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Kitamura, Akira; Akahori, Kuniaki; Nagata, Masanobu*
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 27(2), p.83 - 93, 2020/12
Direct disposal of spent nuclear fuel (SNF) in deep underground repositories (hereafter "direct disposal") is a concept that disposal canisters stored fuel assemblies dispose without reprocessing. Behavior of radionuclide release from SNF must be different from that from vitrified glass. The present study established a methodology on determination of instant release fraction (IRF) of radionuclides from SNF, which is the one of the parameters on radionuclide release based on the latest safety assessment reports in other countries, especially for IRF values proportional to a fission gas release ratio (FGR). Recommended and maximum values of FGR have been estimated using the fuel performance code FEMAXI-7 after collecting FGR values on Japanese SNFs. Furthermore, recommended and maximum values of IRF for Japanese SNFs used in a pressurized water reactor (PWR) have been estimated using the presently obtained FGR values and experimentally obtained IRF values on foreign SNFs. The recommended and maximum IRF values obtained in the present study have been compared with those of the latest safety assessment reports in other countries.
Nishihara, Kenji; Yamagishi, Isao; Yasuda, Kenichiro; Ishimori, Kenichiro; Tanaka, Kiwamu; Kuno, Takehiko; Inada, Satoshi; Goto, Yuichi
Journal of Nuclear Science and Technology, 52(3), p.301 - 307, 2015/03
Times Cited Count：18 Percentile：79.54(Nuclear Science & Technology)After the severe accident at the Fukushima-1 nuclear power plant, large amounts of contaminated stagnant water have accumulated in turbine buildings and their surroundings. This rapid communication reports calculation of the radionuclide inventory in the core, collection of measured inventory in the stagnant water, and estimation of radionuclide release ratios from the core to the stagnant water. This evaluation is based on data obtained before June 3, 2011. The release ratios of tritium, iodine, and cesium were several tens of percent, whereas those of strontium and barium were smaller by one or two orders of magnitude. The release ratios in the Fukushima accident were equivalent to those in the TMI-2 accident.
Kudo, Tamotsu; Hidaka, Akihide*; Fuketa, Toyoshi
Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.883 - 889, 2005/10
The VEGA program have been performed at Japan Atomic Energy Research Institute (JAERI). The program was comprised of series of experiments on radionuclides release from fuel under severe accident conditions and post-test evaluation with numerical calculations. Effects on the release of ambient pressure, fuel temperature, inert or steam environment and MOX-effect were studied in the program. These effects had been hardly investigated in previous studies due to difficulties in experiments with high temperature and pressure conditions. Release of cesium was mitigated at elevated pressure in comparison with atmospheric pressure. Cesium release was enhanced in the temperature region where fuel foaming occurred below the melting point of UO
. Release of cesium and ruthenium under steam condition was greater than that under the inert helium condition. Released mass of plutonium above 2800 K was higher by nearly three orders of magnitude than that in lower temperature than 2800 K.
Hidaka, Akihide; Kudo, Tamotsu; Ishikawa, Jun; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 42(5), p.451 - 461, 2005/05
Times Cited Count：6 Percentile：39.62(Nuclear Science & Technology)The radionuclide release from MOX under severe accident conditions was investigated in VEGA program to contribute to the technical bases for safety evaluation including PSA for LWR using MOX. The MOX specimens irradiated at ATR Fugen were heated up to 3123K in helium at 0.1 and 1.0MPa. The release of volatile FP was slightly enhanced below 2200K compared with that of UO. The volatile FP release at elevated pressure was decreased as in the case with UO. The total fractional release of Cs reached almost 100% while almost no release of low-volatile FP even after the fuel melting. The release rate of plutonium above 2800K increased rapidly although the amount was small. Since the existing models cannot predict this increase, an empirical model was prepared based on the data. There is no large difference in FP inventories between UO and MOX, and the fractional releases from MOX can be mostly predicted by the model for UO. This suggests that the consequences of LWR using MOX are mostly equal to those using UO from a view point of risks.
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. 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, 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 under extremely severe accident conditions. This indicates that the consequences of LWR using MOX are mostly equal to those using UO. The effect of difference between MOX and UO on the consequences will be systematically investigated using the JAERI's source term code, THALES-2.
Fuel Safety Research Laboratory
JAERI-Review 2004-021, 226 Pages, 2004/10
JAERI-Review-2004-021.pdf:14.4MB
Fuel Safety Research Meeting 2004, which was organized by Japan Atomic Energy Research Institute, was held on March 1-2, 2004 at Toranomon Pastoral, Tokyo. Purposes of the meeting are to present and discuss results of experiments and analyses on reactor fuel safety and to exchange views and experiences among the participants. Technical topics of the meeting covered status of fuel safety research activities, fuel behavior under RIA and LOCA conditions, high burnup fuel behavior, and radionuclides release under severe accident conditions. This proceeding contains all the papers presented in the meeting.
-ray measurements of VEGA-5 testHidaka, Akihide; Kudo, Tamotsu; Nakamura, Takehiko; Kanazawa, Toru; Kiuchi, Toshio; Uetsuka, Hiroshi
JAERI-Tech 2003-009, 30 Pages, 2003/03
JAERI-Tech-2003-009.pdf:1.73MB
The VEGA (Verification Experiments of radionuclides Gas/Aerosol release) program is being performed at JAERI to clarify mechanisms of radionuclides release from irradiated fuel during severe accidents and to improve source term predictability. The fifth VEGA-5 test was conducted in January 2002 to confirm the reproducibility of decrease in cesium release under elevated pressure that was observed in the VEGA-2 test and to investigate the release behavior of short-life radionuclides. The PWR fuel of 47GWd/tU after 8.2 years of cooling was re-irradiated at Nuclear Safety Research Reactor (NSRR) for 8 hours before the heat-up test. After that, the two pellets of 10.9g without cladding were heated up to about 2,900K at 1.0MPa under the inert He condition. The experiment reconfirmed the decrease in cesium release under elevated pressure. The release data on short-life radionuclides such as Ru-103 and Ba-140 that has never been observed in the previous VEGA tests without re-irradiation was obtained using the gamma ray measurement.
Kudo, Tamotsu; Hidaka, Akihide; Nakamura, Takehiko; Uetsuka, Hiroshi
Journal of Nuclear Science and Technology, 38(10), p.910 - 911, 2001/10
Times Cited Count：11 Percentile：61.23(Nuclear Science & Technology)This article describes the effects of system pressure on the cesium release obtained in the first two tests of radionuclides release from irradiated fuel, VEGA-1 and VEGA-2, which were conducted at the same maximum temperature of 2773K and different system pressures. The fractional releases of Cs in VEGA-2 test at 1.0MPa were smaller than those in VEGA-1 test at atmospheric pressure. In order to quantify the difference of the release rate in the two tests due to the pressure, the release rate coefficients of Cs were evaluated. The Cs release rate coefficient in VEGA-2 was smaller by a factor of about 2.8 than that in VEGA-1 at temperature above about 1900K.
; ; T.H.Pigford*
JAERI-M 9183, 11 Pages, 1980/11
no abstracts in English
