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Yamano, Hidemasa; Kubo, Shigenobu; Kurisaka, Kenichi; Shimakawa, Yoshio*; Sago, Hiromi*
Proceedings of 2008 International Congress on Advances in Nuclear Power Plants (ICAPP '08) (CD-ROM), p.469 - 504, 2008/06
The conceptual design of an advanced sodium-cooled fast reactor (named JSFR) is currently carried out by the Japan Atomic Energy Agency (JAEA). In general, a large-scale nuclear reactor (approximately 1.5 GWe class) tended to increase in the number of loops (e.g., four loops in Super Ph
nix and APWR), while the JSFR adopts a two-loop cooling system that enables significantly reducing a plant construction cost resulting from decreasing in material amount of the nuclear steam supply system and in the reactor building volume. This paper describes technological feasibility of the two-loop cooling system in JSFR; especially, focused on flow-induced vibration of piping, safety analysis and decay heat removal system.
Sakamoto, Yoshihiko; Kubo, Shigenobu; Kotake, Shoji; Kamishima, Yoshio*
Proceedings of 2008 International Congress on Advances in Nuclear Power Plants (ICAPP '08) (CD-ROM), p.505 - 511, 2008/06
JSFR has the advanced loop type layout. In this paper, advantages of the advanced loop type reactor are presented in terms of reliability on the reactor structure. Compact design of the RV enables its manufacture in a factory which has high quality welding and precise machining. The RV has high reliability against thermal stress due to application of ring-shaped forgings around high stress parts. The in-vessel structures are simple, and this makes the approach to inspection targets easy. In JSFR, sodium boundary area is reduced significantly, which makes double-walled piping design easier, and reduces welding lines. So, the reactor structure of JSFR is desirable to inspect the in-vessel structures efficiently, and there is a prospect of reliable plant operation. Advanced technologies are also under development to inspect the structures immersed in sodium. In addition, the loop type reactor is suitable under severe earthquake condition as a result of comparative evaluation.
Kotake, Shoji; Mihara, Takatsugu; Kubo, Shigenobu; Aoto, Kazumi; Toda, Mikio*
Proceedings of 2008 International Congress on Advances in Nuclear Power Plants (ICAPP '08) (CD-ROM), p.486 - 495, 2008/06
Japan Atomic Energy Agency (JAEA) is now executing "Fast Reactor Cycle Technology Development (FaCT)" project in cooperation with the Japanese electric utilities. In the FaCT project, both the conceptual design study for JAEA sodium-cooled fast reactor (JSFR) and the developments of the innovative technologies are now implemented with paying attention to the consistency between the design and the innovative technologies. The current target is that the development will have been accomplished around 2015, after that a licensing procedure for JSFR demonstration reactor will be launched. This paper describes the design requirements, design characteristics of JSFR and evaluation on the performances for economic competitiveness. Furthermore, the current status of the key technology development for JSFR is briefly introduced.
Nakabayashi, Hiroki; Nagai, Toshihisa
Proceedings of 2008 International Congress on Advances in Nuclear Power Plants (ICAPP '08) (CD-ROM), p.2029 - 2035, 2008/06
We performed a basic design of the solvent extraction test system with centrifugal contactors for the engineering-scale hot examination facility to ensure the development and commercialization of the advanced aqueous reprocessing technology for fast reactor fuels in Japan. The system was designed to have the ability to operate two different flowsheet, the simplified solvent extraction method and the co-processing method, which we proposed as promising solvent extraction processes. In the design, various engineering issues, such as error of flow rate of reagent pumps and a dissolver solution feeder, error of chemical analyses and environmental temperature fluctuation, were delt with. For the design we modified and used the computer code "MIXSET-X" which was developed to simulate solvent extraction system by JAEA in 1999. The validity of the modified code was benchmarked by comparison with an engineering scale uranium test.
Sato, Fuminori; Nakamura, Hirofumi
Proceedings of 2008 International Congress on Advances in Nuclear Power Plants (ICAPP '08) (CD-ROM), p.2046 - 2050, 2008/06
Fukaya, Yuji; Nakano, Yoshihiro; Okubo, Tsutomu
Proceedings of 2008 International Congress on Advances in Nuclear Power Plants (ICAPP '08) (CD-ROM), p.127 - 135, 2008/06
Conceptual design studies on the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) have been performed at Japan Atomic Energy Agency (JAEA), and study on high conversion type core of FLWR (HC-FLWR) for MA recycling was started. Because HC-FLWR is the near-term technology with a conversion ratio around 0.85, it would be useful if HC-FLWR can recycle MA. The design criteria and requirements for the core are set as follows: The void reactivity coefficient should be negative value. And the discharge burn-up should be around 50 GWd/t. The parametrical survey with core burn-up calculations was performed. From these survey calculations, some feasible design ranges have been obtained for MA recycling core. The Puf enrichment of MOX fuel is 13 wt%. The core height is 116 cm. The fuel rod diameter is 9 mm. The core average void fraction is about 50%. In these reactor core specifications, around 2 wt% of Np or Am can be recycled in the HC-FLWR MOX fuel.
Usui, Shinichi; Mihara, Takatsugu; Obata, Hiroyuki; Kotake, Shoji
Proceedings of 2008 International Congress on Advances in Nuclear Power Plants (ICAPP '08) (CD-ROM), p.512 - 518, 2008/06
Refueling operation of sodium fast reactor (SFR) is one of major technical issue due to the chemical activities and opaqueness of sodium coolant properties in comparison with that of LWR. In the Japan Atomic Energy Agency (JAEA) sodium cooled Fast Reactor (JSFR) design study, the further reliable and rational fuel handling system (FHS) has been developing based on the experience of safe and reliable fuel handling operation in the existent SFR plants. Some of advanced concepts for the FHS have being studied in order to increase economic competitiveness further by attempting reduction of the amount of the material and the refueling time, and are scheduled to execute elemental tests and/or mock-up tests to confirm their feasibilities.
Rugama, Y.*; Blomquist, R.*; Brady Raap, M.*; Briggs, B.*; Gulliford, J.*; Miyoshi, Yoshinori; Suyama, Kenya
Proceedings of 2008 International Congress on Advances in Nuclear Power Plants (ICAPP '08) (CD-ROM), p.1391 - 1393, 2008/06
The OECD Nuclear Energy Agency (NEA) started dealing with criticality-safety related subjects during the nineteen-seventies. In the mid-nineties, several activities related to criticality-safety were grouped together into the Working Party on Nuclear Criticality Safety. This working party has since been operating and reporting to the Nuclear Science Committee. Six expert groups coordinate various activities ranging from experimental evaluations to code and data intercomparisons for the study of static and transient criticality behaviours. The paper describes current activities performed in this framework and the achievements of the various expert groups.
Naganuma, Masayuki; Ogawa, Takashi; Oki, Shigeo; Mizuno, Tomoyasu; Kubo, Shigenobu*
Proceedings of 2008 International Congress on Advances in Nuclear Power Plants (ICAPP '08) (CD-ROM), p.526 - 535, 2008/06
In FaCT project, sodium-cooled fast reactor with mixed-oxide fuel was selected as the primary candidate. Present study focused on effects of TRU composition on the design of JSFR. In a transitional stage from LWR to FBR, there is possibility for the JSFR fuel to have high MA content due to recycle of LWR spent fuel. That affects core and fuel designs (core reactivity, material property and so on). Thus, to evaluate the effects quantitatively, design studies for the JSFR cores with two TRU compositions were conducted, one was FBR multi-recycle composition with about 1wt% of MA and the other was LWR recycle one, for which 3wt% of MA was assumed as a typical composition. The results showed that the change from FBR multi-recycle composition to LWR recycle one leads to 10% increase of sodium void reactivity, 1-2% decrease of linear power limit and 5% extension of gas plenum length. As a result, effects of TRU composition on the core and fuel designs were indicated to be benign.
Nakanishi, Shigeyuki; Kubo, Shigenobu*; Takamatsu, Misao; Ikarimoto, Iwao*; Kato, Jungo*; Shimakawa, Yoshio*; Harada, Kiyoshi*
Proceedings of 2008 International Congress on Advances in Nuclear Power Plants (ICAPP '08) (CD-ROM), p.519 - 525, 2008/06
A self-actuated shutdown system (SASS) 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.