Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Yamano, Hidemasa; Tobita, Yoshiharu; Sato, Ikken
Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 13 Pages, 2010/10
A Level 2 PSA methodology is developed for the risk evaluation of sodium-cooled fast reactors (SFRs). For this purpose, the phenomenological event tree is developed as well as a technical basis to quantify the probability of event sequences in the Level 2 PSA, focusing on the transition phase in an unprotected loss of flow (ULOF) accident in this paper. In addition, dominant factors are also identified through parametric analyses using the SIMMER-III code. The experimental findings on the fuel discharge behavior and its driving force formation were summarized from the CABRI and EAGLE experiments. Using past experimental evidences, furthermore, the experimental database is developed to quantify the probability of the Level 2 PSA.
Honda, Kei; Ohira, Hiroaki; Sotsu, Masutake; Yoshikawa, Shinji
Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 12 Pages, 2010/10
In this study, we calculated the thermal hydraulics of the upper plenum of MONJU by the detailed analysis model using commercial FVM code, FrontFlow/Red. The present analysis model simulates all structures with high resolution meshes. The 1st order upwind and 2nd order central difference scheme were applied to the advection and diffusion terms, respectively. And RNG 
-
model was applied to turbulence modeling. These calculation results indicated that the structures installed in the plenum except for UIS did not affect largely to the temperature and velocity, the flow characteristics in the present results had similar tendencies with porous media approached applied to the UCS region and that the difference between the temperature measured in the UCS region and that of SA outlets is relatively small.
Ohno, Shuji; Seino, Hiroshi; Miyahara, Shinya
Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 12 Pages, 2010/10
This research has compiled technical basis which is necessary to carry out a probabilistic safety assessment (Level 2 PSA) for a sodium-cooled fast reactor. The accumulated technical information consists of experimental and analytical information which help ones to understand the loading to a containment vessel, as well as the existing information on dominant factors of important ex-vessel phenomena.
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.
Liu, W.; Takase, Kazuyuki
Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 10 Pages, 2010/10
A system for measuring the surface temperature and the surface heat flux of a heating block was developed that has no need for a sensor set on the surface. The system consisted of two parts: (1) inner block temperatures were measured using micro-thermocouples; (2) using the measured temperatures, an inverse heat conduction problem was solved to get the surface heat flux and surface temperature. For the inner block temperature measurement, special T-type micro thermocouples with a common positive pole were developed. A total of 10 thermocouples were set at a depth of 3.1 micro meter beneath the boiling surface, along a radius of 5 mm. The developed system was verified using a pool boiling experiments. Experiments were performed at atmospheric pressure. The experiments showed that the developed special T-type micro thermocouples can successfully measure the temperature change in a boiling process. By using the measured temperature, a semi-infinite inverse heat conduction problem was solved to get the surface heat flux and surface temperature. The change in surface heat flux and surface temperature in a bubble cycle was derived.
Nishimura, Masahiro; Kamide, Hideki; Sugiyama, Kenichiro*; Otake, Shiro*
Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 11 Pages, 2010/10
The purpose of this study is to understand oxidation behavior of sodium precisely for FR safety against sodium combustion. It was recognized that dendritic oxide took an important role for the combustion reaction such as supplying the sodium to the reaction interface. In this study, we proposed a mechanistic model of supplying liquid sodium through the dendritic oxide based on the observation result of the growing behavior of dendritic oxide during combustion. In this model we made an attention to the kinds of chemical compounds. The formation of sodium peroxides can provide the sodium supplying route in the dendritic oxide. On the other hand the formation of sodium monoxide will block sodium supplying. The kinds of chemical compounds were decided by the Gibbs's free energy of thermodynamics in the reaction field such as temperature and oxygen concentration. This mechanistic model can explain the oxidation behavior consistently with the observation results.
Tobita, Yoshiharu; Yamano, Hidemasa
Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 11 Pages, 2010/10
As part of level-2 PSA methodology development for sodium cooled fast reactors, the phenomenological event tree for the Protected Loss of Heat Sink (PLOHS) is constructed for JSFR (Japan Sodium-cooled Fast Reactor). The early stage of PLOHS accident, where the heating of coolant, coolant boiling and uncovering of core occur, is analyzed by the ARGO code. The event progression in the subsequent core degradation phase, where the meltdown of core and the power transient by re-criticality occurs, is analyzed by the APPLOHS and SIMMER-III codes. The dominant factors, which affect the event progression in PLOHS, are recognized through these analyses and by considering the phenomenological event progression. The event tree of PLOHS is developed assigning these dominant factors as the headings. For each of the headings, available information for judgment of branching probability are reviewed and integrated as database for level-2 PSA.
Nakai, Ryodai; Suzuki, Toru; Kamiyama, Kenji; Seino, Hiroshi; Koyama, Kazuya*; Morita, Koji*
Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 12 Pages, 2010/10
The evaluation technology of Level-2 PSA for Sodium-Cooled Fast Reactors (SFRs) was established in order to systematically assess the core damage sequences. In addition to the existing computational tools for Level-2 PSA, the computational tools, MUTRAN and SIMMER-LT were developed for core material relocation phase. Also the analytical models, CORCON and VANESA, were improved based on newly performed experiments for the ex-vessel phase taking into account the feature of SFRs. The technical information was compiled as technical database used in the construction and quantification of level-2 PSA event trees for SFRs. The technical basis was established for the Level-2 PSA for SFRs.
Uchibori, Akihiro; Ohshima, Hiroyuki; Watanabe, Akira*
Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 10 Pages, 2010/10
To evaluate the behavior of the reaction jet and its effect on the adjacent tubes under the tube failure accident in a steam generator of sodium cooled fast reactors, a computer program called SERAPHIM for the compressible multiphase flow involving sodium-water chemical reaction has been developed. In this study, numerical analysis of supersonic gas jets into liquid pools with or without a chemical reaction was performed as a part of validation of the SERAPHIM program. In the analysis of the horizontal air jet into the water pool, the behavior of the jet was reproduced very well. The vertical injection of the chlorine jet into the mixture of the sodium and the sodium chloride was also analyzed. The numerical results showed that the injected gas disappeared at a certain height by the chemical reaction. The calculated penetration length agreed with the experimental data.
Sato, Ikken; Tobita, Yoshiharu; Yamano, Hidemasa
Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 12 Pages, 2010/10
As part of Level-2 PSA methodology development for sodium cooled fast reactors (SFR), event trees for the initiating phase (IP) of Anticipated Transient Without Scram (ATWS) are constructed. ULOF (Unprotected Loss of Flow), UTOP (Unprotected Transient Overpower) and ULOHS (Unprotected Loss of Heat Sink) are selected as typical and important accident categories. Based on the state-of-the-art knowledge, the headings of these event trees are selected so that dominant factors in accident consequences can be represented appropriately. For each of the headings, available information for judgment are reviewed and integrated as database for Level-2 PSA. It is clarified that the headings of ULOF, for which experimental database and evaluation models have been reasonably established, can be commonly applied to certain part of the different accident categories. While, some points specific for UTOP and ULOHS are identified. ULOHS, in which significant heat up of the primary system is expected before start of the core disruption, necessitates an additional event tree before the core disruption providing various boundary conditions for the core disruption process.
Koshizuka, Seiichi*; Morita, Koji*; Arima, Tatsumi*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Naito, Masanori*; Shirakawa, Noriyuki*; Okada, Hidetoshi*; Uehara, Yasushi*; et al.
Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 11 Pages, 2010/10
In this paper, FY2009 results of the COMPASS code development are reported. Validation calculations for melt freezing and blockage formation, eutectic reaction of metal fuel, duct wall failure (thermal-hydraulic analysis), fuel pin failure and disruption and duct wall failure (structural analysis) are shown. Phase diagram calculations, classical and first-principles molecular dynamics were used to investigate physical properties of eutectic reactions: metallic fuel/steel and control rod material/steel. Basic studies for the particle method and SIMMER code calculations supported the COMPASS code development. COMPASS is expected to clarify the basis of experimentally-obtained correlations used in SIMMER. Combination of SIMMER and COMPASS will be useful for safety assessment of CDAs as well as optimization of the core design.
Suzuki, Soju
no journal, ,