Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Tanaka, Masaaki; Murakami, Satoshi*
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 10 Pages, 2017/07
Thermal striping on the core instrumentation plate (CIP) around the primary control rod (PCR) and backup control rod (BCR) channels and the radial blanket fuel assemblies (RBAs) may be caused. Since the interaction between neighbor areas exists in the UIS and the cold sodium flowing from the RBA is affected by the external flow around the UIS, a spatial connection method consisting of the numerical model for the whole upper plenum and the local target area has been developed. The numerical results were compared with the experimental results to confirm applicability of the method to the practical problem. And, sensitivity of mesh arrangement to the numerical results was discussed by using wide and narrow area models with two different spatial resolutions in each model. Through the examinations, appropriate local model for the spatial connection mothed could be proposed.
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.
Tanaka, Masaaki; Ohno, Shuji; Ohshima, Hiroyuki
Nuclear Engineering and Design, 299, p.174 - 183, 2016/04
Times Cited Count：4 Percentile：36.27(Nuclear Science & Technology)A procedure called as V2UP (Verification and Validation plus Uncertainty quantification and Prediction) was made by referring to the existing guidelines on V&V and the methodologies of the safety assessment (CSAU, ISTIR, EMDAP). The V2UP consisted of five components as follows: (1) phenomena analysis with the Phenomena Identification and Ranking Table (PIRT) method, (2) implementation of the V&V, (3) design and rearrangement of experiments for the V&V, (4) uncertainty quantification in each problem and integration of uncertainties and (5) numerical prediction (estimation) for the target issue. Although the complete application of the procedure has not been performed at this moment, a flow chart of the V2UP procedure was described in this paper with recent results of the examinations.
Tanaka, Masaaki; Miyake, Yasuhiro*
Nihon Kikai Gakkai M&M 2015 Zairyo Rikigaku Kanfarensu Koen Rombunshu (Internet), 3 Pages, 2015/11
A prototype coupling method consisting of the fluid-structure thermal interaction simulation code MUGTHES and the structural thermal stress analysis code FINAS with interface program MUFIN has been developed in order to estimate the thermal fatigue in the SFRs. As a fundamental validation of the coupled method, it was applied to the water experiment for thermal mixing phenomena in a T-junction piping system. In the experiment, thermal interaction between the fluid and the structure made of aluminum installed to the branch pipe side wall was considered. Through the numerical simulations, applicability of the coupled method was confirmed.
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.
Tanaka, Masaaki; Ohno, Shuji; Ohshima, Hiroyuki
Proceedings of OECD/NEA & IAEA Workshop on Application of CFD/CMFD Codes to Nuclear Reactor Safety and Design and their Experimental Validation (CFD4NRS-5) (Internet), 14 Pages, 2014/09
A procedure called as V2UP (Verification and Validation plus Uncertainty quantification and Prediction) was made by referring to the existing guidelines on V&V and the methodologies of the safety assessment (CSAU, ISTIR, EMDAP). The V2UP consisted of five components as follows: (1) phenomena analysis with the Phenomena Identification and Ranking Table (PIRT) method, (2) implementation of the V&V, (3) design and rearrangement of experiments for the V&V, (4) uncertainty quantification in each problem and integration of uncertainties and (5) numerical prediction (estimation) for the target issue. Although the complete application of the procedure has not been performed at this moment, a flow chart of the V2UP procedure was described in this paper with recent results of the examinations.
Tanaka, Masaaki; Miyake, Yasuhiro*
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 13 Pages, 2014/07
In this study, numerical simulation for the WATLON experiment which was the water experiment of a T-junction piping system (T-pipe) was carried out to validate the MUGTHES and to investigate the relation between the mechanism of temperature fluctuation generation and the unsteady motion of large eddy structures. In the numerical simulation, the large eddy simulation (LES) approach with standard Smagorinsky model was employed as eddy viscosity model to simulate large scale eddy motion in the T-pipe. As for uncertainty quantification in the validation process, the modified method of the Grid Convergence Index (GCI) estimation based on the least squire version could successfully quantify uncertainty. Through the numerical simulations, it was indicated that the fine mesh arrangement could improve the temperature distribution in the wake. It could be found that the thermal mixing phenomena in the T-pipe were caused by the mutual interaction of the necklace-shaped vortex around the wake from the front of the branch jet, the horseshoe-shaped vortex and the Karman's vortex motions in the wake.
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.
Kobayashi, Jun; Kimura, Nobuyuki*; Tanaka, Masaaki; Kamide, Hideki; Oyama, Kazuhiro*; Watanabe, Osamu*
no journal, ,
no abstracts in English
Tanaka, Masaaki; Murakami, Satoshi*
no journal, ,
Thermal striping on the core instrumentation plate (CIP) around the primary control rod (PCR) and backup control rod (BCR) channels and the radial blanket fuel assemblies (RBAs) may be caused. Since the interaction between neighbor areas exists in the UIS and the cold sodium flowing from the RBA is affected by the external flow around the UIS, a spatial connection method consisting of the numerical model for the whole upper plenum and the local target area has been developed. The numerical results were compared with the experimental results to confirm applicability of the method to the practical problem. And, sensitivity of mesh arrangement to the numerical results was discussed by using wide and narrow area models with two different spatial resolutions in each model. Through the examinations, appropriate local model for the spatial connection mothed could be proposed.
Kobayashi, Jun; Ezure, Toshiki; Tanaka, Masaaki; Kamide, Hideki
no journal, ,
no abstracts in English
Kobayashi, Jun; Ezure, Toshiki; Tanaka, Masaaki; Kamide, Hideki
no journal, ,
JAEA has been conducting a design study for an advanced large-scale sodium-cooled fast reactor (SFR). Hot sodium from the fuel subassembly can mix with the cold sodium from the control rod (CR) channel and blanket assemblies at the bottom of Upper Internal Structure (UIS). Temperature fluctuation due to the fluid mixing at the core outlet may cause high cycle thermal fatigue at the bottom of UIS. JAEA had performed a water experiment to examine countermeasures for the significant temperature fluctuation generated at the bottom of SFRs UIS. A water experiment was conducted using a 1/3 scale 60
-sector model of the reactor core and upper plenum. The temperature fluctuations near the cold fluid outlets were determined, and several countermeasures were tested. It was confirmed that these countermeasures could reduce the temperature fluctuations at the bottom of the UIS.
Tanaka, Masaaki
no journal, ,
In the lecture course of "Guideline for Credibility Assessment of Nuclear Simulations: 2015", the outline of the V2UP (V&V Plus Uncertainty Quantification and Prediction) procedure for high cycle thermal fatigue in sodium-cooled fast reactor is to be introduced.
Tanaka, Masaaki
no journal, ,
In the lecture course of "Guideline for Credibility Assessment of Nuclear Simulations: 2015", the outline of the V2UP (V&V Plus Uncertainty Quantification and Prediction) procedure for high cycle thermal fatigue in sodium-cooled fast reactor is to be introduced.
Kobayashi, Jun; Ezure, Toshiki; Tanaka, Masaaki; Kamide, Hideki
no journal, ,
Hot sodium from a fuel subassembly can mix with cold sodium from a control rod (CR) channel at a bottom of Upper Internal Structure in Advanced-SFR reactor vessel. It was confirmed by the water experiment that a mitigation countermeasure was applied to the temperature fluctuation intensity, and effective.
Kobayashi, Jun
no journal, ,
no abstracts in English
