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Journal Articles

Transient response of LWR fuels (RIA)

Udagawa, Yutaka; Fuketa, Toyoshi*

Comprehensive Nuclear Materials, 2nd Edition, Vol.2, p.322 - 338, 2020/08

Journal Articles

Numerical simulation of two-phase flow in 4$$times$$4 simulated bundle

Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki

Mechanical Engineering Journal (Internet), 7(3), p.19-00583_1 - 19-00583_12, 2020/06

JAEA is implementing the 3D detailed nuclear-thermal-coupled analysis code to analyze the transition state of the core and to reduce the likelihood of the design. In the development plan, the computational fluid dynamics code based on the VOF method, JUPITER, is applied for TH part of the 3D detailed nuclear-thermal-coupled analysis code.

Journal Articles

Irradiation growth behavior of improved Zr-based alloys for fuel cladding

Amaya, Masaki; Kakiuchi, Kazuo; Mihara, Takeshi

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.1048 - 1056, 2019/09

Journal Articles

Behavior of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

Taniguchi, Yoshinori; Udagawa, Yutaka; Mihara, Takeshi; Amaya, Masaki; Kakiuchi, Kazuo

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.551 - 558, 2019/09

Journal Articles

Study on the two-phase flow in simulated LWR fuel bundle by CFD code

Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.666 - 677, 2019/08

An evaluation methodology of critical heat fluxes (CHFs) based on a mechanism for fuel assemblies in light water reactors (LWRs) is needed in order to design and evaluate the safety for the fuel assemblies in LWRs. In our study, the numerical simulation with surface-tracking will be applied for the two-phase flow in fuel assemblies in order to obtain the detail data relating to the size and velocity of bubbles in the subchannel, which is needed to predict the CHF based on the mechanism. In this study, the numerical simulation of two-phase flow in 4$$times$$4 bundle was implemented by using JUPITER in order to establish the evaluation method of the size and velocity of bubbles by the numerical simulation, which is the multi-physics simulation code and enable to track the gas-liquid surface. The simulation results are validated by the curve of flow regime for air-water under the adiabatic condition. The bubble and velocity of bubbles obtained by simulation results are analyzed.

JAEA Reports

Progress report on Nuclear Safety Research Center (JFY 2015 - 2017)

Nuclear Safety Research Center, Sector of Nuclear Safety Research and Emergency Preparedness

JAEA-Review 2018-022, 201 Pages, 2019/01

JAEA-Review-2018-022.pdf:20.61MB

Nuclear Safety Research Center (NSRC), Sector of Nuclear Safety Research and Emergency Preparedness, Japan Atomic Energy Agency (JAEA) is conducting technical support to nuclear safety regulation and safety research based on the Mid-Long Term Target determined by Japanese government. This report summarizes the research structure of NSRC and the cooperative research activities with domestic and international organizations as well as the nuclear safety research activities and results in the period from JFY 2015 to 2017 on the nine research fields in NSRC; (1) severe accident analysis, (2) radiation risk analysis, (3) safety of nuclear fuels in light water reactors (LWRs), (4) thermohydraulic behavior under severe accident in LWRs, (5) materials degradation and structural integrity, (6) safety of nuclear fuel cycle facilities, (7) safety management on criticality, (8) safety of radioactive waste management, and (9) nuclear safeguards.

JAEA Reports

Development of fuel performance code FEMAXI-8; Model improvements for light water reactor fuel analysis and systematic validation

Udagawa, Yutaka; Yamauchi, Akihiro*; Kitano, Koji*; Amaya, Masaki

JAEA-Data/Code 2018-016, 79 Pages, 2019/01

JAEA-Data-Code-2018-016.pdf:2.75MB

FEMAXI-8 is the latest version of the fuel performance code FEMAXI developed by JAEA. A systematic validation work has been achieved against 144 irradiation test cases, after many efforts have been made, in development of new models, improvements in existing models and the code structure, bug-fixes, construction of irradiation-tests database and other infrastructures.

Journal Articles

Technical basis of accident tolerant fuel updated under a Japanese R&D project

Yamashita, Shinichiro; Nagase, Fumihisa; Kurata, Masaki; Nozawa, Takashi; Watanabe, Seiichi*; Kirimura, Kazuki*; Kakiuchi, Kazuo*; Kondo, Takao*; Sakamoto, Kan*; Kusagaya, Kazuyuki*; et al.

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

In Japan, the research and development (R&D) project on accident tolerant fuel and other components (ATFs) of light water reactors (LWRs) has been initiated in 2015 for establishing technical basis of ATFs. The Japan Atomic Energy Agency (JAEA) has coordinated and carried out this ATF R&D project in cooperation with power plant providers, fuel venders and universities for making the best use of the experiences, knowledges in commercial uses of zirconium-base alloys (Zircaloy) in LWRs. ATF candidate materials under consideration in the project are FeCrAl steel strengthened by dispersion of fine oxide particles(FeCrAl-ODS) and silicon carbide (SiC) composite, and are expecting to endure severe accident conditions in the reactor core for a longer period of time than the Zircaloy while maintaining or improving fuel performance during normal operations. In this paper, the progresses of the R&D project are reported.

Journal Articles

Performance degradation of candidate accident-tolerant cladding under corrosive environment

Nagase, Fumihisa; Sakamoto, Kan*; Yamashita, Shinichiro

Corrosion Reviews, 35(3), p.129 - 140, 2017/08

 Times Cited Count:8 Percentile:51.83(Electrochemistry)

As the lessons learnt from the accident at the Fukushima Daiichi Nuclear Power Station, advanced cladding materials are being developed to enhance accident tolerance comparing with conventional zirconium alloys. The present paper reviews the progress of the development and summarizes subjects to be solved for the enhanced accident-tolerance fuel cladding, focusing on performance degradation under various corrosive environmental conditions that should be considered in designing the LWR fuel.

Journal Articles

Parametric survey on possible impact of partitioning and transmutation of high-level radioactive waste

Oigawa, Hiroyuki; Yokoo, Takeshi*; Nishihara, Kenji; Morita, Yasuji; Ikeda, Takao*; Takaki, Naoyuki*

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

The benefit of implementing Partitioning and Transmutation (P&T) of high-level wastes was parametrically surveyed. The possible reduction of the geological repository area was estimated. By recycling minor actinides (MA), the repository area required for unit spent fuel was reduced significantly in the case of MOX-LWR. This effect was caused by removal of $$^{241}$$Am which is a long-term heat source. By partitioning the fission products, in addition to MA recycling, further 70-80% reduction from the MA-recovery case can be expected for both UO$$_2$$ and MOX. This significant reduction was independent of the cooling time before the partitioning process.

Journal Articles

Predicted two-phase flow structure in a fuel bundle of an advanced light-water reactor

Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Tamai, Hidesada

Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 14 Pages, 2004/10

no abstracts in English

Journal Articles

Update status of benchmark activity for reactor physics study of LWR next generation fuels

Unesaki, Hironobu*; Okumura, Keisuke; Kitada, Takanori*; Saji, Etsuro*

Transactions of the American Nuclear Society, 88, p.436 - 438, 2003/06

In order to investigate the calculation accuracy of the nuclear characteristics of LWR next generation fuels, the Research Committee on Reactor Physics organized by JAERI has proposed "Reactor Physics Benchmark for LWR Next Generation Fuels". The next generation fuels aim at very high burn-up of about 70GWd/t in PWR or BWR with UO$$_{2}$$ or MOX fuels whose fissile enrichments may exceed the Japanese regulatory limitations for the current LWR fuels such as 5wt.% U-235. Until now, twelve organizations have pariticipated in the benchmark activity. From the comparison with the cell burn-up calculation results using different codes and library data, status of the calculation accuracy and future subjects are clarified.

JAEA Reports

Study on core physics characteristics of high burnup full MOX PWR core, 2

Kugo, Teruhiko; Okubo, Tsutomu; *

JAERI-Research 99-057, p.29 - 0, 1999/09

JAERI-Research-99-057.pdf:1.77MB

no abstracts in English

JAEA Reports

Study on nuclear physics of high burnup full MOX PWR core

; Shimada, Shoichiro*; Okubo, Tsutomu; Ochiai, Masaaki

JAERI-Research 98-059, 40 Pages, 1998/10

JAERI-Research-98-059.pdf:1.73MB

no abstracts in English

JAEA Reports

The Isotopic compositions database system on spent fuels in light water reactors (SFCOMPO)

Kurosawa, Masayoshi; Naito, Yoshitaka; ; *

JAERI-Data/Code 96-036, 156 Pages, 1997/02

JAERI-Data-Code-96-036.pdf:3.22MB

no abstracts in English

JAEA Reports

Data book of the isotopic composition of spent fuel in light water reactors

Naito, Yoshitaka; Kurosawa, Masayoshi; *

JAERI-M 94-034, 225 Pages, 1994/03

JAERI-M-94-034.pdf:5.59MB

no abstracts in English

JAEA Reports

Databook of the isotopic composition of spent fuel in light water reactors

Naito, Yoshitaka; Kurosawa, Masayoshi; *

JAERI-M 93-061, 225 Pages, 1993/03

JAERI-M-93-061.pdf:5.45MB

no abstracts in English

JAEA Reports

Control rod effects on reaction rate distributions in tight pitched PuO$$_{2}$$-UO$$_{2}$$ fuel assembly

C-S.Gil*; Okumura, Keisuke;

JAERI-M 91-200, 61 Pages, 1991/11

JAERI-M-91-200.pdf:1.26MB

no abstracts in English

Journal Articles

Study of pellet-cladding interaction on light water reactor fuel, (I); PWR type fuel rod

; *; E.Kolstad*

Nihon Genshiryoku Gakkai-Shi, 28(7), p.641 - 657, 1986/00

 Times Cited Count:4 Percentile:48.61(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Out-pile FP Release Experiment by Heating of Irradiated UO$$_{2}$$ Pellet

;

JAERI-M 85-199, 16 Pages, 1985/12

JAERI-M-85-199.pdf:0.68MB

no abstracts in English

28 (Records 1-20 displayed on this page)