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Shiota, Yuki; Kudo, Junki; Tsuno, Hiromi; Takeuchi, Ryotaro; Ariyoshi, Hideo; Shiohama, Yasutaka; Hamano, Tomoharu; Takagi, Tsuyohiko; Nagaoki, Yoshihiro
JAEA-Technology 2023-002, 87 Pages, 2023/06
JAEA-Technology-2023-002.pdf:8.53MB
In the first stage of Monju decommissioning project, fuel unload work began to be carried out. There are two tasks in this work. One is Fuel Treatment and Storage work that gets rid of sodium on the fuel assemblies unloaded from Ex-Vessel fuel Storage Tank (EVST) and carries it in the storage pool, and the other is Fuel Unloading that the fuel assemblies in the reactor core is replaced with dummy fuels and stored in EVST. Fuel Treatment and Storage work and Fuel Unloading work are performed alternately, and 370 fuel assemblies in the core and 160 fuel assemblies in EVST are all carried to the storage pool. Monju had a large amount of sodium in the reactor vessel and EVST, and there was a residual risk of fuel failure due to the superposition of a large scale sodium fire. Therefore, in the first stage of the Monju decommissioning project, it was decided to take about 5.5 years to remove the residual risk by storing all the fuel rods in the fuel storage pool. There are few Fuel handling system of Sodium Fast Reactor in the world, so the driving record and experience are not enough. So, events that occur even if taken measure are assumed. The following three events apply to this; first, events that are difficult to prevent, events. Second, that are due to lack of experience, and final, events optimization of system is not enough. Plans were taken to suppress these events. This report summarizes the "Monju decommissioning project" work conducted so far in all four campaigns.
Shiota, Yuki; Ariyoshi, Hideo; Shiohama, Yasutaka; Isobe, Yuta; Takeuchi, Ryotaro; Kudo, Junki; Hanaki, Shotaro; Hamano, Tomoharu; Takagi, Tsuyohiko
JAEA-Technology 2022-019, 95 Pages, 2022/09
JAEA-Technology-2022-019.pdf:7.59MB
In the first stage of "Monju" decommissioning project, "Fuel Unloading Operations" have been carrying out. The operations consists of two processes. The first process is "Fuel Treatment and Storage" is that the fuel assemblies unloaded from the Ex-Vessel fuel Storage Tank (EVST) are canned after sodium cleaning, and then transferred to the storage pool. The second process is "Fuel Unloading" that the fuel assemblies in the reactor core are replaced with dummy fuel assemblies and stored in the EVST. "Fuel Treatment and Storage" and "Fuel Unloading" are performed alternately until 370 fuel assemblies in the core and 160 fuel assemblies in the EVST are all transferred to the storage pool. This is a summary of "Fuel Unloading" in the third quarter of "Fuel Unloading Operation". In fiscal 2020, as "Fuel Unloading", 72 fuel assemblies and 74 blanket fuel assemblies were unloaded from the core, and stored in the EVST. From the EVST, 145 dummy fuel assemblies and 1 fixed absorber were loaded in the core instead. During these operations, a total of 36 cases alarming or equipment malfunctions classified into 4 types occurred. However, these events were estimated in advance, there were no significant events that menaces to safety of fuel assemblies and equipment. Therefore, there were no serious problem like fall of fuel assemblies and events that may affect schedule of the project like stick of gripper of ex-vessel fuel transfer machine. When equipment's work or performance fail, the operation continued with safety by elimination of causes of problem. Fuel handling system of Monju has function that is endemic to sodium cooling fast breeding reactor. Because continuous operations of fuel handling system with actual fuel assemblies start recently, we don't have as much experience as PWR and BWR. With estimation of various troubles, reduction of frequency of trouble occurrence and minimization of impacts on schedule performed.
Yokoyama, Sumi*; Tsujimura, Norio; Hashimoto, Makoto; Yoshitomi, Hiroshi; Kato, Masahiro*; Kurosawa, Tadahiro*; Tatsuzaki, Hideo*; Sekiguchi, Hiroshi*; Koguchi, Yasuhiro*; Ono, Koji*; et al.
Journal of Radiation Protection and Research, 47(1), p.1 - 7, 2022/03
Background: In Japan, new regulations that revise the dose limit for the lens of the eye (the lens), operational quantities, and measurement positions for the lens dose were enforced in April 2021. Based on the international safety standards, national guidelines, the results of the Radiation Safety Research Promotion Fund of the Nuclear Regulatory Authority, and other studies, the Working Group of Radiation Protection Standardization Committee, the Japan Health Physics Society (JHPS) developed a guideline for radiation dose monitoring for the lens. Materials and Methods: The Working Group of the JHPS discussed the criteria of non-uniform exposure and the management criteria set to not exceed the dose limit for the lens. Results and Discussion: In July 2020, the JHPS guideline was published. The guideline consists of three parts: main text, explanations, and 26 questions. In the questions, the corresponding answers were prepared, and specific examples were provided to enable similar cases to be addressed. Conclusion: With the development of guideline on radiation dose monitoring of the lens, radiation managers and workers will be able to smoothly comply with revised regulations and optimise radiation protection.
Yokoyama, Sumi*; Iwai, Satoshi*; Tsujimura, Norio; Hashimoto, Makoto; Yoshitomi, Hiroshi; Kato, Masahiro*; Kurosawa, Tadahiro*; Tatsuzaki, Hideo*; Sekiguchi, Hiroshi*; Koguchi, Yasuhiro*; et al.
Proceedings of 15th International Congress of the International Radiation Protection Association (IRPA-15) (Internet), 8 Pages, 2022/00
Hirayama, Hideo*; Nakashima, Hiroshi; Sanami, Toshiya*; Yamaguchi, Yasuhiro; Sato, Osamu*; Takagi, Shunji*; Suzuki, Toshikazu*; Iwai, Satoshi*
Nihon Genshiryoku Gakkai-Shi ATOMO
, 55(2), p.83 - 96, 2013/02
Since the accident at the Fukushima Daiichi Power Plant, units of radiation doses such as Sv have been frequently reported by newspapers. However, it is not exactly understood even by experts on nuclear engineering that there are two different concepts which are the protection quantity such as the effective dose related to a radiation health risk and the practical quantity such as ambient dose equivalents related to measurements. In this special issue, Radiological Protection Quantity, and Operational Quantity and Radiological Protection Quantity, explain the radiation protection dose system including the protection and practical quantities established by ICRP and ICRU. Relation between Dosimeters used at Fukushima etc. and Radiological Protection Quantity describes a method and its problems to measure radiations from Cs on surface and underground using survey meters and personal dosimeters, and convert the measured values to effective doses.
Yamada, Kunimi; Hanamuro, Takahiro; Tagami, Takahiro*; Shimada, Koji; Takagi, Hideo*; Yamada, Ryuji*; Umeda, Koji
Journal of Asian Earth Sciences, 45(2), p.17 - 23, 2012/02
Times Cited Count：5 Percentile：11.19(Geosciences, Multidisciplinary)no abstracts in English
Takagi, Hideo*; Miwa, Shigenori*; Yokomizo, Yoshiyuki*; Nishijima, Kei*; Enjoji, Mamoru*; Mizuno, Takashi; Amano, Kenji
Chishitsugaku Zasshi, 114(7), p.321 - 335, 2008/07
Based on the 3-D orientations of intracrystalline healed, sealed and open extension microcracks in quartz grains in the Late Cretaceous Toki Granite, we discuss the paleostress field and the history of microcracking combining the microthermometry of fluid inclusions in healed microcracks and sealing material identification in sealed microcracks. Twenty one oriented samples are collected mainly from the DH-15 core (240-1000 mabh) drilled by Japan Atomic Energy Agency (JAEA) and additionally from five outcrops in the Tono region. 3-D orientations of healed microcracks indicate the 
orientation of N-S to NW-SE in almost all sites, whereas those of sealed and open microcracks indicate the dominant orientation of E-W. Two or three orthogonal sets of microcracks are common in both healed and sealed microcracks. The formation of these sets can be attributed to the switch of principal stress axis due to stress release just after the crack formation. Healed microcracks probably reflect more regional paleostress field because of consistency of the orientations than the case of sealed and open microcracks that show rather scattering orientations. N-S to NW-SE trending healed microcracks are formed around 60 Ma on the basis of K-Ar biotite ages of the Toki Granite and formation temperature (c. 300-400
C) of fluid inclusions estimated from microthermometry in the case of intrusion depth (3.5 km=100 MPa) of the Toki Granite. Thus the orientation trended NW-SE after the restoration of clockwise rotation of SW Japan at c. 15 Ma. Following the healed microcrack formation, E-W trending high-angle sealed microcracks filled with carbonate mineral are formed. Open microcracks presumably formed at near-surface at the last stage of sealed microcrack formation after c. 20 Ma before the Mizunami Group deposited unconformably on the granite.
Moriyama, Kiyofumi; Takagi, Seiji*; Muramatsu, Ken; Nakamura, Hideo; Maruyama, Yu
JAEA-Research 2007-072, 54 Pages, 2007/11
JAEA-Research-2007-072.pdf:25.89MB
The containment failure probability due to ex-vessel steam explosions was evaluated for BWR and PWR model plants. A stratified Monte Carlo technique (Latin Hypercube Sampling (LHS)) was applied for the evaluation of steam explosion loads, in which a steam explosion simulation code JASMINE was used as a physics model. The evaluation was made for three scenarios: a steam explosion in the pedestal area or in the suppression pool of a BWR model plant with Mark-II containment, and in the reactor cavity of a PWR model plant. The scenario connecting the generation of steam explosion loads and the containment failure was assumed to be displacement of the reactor vessel and pipings, and failure at the penetration in the containment boundary. We evaluated the conditional containment failure probability (CCFP) based on the preconditions of failure of molten core retention within the reactor vessel, relocation of the core melt into the water pool without significant interference, and a strong triggering at the time of maximum premixed mass. The obtained mean and median values of the CCPF were 6.4E-2 (mean) and 3.9E-2 (median) for the BWR suppression pool case, 2.2E-3 (mean) and 2.8E-10 (median) for the BWR pedestal case, and 6.8E-2 (mean) and 1.4E-2 (median) for the PWR cavity case. The evaluation of CCFPs on the basis of core damage needs consideration of probabilities for the above-mentioned preconditions. Thus, the CCFPs per core damage should be lower than the values given above. The specific values of the probability were most dependent on the assumed range of melt flow rate and fragility curve that involved conservatism and uncertainty due to simplified scenarios and limited information. Additionally, the source term significance of the fine particles generated by steam explosions was discussed.
Moriyama, Kiyofumi; Takagi, Seiji*; Muramatsu, Ken; Nakamura, Hideo; Maruyama, Yu*
Journal of Nuclear Science and Technology, 43(7), p.774 - 784, 2006/07
Times Cited Count：17 Percentile：74.61(Nuclear Science & Technology)The containment failure probability due to ex-vessel steam explosions was evaluated for Japanese BWR and PWR model plants. A stratified Monte Carlo technique (Latin Hypercube Sampling (LHS)) was applied for the evaluation of steam explosion loads, in which a steam explosion simulation code JASMINE was used as a physics model. The evaluation was made for three scenarios: a steam explosion in the pedestal area or in the suppression pool of a BWR model plant with a Mark-II containment, and in the reactor cavity of a PWR model plant. The scenario connecting the generation of steam explosion loads and the containment failure was assumed to be displacement of the reactor vessel and pipings, and failure at the penetration in the containment boundary. The mean conditional containment failure probabilities (CCFPs) were 
(mean) and 
(median) for the BWR suppression pool case, 
(mean) and 
(median) for the BWR pedestal case, and 
(mean) and 
(median) for the PWR cavity case. Note that the specific values of the probability are most dependent on assumed range of melt flow rates and on fragility curves that involve conservatism and uncertainty due to simplified scenarios and limited information. Also, note that these CCFPs were based on the preconditions of failure of accident termination within the reactor vessel, relocation of the core melt into the water pool at the place in question without significant interference, and a strong triggering ofa steam explosion with maximized premixed mass for the given premixing condition. The evaluation of CCFPs on the basis of core damage needs consideration of probabilities for these preconditions. Thus, the CCFPs per core damage should be lower than the values given above.
Moriyama, Kiyofumi; Takagi, Seiji; Muramatsu, Ken; Nakamura, Hideo; Maruyama, Yu
Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 9 Pages, 2005/05
The containment failure probability due to ex-vessel steam explosions were evaluated for a BWR Mk-II model plant. The evaluation was made for two scenarios: a steam explosion in the pedestal area, or in the suppression pool. A probabilistic approach, Latin Hypercube Sampling (LHS), was applied for the evaluation of steam explosion loads, in which a steam explosion simulation code JASMINE was used as a physics model. The fragility curves connecting the steam explosion loads and containment failure were developed based on simplified assumptions on the containment failure scenarios. The mean conditional probabilities of containment failure per occurrence of a steam explosion were for suppression pool and for pedestal area. Note that the results depend on the assumed range of input parameters and fragility curves that involve conservatism and simplification.
Miwa, Shigenori*; Takagi, Hideo*; Nishijima, Kei*; Yokomizo, Yoshiyuki*; Enjoji, Mamoru*; Mizuno, Takashi
no journal, ,
no abstracts in English
Yamada, Kunimi; Takagi, Hideo*; Tagami, Takahiro*; Danhara, Toru*
no journal, ,
Taki pseudotachylyte first found along the MTL was formed after mylonitization of the Hatai Tonalite under amphibolite to greenschist facies conditions, and had been post-dated by cataclasis. We dated some zircons from the mylonited Tonalite around the pseudotachylyte using the (U-Th)/He method, the age corresponds approximately to the FT ages of the pseudotachylyte, not the Tonalite. This suggests that the ambient temperature for generating the pseudotachylyte was higher than the closure temperature of ZHe, and this is consistent with cataclasis temperature. In this case we can not say the age of the pseudotachylyte, however, we can constrain something for the age and cooling-denudation history of this fault. Now we continue to date zircons from the pseudotachylyte.
Yamada, Kunimi; Hanamuro, Takahiro; Tagami, Takahiro*; Takagi, Hideo*; Shimada, Koji; Umeda, Koji; Iwano, Hideki*; Danhara, Toru*
no journal, ,
Japan Atomic Energy Agency have jointly set up the system of the (U-Th)/He dating in cooperation with Kyoto University and National Institute for Earth Sciences and Disaster Prevention. Through this two years, helium extraction machine using an electronically-cooled diode laser, micro crusible system using XRF bead sampler and uranium and thorium quantification procedure without artificial spikes are developed. Now we produce new ages of some age known samples. At the present stage, the accuracy and precision are roughly 20 and 10%, respectively, using the age standard zircon from the Fish Canyon Tuff (FC3). In the presentation, we will introduce our systems and the ages detailedly.
Yamada, Kunimi; Hanamuro, Takahiro; Tagami, Takahiro*; Shimada, Koji; Takagi, Hideo*; Iwano, Hideki*; Danhara, Toru*; Umeda, Koji
no journal, ,
no abstracts in English
Yamada, Kunimi; Hanamuro, Takahiro; Tagami, Takahiro*; Shimada, Koji; Takagi, Hideo*; Iwano, Hideki*; Danhara, Toru*; Umeda, Koji
no journal, ,
no abstracts in English
Yamada, Kunimi; Hanamuro, Takahiro; Shimada, Koji; Umeda, Koji; Tagami, Takahiro*; Takagi, Hideo*; Yamada, Ryuji*
no journal, ,
no abstracts in English
Iwamori, Akiyuki*; Takagi, Hideo*; Shimada, Koji; Asahi, Nobutaka*; Sugimori, Tatsuji*; Sasaki, Toshinori*; Aiyama, Kotaro*
no journal, ,
no abstracts in English
Iwamori, Akiyuki*; Ogita, Yasuhiro; Shimada, Koji; Tateishi, Ryo*; Takagi, Hideo*; Ota, Toru*; Kanno, Mizuho*; Wada, Shinya*; Ono, Akihiro*; Otsuka, Yoshiharu*
no journal, ,
We investigated the W value, which is an index showing the progress of weathering, for the fault rocks in the Kojak granite distributed in the eastern Wakasa area. The W value represents the contribution due to weathering calculated from the chemical composition, and along with the M value (contribution of the mafic component) and F value (contribution of the felsic component), a triangular diagram of M + F + W = 100% showing trends from protolith to fault rocks (cataclasite or fault gouge) can be drown. We also investigated on the fault at the geological boundary between the Kojaku granite and the Mino-Tamba metabasalt, and on the difference in characteristics from the fault rock in the Kojaku granite. Granite protolith has an F value of 94.2% and W value of 4.9%, and the fault rock sample has an M value of about 3% regardless of whether it is an active fault or an inactive fault. As weathering progresses, the F value decreases and the W value increases. Metabasalt has an M value of 88.2% and a W value of 6.6%, and the cataclasite has an almost constant F value. As weathering progresses, the M value decreases and the W value increases. Some of the F value increases with the increase of the W value, which is consistent with the contamination of granite-origin quartz fragments found in the basaltic fault gouge. As a result of examination, it was confirmed that Na
O and CaO have a great influence on the increase and decrease of the W value.
Sakai, Toru*; Kametaka, Masao*; Aoki, Kazuhiro; Shimada, Koji; Takagi, Hideo*
no journal, ,
no abstracts in English
