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Journal Articles

Analytical study of SPERT-CDC test 859 using fuel performance codes FEMAXI-8 and RANNS

Taniguchi, Yoshinori; Udagawa, Yutaka; Amaya, Masaki

Annals of Nuclear Energy, 139, p.107188_1 - 107188_7, 2020/05

 Times Cited Count:0

Journal Articles

The Effect of base irradiation on failure behaviors of UO$$_{2}$$ and chromia-alumina additive fuels under simulated reactivity-initiated accidents; A Comparative analysis with FEMAXI-8

Udagawa, Yutaka; Mihara, Takeshi; Taniguchi, Yoshinori; Kakiuchi, Kazuo; Amaya, Masaki

Annals of Nuclear Energy, 139, p.107268_1 - 107268_9, 2020/05

 Times Cited Count:0

Journal Articles

Improvements on evaluation functions of a probabilistic fracture mechanics analysis code for reactor pressure vessels

Lu, K.; Katsuyama, Jinya; Li, Y.

Journal of Pressure Vessel Technology, 142(2), p.021208_1 - 021208_11, 2020/04

JAEA Reports

User's manual for the OSCAAR code package

Radiation Risk Analysis Research Group, Risk Analysis and Applications Research Division

JAEA-Testing 2020-001, 65 Pages, 2020/03


The Nuclear Safety Research Center (NSRC) has developed a Level 3 Probabilistic Risk Assessment (PRA) code, the OSCAAR (Off-Site Consequence Analysis code for Atmospheric Release in Reactor Accident). Using a source-term derived from a severe accident analysis code, OSCAAR calculates the downwind transport, airborne dispersion, and ground deposition of radioactive materials as well as the radiation dose and then can evaluate the effect of early and late protective actions and health effect. The NSRC also developed the OSCAAR Code Package that has a Windows-based interface and framework for performing consequence analysis in order to create or modify input files, execute the OSCAAR and postprocess the output files efficiently and conveniently. The report is a user's manual for the OSCAAR Code Package.

JAEA Reports

Development of database of dose reduction factor for reducing equivalent doses to the thyroid by implementing iodine thyroid blocking (Contract research)

Kimura, Masanori; Munakata, Masahiro; Hato, Shinji*; Kanno, Mitsuhiro*

JAEA-Data/Code 2020-002, 38 Pages, 2020/03


To consider the method of implementing urgent protective actions in a nuclear accident appropriately, the authors have been assessing the effects of reducing doses by taking urgent protective actions using a Level 3 Probabilistic Risk Assessment (PRA) code, the OSCAAR, developed by the JAEA. Iodine thyroid blocking is an effective urgent protective action to reduce equivalent doses to the thyroid due to inhalation of radioactive iodine. However, the timing of the administration of stable iodine (SI) is important to maximize the effectiveness for thyroidal blocking. Therefore, the careful consideration should be given to the most effective way of iodine thyroid blocking when preparing off-site emergency plans. In the present study, the authors developed a new metabolic model for thyroid by combining the respiratory tract model (Publ.66) and gastrointestinal tract model (Publ.30) of the ICRP with a metabolic model for thyroid (Johnson's model) in order to calculate the behavior of radioiodine and stable iodine in the body more realistically. The model is useful to evaluate the effect of the administration of SI for reducing equivalent doses to the thyroid depending on the its timing. We also calculated the reduction factor for equivalent doses to the thyroid in order to the thyroid by using the model, and then develped the its database for the OSCAAR. Consequecly, the OSCAAR can evaluate the effectiveness for thyroidal blocking realistically and promptly.

Journal Articles

Experimental and analytical investigation of formation and cooling phenomena in high temperature debris bed

Hotta, Akitoshi*; Akiba, Miyuki*; Morita, Akinobu*; Konovalenko, A.*; Vilanueva, W.*; Bechta, S.*; Komlev, A.*; Thakre, S.*; Hoseyni, S. M.*; Sk$"o$ld, P.*; et al.

Journal of Nuclear Science and Technology, 57(3-4), p.353 - 369, 2020/03

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Journal Articles

Irradiation growth behavior of improved alloys for fuel cladding

Kakiuchi, Kazuo; Amaya, Masaki

Nippon Genshiryoku Gakkai Wabun Rombunshi, 19(1), p.24 - 33, 2020/03

The irradiation growth behavior of the improved Zr alloys for light-water reactor fuel cladding was investigated. The coupon specimens, which were prepared from fuel cladding tubes with improved Zr alloys, had been irradiated in the Halden reactor in Norway at temperatures of 300 and 320$$^{circ}$$C under a typical water chemistry condition of PWR and 240$$^{circ}$$C under the coolant condition of the Halden reactor up to a fast neutron fluence of $$sim$$ 8$$times$$10$$^{21}$$ (n/cm$$^{2}$$, E $$>$$1 MeV). During and after the irradiation test, the amount of irradiation growth of each specimen was evaluated. The effect of the difference in alloy composition on irradiation growth behavior seemed insignificant if the other conditions e.g. the final heat treatment condition at fabrication, irradiation temperature and the amount of hydrogen pre-charged in the specimen were the same.

Journal Articles

Effects of oxidation and secondary hydriding during simulated Loss-Of-Coolant-Accident tests on the bending strength of Zircaloy-4 fuel cladding tube

Okada, Yuji; Amaya, Masaki

Annals of Nuclear Energy, 136, p.107028_1 - 107028_9, 2020/02

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Journal Articles

Evaluation of the effects of differences in building models on the seismic response of a nuclear power plant structure

Choi, B.; Nishida, Akemi; Muramatsu, Ken*; Takada, Tsuyoshi*

Nippon Jishin Kogakkai Rombunshu (Internet), 20(2), p.2_1 - 2_16, 2020/02

no abstracts in English

Journal Articles

A New probabilistic evaluation model for weld residual stress

Mano, Akihiro; Katsuyama, Jinya; Miyamoto, Yuhei*; Yamaguchi, Yoshihito; Li, Y.

International Journal of Pressure Vessels and Piping, 179, p.103945_1 - 103945_6, 2020/01

Weld residual stress (WRS) is one of the most important factors in the structural integrity assessment of piping welds, and it is considered a driving force for crack growth. It is characterized by large uncertainty. For more rational assessment, it is important to consider the uncertainty of WRS for evaluating crack growth behavior in probabilistic fracture mechanics (PFM) analysis. In existing PFM analysis codes, WRS uncertainty is set by statistically processing the results of multiple finite element analyses. This process depends on the individual performing PFM analysis, which may lead to uncertainties whose sources would be different from the original WRS. In this study, we developed a new WRS evaluation model based on Fourier transformation, and the model was incorporated into PASCAL-SP, which has been developed by Japan Atomic Energy Agency. Through improvements to the code, WRS uncertainty can be considered automatically and appropriately by inputting multiple WRS analysis results directly as input data for PFM analysis.

Journal Articles

Fracture limit of high-burnup advanced fuel cladding tubes under loss-of-coolant accident conditions

Narukawa, Takafumi; Amaya, Masaki

Journal of Nuclear Science and Technology, 57(1), p.68 - 78, 2020/01

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Journal Articles

Liquid film behavior and heat-transfer mechanism near the rewetting front in a single rod air-water system

Wada, Yuki; Le, T. D.; Satou, Akira; Shibamoto, Yasuteru; Yonomoto, Taisuke

Journal of Nuclear Science and Technology, 57(1), p.100 - 113, 2020/01

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Journal Articles

Rapid clogging of high-efficiency particulate air filters during in-cell solvent fires at reprocessing facilities

Ono, Takuya; Tashiro, Shinsuke; Amano, Yuki; Yoshida, Ryoichiro; Abe, Hitoshi

Nuclear Technology, 206(1), p.40 - 47, 2020/01

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Recent Japanese nuclear regulations have focused on the hazards of in-cell solvent fires at reprocessing facilities. In this work, a mixture of tributyl phosphate and dodecane-based solvents was burned to generate an aerosol composed of soot and unburned solvent that was then loaded onto a high-efficiency particulate air filter simulating the ventilation system of reprocessing facilities. A radical increase of differential pressure occurred in the filters during these tests after the dodecane burned out from the solvent in a phenomenon we named as rapid clogging, likely caused by the burnout of dodecane. This relationship provides valuable insight into the establishment of new regulations for reprocessing facilities. This work indicates that clogging of ventilation filters during solvent fires may occur more rapidly than previously estimated.

Journal Articles

Development of an external radiation dose estimation model for children returning to their homes in areas affected by the Fukushima Nuclear Accident

Mori, Airi; Takahara, Shogo; Yoshida, Hiroko*; Sanada, Yukihisa; Munakata, Masahiro

Health Physics, 117(6), p.606 - 617, 2019/12

 Times Cited Count:0 Percentile:100(Environmental Sciences)

Journal Articles

Experimental study on local interfacial parameters in upward air-water bubbly flow in a vertical 6$$times$$6 rod bundle

Han, X.*; Shen, X.*; Yamamoto, Toshihiro*; Nakajima, Ken*; Sun, Haomin; Hibiki, Takashi*

International Journal of Heat and Mass Transfer, 144, p.118696_1 - 118696_19, 2019/12

 Times Cited Count:0 Percentile:100(Thermodynamics)

Journal Articles

Study on dryout and rewetting during accidents including ATWS for the BWR at JAEA

Satou, Akira; Wada, Yuki; Shibamoto, Yasuteru; Yonomoto, Taisuke

Nuclear Engineering and Design, 354, p.110164_1 - 110164_10, 2019/12

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

JAEA has conducted a series of experimental researches on the Post-boiling transition heat transfer, transient critical heat flux and rewetting for BWRs. Experimental data bases covering the anticipated operational conditions was developed; the significance of the precursor cooling was identified. This paper presents approaches of the present research focusing on the anticipated transient without scram, effects of the spacer and physical understanding of the phenomena for development of mechanistic model together with promising results obtained so far.

Journal Articles

Ion-induced irradiation hardening of the weld heat-affected zone in low alloy steel

Ha, Yoosung; Takamizawa, Hisashi; Katsuyama, Jinya; Hanawa, Satoshi; Nishiyama, Yutaka

Nuclear Instruments and Methods in Physics Research B, 461, p.276 - 282, 2019/12

 Times Cited Count:0 Percentile:100(Instruments & Instrumentation)

Journal Articles

The Effect of air fraction in steam on the embrittlement of Zry-4 fuel cladding oxidized at 1273-1573 K

Negyesi, M.; Amaya, Masaki

Oxidation of Metals, 92(5-6), p.439 - 455, 2019/12

 Times Cited Count:0 Percentile:100(Metallurgy & Metallurgical Engineering)

JAEA Reports

Background radiation monitoring using manned helicopter for application of technique of nuclear emergency response in the fiscal year 2018 (Contract research)

Futemma, Akira; Sanada, Yukihisa; Komiya, Tomokazu; Iwai, Takeyuki*; Seguchi, Eisaku*; Matsunaga, Yuki*; Kawabata, Tomoki*; Haginoya, Masashi*; Hiraga, Shogo*; Sato, Kazuhiko*; et al.

JAEA-Technology 2019-017, 95 Pages, 2019/11


By the nuclear disaster of Fukushima Daiichi Nuclear Power Station (FDNPS), Tokyo Electric Power Company (TEPCO), caused by the Great East Japan Earthquake and the following tsunami on March 11, 2011, a large amount of radioactive material was released from the FDNPS. After the nuclear disaster, airborne radiation monitoring using manned helicopter was conducted around FDNPS. We have carried out the background radiation monitoring around the nuclear power stations of the whole country to apply the airborne radiation monitoring technique that has been cultivated in the aerial monitoring around FDNPS against nuclear emergency response. The results of monitoring around Shimane and Hamaoka Nuclear Power Stations in the fiscal 2018 were summarized in this report. In addition, technical issues were described.

JAEA Reports

Progress report on Nuclear Safety Research Center (JFY 2018)

Nuclear Safety Research Center, Sector of Nuclear Safety Research and Emergency Preparedness

JAEA-Review 2019-015, 147 Pages, 2019/11


Nuclear Safety Research Center (NSRC), Sector of Nuclear Safety Research and Emergency Preparedness, Japan Atomic Energy Agency (JAEA) is conducting technical support to nuclear safety regulation and safety research based on the Mid-Long Term Target determined by Japanese government. This report summarizes the research structure of NSRC and the cooperative research activities with domestic and international organizations as well as the nuclear safety research activities and results in JFY 2018 on the nine research fields in NSRC; (1) severe accident analysis, (2) radiation risk analysis, (3) safety of nuclear fuels in light water reactors (LWRs), (4) thermohydraulic behavior under severe accident in LWRs, (5) materials degradation and structural integrity, (6) safety of nuclear fuel cycle facilities, (7) safety management on criticality, (8) nuclear safeguards, and (9) safety of radioactive waste management.

3197 (Records 1-20 displayed on this page)