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Kasahara, Naoto*; Yamano, Hidemasa; Nakamura, Izumi*; Demachi, Kazuyuki*; Sato, Takuya*; Ichimiya, Masakazu*
Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 8 Pages, 2024/03
In this study, we propose failure mitigation methods by application of passive safety structures. The idea of the passive safety structures was applied to next generation fast reactors under high temperature conditions and excessive earthquake conditions.
Kasahara, Naoto*; Yamano, Hidemasa; Nakamura, Izumi*; Demachi, Kazuyuki*; Sato, Takuya*; Ichimiya, Masakazu*
Transactions of the 26th International Conference on Structural Mechanics in Reactor Technology (SMiRT-26) (Internet), 8 Pages, 2022/07
Utilizing fracture control, we are developing a technology to suppress the expansion of damage caused by an event that exceeds the design assumption. We made a plan to develop three issues; (1) Technology for mitigating failure consequence at extremely high temperatures, (2) Technology for mitigating failure consequence against excessive earthquakes, and (3) Methodology for improving reactor structure resilience.
Oikawa, Kenichi; Kiyanagi, Yoshiaki*; Sato, Hirotaka*; Omae, Kazuma*; Pham, A.*; Watanabe, Kenichi*; Matsumoto, Yoshihiro*; Shinohara, Takenao; Kai, Tetsuya; Harjo, S.; et al.
Materials Research Proceedings, Vol.15, p.207 - 213, 2020/02
Hama, Katsuhiro; Sasao, Eiji; Iwatsuki, Teruki; Onoe, Hironori; Sato, Toshinori; Fujita, Tomoo; Sasamoto, Hiroshi; Matsuoka, Toshiyuki; Takeda, Masaki; Aoyagi, Kazuhei; et al.
JAEA-Review 2016-014, 274 Pages, 2016/08
JAEA-Review-2016-014.pdf:44.45MB
We synthesized the research results from the Mizunami/Horonobe Underground Research Laboratories (URLs) and geo-stability projects in the second midterm research phase. This report can be used as a technical basis for the Nuclear Waste Management Organization of Japan/Regulator at each decision point from siting to beginning of disposal (Principal Investigation to Detailed Investigation Phase).
reflector and proton beam windowTeshigawara, Makoto; Kinoshita, Hidetaka; Wakui, Takashi; Meigo, Shinichiro; Seki, Masakazu; Harada, Masahide; Ito, Manabu; Suzuki, Toru; Ikezaki, Kiyomi; Maekawa, Fujio; et al.
JAEA-Technology 2012-024, 303 Pages, 2012/07
JAEA-Technology-2012-024.pdf:46.04MB
3 GeV Protons with 1 MW beam power are irradiated to mercury target of spallation neutron source in Materials and Life science Facility (MLF), which is one of facilities of J-PARC. Irradiated components, such as target container, moderator, reflector and proton beam window, are needed to replace periodically due to irradiation damage of high energy protons and neutrons. These used components are replaced remotely because of highly activated. Maintenance scenario was settled so as to handle these components. Required remote handling machines were designed and installed in hot cell and other room of the MLF. We performed remote handling tests by using actual components to confirm the design. We report results, such as replacement procedure, trouble and its solution, etc., for moderator, reflector and proton beam window in order to provide the handling of actual used components.
pellets for production of 
Mo medical isotopeKimura, Akihiro; Sato, Yuichi*; Tanase, Masakazu*; Tsuchiya, Kunihiko
IOP Conference Series; Materials Science and Engineering, 18(4), p.042001_1 - 042001_4, 2011/10
In the medical field, the radioisotopes are indispensable. Especially, 
Tc is most commonly used as a radiopharmaceutical. However, the supply of Mo in Japan depends fully on the import from foreign countries. JMTR has a plan to produce a medical isotope of Mo, the parent nuclide of Tc by the (n,
) method and a part of the import volume can be covered of the home country. In this plan, it is important to develop the production method of the irradiation targets such as the Molybdenum oxide (MoO) pellets. However, MoO is low sublime temperature and it is difficult to produce the pellets with high density. In this study, MoO pellets were produced by two kinds of production methods. As a result, MoO pellet of about 70% TD was obtained by CIP and MoO pellet of over 95% TD was obtained by SPS.
Tc-MDP using Tc separated from Mo produced by 
Mo(
,2)MoNagai, Yasuki; Hatsukawa, Yuichi; Kin, Tadahiro; Hashimoto, Kazuyuki; Motoishi, Shoji; Konno, Chikara; Ochiai, Kentaro; Takakura, Kosuke; Sato, Yuichi*; Kawauchi, Yukimasa*; et al.
Journal of the Physical Society of Japan, 80(8), p.083201_1 - 083201_4, 2011/08
Times Cited Count：15 Percentile：65.21(Physics, Multidisciplinary)We have for the first time succeeded to separate Tc from a Mo oxide sample irradiated by accelerator neutrons, and to formulate Tc-methylene diphosphonate (Tc-MDP). Mo, the mother nuclide of Tc, was produced by the Mo(,2)Mo reaction using about 14 MeV neutrons provided at the Fusion Neutronics Source of Japan Atomic Energy Agency. The Tc was separated from Mo by the sublimation method, and its radionuclide purity was confirmed to be higher than 99.99%. The labeling efficiency of Tc-MDP was shown to be higher than 99%. These values exceed the United States Pharmacopeia requirements for a fission product, Mo. Consequently, a Tc radiopharmaceutical preparation formed by using the mentioned Mo can be a promising substitute for the fission product Mo. A longstanding problem to ensure a reliable and constant supply of Mo in Japan can be partially mitigated.
Tc master-milker; Comparison with PZC based wet methodIshitsuka, Etsuo; Yamabayashi, Hisamichi*; Tanase, Masakazu*; Fujisaki, Saburo*; Sato, Norihito*; Hori, Naohiko; Awaludin, R.*; Gunawan, A. H.*; Lubis, H.*; Mutalib, A.*
JAEA-Technology 2011-019, 18 Pages, 2011/06
JAEA-Technology-2011-019.pdf:2.61MB
Feasibility study of sublimation type Tc master-milker was carried out as a Mo/T production development with the JMTR. As the feasibility study, the experimental equipment for sublimation method and wet method with PZC based Tc solution were tentatively manufactured, and their properties as the master-milker were investigated by comparing two methods with each other. As a result, it was found that the Tc recovery rate and process time of the sublimation method were about 80% and 1.5 hour, respectively, and the similar values were observed with the wet method. Superior points of the sublimation method are easier operation and reusability of the used MoO comparing with the wet method. On the other hand, disadvantageous point is that the Tc recovery rate decreases with the increase of treating amount of MoO.
Tc extraction techniques from Mo by (n,) reactionKimura, Akihiro; Hori, Naohiko; Tsuchiya, Kunihiko; Ishihara, Masahiro; Yamabayashi, Hisamichi*; Tanase, Masakazu*; Fujisaki, Saburo*; Sato, Yuichi*
JAEA-Review 2010-053, 23 Pages, 2010/11
JAEA-Review-2010-053.pdf:2.52MB
Production techniques of Mo, parent nuclide of Tc, have been developed for the industrial utilization as medical diagnosis medicine after the JMTR refurbishment. The (n,) method is proposed in JMTR because of low-amount radioactive wastes and easy Tc production process. In this study, the production of the high-density MoO pellet and concentration techniques of Tc solution were developed. As the trial test, the MoO pellets with high density were produced by the SPS (Spark Plasma Sintering) method. On the other hands, it was possible to concentrate Tc solution by the solvent extraction using Methyl Ethyl Ketone (MEK). From the result, the Tc concentrating device with more than 80% concentration efficiency, was performed successfully.
Sato, Satoshi; Verzilov, Y.*; Ochiai, Kentaro; Wada, Masayuki*; Kutsukake, Chuzo; Tanaka, Shigeru; Abe, Yuichi; Seki, Masakazu; Oginuma, Yoshikazu*; Kawabe, Masaru*; et al.
Journal of Nuclear Science and Technology, 44(4), p.657 - 663, 2007/04
Times Cited Count：9 Percentile：54.79(Nuclear Science & Technology)Neutronics experiments have been performed for the solid breeder blanket using a DT neutron source at the FNS facility in JAEA. We have applied the blanket mockup composed of two enriched Li
TiO and three beryllium layers, and measured the detailed spatial distribution of the tritium production rate (TPR) using enriched LiCO pellets. TPRs in the pellets have been measured by a liquid scintillation counter. Experiments have been done under a condition with a neutron reflector surrounding the DT neutron source. Numerical simulations have been performed using the MCNP-4C with the FENDL-2.0 and JENDL-3.3. The ranges of ratios of calculation results to experimental ones (C/Es) are 0.97-1.17 concerning with local TPR, and 1.04-1.09 for the integrated tritium production. It is found that the total integrated tritium production, which corresponds to tritium breeding ratio, can be predicted within uncertainty of 10% using the Monte Carlo calculation code and latest nuclear data libraries.
Narita, Osamu; Iwata, Noboru; Isobe, Yoshihiro; Seki, Masakazu; Kadosaka, Hidetake; Ninomiya, Kazushige; Sato, Osamu
JAEA-Technology 2006-037, 102 Pages, 2006/06
JAEA-Technology-2006-037.pdf:7.67MB
We edited and published the Environmental Report according to the law on Promotion of the Environmental Concerning Activity by means of the Publish of Environmental Concerning Data. The report included the environmental concerning activity at the Japan Atomic Energy Research Institute (JAERI) and the Japan Nuclear Fuel Cycle Development Institute (JNC) in the first half year of 2005. This report is the first one which is regulated and obligated by the law. We have made much effort for gathering the data and gained a lot of experience on editing the report. We hope this paper is useful not only for the back data of our environmental report, but also for the organization which is planning to publish the similar environmental report.
Asakura, Toshihide; Sato, Makoto; Matsumura, Masakazu; Morita, Yasuji
JAERI-Conf 2005-007, p.345 - 347, 2005/08
This paper reviews the succeeding development and utilization of Extraction System Simulation Code for Advanced Reprocessing (ESSCAR). From the viewpoint of development, more tests with spent fuel and calculations should be performed with better understanding of the physico-chemical phenomena in a separation process. From the viewpoint of process safety research on fuel cycle facilities, it is important to know the process behavior of a key substance; being highly reactive but existing only trace amount.
Isayama, Akihiko; Iwama, Naofumi*; Showa, Takeshi*; Hosoda, Yosuke*; Isei, Nobuaki; Ishida, Shinichi; Sato, Masayasu
Japanese Journal of Applied Physics, Part 1, 42(9A), p.5787 - 5796, 2003/09
Times Cited Count：3 Percentile：15.61(Physics, Applied)no abstracts in English
Si and response to hydrostatic pressure as studied by neutron diffractionKawarazaki, Shuzo*; Sato, Masugu*; Miyako, Yoshihito*; Chigusa, Nobusato*; Watanabe, Kenji*; Metoki, Naoto; Koike, Yoshihiro; Nishi, Masakazu*
Physical Review B, 61(6), p.4167 - 4173, 2000/02
Times Cited Count：86 Percentile：93.71(Materials Science, Multidisciplinary)no abstracts in English
*; *; *;
JAERI-M 82-077, 150 Pages, 1982/07
no abstracts in English
*; *; *; Daigo, Yoshimichi
PNC TN941 81-05, 235 Pages, 1981/01
Failure propagation tests have been carried out using the Steam Generator Safety Test Facility (SWAT-3) in PNC O-arai Engineering Center to establish the method of safety design of the LMFBR Monju's prototype steam generator with reference to preventing sodium-water reaction accidents. The main object of these tests is to understand how the failure propagation to heat transfer tubes around progresses owing to the water leakage from the initial nozzle. Here reported are the data of the first three failure propagation tests conducted in october 1979. Three injection tests (SWAT-3 Run-8 through 10) were executed whose initial water leak rates were 36 g/s, 6.8 g/s and 570 g/s respectively. (1)Failure propagation progressed in about each one minute in Run-8 and 10, but it did not occur in Run-9. (2)The maximum size of penetration holes is about 5.7 mm
for water tubes, and is 18 mm 
33 mm for gas tubes. (3)The main mechanism of failure seemed to be the wastage. (4)There were some bowing and buldging tubes as well as wastaged tubes in Run-8 and 10. (5)The wastage rate was less than 710
mm/s in accordance with results of intermediate wastage tests.
*; *; *; *; *; *
PNC TN941 80-27, 272 Pages, 1980/02
A series of intermediate leak wastage tests were conducted using Large Leak Sodium-Water Reaction Test Rig (SWAT-1) at O-arai Engineering Center. The purpose of these tests was to clarify the mechanism of failure propagation to adjacent heat transfer tubes due to the flame jet of leak water in the steam generator of LMFBR. Then the value of the design basis leak for the steam generator should be proposed. The wastage rate, the size of secondary failure, and the multi-tube wastage phenomenon were mainly discussed on the basis of eleven test results, in which the water leak rate was in the range of 10
200 g/sec. Findings are as follows ; (1)The wastage rate depends on L/D (L: nozzle to target distance, D : nozzle diameter) and has a maximum value of 7 10 mm/sec at L/D=2030. (2)The time of secondary failure occurring in the tube bunk structure does not depend on the water leak rate. (3)The maximum diameter of the penetration hole obtained in these tests was 19mm. (4)The dominant mechanism of the failure propagation is not overheating but wastage in the intermediate leak region. (5)The extent of multi-tube wastage increases with the increasing leak rate. Six tubes were damaged considerably at the leak rate of 200 g/sec.
Hiroi, Hiroshi*; *; *; *; *; *
PNC TN941 79-04, 274 Pages, 1979/10
Large leak sodium-water reaction tests have been carried out using SWAT-3 facility in PNC O-arai Engineering Center to obtain the data on the safe design of the prototype, LMFBR Monju is steam generator against large leak accident. This report gives the results of SWAT-3 run-5 test. The heat transfer tube bundle of the evaporator used in run-5 test was designed and fabricated by MITSUBISHI HEAVY INDUSTRIES, LTD. The water injection rate into the evaporator was 15 kg/sec, which corresponds to test scale of 5 tubes failure in actural size system according to iso-velocity modeling. Measurements were made of pressure, strain, temperature, sodium level, void, thrust load, acceleration, displacement, flow rate, and so on. Initial spike pressure was 24.6 kg/cm
a nearest to injection point, and the maximun quasi-steady pressure in evaporator was 7.6 kg/cma. The rupture disc of evaporator was bursted at 0.23 sec after water injected, and the pressure relief system was well functioned. No secondary tube failure was observed.
*; *; *; Daigo, Yoshimichi; *
PNC TN941 79-141, 198 Pages, 1979/09
When a large leak sodium-water reaction accident occurs in a steam generator of LMFBR, pressure relief piping might be received thermal shock or blocked by frozen sodium, if it is not preheated. Then, the thermal transient tests were performed using the large leak sodium water reaction test rig SWAT-1. The results are summarized as follows; (1)Four tests were executed. The water injection rate of two tests was equivalent to that of several DEG (double-ended guilotine) failure of heat transfer tubes considering the difference of evapourator inner diameters between "Monju" and SWAT-1, and in other two tests the injection ratio was equivalant to less than that of 1 DEG. (2)Flow pattern in the pressure relief piping of two large injection rate tests was as follows, void fraction was as low as that of sodium single-phase flow in its early stage of 0.20.3 sec., and rapidly increased to about 0.9. In case of the small injection rate tests, the stratified flow had continued for 23 sec., it was followed by hydrogen gas single-phase flow. (3)In the large injection rate tests the maximum value of heat flux was about 110
[kcal/(mh)], and that of heat transfer coefficient was 310
[kcal/(mh
C)] except in its very initial stage. In case of the small tests, they were lower. (4)In the large injection rate tests, stain of outer piping surface was about 8001,50010
, which agrees with the calculation using above value as heat transfer coefficient. (5)Possibility of blockage by frozen sodium is seemed to be very little in SWAT-1 test rig.
