Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Yamashita, Susumu; Takase, Kazuyuki
Dai-19-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.153 - 156, 2014/06
no abstracts in English
Tanaka, Masaaki; Ohno, Shuji
Dai-19-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.247 - 250, 2014/06
A procedure combined V&V of the code and numerical prediction process called as V2UP (Verification and Validation plus Uncertainty quantification and Prediction) has been developed by referring to the existing guidelines. By using numerical results by MUGTHES for the WATLON experiment which was a water experiment to investigate thermal mixing phenomena in a T-junction piping system, applicability of the GCI estimation method and the area validation metric (AVM) method and the modified one (MAVM) in the V2UP were examined. Through the examinations, it was found that the GCI estimation by using the modified least-square version was applicable and the AVM and the MAVM methods were applicable if transient data were obtained in the experiment.
Kikuchi, Shin; Seino, Hiroshi; Ohno, Shuji
no journal, ,
For the purpose of elucidating the mechanism of sodium-concrete reaction in SFR, kinetic study of the sodium-hydroxide (NaOH)-silica (SiO
) reaction was carried out by Differential Scanning Calorimetry (DSC). The parameters, including melting point of NaOH, phase transition temperature of NaOH and SiO, and NaOH-SiO reaction temperature were identified from DSC curves. From visualization test, sample eruption was observed during reaction. It was found that rate of NaOH-SiO reaction was quite fast from DSC curves, which was similar with that of the reaction between NaOH and aggregate of practical concrete. Thermal analysis results indicated that NaOH-SiO reaction could occur in the timeframe of sodium-concrete reaction.
Honda, Kei; Mori, Takero; Sotsu, Masutake; Ohira, Hiroaki
no journal, ,
Thermal-hydraulics analyses of the reactor upper plenum of Monju, Japanese prototype of FBR, were performed in IAEA/Monju-CRP from 2008 to 2012. However, detail temperature and flow rate conditions of the inlets were required for an accurate analysis. In this paper we re-evaluated the inlet boundary condition (subassembly outlets) and performed another thermal-hydraulics analysis with Star-CCM+. The surface of the structures in the upper plenum was precisely modeled. The structures included a fuel transfer machine, in-vessel racks, flow-guide tubes, etc. The result was following: the structure didn't have large influence to the temperature distribution, and the analysis result of the temperature distribution on the thermocouple plug had some difference from the test result.
Mori, Takero; Sotsu, Masutake; Ohira, Hiroaki
no journal, ,
no abstracts in English
Kurihara, Akikazu; Umeda, Ryota; Shimoyama, Kazuhito; Kikuchi, Shin; Ohshima, Hiroyuki; Narabayashi, Tadashi*
no journal, ,
Wastage phenomena on adjacent tubes (target-wastage) arise from water/steam leak in steam generators of sodium-cooled fast reactors. Target-wastage is likely to be caused by liquid droplet impingement erosion (LDI) and flow-accelerated corrosion (FAC) in an environment marked by high-temperature and high-alkali (reaction jet) due to sodium-water reaction. The authors carried out flow-accelerated corrosion experiments as a part of phenomena clarification experiments for target-wastage by using tube material under high-temperature sodium-hydroxide and sodium monoxide conditions which are mainly generated by sodium-water reaction. New wastage correlations were derived from LDI and FAC data based on influencing factors which were formed on the periphery of an adjacent tube, and were confirmed those applicability to water leak event in this report.
Deguchi, Yoshihiro*; Tamura, Kenta*; Muranaka, Ryota*; Kusano, Koji*; Takata, Takashi*; Kikuchi, Shin; Kurihara, Akikazu
no journal, ,
In a sodium-cooled fast reactor (SFR), liquid sodium is used as a heat transfer fluid because of its excellent heat transport capability. One of the design basis accidents of the SFR is the water leakage into the liquid sodium flow by a breach of heat transfer tubes in a steam generator. Therefore the study on sodium-water chemical reactions is of paramount importance for safety reasons. This study aims to clarify the sodium-water reaction mechanisms using laser diagnostics. The sodium-water, sodium-oxygen and sodium-hydrogen counter-flow reactions were measured using laser diagnostics such as Raman, absorption and photo-fragmentation spectroscopies. The measurement results show that the main product of the sodium-water reaction is NaOH. The sodium-water reaction rate is slower than that of the sodium-oxygen reaction and hydrogen does not react noticeably with sodium.
Shimoyama, Kazuhito; Kurihara, Akikazu; Kikuchi, Shin; Umeda, Ryota; Ohshima, Hiroyuki
no journal, ,
Corrosion may occur on the tube surface due to chemical reaction between sodium and water (self-wastage) if water/steam leak proceed through the penetrating crack caused in the steam generator tube of sodium-cooled fast reactor. When the self-wastage goes along up to inside wall of tube, water leak rate will be larger and it will be likely to spread the affected area caused by sodium-water reaction. It is very important to clarify the self-wastage behavior for locally affected region and detection of the water leak in real plant. In this study, we performed the self-wastage experiments under high sodium temperature condition to evaluate the effect of wastage form/geometry and water leak rate on self-wastage rate in the pinhole type micro crack.
Uchibori, Akihiro; Ohshima, Hiroyuki
no journal, ,
A mechanistic computer program called SERAPHIM to calculate compressible multicomponent multiphase flow with sodium-water chemical reaction has been developed as one of the numerical evaluation methods for heat transfer tube failure accident in a steam generator of sodium-cooled fast reactors. In this study, applicability of the SERAPHIM program was investigated through the analysis of the experiment on vertical water vapor discharging into liquid sodium under the actual condition of the steam generator. The high-velocity region of the entrained liquid droplets was formed around the water vapor jet. The impingement position of the liquid droplets on the target tube located above the water vapor discharging tube agreed with the position of the wastage mark confirmed in the experiment. The temperature distribution measured around the reacting jet was also successfully reproduced by the SERAPHIM program.
Nagatake, Taku; Takase, Kazuyuki; Yoshida, Hiroyuki; Kurata, Masaki
no journal, ,
no abstracts in English
Mizutani, Jun*; Ebara, Shinji*; Hashizume, Hidetoshi*; Yamano, Hidemasa
no journal, ,
It is quite likely that a complex turbulent flow field and large pressure fluctuation induced by separation vortex shed from the intrados of the elbows are seemed to appear in the cold leg piping of the primary cooling system of Japan Sodium-cooled Fast Reactor. This study researched the influence of the inflow condition upon the flow separation especially in the 3rd elbow. In this study, the inflow condition to the 3rd elbow was imposed by changing the distance between the 2nd and 3rd elbows from 6.4D (original design) to 9.4D. The visualization experiment showed that the flow separation appeared in the intrados of the 3rd elbow as was the case with the original design and the separated regions became larger than that in the original one. This is because a swirling flow observed at the inlet of the 3rd elbow became weaker than that in the original case.
Ohshima, Hiroyuki; Kurihara, Akikazu; Yamaguchi, Akira*; Takata, Takashi*; Narabayashi, Tadashi*; Deguchi, Yoshihiro*
no journal, ,
When a heat transfer tube is failed in a steam generator (SG) of a sodium-cooled fast reactor (SFR), pressurized water and/or water vapor leaks into liquid sodium surrounding the tube and forms a reacting jet with high temperature and high alkali. This reacting jet might cause the secondary failure of adjacent heat transfer tubes due to wastage or over-heating tube rapture resulting in undesirable failure propagation. Therefore, the sodium-water reaction phenomenon (SWR) is one of the most important issues for the design and safety assessment of SFRs. The authors have carried out systematic experiments for the elucidation of SWR and developed a new multi-physics numerical simulation system which is based on mechanistic and theoretical modeling of SWR rather than empirical modeling and can contribute to detailed and quantitative evaluations of SWR in any types of SGs. This paper summarizes the results of four years' R&D activities.