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JAEA Reports

Study on the radioactivity evaluation method of biological shielding concrete of JPDR for near surface disposal

Kochiyama, Mami; Okada, Shota; Sakai, Akihiro

JAEA-Technology 2021-010, 61 Pages, 2021/07

JAEA-Technology-2021-010.pdf:3.56MB
JAEA-Technology-2021-010(errata).pdf:0.75MB

It is necessary to evaluate the radioactivity inventory in wastes in order to dispose of radioactive wastes generated from dismantling nuclear reactor in the shallow ground. In this report, we examined radioactivity evaluation method for near surface disposal about biological shield concrete near the core generated from the dismantling of JPDR. We calculated radioactive concentration of the target biological concrete using the DORT code and the ORIGEN-S code, and we estimated radioactivity concentration Di (Bq/t). For DORT calculation, the cross-section library created from the MATXSLIB-J40 file from JENDL-4.0 was used, and for ORIGEN-S, the attached library of SCALE6.0 was used. As a result of comparing the calculation results of the radioactivity concentration with the past measured values in the radial direction and the vertical direction, we found that the trends were generally the same. We calculated radioactive concentration of the target biological concrete Di (Bq/t), and we compared with the estimated Ci (Bq/t) equivalent to the dose criteria of trench disposal calculated for 140 nuclides. As a result we inferred that the except for about 2% of target waste could be disposed of in the trench disposal facility. We also preselected important nuclides for trench disposal based on the ratios (Di/Ci) for each nuclide, H-3, C-14, Cl-36, Ca-41, Co-60, Sr-90, Eu-152 and Cs-137 were selected as important nuclides.

Journal Articles

Determination of $$^{36}$$Cl in biological shield concrete using pyrohydrolysis and liquid scintillation counting

Ito, Mitsuo; Watanabe, Kazuo; Hatakeyama, Mutsuo; Tachibana, Mitsuo

Analyst, 127(7), p.964 - 966, 2002/06

 Times Cited Count:10 Percentile:32.54(Chemistry, Analytical)

A method for the determination of Cl-36 in biological shield concrete of nuclear reactor was developed. Cl in the concrete sample was extracted quantitatively by pyrohydrolysis at 900 ºC and recovered in Na2CO3 solution for subsequent measurement of Cl-36 by liquid scintillation counting. WO3 was used as an accelerator in the pyrohydrolysis. The Cl extraction procedure was optimized by investigating experimental conditions with the use of ion chromatography and its recovery was evaluated by the analysis of the geochemical reference samples. Detection limit of Cl-36 was 0.02 Bq g-1 for sample weight of 2g. Relative standard deviation was 3 – 7 % for the samples containing 0.5 Bq g-1 levels of 36Cl. The newly developed method was applied to determine Cl-36 in biological shield concrete of the Japan Power Demonstration Reactor.

JAEA Reports

Study on bulk shielding for a spallation neutron source facility in the high-intensity proton accelerator project

Maekawa, Fujio; Teshigawara, Makoto; Takada, Hiroshi; Furusaka, Michihiro*; Watanabe, Noboru

JAERI-Tech 2002-035, 68 Pages, 2002/03

JAERI-Tech-2002-035.pdf:4.76MB

no abstracts in English

Journal Articles

Determination of $$^{41}$$Ca in biological-shield concrete by low-energy X-ray spectrometry

Ito, Mitsuo; Watanabe, Kazuo; Hatakeyama, Mutsuo; Tachibana, Mitsuo

Analytical and Bioanalytical Chemistry, 372(5-6), p.532 - 536, 2002/03

 Times Cited Count:15 Percentile:44.09(Biochemical Research Methods)

An X-ray spectrometry method has been developed for the determination of Ca-41 in the biological shield concrete of nuclear reactors. To isolate Ca from other elements, the concrete sample was first decomposed with nitric, hydrofluoric and perchloric acids. Calcium was then separated from other coexisting radionuclides by ion-exchange chromatography and recovered as an oxalate precipitate. X rays at 3.3 keV from Ca-41 in the calcium oxalate pellet were measured. Detection efficiency of the X-ray measurement at 3.3 keV was calculated from those obtained by measuring Fe-55 standard pellets at 5.9 keV using mass absorption coefficients of the calcium oxalate pellet at each X-ray energy value. A lower limit of determination of 8 Bq g-1 was obtained for a sample weight of 1 g.

JAEA Reports

Study on residual radioactive inventory estimation in reactor decommissioning program (Contract research)

Sukegawa, Takenori; Hatakeyama, Mutsuo; Yanagihara, Satoshi

JAERI-Tech 2001-058, 81 Pages, 2001/09

JAERI-Tech-2001-058.pdf:5.98MB

In general, neutron transport and activation calculation codes are used for residual radioactive inventory estimation; however, it is essential to verify calculations by measurement results because of geometrical complexity of the reactor and so on. The comparison between measured and calculated radioactivity in the JPDR core components showed a relatively good agreement (factor of 2), and it was cleared that water content and weight ratio of steel bars to concrete materials significantly influenced the neutron flux distribution in the biological shield (factor of 2-10 error). The measured radioactivity inside of the reactor pressure vessel wall and at the inner part of the biological shield was compared in detail with the calculations to verify the methodology applied to calculations of radioisotope production. Then it was found that the radioactive inventory could be estimated accurately with combination of calculations and measurement of radioactivity in samples and dose rate distribution for planning of dismantling activities.

Journal Articles

Cutting technique and system for biological shield

Nakamura, Hisashi; *; Yanagihara, Satoshi

Nuclear Technology, 86, p.168 - 178, 1989/08

 Times Cited Count:5 Percentile:55.75(Nuclear Science & Technology)

no abstracts in English

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