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Aoya, Juri; Mori, Amami; Sato, Hinata; Kono, Soma; Morokado, Shiori; Horigome, Kazushi; Goto, Yuichi; Yamamoto, Masahiko; Taguchi, Shigeo
JAEA-Technology 2023-008, 34 Pages, 2023/06
JAEA-Technology-2023-008.pdf:1.92MB
Flush-out, by which nuclear materials in the Tokai Reprocessing Plant process are recovered, has been started in June 2022 as the first step of decommissioning. Flush-out consists of removal of spent fuel sheared powder, plutonium solution, uranium solution, and the other nuclear materials. Removal of spent fuel sheared powder has been completed in September 2022. During removal of spent fuel sheared powder, uranium concentration, plutonium concentration, acid concentration, radioactivity concentration, and solution density have been analyzed for process control. For nuclear material accountancy, uranium concentration, plutonium concentration, isotope ratio, and solution density have been analyzed. Analysis work including sample pretreatment before transportation to IAEA analytical facility for safeguards, and the other operations related to Flush-out such as calibration of analytical instruments, education, and training of operators are reported.
for production of 
Tc radioactive isotopeSuematsu, Hisayuki*; Sato, Soma*; Nakayama, Tadachika*; Suzuki, Tatsuya*; Niihara, Koichi*; Nanko, Makoto*; Tsuchiya, Kunihiko
Journal of Asian Ceramic Societies (Internet), 8(4), p.1154 - 1161, 2020/12
Times Cited Count：3 Percentile：17(Materials Science, Ceramics)Pulsed electric current sintering of molybdenum trioxide (MoO) was carried out by one- and two-step pressuring methods for fabrication of irradiation target using production of 
Mo and 
Tc nuclear medicine. At 550
C by the two-step pressurizing method, a relative density of 93.1% was obtained while, by the one-step pressurization method, the relative density was 76.9%. Direct sample temperature measurements were conducted by inserting a thermocouple in a punch. By the two-step pressurizing method, the sample temperature was higher than that by the one-step pressurizing method even almost the same die temperature. From voltage and current waveforms, it was thought that the conductivity of the sample increased by the two-step pressurizing method to increase the sample temperature and the relative density. The two-step pressurization method enables us to prepare dense targets at a low temperature from recycled and coarse-grained 
Mo enriched MoO powder.
Yamamoto, Masahiro; Sato, Tomonori; Igarashi, Takahiro; Ueno, Fumiyoshi; Soma, Yasutaka
Proceedings of European Corrosion Congress 2017 (EUROCORR 2017) and 20th ICC & Process Safety Congress 2017 (USB Flash Drive), 6 Pages, 2018/09
The authors have studied the differences between outer surface and the crevice-like portion of SUS316L in high pressurized and high temperature water containing dissolved oxygen. We have already introduced that changes in the characteristics of corrosion products along the crevice directions and gap width. It is suggested that the environmental conditions are different with the features of crevice from these results. In this report, we introduce the changes in oxide films with crevice gaps and comparison with the numerical simulation data utilizing of FEM calculation.
Negishi, Kazuo; Hosoya, Takusaburo; Sato, Kenichiro*; Somaki, Takahiro*; Matsuo, Ippei*; Shimizu, Katsusuke*
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9418_1 - 9418_7, 2009/05
An innovative containment vessel, namely Steel plate reinforced Concrete Containment Vessel (SCCV) is developed for Japan Sodium-Cooled Fast Reactor (JSFR). Reducing plant construction cost is one of the most important issues for commercialization of fast reactors. This study investigated construction issues including the building structure and the construction method as well as design issues in terms of the applicability of SCCV to fast reactors. An experimental study including loading and/or heating tests has been carried out to investigate the fundamental structural features, which would be provided to develop methodology to evaluate the feasibility of SCCV under the severe conditions. In this paper, the test plan is described as well as the first test results.
Konomura, Mamoru; Ogawa, Takashi; Okano, Yasushi; Yamaguchi, Hiroyuki; Murakami, Tsutomu; Takaki, Naoyuki; Nishiguchi, Youhei; Sugino, Kazuteru; Naganuma, Masayuki; Hishida, Masahiko; et al.
JNC TN9400 2004-035, 2071 Pages, 2004/06
JNC-TN9400-2004-035.pdf:76.42MB
The attractive concepts for Sodium-, lead-bismuth-, helium- and water-cooled FBRs have been created through using typical plant features and employing advanced technologies. Efforts on evaluating technological prospects of feasibility have been paid for these concepts. Also, it was comfirmed if these concepts satisfy design requierments of capability and performance presumed in the feasibilty study on commertialization of Fast Breeder Reactor Systems. As results, it was concluded that the selection of sodium-cooled reactor was most rational for practical use of FBR technologies in 2015.
Sato, Tatsuhiko; Fujii, Katsutoshi; Murayama, Takashi; Sakamoto, Yukio; Yamaguchi, Yasuhiro; Sato, Yukio*; Soma, Nobuyuki*; Fujisaki, Noboru*; Hara, Satoshi*; Aikawa, Yukio*; et al.
JAERI-Tech 2002-028, 20 Pages, 2002/03
JAERI-Tech-2002-028.pdf:5.01MB
Tokyo Fire Department developed an armored car against radiation accidents. Dose attenuation factors of the radiation shields had been determined by a simple estimation, and a more precise evaluation was required. By request from Tokyo Fire Department, a precise evaluation of the dose attenuation factor was carried out. The evaluation was done by a Monte Carlo radiation transport simulation code MCNP4B. Benchmark experiments using neutron and gamma ray sources were also performed for ensuring the evaluation method. As a result, it was found out that doses of neutron and gamma ray were attenuated to approximately 10% and 25% by the thickest shield, respectively. These values were close to the ones which had already obtained by the simple estimation.
Sato, Kazuyoshi; Neyatani, Yuzuru; Maruo, Takeshi; Mukai, Satoru*; Uchida, Shoji*; Soman, Yoshindo*
no journal, ,
no abstracts in English
Kato, Atsushi; Negishi, Kazuo; Sato, Kenichiro*; Akiyama, Yo*; Hara, Hiroyuki*; Iwasaki, Mikinori*; Abe, Ganji*; Tokiyoshi, Takumi*; Okafuji, Takashi*; Umeki, Katsuhiko*; et al.
no journal, ,
Report research and development activities related to steel plate reinforced concrete containment vessel for the JSFR conducted as a part of METI commissioned research.
for production of Tc by neutron irradiationSuematsu, Hisayuki*; Sato, Soma*; Nanko, Makoto*; Tsuchiya, Kunihiko; Nishikata, Kaori; Suzuki, Tsuneo*; Nakayama, Tadachika*; Niihara, Koichi*
no journal, ,
Spark plasma sintering of MoO was carried out for production of Tc from Mo by the (n,
) method in a nuclear reactor. Powder of MoO with an average grain size of 0.8
m and a purity of 99.99% was pressed in a graphite die with a diameter of 20 mm. Then, the green compact was heated in a spark plasma sintering apparatus with heating rates of 100 
200C/min to 500 600C in vacuum. After holding the temperature for 5 min, the sample was quenched. The sintered samples were characterized by powder X-ray diffraction for phase identifications, electron energy loss spectroscopy for compositional analyses and scanning electron microscopy for grain size measurements. After sintering at 550C, a sintered bulk of MoO with a relative density of 98% was obtained. These properties are good enough for separation of Tc and recycle of Mo.
powder for a radioisotope productionSato, Soma*; Nanko, Makoto*; Suzuki, Tsuneo*; Nakayama, Tadachika*; Suematsu, Hisayuki*; Niihara, Koichi*; Tsuchiya, Kunihiko
no journal, ,
no abstracts in English
for production of radioactive isotopesSuematsu, Hisayuki*; Seki, Misaki*; Sato, Soma*; Nanko, Makoto*; Tsuchiya, Kunihiko; Nishikata, Kaori; Suzuki, Tsuneo*; Nakayama, Tadachika*; Niihara, Koichi*
no journal, ,
no abstracts in English
for Production of Radioactive IsotopesSuematsu, Hisayuki*; Sato, Soma*; Seki, Misaki*; Nanko, Makoto*; Nishikata, Kaori; Suzuki, Yoshitaka; Tsuchiya, Kunihiko; Suzuki, Tsuneo*; Nakayama, Tadachika*; Niihara, Koichi*
no journal, ,
Tc has been utilized as a radioactive isotope in medical applications. The majority of this isotope has been separated from nuclear fission products in testing reactors with highly enriched 
U fuel. However, these reactors have been shut down because of the age and the nuclear security reasons. On the other hand, a nuclear reaction method has been proposed. This method is to irradiate Mo by neutrons in a reactor to form Mo and then to decay to Tc. As the target, MoO pellets are required. However, because of the low evaporation temperature (700 C) and coarse grain size of Mo enriched powder, it was difficult to obtain high density MoO pellets. To overcome this problem, a two-step loading method in pulsed electric current sintering was carried out in this study.
Mohamad, A. B.; Nemoto, Yoshiyuki; Soma, Yasutaka; Ishijima, Yasuhiro; Sato, Tomonori; Ioka, Ikuo; Pham, V. H.; Miwa, Shuhei; Nakajima, Kunihisa; Kaji, Yoshiyuki; et al.
no journal, ,
no abstracts in English
Yamashita, Shinichiro; Mohamad, A. B.; Nemoto, Yoshiyuki; Soma, Yasutaka; Ishijima, Yasuhiro; Sato, Tomonori; Ioka, Ikuo; Pham, V. H.; Miwa, Shuhei; Nakajima, Kunihisa; et al.
no journal, ,
JAEA is conducting research on various coating technologies for fuel cladding tubes aimed at improving accident resistance. In the lecture, we will introduce new equipment development at JAEA aimed at using these studies in addition to an overview of the entire project.
Mohamad, A. B.; Soma, Yasutaka; Nemoto, Yoshiyuki; Abe, Yosuke; Ioka, Ikuo; Sato, Tomonori; Ishijima, Yasuhiro; Miwa, Shuhei; Nakajima, Kunihisa; Kaji, Yoshiyuki; et al.
no journal, ,
Japan Atomic Energy Agency (JAEA) has launched fundamental researches on zircalloy with accident tolerance since 2019. The main purposes of the fundamental researches are to deepen the understanding of the zircalloy behavior under long-term normal operation or Loss of Coolant Accident (LOCA), beyond design basis accident (B-DBA) and severe accident (SA) conditions, and to support the implementation of Cr-coated zircalloy which is being developed by Japanese vendor. JAEA has also been conducted basic technology developments which is necessary for the understanding of the behavior of accident tolerant coated-zircalloy under normal operation, LOCA, B-DBA and SA conditions. For example, the ion irradiation technique combined with light water reactor (LWR) coolant conditions is being developed to simulate the normal operation condition. In addition, to understand LOCA phenomena, the results obtained from the LOCA test are implemented in the machine learning to understand in more detail the cladding fracture and ballooning. Furthermore, a separate effect test, such as the high temperature oxidation test, is also carried out. The fission product release during the B-DBA and SA are also included in the research program. The research results obtained by using these basic technologies will be integrated and implemented into the fuel performance analysis code to predict the fuel performance under reactor operating conditions.
