Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Ha, Yoosung; Shimodaira, Masaki; Katsuyama, Jinya
Transactions of 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 7 Pages, 2024/03
Heat-affected zone (HAZ) produced by butt-welding in reactor pressure vessel (RPV) steel is one of the representative materials for surveillance program. The fracture toughness values of HAZ may show a large uncertainty due to inhomogeneous metallurgical structures. Also, the inhomogeneous microstructure in HAZ may influence on the degree of uncertainty in the fracture toughness and the sensitivity to irradiation embrittlement. We investigated the fracture toughness in HAZ of unirradiated material with respect to its distance from the fusion line of welds, where the amount of mixed microstructures change due to the thermal history during the welding. Mini-C(T) specimens of HAZ were harvested from the crack position at 0.5 mm, 1 mm and 2 mm from the fusion line of welds. The uncertainty of fracture toughness in HAZ, from the fusion line at 0.5 mm in particular, was larger than those of base metal at a quarter thickness. From the results of fracture toughness evaluation considering the standard deviation, there was the difference of reference temperature, 
in each position of HAZ. in all positions of HAZ was significantly lower than that of base metal, which means the fracture toughness in HAZ was greater than that of base metal at a quarter thickness.
Taniguchi, Yoshinori; Mihara, Takeshi; Kakiuchi, Kazuo; Udagawa, Yutaka
Annals of Nuclear Energy, 195, p.110144_1 - 110144_11, 2024/01
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Li, F.; Mihara, Takeshi; Udagawa, Yutaka
Journal of Nuclear Science and Technology, 11 Pages, 2024/00
Times Cited Count：0Nanjo, Kotaro; Shiotsu, Hiroyuki; Maruyama, Yu; Sugiyama, Tomoyuki; Okamoto, Koji*
Journal of Nuclear Science and Technology, 60(7), p.816 - 823, 2023/07
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)
fuel with additives under reactivity-initiated accident conditionsMihara, Takeshi; Kakiuchi, Kazuo; Taniguchi, Yoshinori; Udagawa, Yutaka
Journal of Nuclear Science and Technology, 60(5), p.512 - 525, 2023/05
Times Cited Count：1 Percentile：0.01(Nuclear Science & Technology)Kakiuchi, Kazuo; Amaya, Masaki; Udagawa, Yutaka
Journal of Nuclear Materials, 573, p.154110_1 - 154110_7, 2023/01
Times Cited Count：0 Percentile：0.01(Materials Science, Multidisciplinary)Li, F.; Narukawa, Takafumi; Udagawa, Yutaka
Journal of Nuclear Science and Technology, 12 Pages, 2023/00
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Li, F.; Mihara, Takeshi; Udagawa, Yutaka
Journal of Nuclear Science and Technology, 59(12), p.1455 - 1464, 2022/12
Times Cited Count：2 Percentile：31.61(Nuclear Science & Technology)Sun, Haomin; Leblois, Y.*; Gelain, T.*; Porcheron, E.*
Journal of Nuclear Science and Technology, 59(11), p.1356 - 1369, 2022/11
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)In severe accident scenarios of PWR, containment spray can be employed to washout the aerosol of radioactive materials, retaining them in the containment. Therefore, it is crucial to correctly predict the washout efficiency for safety assessment. For a PWR, a high spray coverage ratio (
84%-95%) is required. However, experimental studies on the washout with such a high coverage ratio in a large vessel are quite limited. To understand such a washout phenomenon for model development, aerosol washout experiments are performed in a large vessel with not only aerosol measurements but also spray droplet characterizations. The spray coverage ratios are experimentally confirmed to be compatible with a real PWR. The washout features are investigated in detail. The model in MELCOR is examined using the measured aerosol removal rate, showing the removal rate tendency against particle diameters being reproduced. Although a significant underestimation occurs for large particles, a satisfactory agreement is obtained for smaller ones (
0.52 
m in diameter) corresponding to the minimum removal rate and around.
Okada, Yuji; Magome, Hirokatsu; Matsui, Yoshinori
JAEA-Technology 2022-014, 113 Pages, 2022/09
JAEA-Technology-2022-014.pdf:15.79MB
Material irradiation test system had been newly installed in JMTR (Japan Materials Testing Reactor) with taking 5 years which was from 2008 through 2013. The aim of material irradiation test system is to conduct IASCC (Irradiation Assisted Stress Corrosion Cracking) evaluation study. This system is mainly consist of water control unit, which can simulate elevated temperature and pressure of the light water reactor environment in the reactor, and load control unit, which can perform the crack propagation examination under irradiation. This load control unit gives a load to CT (Compact Tension) specimen, and perform the crack propagation examination. The principle of loading to CT specimen is using pressure difference between pressure generated by high temperature and high pressure water by water control unit in capsule and pressure generated by load gas pressure supplied by helium gas cylinder in bellows installed in load control unit. In 2013, the commissioning of material irradiation test system was carried out for adjustment. During this commissioning, the correlation between the differential pressure in load control unit and the load was confirmed by using the test container connected to load control unit with load cell. From the results of commissioning, the problem, which the load change speeds at loading and unloading were different due to different pressure change speeds by the piping resistance performance in the periodic loading test in which load from minimum to maximum repeatedly applied, was confirmed. This report summarizes the problem of load change speed due to the piping resistance performance, which was confirmed in 2013, the improvement and performance test of load control unit for solving the problem described above, which were carried out from 2014 to 2015, and operating procedure.
Yamaguchi, Akinori*; Yokotsuka, Muneyuki*; Furuta, Masayo*; Kubota, Kazuo*; Fujine, Sachio*; Mori, Kenji*; Yoshida, Naoki; Amano, Yuki; Abe, Hitoshi
Nihon Genshiryoku Gakkai Wabun Rombunshi (Internet), 21(4), p.173 - 182, 2022/09
Risk information obtained from probabilistic risk assessment (PRA) can be used to evaluate the effectiveness of measures against severe accidents in nuclear facilities. The PRA methods used for reprocessing facilities are considered immature compared to those for nuclear power plants, and to make the methods mature, reducing the uncertainty of accident scenarios becomes crucial. In this paper, we summarized the results of literature survey on the event progression of evaporation to dryness caused by boiling of high-level liquid waste (HLLW) which is a severe accident in reprocessing facilities and migration behavior of associated radioactive materials. Since one of the important characteristics of Ru is its tendency to form volatile compounds over the course of the event progression, the migration behavior of Ru is categorized into four stages based on temperature. Although no Ru has been released in the waste in the high temperature region, other volatile elements such as Cs could be released. Sufficient experimental data, however, have not been obtained yet. It is, therefore, necessary to further clarify the migration behavior of radioactive materials that predominantly depends on temperature in this region.
Kakiuchi, Kazuo; Udagawa, Yutaka; Yamauchi, Akihiro*
JAEA-Research 2022-001, 21 Pages, 2022/06
JAEA-Research-2022-001.pdf:1.84MB
The primary cause of cladding embrittlement during loss-of-cool ant accident (LOCA) is the increase in oxygen concentration in the metallic layer and associated microstructural change due to oxidation. In the case of cladding high temperature rupture, inner surface oxidation by the steam ingress and the consequent increase in hydrogen partial pressure result in hydrogen absorption (secondary hydriding) localized in the axial direction at the distance apart from the rupture opening as is well known from preceding studies. In order to understand the effect of cladding microstructural changes on mechanical property of a fuel rod under LOCA conditions in a more precise and quantitative manner, the nanoindentation method has been applied to evaluation of mechanical properties of a cladding specimen after a LOCA simulated test; results for two samples taken from the rupture opening part and secondary hydriding part were compared with each other. The fraction of plastic work during the indentation was evaluated from the load-displacement curve in addition to hardness and Young's modulus. The plastic work fraction at the secondary hydriding part was found to be clearly lower than that at the rupture opening part and rather close to that in the ZrO and 
-Zr(O) layers, suggesting the significant ductility reduction of the secondary hydriding part despite its relatively low oxygen concentration.
Kakiuchi, Kazuo; Amaya, Masaki; Udagawa, Yutaka
Annals of Nuclear Energy, 171, p.109004_1 - 109004_9, 2022/06
Times Cited Count：4 Percentile：78.52(Nuclear Science & Technology)Shimada, Taro; Nemoto, Hiromi*; Takeda, Seiji
Hoken Butsuri (Internet), 57(1), p.5 - 29, 2022/03
Of the asbestos-containing wastes arising from the dismantling activities of nuclear facilities, those with radioactive concentrations that do not need to be treated as radioactive substances will be cleared from the nuclear regulatory control. Those will be disposed of or recycled as specially controlled industrial waste based on the Waste Management and Public Cleansing Act. The authors constructed evaluation scenarios according to the treatment manual for asbestos-containing waste and evaluated public exposure doses per year for 33 radionuclides. Based on the evaluated doses, the radioactive concentration corresponding to the dose criteria of 10 Sv/y for clearance was calculated for each radionuclide and scenario. As a result, the evaluated concentration was equal to or higher than the current clearance level. It was confirmed that the application of the current clearance level for asbestos-containing wastes did not affect safety.
Takamizawa, Hisashi; Hata, Kuniki; Nishiyama, Yutaka; Toyama, Takeshi*; Nagai, Yasuyoshi*
Journal of Nuclear Materials, 556, p.153203_1 - 153203_10, 2021/12
Times Cited Count：3 Percentile：31.78(Materials Science, Multidisciplinary)Solute clusters (SCs) formed in pressurized water reactor surveillance test specimens neutron-irradiated to a fluence of 1 
10
n/cm
were analyzed via atom probe tomography to understand the effect of silicon on solute clustering and irradiation embrittlement of reactor pressure vessel steels. In high-Cu bearing materials, Cu atoms were aggregated at the center of cluster surrounded by the Ni, Mn, and Si atoms like a core-shell structure. In low-Cu bearing materials, Ni, Mn, and Si atoms formed cluster and these solutes were not comprised core-shell structure in SCs. While the number of Cu atoms in clusters was decreased with decreasing nominal Cu content, the number of Si atoms had clearly increased. The cluster radius (
) and number density (
) decreased and increased, respectively, with increasing nominal Si content. The shift in the reference temperature for nil-ductility transition (
RT
) showed a good correlation with the square root of volume fraction (
) multiplied by r (
). This suggested that the dislocation cutting through the particles mechanism dominates the precipitation hardening responsible for irradiation embrittlement. The negative relation between the nominal Si content and RT indicated that increasing of nominal Si content reduces the degree of embrittlement.
Takamizawa, Hisashi; Nishiyama, Yutaka
Journal of Pressure Vessel Technology, 143(5), p.051502_1 - 051502_8, 2021/10
Times Cited Count：3 Percentile：30.36(Engineering, Mechanical)no abstracts in English
Narukawa, Takafumi; Udagawa, Yutaka
Proceedings of TopFuel 2021 (Internet), 10 Pages, 2021/10
Sun, Haomin; Shibamoto, Yasuteru; Hirose, Yoshiyasu; Kukita, Yutaka
Journal of Nuclear Science and Technology, 58(9), p.1048 - 1057, 2021/09
Times Cited Count：4 Percentile：56.94(Nuclear Science & Technology)no abstracts in English
Kakiuchi, Kazuo; Udagawa, Yutaka; Amaya, Masaki
Annals of Nuclear Energy, 155, p.108171_1 - 108171_11, 2021/06
Times Cited Count：1 Percentile：16.35(Nuclear Science & Technology)Tanimura, Yoshihiko; Yoshitomi, Hiroshi; Nishino, Sho; Takahashi, Masa
Radiation Measurements, 137, p.106389_1 - 106389_5, 2020/09
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)A portable thyroid dose monitoring system has been developed at the Japan Atomic Energy Agency (JAEA) to assess the thyroid equivalent dose for workers and members of the public in a high dose rate environment. The background (B.G.) photon correction is required for an accurate measurement in a high dose rate environment at an early stage after a nuclear accident. We developed the B.G. photon correction method using cylindrical PMMA phantoms.
