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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.

Journal Articles

Leaching behavior of radionuclides from samples prepared from spent fuel rod comparable to core debris in the 1F NPS

Onishi, Takashi; Maeda, Koji; Katsuyama, Kozo

Journal of Nuclear Science and Technology, 58(4), p.383 - 398, 2021/04

 Times Cited Count:1 Percentile:39.17(Nuclear Science & Technology)

Journal Articles

Post-irradiation examinations of annular mixed oxide fuels with average burnup 4 and 5% FIMA

Cappia, F.*; Tanaka, Kosuke; Kato, Masato; McClellan, K.*; Harp, J.*

Journal of Nuclear Materials, 533, p.152076_1 - 152076_14, 2020/05

 Times Cited Count:1 Percentile:39.17(Materials Science, Multidisciplinary)

Journal Articles

Thresholds for failure of high-burnup LWR fuels by pellet cladding mechanical interaction under reactivity-initiated accident conditions

Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki

Journal of Nuclear Science and Technology, 56(12), p.1063 - 1072, 2019/12

 Times Cited Count:3 Percentile:57.07(Nuclear Science & Technology)

no abstracts in English

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

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:57.07(Nuclear Science & Technology)

no abstracts in English

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

Current status of the next generation fast reactor core & fuel design and related R&Ds in Japan

Maeda, Seiichiro; Oki, Shigeo; Otsuka, Satoshi; Morimoto, Kyoichi; Ozawa, Takayuki; Kamide, Hideki

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06

The next generation fast reactor is being investigated in Japan, aiming at several targets such as "safety", "reduction of environmental burden" and "economic competitiveness". As for the safety aspect, FAIDUS concept is adopted to avoid re-criticality in core destructive accidents. The uranium-plutonium mixed oxide fuel, in which minor actinide elements are included, will be applied to reduce the amount and potential radio-toxicity of radioactive wastes. The high burn-up fuel is pursued to reduce fuel cycle cost. The candidate concept of the core and fuel design, which could satisfy various design criteria by design devisals, has been established. In addition, JAEA is investigating material properties and irradiation behavior of MA-MOX fuel. JAEA is developing the fuel design code especially for the fuel pin with annular pellets of MA-bearing MOX. Furthermore, JAEA is developing oxide dispersion strengthened (ODS) ferritic steel cladding for the high burnup fuel.

Journal Articles

Thermal expansion measurement of (U,Pu)O$$_{2-x}$$ in oxygen partial pressure-controlled atmosphere

Kato, Masato; Ikusawa, Yoshihisa; Sunaoshi, Takeo*; Nelson, A. T.*; McClellan, K. J.*

Journal of Nuclear Materials, 469, p.223 - 227, 2016/02

 Times Cited Count:7 Percentile:66.89(Materials Science, Multidisciplinary)

Thermal expansion of (U$$_{0.7}$$Pu$$_{0.3}$$)O$$_{2-x}$$ (x = 0, 0.01, 0.02, 0.03) and (U$$_{0.52}$$Pu$$_{0.48}$$)O$$_{2.00}$$ was measured with a dilatometer in an oxygen partial pressure-controlled atmosphere. The oxygen partial pressure was controlled to hold a constant oxygen-to-metal ratio in the (U,Pu)O$$_{2-x}$$ during the measurement. Thermal expansion slightly increased with the decrease in oxygen-to-metal ratio. The relationship was derived to describe thermal expansion.

Journal Articles

Early-in-life fuel restructuring behavior of Am-bearing MOX fuels

Tanaka, Kosuke; Sasaki, Shinji; Katsuyama, Kozo; Koyama, Shinichi

Transactions of the American Nuclear Society, 113(1), p.619 - 621, 2015/10

In order to evaluate the microstructural change behavior of Am-MOX fuels at the initial stage of irradiation, detailed investigations using image analysis were performed on X-ray Computed Tomography (X-ray CT) images and on ceramographs from fuels irradiated in both B11 and B14.

Journal Articles

Extension of effective cross section calculation method for neutron transport calculations in particle-dispersed media

Yamamoto, Toshihiro; Miyoshi, Yoshinori; Takeda, Toshikazu*

Journal of Nuclear Science and Technology, 43(1), p.77 - 87, 2006/01

 Times Cited Count:6 Percentile:43.58(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Investigation on Innovative Water Reactor for Flexible Fuel Cycle (FLWR), 1; Conceptual design

Uchikawa, Sadao; Okubo, Tsutomu; Kugo, Teruhiko; Akie, Hiroshi; Nakano, Yoshihiro; Onuki, Akira; Iwamura, Takamichi

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

no abstracts in English

Journal Articles

Present status of PSA methodology development for MOX fuel fabrication facilities

Tamaki, Hitoshi; Hamaguchi, Yoshikane; Yoshida, Kazuo; Muramatsu, Ken

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

A PSA procedure for MOX fuel fabrication facilities is being developed at the JAERI. This procedure consists of four steps, which are hazard analysis, accident scenario analysis, frequency evaluation and consequence evaluation. The proposed procedure is characterized by the hazard analysis step. The Hazard analysis step consists of two sub-steps. In the first sub-step, a variety of functions of equipment composing the facility are analyzed to identify potential abnormal events exhaustively. In the second sub-step, these potential events are screened to select abnormal events by using a risk matrix based on the rough estimation of likelihood and maximum unmitigated release of radioactive material. One of the unique technical issues in this research is the estimation of likelihood of criticality event. A method is also proposed as a part of PSA procedure taking into consideration of failure of a computerized control system for MOX powder handling process. The applicability of the PSA procedure was demonstrated through the trial application of it to a model plant of MOX fuel fabrication facility.

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

Study on safety evaluation for nuclear fuel cycle facility under fire accident conditions

Tashiro, Shinsuke; Abe, Hitoshi; Morita, Yasuji

JAERI-Conf 2005-007, p.348 - 350, 2005/08

Hot test at Rokkasho Reprocessing plant has been started since last year. In addition, construction of the MOX fuel fabrication facility at Rokkasho site is planning. So, the importance of safety evaluation of the nuclear fuel cycle facility is increasing. Under the fire accident, one of the serious postulated accidents in the nuclear fuel cycle facility, the equipments (glove-box, ventilation system, ventilation filters etc.) for the confinement of the radioactive materials within the facility could be damaged by a large amount of heat and smoke released from the combustion source. Therefore, the fundamental data and models calculating for the amount of heat and smoke released from the combustion source under such accident are important for the safety evaluation of the facility. In JAERI, the study focused on the evaluation of amount of heat and smoke released from the combustion source is planning. In this paper, the outline of experimental apparatus, measurement items and evaluation terms are described.

Journal Articles

Concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR)

Iwamura, Takamichi; Uchikawa, Sadao; Okubo, Tsutomu; Kugo, Teruhiko; Akie, Hiroshi; Nakatsuka, Toru

Proceedings of 13th International Conference on Nuclear Engineering (ICONE-13) (CD-ROM), 8 Pages, 2005/05

In order to ensure sustainable energy supply in the future based on the matured Light Water Reactor (LWR) and coming LWR-Mixed Oxide (MOX) technologies, a concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in Japan Atomic Energy Research Institute (JAERI). The concept consists of two parts in the chronological sequence. The first part realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR without significant gaps in technical point of view. The second part represents the Reduced-Moderation Water Reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The key point is that the two core concepts utilize the compatible and the same size fuel assemblies, and hence, the former concept can proceed to the latter in the same reactor system, based flexibly on the fuel cycle circumstances.

Journal Articles

Development of a criticality evaluation method considering the particulate behavior of nuclear fuel

Sakai, Mikio; Yamamoto, Toshihiro; Murazaki, Minoru; Miyoshi, Yoshinori

Nuclear Technology, 149(2), p.141 - 149, 2005/02

 Times Cited Count:1 Percentile:11.17(Nuclear Science & Technology)

In the conventional criticality evaluation of the nuclear powder system, the effects of particulate behavior have not been considered. In other words, it is difficult to reflect the particle behavior into the conventional criticality evaluation. We have developed a novel criticality evaluation code to resolve this issue. The criticality evaluation code, coupling a Discrete Element Method simulation code with a continuous-energy Monte Carlo transport code, makes it possible to study the effect of the particulate behavior on a criticality evaluation. The criticality evaluation code has been applied to the powder system of the MOX fuel powder agitation process. The criticality evaluations have been performed under mixing the MOX fuel powder in a stirred vessel to investigate the effects of the powder boundary deformation and particulate mixture conditions on the criticality evaluation. The evaluation results revealed that the powder uniformity mixture condition and the boundary deformation could make the neutron effective multiplication factor decrease.

Journal Articles

Hazard analysis approach with functional FMEA in PSA procedure for MOX fuel fabrication facility

Tamaki, Hitoshi; Yoshida, Kazuo; Watanabe, Norio; Muramatsu, Ken

Proceedings of International Topical Meeting on Probabilistic Safety Analysis (PSA '05) (CD-ROM), 11 Pages, 2005/00

A probabilistic safety assessment (PSA) procedure for Mixed Oxide (MOX) fuel fabrication facilities is being developed applicable to nuclear facilities at Japan Atomic Energy Research Institute (JAERI). As part of the PSA procedure, the approach to hazard analysis was established, which consists of two analysis steps: Functional Failure Modes and Effects Analysis (Functional FMEA) and Risk Matrix Analysis. In the Functional FMEA, a variety of functions of equipment composing the facility are analyzed to identify potential abnormal events exhaustively. In the second step, these potential events are screened to select abnormal events as candidate events to be analyzed for frequency and consequence by using two-dimensional matrix based on the rough estimation of likelihood and maximum unmitigated release of radioactive material. The applicability of the hazard analysis approach established was demonstrated through the trial application of the PSA procedure being developed to model plant of MOX fuel fabrication facility.

Journal Articles

Radionuclide release from mixed-oxide fuel under severe accident conditions

Hidaka, Akihide; Kudo, Tamotsu; Fuketa, Toyoshi

Transactions of the American Nuclear Society, 91, p.499 - 500, 2004/12

The radionuclides release from MOX under severe accident conditions was investigated in the VEGA program to prepare the technical bases for safety evaluation including PSA for LWR using MOX. The MOX specimen irradiated at ATR Fugen was heated up to 3123K in He at 0.1MPa. The Cs release started at about 1000K and was enhanced below 2200K compared with that of UO$$_{2}$$. The possible reason is due to the formation of cracks connected to the high burn-up Pu spots. The total fractional releases were evaluated by alpha-ray, gamma-ray and ICP-AES and compared with the ORNL-Booth model. Although the model was prepared based on the tests with UO$$_{2}$$, the predictions are in reasonable agreement with the measurements. The VEGA test showed that the total releases from MOX are almost the same as those from UO$$_{2}$$ under extremely severe accident conditions. This indicates that the consequences of LWR using MOX are mostly equal to those using UO$$_{2}$$. The effect of difference between MOX and UO$$_{2}$$ on the consequences will be systematically investigated using the JAERI's source term code, THALES-2.

JAEA Reports

A Set of ORIGEN2 cross section libraries based on JENDL-3.3 library; ORLIBJ33

Katakura, Junichi; Kataoka, Masaharu*; Suyama, Kenya; Jin, Tomoyuki*; Oki, Shigeo*

JAERI-Data/Code 2004-015, 115 Pages, 2004/11

JAERI-Data-Code-2004-015.pdf:16.52MB

A set of cross section libraries for ORIGEN2 code, ORLIBJ33, has been produced based on the latest Japanese Evaluated Nuclear Data Library JENDL-3.3. The produced libraries are for LWR's which include PWR, BWR and their MOX fuels. The libraries for FBR's are also produced. Using the libraries for LWR, comparisons with old libraries based on JENDL-3.2 were performed. The comparisons with measured PIE data were also carried out. For the libraries for FBR, the comparisons with the calculations using the old libraries were performed and the effects using different libraries were discussed.

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