were subjected to solution annealing, thermal sensitization, or cold working prior to neutron irradiation. The solution annealed material exhibited a combination of transgranular stress corrosion cracking (TGSCC) and ductile fracture, and almost all intergranular stress corrosion crackings were observed in the thermally-sensitized material. In the cold-worked material, cracking was introduced before the maximum stress was reached, and the fracture mode changed from TGSCC to ductile fracture to transgranular cracking together with the progress of crack growth in one direction.

It is important to evaluate the effect of irradiation on the corrosion behavior of materials to be used in spallation neutron sources. Solution annealed high purity Fe–18Cr–12Ni specimens were used in this study. Ni3+ and H+ or He2+ ions were injected at 473–773 K. After corrosion test, the specimens were examined with atomic force microscope (AFM) to evaluate the corrosion behavior. It was shown that the corrosion rate of the irradiated area increased with increasing dose and temperature. H implantation accelerated corrosion. On the other hand, He implantation seemed to suppress corrosion. Mechanisms for these effects of the different irradiation conditions on the corrosion behavior are discussed.

Stress-relieved specimens and recrystallized specimens of pure Mo and Mo–Re alloys with Re contents of 2, 4, 5, 10, 13 and 41 wt% were neutron irradiated up to 20 dpa at temperatures from 681 to 1072 K. On microstructural observation, sigma phase and chi phase precipitates were found in all irradiated Mo–Re alloys. Voids were observed in all irradiated specimens, and dislocation loops and dislocations were observed in the specimens that were irradiated at lower temperatures. On Vickers hardness testing, all of the irradiated specimens showed hardening. Especially Mo–41Re were drastically embrittled after irradiation at 874 K or below. From these results, the authors discuss about the relation between microstructure development and radiation hardening and embrittlement, and propose the optimum Re content and thermal treatment for Mo–Re alloys to be used under irradiation conditions.

Type 316L(N) stainless steel of the international thermonuclear experimental reactor grade (316L(N)-IG SS) is being considered for the first wall/blanket module. Hot isostatic pressing (HIP) technique is expected for the fabrication of the module. To evaluate the integrity and susceptibility to stress corrosion cracking (SCC) of HIPed 316L(N)-IG SS, tensile tests in vacuum and slow strain rate tests in high temperature water were performed. Specimen with the HIPed joint had similar tensile properties to specimens of 316L(N)-IG SS, and did not show susceptibility to SCC in oxygenated water at 423 K. Thermally sensitized specimen was low susceptible to SCC even in the creviced condition. It is concluded that the tensile properties of HIPed SS are as high as those of the base alloy and the HIP process caused no deleterious effects.

Tungsten (W) alloys are candidate materials to be used as high-heat-flux materials in fusion reactors. In our previous work, W–26 wt% Re showed drastic hardening and embrittlement after the neutron irradiation. In this study, to clarify the irradiation hardening and embrittlement behavior of W–26 wt% Re, from the viewpoint of microstructural development, the microstructure observation of the neutron irradiated W–26 wt% Re was carried out using transmission electron microscope (TEM). The specimens were irradiated at the materials open test assembly of the fast flux test facility (FFTF/MOTA-2A cycle 11) up to ∼1×10 27 n/m 2 , (En>0.1 MeV). The irradiation temperatures were 646, 679, 792, 873 and 1073 K. In all neutron irradiated W–26 wt% Re samples, sigma-phase precipitates and chi-phase precipitates were observed, while in the thermally aged specimen, only sigma-phase precipitates were observed. Irradiation effects on microstructural development are discussed.

Vanadium alloys are considered as candidate structural materials for fusion reactor system. When vanadium alloys are used in fusion reactor system, joining with ceramics for insulating is one of material issues to be solved to make component of fusion reactor. In the application of ceramics/metal jointing and coating, residual stress caused by difference of thermal expansion rate between ceramics and metals is an important factor in obtaining good bonding strength and soundness of coating. In this work, residual stress distribution in direct diffusion bonded vanadium/alumina joint (jointing temperature: 1400°C) was measured by small area X-ray diffraction method. And the comparison of Finite Element Method (FEM) analysis and actual stress distribution was carried out. Tensile stress concentration at the edge of the boundary of the joint in alumina was observed. The residual stress concentration may cause cracks in alumina, or failure of bonding. Actually, cracks in alumina caused by thermal stress after bonding at 1500°C was observed. The stress concentration of the joint must be reduced to obtain good bonded joint. Lower bonding temperature or to devise the shape of the outer surface of the joint will reduce the stress concentration.