Study on severe accidents of oxide-fueled sodium-cooled fast reactors, 1; Project overview and progress until 2025

Yamano, Hidemasa; Matsuba, Kenichi; Sasa, Kyohei*; Kinoshita, Enki*; Morita, Koji*; Saito, Yasushi*; Ito, Kei*; Suzuki, Toru*

Dai-30-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2026/06

Seismic fragility evaluation of a thin-walled vessel in a sodium-cooled fast reactor

Watakabe, Tomoyoshi; Yamano, Hidemasa; Futagami, Satoshi

Transactions of the 28th International Conference on Structural Mechanics in Reactor Technology (SMiRT28) (Internet), 10 Pages, 2025/08

The fragility evaluation method of the RV in the SFR has been developed for both failure mode buckling and fatigue. The past studies showed the fragility using the RV of a loop-type SFR: radius R / thickness t is about 100. Recently, Japan Atomic Energy Agency together with Japanese industrial partners has been conducting a pool-type SFR design. The fragility of the pool-type RV has never been evaluated by the proposed method, which is based on the fatigue failure mode instead of the buckling. To investigate the basic behaviour of the pool-type RV under excessive seismic loading, this report investigated the fragility by using a simple cylindrical model with a wall thickness close to the design of the pool-type RV: R/t is 200, as the first trial.

Data report of ROSA/LSTF experiment SB-PV-03; 0.2% pressure vessel bottom break LOCA with SG depressurization and gas inflow

Takeda, Takeshi

JAEA-Data/Code 2024-014, 76 Pages, 2024/12

JAEA-Data-Code-2024-014.pdf:4.0MB

An experiment denoted as SB-PV-03 was conducted on November 19, 2002 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-PV-03 simulated a 0.2% pressure vessel bottom small-break loss-of-coolant accident in a pressurized water reactor (PWR). The test assumptions included total failure of high pressure injection system of emergency core cooling system (ECCS) and noncondensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of ECCS. Secondary-side depressurization of both steam generators (SGs) as an accident management (AM) action to achieve the depressurization rate of 55 K/h in the primary system was initiated 10 min after the generation of a safety injection signal, and continued afterwards. Auxiliary feedwater injection into the secondary-side of both SGs was started for 30 min with some delay after the onset of the AM action. The AM action was effective on the primary depressurization until the ACC tanks began to discharge nitrogen gas into the primary system. The core liquid level recovered in oscillative manner because of intermittent coolant injection from the ACC system into both cold legs. Therefore, the core liquid level remained at a small drop. The pressure difference between the primary and SG secondary sides became larger after nitrogen gas ingress. Core uncovery occurred by core boil-off during reflux condensation in the SG U-tubes under nitrogen gas influx. When the maximum cladding surface temperature of simulated fuel rods exceeded the pre-determined value of 908 K, the core power was automatically reduced to protect the LSTF core. After the automatic core power reduction, coolant injection from low pressure injection (LPI) system of ECCS into both cold legs led to the whole core quench. After the continuous core cooling was confirmed through the actuation of the LPI system, the experiment was terminated.

Fukushima Daiichi Nuclear Power Plant accident analysis considering the thermal stratification and containment leakage

Nakamura, Yuki*; Kojima, Yoshihiro*; Yamashita, Takuya; Shimomura, Kenta; Mizokami, Shinya

Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NUTHOS-14) (Internet), 12 Pages, 2024/08

Development of the buckling evaluation method for large scale vessels in fast reactors made of grade 91 steel and austenitic stainless steel with large initial imperfections

Okafuji, Takashi*; Miura, Kazuhiro*; Sago, Hiromi*; Murakami, Hisatomo*; Watakabe, Tomoyoshi; Ando, Masanori; Miyazaki, Masashi

Proceedings of the ASME 2024 Pressure Vessels & Piping Conference (PVP 2024) (Internet), 8 Pages, 2024/07

We have developed the buckling strength equations of vessels for fast reactors with seismic isolation system. The applicability of the buckling equations was confirmed by a series of buckling tests and analyses under monotonic or cyclic axial compressive load accompanied with constant horizontal load in the previous reports. In this report, we proposed a correction factor to reduce the buckling strength calculated by the buckling equations for large initial imperfections. A series of elastic-plastic buckling analyses considering large displacement and large strain theories was conducted to Grade 91 steel and austenitic stainless steel vessels which has a wide range of dimensions, initial imperfection amplitude, and vertical/horizontal load ratio. The simulation results showed that the correction factor generally shows a reduction tendency of buckling strength corresponding to initial imperfection amplitude, and the modified buckling equations are applicable to the vessels in fast reactors even for large initial imperfection amplitude which exceeds half the wall thickness.

Formulation of material property formula for calculation of damage in reactor pressure vessel during accident evaluation

Shimomura, Kenta; Yamashita, Takuya; Nagae, Yuji

Proceedings of 11th European Review Meeting on Severe Accident Research Conference (ERMSAR 2024) (Internet), 12 Pages, 2024/05

https://doi.org/10.5445/IR/1000174165/v2

Large-eddy simulation on gas mixing induced by the high-buoyancy flow in the CIGMA facility

Abe, Satoshi; Shibamoto, Yasuteru

Nuclear Engineering and Technology, 55(5), p.1742 - 1756, 2023/05

https://doi.org/10.1016/j.net.2023.01.019

CFD analysis on stratification dissolution and breakup of the air-helium gas mixture by natural convection in a large-scale enclosed vessel

Hamdani, A.; Abe, Satoshi; Ishigaki, Masahiro; Shibamoto, Yasuteru; Yonomoto, Taisuke

Progress in Nuclear Energy, 153, p.104415_1 - 104415_16, 2022/11

https://doi.org/10.1016/j.pnucene.2022.104415

Preliminary deformation analysis of the reactor vessel due to core debris accumulation onto the reactor vessel bottom for sodium-cooled fast reactor

Onoda, Yuichi; Yamano, Hidemasa

Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 9 Pages, 2022/10

In Japan, sodium-cooled fast reactor design takes In-Vessel Retention (IVR) strategy to stably cool damaged core materials in the reactor vessel during a severe accident with various design measures. Although a possibility to fail IVR is extremely low, a probabilistic risk assessment study needs a wide variety of scenarios including the IVR failure. Therefore, in order to study a wide range of event spectra related to stable cooling of debris in the reactor vessel, this study numerically investigated the deformation and failure behavior of the reactor vessel due to the debris deposited onto the skirt of the core catcher using the FINAS-STAR structural analysis code. The analyses are conducted in two cases of power density with the aim of investigating failure conditions of the bottom of the reactor vessel. Reactor vessel deforms significantly when the temperature reaches about 1100 C and the reactor vessel reaches the failure criteria in high-power-density case.

Estimation of long-term ex-vessel debris cooling behavior in Fukushima Daiichi Nuclear Power Plant unit 3

Sato, Ikken; Yamaji, Akifumi*; Li, X.*; Madokoro, Hiroshi

Mechanical Engineering Journal (Internet), 9(2), p.21-00436_1 - 21-00436_17, 2022/04

https://doi.org/10.1299/mej.21-00436