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Nakashio, Nobuyuki; Osugi, Takeshi; Iseda, Hirokatsu; Tohei, Toshio; Sudo, Tomoyuki; Ishikawa, Joji; Mitsuda, Motoyuki; Yokobori, Tomohiko; Kozawa, Kazushige; Momma, Toshiyuki; et al.
Journal of Nuclear Science and Technology, 53(1), p.139 - 145, 2016/01
Times Cited Count：1 Percentile：9.88(Nuclear Science & Technology)no abstracts in English
Sudo, Tomoyuki; Mimura, Ryuji; Ishihara, Keisuke; Satomi, Shinichi; Myodo, Masato; Momma, Toshiyuki; Kozawa, Kazushige
JAEA-Technology 2011-015, 24 Pages, 2011/06
JAEA-Technology-2011-015.pdf:2.28MB
The super compactor in the Advanced Volume Reduction Facilities (AVRF) treats metal wastes mainly generated from research reactors in the Nuclear Science Research Institute of JAEA. Those wastes are compacted from one third to one fourth with maximum 2,000-ton force. In the trial operation using simulated wastes, some technical problems were found to be improve for the stable operation. One problem is the motion mechanism for carrying wastes before and after compaction. The other problem is the mechanism for treating the irregular supercompacted products. In this report, we describe the detail and the result of improvement on those problems for the stable operation in the super compactor.
Sudo, Tomoyuki; Nakashio, Nobuyuki; Osugi, Takeshi; Mimura, Ryuji; Ishihara, Keisuke; Satomi, Shinichi; Myodo, Masato; Momma, Toshiyuki; Kozawa, Kazushige
JAEA-Technology 2010-041, 38 Pages, 2011/01
JAEA-Technology-2010-041.pdf:4.73MB
The super compactor in the AVRF treats compactible metal wastes mainly generated from research reactors in the Nuclear Science Research Institute of JAEA. Those wastes are compacted with the maximum about 2,000-ton force. The supercompacted wastes are packed into the container and then immobilized with cementitious materials. The solidified wastes (containing supercompacted wastes) become an object for near surface disposal with artificial barrier. For disposal, the solidified wastes must be satisfied the technical criteria. One of the important indicators is the void ratio in the solidified wastes. In this report, we manufactured the supercompacted wastes with the ordinary treatment method for actual wastes treated in the AVRF and immobilized with a mortar grout. The void ratio of the solidified wastes were evaluated in consideration for concrete vault disposal. As a result, We confirmed the integrity of the solidified wastes from a point of view of void ratio.
Higuchi, Hidekazu; Osugi, Takeshi; Nakashio, Nobuyuki; Momma, Toshiyuki; Tohei, Toshio; Ishikawa, Joji; Iseda, Hirokatsu; Mitsuda, Motoyuki; Ishihara, Keisuke; Sudo, Tomoyuki; et al.
JAEA-Technology 2007-038, 189 Pages, 2007/07
JAEA-Technology-2007-038-01.pdf:15.13MBJAEA-Technology-2007-038-02.pdf:38.95MBJAEA-Technology-2007-038-03.pdf:48.42MBJAEA-Technology-2007-038-04.pdf:20.53MBJAEA-Technology-2007-038-05.pdf:10.44MB
The Advanced Volume Reduction Facilities (AVRF) is constructed to manufacture the waste packages of radioactive waste for disposal in the Nuclear Science Research Institute of the Japan Atomic Energy Agency. The AVRF is constituted from two facilities. The one is the Waste Size Reduction and Storage Facility (WSRSF) which is for reducing waste size, sorting into each material and storing the waste package. The other is the Waste Volume Reduction Facility (WVRF) which is for manufacturing the waste package by volume reducing treatment and stabilizing treatment. WVRF has an induction melting furnace, a plasma melting furnace, an incinerator, and a super compactor for treatment. In this report, we summarized about the basic concept of constructing AVRF, the constitution of facilities, the specifications of machineries and the state of trial operation until March of 2006.
Nakashio, Nobuyuki; Higuchi, Hidekazu; Momma, Toshiyuki; Kozawa, Kazushige; Tohei, Toshio; Sudo, Tomoyuki; Mitsuda, Motoyuki; Kurosawa, Shigenobu; Hemmi, Ko; Ishikawa, Joji; et al.
Journal of Nuclear Science and Technology, 44(3), p.441 - 447, 2007/03
Times Cited Count：9 Percentile：53.98(Nuclear Science & Technology)The Japan Atomic Energy Agency (JAEA) constructed the Advanced Volume Reduction Facilities (AVRF), in which volume reduction techniques are applied and achieved high volume reduction ratio, homogenization and stabilization by means of melting or super compaction processes for low level solid wastes. It will be able to produce waste packages for final disposal and to reduce the volume of stored wastes by operating the AVRF. The AVRF consist of the Waste Size Reduction and Storage Facilities (WSRSF) and the Waste Volume Reduction Facilities (WVRF); the former have cutting installations for large size wastes and the latter have melting units and a super compactor. Cutting installations in the WSRSF have been operating since July 1999. Radioactive wastes treated so far amount to 750 m
and the volume reduction ratio is from 1.7 to 3.7. The WVRF have been operating with non-radioactive wastes since February 2003 for the training and the homogeneity investigation in the melting processes. The operation of the pretreatment system in the WVRF with radioactive wastes has partly started in FY2005.
Higuchi, Hidekazu; Momma, Toshiyuki; Nakashio, Nobuyuki; Kozawa, Kazushige; Tohei, Toshio; Sudo, Tomoyuki; Mitsuda, Motoyuki; Kurosawa, Shigenobu; Hemmi, Ko; Ishikawa, Joji; et al.
Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10
The JAERI constructed the Advanced Volume Reduction Facilities(AVRF). The AVRF consists of the Waste Size Reduction and Storage Facilities(WSRSF) and the Waste Volume Reduction Facilities(WVRF). By operating the AVRF, it will be able to produce waste packages for final disposal and to reduce the amount of the low level solid wastes. Cutting installations for large wastes such as tanks in the WSRSF have been operating since June 1999. The wastes treated so far amount to 600 m and the volume reduction ratio is around 1/3. The waste volume reduction is carried out by a high-compaction process or melting processes in the WVRF. The metal wastes from research reactors are treated by the high-compaction process. The other wastes are treated by the melting processes that enable to estimate radioactivity levels easily by homogenization and get chemical and physical stability. The WVRF have been operating with non-radioactive wastes since February 2003 for the training and the homogeneity investigation in the melting processes. The operation with radioactive wastes will start in FY2005.
Shirai, Nobutoshi; Sudo, Toshiyuki; Nojiri, Ichiro
Monte Karuroho Niyoru Ryoshi Shimyureshon No Genjo To Kadai, p.235 - 249, 2002/00
None
Shirasu, Noriko; Yamashita, Toshiyuki; Kanazawa, Hiroyuki; Kimura, Yasuhiko; Sudo, Kenji; Magara, Masaaki; Inagawa, Jun; Kono, Nobuaki; Nakahara, Yoshinori
JAERI-Research 2001-018, 23 Pages, 2001/03
JAERI-Research-2001-018.pdf:1.48MB
no abstracts in English
Ishida, Michihiko; Sudo, Toshiyuki; Omori, Eiichi; Nojiri, Ichiro
Vol.1 No.040, 1(40), 0 Pages, 2001/00
None
Shirai, Nobutoshi; ; ; Shirozu, Hidetomo; Sudo, Toshiyuki; Hayashi, Shinichiro;
JNC TN8410 2000-006, 116 Pages, 2000/04
JNC-TN8410-2000-006.pdf:2.77MB
Criticality limits for equipments in Tokai Reprocessing Plant which handle fissile material solution and are under shape and dimension control were reevaluated based on the guideline No.10 "Criticality safety of single unit" in the regulatory guide for reprocessing plant safety. This report presents criticality safety evaluation of each equipment as single unit. Criticality safety of multiple units in a cell or a room was also evaluated. The evaluated equipments were ones in dissolution, separation, purification, denitration, Pu product storage, and Pu conversion processes. As a result, it was reconfirmed that the equipments were safe enough from a view point of criticality safety of single unit and multiple units.
Shirai, Nobutoshi; Taguchi, Katsuya; Iitsuka, Shoji; ; ; Sudo, Toshiyuki
JNC TN8410 99-055, 69 Pages, 1999/09
As a part of the safety confirmation work of Tokai Reprocessing Plant, an assessment of the basic data for criticality safety and shielding design has already been reported. In that report, two plutonium solution storage cells were evaluated to be safe enough from a viewpoint of multiple unit criticality safety. In this report, additionai evaluation of multiple unit criticality safety was made for the main plant and the UO
storage which were designed in 1960's and constructed in 1970's. The evaluated cells and rooms are enriched uranium dissolution cell, adjustment and feeding cell, two second extraction cycle cells, third uranium cycle cell, uranium concentration and denitration rooms, third plutonium cycle cell, plutonium concentration cell, two plutonium solution storage cells, rework cell, and UO storage room. As a result, it was confirmed that these cells and rooms were safe enough from a viewpoint of multiple unit criticality safety.
Naito, Motoyuki; Sudo, Toshiyuki; Asakawa, Kazuhiro*; Kashiwagi, Eisuke*
JNC TN8400 99-005, 273 Pages, 1999/03
MIXSET is a FORTRAN code developed in about 1980 to simulate the PUREX solvent extraction process using mixer-settler type extractors. A rebuilt version of it, MIXSET-X, has been developed to give new features listed below. (1)Multiple distribution coefficients of 31 constituents including actinides and FPs and 45 chemical reactions are built in. (2)Users can select any distribution coefficients and chemical reactions to be handled and addust each reaction rate constants by the input data. (3)Recycle flow through buffer tank and TBP degradation calculation can be treated. MIXSET-X can simulate the whole solvent extraction processes of Tokai reprocessing plant at one calculation run. Program structure and mathematical modeling of the system have been entirely changed to get calculational stability and program readability. This report describes detailed features and program list of the code. For further study, validation for calculation resultes of MIXSET-X will be performed.
Omori, Eiichi; Sudo, Toshiyuki; ; Kosaka, Ichiro; ; ;
JNC TN8410 99-005, 274 Pages, 1999/02
Through the investigation of the cause of the fire and explosion incident at Bituminization Demonstration Facility of JNC Tokai Works, the lesson learned is that the safety assessment is necessary even for the licensed facilities by recent knowledge. The safety assessment has been conducted for the facilities in Tokai Reprocessing Plant by recent knowledge and operational experience. This report describes the evaluation results of the incident mitigation systems and the hypothetical accidents. In the evaluation of the incident mitigation system, supposed incidents were solvent fire, rapid reaction of hydrazine decomposition, leakage of radioactive material and loss of power supply. The evaluation was focused on the integrity of the filters in case of the fire and the rapid leaction, the availability of the recovery system in case of the radioactive leakage, and so on. As a result of evaluation, several improvements were pointed out to be necessary for the prevention of incident magnification. In the evaluation of the hypothetical accidents, criticality at a dissolver and fire at solvent extraction mixer-settlers were hypothesized. It was confirmed that the Tokai Reprocessing Plant is still distant enough from the population.
Sudo, Toshiyuki; ; ; Nojiri, Ichiro; Maki, Akira; Yamanouchi, Takamichi
JNC TN8410 99-003, 69 Pages, 1998/11
As a part of the safety confirmation work of Tokai Reprocessing Plant, the appropriateness was checked on the basic data used in criticality safety and shielding design of early-designed facilities in the plant on the basis of recent knowledge and safety evaluation methods. In the criticality safety design, it was confirmed that critical and subcritical values concerning mass and concentration of U and Pu and equipment dimension were appropriate. In the shielding design, it was found that the relation between shielding thickness and permissible radioactivity might give underestimated results of shielding thickness necessary to limit dose rate to the designated one on some condition. In this cases, however, it was confirmed that necessary shielding thickness has been secured because of the conservative calculation conditions for the real conditions except the operation test laboratory (OTL). However, the amount of radioactivity handled at OTL needs to be limited. From a viewpoint of criticality safety, operational control for U and Pu transfer was also investigated, As a result of it, at the transfer route where erroneous batch-wise transfer of process solution might lead to a criticality accident, the reliability of U and Pu concentration measurement needs to be improved by multiple measurements. At other transfer routes, it was confirmed that single failure of equipment or operation error would not lead to a criticality problem.
Omori, Eiichi; Sudo, Toshiyuki; ; ; ; ; Maki, Akira
JNC TN8410 99-002, 205 Pages, 1998/11
Through the investigation of the cause of the fire and explosion incident at Bituminization Demonstration Facility of JNC Tokai Works, the lesson learned is that the safety assessment is necessary even for the licensed facilities by recent knowledge. The safety assessment has been conducted for the facilities in Tokai Reprocessing Plant by recent knowledge and operational experience. This report describes the review of the design data for safety assessment of Tokai Reprocessing Plant. The spent fuel inventory, the radio activity balance in the processes, the inventory contained in each equipment, and the evaluation method of public dose were evaluated with the recent data, the new calculation method and the data obtained through the plant operation. The important equipments were selected for safety assessment as for the public dose. The hydrogen generation from radiolysis of solution was also evaluated, and the hydrogen concentration in each equipment was kept lower than the flammable limit except several equipments that are to be improved.
Koyama, Tomozo; ; Sano, Yuichi; Omori, Eiichi; ; ; Kitatani, Fumito; Kosugi, Kazumasa; Sudo, Toshiyuki; Kikuchi, Naoki; et al.
Nihon Genshiryoku Gakkai-Shi, 40(10), p.740 - 766, 1998/00
Times Cited Count：1 Percentile：14.87(Nuclear Science & Technology)None
Sudo, Toshiyuki; ; ;
Donen Giho, (99), p.99 - 104, 1996/09
None
Sudo, Toshiyuki; Takahashi, Yuki*
PNC TN8450 96-006, 91 Pages, 1996/08
The 27-group ENDF/B-IV cross section library included in the SCALE code system is a general-purpose criticality analysis library. This library has been extensively vilidated against critical experiments and has been widely used in the world. The purpose of this report is to present the cross sections in a graphic manner so that the report can be useful for criticality analysts to understand the characteristics of the library. The graphic plots include total, absorption, capture cross sections and fission cross section multiplied by 
-value. In addition, numerical data of these cross sections and elastic, fission cross section and -value are tabulated for convenience.
Shirai, Nobutoshi; Sudo, Toshiyuki
PNC TN8460 95-001, 92 Pages, 1995/09
no abstracts in English
