Numerical quantification of self-wastage phenomena in sodium-cooled fast reactor

Jang, S.*; Takata, Takashi; Yamaguchi, Akira*; Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 8 Pages, 2014/11

Numerical quantification of the self-wastage phenomenon has been carried out using a multi-dimensional computational code: SERAPHIM. The width of the completely enlarged crack was investigated in this study. Several steps of numerical calculations were devised to reproduce transient self-wastage phenomenon caused by Sodium Water Reaction (SWR). In the analyses, 2-dimensional calculation was carried out to obtained thermal hydraulic properties in the reaction zone. The wastage amount was evaluated based on hypothetical Arrhenius equation by using the temperature and molar concentration of Sodium hydroxide. New analytical grid was created by exchanging the solid cells to fluid cells in the reaction based on the wastage amount evaluation. These series of procedure have been repeated. The width and the shape of the enlarged crack showed good agreement with the experimental results.

U-RANS simulation of elbow flow with a 1/7 scale experimental loop simulating JSFR cold-leg piping

Yamano, Hidemasa; Kaneko, Tetsuya*; Aizawa, Kosuke; Tanaka, Masaaki; Ebara, Shinji*; Hashizume, Hidetoshi*

Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 10 Pages, 2014/11

In this study, Unsteady Reynolds Averaged Navier-Stokes equations (U-RANS) approach with the Reynolds Stress Model using a commercial computational fluid dynamics code has been applied to the flow through single and double short-elbows under a high Reynolds number condition (1.010). This numerical simulation has utilized a 1/7 scaled water experiment simulating the JSFR cold-leg piping in order to investigate the flow pattern in a three-dimensionally connected double elbow. As a result, validity and applicability of the U-RANS numerical simulation method were confirmed by comparison between the numerical and experimental results. The simulation has also shown that unsteady flow fields including vortex structures in the 1st elbow were very different from that in the 2nd elbow. This is because horseshoe vortexes generated in the 1st elbow downstream enter sideways the 2nd elbow. In addition, the simulation has indicated that the influence of distance between two elbows in the double elbow flow addressed in this paper was small.

Study on flow in the subchannels of pin bundle with wrapping wire

Nishimura, Masahiro; Hiyama, Tomoyuki; Kamide, Hideki; Ohshima, Hiroyuki; Nagasawa, Kazuyoshi*; Imai, Yasutomo*

Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 7 Pages, 2014/11

The Effect of profile of inlet velocity on the pressure fluctuation on the inside wall of short-elbow

Ono, Ayako; Tanaka, Masaaki; Kobayashi, Jun; Kamide, Hideki

Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 7 Pages, 2014/11

Experimental study on vortex cavitation in scaled upper plenum model of Japan Sodium-cooled Fast Reactor, 1; Evaluation of circulation and vortex cavitation occurrences using vortex model

Ezure, Toshiki; Ito, Kei; Kameyama, Yuri*; Hagiwara, Hiroyuki*; Kamide, Hideki

Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 7 Pages, 2014/11

Numerical study on inert gas behavior in fast reactor primary coolant system; Inert gas accumulation at HPP and consideration of gas elimination system

Takata, Takashi*; Konaka, Yuji*; Yamaguchi, Akira*; Ito, Kei; Ohno, Shuji; Ohshima, Hiroyuki

Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 7 Pages, 2014/11

In this paper, a model of bubble transport at entrance nozzle has been developed based on theoretical and computational methods. For this purpose, a three-dimensional analysis of dynamic bubbles behavior has been carried out by using one-way bubble tracking method which is specified to a bubble flow under a low void fraction. A commercial CFD tool, FLUENT Ver. 6.3.26 is used for three-dimensional flow field analyses. Then the gas accumulation at a high pressure plenum (HPP) of sodium-cooled fast rector has been quantitatively-assessed by implementing the developed model in SYRENA code. Furthermore, gas elimination (bubble removal) systems have been considered and the quantitatively-assessment has been carried out in order to investigate the effect of each measure to the accumulation of gas at the HPP.

Development of a wastage environment evaluation model for a sodium-water reaction analysis code SERAPHIM

Uchibori, Akihiro; Ohshima, Hiroyuki

Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 6 Pages, 2014/11

A computer code called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction has been developed to evaluate wastage environment under tube failure accident in a steam generator of sodium-cooled fast reactors. In this study, the numerical model for liquid droplet entrainment and its transport was developed. The applicability of the model was investigated through the analysis of the basic experiment. It was demonstrated that our numerical model could reproduce the time to end of entrainment and the pressure variation during the occurrence of entrainment.

International reviews on safety design criteria and development of safety design guidelines for Generation-IV Sodium-cooled Fast Reactor

Okano, Yasushi; Nakai, Ryodai; Kubo, Shigenobu

no journal, , 

The "SDC Phase I Report" was issued in May 2013 and was circulated for review to international organizations and regulatory bodies of the SFR developing countries under the GIF. The important feedbacks via the international reviews are mainly about "clarification of the requirements and definitions", "further specification of the detailed requirements specific to the SFR system", and "common/different points between the SFR and LWR". The second term activity to develop "Safety Design Guidelines (SDG)" was started. The SDG is below the SDC in a hierarchy of safety standards, in analogy with the IAEA NS-G series below the IAEA SSR 2/1. There are two SDGs to be developed by the GIF. The 1st SDG is on "Safety Approach" and the 2nd SDG is on "Key Structures, Systems and Components (SSCs)". The time frame of the summarization will be in 2016.