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Journal Articles

A Modelling study on water radiolysis for primary coolant in PWR

Mukai, Satoru*; Umehara, Ryuji*; Hanawa, Satoshi; Kasahara, Shigeki; Nishiyama, Yutaka

Proceedings of 20th International Conference on Water Chemistry of Nuclear Reactor Systems (NPC 2016) (USB Flash Drive), 9 Pages, 2016/10

In Japanese PWR, the concentration of dissolved hydrogen in the primary coolant is controlled in the range from 25 cc/kg-H $_{2}$ O to 35 cc/kg-H $_{2}$ O for suppression of water decomposition. However this concentration is desired to reduce for the purpose of radiation source reduction in Japan. So, the concentration due to water radiolysis in primary coolant was evaluated at lower hydrogen concentration by the water radiolysis model in consideration of ray, fast neutron and alpha ray due to the reaction $^{10}$ B(n,) $^{7}$ Li. The results of evaluation showed that the water radiolysis was suppressed even if the hydrogen concentration was decreased to 5 cc/kg-H $_{2}$ O. The effects of the different G-value and the rate constants of major reaction on the concentration of H $_{2}$ O $_{2}$ and O $_{2}$ were studied under hydrogen addition. We also focused on the effect of the alpha radiolysis in boron acid water.

JAEA Reports

Study on the evaluation method to determine the radioactivity concentration waste generated from post-irradiation examination facilities

Hoshino, Yuzuru; Sakamoto, Yoshiaki; Muroi, Masayuki*; Mukai, Satoru*

JAEA-Technology 2015-015, 96 Pages, 2015/07

JAEA-Technology-2015-015.pdf:20.34MB

In order to dispose of the radioactive waste which generates from post-irradiation examination (PIE) facilities, the common evaluation method of radioactivity in wastes from PIE should be established by the actual data such as radioactivity values and the theoretical calculation. In this study, the radioactivity concentrations of 17 nuclides (H-3, C-14, Co-60, Ni-63, Sr-90, Tc-99, Cs-137, Eu-154, U-234, U-235, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Am-241, Cm-244) in combustible wastes stored in NUCLEAR DEVELOPMENT CORPORATION were measured from 3 samples and the radioactivity was calculated by ORIGEN-2 based on initial contents and operation record of the spent fuel. From the comparison of the obtained data by the radiological measurement with the calculated values, the subject to be solved for establishment of the radioactivity evaluation method for PIE was extracted.

JAEA Reports

Study on diffusion depth of matrix in granitic rock

Tochigi, Yoshikatsu; Kogawa, Noritaka*; Mukai, Satoru*; Jintoku, Takashi*; Sasamoto, Hiroshi; Shibata, Masahiro; Yui, Mikazu

JAEA-Research 2007-024, 27 Pages, 2007/03

JAEA-Research-2007-024.pdf:1.83MB

In the safety assessment of high-level radioactive waste (HLW) geological disposal, it is important to understand diffusion behavior of radionuclides for evaluation of retardation capacity in the host rock. In this study, non steady state diffusion experiments with non-sorbing ion ware carried out using several granitic rock samples of different length and the matrix diffusion depth was evaluated. As the results of experiment, it is confirmed that the non-sorbing ion can diffuse into the unaltered zone of granitic rock at least about 200 mm in depth and almost same diffusion coefficient is estimated for all rock samples independent of length. Based on these results, it is expected that matrix diffusion of nuclides is acting on a scale of several dozens of centimeters in the granitic rock.

JAEA Reports

Study on the nuclide behavior in the engineered barrier for TRU waste (V)

Yanagisawa, Ichiro*; Katsurai, Kiyomichi*; Mukai, Satoru*; *

JNC TJ8400 2000-028, 224 Pages, 2000/02

JNC-TJ8400-2000-028.pdf:4.06MB

In this study, in-diffusion experiment was conducted under reducing condition to measure apparent diffusion coefficients and distribution coefficients of Uranium and Neptunium for bentonite materials as engineered barrier. The summary of this study is as follows: (1)Apparent diffusion coefficients of U and Np under reducing condition were one or two orders magnitude lower than those of the results obtained in the previous studies. From this result, the valence of U and Np was estimated to be +4 charge during diffusion experiment. (2)Distribution coefficients of U and Np for the system of equilibriated water with bentonite at density of 1.6 g/cm $^{3}$ were a little lower than those for the other synthetic groundwater system. (3)Distribution coefficients of Cs and Sr decreased as ion strength of pore water in bentonite increased. Distribution coefficient of Sr increased for alternation of bentonite. (4)Distribution coefficients of Ni, Sn, Zr and Nb were little difference for any type of bentonite or synthetic groundwater. (5)Distribution coefficients or Am for Na-type of bentonite was higher than that of Ca-type of bentonite. But the difference of distribution coefficients for any synthetic groundwater was small. (6)Data base of apparent diffusion coefficients and distribution coefficients was renewed.

JAEA Reports

Study on the nuclide behavior in the engineered barrier for TRU waste(V)

Yanagisawa, Ichiro*; Katsurai, Kiyomichi*; Mukai, Satoru*; *

JNC TJ8400 2000-027, 28 Pages, 2000/02

JNC-TJ8400-2000-027.pdf:0.76MB

no abstracts in English

JAEA Reports

Study on the nuclide behavior in the engineered barrier for TRU wasete, IV

Yanagisawa, Ichiro*; Naito, Taisei*; Kitao, Hideo*; Mukai, Satoru*; Doi, Hideo*

JNC TJ8400 99-020, 25 Pages, 1999/02

JNC-TJ8400-99-020.pdf:0.64MB

In this study, apparent diffusion coefficients and distribution coefficient of some nuclides were measured for bentonite materials as engineered barrier. The summary of this study is as follows: (1)Apparent diffusion coefficients of nuclides in compacted bentonite were measured. The data so far were summarized as database of apparent diffusion coefficients and distribution coefficients. (2)In diffusion tests for mixture of Ca-type altered bentonite and analcime, apparent diffusion coefficient of Sr was a little smaller than that for Ca-type altered bentonite. (3)In diffusion tests for Ca-type altered bentonite, apparent diffusion coefficients and distribution coefficients of nuclides showed no significant difference between bentonite-water system and bentonite-cementitious solution systems. (4)In Na-type bentonite-solution with nitrate system, apparent diffusion coefficients of Cs and Sr were smaller than that for other bentonite-solution systems. (5)Apparent diffusion coefficients of tritium showed no significant difference between Ca-type altered bentonite and Na-type bentonite in any solution systems.

Journal Articles

Sato, Kazuyoshi; Neyatani, Yuzuru; Maruo, Takeshi; Mukai, Satoru*; Uchida, Shoji*; Soman, Yoshindo*

no journal, ,

no abstracts in English

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A Modelling study on water radiolysis for primary coolant in PWR

Study on the evaluation method to determine the radioactivity concentration waste generated from post-irradiation examination facilities

Study on diffusion depth of matrix in granitic rock

Study on the nuclide behavior in the engineered barrier for TRU waste (V)

Study on the nuclide behavior in the engineered barrier for TRU waste(V)

Study on the nuclide behavior in the engineered barrier for TRU wasete, IV

Diffusion of some radionuclides in compacted Ca-bentonite under reducing condition

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None

None

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Integration study (IV) for performance assessment system of geological disposal

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Chemistry and radiochemistry tests of N.S.MUTSU; Reports on N.S.MUTSU power-up tests

Study of the radiolysis for ITER cooling water at 14MeV neutron