Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 157

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

JAEA Reports

Technical basis of ECCS acceptance criteria for light-water reactors and applicability to high burnup fuel

Nagase, Fumihisa; Narukawa, Takafumi; Amaya, Masaki

JAEA-Review 2020-076, 129 Pages, 2021/03

JAEA-Review-2020-076.pdf:3.9MB

Each light-water reactor (LWR) is equipped with the Emergency Core Cooling System (ECCS) to maintain the coolability of the reactor core and to suppress the release of radioactive fission products to the environment even in a loss-of-coolant accident (LOCA) caused by breaks in the reactor coolant pressure boundary. The acceptance criteria for ECCS have been established in order to evaluate the ECCS performance and confirm the sufficient safety margin in the evaluation. The limits defined in the criteria were determined in 1975 and reviewed based on state-of-the-art knowledge in 1981. Though the fuel burnup extension and necessary improvements of cladding materials and fuel design have been conducted, the criteria have not been reviewed since then. Meanwhile, much technical knowledge has been accumulated regarding the behavior of high-burnup fuel during LOCAs and the applicability of the criteria to the high-burnup fuel. This report provides a comprehensive review of the history and technical bases of the current criteria and summarizes state-of-the-art technical findings regarding the fuel behavior during LOCAs. The applicability of the current criteria to the high-burnup fuel is also discussed.

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

Four-point-bend tests on high-burnup advanced fuel cladding tubes after exposure to simulated LOCA conditions

Narukawa, Takafumi; Amaya, Masaki

Journal of Nuclear Science and Technology, 57(7), p.782 - 791, 2020/07

 Times Cited Count:2 Percentile:66.08(Nuclear Science & Technology)

Journal Articles

Fracture-mechanics-based evaluation of failure limit on pre-cracked and hydrided Zircaloy-4 cladding tube under biaxial stress states

Li, F.; Mihara, Takeshi; Udagawa, Yutaka; Amaya, Masaki

Journal of Nuclear Science and Technology, 57(6), p.633 - 645, 2020/06

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Analytical study of SPERT-CDC test 859 using fuel performance codes FEMAXI-8 and RANNS

Taniguchi, Yoshinori; Udagawa, Yutaka; Amaya, Masaki

Annals of Nuclear Energy, 139, p.107188_1 - 107188_7, 2020/05

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

The Effect of base irradiation on failure behaviors of UO$$_{2}$$ and chromia-alumina additive fuels under simulated reactivity-initiated accidents; A Comparative analysis with FEMAXI-8

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

Annals of Nuclear Energy, 139, p.107268_1 - 107268_9, 2020/05

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Fracture limit of high-burnup advanced fuel cladding tubes under loss-of-coolant accident conditions

Narukawa, Takafumi; Amaya, Masaki

Journal of Nuclear Science and Technology, 57(1), p.68 - 78, 2020/01

 Times Cited Count:1 Percentile:24.18(Nuclear Science & Technology)

Journal Articles

Behavior of high-burnup advanced LWR fuel cladding tubes under LOCA conditions

Narukawa, Takafumi; Amaya, Masaki

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.912 - 921, 2019/09

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

Overview of accident-tolerant fuel R&D program in Japan

Yamashita, Shinichiro; Ioka, Ikuo; Nemoto, Yoshiyuki; Kawanishi, Tomohiro; Kurata, Masaki; Kaji, Yoshiyuki; Fukahori, Tokio; Nozawa, Takashi*; Sato, Daiki*; Murakami, Nozomu*; et al.

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.206 - 216, 2019/09

After the nuclear accident at Fukushima Daiichi Power Plant, research and development (R&D) program for establishing technical basis of accident-tolerant fuel (ATF) started from 2015 in Japan. Since then, both experimental and analytical studies necessary for designing a new light water reactor (LWR) core with ATF candidate materials are being conducted within the Japanese ATF R&D Consortium for implementing ATF to the existing LWRs, accompanying with various technological developments required. Until now, we have accumulated experimental data of the candidate materials by out-of-pile tests, developed fuel evaluation codes to apply to the ATF candidate materials, and evaluated fuel behavior simulating operational and accidental conditions by the developed codes. In this paper, the R&D progresses of the ATF candidate materials considered in Japan are reviewed based on the information available such as proceedings of international conference and academic papers, providing an overview of ATF program in Japan.

Journal Articles

Benchmark of fuel performance codes for FeCrAl cladding behavior analysis

Pastore, G.*; Gamble, K. A.*; Cherubini, M.*; Giovedi, C.*; Marino, A.*; Yamaji, Akifumi*; Kaji, Yoshiyuki; Van Uffelen, P.*; Veshchunov, M.*

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.1038 - 1047, 2019/09

Oxidation-resistant iron-chromium-aluminum (FeCrAl) steels have been proposed for application as cladding materials in light water reactor fuel rods with improved accident tolerance. Within the Coordinated Research Project ACTOF of the International Atomic Energy Agency (IAEA), a fuel performance modeling benchmark for FeCrAl cladding behavior was conducted. During this effort, calculations were performed with various fuel performance codes for a set of fuel rod problems with FeCrAl steel as cladding material, and results were compared to each other.

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

 Times Cited Count:6 Percentile:82.71(Nuclear Science & Technology)

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

 Times Cited Count:3 Percentile:58.77(Nuclear Science & Technology)

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

 Times Cited Count:3 Percentile:47(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

157 (Records 1-20 displayed on this page)