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Doda, Norihiro; Ohshima, Hiroyuki; Kamide, Hideki; Watanabe, Osamu*; Okubo, Yoshiyuki*
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) (CD-ROM), 13 Pages, 2011/09
Toward the commercialization of fast reactors, a design study of Japan Sodium-cooled Fast Reactor (JSFR) is being performed, in which fully natural circulation system is adopted as the decay heat removal system. A new evaluation method of core hot spot which can be applied to natural circulation decay heat removal has been developed. The new method consists of three-step thermal hydraulics analyses in order to consider the effects of physical phenomena particular to natural circulation, such as inter-fuel-assembly heat transfer and flow redistribution in the core due to buoyancy force. From the viewpoint of calculation cost reduction, we have also developed a simplified model substituting for the third step analysis (subchannel analysis). The new method was applied to the evaluations of a loss-of-external-power event and of a sodium leakage accident in a secondary loop of a large scale reactor.
Okubo, Tsutomu; Oki, Shigeo; Ogura, Masashi*; Okubo, Yoshiyuki*; Kotake, Shoji*
Proceedings of 2011 International Congress on Advances in Nuclear Power Plants (ICAPP '11) (CD-ROM), p.479 - 486, 2011/05
A conceptual design study and related R&D on a commercial-base large-scale Japan Sodium-cooled Fast Reactor (JSFR) have been carried out in the framework of the Fast Reactor Cycle Technology development (FaCT) project. As a next generation plant, JSFR adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability and safety. This paper describes the current results of the ongoing conceptual design study on the JSFR core. The most important point in the core design is to achieve a high core average burn-up around 150 GWd/t, assuming the ODS steel utilization as the cladding material. Another design target for the breeding ratio is intended to have some flexibility and is set at from around 1.0 to 1.2 under the design philosophy of the compatible fuel assembly among them. Also, the fuel composition is considered to have some variation range based on the wide variety of the spent fuel composition expected to be treated during the LWR to FBR transition period. The core design study performed in the FaCT project has clarified the feasibility of the JSFR core concept, which is based on the high internal conversion ratio type core using a large fuel rod diameter around 10 mm and satisfies a number of design targets and requirements including ones mentioned above.
Doda, Norihiro; Ohshima, Hiroyuki; Kamide, Hideki; Watanabe, Osamu*; Okubo, Yoshiyuki*
Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 11 Pages, 2010/10
Toward the commercialization of fast reactors, design study of JSFR is being performed. Adoption of fully natural circulation system is being examined as the decay heat removal system. In order to confirm feasibility of the system, we are developing a new evaluation method of core hot spot in transition from rated operation to natural circulation decay heat removal conditions, which requires uncertainty factor assessment for the natural circulation conditions as well as for the rated operation conditions. In this paper, we focus on effects of wire-spacer contact and pellet- cladding eccentricity on the peak cladding temperature as typical uncertainty factors and evaluated these two effects under natural circulation conditions quantitatively.
Ohshima, Hiroyuki; Doda, Norihiro; Kamide, Hideki; Watanabe, Osamu*; Okubo, Yoshiyuki*
Nihon Kikai Gakkai Rombunshu, B, 76(763), p.448 - 450, 2010/03
Toward the commercialization of fast reactors, a design study of Japan Sodium Cooled Fast Reactor is being performed. In this design study, the adoption of decay heat removal system operated by fully natural circulation is being examined from viewpoints of economic competitiveness and passive safety. This paper describes a new evaluation method of core hot spot that is necessary for confirming feasibility of the fully natural circulation decay heat removal system. The new method consists of three analysis steps in order to include effects of thermal hydraulic phenomena particular to the natural circulation decay heat removal and therefore it enables more rational hot spot evaluation rather than conventional ones. This method was applied to a hot spot evaluation of loss-of-external-power event and the result was compared with those by conventional simulations. It was confirmed that the proposed method can estimate the hot spot with reasonable degree of conservativeness.
Chikazawa, Yoshitaka; Okano, Yasushi; Hori, Toru*; Okubo, Yoshiyuki*; Shimakawa, Yoshio*; Tanaka, Toshihiko*
Journal of Nuclear Science and Technology, 43(8), p.829 - 843, 2006/08
Times Cited Count：4 Percentile：30.64(Nuclear Science & Technology)In phase II of the feasibility study of commercialized fast reactor cycle systems, we make a concept of a small sodium cooled reactor for a power source of a city with various requirements, such as, safety and economical competitiveness. Various reactor concepts are surveyed and a tank type reactor whose intermediate heat exchanger and primary main pumps are arranged in series is selected. In this study, a compact long life core and a simple reactor structure designs are pursued. The core type is three regional Zr concentration with one Pu enrichment core, the reactor outlet temperature achieves 550
C and the reactor electric output increases from 150MWe to 165MWe. The construction cost is much higher than the economical goal in the case of FOAK. But the construction cost in the case of NOAK is estimated to be 85.6% achieving the economical goal.
Konomura, Mamoru; SAIGUSA, Toshiie; Mizuno, Tomoyasu; OHKUBO, Yoshiyuki*
Fast Spectrum Reactors, 0 Pages, 2003/00
In Feasibility Studies on Commercialized Fast Reactor (FR) Systems, examining about the subject of three gas cooled FR concepts, (1) carbon dioxide cooled FR using pin type fuel, (2) helium cooled FR using pin type fuel, (3) helium cooled FR using coated particle fuel, a promising concept has been selected from three concepts. From a viewpoint of economic competitiveness and ensuring safety, etc, "helium cooled FR using coated particle fuel" has been selected as a promising concept of gas cooled FR. About fuel assembly concept of helium cooled FR using coated particle fuel, block type vertical flow cooling concept with 2nd boundaries was also examined, other than horizontal flow cooling concept with directly cooling system. About selected helium cooled FR using coated particle fuel, it drew up R&D plan about the most important R&D items influencing on the feasibility of the design concept.
Fujii, Tadashi; ; ; Sakai, Takaaki; ; Oki, Yoshihisa;
JNC TN9400 2002-049, 78 Pages, 2002/09
JNC-TN9400-2002-049.pdf:3.72MB
The conceptual design study of the large-scale sodium-cooled reactor is in progress in the "Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)". The direct water pool storage system is being examined as a candidate concept to simplify the fuel handling facility for the sodium-cooled reactor. In this concept, the decay heat of a fuel subassembly is relatively higher (18kW which is about 4.5 times of the Ex-vessel Storage Tank system in the demonstration FBR). Therefore, the information about the cooling characteristics of the fuel subassembly are lacking in cases of submergence process at a normal operation and cooling water injection process from upper part of the subassembly at a transfer accident. Accordingly, the understanding of the cooling characteristics of the fuel subassembly in higher decay heat condition was pointed out as one of the thermal hydraulic problems which influence the realization of the plant concept. Using the single heater pin equipment, fundamental tests were conducted with the parameters of the thermal conditions of a fuel pin, the outlet shapes of it, the submergence speeds and so on. Then, following basic data were acquired to be reflected in the actual plant design. (1)Cooling modes of the normally submergence tests and water injection tests were identified by visualization of the boiling behavior in the test section and the temperature change of the heater pin. (2)The initial temperature of the heater surface and the blockage size of the outlet of test section were dominating factors to the cooling completion time. (3)Maximum temperature rise of the heater surface was about 4K in normally submergence tests and 6K in water injection tests, respectively. Therefore, the heater was well cooled without significant temperature rise.(4)In the normally submergence tests, the pressure of the upper part of the test section did not exceed the lower part pressure and a water level rise in the test section was not obstructed ...
*; Mizuno, Mineo*; *; Ito, Kunihiro*; *; *
PNC TJ9678 95-003, 195 Pages, 1995/03
In preceding years, as the method of TRU transmutation (TRU means Minor Actinide such as Np, Am and Cm in this report) in FBR, we have evaluated core characteristics and TRU transmuting characteristics for two TRU loading methods : homogeneous TRU-loading method where the TRU fuel is dispersed uniformly throughout the core; and heterogeneous TRU-loading method where a few number of subassemblies with concentrated TRU fuel (target S/As) are loaded in the core. Also, as the research on the TRU transmutation by FBR plant, the survey have been conducted for the effect upon the core characteristics and the TRU transmuting characteristics affected by RE(Rare Earth) which is entrapped when TRU is separated from the high level waste. At the same time, the effect have been surveyed in the case where the TRU recycle was executed in FBR. In this fiscal year, the investigation have been conducted for a core concept in which TRU are loaded separately between the group of Np and the group of Am + Cm + RE , according to the results of advancement of nuclear fuel reprocessing technology. The core concepts with excellent TRU transmuting characteristic have been investigated for the separate loading method of Np and Am + Cm by adding only Np in the core region and adding Am, Cm and RE in the target assemblies located at radial blanket location of the core periphery. As the result of investigation, it has been found that the total amount of TRU up to around 20 % is possible in the case of loading 72 target assemblies composed of 271 pins with the pin diameter almost equal to that of the core fuel, and the TRU amount transmuted per cycle becomes 580 kg which is 3 
4 times higher than the conventional TRU loading methods such as homogeneous loading and heterogeneous loading. The main core characteristics of this core concept are as follows, and the core design is feasible, : ...
*; *; *; *
PNC TJ9678 95-002, 180 Pages, 1995/03
Until now, the investigation was conducted for the concept of various kinds of core (high-Pu enrichment core with increased neutron leakage, absorber addition type high-Pu enrichment core, etc.) in LMFRs to improve the Pu transmutation characteristics without damaging main characteristics of the core, and the main core characteristics including reactivity characteristics were obtained. In preceding year, the reference core with core height of 60cm, core equivalent diameter of 5.3 m and refueling interval (123 days) was selected on 800MWe core, and the effect of the core parameters such as core height, fuel volume fraction, etc. upon the core characteristic was analyzed and evaluated. In this year, the analysis was performed under the below-mentioned specification conditions, and the concept of the Pu burner fast reactor was further clarified. The main specifications are as follows: (Core hight: 60cm) (Equivalent core diameter: 4 m or shorter) (Refueling interval: around half year) The core concept of investigation object is: (1)High-Pu enrichment core 1 (Pu enrichment:45w/o at maximum) (2)High-Pu enrichment core 2 (Pu enrichment:30w/o as average) (3)B
C addition type core Also, based on the high-Pu enrichment core 1, the effect was surveyed for (1)power density, (2)MA addition and (3)Pu composition ratio change, Further, the transient characteristics of the cores of (1) and (2) were analyzed under the ULOF and UTOP events, and, at the same time, the comparison with the conventional MOX core (600MWe FBR) was also conducted. .....
*
PNC TN9410 90-180, 208 Pages, 1990/12
Results of the design study performed in the first half of the FY1990 on 600MWe class large scale FBR plant is reported promptly. The design study was performed with the principal object of proposing plant concept which calls forth constructing will of demonstration plant, laying stress on engineering reliability. The results are summarized in the following design fields. (1)System design. (2)Core and fuel design. (3)Component and structural design. (4)Safety design and safety evaluation. (5)R&Dissues. Contents of the technological results are referred to the collection of OHP using at the interim briefing session. Fundamental feasibility is confirmed on the 600 MWe plant based on the essential design concept of head-access-piping system. Concrete and detail design of components, systems, etc. is the problem to be investigated in the latter half of the year.
approximation*; ;
Journal of Nuclear Science and Technology, 11(8), p.348 - 352, 1974/08
Times Cited Count：1no abstracts in English
; ; ; Shikakura, Sakae;
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles, ,
None
; ;
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR '91), ,
None
Aizawa, Kosuke; Chikazawa, Yoshitaka; Shiraishi, Tadashi*; Sakata, Hideyuki*; Okubo, Yoshiyuki*
no journal, ,
no abstracts in English
Ohshima, Hiroyuki; Kamide, Hideki; Tanaka, Masaaki; Watanabe, Osamu*; Okubo, Yoshiyuki*
no journal, ,
In the design study of commercialized sodium-cooled fast reactor, the adoption of decay heat removal by entire natural circulation is being examined from the viewpoint of enhancing economical competitiveness and safety. In this study, an evaluation method is proposed, in which the hot spot in the core can be rationally evaluated under transient conditions from rated operation to natural circulation decay heat removal.
Doda, Norihiro; Ohshima, Hiroyuki; Kamide, Hideki; Watanabe, Osamu*; Okubo, Yoshiyuki*
no journal, ,
no abstracts in English
Ohshima, Hiroyuki; Kamide, Hideki; Doda, Norihiro; Watanabe, Osamu*; Okubo, Yoshiyuki*
no journal, ,
In the conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR), adoption of a decay heat removal system (DHRS) utilizing passive natural circulation is being examined as one of innovative technologies to improve economical competitiveness and reactor safety at the same time. In order to adopt such a passive DHRS in a large scale SFR, it is necessary to clarify the core cooling capability. In this presentation, a new evaluation method of core peak temperature is proposed, which takes account of buoyancy force effects (e.g., flow redistribution in core/fuel assemblies and inter-assembly heat transfer) and uncertainty factors under the natural circulation conditions.
Doda, Norihiro; Ohshima, Hiroyuki; Kamide, Hideki; Watanabe, Osamu*; Okubo, Yoshiyuki*
no journal, ,
Toward the commercialization of fast reactors, adoption of fully natural circulation system is being examined as the decay heat removal system from the viewpoints of economic competitiveness and passive safety. This paper introduces a new evaluation method for hot spot in core of fast reactor under natural circulation. The new method consists of three step analysis in order to include the effects of thermal hydraulic phenomena particular to natural circulation decay heat removal, inter-assembly heat transfer and flow redistribution in fuel assemblies and in the core by buoyancy force. The method was applied to an analysis of loss-of-external-power event and the result was compared with those by a conventional method and a detailed 3D simulation. It was confirmed that the proposed method can estimate the hot spot with a reasonable degree of conservativeness.
Doda, Norihiro; Ohshima, Hiroyuki; Kamide, Hideki; Watanabe, Osamu*; Okubo, Yoshiyuki*
no journal, ,
Toward the commercialization of sodium cooled fast reactors, adoption of fully natural circulation system is being examined as the decay heat removal system. This paper introduces investigations on a simplified evaluation method for hot spot in core of fast reactor under natural circulation.
