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JAEA Reports

Utilization of gamma ray irradiation at the WASTEF Facility

Sano, Naruto; Yamashita, Naoki; Watanabe, Masaya; Tsukada, Manabu*; Hoshino, Kazutoyo*; Hirai, Koki; Ikegami, Yuta*; Tashiro, Shinsuke; Yoshida, Ryoichiro; Hatakeyama, Yuichi; et al.

JAEA-Technology 2023-029, 36 Pages, 2024/03

JAEA-Technology-2023-029.pdf:2.47MB

At the Waste Safety Testing Facility (WASTEF), the gamma ray irradiation device "Gamma Cell 220" was relocated from the 4th research building of the Nuclear Science Research Institute in FY2019, and the use of gamma ray irradiation has begun. Initially, Fuel Cycle Safety Research Group, Fuel Cycle Safety Research Division, Nuclear Safety Research Center, Sector of Nuclear Safety Research and Emergency Preparedness, the owner of this device, conducted the tests as the main user, but since 2022, other users, including those outside the organization, have started using it. The gamma ray irradiation device "Gamma Cell 220" is manufactured by Nordion International Inc. in Canada. Since it was purchased in 1989, the built-in 60Co radiation source has been updated once, and safety research related to nuclear fuel cycles, etc. It is still used for this purpose to this day. This report summarizes the equipment overview of the gamma ray irradiation device "Gamma Cell 220", its permits and licenses at WASTEF, usage status, maintenance and inspection, and future prospects.

JAEA Reports

R&D and maintenance management of the WASTEF Facility (FY2021)

Sano, Naruto; Yamashita, Naoki; Hoshino, Kazutoyo*; Tsukada, Manabu*; Sawauchi, Fumiya*; Otake, Yoshinori; Ichise, Kenichi; Tagami, Susumu

JAEA-Technology 2022-034, 47 Pages, 2023/03

JAEA-Technology-2022-034.pdf:2.81MB

The Waste Safety Testing Facility (WASTEF) was established in 1982 as an experimental facility for long-term storage of solidified high-level radioactive waste generated in the reprocessing of spent light water reactor fuel and the subsequent safety assessment of geological disposal. It is a historic facility that started operation in 1982. This facility consists of 5 concrete cells, 1 lead cell, 6 glove boxes, and 7 hoods, and is a large-scale facility that can use nuclear fuel materials including uranium and plutonium and radioactive isotopes including TRU. In this facility, research and development requested by the research department is carried out in the Hot Material Examination Section. In addition, patrol inspections, self-inspections, etc. are also carried out as maintenance management based on safety regulations. This report summarizes the overview of WASTEF facilities, the results of operation, maintenance and management work in FY2021, and the future outlook.

Journal Articles

Erratum; Effects of OH$$^{-}$$ activity and temperature on the dissolution rate of compacted montmorillonite under highly alkaline conditions [Clay Minerals, vol.51, p.275 (2016), Corrected Fig. 7.]

Sawaguchi, Takuma; Tsukada, Manabu; Yamaguchi, Tetsuji; Mukai, Masayuki

Clay Minerals, 51(5), P. 815, 2016/12

ERRATUM; Effects of OH$$^{-}$$ activity and temperature on the dissolution rate of compacted montmorillonite under highly alkaline conditions [Clay Minerals, vol.51, p.275 (2016), Corrected Fig. 7.]

Journal Articles

Effects of OH$$^{-}$$ activity and temperature on the dissolution rate of compacted montmorillonite under highly alkaline conditions

Sawaguchi, Takuma; Tsukada, Manabu; Yamaguchi, Tetsuji; Mukai, Masayuki

Clay Minerals, 51(2), p.267 - 278, 2016/05

 Times Cited Count:6 Percentile:24.72(Chemistry, Physical)

The dependences of the dissolution rate of compacted montmorillonite on activity of OH$$^{-}$$ (a$$_{rm OH}$$-) and temperature (T) were investigated. The dissolution rate of montmorillonite ($$R_{rm A}$$) in compacted pure montmorillonite, which was formulized as $$R_{rm A}$$ = 10$$^{4.5}$$ (a$$_{rm OH}$$-)$$^{1.3}$$ e$$^{-55000/RT}$$, was higher than that in the compacted sand-bentonite mixtures: $$R_{rm A}$$ = 3500 (a$$_{rm OH}$$-)$$^{1.4}$$ e$$^{-51000/RT}$$. The difference can be explained by considering the decrease in a$$_{rm OH}$$- in the mixtures accompanied by dissolution of accessory minerals such as quartz and chalcedony. The dissolution rate model developed for pure montmorillonite is expected to be applied to bentonite mixtures if quantification of the decrease in a$$_{rm OH}$$- is achieved somehow.

JAEA Reports

Procedures of chemical analysis of radioactive waste for decommissioning of TEPCO'S Fukushima Daiichi Nuclear Power Station

Yonekawa, Minoru; Iwasaki, Maho; Shimada, Kozue; Yanagiya, Shoko; Tsukada, Manabu; Iizuka, Yoshiyuki; Kaneko, Munenori; Unno, Toshimichi

JAEA-Testing 2015-002, 151 Pages, 2016/03

JAEA-Testing-2015-002.pdf:4.29MB
JAEA-Testing-2015-002-appendix(CD-ROM).zip:5.7MB

Preparatory Office for Hot Laboratory Operation Management in Fukushima Research Infrastructural Creation Center has advanced research and development for decommissioning of TEPCO'S Fukushima Daiichi Nuclear Power Station. For this purpose, work procedure manual of chemical analysis for safety evaluation on processing, disposal and management of radioactive waste such as low dose level rubbles and fuel debris has been prepared. The manual will be used for personnel training and animation function of PowerPoint was used as the beginner of the chemical analysis to understand easily. This report describes about nuclides which were established analysis method and completed to make animation of work procedure.

Journal Articles

Mineralogical changes and associated decrease in tritiated water diffusivity after alteration of cement-bentonite interfaces

Yamaguchi, Tetsuji; Sawaguchi, Takuma; Tsukada, Manabu; Hoshino, Seiichi*; Tanaka, Tadao

Clay Minerals, 51(2), p.279 - 287, 2016/02

 Times Cited Count:7 Percentile:24.72(Chemistry, Physical)

Alteration of bentonite-cement interfaces and accompanying changes in diffusivity of tritiated water was experimentally investigated using intact hardened cement specimens. The alteration by carbonate solution was accompanied by mineralogical changes at the interface and a decrease in the diffusivity to 70% of the initial value after 180-day period. Another alteration under silicate system contacting hardened cement and compacted bentonite was accompanied by mineralogical changes at the interface and a decrease in the diffusivity to 71% of the initial value after 600-day period. The changes in the diffusivity were much less than those observed for mixed specimens of granulated hardened cement and bentonite where the diffusivity decreased down to 20% of the initial value over 180 days. The results were extrapolated to 15 years under simple assumptions and showed good agreement with those observed in the cement-argillite interface at Tournemire URL. Such an explanation enhances our confidence in our assessment of alteration of cement-bentonite systems and can be a base for using our data and models in long term assessment of radioactive waste disposal.

Journal Articles

Changes in hydraulic conductivity of sand-bentonite mixtures accompanied by alkaline alteration

Yamaguchi, Tetsuji; Sawaguchi, Takuma; Tsukada, Manabu; Kadowaki, Mitsushi*; Tanaka, Tadao

Clay Minerals, 48(2), p.403 - 410, 2013/05

 Times Cited Count:7 Percentile:23.85(Chemistry, Physical)

Montmorillonite is the main constituent of bentonite clay buffer materials in radioactive waste repositories. Highly alkaline environments induced by cement based materials are likely to alter montmorillonite, and to deteriorate the physical and/or chemical properties of the buffer materials. The deterioration may cause variation in hydraulic conductivity of the buffer and induce major uncertainties in the radionuclide migration analysis. Empirical data on the variation of hydraulic conductivity are, however, scarce mainly because the alteration of compacted buffer materials, sand-bentonite mixture specimen, is extremely slow. In this study, laboratory experiments were performed to observe changes in hydraulic conductivity of sand-bentonite mixtures accompanied with their alkaline alteration using NaOH based solutions at 80-90 $$^{circ}$$C. Three types of experiments proved that the alkaline alteration of bentonite buffer can increase the hydraulic conductivity. The data obtained in this study are useful for verification of the code that will be used for assessing the alteration.

Oral presentation

Dependence of the dissolution rate of compacted montmorillonite on OH$$^{-}$$ activity under highly alkaline conditions

Sawaguchi, Takuma; Kadowaki, Mitsushi*; Mukai, Masayuki; Tsukada, Manabu; Maeda, Toshikatsu; Tanaka, Tadao

no journal, , 

no abstracts in English

Oral presentation

Development on evaluation method for hydraulic conductivity of compacted bentonite under alkaline conditions, 1; Outline of long-term hydraulic conductivity experiments

Tsukada, Manabu; Kadowaki, Mitsushi*; Mukai, Masayuki; Sawaguchi, Takuma; Kataoka, Masaharu; Maeda, Toshikatsu; Tanaka, Tadao

no journal, , 

no abstracts in English

Oral presentation

Development on evaluation method for hydraulic conductivity of compacted bentonite under alkaline conditions, 2; Analysis using coupled mass-transport/chemical-reaction code

Kataoka, Masaharu; Mukai, Masayuki; Sawaguchi, Takuma; Tsukada, Manabu; Kadowaki, Mitsushi*; Maeda, Toshikatsu; Tanaka, Tadao

no journal, , 

no abstracts in English

Oral presentation

Study on hydraulic conductivity of conpacted bentonite considered with alkaline alteration

Kataoka, Masaharu; Tsukada, Manabu; Mukai, Masayuki; Sawaguchi, Takuma; Kadowaki, Mitsushi*; Maeda, Toshikatsu; Tanaka, Tadao

no journal, , 

no abstracts in English

Oral presentation

Development of evaluation methodology on hydraulic performance of bentonite with long-term degradation

Mukai, Masayuki; Sawaguchi, Takuma; Kataoka, Masaharu; Tsukada, Manabu; Maeda, Toshikatsu; Tanaka, Tadao

no journal, , 

In the Japanese HLW disposal program, vitrified HLW is encapsulated in a steel container called overpack, and placed in a repository surrounded with bentonite clay material in an underground site. Bentonite is planed to be used as buffer material to prevent from contacting overpack with huge volume of groundwater for extended period of post-closure. Cementinious material affects on surrounding groundwater to alkalize and that probably deteriorates the low hydraulic conductivity of bentonite. JAEA, therefore, has been developed evaluation methodology to calculate long-term performance of bentonite in alkaline groundwater condition by means of concurrent calculations of chemical-reaction and mass-transport, combining with key models formulated mainly on experimentally basis.

Oral presentation

Study on evaluation for long-term alteration behavior of bentonite buffer materials

Sawaguchi, Takuma; Kadowaki, Mitsushi*; Mukai, Masayuki; Tsukada, Manabu; Kataoka, Masaharu; Maeda, Toshikatsu; Tanaka, Tadao

no journal, , 

no abstracts in English

Oral presentation

Study on hydraulic conductivity of compacted bentonite considered with Long-term degradation

Kataoka, Masaharu; Tsukada, Manabu; Mukai, Masayuki; Sawaguchi, Takuma; Kadowaki, Mitsushi*; Maeda, Toshikatsu; Tanaka, Tadao

no journal, , 

no abstracts in English

Oral presentation

An Experiment on evolution of mineralogy and transport property of compacted bentonite-hardened cement system

Tsukada, Manabu; Hoshino, Seiichi; Yamaguchi, Tetsuji; Sawaguchi, Takuma; Mukai, Masayuki; Maeda, Toshikatsu; Tanaka, Tadao

no journal, , 

no abstracts in English

Oral presentation

Development of an evaluation method for hydraulic conductivity of bentonite buffer materials altered under alkaline conditions

Sawaguchi, Takuma; Tsukada, Manabu; Mukai, Masayuki; Yamaguchi, Tetsuji

no journal, , 

no abstracts in English

Oral presentation

Tensile and ductile-brittle properties of the MEGAPIE samples evaluated by small punch (SP) tests

Saito, Shigeru; Wakui, Takashi; Tsukada, Manabu*; Yamashita, Naoki; Sano, Naruto; Dai, Y.*; Futakawa, Masatoshi

no journal, , 

Post irradiation examination (PIE) of the MEGAPIE (MEGAwatt Pilot Experiment) project has been performed on the samples allotted to JAEA. In the experiments, specimens for tensile test are prepared by taking from the component of the MEGAPIE target such as the beam window (BW) of T91. The irradiation temperatures were about 250 $$^{circ}$$C and the displacement-damage levels ranged between 0.75 and 1.74 dpa. After the tensile tests, SP specimens are prepared from the grip part of the tested tensile specimens. Testing temperatures for SP tests ranged from -150$$^{circ}$$C to 250$$^{circ}$$C. From the area of the load-displacement curves (LDC) of the SP tests, ductile-brittle transition temperature (DBTT) was evaluated. Comparison of these data with previous STIP data showed similar values. In addition, the yield stress and ultimate tensile strength were calculated from the SP tests results by applying the conversion equation. These were compared with the results of the tensile tests and found to be in good agreement.

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