The Role of silicon on solute clustering and embrittlement in highly neutron-irradiated pressurized water reactor surveillance test specimens

Takamizawa, Hisashi; Hata, Kuniki; Nishiyama, Yutaka; Toyama, Takeshi*; Nagai, Yasuyoshi*

Journal of Nuclear Materials, 556, p.153203_1 - 153203_10, 2021/12

https://doi.org/10.1016/j.jnucmat.2021.153203

Solute clusters (SCs) formed in pressurized water reactor surveillance test specimens neutron-irradiated to a fluence of 1 10 n/cm were analyzed via atom probe tomography to understand the effect of silicon on solute clustering and irradiation embrittlement of reactor pressure vessel steels. In high-Cu bearing materials, Cu atoms were aggregated at the center of cluster surrounded by the Ni, Mn, and Si atoms like a core-shell structure. In low-Cu bearing materials, Ni, Mn, and Si atoms formed cluster and these solutes were not comprised core-shell structure in SCs. While the number of Cu atoms in clusters was decreased with decreasing nominal Cu content, the number of Si atoms had clearly increased. The cluster radius () and number density () decreased and increased, respectively, with increasing nominal Si content. The shift in the reference temperature for nil-ductility transition (RT) showed a good correlation with the square root of volume fraction () multiplied by r (). This suggested that the dislocation cutting through the particles mechanism dominates the precipitation hardening responsible for irradiation embrittlement. The negative relation between the nominal Si content and RT indicated that increasing of nominal Si content reduces the degree of embrittlement.

Development of seismic safety assessment method for piping in long-term operated nuclear power plant

Yamaguchi, Yoshihito; Li, Y.

Haikan Gijutsu, 63(12), p.22 - 27, 2021/10

no abstracts in English

Bayesian analysis of Japanese pressurized water reactor surveillance data for irradiation embrittlement prediction

Takamizawa, Hisashi; Nishiyama, Yutaka

Journal of Pressure Vessel Technology, 143(5), p.051502_1 - 051502_8, 2021/10

https://doi.org/10.1115/1.4050317

no abstracts in English

Comparison analysis between U.S. and Japan on Evacuation Time Estimation for nuclear emergency planning zones

Shimada, Kazumasa; Takahara, Shogo

JAEA-Review 2021-013, 142 Pages, 2021/09

JAEA-Review-2021-013.pdf:4.74MB

In this report, the authors reviewed the published reports of Evacuation Time Estimation (ETE) conducted in Japan and United States and examined the issues of ETE in Japan. The authors obtained public ETE reports in Japan from 16 prefectures up to February 2020. In addition, the authors obtained 58 ETE reports in United States from 2011 to 2018. Next, the overview of ETE for the Emergency Planning Zone (EPZ) around the nuclear power plant in United States was described based on the NUREG/CR-7002 of the U.S Nuclear Regulatory Commission (NRC). Then, based on the guidance of the ETE of the Cabinet Office of Japan, the overview of ETE in Japan for the Precautionary Action Zone (PAZ) and the Urgent Protective Action Planning Zone (UPZ) was described and compared with the ETE in United States. It was found that ETE in Japan often outputs only the time of 90% or 100% that population complete evacuation. Therefore, in order to use ETE in Japan for emergency decision-making, it is necessary to unify the inputs and outputs of ETE as in the United States' ETE reports.

The Dependence of pool scrubbing decontamination factor on particle number density; Modeling based on bubble mass and energy balances

Sun, Haomin; Shibamoto, Yasuteru; Hirose, Yoshiyasu; Kukita, Yutaka

Journal of Nuclear Science and Technology, 58(9), p.1048 - 1057, 2021/09

https://doi.org/10.1080/00223131.2021.1907254

no abstracts in English

Evaluation of risk dilution effects in dynamic probabilistic risk assessment of nuclear power plants

Kubo, Kotaro; Tanaka, Yoichi

Proceedings of 31st European Safety and Reliability Conference (ESREL 2021) (Internet), p.810 - 817, 2021/09

https://doi.org/10.3850/978-981-18-2016-8_477-cd

Probabilistic risk assessment (PRA) is a method of effectively evaluating risks in nuclear power plants and is used in various agencies. Dynamic PRA is attracting considerable attention, as it enables realistic assessment by reducing the assumptions and engineering judgments related to time-dependent failure probability and/or human action reliability. However, it is difficult to remove all assumptions and engineering judgments. Therefore, their effects on assessment results should be understood. This study focuses on the "risk dilution effect," which arises from assumptions about uncertainty. Results showed that this effect causes a difference of about 10% to 20% in the relative change of the conditional core damage probability in the station blackout scenario. This effect should be fully considered when using dynamic PRA in critical decision-making, such as that on regulations.

Analysis of behavior of Ru with nitrogen oxide chemical behavior in accident of evaporation to dryness by boiling of reprocessed high level liquid waste

Yoshida, Kazuo; Tamaki, Hitoshi; Hiyama, Mina*

JAEA-Research 2021-005, 25 Pages, 2021/08

JAEA-Research-2021-005.pdf:2.91MB

An accident of evaporation to dryness by boiling of high level liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (Ru) are released from the tanks with water and nitric-acid mixed vapor into atmosphere. Accurate quantitative estimation of released Ru is one of the important issues for risk assessment of those facilities. To resolve this issue, an empirical correlation equation of Ru mass transfer coefficient across the vapor-liquid surface, which can be useful for quantitative simulation of Ru mitigating behavior, has been obtained from data analyses of small-scale experiments conducted to clarify gaseous Ru migrating behavior under steam-condensing condition. A simulation study has been also carried out with a hypothetical typical facility building successfully to demonstrate the feasibility of quantitative estimation of amount of Ru migrating in the facility using the obtained correlation equation implemented in SCHERN computer code which simulates chemical behaviors of nitrogen oxide based on the condition also simulated thermal-hydraulic computer code.

Mechanical failure of high-burnup fuel rods with stress-relieved annealed and recrystallized M-MDA cladding under reactivity-initiated accident conditions

Mihara, Takeshi; Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki

Journal of Nuclear Science and Technology, 58(8), p.872 - 885, 2021/08

https://doi.org/10.1080/00223131.2021.1883143

An Approach toward evaluation of long-term fission product distributions in the Fukushima Daiichi Nuclear Power Plant after the severe accident

Uchida, Shunsuke; Karasawa, Hidetoshi; Kino, Chiaki*; Pellegrini, M.*; Naito, Masanori*; Osaka, Masahiko

Nuclear Engineering and Design, 380, p.111256_1 - 111256_19, 2021/08

https://doi.org/10.1016/j.nucengdes.2021.111256

It is essential to grasp the long-term distributions of FP as well as fuel debris all over the Fukushima Daiichi Nuclear Power Plant (1F) for safe completion of its decommissioning projects. The fuel debris is going to be removed from the plant under the severe conditions of FP being scattered during major decommissioning work, and then, the decommissioning projects are going to be terminated by storing safely the removed debris as recovered fertile materials or as materials for final radioactive disposal. In order to determine the FP distribution in the plant for the long period from the accident occurrence to the termination of the plant decommissioning, procedures for analyzing multi-term FP behaviors were proposed. The proposed procedures should be improved by applying the FP data measured in the plant and validated based on the feedback data. Then, the accuracy-improved procedures should be applied to estimate FP distribution during each period of the decommissioning projects.

Improvement of intragranular fission gas behavior model for fuel performance code FEMAXI-8

Udagawa, Yutaka; Tasaki, Yudai

JAEA-Data/Code 2021-007, 56 Pages, 2021/07

JAEA-Data-Code-2021-007.pdf:5.05MB

Japan Atomic Energy Agency (JAEA) has released FEMAXI-8 in 2019 as the latest version of the fuel performance code FEMAXI, which has been developed to analyze thermal and mechanical behaviors of a single fuel rod in mainly normal operation conditions and anticipated transient conditions. This report summarizes a newly developed model to analyze intragranular fission gas behaviors considering size distribution of gas bubbles and their dynamics. While the intragranular fission gas behavior models implemented in the previous FEMAXI versions have supported only treating single bubble size for a given fuel element, the new model supports multiple gas groups according to their size and treats their dynamic behaviors, making the code more versatile. The model was first implemented as a general module that takes arbitrary number of bubble groups and spatial mesh division for a given fuel grain system. An interface module to connect the model to FEMAXI-8 was then developed so that it works as a 2 bubble groups model, which is the minimum configuration of the multi-grouped model to be operated with FEMAXI-8 at the minimum calculation cost. FEMAXI-8 with the new intragranular model was subjected to a systematic validation calculation against 144 irradiation test cases and recommended values for model parameters were determined so that the code makes reasonable predictions in terms of fuel center temperature, fission gas release, etc. under steady-state and power ramp conditions.