Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Uesawa, Shinichiro; Horiguchi, Naoki; Shibata, Mitsuhiko; Yoshida, Hiroyuki
Dai-22-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 6 Pages, 2017/06
no abstracts in English
Horiguchi, Naoki; Yoshida, Hiroyuki; Uesawa, Shinichiro; Abe, Yutaka*
Dai-22-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2017/06
Venturi scrubber is installed in nuclear power plant as a component of filtered venting system and used to remove small aerosols with fission products. There is, however, no method to estimate its decontamination performance in the assumed operating pressure range. In this study, we establish a method to estimate the decontamination performance based on mechanistic thermal hydrodynamic simulation code and report simulated results of the thermal hydrodynamics and decontamination performance in the pressure range in the pressure range in actual environments. With decrease in inlet pressure of the Venturi scrubber, gas flow velocity at the throat of it is suppressed, liquid flow velocity in a hole of it by self-priming changes and the decontamination factor changes were obtained. Also, with larger the aerosol diameter, the decontamination factor become larger was obtained.
Tanaka, Masaaki; Kobayashi, Jun; Nagasawa, Kazuyoshi*
Dai-22-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2017/06
In JAEA, a numerical simulation code named MUGTHES which can deal with conjugate heat transfer between the fluid and the structure parts has been developed for estimation of the thermal fatigue issue. In fundamental validation, the benchmark analysis was considered using the experiment of planar triple parallel jet sodium test (PLAJEST). Three specific experimental conditions at Vr=1, 1.56, and 5.56 were employed for the benchmark analyses according to the knowledge in the literatures. Through the benchmarks, applicability of the large eddy simulation (LES) approach with the standard Smagorinsky model in MUGTHES to simulate thermal striping phenomena was potentially confirmed and issues to be modified in the future works were indicated.
Yada, Hiroki; Takaya, Shigeru; Wakai, Takashi; Machida, Hideo*
no journal, ,
In the in-service inspection of sodium cooled fast reactors, an examination by continuous leak monitoring is considered for components constituting a sodium boundary. The continuous leak monitoring examinations are premises in a LBB (leak before break) being established. In previous LBB evaluation, the axial crack of the elbow flank has been evaluated, because in the piping system of the fast reactor with high thermal expansion, high stress is generated in the elbows flank. However, the LBB evaluations for circumferential cracks are important in the point of continuous leak monitoring in the in-service inspection. In this study, the load conditions in the LBB assessments for circumferential cracks were examined and LBB characteristics of 1 class 1 pipes of the prototype fast breeder reactor "Monju" were evaluated as an example.
Kurihara, Akikazu; Umeda, Ryota; Shimoyama, Kazuhito; Kikuchi, Shin
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 Na-Fe composite oxidation type corrosion with flow (COCF) in an environment marked by high temperature and high-alkali (reaction jet) due to sodium-water reaction. The authors derived new wastage correlations from COCF and LDI data based on influencing factors which were formed on the periphery of an adjacent tube. In this report, the authors established that the new wastage correlations were applicable to each tube of tube bundle in sodium-water reaction test with straight vertical tube bundle under practical steam generator operation condition, and the time progress of wastage was qualitatively investigated for the two penetrated tubes in the period including the blowdown.
Yoshikawa, Ryuji; Imai, Yasutomo*; Tanaka, Masaaki; Ohshima, Hiroyuki
no journal, ,
Accurate evaluation of three-dimensional temperature distribution is essentially required for the structural integrity analysis of large-sized straight tube steam generator (SG) of an advanced sodium-cooled fast reactor (SFR). Numerical simulation system TSG to analyze three-dimensional thermal-hydraulics in the straight tube SG has been developed. The three-dimensional simulation on sodium side by the CFD code FLUENT with porous body model was coupled with the multi-channel simulation on water/steam side. Fundamental validation of TSG code with the 1MWt straight tube SG test was performed. Temperature distribution of large-sized SG under the condition with plugged tubes was analyzed and the temperature deviation of heat transfer tubes was evaluated. Through the numerical simulation, local temperature increase near the plugged tubes was quantitatively evaluated, and the applicability of the simulation results to the structural integrity analysis of straight tube SG was indicated.
Aoyagi, Mitsuhiro; Takata, Takashi; Ohno, Shuji; Uno, Masayoshi*
no journal, ,
We have been conducting an R&D project to develop detailed numerical simulation methods for sodium fire consequences which is one of risks to put heat and pressure loads to containment vessel of sodium-cooled fast reactor. We carried out sodium fire analysis of an upward spray experiment for integrated validation of the analysis code. This paper describes detailed influencing factors in the validation.
Uchibori, Akihiro; Watanabe, Akira*; Takata, Takashi; Ohshima, Hiroyuki
no journal, ,
For assessment of the wastage environment under tube failure accident in a steam generator of sodium-cooled fast reactors, a mechanistic computer code called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction has been developed. The original SERAPHIM code is based on the finite difference method. In this study, unstructured mesh-based numerical method was developed and introduced into the SERAPHIM code to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Validity of the unstructured mesh-based SERAPHIM code was investigated through the analysis of an underexpanded jet experiment. The calculated pressure profile showed good agreement with the experimental data. Numerical analysis of water vapor discharging into liquid sodium was also performed. It was demonstrated that the proposed numerical method could be applicable to evaluation of the sodium-water reaction phenomenon.