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Liu, W.
Dai-20-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.391 - 392, 2015/06
Subcooled flow boiling is a boiling that begins and develops even though the mean enthalpy of liquid phase is lower than saturation. This forced convective boiling is one of the most efficient ways for the removal of high heat flux. It is widely used in the high heat flux components such as nuclear reactor cores, accelerator targets and fusion reactor components. The thermal outputs of these systems are restricted by Critical Heat Flux(CHF). Because of the importance of the CHF, lots of researches, including both experimental and mechanistic modelling, have been performed. However, the CHF prediction for rod bundles still remains unsolved. As the first step for the CHF prediction in rod bundles, in this paper, we tried to predict the CHF in annulus, which is the most basic flow geometry simplified from a rod bundle. We performed the CHF prediction by using liquid sublayer dryout model, combined with Nouri single phase velocity distribution correlation for annulus. The results show that the CHF in annulus can be predicted in an accuracy of about 
20%.
Jiao, L.; Yoshida, Hiroyuki; Takase, Kazuyuki
Dai-20-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.387 - 388, 2015/06
In Japan Atomic Energy Agency, the detailed two-phase flow analysis code TPFIT has been developed to simulate and evaluate two-phase flow characteristics in nuclear systems. In this study, a numerical simulation of bubbly flow in a vertical square channel was performed to validate the applicability of TPFIT code on bubbly flow simulation. By checking bubble distribution development in the flow direction, the calculation of the forces acting on bubbles was validated through comparing simulation results and experimental results from Matos et al. (2004). Comparisons between the experimental and numerical data revealed, in general, good agreement except serious bubble coalescence appeared in numerical simulation.
Tanaka, Masaaki
Dai-20-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.55 - 58, 2015/06
Numerical estimation method for high cycle thermal fatigue on a structure has been developed in JAEA. In development of numerical simulation codes and application of the codes to plant design, implementation of verification and validation (V&V) is indispensable. A procedure called as V2UP (Verification and Validation plus Uncertainty quantification and Prediction) has been made by referring to the existing guidelines on V&V. The PIRT (Phenomena Identification and Ranking Table) method based on the nine-step process used by the USNRC for the next generation nuclear plant development was employed at the first step of the V2UP. Through the first step of the V2UP with PIRT method, the conceptual model for the numerical estimation of high cycle thermal fatigue was successfully constructed.
Umeda, Ryota; Kurihara, Akikazu; Shimoyama, Kazuhito
no journal, ,
Overheating tube rupture of adjacent tubes arises from water/steam leak in steam generators of sodium-cooled fast reactors. It is very important to predict the tube wall stress (tube wall temperature) with a high degree of accuracy on evaluation of overheating tube rupture, and is crucial to estimate quantitatively material strength standard which is one of the major influencing factor. Therefore, in present study, the authors carried out tube rupture experiments with rapidly-heating which were simulated the tube thermally-affected by sodium-water reaction jet, and evaluated quantitatively failure hoop stress and failure time. Then, the authors confirmed that existing stress strength standard was applicable to thin diameter and thick-walled single tube in case of sodium-water reaction exceeding 1300
C under practical steam generator operation conditions.
Shimoyama, Kazuhito; Kurihara, Akikazu; Kikuchi, Shin; Umeda, Ryota
no journal, ,
Self-wastage comes from water/steam leak through the penetrating crack caused in the steam generator tube of sodium-cooled fast reactor. When the self-wastage proceeds to inside wall of tube, breach area and water leak rate will be larger, then, 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 the 1st report, the authors have performed the self-wastage experiments for the pinhole type micro crack. In this report, fatigue crack type self-wastage experiments were carried out to evaluate the effect of wastage form/geometry and water leak rate, it was confirmed that initial defect geometry, such as pinhole and fatigue crack, does not strongly influence to self-wastage rate.
Nakamura, Hideo; Arai, Kenji*; Oikawa, Hirohide*; Umezawa, Shigemitsu*; Onuki, Akira*; Fujii, Tadashi*; Nishi, Yoshihisa*; Abe, Yutaka*; Sugimoto, Jun*; Koshizuka, Seiichi*; et al.
no journal, ,
no abstracts in English
Nakamura, Hideo; Yamamoto, Yasushi*; Yamada, Hidetomo*; Nagayoshi, Takuji*; Nishi, Yoshihisa*
no journal, ,
no abstracts in English
Kurihara, Akikazu; Kikuchi, Shin; Umeda, Ryota; Shimoyama, Kazuhito
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 applicability of new wastage correlations were confirmed by using data of sodium-water reaction test with tube bundle under practical steam generator operation condition.
Nishi, Yoshihisa*; Arai, Kenji*; Oikawa, Hirohide*; Fujii, Tadashi*; Umezawa, Shigemitsu*; Yamada, Hidetomo*; Nakamura, Hideo
no journal, ,
no abstracts in English
Onuki, Akira*; Umezawa, Shigemitsu*; Yamada, Hidetomo*; Nishi, Yoshihisa*; Arai, Kenji*; Oikawa, Hirohide*; Fujii, Tadashi*; Nakamura, Hideo
no journal, ,
no abstracts in English
Fujii, Tadashi*; Arai, Kenji*; Oikawa, Hirohide*; Umezawa, Shigemitsu*; Nishi, Yoshihisa*; Nakamura, Hideo
no journal, ,
no abstracts in English
Oikawa, Hirohide*; Arai, Kenji*; Fujii, Tadashi*; Umezawa, Shigemitsu*; Nishi, Yoshihisa*; Nakamura, Hideo
no journal, ,
no abstracts in English