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Sugaya, Atsushi; Tanaka, Kenji; Akutsu, Shigeru
Proceedings of International Waste Management Symposia 2011 (WM2011) (CD-ROM), 11 Pages, 2011/02
The main component of the liquid wastes is sodium nitrate. Nitrate ion decomposition technology is under development to conserve a circumference environment of a disposal site. To investigate the methods for decomposing nitrate ion, several small-scale trials were performed using reductants and a catalyst in sodium nitrate solutions. It will be reported that the cement based encapsulation trials to immobilize the sodium carbonate based liquid waste, which was performed under non-radioactive condition at both small and full scale to investigate the optimum cement formulation.
Tanai, Kenji; Kikuchi, Hirohito; Nakamura, Kunihiko*; Tanaka, Yukihisa*; Hironaga, Michihiko*
JAEA-Research 2010-025, 186 Pages, 2010/08
JAEA-Research-2010-025.pdf:9.01MB
Bentonite-based material is used as one of the components of the LLW, TRU and HLW disposal facilities. Required characteristics of bentonite-based material are low permeability, swelling property, etc. Those are evaluated in many cases by laboratory test results. However the uncertainty exists in the evaluation of those characteristics in construction. Because even if the index of the dry density etc is the same, laboratory test results have variability. In addition, the uncertainty in construction has the possibility to increase the uncertainty of long-term evaluation of characteristics. On the other hand, several of laboratory test methods of bentonite are not standardized. So, this is a possibility that is one of the uncertain causes of the evaluation of characteristics of the bentonite. Therefore, it is hoped that the laboratory tests of bentonite are standardized. Therefore, this study analysis the uncertainty on the physical properties by laboratory tests and put together the problem and ponts of concern in the tests.
Horiguchi, Kenichi; Sugaya, Atsushi; Saito, Yasuo; Tanaka, Kenji; Akutsu, Shigeru; Hirata, Toshiaki
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9411_1 - 9411_9, 2009/05
The low-level radioactive Waste treatment Facility (LWTF) was constructed at the Tokai Reprocessing Plant (TRP) and cold test has been carried out since 2006. The waste which is treated in the LWTF is combustible/incombustible solid waste and liquid waste. In the LWTF, The combustible/incombustible solid waste will be incinerated. The liquid waste will be treated by the radio-nuclides removal process subsequently solidified by cement materials. This report describes the essential technologies of the LWTF and results of R&D work for the nitrate-ion decomposition technology for the liquid waste.
Sugaya, Atsushi; Horiguchi, Kenichi; Tanaka, Kenji; Kobayashi, Kentaro
Materials Research Society Symposium Proceedings, Vol.1107, p.173 - 179, 2008/00
In Nuclear Fuel Reprocessing Plant, it is necessary to dispose of a large amount of low level radioactive effluent containing nitrate as a major ingredient, safely and economically. Therefore, engineering developments concerning a cement based encapsulation process have been carried out in JAEA. From the viewpoint of disposal cost decrease, a low level radioactive effluent is passed through the nuclide separation process before cementation to concentrate the radioactivity into the minimum volume for conditioning and disposal. Two kinds of effluents are generated as a result of the nuclide separation; Non-radioactive simulants were prepared for each of these waste streams, and used in encapsulation trials to investigate a special slag cement, on a beaker scale and at full scale (200-litres). The results have confirmed that the nitrate effluent, evaporated up to a predetermined density, can be successfully encapsulated at a salt filling rate of 50wt%, to produce a wasteform which satisfies the required conditions. In the slurry effluent, the strength of the product decreased when carbonate concentration was high. However, it was confirmed that the product made at salt filling rate 50wt% satisfied the required conditions, if the carbonate concentration in the effluent was decreased to 10 g/L or less.
Saito, Yasuo; Takano, Masato; Tanaka, Kenji; Kobayashi, Kentaro; Otani, Yoshikuni
Proceedings of International Symposium on Radiation Safety Management 2007 (ISRSM 2007), p.275 - 280, 2007/11
The Low-radioactive Waste Treatment Facility (LWTF), which aims to provide the safe, efficient and economic treatment and disposal of Low-level Liquid Waste (LLW) generated from LWR spent fuel reprocessing, was constructed at the Tokai Reprocessing Plant (TRP), and a cold test is now being carried out. New treatment processes such as a removal process for radio-nuclides and the ROBE (BORSAURE EINENGUNG ANLAGE) solidification process are being implemented in the LWTF. In order to treat the large amount of sodium nitrate contained in the concentrated LLW with higher safety and economy, R&D work on nitrate-ion decomposition technology using a catalytic reduction method and on the solidification process by cementation is being undertaken. The results of this R&D will be adopted in the LWTF in the near future. This report describes an outline of liquid waste treatment in the LWTF and new treatment technologies for LLW to achieve safe, efficient and economic treatment and disposal.
; ;
Proceedings of 8th International Conference on Radioactive Waste Management and Environmental Remediation (ICEM '01) (CD-ROM), 20A(15), 0 Pages, 2001/00
None
; ; Ishibashi, Takashi;
Asufaruto Koka No Anzen To Seino Hyoka Ni Kansuru Kokusai Wakushoppu, 0 Pages, 1999/00
None
Maeda, Munehiro*; Tanai, Kenji; Ito, Masaru; Mihara, Morihiro; Tanaka, M.*
PNC TN8410 98-021, 136 Pages, 1998/03
The buffer material component of the engineered barrier system of the radioactive waste repository, functions to maintain low groundwater flow and mechanical stability in the repository for long periods of time. If Na bentonite is used as a buffer material, it is possible that the Na bentonite will change to Ca bentonite by exposure to Ca ions derived from calcite in the ground water. In the TRU waste disposal repository if cementitious materials are used, the change from Na to Ca bentonite may be almost immediate. Therefore it is important to investigate the mechanical properties of Ca bentonite as part of TRU waste disposal research and development. This paper reports the results of swelling pressure, water permeability and compressive strength tests for compacted Ca bentonite and for bentonite which has undergone Na-Ca exchange. Maximum and equilibrium swelling pressure tests and water permeability tests were performed. Compressive strength tests produced compressive strength and elastic modulus values for unsaturated compacted bentonite. The results of these test correlate with the dry density and sand mixing contents of bentonite. Finally, the relative properties of Ca, Na and Ca-Na exchanged bentonite were compared. It is clear that the maximum and equilibrium swelling pressures and hydraulic conductivity of compacted Ca-Na exchanged bentonite are the same as Na bentonite when both their dry density was about 1.8 g/cm
. The compressive strength and elastic modulus of compacted Ca-Na exchanged bentonite are a little higher than Na bentonite. When compacted, results are the same for either Ca bentonite (dry density 1.4g/cm) or Ca-Na exchanged bentonite compared with Na bentonite (dry density 1.6-1.8g/cm).
Maki, Yasuro*; Kitano, Koichi*; Inoue, Daiei*; Onuma, Hiroshi*; Komada, Hiroya*; Yamaji, Kenji*; Osumi, Takashi*; Tanaka, Hiroshi*; Imazu, Masanori*
JNC TJ1400 2005-005, 98 Pages, 1989/03
JNC-TJ1400-2005-005.pdf:2.98MB
no abstracts in English
Sugaya, Atsushi; Horiguchi, Kenichi; Tanaka, Kenji; Kobayashi, Kentaro
no journal, ,
In Nuclear Fuel Reprocessing Plant, it is necessary to dispose of a large amount of low level radioactive effluent containing nitrate as a major ingredient, safely and economically. Therefore, engineering developments concerning a cement based encapsulation process have been carried out in JAEA. From the viewpoint of disposal cost decrease, a low level radioactive effluent is passed through the nuclide separation process before cementation to concentrate the radioactivity into the minimum volume for conditioning and disposal. Two kinds of effluents are generated as a result of the nuclide separation; a nitrate effluent of which the principal ingredient is nitrate with a comparatively low radiation level, and a slurry effluent including several kinds of salts with a comparatively high radiation level. Non-radioactive simulants were prepared for each of these waste streams, and used in encapsulation trials to investigate a special slag cement, on a beaker scale and at full scale (200-litres). The results have confirmed that the nitrate effluent, evaporated up to a predetermined density, can be successfully encapsulated at a salt filling rate of 50wt%, to produce a wasteform which satisfies the required conditions. In the slurry effluent, the strength of the product decreased when carbonate concentration was high. However, it was confirmed that the product made at salt filling rate 50wt% satisfied the required conditions, if the carbonate concentration in the effluent was decreased to 10g/L or less.
Takano, Masato; Horiguchi, Kenichi; Tanaka, Kenji; Kobayashi, Kentaro
no journal, ,
no abstracts in English
Takano, Masato; Tanaka, Kenji; Kobayashi, Kentaro; Tsukamoto, Ryosuke*
no journal, ,
no abstracts in English
Horiguchi, Kenichi; Sugaya, Atsushi; Tanaka, Kenji; Kobayashi, Kentaro; Sasaki, Tadashi*
no journal, ,
no abstracts in English
Takano, Masato; Kojima, Hiroshi; Tanaka, Kenji; Kobayashi, Kentaro; Tsukamoto, Ryosuke*
no journal, ,
no abstracts in English
Horiguchi, Kenichi; Sugaya, Atsushi; Tanaka, Kenji; Kobayashi, Kentaro; Sasaki, Tadashi*
no journal, ,
In Nuclear Fuel Reprocessing Plant, it is necessary to dispose of a large amount of low level radioactive effluent containing nitrate as a major ingredient, safely and economically. Therefore, engineering developments concerning a cement based encapsulation process have been carried out in JAEA. From the view point of disposal cost decrease, a low level radioactive effluent is passed through the nuclide separation process before cementation to concentrate the radioactivity into the minimum volume for conditioning and disposal. Two kinds of effluents are generated as a result of the nuclide separation; A nitrate effluent of which the principal ingredient is nitrate with a comparatively low radiation level, and; A slurry effluent including several kinds of salts with a comparatively high radiation level. Non-radioactive stimulants were prepared for each of these waste streams, and used in encapsulation trials to investigate special slag cement, on a beaker scale and full scale(200-litres). Furthermore, JAEA has carried out hazardous material judgment for cement products and leaching test of the cement products which encapsulated actual effluent. I will report that result of there development trials.
Takano, Masato; Kojima, Hiroshi; Tanaka, Kenji; Kobayashi, Kentaro; Tsukamoto, Ryosuke*
no journal, ,
no abstracts in English
Horiguchi, Kenichi; Sugaya, Atsushi; Tanaka, Kenji; Kobayashi, Kentaro; Sasaki, Tadashi*
no journal, ,
In Nuclear Fuel Reprocessing Plant, it is necessary to dispose of a large amount of low level radioactive effluent safely and economically. In JAEA engineering developments concerning a cement solidification process have been carried out. The phosphate effluent occurring from solvent waste treatment facility is based on sodium dihydrogen phosphate. The acidity of this effluent (pH 4) requires a pre-treatment process before cement solidification. Phosphate effluent interfere with cementing reactions by retard of reaction rate and loss of strength, because it is combined with calcium that is element of cement material. It reports on the result of Non-radioactive simulant was prepared for the phosphate effluent, and used in cementation trials to investigate a special slag cement, on a beaker scale and at full scale (200-litres).
