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Journal Articles

Development methodology on determination of instant release fractions for generic safety assessment for direct disposal of spent nuclear fuel

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.

Journal Articles

Effect of carbonate concentration on the dissolution rates of UO$$_{2}$$ and spent fuel; A Review

Kitamura, Akira; Akahori, Kuniaki*

Advances in Materials Science for Environmental and Energy Technologies, 6, p.133 - 144, 2017/10

The Japanese geological disposal program has started researching disposal of spent nuclear fuel (SF) in deep geological strata as an alternative management option other reprocessing followed by vitrification and geological disposal of high-level radioactive waste. One of the key parameters for SF disposal other than the disposal of high-level radioactive waste is the fuel dissolution rate. Carbonate concentration in the simulated water composition with contact to SF after canister breaching in the Japanese SF disposal system is around 10$$^{-2}$$ mol dm$$^{-3}$$, which is one order of magnitude larger than those in some countries in Europe. The SF dissolution rate will be depend on carbonate concentration due to promoting oxidative dissolution of SF by formation of carbonate complexes of uranium(VI). For evaluation of reliable SF dissolution rate in the Japanese SF disposal system as an alternative management option, effect of carbonate concentration on dissolution rate of UO$$_{2}$$ and spent fuel has been reviewed.

Journal Articles

Evaluation of source term parameters for spent fuel disposal in foreign countries, 2; Dissolution rates of spent fuel matrices and construction materials for fuel assemblies

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.

Journal Articles

Evaluation of source term parameters for spent fuel disposal in foreign countries, 1; Instant release fraction from spent fuel matrices and composition materials for fuel assemblies

Nagata, Masanobu; Chikazawa, Takahiro*; Akahori, Kuniaki*; Kitamura, Akira; Tachi, Yukio

Genshiryoku Bakkuendo Kenkyu (CD-ROM), 23(1), p.31 - 54, 2016/06

Although spent nuclear fuel is planned to be disposed after reprocessing and vitrification of high-level radioactive waste (HLW), feasibility study on direct disposal of spent nuclear fuel (SF) has been started as an alternative option to flexibly apply change of future energy situation in Japan. Radionuclide inventories and their release behavior after breaching spent fuel container should be assessed to confirm safety of the SF disposal. However, these detailed studies have not been performed in Japan. Therefore, we investigated some foreign safety assessment reports on direct disposal of spent nuclear fuel by focusing on the source term of the fast release of radionuclides (i.e. instant release fraction; IRF) for the purpose of contributing to the safety assessment of Japanese SF disposal system. As a result of comparison between the safety assessment reports in foreign countries, although some fundamental data have been referred to the reports in common, the final source term dataset (IRF) was seen differences between countries in the result of taking into account the national circumstances (Reactor type and burnup, etc.). We also found the difference of assignment of uncertainties among the investigated reports; a report selected pessimistic values and another report selected mean values and their deviations. 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.

JAEA Reports

The Conceptual Design of Waste Repository for Radioactive Waste from Medical, Industrial and Research Facilities Containing Comparatively High Radioactivity III (Summary Report)

Kageyama, Hitoshi*; Akahori, Kuniaki*

JNC TJ8400 2003-085, 82 Pages, 2004/02

JNC-TJ8400-2003-085.pdf:0.72MB

Advisory Committee on Nuclear Fuel Cycle Backend Policy reported the basic approach to the RI and Institute etc. wastes on March 2002. According to it, radioactive waste form medical, industrial and research facilities should be classified by their radioactivity properties and physical and chemical properties, and should be disposed in the appropriate types of repository with that classification. For the radioactive waste containing comparatively high radioactivity generated from reactors, NSC has established the Concentration limit for disposal. NSC is now discussing about the limit for the radioactive waste from medical, industrial and research facilities containing comparatively high radioactivity. Japan Nuclear Cycle Development Institute (JNC) has studied about the feasibility and the cost of the disposal for radioactive waste from medical, industrial and research facilities. This study was started to renew to latest data of the radioactive waste. And at the point of shielding from radiation, the waste was categorized by activity of nuclide in waste container. Th

JAEA Reports

The Conceptual Design of Waste Repository for Radioactive Waste from Medical, Industrial and Research Facilities Containing Comparatively High Radioactivity III

Kageyama, Hitoshi*; Akahori, Kuniaki*

JNC TJ8400 2003-084, 206 Pages, 2004/02

JNC-TJ8400-2003-084.pdf:3.38MB

Advisory Committee on Nuclear Fuel Cycle Backend Policy reported the basic approach to the RI and Institute etc. wastes on March 2002. According to it, radioactive waste form medical, industrial and research facilities should be classified by their radioactivity properties and physical and chemical properties, and should be disposed in the appropriate types of repository with that classification. For the radioactive waste containing comparatively high radioactivity generated from reactors, NSC has established the Concentration limit for disposal. NSC is now discussing about the limit for the radioactive waste from medical, industrial and research facilities containing comparatively high radioactivity. Japan Nuclear Cycle Development Institute (JNC) has studied about the feasibility and the cost of the disposal for radioactive waste from medical, industrial and research facilities. This study was started to renew to latest data of the radioactive waste. And at the point of shielding from radiation, the waste was categorized by activity of nuclide in waste container. Then the safety assessment and the prediction of cost of the disposal performed.

JAEA Reports

None

Akahori, Kuniaki; Uchida, Masahiro

PNC TN8410 97-004, 58 Pages, 1997/03

PNC-TN8410-97-004.pdf:5.59MB

None

Oral presentation

Analysis of data perturbation found in the sorption database; Explanation by adopting mechanistic sorption models

Oe, Toshiaki*; Nagasaki, Shinya*; Kimura, Hideo; Takeda, Seiji; Sekioka, Yasushi; Kato, Hiroyasu*; Akahori, Kuniaki*

no journal, , 

no abstracts in English

Oral presentation

Estimation of source term for spent fuel disposal, 2; Effect of carbonate concentration on dissolution rate of UO$$_{2}$$ and spent fuel; A Review

Kitamura, Akira; Akahori, Kuniaki*

no journal, , 

Since dissolution rate of UO$$_{2}$$ matrices will be depend on carbonate concentration due to promoting oxidative dissolution of spent nuclear fuel by formation of carbonate complexes of uranium(VI), effect of carbonate concentration on dissolution rate of UO$$_{2}$$ and spent nuclear fuel has been reviewed. It is found that a systematic study on dissolution rate of UO$$_{2}$$ and/or spent fuel as a function of carbonate concentration is recommended.

Oral presentation

Estimation of source term for spent fuel disposal, 1; A Review of instant release parameters in foreign countries, and a study of provisional parameters for domestic spent fuel

Nagata, Masanobu; Kitamura, Akira; Tachi, Yukio; Akahori, Kuniaki*; Chikazawa, Takahiro*

no journal, , 

no abstracts in English

Oral presentation

Estimation of source term for spent fuel disposal, 3; Development of methodology on estimation of fission gas release ratios for Japanese spent fuels

Nagata, Masanobu; Akahori, Kuniaki*; Kitamura, Akira; Tachi, Yukio; Chikazawa, Takahiro*

no journal, , 

no abstracts in English

11 (Records 1-11 displayed on this page)
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