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Takano, Kazuya; Oki, Shigeo; Ozawa, Takayuki; Yamano, Hidemasa; Kubo, Shigenobu; Ogura, Masashi*; Yamada, Yumi*; Koyama, Kazuya*; Kurita, Koichi*; Costes, L.*; et al.
EPJ Nuclear Sciences & Technologies (Internet), 8, p.35_1 - 35_9, 2022/12
The France and Japan teams have carried out collaborative works to have common technical views regarding a sodium-cooled fast reactor concept. Japan has studied the feasibility of an enhanced high burnup low-void effect (CFV) core and fuel using oxide dispersion-strengthened steel cladding in ASTRID 600. Regarding passive shutdown capabilities, Japan team has performed a preliminary numerical analysis for ASTRID 600 using a complementary safety device, called a self-actuated shutdown system (SASS), one of the safety approaches of Japan. The mitigation measures of ASTRID 600 against a severe accident, such as a core catcher, molten corium discharge assembly, and the sodium void reactivity features of the CFV core, are promising to achieve in-vessel retention for both countries. The common design concept based on ASTRID 600 is feasible to demonstrate the SFR core and safety technologies for both countries.
Yamano, Hidemasa; Kubo, Shigenobu; Tokizaki, Minako*; Nakamura, Hironori*
Proceedings of International Conference on Topical Issues in Nuclear Installation Safety; Strengthening Safety of Evolutionary and Innovative Reactor Designs (TIC 2022) (Internet), 12 Pages, 2022/10
Specific design features of advanced sodium-cooled fast reactors (SFRs) designed in Japan are a passive reactor shutdown system, a passive decay heat removal system (DHRS), and an in-vessel retention (IVR) concept against an anticipated transients without scram (ATWS) in design extension condition (DECs). The present paper describes numerical analysis methodologies for event sequences studied in Japan and some numerical analyses of DECs to show the effectiveness of the passive shutdown system against a typical ATWS and severe accident mitigation measures for the IVR of molten core. For the passive shutdown capability, the numerical analysis has demonstrated the effectiveness of a self-actuated shutdown system against a severe ATWS event, for which the temperature response time was separately evaluated by a computational fluid dynamics (CFD) code. A recently developed debris-bed cooling analysis methodology coupled with a CFD code and a debris-bed module has successfully simulated a three-dimensional coolant flow field near the debris bed with the passive DHRS capability in order to demonstrate the debris-bed coolability on a core catcher.
Horiguchi, Naoki; Yoshida, Hiroyuki; Abe, Yutaka*
Journal of Nuclear Science and Technology, 56(3), p.278 - 290, 2019/03
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Venturi scrubber (VS) has been installed in NPPs as one of the components of filtered venting systems. It can eliminate fine aerosol particles including fission products from polluted gas. Under extremely high-velocity conditions during severe accident, the data related to the droplet in the VS, which affects decontamination performance, is insufficient. This objective is to obtain the diameter of the extremely high-speed droplets and clarify an applicability of the existing correlations for the diameter. To visualize the extremely high-velocity droplet, an optical system with high frame rate and high resolution was developed. Visualization experiment using the VS under air-water condition was conducted, and droplet diameter distributions and Sauter mean diameter (SMD) were obtained. By comparing experimental data with values evaluated by existing correlations, it was clarified that the Nukiyama-Tanasawa equation can evaluate the SMD with good accuracy in the gas velocity range 82-250 m/s.
Yoshizawa, Atsufumi*; Oba, Kyoko; Kitamura, Masaharu*
Ningen Kogaku, 54(3), p.124 - 134, 2018/06
Fukushima Daiichi Nuclear Power Plant caused a severe accident which released a large amount of radioactivity triggered by the Great East Japan Earthquake. The existing investigation reports of the accident prepared by several institutions pay attention only to the process which caused the accident but not much to the accident mitigation or the recovery process. This study focused on Unit 3 of Fukushima Daiichi Nuclear Power Plant, including its recovery process from the accident. Based on the public data, the time sequences for the recovery process between the accident occurrence and the state of cold shutdown were classified. Then, the groups of actions were sorted out in terms of ergonomics viewpoint. The important responses in the recovery process were identified and analyzed referring to the m-SHEL model. As a result, new lessons were learned from the accident case regarding the actions required for recovering from the accident.
Sono, Hiroki; Tsukamoto, Michio; Aizawa, Eiju; Takeuchi, Masaki; Fukaya, Yuji; Iseda, Hirokatsu*; Ogawa, Kazuhiko; Sakuraba, Koichi; Tonoike, Kotaro
UTNL-R-0446, p.3_1 - 3_10, 2005/03
This report describes an investigation and countermeasures of the unexpected reactor shutdown of the TRACY facility at the Japan Atomic Energy Research Institute on June 17, 2004, caused by a malfunction of a safety rod of the facility. As a result of the investigation, the principal cause of the malfunction was ascribed to a drop in attraction of a binding magnet for the safety rod by a small piece of a foreign material. The following countermeasures were taken against its recurrence: prevention of incidence of foreign materials, reduction of the chance of intrusion of foreign materials into devices, and certain detection of foreign materials. This lesson learned through the experience is expected to be shared with the staffs concerned in reactor operation and maintenance.
Takamatsu, Kuniyoshi; Nakagawa, Shigeaki
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(4), p.369 - 380, 2004/12
This paper describes the validation of the TAC-NC code using the measured value of the test by tripping of one and two out of three gas circulators at 30% (9MW). By upgrading the code, the analytical result can evaluate accurately the measured value of the transient temperature distribution within 20
C. Also, the improved code can analyze the maximum fuel temperature and temperature distributions of the test by tripping all the three gas circulators. The result of this study can be applied to development for not only the commercial HTGRs but also the Very High Temperature Reactor (VHTR) such as one of the Generation IV reactors.
Katanishi, Shoji; Kunitomi, Kazuhiko; Tsuji, Nobumasa*; Maekawa, Isamu*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(3), p.257 - 267, 2004/09
no abstracts in English
Hori, Toshihiko*; Chishiro, Etsuji; Yamazaki, Masayoshi*; Suzuki, Hiroyuki*; Hasegawa, Kazuo
KEK Proceedings 2003-16 (CD-ROM), 4 Pages, 2004/02
no abstracts in English
Sono, Hiroki; Fukaya, Yuji; Yanagisawa, Hiroshi; Miyoshi, Yoshinori
JAERI-Tech 2003-065, 61 Pages, 2003/07
JAERI-Tech-2003-065.pdf:3.11MB
A series of critical experiments using a heterogeneous core of the Static Experiment Critical Facility (STACY) in the Japan Atomic Energy Research Institute is planned in F.Y. 2003. In the experiment, the core is composed of uranyl nitrate solution (
U enrichment 6 wt%) and 333 pins of uranium dioxide (U enrichment 5 wt%) loaded in lattice-pitch of 1.5 cm. Prior to the experiment, neutronic characteristics are analyzed to evaluate neutronic safety and criticality limitations of the core. The analyzed items are the parameters on criticality, reactivity and reactor shutdown margins. In the analyses, a Monte Carlo code, MVP, and a neutronics code system, SRAC, have been used with an evaluated nuclear data library, JENDL-3.3. By using the calculated characteristics, simplified equations to interpolate these values and criticality limitations of the core are evaluated. It has been also confirmed that the reactor shutdown margins will comply with safety criteria under all fuel conditions in the experiments.
Yanagisawa, Hiroshi; Sono, Hiroki
JAERI-Tech 2003-057, 39 Pages, 2003/06
JAERI-Tech-2003-057.pdf:2.76MB
no abstracts in English
Ida, Mizuho*; Nakamura, Hideo; Nakamura, Hiroshi*; Nakamura, Hiroo; Ezato, Koichiro; Takeuchi, Hiroshi
Fusion Engineering and Design, 63-64, p.333 - 342, 2002/12
Times Cited Count：43 Percentile：91.31(Nuclear Science & Technology)no abstracts in English
Odano, Naoteru; Ishida, Toshihisa; Ochiai, Masaaki
JAERI-Research 2001-044, 53 Pages, 2001/10
JAERI-Research-2001-044.pdf:2.35MB
A very small reactor, MR-1G with a thermal output of 1 MW, is the integral-pressurized type reactor to be used for heat supply to an office building in a city. Neutronic study has been carried out for design of the MR-1G, the core of which could achieve continuous long-term operation without refueling for 10 years assuming a load factor of the core of 44 %. In the present study, arrangement of fuel rods and U enrichment of UO
fuel rods were surveyed as design parameters. The U enrichment was determined to be 8.5 wt% to satisfy design requirement. Reactor physics parameters including reactivity coefficients and power distributions were evaluated for the determined core specifications. Reactor physics parameters related to core safety were also analyzed. Results of the safety analysis indicated that the determined core specifications satisfied design conditions. Reactor shutdown performance by dropping the reflector, which was adopted as a passive reactor shutdown system, was analyzed and confirmed it's availability from a viewpoint of reactor physics.
*
JNC TN9440 2000-008, 79 Pages, 2000/08
JNC-TN9440-2000-008.pdf:2.33MB
This report summarizes the operating and irradiatlon data of the experimental reactor "JOYO" 35th cycle. Irradiation tests in the 35th cycle are as follows: (1)C-type irradiation rig (C4F) (a)High burnup performance test of advanced austenitic stainless steel cladding fuel pins (in collaboration with France) (2)C-type irradiation rig (C6D) (a)Large diameter fuel pins irradiation tests (3)Core Materials Irradiation Rig (CMIR-5) (a)Cladding tube materials irradiation tests for "MONJU" (4)Structure Materials Irradiation Rigs (SMIR) (a)Decision of material design base standard of structure materials for prototype reactor and large scale reactor (5)Upper core structure irradiation Plug Rig (UPR-1-5) (a)Upper core neutron spectrum effect and accelerated irradiation effect (6)SurVeillance un-instrument Irradiation Rig (SVIR) (a)Confimation of surveillance irradiation condition for "JOYO" (b)Material irradiation tests (based on a contract with universities) The maximum burnup driver assembly "PFD253" reached 67,600 MWd/t (pin average).
JNC TN9400 2000-038, 98 Pages, 2000/04
JNC-TN9400-2000-038.pdf:7.49MB
As an effort in the feasibility study on commercialized Fast Breeder Reactor cycle systems, an evaluation of the measures to prevent the energetic re-criticality in sodium-cooled large MOX core, which is one of the candidates for the commercialized reactor, has been performed. The core disruptive accident analysis of Demonstration FBR showed that the fuel compaction of the molten fuel by radial motion in a large molten core pool had a potential to drive the severe super-prompt re-criticality phenomena in ULOF sequence. ln order to prevent occurrence of the energetic re-criticality, a subassembly with an inner duct and the removal of a part of LAB are suggested based on CMR (Controlled Material Relocation) concept. The objective of this study is the comparison of the effectiveness of CMR among these measures by the analysis using SIMMER-III. The molten fuel in the subassembly with inner duct flows out faster than that from other measures. The subassembly with inner duct will work effectively in preventing energetic re-criticality. Though the molten fuel in the subassembly without a part of LAB flows out a little slower, it is still one of the promising measures. However, the UAB should be also removed from the same pin to prevent the fuel re-entries into the core region due to the pressurization by FCl below the core, unless it disturbs the core performance. The effect of the axial fuel length of the center pin to CMR behavior is small, compared to the effect of the existence of UAB.
Yoshino, Ryuji; Ozeki, Takahisa; Tokuda, Shinji; Iio, Shunji*; Yamazaki, Kozo*
Purazuma, Kaku Yugo Gakkai-Shi, 76(2), p.116 - 126, 2000/02
no abstracts in English
Tanaka, Toshiyuki
Kaku Deta Nyusu (Internet), (63), p.2 - 8, 1999/00
no abstracts in English
Kaieda, Keisuke; ; Arigane, Kenji
Genshiryoku Nenkan '98/'99-Nen Ban, P. 216, 1998/12
no abstracts in English
Yabuuchi, Noriaki; Nakazawa, Toshio; Takahashi, Hiroki; Shimazaki, Junya; Hoshi, Tsutao
JAERI-Tech 97-057, 54 Pages, 1997/11
no abstracts in English
Kawano, Yasunori; Yoshino, Ryuji; Neyatani, Yuzuru; Kondoh, Takashi; Isei, Nobuaki; Ishida, Shinichi; Tobita, Kenji; Hatae, Takaki; Itami, Kiyoshi; Sakasai, Akira; et al.
24th EPS Conf. on Controlled Fusion and Plasma Physics, 21A, p.501 - 504, 1997/00
no abstracts in English
Kawano, Yasunori; Yoshino, Ryuji; Neyatani, Yuzuru; Kondoh, Takashi; Isei, Nobuaki; Ishida, Shinichi; Tobita, Kenji; Hatae, Takaki; Itami, Kiyoshi; Sakasai, Akira; et al.
Fusion Energy 1996, Vol.1, p.345 - 357, 1997/00
no abstracts in English
