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Tanaka, Masaru*; Aoyama, Isao*; Ishizaka, Kaoru*; Ohata, Yuki*; Fukuike, Iori*; Kawase, Keiichi; Watanabe, Masanori; Tokizawa, Takayuki; Miyagawa, Hiroshi*; Ishimori, Yuu
JAEA-Research 2017-003, 65 Pages, 2017/06
JAEA-Research-2017-003.pdf:2.92MB
JAEA Ningyo-toge Environmental Engineering Center and Fukushima Environmental Safety Center have same challenges in risk communication. As reference, similar domestic cases were investigated by our two Centers, and requirements for building long-term relationship were clarified. As follows; (1) Develop new relationship with various stakeholders in the region. (2) Make better use of existing resources (personnel, land and facilities, etc.). (3) Make a concerted effort to create new values with local stakeholders. (4) Make an opportunity which local stakeholders confirm safety and build confidence to the project. These efforts will enhance the opportunities for operators and residents to learn about environment management and environmental protection.
Tanaka, Masaru*; Aoyama, Isao*; Ishizaka, Kaoru*; Ohata, Yuki*; Fukuike, Iori*; Miyagawa, Hiroshi*; Ishimori, Yuu
JAEA-Research 2016-017, 76 Pages, 2017/01
JAEA-Research-2016-017.pdf:10.57MB
From 1955 to 2001, Japan Atomic Energy Agency (JAEA) carried out research and development projects related to uranium exploration, mining refining, conversion and enrichment at/around Ningyo-toge in Japan. Subsequently, JAEA has been conducting projects related to decommissioning of nuclear fuel cycle facilities and remediation of closed mine sites. JAEA had opportunities of communication with local stakeholders through the projects. Consensus building with local stakeholders and maintain it for over decades are important challenges for JAEA Ningyo-toge Environmental Engineering Center. For this aim, similar domestic case were investigated and compared, and evaluated about required measures for long term relationship with local stakeholder.
Nakanishi, Shigeyuki*; Hosoya, Takusaburo; Kubo, Shigenobu*; Kotake, Shoji; Takamatsu, Misao; Aoyama, Takafumi; Ikarimoto, Iwao*; Kato, Jungo*; Shimakawa, Yoshio*; Harada, Kiyoshi*
Nuclear Technology, 170(1), p.181 - 188, 2010/04
Times Cited Count：14 Percentile：67.43(Nuclear Science & Technology)A self-actuated shutdown system (SASS) for sodium cooled fast reactor (SFR) is a passive safety feature which inserts control rods by the gravity force, where the detachment of the rods would be achieved by the coolant temperature rise under anticipated transient without scram (ATWS) conditions. Various out-of-pile tests have already carried out to investigate the basic characteristics of SASS, and a demonstration test of holding stability under the reactor operation condition has been performed, where a function test of the driving system to re-connect and of pulling out the control rod have been done in the experimental reactor JOYO. The element irradiation tests have been also conducted to confirm that no impact will be foreseen by the irradiation. The effectiveness of SASS for a reference core design of JSFR has been evaluated through all types of ATWS. As a result, it is ensured that JSFR will have a reliable passive shutdown system.
Ishikawa, Koki; Takamatsu, Misao; Kawahara, Hirotaka; Mihara, Takatsugu; Kurisaka, Kenichi; Terano, Toshihiro; Murakami, Takanori; Noritsugi, Akihiro; Iseki, Atsushi; Saito, Takakazu; et al.
JAEA-Technology 2009-004, 140 Pages, 2009/05
JAEA-Technology-2009-004.pdf:2.0MB
Probabilistic safety assessment (PSA) has been applied to nuclear plants as a method to achieve effective safety regulation and safety management. In order to establish the PSA standard for fast breeder reactor (FBR), the FBR-PSA for internal events in rated power operation is studied by Japan Atomic Energy Agency (JAEA). The level1 PSA on the experimental fast reactor Joyo was conducted to investigate core damage probability for internal events with taking human factors effect and dependent failures into account. The result of this study shows that the core damage probability of Joyo is 5.0
10
per reactor year (/ry) and that the core damage probability is smaller than the safety goal for existed plants (10 ry) and future plants (10
/ry) in the IAEA INSAG-12 (International Nuclear Safety Advisory Group) basic safety principle.
Soga, Tomonori; Sekine, Takashi; Takamatsu, Misao; Kitamura, Ryoichi; Aoyama, Takafumi
UTNL-R-0453, p.13_1 - 13_8, 2006/03
no abstracts in English
Kono, Naomi; Aoyama, Takafumi; Sekine, Takashi; Ooka, Makoto; Maeda, Shigetaka; Takamatsu, Misao
JNC TN9200 2003-003, 103 Pages, 2004/03
JNC-TN9200-2003-003.pdf:19.5MB
None
Takamatsu, Misao; Aoyama, Takafumi; Masui, Tomohiko*
PNC TN9410 98-011, 46 Pages, 1997/11
Neutron intensity of spent fuel is important not only for the shielding design and dose evaluation of the reprocessing plant and the transportation of the mixed oxide (MOX) fuel, but also for the core management, because it contains more minor actinides than that of LWR fuel. In order to obtain the experimental data and to improve the accuracy of burnup calculation, the neutron counting rate from a spent fuel subassembly of the JOYO MK-II core with a burnup of 62,5ooMWd/t and cooling time of 5.2 years was measured in the spent fuel storage pond at JOYO. The measured neutron counting rate was then converted to the neutron intensity using the detector response which was obtained by the Monte Carlo calculation code "MCNP-4A". And the neutron intensity was compared with the calculated value based on the 3-D diffusion theory with 7 energy groups using the JOYO core management code system "MAGI". The major results obtained in this study are summarized as follows; (1) The measured neutron intensity per fuel subassembly was about 2.710
n/s and is about 3 times as much as that of fresh (unirradiated) fuel. (2) The average C/E value of neutron intensity was about 1.07. (3)It was found that the axial neutron intensity didn't simply follow the burnup distribution, and the neutron intensity was locally increased at the bottom end of the fuel region due to an accumulation of 
Cm.
Arigane, Kenji; ; ; Aoyama, Isao; Seguchi, Tadao; Takahashi, Hidetake
JAERI-M 92-078, 22 Pages, 1992/06
no abstracts in English
Aoyama, Isao; Komaki, Akira
Proc. of 2nd Asian Symp. on Research Reactors,Vol. 2, p.129 - 144, 1992/00
no abstracts in English
Shirai, Eiji; Aoyama, Isao
Genshiryoku Kogyo, 35(12), p.41 - 50, 1989/12
no abstracts in English
Aoyama, Isao; Shirai, Eiji;
50 Years with Nuclear Fission, p.843 - 848, 1989/00
no abstracts in English
Aoyama, Isao
Japan-China Symp. on Reseach and Test Reactors, 12 Pages, 1988/00
no abstracts in English
Aoyama, Isao; Sakurai, Hiroshi;
Multipurpose Research Reactors, p.435 - 445, 1988/00
no abstracts in English
Takamatsu, Misao; Tobita, Shigeharu; Sekine, Takashi; Kitamura, Ryoichi; Aoyama, Takafumi
no journal, ,
no abstracts in English
Takamatsu, Misao; Itagaki, Wataru; Soga, Tomonori; Sekine, Takashi; Aoyama, Takafumi
no journal, ,
no abstracts in English
Sekine, Takashi; Takamatsu, Misao; Aoyama, Takafumi
no journal, ,
Self actuated shutdown system (SASS) with a Curie point electromagnet (CPEM) has been developed for use in a large-scale liquid metal cooled fast breeder reactor (LMFBR) in order to establish the passive shutdown capability against anticipated transient without scram (ATWS) events. The basic characteristics of SASS have already been investigated by various out-of-pile tests for material elements. As the final stage of the development, the demonstration test of holding stability using the reduced-scale experimental equipment of SASS was conducted in the 1st and 2nd operational cycles of the experimental fast reactor Joyo MK-III. The rod-holding stability and the rod-recovering functions of the driving system to re-connect and pull out the separated control rod were fully confirmed. The results also indicate there is no essential problem for the practical use of SASS about its operational trouble involving the unexpected drop during reactor operation.
Takamatsu, Misao; Kuroha, Takaya*; Aoyama, Takafumi
no journal, ,
Studies for passive safety tests to be conducted by using Joyo have been carried out to demonstrate the inherent safety of LMFBR. In these studies, emphasis has been placed on the improvement of the calculation accuracy of the feedback reactivity. Therefore, in the passive safety tests, the feedback reactivity measurement is planned under the condition simulating ATWS event. The Mimir-N2 which has been developed to analyze Joyo plant dynamic is used for predicting plant behavior during transients. In order to determine the passive safety tests plan, the improvement of the accuracy of Mimir-N2 was required. In this study, transient tests including manual reactor shutdown test, loss of power supply test and preliminary transient over power test were conducted to verify the updated model of the Joyo MK-III core and the heat transport system of Mimir-N2. As a result, a good agreement was obtained between calculated and measured sodium temperatures.
Kawahara, Hirotaka; Takamatsu, Misao; Aoyama, Takafumi; Kuroha, Takaya*
no journal, ,
no abstracts in English
Kawahara, Hirotaka; Iseki, Atsushi; Yamazaki, Manabu; Yamamoto, Masaya; Takamatsu, Misao; Ishikawa, Koki; Kurisaka, Kenichi; Aoyama, Takafumi
no journal, ,
no abstracts in English
