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Kam, D. H.*; Grabaskas, D.*; Starkus, T.*; Bucknor, M.*; Uchibori, Akihiro
Transactions of the American Nuclear Society, 126(1), p.536 - 539, 2022/06
Removal of gaseous radionuclides from the bubbles released into the sodium pool is an important consideration of fuel pin failure accident in sodium-cooled fast reactors. To support modeling of this phenomenon as a part of development of the SRT (Simplified Radionuclide Transport) code in Argonne National Laboratory, numerical analysis of experiment on Iodine gas transport to sodium pool was performed. A proposed evaluation method can be regarded to be reasonably predicting the measured decontamination factors.
Nagai, Haruyasu
Fission Product Behavior under Severe Accident, p.112 - 116, 2021/05
no abstracts in English
Tanaka, Kosuke; Sato, Isamu*; Onishi, Takashi; Ishikawa, Takashi; Hirosawa, Takashi; Katsuyama, Kozo; Seino, Hiroshi; Ohno, Shuji; Hamada, Hirotsugu; Tokoro, Daishiro*; et al.
Journal of Nuclear Materials, 536, p.152119_1 - 152119_8, 2020/08
Times Cited Count：0 Percentile：0.01(Materials Science, Multidisciplinary)In order to obtain the release rate coefficients from fuels for fast reactors (FRs), heating tests and the subsequent analyses of the fission products (FPs) and actinides that are released were carried out using samples of uranium-plutonium mixed oxide (MOX) fuel pellets irradiated at the experimental fast reactor Joyo. Three heating tests targeting temperatures of 2773, 2973 and 3173 K were conducted using an FP release behavior test apparatus equipped with a high-frequency induction furnace and solid FP sampling systems consisting of a thermal gradient tube (TGT) and filters. Irradiated fuel pellets were placed into a tungsten crucible, then loaded into the induction furnace. The temperature was raised continuously at a heating rate of 10 K/s to the targeted temperature and maintained for 500 s in a flowing argon gas atmosphere. The FPs and actinides released from the MOX fuels and deposited in the TGT and filters were quantified by gamma-ray spectrometry and inductively coupled plasma mass spectrometry (ICP-MS) analysis. Based on the analysis, the release rates of radionuclides from MOX fuels for FR were obtained and compared with literature data for light water reactor (LWR) fuels. The release rate coefficients of FPs obtained in this study were found to be similar to or lower than the literature values for LWR fuels. It was also found that the release rate coefficient data for actinides were within the range of variation of literature values for LWR fuels.
C/steel interaction and oxidation during severe accidentsIshikawa, Jun; Shiotsu, Hiroyuki; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07
Kitamura, Akira; Chikazawa, Takahiro*; Akahori, Kuniaki*; Tachi, Yukio
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 23(1), p.55 - 72, 2016/06
The Japanese geological disposal program has started researching disposal of spent nuclear fuel (SF) in deep geological strata (hereafter "direct disposal of SF") as an alternative management option other reprocessing followed by vitrification and geological disposal of high-level radioactive waste. We conducted literature survey of dissolution rate of SF matrix and constructing materials (e.g. zircaloy cladding and control rods) selected in safety assessment reports for direct disposal of SF in Europe and United States. We also investigated basis of release rate determination and assignment of uncertainties in the safety assessment reports. Furthermore, we summarized major conclusions proposed by some European projects governed by European Commission. It was found that determined release rates are fairly similar to each other due to use of similar literature data in all countries of interest. It was also found that the determined release rates were including conservativeness because it was difficult to assign uncertainties quantitatively. It is expected that these findings are useful as fundamental information for determination of the release rates for the safety assessment of Japanese SF disposal system.
Hidaka, Akihide
Enerugi Rebyu, 35(9), p.20 - 24, 2015/09
Operation of nuclear power plant causes accumulation of radionuclides in fuel rods as a result of nuclear fission of uranium and plutonium. During severe accidents, large amount of radionuclides are released from fuel and transport in the reactor coolant system and/or the containment. When the containment fails or its confinement function is lost, radionuclides could be released into the environment. Meanwhile, radionuclides can be removed by condensation onto wall, natural deposition such as gravitational settling, the engineered safety features (ESF) such as containment spray and so on. After various processes described above, the species, amounts and timing of radionuclide released into the environment is called source terms. The behavior of radionuclide can be described mechanistically by condensation or evaporation of gaseous radionuclide, deposition, growth and removal of aerosol by ESF. Present paper summarizes the radionuclide behavior during severe accidents.
Sato, Isamu; Onishi, Takashi; Tanaka, Kosuke; Iwasaki, Maho; Koyama, Shinichi
Journal of Nuclear Materials, 461, p.22 - 28, 2015/06
We observed one of the simplified processes by conducting primitive experiments. CsI was heated at 1323 K to be vaporized and deposited on sampling parts with a temperature range of 1023 - 423 K and then B
O
was vaporized at 1973 K to be reacted with Cs/I there. After heating tests, each sampling part was soaked into alkali water to dissolve the surface-deposits for ICP-MS analysis. The results showed that CsI deposited at the sampling parts kept above approx. 850 K was striped by BO vapour. This behaviour will be thermodynamically discussed to study the Cs/I/B chemistry in the severe accidents.
Sato, Isamu; Onishi, Takashi; Tanaka, Kosuke; Iwasaki, Maho; Koyama, Shinichi
Journal of Nuclear Materials, 461, p.22 - 28, 2015/06
Times Cited Count：8 Percentile：56.13(Materials Science, Multidisciplinary)In order to evaluate B influence on the release and transport of Cs and I during severe accidents, basic experiments have been performed on the interaction between deposited Cs/I compounds and vapor/aerosol B compounds. CsI and BO were utilized as a Cs/I compound and a B compound, respectively. Deposited CsI on the thermal gradient tube (TGT), which is exposed to temperatures ranging from 423 K to 1023 K was reacted with vapor/aerosol BO, and then observed to determine how it changed Cs/I decomposition profiles. As a result, vapor/aerosol BO stripped a portion of deposited CsI within a temperature range from 830 K to 920 K to make gaseous CsBO and I. In addition, gaseous I was re-deposited at a temperature range from 530 K to 740 K, while CsBO travelled through the sampling tubes and filters without deposition. It is implied that B influences Cs carriers such as CsBO to transport Cs to the colder regions.
Ishikawa, Jun; Ito, Hiroto; Maruyama, Yu
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
Miyahara, Shinya; Seino, Hiroshi; Ohno, Shuji; Konishi, Kensuke
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05
A CONTAIN-LMR code has been developed in JAEA for application to PRA of LMFRs since the original CONTAIN code had been introduced from SNL of U.S. in 1982. The code is a best-estimate, integrated analysis tool for predicting the physical, chemical and radiological conditions inside a containment building of LMFRs following a severe accident with reactor vessel melt-through. The code is also able to predict the source term to the environment in the accident. This code can treat many important phenomena consistently such as sodium fire, radioactive aerosol behavior, hydrogen burn, sodium-concrete reaction and core debris-concrete interaction occurred in the accident with inter-cell heat and mass flow under the multiple cell geometry. This paper describes the chronology of the code development in JAEA briefly as an introduction, and after that, the outline of computational models in the code, the examples of the code validation, and the future plan of the code application are described.
Hidaka, Akihide*; Kudo, Tamotsu; Kida, Mitsuko; Fuketa, Toyoshi
JAERI-Research 2005-001, 67 Pages, 2005/02
JAERI-Research-2005-001.pdf:3.38MB
In the VEGA program to investigate radionuclides release from irradiated fuel during severe accidents, the analyses are being performed with VICTORIA2.0 code for comprehensive understanding of radionuclides release and transport phenomena. The VEGA-1 and -3 tests were analyzed in the present study. The correlation for Cs diffusion coefficient in fuel grain obtained from VEGA-1 was applied to the release analysis of VEGA-3. The calculated release of Cs agreed well with the measurement. The correlation was applied to subsequent Cs transport and deposition analyses. The calculation underpredicted the total mass of Cs deposited onto the test apparatuses because nucleation of aerosol and its growth were underestimated due to the consideration of aerosol nucleation originated only from released FP in VICTORIA2.0. A sensitivity analysis with aerosol seeds for heterogeneous nucleation showed a reasonable agreement with the measured Cs distribution. It turned out that additional aerosol seeds besides the released FP be considered when the VICTORIA2.0 code is applied to the VEGA test analyses.
Hidaka, Akihide
Nihon Genshiryoku Gakkai-Shi, 45(8), p.493 - 496, 2003/08
In order to investigate the radionuclides release from irradiated fuel under severe accident conditions of LWR, VEGA experimental facility that realizes the highest temperature and pressure conditions was designed and constructed at JAERI. The effect of ambient pressure on radionuclides release was uniquely quantified by using this facility. A model that explains the observed pressure effect was also proposed based on the experimental results. For this effort, the atomic energy society of Japan gave us the preeminent monograph award in FY 2002. This paper describes my encounter with the research awarded this time, fascination of this research, hard-worked points, future plans and so on.
Abe, Hitoshi; Watanabe, Koji*; Tashiro, Shinsuke; Takada, Junichi; Uchiyama, Gunzo
JAERI-Research 2001-052, 18 Pages, 2001/11
JAERI-Research-2001-052.pdf:1.83MB
no abstracts in English
Abe, Hitoshi; Tashiro, Shinsuke; Koike, Tadao; Okagawa, Seigo; Uchiyama, Gunzo
Proceedings of the 2001 Topical Meeting on Practical Implementation of Nuclear Criticality Safety (CD-ROM), 8 Pages, 2001/11
no abstracts in English
Hayashida, Retsu*; Hidaka, Akihide; Nakamura, Takehiko; Kudo, Tamotsu; Otomo, Takashi; Uetsuka, Hiroshi
JAERI-Tech 2001-029, 161 Pages, 2001/03
JAERI-Tech-2001-029.pdf:9.33MB
no abstracts in English
Hidaka, Akihide
RIST News, (30), p.2 - 14, 2000/10
no abstracts in English
Kobayashi, Kaoru*; Kaminaga, Masanori; Haga, Katsuhiro; Kinoshita, Hidetaka; Aso, Tomokazu; Kogawa, Hiroyuki; Hino, Ryutaro
JAERI-Tech 2000-050, 43 Pages, 2000/08
JAERI-Tech-2000-050.pdf:2.36MB
no abstracts in English
Yamawaki, Michio*; Yamaguchi, Kenji*; Ono, Futaba*; Huang, J.*; Harada, Yuhei; Hidaka, Akihide; Sugimoto, Jun
JAERI-Tech 2000-015, p.38 - 0, 2000/03
JAERI-Tech-2000-015.pdf:1.31MB
no abstracts in English
Hidaka, Akihide; Nakamura, Takehiko; Kudo, Tamotsu
Genshiryoku eye, 46(3), p.79 - 83, 2000/03
no abstracts in English
Hidaka, Akihide; Kudo, Tamotsu; Nakamura, Takehiko; Hayashida, Retsu*; Otomo, Takashi; Nakamura, Jinichi; Uetsuka, Hiroshi
JAERI-Research 99-066, p.38 - 0, 1999/12
JAERI-Research-99-066.pdf:6.68MB
no abstracts in English
