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CO
on C transfer to plants impacted by below-ground CH
releaseOta, Masakazu; Tanaka, Taku*
Journal of Environmental Radioactivity, 201, p.5 - 18, 2019/05
Percentile：100(Environmental Sciences)CH released from deep underground radioactive waste disposal facilities can be a belowground source of CO owing to microbial oxidation of CH to CO in soils. Environmental C models assume that the transfer of CO from soil to plant occurs via foliar uptake of CO. Nevertheless, the importance of CO root uptake is not well understood. In the present study, belowground transport and oxidation of CH were modeled and incorporated into an existing land surface CO model (SOLVEG-II) to assess the importance of root uptake on CO transfer to plants. Performance of the model in calculating the belowground dynamics of CH was validated by simulating a field experiment of 
CH injection into subsoil. The model was then applied to C transfer in a hypothetical ecosystem impacted by continuous CH input from the water table (bottom of one-meter thick soil). In a shallowly rooted ecosystem with rooting depth of 11 cm, foliar uptake of CO was significant, accounting for 80% of the C accumulation in the leaves. In a deeply rooted ecosystem (rooting depth of 97 cm), where the root penetrated to depths close to the water-table, more than half (63%) the C accumulated in the leaves was transferred by the root uptake. We found that CO root uptake in this ecosystem depended on the distribution of methane oxidation in the soil; all C accumulated in the leaves was transferred by the root uptake when methane oxidation occurred at considerable depths (e-folding depths of 20 cm, or 80 cm). These results indicate that CO root uptake contributes significantly to CO transfer to plants if CH oxidation occurs at great depths and roots penetrate deeply into the soil.
Amamoto, Ippei
Journal of the Society of Inorganic Materials, Japan, 24(391), p.393 - 401, 2017/11
Glass is a non-crystalline solid, as such, it is relatively easy to change its composition to control its characteristics. The borosilicate glass, which is produced by the addition of boron oxide into sodium-lime glass, possesses excellent heat-resistant properties and mechanical strength. It has a wide variety of uses. The borosilicate glass is applied as the vitrified medium for radioactive wastes to immobilize and stabilize them for long term. The glass form which is loaded with high-level radioactive waste is called the vitrified waste. This paper classified the radioactive waste and describes treatment and production methods of vitrified waste, its characteristics, disposal method and also introduces alternative vitrified medium.
Tsuji, Tomoyuki; Hoshino, Yuzuru; Sakai, Akihiro; Sakamoto, Yoshiaki; Suzuki, Yasuo*; Machida, Hiroshi*
JAEA-Technology 2017-010, 75 Pages, 2017/06
JAEA-Technology-2017-010.pdf:2.31MB
It is necessary for reasonable disposal to be studied on evaluation methods to determine radioactivity concentrations in the radioactive wastes, which is generated from post-irradiation examination (PIE) facilities, for establishment of reasonable confirmation methods concerning radioactive wastes generated from research, industrial, and medical facilities. It has been chosen the PIE facilities of NUCLEAR DEVELOPMENT CORPORATION as a model for this study. As a result, it has been confirmed that the theoretical methods are applied for the important 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 and Cm-244).
Okada, Shota; Izumo, Sari; Nakata, Hisakazu; Tsuji, Tomoyuki; Sakai, Akihiro; Amazawa, Hiroya
JAEA-Technology 2016-023, 129 Pages, 2016/11
JAEA-Technology-2016-023.pdf:8.95MB
Waste packages must meet the technical requirements. This is because JAEA has been preparing an operating procedure manual for quality control of radioactive waste disposal to be applied to the processing of the waste packages. Raw wastes generated by JAEA are segregated and stored by a method specified in the manual. The composition of raw wastes was characterized on the basis of records of the segregation process. Simulated waste packages were produced by placing the waste materials in a 200 liter drum, which was then filled with mortar, followed by curing in a controlled manner. The static load test was conducted to measure deformation and strain performance of the simulated waste package. Compression apparatuses which can imitate loading conditions in pit-type and trench-type facility that are planned by JAEA were used. Based on the test result, waste packages produced in accordance with the manual met the technical requirement under the condition.
Kitamura, Akira; Kirishima, Akira*
Journal of Nuclear Science and Technology, 52(3), p.448 - 450, 2015/03
Times Cited Count：1 Percentile：8.26(Nuclear Science & Technology)The Journal of Nuclear Science and Technology covers a variety of subjects in the field of nuclear waste management, which includes radioactive waste treatment, radioactive waste disposal and environment, decommissioning and dismantling. This summary introduces activities presented in recent years.
Tanaka, Tadao; Sakamoto, Yoshifumi; Yamaguchi, Tetsuji; Takazawa, Mayumi; Akai, Masanobu; Negishi, Kumi; Iida, Yoshihisa; Nakayama, Shinichi
JAERI-Conf 2005-007, p.105 - 110, 2005/08
Highly alkaline environments induced by cementitious materials in radioactive waste repositories are likely to dissolve and to alter montmorillonite, the main constituent of bentonite buffer materials. For the prediction of the long-term variations in permeability of compacted sand-bentonite mixtures, long-term alteration of bentonite should be quantified based on information accumulated by using the compacted or powdered bentonite materials, with batch experiments or column experiments. In this study, we summarize distinctive information obtained from various experimental systems, and propose functional and effective integration of experimental approaches to prediction of bentonite alteration.
Akai, Masanobu; Ito, Nobuyuki*; Yamaguchi, Tetsuji; Tanaka, Tadao; Iida, Yoshihisa; Nakayama, Shinichi; Inagaki, Shingo*
JAERI-Tech 2004-058, 47 Pages, 2004/09
JAERI-Tech-2004-058.pdf:7.27MB
Geochemistry Research Equipment for TRU Waste Elements has been installed in Back-end Cycle Key Elements Research Facility (BECKY) of Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF). This equipment is designed to study geochemical behavior of TRU elements and other radionuclides contained in TRU waste (TRU waste elements) and to acquire data for safety assessments of radioactive wastes disposal. The equipment consists of anaerobic glove box systems, aerobic glove box systems equipped with built-in barrier performance testing apparatus, and analytical instruments. This report describes principles, structure, performance and safety designs of each component of the equipment, and results of research performed in the equipment.
7th NUCEF Seminar Working Group
JAERI-Conf 2004-011, 166 Pages, 2004/07
JAERI-Conf-2004-011.pdf:30.39MB
no abstracts in English
6th NUCEF Seminar Working Group
JAERI-Conf 2003-018, 132 Pages, 2003/10
JAERI-Conf-2003-018.pdf:10.41MB
no abstracts in English
Onuki, Toshihiko
Genshiryoku Bakkuendo Kenkyu, 9(1), p.35 - 42, 2002/09
Various microorganisms have been observed in deep geologic formation. The effects of such microorganisms on the performance of HLW disposal are still unknown. This paper reviews the studies of microbial effects on the long-term contaiment of HLW disposal, and discusses the future work to be carried out. Microbial reduction and oxidation and byproducts derived from microbial activities affect performance of HLW repository and have a potential to enhance actinides migration in geologic formation (degradation of the materials of repository, complex-formation, dissolution of actinides precipitates and occurrence of nm scale colloid formation). Potential microbial perturbation of performance of the barriers may enhance confinement of actinides by biomineralization, bioadsorption, bioaccumulation and precipitation. These studies indicate that further experiments are required to elucidate microbial effects on the performance of HLW disposal.
; Kawakami, Susumu; *
JNC-TN8400 2001-025, 27 Pages, 2002/03
JNC-TN8400-2001-025.pdf:0.82MB
It is essential to understand the corrosion type of carbon steel under the repository conditions for the lifetime assessment of carbon steel overpack used for geological isolation of high-level radioactive waste. According to the previous study, carbon steel is hard to passivate in buffer material assuming a chemical condition range of groundwater in Japan. However, concrete support will be constructed around the overpack in the case of repository in the soft rock system and groundwater having a higher pH may infiltrate to buffer material. There is a possibility that the corrosion type of carbon steel will be influenced by the rise of the pH in groundwater. In this study, anodic polarization experiments were performed to understand the passivation condition of carbon steel in buffer material saturated with water contacted with concrete. An ordinary concrete and a low-alkalinity concrete were used in the experiment. The results of the experiments showed that the carbon steel can passivate under the condition that water having pH 
13 infiltrate to the buffer material assuming present property of buffer material. If the low-alkalinity concrete is selected as the support material, passivation can not occur on carbon steel overpack. The effect of the factors of buffer material such as dry density and mixing ratio of sand on the passivation of carbon steel was also studied. The results of the study showed that the present property of buffer material is enough to prevent passivation of carbon steel.
Bamba, Tsunetaka
Genshiryoku Bakkuendo Kenkyu, 8(2), p.205 - 206, 2002/03
no abstracts in English
Sugita, Yutaka; Yui, Mikazu
JNC-TN8450 2001-007, 16 Pages, 2002/02
JNC-TN8450-2001-007.pdf:0.78MB
This report summary the dataset of the relationship between unconfined compressive strength and tensile strength of the rock mass described in supporting report 2; repository design and engineering technology of second progress report (H12 report) on research and development for the geological disposal of HLW in Japan.
*; Tanai, Kenji; *
JNC-TN8420 2001-007, 86 Pages, 2002/02
JNC-TN8420-2001-007.pdf:6.04MB
The objectives of this study is to identify the research issues, which are to be conducted in the future underground research laboratory, about operation and logistics systems for the planning of future research and development program. The research programs and experiments,etc. were investigated for the geological disposal projects in overseas sedimentary rocks and coastal geological environments aiming to reflect in the future underground research facility plan in Japan. In the investigation, information on the engineered-barrier performance, design and construction of underground facilities, tunnel support, transportation and emplacement, and backfilling technology, etc. were collected. Based on these informations, the purpose, the content, and the result of each investigations and tests were arranged. The strategy and the aim in the entire underground research facility, and the flow of investigations and tests, etc. were also arranged from the purpose, the relations and the sequence of each investigation and experiment, and the usage of results, etc.
*; ;
JNC-TN8430 2001-007, 56 Pages, 2002/01
JNC-TN8430-2001-007.pdf:13.13MB
In the geological disposal concept of radioactive wastes, a kind of clay with sorption ability and low permeability, called bentonite, is envisaged as an engineered barrier system in the geological repository. Also, the cemetitious material is envisaged as the backfill material in the vaults and the structure material of the vaults. The groundwater in contact with the cementitious material will promote hyperalkaline conditions in the repository environment and these conditions will affect the performance of the bentonite. Therefore, it is necessary to investigate the interaction between the cementitious material and the bentonite for the evaluation of long term stability of the disposal system. In this study, for the identification and the investigation of the secondary minerals, the batch immersion experiments of the powder bentonite were carried out using synthetic cement leachates (pH=7, 12.5, 14) at 200
C. As the results, it was confirmed that Na as exchangeable cations in the bentonite can exchange relatively easily with Ca in the solution from the experiment results. And the ratio of cation exchange was estimated to be about 25% based on the amount of exchangeable cations Ca
between layers. Furthermore, it was concretely shown that the generation of analcime might be affected by the Na concentration from results of the solution analyses and a stability analysis of analcime using the chemical equilibrium model, in addition to the pH in the solution.
Kato, Hiroshige*; Mine, Tatsuya*; Mihara, Morihiro; Oi, Takao; Honda, Akira
JNC-TN8400 2001-029, 63 Pages, 2002/01
JNC-TN8400-2001-029.pdf:1.81MB
Cementitious materials will be used for the TRU waste repository as a component of engineered barrier system. The distribution coefficients which represent the retardation of radionuclides migration for the cementitious materials would be one of the important parameter for the safety assessment. The much information of radionuclide sorption onto the cementitious materials has been accumulated through the study in the world. Therefore it is necessary to compile the information and Kd of the radionuclides reported in previous studies. In this report, the Kd of the important radionuclides, such as C, Ni, Se, Sr, Zr, Nb, Mo, Tc, Sn, I, Cs, Sm, Pb, Ra, Ac, Th, Pa, U, Np, Pu, Am, Cm, for the cementitious materials were compiled as the Sorption Database (SDB). For radionuclides to be sensitive to the redox potential, e.g. Se, Tc, Pa, U, Pu and Np, some Kds measured under the controlled atmosphere had been reported, and few Kds measured under the controlled redox potential had been reported. For Se, Mo, Sm, Cm and Ac, the distribution coefficients had not been reported, therefore distribution coefficients of Se and Mo for OPC (Ordinary Portland Cement) pastes were measured by batch sorption experiments and these data were added into the SDB.
Ito, Akira; Kawakami, Susumu; Yui, Mikazu
JNC-TN8400 2001-028, 38 Pages, 2002/01
JNC-TN8400-2001-028.pdf:2.35MB
In a repository for high-level radioactive waste, coupled thermo -hydro -mechanical and chemical (THMC) processes will ocurr, involving the interactive processes between radioactive decay heat from the vitrified waste, infiltration of groundwater, swelling pressure generation and chemical evolution of the buffer material and porewater chemistry. In this program, numerical experiment system for the coupled THMC processes will be developed in order to predict the long-term performance of the near-field (engineered barrier and host rock) for various geological environments. The simulation code development has been started in FY 2001 and three development steps are planned, because (1)development will be continued for some years, (2)feasibility of numerical experiment have to be confirmed by using existing tools. This report presents the following items of the simulation code development for the coupled THMC processes. (1)First step of the simulation code development (2)Mass transport passways in compacted bentonite (3)Parallelization of the simulation code
; ; Shigetome, Yoshiaki
JNC-TN8200 2001-006, 19 Pages, 2001/12
JNC-TN8200-2001-006.pdf:0.92MB
None
*;
JNC-TN8400 2001-027, 131 Pages, 2001/11
In order to document a basic manual about input data, output data, execution of computer code on groundwater flow and radionuclide transport calculation in heterogeneous porous rock, we investigated the theoretical background about geostastical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport which calculates water flow in three dimension, the path of moving radionuclide, and one dimensional radionuclide migration. In this report, based on above investigation we describe the geostastical background about simulating heterogeneous permeability field. And we describe construction of files, input and output data, a example of calculating of the programs which simulates heterogeneous permeability field, and calculates groundwater flow and radionuclide transport. Therefore, we can document a manual by investigating the theoretical background about geostastical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport calculation. And we can model heterogeneous porous rock and analyze groundwater flow and radionuclide transport by utilizing the information from this report.
; ; *
JNC-TN8440 2001-024, 210 Pages, 2001/08
JNC-TN8440-2001-024.pdf:24.99MB
In order to make this book reflect in the investigation which turned the bitumen solidification object to maintenance of the abandonment object technical standard on condition of carrying out subterranean disposal in the future - solidification - it created for the purpose of utilizing as precious sources of information, such as a nuclide inventory in the living body, group-izing of the past campaign required for typical solidification object selection, and information offer at the time of disposal examination. A development operation history collected so that histories including the shift action in an institution of the formation of discharge reduction of the characteristic of solidification object manufacture outlines, such as composition of the process of an institution and a solidification object and a storage actual result, the contents of an examination of the past campaign, and the solidification object manufactured based on topics or radioactive iodine and radioactive carbon etc., such as the past contents of an examination / operation, may grasp comprehensively in creation, and it carried out as the composition stared the trend of future disposal fixedly. It was a period (for 16 years) until an bituminization demonstration facility processing institution will start a cold examination from April (Showa 57), 1982, and it starts a hot examination from May 4, it starts solidification processing technical development operation from october 6 and it results in the fire explosion accident on March 11 (Heisei 9), 1997, and low level radioactivity concentration waste fluid was processed 7,438 m
, and 29,967 bitumen solidification objects were manufactured. According to the accident, it is necessary to hand it down to future generations with processing technology while the bitumen solidification object manufactured in 15 years although the bituminization demonstration facility processing institution came to close the mission holds information precious ...
