Critical power of boiling water flows in 33 rod bundle simulating BWR subchannels at elevated pressure conditions

Wada, Yuki; Shibamoto, Yasuteru; Hibiki, Takashi*

International Communications in Heat and Mass Transfer, 179(Part.3), p.112282_1 - 112282_22, 2026/10

https://doi.org/10.1016/j.icheatmasstransfer.2026.112282

Numerical analysis of a potential Reactor Pressure Vessel (RPV) boundary failure mechanism in Fukushima Daiichi Nuclear Power Station Unit-2

Li, X.; Yamaji, Akifumi*; Sato, Ikken*; Yamashita, Takuya

Annals of Nuclear Energy, 214, p.111217_1 - 111217_13, 2025/05

https://doi.org/10.1016/j.anucene.2025.111217

Scenario analysis of future nuclear energy use in Japan, 3; Promotion of Plutonium utilization by RBWR-Backfit

Nishihara, Kenji; Oizumi, Akito; Hino, Tetsushi*; Soneda, Hideo*

Progress in Nuclear Science and Technology (Internet), 7, p.305 - 310, 2025/03

https://doi.org/10.15669/pnst.7.305

Hitachi-GE is conducting research and development of new fuel design named resource-renewable boiling water reactor-backfit (RBWR-BF) to enhance plutonium utilization in BWRs. In this study, the effects of installing RBWR-BF were estimated using the NMB code, a nuclear fuel cycle simulator, under the assumption of a future nuclear energy utilization scenario based on light water reactor (LWR) in Japan. As a result, it was shown that burden of MOX loading to LWRs can be minimized and fissionability of remaining plutonium in 2100 was improved by RBWR-BF introduction from 2040.

Analysis methodology of the calculation code, WRAC-JAEA, for determining major indexes of corrosive circumstance in light water reactors

Uchida, Shunsuke; Hata, Kuniki; Hanawa, Satoshi

JAEA-Data/Code 2024-003, 119 Pages, 2025/01

JAEA-Data-Code-2024-003.pdf:11.29MB
JAEA-Data-Code-2024-003-appendix(CD-ROM).zip:0.28MB

The calculation code for determining corrosive circumstance in light water reactors, WRAC-JAEA, has been developed based on water radiolysis calculation codes for BWR. The code has involved several new calculation functions to apply it for PWR, i.e., (1) high temperature pH (pH), (2) pH effects on water radiolysis, (3) electrochemical corrosion potential (ECP) based on the mixed potential theory, and (4) ECP based on the water radiolysis calculation results. Moderation of corrosive conditions in the primary cooling systems is one of the promising procedures to mitigate the loss of reliabilities of major components in the systems, especially in aging plants. However, water chemistry control for corrosive environment mitigation procedures are much different in BWRs and PWRs. In BWRs, intergranular stress corrosion cracking (IGSCC) of stainless steel is the dominant causes for determining plant reliability. It is difficult to increase pH and injected hydrogen amounts due to direct power cycle operation. So, precise control of hydrogen injection with supported by water radiolysis and ECP analyses has been carried out to keep material reliability. In PWRs, it is possible to maintain stable control of corrosive circumstances with higher pH and sufficiently large hydrogen concentration. Recently, it was pointed out that one of the major causes of primary water stress corrosion cracking (PWSCC) of nickel alloys was hydrogen. The optimal hydrogen concentration should be evaluated to mitigate ECP without increasing hydrogen concentration. For this, a combined water radiolysis and ECP analysis code is required to determine the suitable hydrogen concentration and ECP. WRAC-JAEA can contribute not only to evaluation of corrosive conditions each of BWR and PWR, but also to prepare for suitable countermeasures for both BWR and PWR by cross-talking the knowledge and experience with assistance of the code results.

A Coupled analyses of water radiolysis and ECP for evaluation of the corrosive conditions in BWRs and PWRs

Hata, Kuniki; Uchida, Shunsuke; Hanawa, Satoshi; Chimi, Yasuhiro; Sato, Tomonori

Proceedings of 21st International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors (Internet), 14 Pages, 2023/08

Evaluation of core material energy change during the in-vessel phase of Fukushima Daiichi Unit 3 based on observed pressure data utilizing GOTHIC code analysis

Sato, Ikken; Arai, Yuta*; Yoshikawa, Shinji

Journal of Nuclear Science and Technology, 58(4), p.434 - 460, 2021/04

https://doi.org/10.1080/00223131.2020.1862718

Thresholds for failure of high-burnup LWR fuels by pellet cladding mechanical interaction under reactivity-initiated accident conditions

Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki

Journal of Nuclear Science and Technology, 56(12), p.1063 - 1072, 2019/12

https://doi.org/10.1080/00223131.2019.1637795

no abstracts in English

Empirical equations of crack growth rates based on data fitting of neutron irradiated stainless steel under high temperature water simulating boiling water reactor core conditions

Kasahara, Shigeki; Chimi, Yasuhiro; Hata, Kuniki; Fukuya, Koji*; Fujii, Katsuhiko*

Proceedings of 19th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors (Internet), p.1345 - 1355, 2019/08

This paper describes empirical equation development of crack growth rates (CGR) in consideration of IASCC of neutron irradiated stainless steel to contribute to structural integrity assessment of BWR reactor internals. Empirical equations of CGR (da/dt) were developed based on a formula of da/dt = MK, assuming that "M" and "n" tend to be saturated with increasing neutron fluence. To obtain the empirical equations for normal water chemistry (NWC) and hydrogen water chemistry (HWC) of BWR, a data fitting with least square method was applied to the datasets consisting of F, K and CGR from post irradiation examinations of neutron irradiated stainless steel under simulated NWC and HWC conditions from open literature. As a result, calculated results by the equation for NWC showed good agreement with measured CGR data, meanwhile those for HWC did not. The above difference was seemed to be attributed that CGR data obtained under HWC conditions were scattered extensively.

An Interpretation of Fukushima-Daiichi Unit 3 plant data covering the two-week accident-progression phase based on correction for pressure data

Sato, Ikken

Journal of Nuclear Science and Technology, 56(5), p.394 - 411, 2019/05

https://doi.org/10.1080/00223131.2019.1588798

Water columns were adopted in the pressure measurement system of Fukushima-Daiichi Unit-3. Part of these water columns evaporated during the accident condition jeopardizing correct understanding on actual pressure. Through comparison of RPV (Reactor Pressure Vessel) and S/C pressures with D/W pressure, such water-column effect was evaluated. Correction for this effect was developed enabling clarification of slight pressure difference among RPV, S/C and D/W. This information was then integrated with other available data such as, water level, CAMS and environmental dose rate, into an interpretation of accident focusing on RPV and PCV pressurization/depressurization and radioactive material release to environment. It is suggested that dryout of in-vessel and ex-vessel debris was likely causing pressure decrease. S/C water poured into pedestal heated by relocated debris was the likely cause of pressurization. Cyclic reflooding of pedestal debris and dryout was likely.

Mass transfer inside narrow crevice of SUS316L in high temperature water

Yamamoto, Masahiro; Soma, Yasutaka; Igarashi, Takahiro; Ueno, Fumiyoshi

Proceedings of Annual Congress of the European Federation of Corrosion (EUROCORR 2018) (USB Flash Drive), 7 Pages, 2018/09

In order to clarify the SCC behavior of SUS316L under BWR environment, mass transfer inside crevice of SUS316L in high temperature water using various crevice gap samples was investigated. The samples were prepared by put together two SUS316L sheets. Crevice gap differs from 0.005 mm to 0.1 mm. Corrosion tests were conducted in 8 ppm dissolved oxygen (DO) conditions. Surface oxide film was analysed by laser Raman spectroscopy (LRS) after immersion. Numerical simulations were also conducted by using COMSOL Maltiphysics. Diffusion process of DO and the other chemical species were calculated with connected to electrochemical process. Electrical conductivities inside the crevice were 100 times larger than these of outer water. The reason of high conductivity is existence of Fe ions at the DO depletion crevice.