Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Hamamoto, Takafumi*; Ishida, Keisuke*; Shibutani, Sanae*; Fujisaki, Kiyoshi*; Tachi, Yukio; Ishiguro, Katsuhiko*; McKinley, I. G.*
Proceedings of 2019 International High-Level Radioactive Waste Management Conference (IHLRWM 2019) (USB Flash Drive), p.77 - 82, 2019/04
Goto, Takahiro*; Mitsui, Seiichiro; Takase, Hiroyasu*; Kurosawa, Susumu*; Inagaki, Manabu*; Shibata, Masahiro; Ishiguro, Katsuhiko*
MRS Advances (Internet), 1(63-64), p.4239 - 4245, 2016/00
NUMO and JAEA have conducted a joint research since FY2011, which is designed to enhance the methodology of repository design and performance assessment in preliminary investigation stage for deep geological disposal of radioactive waste. As a part of this joint research, we have been developing glass dissolution models which consider various processes in EBS, such as precipitation of Fe-silicates associated with iron overpack corrosion, and Si transport through corrosion products in the cracked overpack. The objectives of the modeling work are to evaluate relative importance of relevant processes and to identify further R&D issues towards development of a convincing safety case. Sensitivity analyses suggested that predicted glass dissolution time ranges from 1
10
to 110
years or more due to uncertainties in the current understanding of the key processes, namely precipitation of Fe-silicates and transport characteristics of the altered glass layer.
Shibata, Masahiro; Sawada, Atsushi; Tachi, Yukio; Hayano, Akira; Makino, Hitoshi; Wakasugi, Keiichiro; Mitsui, Seiichiro; Oda, Chie; Kitamura, Akira; Osawa, Hideaki; et al.
JAEA-Research 2013-037, 455 Pages, 2013/12
JAEA-Research-2013-037.pdf:42.0MB
Following FY2011, JAEA and NUMO have conducted a collaborative research work which is designed to enhance the methodology of repository design and performance assessment in preliminary investigation stage. With regard to (1) study on rock suitability in terms of hydrology, the tree diagram of methodology of groundwater travel time has been extended for crystalline rock, in addition, tree diagram for sedimentary rock newly has been organized. With regard to (2) study on scenario development, the existing approach has been improved in terms of a practical task, and applied and tested for near field focusing on the buffer. In addition, the uncertainty of some important processes and its impact on safety functions are discussed though analysis. With regard to (3) study on setting radionuclide migration parameters, the approaches for parameter setting have been developed for sorption for rocks and solubility, and applied and tested through parameter setting exercises for key radionuclides.
Shibata, Masahiro; Sawada, Atsushi; Tachi, Yukio; Makino, Hitoshi; Hayano, Akira; Mitsui, Seiichiro; Taniguchi, Naoki; Oda, Chie; Kitamura, Akira; Osawa, Hideaki; et al.
JAEA-Research 2012-032, 298 Pages, 2012/09
JAEA-Research-2012-032.pdf:33.68MB
JAEA and NUMO have conducted a collaborative research work which is designed to enhance the methodology of repository design and performance assessment in preliminary investigation phase. The topics and the conducted research are follows; (1) Study on selection of host rock: in terms of hydraulic properties, items for assessing rock property, and assessment methodology of groundwater travel time has been organized with interaction from site investigation. (2) Study on development of scenario: the existing approach has been embodied, in addition, the phenomenological understanding regarding dissolution of and nuclide release from vitrified waste, corrosion of the overpack, long-term performance of the buffer are summarized. (3) Study on setting nuclide migration parameters: the approach for parameter setting has been improved for sorption and diffusion coefficient of buffer/rock, and applied and tested for parameter setting of key radionuclides. (4) Study on ensuring quality of knowledge: framework for ensuring quality of knowledge has been studied and examined aimed at the likely disposal facility condition.
Ishikawa, Masumi*; Kaneko, Satoru*; Kitayama, Kazumi*; Ishiguro, Katsuhiko*; Ueda, Hiroyoshi*; Wakasugi, Keiichiro*; Shinohara, Nobuo; Okumura, Keisuke; Chino, Masamichi; Moriya Noriyasu*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 8(4), p.304 - 312, 2009/12
Since quality control issues for vitrified waste are defined mainly with the focus on the transport and storage of the waste rather than the long-term safety of geological disposal, they do not cover inventories of long-lived nuclides which are of most interest in the safety assessment of geological disposal. Therefore we suggest a flow chart for assessment of inventories of long-lived nuclides in the vitrified waste focusing on measured value. We started a programme to examine the applicability as well as to improve reliability of nuclide generation/decay code and nuclear data library using liquid waste from spent fuel with clear irradiation history. To solve the issue of quality control for vitrified waste, comprehensive study is needed in aspects not only of geological disposal field but also of operation of nuclear power plant, reprocessing of spent fuel and vitrification of liquid waste. This study is a pioneering study to integrate them.
Ishiguro, Katsuhiko*; Ueda, Hiroyoshi*; Wakasugi, Keiichiro*; Sakabe, Yasushi*; Kitayama, Kazumi*; Umeki, Hiroyuki; Takase, Hiroyasu*
Engineered Barrier Systems (EBS) in the Safety Case; The Role of Modelling, p.167 - 180, 2007/02
Japanese siting approach calls for volunteer host municipalities for an HLW repository and places particular emphasis on design flexibility. The repository concept should be developed to be tailored to the given siting environments. Starting from the H12 repository concept, NUMO has been examining a range of possible repository design options including EBS. The requirements and strategy of model development for performance assessment and process understanding have been discussed taking into account the step-wise, iterative process of development of repository concepts. The areas further to develop the models and databases in the long-term R&D programme have been identified as a wish list in order to evaluate a range of potential repository concepts, focusing on the near-field for the early stages of development process. Among the issues in the list, NUMO has started the development of a flexible computer code for three-dimensional mass transport model to evaluate various design options and components of the EBS. This tool has been applied for the analysis of the barrier effects of the tunnel plugs placed in fractured rock media.
Inagaki, Yaohiro*; Mitsui, Seiichiro*; Makino, Hitoshi*; Ishiguro, Katsuhiko*; Kamei, Gento*; Kawamura, Kazuhiro*; Maeda, Toshikatsu; Ueno, Kenichi*; Bamba, Tsunetaka*; Yui, Mikazu*
Genshiryoku Bakkuendo Kenkyu, 10(1-2), p.69 - 84, 2004/03
no abstracts in English
Inagaki, Yaohiro*; Mitsui, Seiichiro; Makino, Hitoshi; Ishiguro, Katsuhiko; Kamei, Gento; Kawamura, Kazuhiro; Maeda, Toshikatsu*
JNC TN8400 2003-036, 53 Pages, 2003/12
JNC-TN8400-2003-036.pdf:0.51MB
Obtain of sufficient data for the performance of high-level radioactive waste(HLW) glass and verification of a model for the radionuclide release from the HLW glass in the disposal condition are required in order to show the objective reliability. In this paper, some reports about performance assessment of HLW glass are reviewed and we clarify the problems to raise reliability comparing these reports.
WAKASUGI, Keiichiro; Miyahara, Kaname; Makino, Hitoshi; Ishiguro, Katsuhiko; Sawamura, Hidenori*; Neyama, Atsushi*; Nishimura, Kazuya*
JNC TN8400 2003-022, 84 Pages, 2003/11
JNC-TN8400-2003-022.pdf:9.58MB
The radiation dose from the vitrified waste which is the same specification set in the Reference Case of the second progress report (H12 report) was evaluated quantitatively taking into account of the shield of the canister and the overpack. In order to understand the feature of radiation dose from the vitrified waste in terms of shielding, the thickness of the concrete shield to decrease less than safety standard for a radiation controlled zone was evaluated. Main results are summarized as follows. (1)The effective dose rates in the case considering the vitrified waste and the canister decrease approximate 4
5 orders of magnitude during the period of 1,000 years after vitrification due to decay of short half-life radionuclides. The effective dose rate doesn't decrease from 1,000 to 10,000 years. (2)The effective dose rates at the outside of overpack in the case considering the vitrified waste, canister and overpack are smaller than those inside of overpack approximate 5 orders of magnitude during the period of 100 years due to shielding effect of the overpack. However this difference is relatively small after 100 years since the contribution of 
radiation to total effective dose rates decrease due to decay of fission products. (3)Excepting a few cases, the result using the old law (dose equivalent rate) is larger than the result using the new law (effective dose rate). However the difference between these results is less than factor of 1.2. (4)The thickness of the concrete shield required to attenuate the effective dose during the period of 50 years less than safety standard for a radiation controlled zone is calculated as approximate 0.8m1.5m. The important factors to determine the thickness of the concrete shield are the radiation in the case of vitrified waste and the canister, and the neutron radiation in the case of vitrified waste, canister and overpack.
; ; Suzuki, Yuji*; ; ;
Health Physics, 0 Pages, 2002/00
None
Kato, Tomoko; ; ; G.Smith*
T-8B-2, 0 Pages, 2002/00
None
Kato, Tomoko; Suzuki, Yuji*; Ishiguro, Katsuhiko; ; Ikeda, Takao*; Richard, L.*
JNC TN8400 2001-013, 100 Pages, 2001/03
JNC-TN8400-2001-013.pdf:5.48MB
In the safety assessment of a high-level radioactive waste (HLW) disposal system, it is required to estimate radiological impacts on future human beings arising from potential radionuclide releases from a deep repository into the surface environment. In order to estimate the impacts, a biosphere model is developed by reasonably assuming radionuclide migration processes in the surface environment and relevant humman lifestyles. It is important to modify the present biosphere models or to develop alternative biosphere models applying the biosphere models according to quality and quantify of the information acquired through the siting process for constructing the repository. In this study, alternative biosphere models were developed taking geosphere-biosphere interface of marine environment into account. Moreover, the flux to dose conversion factors calculated by these alternative biosphere models was compared with those by the present basic biosphere models.
Kato, Tomoko; ; Suzuki, Yuji*; ; Ishiguro, Katsuhiko; Ikeda, Takao*; Richard, L.*
JNC TN8400 2001-003, 128 Pages, 2001/03
JNC-TN8400-2001-003.pdf:6.09MB
In the safety assessment of a high-level radioactive waste (HLW) disposal system, it is required to estimate radiological impacts on future human beings arising from potential radionuclide releases from a deep repository into the surface environment. In order to estimate the impacts, a biosphere model is developed by reasonably assuming radionuclide migration processes in the surface environment and relevant human lifestyles. Releases from the repository might not occur for many thousands of years after disposal. Over such timescales, it is anticipated that the considerable climatic change, for example, induced by the next glaciation period expected to occur in around ten thousand years from now, will have a significant influence on the near surface environment and associated human lifestyles. In case of taking these evolution effects into account in modeling, it is reasonable to develop several alternative models on biosphere evolution systems consistent with possible future conditions affected by expected climatic changes. In this study, alternative biosphere models were developed taking effects of possible climatie change into account. In the modeling, different climatic states existing in the world from the present climate condition in Japan are utilized as an analogy. Estimation of net effects of the climatic change on biosphere system was made by comparing these alternative biosphere models with a constant biosphere model consistent with the present climatic state through flux to dose conversion factors derived from each one.
; ; Makino, Hitoshi; Ishiguro, Katsuhiko
JNC TN8400 2000-002, 132 Pages, 2000/01
JNC-TN8400-2000-002.pdf:33.9MB
The evaluation of the effects of buffer thickness and dry density, one of the buffer design, on radionuclides migration behavior is important from the viewpoint of performance assessment since they have relation to radionuclides migration retardation. It is also considered to help investigation of buffer design that satisfy both safety and economy to condition of the disposal site, which may be required with development of disposal project in the future. Therefore we have performed a sensitivity analysis used buffer thickness and dry density as parameter and considered their combination in this report. Based on this, we have evaluated the effects of buffer thickness and dry density on radionuclides migration in engineered barrier system. And, we have considered about radionuclides migration retardation quality of the buffer which is based on the design (relationship between thickness and dry density) set in the second progress report on research and development for the geological disposal of HLW in Japan. In results, the maximum release rates from the engineered barrier system for the nuclides which have high distribution coefficients and short half lives are sensitive to changes in buffer thickness and dry density. And, using dose converted from the nuclide release rates from the engineered barrier system as a convenient index, it is almost shown that the maximum of total dose is less than 10
SV/y in the cases which buffer thickness and dry density are based on the buffer design set in the rsecond progress report on research and development for the geological disposal of HLW in Japan. These can be used as an information when design of bufer thickness and dry density is set by synthetically judgement of balance of safety and economy.
; ; Ishiguro, Katsuhiko; Nakajima, Kunihiko*;
JNC TN8400 99-087, 41 Pages, 1999/11
Corrosion of the carbon steel overpack leads to a volume expansion since the specific gravity of corrosion products is smaller than carbon steel. The buffer material is compressed due to the corrosive swelling, reducing its thickness and porosity. On the other hand, Buffer material may be extruded into fractures of the surrounding rock and this may lead to a deterioration of the planned functions of the buffer, including retardation of nuclides migration and colloid filtration. In this study, the sensitivity analyses for the effect of volume expansion and intrusion of the buffer material on nuclide migration in the engineering barrier system are carried out. The sensitivity analyses were performed on the decrease in the thickness of the buffer material in the radial direction caused by the corrosive swelling, and the change in the porosity and dry density of the buffer caused by both compaction due to corrosive swelling and intrusion of buffer material. As results, it was found the maximum release rates of relatively shorter half-life nuclides from the outside of the buffer material decreased for taking into account of a volume expansion due to overpack corrosion. On the other hand, the maximum release rates increased when the intrusion of buffer material was also taking into account. It was, however, the maximum release rates of longer half-life nuclides, such as Cs-137 and Np-237, were insensitive to the change of buffer material thickness, and porosity and dry density of buffer.
; Ishiguro, Katsuhiko; *
JNC TN8400 99-086, 17 Pages, 1999/11
Se-79 is one of key radionuclides in the performance assessment of the geological disposal system. Based on recent measurements, it is possible that the half-life of Se-79 will be changed longer than the present value in most handbooks and tables of isotopes. This study presents performance assessment calculations to investigate the overall effect of change in half-life of Se-79 on the repository system safety. The total system performance analyses for Se-79 were carried out, which focussed on the Reference-Case of the safety assessment in the H12 Project. As results, the maximum release rate in Becquerel unit of Se-79 from the engineered barrier system with new half-life decreases about one order of magnitude than that with half-life used so far. It is, however, that the maximum release rate in Becquerel unit of Se-79 from the natural barrier system is almost same for both half-life because of the channelling effects of groundwater flow. Consequently, the calculated maximum dose rate of Se-79 with new half-life does not change. It can be concluded that the change in half-life of Se-79 does not affect overall safety of the H12 disposal concept.
; Makino, Hitoshi; ; Ishiguro, Katsuhiko; Miyahara, Kaname; ; *
JNC TN8400 99-085, 88 Pages, 1999/11
Spent fuel removed from nuclear power plants in Japan is reprocessed at home and abroad (Japan Nuelear Fuel Limited [JNFL] reprocessing plant, Tokai Vitrification Facility [TVF] of JNC, COGEMA in France and BNFL in England) and then vitrified. The properties of vitrified waste change depending on the fuel type, fuel burn-up in the reactor, and operating conditions at the reprocessing Plants. However, properties of vitrified waste, such as heat generation and nuclide inventories, are important information for the design study of repository and the performance assessment in the geological disposal system. For the objectives of repository design and safety assessment, it is necessary that the model-vitrified waste is determined from the four types of waste. In this study, for supporting the determination of the model vitrified waste in the Research and Development for the Geological Disposal of HLW in Japan (H12 Project), the calculation and comparison of inventories for the four types of waste were performed. As results, it is found that there are no significant differences in the properties (i.e., radioactivity, heat generation, hazard index and nuclide inventories per one package) of the four types of vitrified waste from JNFL, COGEMA, BNFL and TVF.
Baba, Tomoko; ; Suzuki, Yuji*; ; Ishiguro, Katsuhiko; Ikeda, Takao*; Richard*
JNC TN8400 99-084, 254 Pages, 1999/11
In the safety assessment of a hifh-level radioactive waste disposal system, it is required to estimate future radiological impacts on human beings. Consideration of living habits and the human environment in the future involves a larg degree of uncertainty. To avoid endless speculation aimed at reducing such uncertainty, an approach is applied for identifying and justifying a "reference biosphere" for use in safety assessment in Japan. Considering a wide rang of Japanese geological environments, some specific "reference biospheres" are developed using an approach consistent with the BIOMOVS II reference biosphere methodology. The models represent the components of the surface environment using compartments between which fluxes of materials (solid/water) and radionuclides are defined by transfer factors. A range of exposure pathways via which such radionuclides enter the food-chain, along with uptake and concentration factors, are also defined. The response to a step function of unit flux from the geosphere is determined for each model. The results show that it is reasonable to use steady-state biosphere responses to a unit-input flux to define nuclide-dependent factors for converting fluxes from the geosphere to doses. This simplifies safety assessment calculations, which then require only look-up tables for such flux to dose conversion rather than fully coupled biosphere models.
