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Journal Articles

Damage evaluations for BWR lower head in severe accident based on multi-physics simulations

Katsuyama, Jinya; Yamaguchi, Yoshihito; Nemoto, Yoshiyuki; Furuta, Takuya; Kaji, Yoshiyuki

Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 9 Pages, 2022/07

Journal Articles

Creep damage evaluations for BWR lower head in severe accident

Katsuyama, Jinya; Yamaguchi, Yoshihito; Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Osaka, Masahiko

Transactions of the 24th International Conference on Structural Mechanics in Reactor Technology (SMiRT-24) (USB Flash Drive), 11 Pages, 2017/08

It is difficult to assess rupture behavior of the lower head of reactor pressure vessel in boiling-water-type nuclear power plants due to severe accident like Fukushima Daiichi because Boiling Water Reactor (BWR) lower heads have geometrically complicated structure with a lot of penetrations. Therefore, we have been developing an analysis method to predict time and location of RPV lower head rupture of BWRs considering creep damage mechanisms based on coupled analysis of three-dimensional Thermal-Hydraulics (TH) and thermal-elastic-plastic-creep analyses. In this study, we performed creep damage evaluations to investigate the effects of the debris depth and heat generation locations on failure behavior of lower head. From the analysis results, we discussed the outflow paths of the relocated molten core to the containment, and it was concluded that failure regions of BWR lower head are only the control rod guide tubes or stub tubes under simulated conditions.

Journal Articles

Development of failure evaluation method for BWR lower head in severe accident, 2; Applicability evaluation of the FEM using uni-axial material data for multi-axial deformation analysis

Nemoto, Yoshiyuki; Kato, Hitoshi; Kaji, Yoshiyuki; Yoshida, Hiroyuki

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05

For the evaluation of reactor pressure vessel (RPV) lower head rupture probably occurred during the severe accident in Fukushima Daiichi Nuclear Power Plants, JAEA is conducting the thermal-hydraulics / mechanical coupling analysis. In the mechanical analysis based on the finite element method (FEM), material property data previously obtained from uni-axial material tests are applied. The lower head of BWR such as Fukushima NPP, has complicated structure compared to PWR, with control rod guide tubes, stub tubes, etc., therefore the mechanical analyses need to treat multi-axial deformation of the materials. To perform such mechanical analysis, the applicability of the analytical model using uni-axial data for multi-axial deformation analysis must be validated. In this study, the internal pressure creep tests were performed because which can realize the multi-axial deformation condition. In addition, mechanical analyses were conducted and the analytical results were compared with the experimental data.

JAEA Reports

Long-term effect of creep displacement of host-rock on stability of engineered barrier system for TRU waste; Two-dimensional analysis by the non-linear viscoelasticity model

Aoyagi, Takayoshi*; *; Mihara, Morihiro; Okutsu, Kazuo*; Maeda, Munehiro*

JNC TN8400 2001-024, 103 Pages, 2001/06

JNC-TN8400-2001-024.pdf:8.84MB

In the disposal concept of TRU waste, concentrated disposal of wastes forms in large cross-section underground cavities is envisaged, because most of TRU waste is no-heat producing in spite of large generated volume as compared with HLW. In the design of engineered barrier system based on large cross-section cavities, it is necessary to consider the long-term mechanical process such as creep displacement of the host rock from the viewpoint of the stability of engineered barrier system. In this study, the long-term creep displacement of the host rock was calculated using the non-linear viscoelasticity model and the effects on the stability of engineered barrier system was evaluated. As a result, in the disposal concept of crystalline rock, no creep displacement occurred at the time after 1 milion year. On the other hand, in the disposal concept of sedimentary rock, creep displacement of 8090mm occurred at the time after 1 milion year. Also, in this calculation, a maximum reduction of 45mm concerned with the thickness of buffer material was estimated. But these values resulted within allowance of design values. Therefore, these results show that the effects of the creep displacement on the stability of engieered barrier system would not be significant.

JAEA Reports

Irradiation creep of modified 316 and 15Cr-20Ni base austenitic S.S. fuel pins (MFA-1, 2) irradiated in FFTF

; ; Mizuta, Shunji

JNC TN9400 2000-023, 126 Pages, 2000/02

JNC-TN9400-2000-023.pdf:2.94MB

Modified 316 and 15Cr-20Ni base austenitic stainless steels had been developed by Japan Nuclear Cycle Development lnstitute as the candidate materials for Monju and Demonstration fast breeder reactor. Previously, irradiation creep correlation of modified 316 and 15Cr-20Ni had been evaluated using pressurized tubes irradiated in FFTF/MOTA. 0n the other hand, for other austenitic S.S. developed abroad, it was reported that irradiation creep behavior of fuel pin could not be sufficiently described using results of pressurized tube experiments. ln this study, irradiation creep properties of modified 316 and 15Cr-20Ni fuel pins (MFA-I, 2) irradiated in FFTF were evaluated. And irradiation deformation of MFA-1, 2 fuel pins were estimated using the irradiation creep correlation based on MOTA data. The results are summarized as follows : (1)Irradiation creep compliance B $_{0}$ calculated from MFA-I, 2 data are 5.6 15.010 $^{-6}$ [(I0 $^{26}$ n/m $^{2}$ , E>0.1Mev) $^{-1}$ (MPa) $^{-1}$ ], Which are larger than B $_{0}$ based on MOTA data of 2.26.410 $^{-6}$ and are within the range of B $_{0}$ of other austenitic S.S. abroad. (2)Creep-swelling coupling coefficient D derived from MFA-1, 2 data tend to decrease with increasing swelling rate. And the range of D based on MFA-1, 2 data include values calculated from MOTA data of 3.88.210 $^{-3}$ [(MPa) $^{-1}$ ] and for other austenitic S.S. abroad. (3)As the result that irradiation creep deformation of MFA-1, 2 fuel pins could be appropriately estimated using the irradiation creep correlation derived from MOTA data, it is considered that the creep, correlation based on MOTA data can be applied to estimation of fuel pin deformation.

JAEA Reports

Evaluation of long-term mechanical stability of near field

Takachi, Kazuhiko; Sugino, Hiroyuki

JNC TN8400 99-043, 52 Pages, 1999/11

JNC-TN8400-99-043.pdf:5.2MB

In the near field, as tunnels and pits are excavated, a redistribution of stresses in the surrounding rock will occur. For a long period of time after the emplacement of waste packages various events will take place, such as the swelling of the buffer, sinking of the overpack under its own weight, deformation arising from expansion of overpack corrosion products and the creep deformation of the rock mass. The evaluation of what effects these changes in the stress-state will have on the buffer and rock mass is a major issue from the viewpoint of safety assessment. Therefore, rock creep analysis, overpack corrosion expansion analysis and overpack sinking analysis have been made in order to examine the longterm mechanical stability of the near field and the interaction of various events that may affect the stability of the near field over a long period of time. As the results, rock creep behavior, the variations of the stress-state and the range of the influence zone differ from the rock strength, strength of buffer in the tunnel and side pressure coefficient etc. about the hard rock system and soft rock system established as basic cases. And the magnitude of the stress variations for buffer by the overpack sinking and rock creep deformation is negligible compared with it by the overpack corrosion expansion. Furthermore, though very limited zone of buffer around the overpack is close to the critical state by the overpack corrosion expansion, the engineered barrier system attains a comparatively stable state for a long period of time.

JAEA Reports