Numerical analysis of windowless target in accelerator driven system by use of TPFIT

Yoshida, Hiroyuki; Suzuki, Takayuki*; Takase, Kazuyuki

Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 7 Pages, 2010/11

Sodium experiments on decay heat removal system of Japan sodium cooled fast reactor; Start-up transient of decay heat removal system

Kamide, Hideki; Kobayashi, Jun; Ono, Ayako; Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Watanabe, Osamu*

Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 8 Pages, 2010/11

Fully natural circulation system is adopted in a decay heat removal system (DHRS) of Japan Sodium Cooled Fast Reactor (JSFR). The JSFR has two loops of the main heat transport system in order to reduce number of components and the construction cost. The DHRS of JSFR consists of one DRACS and two PRACS, which have a heat exchanger in a primary-side inlet plenum of IHX in each loop. Sodium experiments were carried out for heat transfer characteristics of a sodium-sodium heat exchanger of PRACS and also start-up transient of the DHRS loop. Heat transfer coefficient on the tube outer surface was in good agreement with a conventional correlation under operation condition in the reactor. The transient experiments for the start-up of DHRS loop showed that smooth increase of natural draft in the air duct followed by the sodium flow rate in the DHRS loop. Some delay of the flow rate increase was recognized in the DHRS loop as compared with that of the natural draft in the air cooler.

SIMMER-III analysis of eagle-1 in-pile tests focusing on heat transfer from molten core material to steel-wall structure

Toyooka, Junichi; Kamiyama, Kenji; Konishi, Kensuke*; Tobita, Yoshiharu; Sato, Ikken

Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 7 Pages, 2010/11

In this study, the heat flux from the molten core materials to the outer surface of the inner duct (the pool-to-duct heat flux) was evaluated based on all the EAGLE-1 in-pile experiments available. Through the evaluation, it was understood that the pool-to-duct heat flux was so high in all the in-pile experiments that the duct wall failed without coolant boiling in its behind. It was also indicated that the presence of steel in the molten core mixture played a key role in this high heat flux. Application of the SIMMER-III code for these tests suggested that some model improvements were necessary to simulate pool-to-duct heat transfer behavior in a mechanistic way.

Development of technical database in the unprotected events for level 2 PSA of sodium-cooled fast reactors

Yamano, Hidemasa; Tobita, Yoshiharu; Sato, Ikken

Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 10 Pages, 2010/11

Current research activities on thermal hydraulics in Japan Sodium-cooled Fast Reactor (JSFR)

Kamide, Hideki; Hayafune, Hiroki

Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 7 Pages, 2010/11

Design study and related researches are carried out for Japan Sodium-cooled Fast Reactor (JSFR). Several innovative technologies, e.g., compact reactor vessel, two-loop system, fully natural circulation decay heat removal, and recriticality free core, have been investigated in order to reduce construction cost and to achieve higher level of reactor safety. Preliminary evaluations of innovative technologies to be applied to JSFR are undergoing. Here progress of design study is introduced. Then, research and development activities on the thermal hydraulics related to the innovative technologies are briefly reviewed.

Investigation of temperature gradient characteristics for thermal stratification interface in upper plenum of fast reactors

Ohno, Shuji; Sugahara, Akihiro*; Oki, Hiroshi*; Ohshima, Hiroyuki

Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 6 Pages, 2010/11

Parametric analyses were carried out as numerical experiments to clarify the basic characteristics of thermal stratification behavior in upper plenum of fast reactors. The analyses were performed changing hypothetically the analytical conditions of sodium flow rate, sodium temperature, and the size of analyzed upper plenum, by using a commercial CFD code FLUENT Ver. 6.3 and the RNG k- turbulence model. The results provided the suggestion that the averaged sodium ascending velocity in the reactor upper plenum region and the sodium temperature difference before and after the transient initiation would be the dominant factors to determine temperature gradient of thermal stratification interface. Further, it was implied that appropriate spatial mesh arrangement in vertical direction around the stratification interface is significant to obtain the accurate numerical solution of interface temperature gradient.