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Journal Articles

Behavior of fuel with zirconium alloy cladding in reactivity-initiated accident and loss-of-coolant accident

Fuketa, Toyoshi*; Nagase, Fumihisa

Zirconium in the Nuclear Industry; 18th International Symposium (ASTM STP 1597), p.52 - 92, 2018/01

Extensive research programs have been performed for more than two decades in JAEA and a better understanding has been developed for fuel behavior under accident conditions. The program is comprised of: RIA studies including pulse-irradiation experiments in the NSRR, cladding mechanical tests, and development and verification of a computer code RANNS; LOCA tests including integral thermal shock tests, oxidation rate measurements, and cladding mechanical tests; development and verification of a computer code FEMAXI-6, etc. Data and findings from the research programs provided technical basis directly and indirectly for regulatory criteria in Japan and other countries. This paper reviews and summarizes the major outcome from the research programs and identifies further research needs, as the acceptance technical paper for the Kroll Medal award of ASTM.

Journal Articles

Fuel behavior analysis for accident tolerant fuel with sic cladding using adapted FEMAXI-7 code

Shirasu, Noriko; Saito, Hiroaki; Yamashita, Shinichiro; Nagase, Fumihisa

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 8 Pages, 2017/09

Silicon carbide (SiC) is an attractive candidate of accident tolerant fuel (ATF) cladding material because of its high chemical stability, high radiation resistance and low neutron absorption. FEMAXI-ATF has been developed to analysis SiC cladding fuel behaviors. The thermal, mechanical and irradiation property models were implemented to FEMAXI-7, which is a fuel behavior analysis code being developed in JAEA. Fuel rod behavior analysis was performed under typical boiling water reactor (BWR) operating conditions with a model based on a 9$$times$$9 BWR fuel (Step III Type B), in which the cladding material was replaced from Zircaloy to SiC. The SiC cladding shows large swelling by irradiation. It increases the gap size and decreases cladding thermal conductivity. The mechanism of relaxation of stress is also different from the Zircaloy cladding. The experimental data for SiC materials are still insufficient to construct the models, especially for evaluating fracture behavior.

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

Improving the corrosion resistance of silicon carbide for fuel in BWR environments by using a metal coating

Ishibashi, Ryo*; Tanabe, Shigetada*; Kondo, Takao*; Yamashita, Shinichiro; Nagase, Fumihisa

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

For improving the corrosion resistance of silicon carbide (SiC) in boiling-water-reactor environments, corrosion-resistant coatings on SiC were evaluated. Due to its hydrogen-generation rate and reaction heat being lower than those of conventional Zircaloy, SiC is expected to be an appropriate material for accident-tolerant fuels. However, there are still many critical issues with the practical application of SiC fuel cladding and fuel channel boxes, one of which is hydrothermal corrosion. Silicon carbide is chemically stable, but silicon oxide formed by oxidation of SiC dissolves in high temperature water. Although the rate of SiC dissolution is very small, the dissolution must be suppressed to comply with regulations for dissolved silica concentration in reactor coolant. In this study, the corrosion behavior of candidate coatings for SiC substrates were evaluated before and after exposure to unirradiated high-purity-water environments.

Journal Articles

Safety evaluation of accident tolerant fuel with SiC/SiC cladding

Sato, Hisaki*; Takeuchi, Yutaka*; Kakiuchi, Kazuo*; Yamashita, Shinichiro; Nagase, Fumihisa

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 9 Pages, 2017/09

Since JFY2015, new Japanese national program has been initiated for the purpose of establishing the technical basis to apply ATF for the existing LWRs. SiC is one of ATF candidates material and the comprehensive applicability is being studied in the program, such as fuel rod design, core and plant design, safety evaluation for design basis accident (DBA) and severe accident (SA) as well. As one of the works in the program, the new procedure including fuel rod performance analysis during DBA was developed and the preliminary analysis was conducted. As a result, it was concluded that the typical transient and LOCA behavior between Zircaloy and SiC was not so much different.

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:7 Percentile:38.45(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

Overview and outcomes of the OECD/NEA benchmark study of the accident at the Fukushima Daiichi Nuclear Power Station

Nagase, Fumihisa; Gauntt, R. O.*; Naito, Masanori*

Nuclear Technology, 196(3), p.499 - 510, 2016/12

 Times Cited Count:11 Percentile:13.47(Nuclear Science & Technology)

The OECD/NEA Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) project was established in November 2012. The primary objectives of this benchmark study are to estimate accident progression and status inside the nuclear reactors, including fuel debris distribution, and consequently to contribute to the decommissioning activity at the Fukushima Daiichi Nuclear Power Plant. Fifteen organizations of eight countries calculated thermo-hydraulic behavior inside the three reactors for the time span of about six days from the occurrence of the earthquake with their severe accident integral codes. The submitted results were compared on coolant level change, hydrogen generation, initiation and progression of melt in fuel bundle and control blade, failure of reactor pressure vessel, distribution and composition of molten and solidified materials, and progression of molten core concrete interaction. This issue summarizes the results of the comparison and discussion with still remaining uncertainties and data needs as the output from the project.

Journal Articles

Development of air cooling performance evaluation method for fuel debris on retrieval of Fukushima Daiichi NPS by dry method, 1; Outline of research project

Yoshida, Hiroyuki; Uesawa, Shinichiro; Yamashita, Susumu; Nagase, Fumihisa

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11

Journal Articles

Establishment of technical basis to implement accident tolerant fuels and components to existing LWRs

Yamashita, Shinichiro; Nagase, Fumihisa; Kurata, Masaki; Kaji, Yoshiyuki

Proceedings of Annual Topical Meeting on LWR Fuels with Enhanced Safety and Performance (TopFuel 2016) (USB Flash Drive), p.21 - 30, 2016/09

Fuel rod, channel box, and control rod designed with new materials and concepts have been developed in Japan for increasing accident tolerance of LWRs. In order to efficiently and properly implement the accident tolerant fuels (ATFs) and the other components, it is necessary not only to accumulate fundamental and practical data but also to consider technology readiness, recognize knowledge gaps, and establish strategy for design and fabrication. The Japan Atomic Energy Agency (JAEA) has established the above "technical basis" and drafted a research plan towards implementation of the ATFs and components as a program sponsored and organized by the Ministry of Economy, Trade and Industry (METI). It is useful to take advantage of the experiences in commercial uses of zirconium-base alloys in LWRs and, therefore, JAEA has conducted this METI project in cooperation with power plant providers, fuel venders, research institutes and universities who have been involved in the development of the ATF materials. The present paper describes the main results of the project conducted to establish the technical basis of the ATFs and components.

Journal Articles

Development of accident tolerant control rod for light water reactors

Ota, Hirokazu*; Nakamura, Kinya*; Ogata, Takanari*; Nagase, Fumihisa

Proceedings of Annual Topical Meeting on LWR Fuels with Enhanced Safety and Performance (TopFuel 2016) (USB Flash Drive), p.159 - 168, 2016/09

Control rods can be disintegrated and neutron absorber would be removed from the core region before most of the fuel pins are still not damaged seriously in severe accidents of LWRs. The present study investigates a concept of accident tolerant control rod (ATCR) with the following characteristics; (1) sufficiently-high melting and eutectic temperatures, (2) high miscibility with molten and solidified fuel materials, and (3) enough control rod worth. It has been shown that rare-earth sesqui-oxides are expected to be compatible with iron up to higher temperatures than the melting points of structure materials of control rods, and that Sm$$_{2}$$O$$_{3}$$, Eu$$_{2}$$O$$_{3}$$, Gd$$_{2}$$O$$_{3}$$, Dy$$_{2}$$O$$_{3}$$ or their mixtures with HfO$$_{2}$$ are available as alternative neutron absorbers to conventional Ag-In-Cd alloy.

Journal Articles

Overview and outcomes of Benchmark Study of the Accident at the Fukushima Daiichi NPS (OECD/NEA BSAF Project)

Nagase, Fumihisa; Gauntt, R. O.*; Naito, Masanori*

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.7033 - 7045, 2015/08

The OECD/NEA Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF) Project has been established in November 2012. Fifteen organizations of eight countries calculated thermo-hydraulic behavior with severe accident integral codes. The primary objective of this benchmark study is to estimate accident progression, status in the reactor pressure vessels and primary containment vessels, and status of debris distribution for a debris removal plan. Finally the calculated results submitted by the participants were compared and evaluated to estimate the accident progression and status inside the reactors though the results showed wide variations. Still remaining uncertainties and data needs that are useful to the communication between analysts and decommissioning activities are also summarized as the output from the project.

Journal Articles

Research program for the evaluation of fission product and actinide release behaviour, focusing on their chemical forms

Miwa, Shuhei; Yamashita, Shinichiro; Ishimi, Akihiro; Osaka, Masahiko; Amaya, Masaki; Tanaka, Kosuke; Nagase, Fumihisa

Energy Procedia, 71, p.168 - 181, 2015/05

BB2013-2241.pdf:0.88MB

 Times Cited Count:11 Percentile:0.42

A basic study towards enhanced safety management of irradiated fuels and materials from a severe accident is underway utilizing JAEA's hot laboratory complex in Oarai. The present study that consists of three basic research programs is aimed at contributing to building enhanced safety management measures (including radioactive decontamination, evaluation measurements, safekeeping, treatment and disposal) of irradiated fuels and materials from the severe accident. In this paper, not only the overview of activities of individual research programs but also the several preliminary results were shown together with future plans.

Journal Articles

Evaluation of fracture resistance of ruptured, oxidized, and quenched Zircaloy cladding by four-point-bend tests

Yamato, Masaaki; Nagase, Fumihisa; Amaya, Masaki

Journal of Nuclear Science and Technology, 51(9), p.1125 - 1132, 2014/09

 Times Cited Count:6 Percentile:50.38(Nuclear Science & Technology)

To evaluate fracture resistance of LWR fuel rods under LOCA and post-LOCA cooling conditions, four-point-bend tests were performed on non-irradiated Zircaloy cladding samples that were ruptured, oxidized in high-temperature steam, and quenched in flooding water. The bend test methodology was designed to apply a uniform bending moment to the entire rupture region and to generate tensile stress on the ruptured side. The fracture bending moment of the cladding decreased with oxidation temperature and hydrogen concentration as well as oxidation amount. Comparison with bending moments estimated from design basis seismic ground motion indicated that the cladding is unlikely to be fractured by seismic loads during post-LOCA cooling if high-temperature oxidation is kept below 15% ECR, the oxidation limit of the Japanese LOCA criteria.

Journal Articles

Research program for the evaluation of fission product release and transport behavior focusing on FP chemistry

Sato, Isamu; Miwa, Shuhei; Tanaka, Kosuke; Nakajima, Kunihisa; Hirosawa, Takashi; Iwasaki, Maho; Onishi, Takashi; Osaka, Masahiko; Takai, Toshihide; Amaya, Masaki; et al.

Proceedings of 2014 Water Reactor Fuel Performance Meeting/ Top Fuel / LWR Fuel Performance Meeting (WRFPM 2014) (USB Flash Drive), 6 Pages, 2014/09

A new research program on severe accidents is lunched for the evaluation of FP release and transport behavior in BWR system. The purpose of the program is to improve the FP release and transport model using experimental database about FP chemistry focusing on Cs and I chemistry. In this program, effects of B including in control rod materials, B$$_{4}$$C for the Cs and I chemistry are paid attention. The experimental database used for the improvement will consist of results to obtain with newly-prepared test device under atmosphere with broad-ranging oxygen and/or steam partial pressure simulated those in BWR. The state of preparation for these experimental studies and analyses is introduced. In addition, the preliminary test was moved into action to show B chemical effect on Cs and I transport under one of the processes, which is deposited Cs compounds and B vapor and aerosol interaction. In this experiment, a "B stripping effect" to deposited CsI was observed.

Journal Articles

Thermal-hydraulic experiments with sodium chloride aqueous solution

Jiao, L.; Liu, W.; Nagatake, Taku; Takase, Kazuyuki; Yoshida, Hiroyuki; Nagase, Fumihisa

Proceedings of 15th International Heat Transfer Conference (IHTC 2014) (USB Flash Drive), 11 Pages, 2014/08

In Fukushima Daiichi nuclear disaster, seawater was injected into the nuclear core, which may change the heat transfer characteristics in the reactor pressure vessels (RPV) due to the different physical properties of seawater and pure water. To remove molten fuel from the Fukushima Daiichi Nuclear Power Plants, it is necessary to know the current status of the reactors. Therefore, in this paper, we measured the basic thermal-hydraulic data in an annular tube with a co-axial heater, which includes the heat transfer rate and the pressure drop, using the sodium chloride aqueous solutions and the synthetic seawater as working fluids. The experiments were performed under atmosphere pressure, with the salinity, the fluid mass flux, the inlet temperature and the heat flux used as the parameters. The experimental results and analyses are reported in this paper and the basic influence of the salinity on the heat transfer and the hydraulic characters are proposed.

Journal Articles

Development of numerical evaluation method for fluid dynamics effects on jet breakup phenomena in BWR lower plenum

Suzuki, Takayuki; Yoshida, Hiroyuki; Nagase, Fumihisa

Journal of Nuclear Science and Technology, 51(7-8), p.968 - 976, 2014/07

AA2013-0900.pdf:0.61MB

 Times Cited Count:4 Percentile:57.92(Nuclear Science & Technology)

Journal Articles

Subjects on criticality safety for fuel debris retrieval; Knowledge and consideration about characteristics of fuel debris

Nagase, Fumihisa

Nippon Genshiryoku Gakkai-Shi, 56(4), p.235 - 239, 2014/04

To proceed the decommissioning of the Fukushima Daiichi NPP, management of criticality safety is required during and after the fuel debris retrieval. The information about size of fuel debris, degree of mixture of fuel and structure materials, fuel burnup distribution, FP retention in fuel debris, etc. is important in the R&D for the management of criticality safety. The present issue review the knowledge on fuel and core materials behavior during a severe accident and characteristics of TMI-2 debris, and discusses composition and morphology of fuel debris in the Fukushima Daiichi NPP.

Journal Articles

Reduction in the onset time of breakaway oxidation on zircaloy cladding ruptured under simulated LOCA conditions

Yamato, Masaaki; Nagase, Fumihisa; Amaya, Masaki

Journal of Nuclear Materials, 445(1-3), p.78 - 83, 2014/02

 Times Cited Count:3 Percentile:66.96(Materials Science, Multidisciplinary)

Breakaway oxidation, which is characterized by sudden increases in oxidation rate and significant hydrogen absorption after longer exposure in high-temperature steam, was examined for the ballooned and ruptured cladding. As a result, the onset time of breakaway oxidation for ruptured claddings was significantly reduced compared with that for non-ruptured claddings. The onset time of breakaway oxidation was reduced only in the case where thick and porous oxide was formed on the inner surface of the cladding after rupture. Results of the parametric experiments suggest that the tensile stress in the circumferential direction generated by the thick oxide film formed on the inner surface is a main cause of the reduction of the onset time of breakaway oxidation.

JAEA Reports

Research program for the evaluation of fission product and actinide release behavior, focusing on their chemical forms; Position and plan of the research program

Miwa, Shuhei; Amaya, Masaki; Tanaka, Kosuke; Osaka, Masahiko; Nagase, Fumihisa

JAEA-Review 2013-034, 42 Pages, 2013/12

JAEA-Review-2013-034.pdf:7.43MB

We have launched a new research program for the evaluation of fission product and actinide release behaviors in the severe accident, focusing on the chemical forms by improving the severe accident analysis code in response to the strong needs for reinforcement of nuclear safety as well as for implementation of the 1F decommissioning R&D project. In the research program, the CORSOR-M model incorporated in THALES-2 code will be improved through implementing the effects of chemical form and high-temperature chemical interaction between fuel and materials based on the investigation of the release kinetics and chemical stabilities of fission products and actinides.

JAEA Reports

Input/output manual of light water reactor fuel performance code FEMAXI-7 and its related codes

Suzuki, Motoe; Saito, Hiroaki*; Udagawa, Yutaka; Nagase, Fumihisa

JAEA-Data/Code 2013-009, 306 Pages, 2013/10

JAEA-Data-Code-2013-009.pdf:5.73MB

A light water reactor fuel analysis code FEMAXI-7 has been developed, as an extended version from the former version FEMAXI-6, for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which are fully disclosed in the code model description published in the form of another JAEA-Data/Code report. The present manual, which is the very counterpart of this description document, gives detailed explanations of files and operation method of FEMAXI-7 code and its related codes, methods of input/output, sample Input/output, methods of source code modification, subroutine structure, and internal variables in a specific manner in order to facilitate users to perform fuel analysis by FEMAXI-7.

222 (Records 1-20 displayed on this page)