Journal of Nuclear Science and Technology, 56(9-10), p.831 - 841, 2019/09
The insoluble Cs particles (Type A) were firstly observed in Tsukuba-city on the morning of March 15. The particles have been considered to be generated in RPV of Unit 2 by evaporation/condensation based on the measured Cs/Cs ratio and the core temperatures of each unit. However, the Type A particles with smaller diameter than the Type B particles of Unit 1 origin, are covered by almost pure silicate glass and have a trace of the quenching. This indicates that the particles could have been generated due to the melting of the HEPA filter in SGTS by the fire of H detonation at Unit 3, and atomization followed by quenching of the molten materials by air blast of the explosion. Although the particles were mostly dispersed to the sea because of the wind direction, some of them deposited onto the lower elevation of R/B at Unit 3, could have been subsequently re-suspended and released into the environment, by the steam flow in the R/B caused by restart of the Unit 3 core cooling water injection at 2:30 of March 15.
Takahashi, Sentaro*; Kawashima, Shigeto*; Hidaka, Akihide; Tanaka, Sota*; Takahashi, Tomoyuki*
Nuclear Technology, 205(5), p.646 - 654, 2019/05
Hidaka, Akihide; Himi, Masashi*; Addad, Y.*
Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR 2019) (Internet), 4 Pages, 2019/05
no abstracts in English
Genshiryoku No Ima To Ashita, p.264 - 265, 2019/03
no abstracts in English
Hidaka, Akihide; Yokoyama, Hiroya
Proceedings of Symposium on Water Chemistry and Corrosion in Nuclear Power Plants in Asia 2017 (AWC 2017) (USB Flash Drive), p.29 - 42, 2017/09
no abstracts in English
Hidaka, Akihide; Yokoyama, Hiroya
Journal of Nuclear Science and Technology, 54(8), p.819 - 829, 2017/08
To clarify what happened during the Fukushima accident, the phenomena within RPV and the discussion of ties with the environmental monitoring are very important. However, the previous study has not necessarily advanced until the present that passed almost six years from the accident. The present study investigated I and Cs release behaviors during the late phase of the accident based on I/Cs ratio of the source terms that were recently evaluated backward by WSPEEDI code based on environmental monitoring data. The I release from the contaminated water in the basement of 1F2 and 1F3 reactor buildings was evaluated to be about 10% of I source term. The increase in Cs release from March 21 to 23 and from March 30 to 31 could be explained by the release of CsBO which is formed as a result of chemical reactions of Cs with BC due to re-ascension of the core temperature caused by slight shortage of the core cooling water.
Hidaka, Akihide; Yokoyama, Hiroya
Journal of Nuclear Science and Technology, 54(8), P. i, 2017/08
no abstracts in English
Yamaguchi, Mika; Hidaka, Akihide; Ikuta, Yuko; Murakami, Kenta*; Tomita, Akira*; Hirose, Hiroya*; Watanebe, Masanori*; Ueda, Kinichi*; Namaizawa, Ken*; Onose, Takatoshi*; et al.
JAEA-Review 2017-002, 60 Pages, 2017/03
Since 2010, IAEA has held the NEM School to develop future leaders who plan and manage nuclear energy utilization in their county. Since 2012, JAEA together with Japan Nuclear HRD Network, University of Tokyo, Japan Atomic Industrial Forum and JAIF International Cooperation Center have cohosted the school in Japan in cooperation with IAEA. Since then, the school has been held in Japan every year. In 2006, Japanese nuclear technology and experience, such as lessons learned from the Fukushima Daiichi Nuclear Power Plant accident, were provided to offer a unique opportunity for the participants to learn about particular cases in Japan. Through the school, we contributed to the internationalization of Japanese young nuclear professionals, development of nuclear human resource of other countries including nuclear newcomers, and enhanced cooperative relationship with IAEA. Additionally, collaborative relationship within the network was strengthened by organizing the school in Japan.
Hidaka, Akihide; Nakano, Yoshihiro; Watanabe, Yoko; Arai, Nobuyoshi; Sawada, Makoto; Kanaizuka, Seiichi*; Katogi, Aki; Shimada, Mayuka*; Ishikawa, Tomomi*; Ebine, Masako*; et al.
JAEA-Review 2016-011, 208 Pages, 2016/07
JAEA has been conducting the Instructor Training Program (ITP) since 1996 under the auspices of MEXT to contribute to human resource development in currently 11 Asian countries in the field of radiation utilization for seeking peaceful use of nuclear energy. ITP consists of Instructor Training Course (ITC), Follow-up Training Course (FTC) and Nuclear Technology Seminars. In the ITP, trainings or seminars relating to technology for nuclear utilization are held in Japan by inviting nuclear related people from Asian countries to Japan and after that, the past trainees are supported during FTC by dispatching Japanese specialists to Asian countries. News Letter is also prepared to provide the broad range of information obtained through the trainings for local people near NPPs in Japan. The present report describes the activities of FY2014 ITP and future challenges for improving ITP more effectively.
Enerugi Rebyu, 35(9), p.20 - 24, 2015/09
Operation of nuclear power plant causes accumulation of radionuclides in fuel rods as a result of nuclear fission of uranium and plutonium. During severe accidents, large amount of radionuclides are released from fuel and transport in the reactor coolant system and/or the containment. When the containment fails or its confinement function is lost, radionuclides could be released into the environment. Meanwhile, radionuclides can be removed by condensation onto wall, natural deposition such as gravitational settling, the engineered safety features (ESF) such as containment spray and so on. After various processes described above, the species, amounts and timing of radionuclide released into the environment is called source terms. The behavior of radionuclide can be described mechanistically by condensation or evaporation of gaseous radionuclide, deposition, growth and removal of aerosol by ESF. Present paper summarizes the radionuclide behavior during severe accidents.
Suehiro, Shoichi*; Sugimoto, Jun*; Hidaka, Akihide; Okada, Hidetoshi*; Mizokami, Shinya*; Okamoto, Koji*
Nuclear Engineering and Design, 286, p.163 - 174, 2015/05
The severe accident evaluation committee of AESJ (Atomic Energy Society of Japan) developed the thermal hydraulic PIRT (Phenomena Identification and Ranking Table) and the source term PIRT based on findings during the Fukushima Daiichi NPPs accident. These PIRTs aimed to explore the debris distribution and the current condition in the NPPs with high accuracy and to extract higher priority from the aspect of the sophistication of the analytical technology to predict the severe accident phenomena by the code. The ST PIRT was divided into 3 phases for the time domain and 9 categories for the spatial domain. The 68 phenomena were extracted and the importance from viewpoint of the source term was ranked through brainstorming and discussion. This paper described the developed ST PIRT list and summarized the high ranked phenomena in each phase.
Hidaka, Akihide; Nakamura, Kazuyuki; Watanabe, Yoko; Yabuuchi, Yukiko; Arai, Nobuyoshi; Sawada, Makoto; Yamashita, Kiyonobu; Sawai, Tomotsugu; Murakami, Hiroyuki
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 9 Pages, 2015/05
Takahara, Shogo; Hidaka, Akihide; Ogino, Takashi*
JAEA-Data/Code 2015-001, 65 Pages, 2015/03
HEINPUT code is one of the preprocessor for probabilistic accident consequence assessment code OSCAAR, and estimates the risk of incidence and death due to radiation induced cancers. HEINPUT code currently uses two models developed by U.S. Nuclear Regulatory Commission (USNRC, 1985; 1993) and U.S. Environmental Protection Agency (EPA, 1994). In this report, the code was improved to enable to make calculation using the new EPA model. In addition, in order to reduce user's burden, we developed input data generator which can provide the input data for three estimation models implemented in HEINPUT-GUI based on the statistical information published.
Nippon Genshiryoku Gakkai Wabun Rombunshi, 14(1), p.51 - 61, 2015/03
BC used mainly for BWR and EPR absorbers could cause phenomena which never happen in PWR with Ag-In-Cd absorbers during severe accident. BC would make a eutectic interaction with stainless steel and enhance melt relocation. Boron oxidation could increase H generation and change of liberated carbon to CH could enhance CHI generation. HBO generated during BC oxidation could be changed to CsBO by combining with Cs. This may increase Cs deposition in reactor coolant system. There could be differences in configuration, surface area, stainless steel-BC weight ratio between BC powder and pellet absorbers. Present issue is to clarify effect of these differences on full scale melt progression, BC oxidation and source term. Advancement of this research domain could contribute to further sophistication of prediction tool for melt progression and source terms, and treatment of organic iodide formation in safety evaluation.
Proceedings of 2014 Nuclear Plant Chemistry Conference (NPC 2014) (USB Flash Drive), 12 Pages, 2014/10
During core cooling at Fukushima Daiichi NPP accident, large amount of contaminated water was accumulated in the basements of reactor buildings at Units 1 to 4. The estimated ratios of I-131 and Cs-137 quantities in water to the core inventories are 0.51%, 0.85% at Unit 1, 74%, 38% at Unit 2 and 26%, 18% at Unit 3, respectively. According to the Henry's law, certain fraction of iodine in water could be released to atmosphere. Many evaluations for I-131 release have been performed so far by MELCOR or the reverse estimation with SPEEDI. The SPEEDI reverse predicted significant release until March 26 while no prediction in MELCOR after March 17. The present study showed that iodine release from accumulated water due to radiolytic conversion from I to I and gas-liquid partition of I may explain the release between March 17 and 26. This strongly suggests a need for improvement of current MELCOR approach which treats the release only from containment breaks.
Hidaka, Akihide; Ishikawa, Jun
Journal of Nuclear Science and Technology, 51(4), p.413 - 424, 2014/04
During the Fukushima Daiichi Nuclear Power Plants accident, large amount of contaminated water was accumulated in the basements of the reactor buildings at Units 1 to 4. The present study estimated the quantities of I-131 and Cs-137 in the water as of late March based on the press-opened data by NISA. The estimated ratios of I-131 and Cs-137 quantities to the core inventories are 0.51%, 0.85% at Unit 1, 74% and 38% at Unit 2, 26% and 18% at Unit 3, respectively. According to the Henry's law, certain fraction of radionuclides in the water could be released to atmosphere due to the gas-liquid partition and contribute to increase in the source terms. Iodine-131 source term evaluation has been performed so far by the MELCOR calculation or the SPEEDI reverse estimation. The SPEEDI reverse predicted significant release until around March 26 while no prediction in MELCOR. A simplified model study showed that the release from accumulated water may explain the release after Mar.17. This strongly suggests a need for improvement of the current MELCOR approach which treats the release only from breaks of drywell or S/C for several days after the core melt.
ERI/NRC 11-211, p.60_69 - 111_117, 2011/12
NRC recently prepared the draft of revised NUREG-1465 for high burnup or MOX fuel and is peer-reviewing it by US, French specialists including myself. In the draft, the containment source terms were evaluated by the MELCOR code with implementation of the Booth model adjusted based on the IRSN's VERCORS test results. However, only the diffusion coefficient for Cs was changed and the original class scale factor of the ORNL-Booth model was used and therefore there was no large difference in the results between the newly revised and existing NUREG-1465. It was proposed that the class scale factor be reevaluated so that the final release rates of radionuclide other than Cs in the VERCORS tests might correspond to the calculation by the Booth model with change of only the diffusion coefficient for Cs.
Journal of Nuclear Science and Technology, 48(1), p.85 - 102, 2011/01
In VEGA program on radionuclide release from irradiated fuel under severe accident condition, totally 10 tests were performed at JAEA from 1999 to 2004 under inert and steam atmospheres including the highest pressure or temperature conditions. These tests showed that increase in release rate at high temperature around UO melting point and decrease in release rate under elevated pressure that was a first observation in the world. The mechanism of pressure effect on release was examined and the release model with pressure effect was proposed. The data on low-volatile radionuclide release, release from MOX fuel, the effect of fuel oxidation and eutectic reaction with cladding on release were obtained through the tests. In addition, the effect of obtained results on source term evaluation and effectiveness of the accident management measures were investigated. This paper summarizes the above described outcomes, limitations of VEGA program, and future issues.
NEA/CSNI/R(2010)10/PART1 (Internet), 12 Pages, 2010/12
In VEGA program on FP release from fuel during severe accidents, 10 tests were conducted under the highest pressure and/or temperature conditions. Tests with PWR fuel at 1.0MPa showed first that Cs release was suppressed by about 30% compared with that at 0.1MPa. This was reproduced by 2-stage diffusion model in UO grains & pores, and a simplified 1/P**0.5 CORSOR-M model. In BWR and MOX fuel tests, however, this effect was not observed clearly due to higher fuel temperature during normal operation and differences in test conditions. The pressure effect may affect PWR source terms and AM measures such as intentional depressurization. Analyses with THALES-2 suggested that the depressurization has many advantages such as delay in accident progression and mitigation of source terms at early CV failure despite increase in FP release into RCS. The effect of pressure on consequences needs to be evaluated systematically for various accident sequences with AM measures.
Nippon Genshiryoku Gakkai "Shibiaakushidentoji No Kakuno Yokinai No Genjitsuteki Sosutamu Hyoka" Tokubetsu Senmon Iinkai Hokokusho, p.3.1_1 - 3.1_38, 2010/04
In VEGA program on radionuclide release from fuel under severe accident condition, totally 10 tests were performed under inert and steam atmospheres including the highest pressure or temperature conditions from 1999 to 2004. These tests showed that increase in release rate at high temperature around UO melting point and decrease in release rate under elevated pressure that was a first observation in the world. The data on low-volatile radionuclide release, release from MOX fuel, the effect of fuel oxidation and eutectic reaction with cladding on radionuclide release were obtained through the tests. In addition, the effect of obtained results on the source term evaluation and effectiveness of the accident management measures were examined. This manuscript summarizes the above described outcomes of VEGA program that have been already submitted to academic societies or international conferences.