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JAEA Reports

Extension of fuel performance code package FEMAXI; Development and validation of reactivity-initiated accident analysis module RANNS for light water reactor fuels

Tasaki, Yudai; Udagawa, Yutaka

JAEA-Data/Code 2024-012, 76 Pages, 2024/12

JAEA-Data-Code-2024-012.pdf:9.25MB

Japan Atomic Energy Agency (JAEA) has been developing a fuel performance code, FEMAXI, to evaluate the behavior of LWR fuels under normal operation and transient conditions. In March 2019, FEMAXI-8, the first systematically validated and performance evaluated code, was released. Since then, the code has undergone various improvements. In parallel, since the 2000s, JAEA has been developing the RANNS module as a branch for design basis accident (DBA) analysis, with a particular emphasis on computational stability, so that fuel behavior can be tracked even for very steep transients, in this case mainly reactivity-initiated accidents (RIAs). The specific models include boiling heat transfer, fission gas release by grain boundary failure, and cladding failure determination based on fracture mechanics parameters, which are essential for predicting such transient behavior. In this report, prior to the release of RANNS, we present a description of the models for accident behavior analysis, the relationship with FEMAXI-8 in terms of the design and structure of the program, and the results of a large-scale validation using the extensive database of RIA experiments conducted and accumulated by JAEA, to evaluate the overall RIA analysis performance. The code will be made available to users as a packaged FEMAXI/RANNS, enabling them to analyze fuel behavior under various conditions. The model parameter sets determined through the above validation analyses are also presented in this report, and by referring to them, the analysis can be easily performed with almost no change in usability from the previously released FEMAXI-8.

JAEA Reports

Report on the evaluation of research and development activities in FY2014; Issue: "Research and Development on Reprocessing of Nuclear Fuel Materials" (Ex-post evaluation)

Tokai Reprocessing Technology Development Center

JAEA-Evaluation 2015-012, 83 Pages, 2015/12

JAEA-Evaluation-2015-012.pdf:6.67MB

Japan Atomic Energy Agency (hereafter referred as "JAEA") consulted the "Evaluation Committee of Research and Development Activities for Fast Reactor Cycle" to assess the issue on "Research and Development on Reprocessing of Nuclear Fuel Materials" conducted by JAEA during the period from FY2010 to FY2014. In response to the JAEA's request, the committee assessed the R&D programs and the activities of JAEA related to the issue and concluded the mission was accomplished. This evaluation was performed based on the "General guideline for the evaluation of government R&D activities", the "Guideline for evaluation of R&D in Ministry of Education, Culture, Sports, Science and Technology (MEXT)" and the "Operational rule for evaluation of R&D activities" by JAEA.

Journal Articles

NSRR RIA-simulating experiments on high burnup LWR fuels

Fuketa, Toyoshi; Sugiyama, Tomoyuki; Sasajima, Hideo; Nagase, Fumihisa

Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.633 - 645, 2005/10

LWR fuel behaviors during a reactivity initiated accident (RIA) are being studied in the NSRR program. Results from recent NSRR experiments, no failures in Tests OI-10 and -12 and the higher failure enthalpy in Test OI-11, reflect the better performance of the new cladding materials in terms of corrosion during PWR operations. Accordingly, these rods with improved corrosion resistance have larger safety margin than conventional Zircaloy-4 rods. In addition, the smaller inventory of inter-granular gas in the large grain pellet could reduce the fission gas release in RIA as observed in the OI-10. Test VA-1 was conducted with an MDA sheathed 78 MWd/kgU PWR fuel rod. Despite of the higher burnup and thicker oxide layer of 81m, the enthalpy at failure remained in a same level as those for rods with of 40m-oxide at 50 - 60 MWd/kgU. This result suggests high burnup structure (rim structure) in pellet periphery does not have strong effect on the failure enthalpy reduction because the PCMI load is produced primarily by solid thermal expansion of the pellet.

Journal Articles

LOCA and RIA studies at JAERI

Sugiyama, Tomoyuki; Nagase, Fumihisa; Nakamura, Jinichi; Fuketa, Toyoshi

HPR-362, Vol.2, 12 Pages, 2004/05

To provide a data base for the regulatory guide of light water reactors, behavior of reactor fuels during off-normal and postulated accident conditions such as loss of coolant accident (LOCA) and reactivity-initiated accident (RIA) is being studied at the Japan Atomic Energy Research Institute (JAERI). The LOCA program consists of integral thermal shock tests and other separate tests for oxidation rate and mechanical property of fuel claddings. Prior to the tests on irradiated claddings, the tests have been conducted on non-irradiated claddings to examine separate effects of corrosion and hydrogen absorption during reactor operation. The tests on irradiated claddings have recently been started and results have been obtained. As for an RIA study, a series of experiments with high burnup fuel rods is being performed by using pulse irradiation capability of the NSRR. This paper presents recent results obtained from the LOCA and RIA studies at JAERI.

JAEA Reports

Fuel irradiation research of Japan at Halden reactor; Achievement of cooperative researches between JAERI and several organizations in the period from 2000 to 2002 (Joint research)

Committee of the Halden Joint Research Programme

JAERI-Tech 2004-023, 38 Pages, 2004/03

JAERI-Tech-2004-023.pdf:1.85MB

JAERI has performed cooperative researches with several Japanese organizations utilizing the Halden Boiling Heavy Water Reactor(HBWR) which is located at Halden in Norway. These researches are carried out based on the contracts of the cooperative researches, which are revised every three years, in accordance with the renewal of the participation of JAERI to the OECD Halden Reactor Project. This report summarizes the objectives, contents and the outlines of the achievements of the cooperative researches during the three years from 2000 January to 2002 December. During the period, seven cooperative researches had been carried out. Two of them had been completed and other five researches have been continued to the next three years period. Most of them are irradiation test researches of advanced fuel and cladding in order to prepare the higher burnup utilization and introduction of LWR fuel and MOX fuel in LWRs of Japan.

Journal Articles