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Journal Articles

Oxidation behavior of high-burnup advanced fuel cladding tubes in high-temperature steam

Narukawa, Takafumi; Amaya, Masaki

Journal of Nuclear Science and Technology, 56(7), p.650 - 660, 2019/07

Journal Articles

Model updates and performance evaluations on fuel performance code FEMAXI-8 for light water reactor fuel analysis

Udagawa, Yutaka; Amaya, Masaki

Journal of Nuclear Science and Technology, 56(6), p.461 - 470, 2019/06

no abstracts in English

JAEA Reports

Mechanical properties database of reactor pressure vessel steels related to fracture toughness evaluation

Tobita, Toru; Nishiyama, Yutaka; Onizawa, Kunio

JAEA-Data/Code 2018-013, 60 Pages, 2018/11

JAEA-Data-Code-2018-013.pdf:1.67MB

Mechanical properties of materials including fracture toughness are extremely important for evaluating the structural integrity of reactor pressure vessels (RPVs). In this report, the published data of mechanical properties of nuclear RPVs steels, including neutron irradiated materials, acquired by the Japan Atomic Energy Agency (JAEA), specifically tensile test data, Charpy impact test data, drop-weight test data, and fracture toughness test data, are summarized. There are five types of RPVs steels with different toughness levels equivalent to JIS SQV2A (ASTM A533B Class 1) containing impurities in the range corresponding to the early plant to the latest plant. In addition to the base material of RPVs, the mechanical property data of the two types of stainless overlay cladding materials used as the lining of the RPV are summarized as well. These mechanical property data are organized graphically for each material and listed in tabular form to facilitate easy utilization of data.

Journal Articles

Effects of ballooning and rupture on the fracture resistance of Zircaloy-4 fuel cladding tube after LOCA-simulated experiments

Yumura, Takanori; Amaya, Masaki

Annals of Nuclear Energy, 120, p.798 - 804, 2018/10

 Percentile:100(Nuclear Science & Technology)

Journal Articles

Corrosion behaviour of FeCrAl-ODS steels in nitric acid solutions with several temperatures

Takahatake, Yoko; Ambai, Hiromu; Sano, Yuichi; Takeuchi, Masayuki; Koizumi, Kenji; Sakamoto, Kan*; Yamashita, Shinichiro

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 9 Pages, 2018/10

The corrosion behaviour of FeCrAl-ODS steels for the accident tolerant fuel cladding of LWRs were investigated in nitric acid solutions for the reprocessing process of spent fuels. The corrosion tests were carried out at 60$$^{circ}$$C, 80$$^{circ}$$C and the boiling point of the solutions, and the specimens were then analysed by XPS. The corrosion remarkably progressed at the boiling point, and the highest corrosion rate was 0.22 mm/y. In the oxide film, the atomic concentration of Fe was lower, than that in the base material, and those of Cr and Al were higher. The results show that the corrosion of FeCrAl-ODS steels in hot nitric acid solution is not severe because of the high corrosion resistance of the oxide film formed on the material; hence, the corrosion resistance of the new cladding materials in the dissolution process of spent fuel is acceptable for reprocessing operations.

Journal Articles

OECD/NEA benchmark on pellet-clad mechanical interaction modelling with fuel performance codes; Impact of number of radial pellet cracks and pellet-clad friction coefficient

Dost$'a$l, M.*; Rossiter, G.*; Dethioux, A.*; Zhang, J.*; Amaya, Masaki; Rozzia, D.*; Williamson, R.*; Kozlowski, T.*; Hill, I.*; Martin, J.-F.*

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 10 Pages, 2018/10

Journal Articles

Application of Bayesian optimal experimental design to reduce parameter uncertainty in the fracture boundary of a fuel cladding tube under LOCA conditions

Narukawa, Takafumi; Yamaguchi, Akira*; Jang, S.*; Amaya, Masaki

Proceedings of 14th International Conference on Probabilistic Safety Assessment and Management (PSAM-14) (USB Flash Drive), 10 Pages, 2018/09

Journal Articles

Coupled computer code study on irradiation performance of a fast reactor mixed oxide fuel element with an emphasis on the fission product cesium behavior

Uwaba, Tomoyuki; Nemoto, Junichi*; Ishitani, Ikuo*; Ito, Masahiro*

Nuclear Engineering and Design, 331, p.186 - 193, 2018/05

 Percentile:100(Nuclear Science & Technology)

A computer code for the analysis of the overall irradiation performance of a fast reactor mixed-oxide (MOX) fuel element was coupled with a specialized code for the analysis of fission product cesium behaviors in a MOX fuel element. The coupled code system allowed for the analysis of the radial and axial Cs migrations, the generation of Cs chemical compounds and fuel swelling due to Cs-fuel-reactions in association with the thermal and mechanical behaviors of the fuel element. The coupled code analysis was applied to the irradiation performance of a fast reactor MOX fuel element attaining high burnup for discussion on the axial distribution of Cs, fuel-to-cladding mechanical interaction owing to the Cs-fuel-reactions by comparing the calculated results with post irradiation examinations.

JAEA Reports

Cladding tube burst experiment assumed MA fuel pin for transmutation physics experimental facility

Sugawara, Takanori; Tsujimoto, Kazufumi

JAEA-Research 2017-011, 35 Pages, 2017/10

JAEA-Research-2017-011.pdf:4.88MB

The construction of Transmutation Physics Experimental Facility (TEF-P) is planned in the J-PARC project. TEF-P is a critical assembly and it will treat minor actinide (MA) fuel in the experiment. The temperature when the air cooling for the TEF-P core would stop was estimated but there were no data to evaluate the soundness of the MA fuel pin. To set a tentative limit temperature for the TEF-P core, cladding tube burst experiment was performed. As the result, the cladding tube burst occurred at 660$$^{circ}$$C as the severest case. Through these results and the estimation of creep rupture time, the tentative limit temperature for the TEF-P core was set to 600$$^{circ}$$C.

Journal Articles

Oxidation kinetics of Zry-4 fuel cladding in mixed steam-air atmospheres at temperatures of 1273 - 1473 K

Negyesi, M.; Amaya, Masaki

Journal of Nuclear Science and Technology, 54(10), p.1143 - 1155, 2017/10

 Times Cited Count:2 Percentile:42.02(Nuclear Science & Technology)

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

Welding technology R&D of Japanese accident tolerant fuel claddings of FeCrAl-ODS steel for BWRS

Kimura, Akihiko*; Yuzawa, Sho*; Sakamoto, Kan*; Hirai, Mutsumi*; Kusagaya, Kazuyuki*; Yamashita, Shinichiro

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

The effect of Al addition on the PRW weldability of ODS steel is shown with the discussion focusing on the microstructure changes by the welding. The ordinary welding methods including electron beam (EB) welding and tungsten inert gas (TIG) welding were also applied to the SUS430 endcap welding to cladding tube made of FeCrAl-ODS steel. The endcap welded ODS steel tube samples were tensile tested at RT. The EB welded FeCrAl-ODS/SUS430 samples broke in the ODS steel tube, indicating that the weld bond is stronger than the ODS base metal. However, the TIG welded FeCrAl-ODS/SUS430 samples broke at a weld bond. X-ray CT scan analysis was performed for the weld bond, and the bonding strength was correlated with the X-ray CT results in order to assess the feasibility of those welding methods for ATF-ODS steel cladding.

Journal Articles

The Applicability of SiC-SiC fuel cladding to conventional PWR power plant

Furumoto, Kenichiro*; Watanabe, Seiichi*; Yamamoto, Teruhisa*; Teshima, Hideyuki*; Yamashita, Shinichiro; Saito, Hiroaki; Shirasu, Noriko

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

Since 2015, Mitsubishi Nuclear Fuel (MNF) has joined in a Japanese R&D project of ATF founded by the Ministry of Economy, Trade and Industry (METI) as a subcontractor to Japan Atomic Energy Agency (JAEA) which is the prime contractor to METI. In this program, MNF plans to evaluate an influence of Silicon Carbide (SiC) composite cladding upon fuel rod behavior in current pressurized water reactors (PWR). This paper reports the evaluation result of the applicability of fuel rod with SiC composite cladding for a conventional PWR. For the applicability evaluations of SiC composite to conventional PWR, both of analytical evaluations and out-of-pile tests for SiC composite were conducted. Analytical evaluations were performed by Mitsubishi's own fuel rod design code and the fuel rod behavior evaluation code developed by JAEA. These codes were modified to evaluate the behavior of the fuel rod with SiC composite cladding. As out-of-pile tests, thermal diffusivity measurement and autoclave corrosion test for SiC composite samples were performed. Test apparatus were developed for evaluation of performance of SiC composite under the condition simulated design basis accident (DBA).

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

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

Ultrasonic guided wave approach for inspecting concave surface of the laser butt-welded pipe

Furusawa, Akinori; Nishimura, Akihiko; Takebe, Toshihiko*; Nakamura, Masaki*; Takenaka, Yusuke*; Saijo, Shingo*; Nakamoto, Hiroyuki*

E-Journal of Advanced Maintenance (Internet), 9(2), p.44 - 51, 2017/08

The aim of this work is to investigate the applicability of ultrasonic guided wave for evaluation of laser beam butt-welding quality. Ten in total test pipes having welding seam is prepared. Two piece of pipe are jointed and continuous laser beam is irradiated on the edges, varying laser irradiation power, welding side and surface profile of the adjacent edges of the pipe. Ultrasonic guided wave testing experiment is performed on the pipes. Torsional mode guided wave is excited by EMAT. The experimental results are analyzed and issues are discussed. The reflection wave bullet from the poor interface of the welding seam is clearly observed, whereas no reflection from fine welded line. From the aspect of laser irradiation power, welding side and surface profile of the adjacent edges, the relation between the interface condition and detection wave bullet are analyzed. It is found that the ultrasonic guided wave technologies have the potential for evaluating laser beam butt-welding seam.

Journal Articles

Development of core and structural materials for fast reactors

Asayama, Tai; Otsuka, Satoshi

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 15 Pages, 2017/06

This paper summarizes ongoing efforts in Japan Atomic Energy Agency on the development of core and structural materials for sodium-cooled fast reactors. For core materials, oxide dispersion strengthened (ODS) steels and 11Cr ferritic steel (PNC-FMS) will be used for the fuel pin cladding and wrapper tube, respectively. As for ODS steel, 9Cr- and 11Cr-ODS steels have been extensively developed. Their laboratory-scale manufacturing technology has been developed including reliability improvement in tube microstructure and strength homogeneity. As for the PNC-FMS wrapper tube, the development of a dissimilar joining technique with type 316 steel and properties evaluation of dissimilar welds have been carried out. For structural materials, codification of 316FR stainless steel and Modified 9Cr-1Mo steel is ongoing. Acquisition and collection of long-term data of base metal and welded joints are continued and evaluation methodologies are being developed to establish a technical basis for 60-year design.

Journal Articles

The Effect of azimuthal temperature distribution on the ballooning and rupture behavior of Zircaloy-4 cladding tube under transient-heating conditions

Narukawa, Takafumi; Amaya, Masaki

Journal of Nuclear Science and Technology, 53(11), p.1758 - 1765, 2016/11

 Times Cited Count:3 Percentile:42.29(Nuclear Science & Technology)

Journal Articles

Seawater effects on the soundness of spent fuel cladding tube

Motooka, Takafumi; Ueno, Fumiyoshi; Yamamoto, Masahiro

Proceedings of 2016 EFCOG Nuclear & Facility Safety Workshop (Internet), 6 Pages, 2016/09

At the Fukushima Daiichi nuclear accident, seawater was injected into spent fuel pools of Unit 2-4 for the emergency cooling. Seawater might cause localized corrosion of spent fuel cladding. This would lead to leakage of not only fissile materials but also fission products from fuel cladding. The behavior, however, is not understood well. In this paper, the effects of seawater on corrosion behavior and mechanical property of were studied by using a spent fuel cladding from a BWR. We immersed the spent cladding tubes in diluted artificial seawater for 300h at 353 K, and conducted their visual, cross-sectional and strength examinations. As a localized corrosion index, the pitting potentials of specimens fabricated from the cladding were measured as functions of chloride ion concentration ranging from 20 to 2500 ppm. The visual examination showed that localized corrosion has not occurred, and cross-sectional examination showed no cracks. The strength of immersed tubes was comparable to that of non-immersed tubes. Additionally, pitting potential could not be measured over 1.0 V; pitting corrosion was hardly occurred. These results suggested that the specimens from the spent fuel cladding tube was very resistant to localized corrosion.

Journal Articles

Recent research activities using NSRR on safety related issues

Udagawa, Yutaka; Sugiyama, Tomoyuki*; Amaya, Masaki

Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1183 - 1189, 2016/04

Journal Articles

The Effect of oxidation and crystal phase condition on the ballooning and rupture behavior of Zircaloy-4 cladding tube-under transient-heating conditions

Narukawa, Takafumi; Amaya, Masaki

Journal of Nuclear Science and Technology, 53(1), p.112 - 122, 2016/01

 Times Cited Count:3 Percentile:42.29(Nuclear Science & Technology)

142 (Records 1-20 displayed on this page)