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Journal Articles

Numerical analysis for FP speciation in VERDON-2 experiment; Chemical re-vaporization of iodine in air ingress condition

Shiotsu, Hiroyuki; Ito, Hiroto*; Sugiyama, Tomoyuki; Maruyama, Yu

Annals of Nuclear Energy, 163, p.108587_1 - 108587_9, 2021/12

 Times Cited Count:0

Journal Articles

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

Solute clusters (SCs) formed in pressurized water reactor surveillance test specimens neutron-irradiated to a fluence of 1 $$times$$ 10$$^{20}$$ n/cm$$^{2}$$ 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 ($$r$$) and number density ($$N_{d}$$) decreased and increased, respectively, with increasing nominal Si content. The shift in the reference temperature for nil-ductility transition ($$Delta$$RT$$_{NDT}$$) showed a good correlation with the square root of volume fraction ($$V_{f}$$) multiplied by r ($$sqrt{V_{f}times {r}}$$). 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 $$Delta$$RT$$_{NDT}$$ indicated that increasing of nominal Si content reduces the degree of embrittlement.

Journal Articles

Coping with electrode polarization for development of DC-driven electrical impedance tomography

Hirose, Yoshiyasu; Sagawa, Jun*; Shibamoto, Yasuteru; Kukita, Yutaka

Flow Measurement and Instrumentation, 81, p.102006_1 - 102006_9, 2021/10

Journal Articles

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

Journal Articles

Flow regime and void fraction predictions in vertical rod bundle flow channels

Han, X.*; Shen, X.*; Yamamoto, Toshihiro*; Nakajima, Ken*; Sun, Haomin; Hibiki, Takashi*

International Journal of Heat and Mass Transfer, 178, p.121637_1 - 121637_24, 2021/10

 Times Cited Count:0 Percentile:0.01(Thermodynamics)

Journal Articles

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

 Times Cited Count:0 Percentile:0(Engineering, Mechanical)

no abstracts in English

Journal Articles

Effect of moderation condition on neutron multiplication factor distribution in $${1/f^beta}$$ random media

Araki, Shohei; Yamane, Yuichi; Ueki, Taro; Tonoike, Kotaro

Nuclear Science and Engineering, 195(10), p.1107 - 1117, 2021/10

 Times Cited Count:0 Percentile:0.02(Nuclear Science & Technology)

Criticality control of random media such as fuel debris is one of the most important safety issues in post-accident management. $$1/f^beta$$ spectrum randomizing model is expected to simulate such random media because it is well known that the $$1/f^beta$$ noise can describe a diverse range of random and disordered natural phenomena. In this paper, we focused on the relationship between the multiplication factor and moderation condition in the $$1/f^beta$$ random media. A number of random media were realized with the $$1/f^beta$$ spectrum randomizing model that is based on the Randomized Weierstrass function (RWF). The volume ratio of concrete to fuel was adopted as an index for the moderation condition. The multiplication factors were calculated with a two-energy group Monte Carlo calculation. The calculation results were analyzed by using variance, skewness, and kurtosis. Those statistical parameters had an extreme value around the optimum moderation condition. This result suggested that it is possible to predict the rough trend of variation range, distortion, and outlier of multiplication factors in the $$1/f^beta$$ random media.

JAEA Reports

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.

Journal Articles

A Numerical investigation on the heat transfer and turbulence production characteristics induced by a swirl spacer in a single-tube geometry under single-phase flow condition

Abe, Satoshi; Okagaki, Yuria; Satou, Akira; Shibamoto, Yasuteru

Annals of Nuclear Energy, 159, p.108321_1 - 108321_12, 2021/09

 Times Cited Count:0 Percentile:0.02(Nuclear Science & Technology)

Journal Articles

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

 Times Cited Count:0 Percentile:0.02(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Penetration factor and indoor deposition rate of elementary and particulate iodine in a Japanese house for assessing the effectiveness of sheltering for radiation exposures

Hirouchi, Jun; Takahara, Shogo; Komagamine, Hiroshi*; Kato, Nobuyuki*; Matsui, Yasuto*; Yoneda, Minoru*

Journal of Radiological Protection, 41(3), p.S139 - S149, 2021/09

 Times Cited Count:0 Percentile:0.02(Environmental Sciences)

Sheltering is one of the countermeasures for protection against radiation exposures in nuclear accidents. The effectiveness of sheltering is often expressed by the reduction factor, that is the ratio of the indoor to the outdoor cumulative radioactivity concentrations or doses. The indoor concentration is mainly controlled by the air exchange rate, penetration factor, and indoor deposition rate. The penetration factor and indoor deposition rate depend on the surface and opening materials. We investigated experimentally these parameters of I$$_{2}$$ and particles. The experiment was performed in two apartment houses, three single-family houses, and chambers. The obtained penetration factor ranged 0.3 $$sim$$ 1 for particles of 0.3 $$sim$$ 1 $$mu$$m and 0.15 $$sim$$ 0.7 for I$$_{2}$$ depending on the air exchange rate. The indoor deposition rate for a house room ranged 0.007 $$sim$$ 0.2 h$$^{-1}$$ for particles of 0.3$$sim$$1 $$mu$$m and 0.2$$sim$$1.5 h$$^{-1}$$ for I$$_{2}$$ depending on floor materials.

Journal Articles

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

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.

JAEA Reports

Transfer and operation of WSPEEDI-II automatic calculation system for responses to nuclear tests by North Korea

Nemoto, Miho*; Ebine, Noriya; Okamoto, Akiko; Hosaka, Yasuhisa*; Tsuzuki, Katsunori; Terada, Hiroaki; Hayakawa, Tsuyoshi; Togawa, Orihiko

JAEA-Technology 2021-013, 41 Pages, 2021/08

JAEA-Technology-2021-013.pdf:2.52MB

When North Korea has carried out nuclear tests, Nuclear Emergency Assistance and Training Center (NEAT) predicts atmospheric dispersion of radionuclides by using the WSPEEDI-II upon requests from Nuclear Regulation Authority (NRA) and submits the predicted results to NRA in cooperation with Nuclear Science and Engineering Center (NSEC). This is a part of the activity of NEAT supporting the Japanese Government in emergency responses. The WSPEEDI-II automatic calculation system specialized for responses to nuclear tests by North Korea was developed by NSEC and was used for responses to three nuclear tests from February 2013 to September 2017. This report describes the transfer and installation of the calculation system to NEAT, and the subsequent maintenance and operation. Future issues for responses to nuclear tests are also described in this report.

JAEA Reports

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.

JAEA Reports

SCHERN-V2: Technical guide of computer program for chemical behavior in accident of evaporation to dryness by boiling of reprocessed high level liquid waste in Fuel Reprocessing Facilities

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

JAEA-Data/Code 2021-008, 35 Pages, 2021/08

JAEA-Data-Code-2021-008.pdf:3.68MB

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. In addition to this, nitrogen oxides (NO$$_{rm x}$$) are also released formed by the thermal decomposition of metal nitrates of fission products (FP) in HLLW. It has been observed experimentally that NOx affects to the migration behavior of Ru at the anticipated atmosphere condition in cells and/or compartments of the facility building. Chemical reactions of NO$$_{rm x}$$ with water and nitric acid are also recognized as the complex phenomena to undergo simultaneously in the vapor and liquid phases. The analysis program, SCHERN has been under developed to simulate chemical behavior including Ru coupled with the thermo-hydraulic condition in the flow paths in the facility building. This technical guide for SCHERN-V2 presents the overview of covered accident, analytical models including newly developed models, differential equations for numerical solution, and user instructions.

Journal Articles

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

 Times Cited Count:0 Percentile:0.02(Nuclear Science & Technology)

Journal Articles

Verification of probabilistic fracture mechanics analysis code for reactor pressure vessel

Li, Y.; Katsumata, Genshichiro*; Masaki, Koichi; Hayashi, Shotaro*; Itabashi, Yu*; Nagai, Masaki*; Suzuki, Masahide*; Kanto, Yasuhiro*

Journal of Pressure Vessel Technology, 143(4), p.041501_1 - 041501_8, 2021/08

 Times Cited Count:0 Percentile:0(Engineering, Mechanical)

Journal Articles

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

 Times Cited Count:0 Percentile:0.02(Nuclear Science & Technology)

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.

Journal Articles

Conversion factors bridging radioactive fission product distributions in the primary containment vessel of Fukushima Daiichi NPP and dose rates measured by the containment atmosphere monitoring system

Uchida, Shunsuke; Pellegrini, M.*; Naito, Masanori*

Nuclear Engineering and Design, 380, p.111303_1 - 111303_11, 2021/08

 Times Cited Count:1 Percentile:83.53(Nuclear Science & Technology)

Multi-term FP analysis procedures were developed to determine FP distribution all over F1 not only for analyzing accident propagation but also for planning its decommissioning projects. They should be validated based on the measured FP data. One of the useful tools for their validation was application of the dose rate data monitored by the containment atmosphere monitoring system (CAMS). However, in order to compare the data with different characteristics and dimensional units, e.g., FP distribution (kg, Bq) and dose rate (Sv/h), application of the conversion factors bridging them would be effective and useful. In order to prepare speedy, easy-to-handle and tractable procedures to calculate radiation dose rates at the CAMS detector locations, dose rate conversion factors were determined for major source locations and major radionuclides. The dose rates could be easily calculated by multiplying FP amounts obtained with the multiterm FP analysis procedures by the conversion factors.

JAEA Reports

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.

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