Interaction of solute manganese and nickel atoms with dislocation loops in iron-based alloys irradiated with 2.8 MeV Fe ions at 400 C

Nguyen, B. V. C.*; Murakami, Kenta*; Chena, L.*; Phongsakorn, P. T.*; Chen, X.*; Hashimoto, Takashi; Hwang, T.*; Furusawa, Akinori; Suzuki, Tatsuya*

Nuclear Materials and Energy (Internet), 39, p.101639_1 - 101639_9, 2024/06

https://doi.org/10.1016/j.nme.2024.101639

Recent improvements of probabilistic fracture mechanics analysis code PASCAL for reactor pressure vessels

Lu, K.; Takamizawa, Hisashi; Katsuyama, Jinya; Li, Y.

International Journal of Pressure Vessels and Piping, 199, p.104706_1 - 104706_13, 2022/10

https://doi.org/10.1016/j.ijpvp.2022.104706

Numerical investigations on the coolability and the re-criticality of a debris bed with the density-stratified configuration

Li, C.-Y.; Uchibori, Akihiro; Takata, Takashi; Pellegrini, M.*; Erkan, N.*; Okamoto, Koji*

Dai-25-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2021/07

The capability of stable cooling and avoiding re-criticality on the debris bed are the main issues for achieving IVR (In-Vessel Retention). In the actual situation, the debris bed is composed of mixed-density debris particles. Hence, when these mixed-density debris particles were launched to re-distribute, the debris bed would possibly form a density-stratified distribution. For the proper evaluation of this scenario, the multi-physics model of CFD-DEM-Monte-Carlo based neutronics is established to investigate the coolability and re-criticality on the heterogeneous density-stratified debris bed with considering the particle relocation. The CFD-DEM model has been verified by utilizing water injection experiments on the mixed-density particle bed in the first portion of this research. In the second portion, the coupled system of the CFD-DEM-Monte-Carlo based neutronics model is applied to reactor cases. Afterward, the debris particles' movement, debris particles' and coolant's temperature, and the k-eff eigenvalue are successfully tracked. Ultimately, the relocation and stratification effects on debris bed's coolability and re-criticality had been quantitatively confirmed.

Application of probabilistic fracture mechanics to reactor pressure vessel using PASCAL4 code

Lu, K.; Katsuyama, Jinya; Li, Y.; Yoshimura, Shinobu*

Journal of Pressure Vessel Technology, 143(2), p.021505_1 - 021505_8, 2021/04

https://doi.org/10.1115/1.4048103

Development of ex-vessel phenomena analysis model for multi-scenario simulation system, spectra

Uchibori, Akihiro; Aoyagi, Mitsuhiro; Takata, Takashi; Ohshima, Hiroyuki

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 7 Pages, 2020/08

https://doi.org/10.1115/ICONE2020-16818

The multi-scenario simulation system named SPECTRA has been developed for integrated analysis of in- and ex-vessel phenomena during a severe accident in sodium-cooled fast reactors. The base module computing ex-vessel compressible gas behavior by a lumped mass model and a sodium-concrete interaction module were verified through the basic analyses individually. A validity of the system including the base module and the individual physical module such as the sodium-concrete interaction module was confirmed through the analysis assuming sodium leakage from a reactor vessel and a primary cooling loop.

Recent verification activities on probabilistic fracture mechanics analysis code PASCAL4 for reactor pressure vessel

Lu, K.; Katsuyama, Jinya; Li, Y.; Miyamoto, Yuhei*; Hirota, Takatoshi*; Itabashi, Yu*; Nagai, Masaki*; Suzuki, Masahide*; Kanto, Yasuhiro*

Mechanical Engineering Journal (Internet), 7(3), p.19-00573_1 - 19-00573_14, 2020/06

https://doi.org/10.1299/mej.19-00573

Improvements on evaluation functions of a probabilistic fracture mechanics analysis code for reactor pressure vessels

Lu, K.; Katsuyama, Jinya; Li, Y.

Journal of Pressure Vessel Technology, 142(2), p.021208_1 - 021208_11, 2020/04

https://doi.org/10.1115/1.4045512

Code-B-2.5.2 for stress calculation for SiC-TRISO fuel particle

Aihara, Jun; Goto, Minoru; Ueta, Shohei; Tachibana, Yukio

JAEA-Data/Code 2019-018, 22 Pages, 2020/01

JAEA-Data-Code-2019-018.pdf:1.39MB

Concept of Pu-burner high temperature gas-cooled reactor (HTGR) was proposed for purpose of more safely reducing amount of recovered Pu. In Pu-burner HTGR concept, coated fuel particle (CFP), with ZrC coated yttria stabilized zirconia (YSZ) containing PuO (PuO-YSZ) small particle and with tri-structural isotropic (TRISO) coating, is employed for very high burn-up and high nuclear proliferation resistance. ZrC layer is oxygen getter. On the other hand, we have developed Code-B-2.5.2 for prediction of pressure vessel failure probabilities of SiC-tri-isotropic (TRISO) coated fuel particles for HTGRs under operation by modification of an existing code, Code-B-2. The main purpose of modification is preparation of applying code for CFPs of Pu-burner HTGR. In this report, basic formulae are described.

Effect of coolant water temperature of ECCS on failure probability of RPV

Katsuyama, Jinya; Masaki, Koichi; Lu, K.; Watanabe, Tadashi*; Li, Y.

Proceedings of 2019 ASME Pressure Vessels and Piping Conference (PVP 2019) (Internet), 7 Pages, 2019/07

For reactor pressure vessel (RPV) of pressurized water reactor, temperature of coolant water in emergency core cooling system (ECCS) may have influence on the structural integrity of RPV during pressurized thermal shock (PTS) events. Focusing on a mitigation measure to raise the coolant water temperature of ECCS for aged RPVs in order to reduce the effect of thermal shock due to PTS events, we performed thermal hydraulic analyses and probabilistic fracture mechanics analyses by using RELAP5 and PASCAL4, respectively. From the analysis results, it was shown that the failure probability of RPV was dramatically reduced when the coolant temperature in accumulator as well as high and low pressure injection systems (HPI/LPI) was raised, although raising the coolant temperature of HPI/LPI only did not cause reduction in the failure probability.

Preliminary analysis of sodium experimental apparatus PLANDTL-2 for development of evaluation method for thermal hydraulics in reactor vessel of sodium fast reactor under decay heat removal system operation condition

Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05

https://doi.org/10.1299/jsmeicone.2019.27.2318

Decay heat removal system (DHRS) by using the natural circulation without depending on the pump as the mechanical equipment is recognized as one of the most effective methodologies for the sodium-cooled fast reactor from the viewpoint of the safety enhancement. In this paper, the numerical simulation results of the preliminary analysis for the sodium experiment with the apparatus of PLANDTL-2, in which the core and the upper plenum with a dipped-type direct heat exchanger (DHX) were modeled, were discussed, in order to establish appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX.