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JAEA Reports

Horonobe Underground Research Laboratory Project; Synthesis of Phase II (Construction Phase) investigations to a depth of 350m

Sato, Toshinori; Sasamoto, Hiroshi; Ishii, Eiichi; Matsuoka, Toshiyuki; Hayano, Akira; Miyakawa, Kazuya; Fujita, Tomoo*; Tanai, Kenji; Nakayama, Masashi; Takeda, Masaki; et al.

JAEA-Research 2016-025, 313 Pages, 2017/03

JAEA-Research-2016-025.pdf:45.1MB

The Horonobe Underground Research Laboratory (URL) Project is being pursued by the Japan Atomic Energy Agency (JAEA) to enhance the reliability of relevant disposal technologies through investigations of the deep geological environment within the host sedimentary formations at Horonobe, northern Hokkaido. This report summarizes the results of the Phase II investigations carried out from April 2005 to June 2014 to a depth of 350m. Integration of work from different disciplines into a "geosynthesis" ensures that the Phase II goals have been successfully achieved and identifies key issues that need to made to be addressed in the Phase II investigations Efforts are made to summarize as many lessons learnt from the Phase II investigations and other technical achievements as possible to form a "knowledge base" that will reinforce the technical basis for both implementation and the formulation of safety regulations.

JAEA Reports

The calculation and estimation of wastes generated by decommissioning of nuclear facilities

; ; ; Takeda, Seiichiro

JNC TN8420 2001-008, 134 Pages, 2001/07

JNC-TN8420-2001-008.pdf:4.4MB

This investigation was conducted as a part of planning the low-level radioactive waste management program (LLW management program). The aim of this investigation was contributed to compile the radioactive waste database of JNC's LLW management program. All nuclear facilities of the Tokai works and Ningyo-toge Environmental Engineering Center were investigated in this work. The wastes generated by the decomissioning of each nuclear facility were classified into radioactive waste and others (exempt waste and non-radioactive waste), and the amount of the wastes was estimated. The estimated amounts of radioactive wastes generated by decomissioning of the nuclear facilities are as follows. (1)Tokai works. The amount of waste generated by decommissioning of nuclear facilities of the Tokai works is about 1,079,100 ton. The amount of radioactive waste is about 15,400 ton. The amount of exempt waste and non-radioactive waste is about 1,063,700 ton. (2)Ningyo-toge Environmental Engineering Center. The amount of waste generated by decommissioning of nuclear facilities of Ningyo-toge Environmental Engineering Center is about 112,500 ton. The amount of radioactive waste is about 7,800 ton. The amount of exempt waste and non-radioactive waste is about 104,700 ton.

JAEA Reports

None

; Inagaki, Tatsutoshi*

JNC TY1400 2000-004, 464 Pages, 2000/08

JNC-TY1400-2000-004.pdf:19.55MB

None

JAEA Reports

None

JNC TN1440 2000-005, 214 Pages, 2000/08

JNC-TN1440-2000-005.pdf:13.81MB

no abstracts in English

JAEA Reports

Investigation of molten salt fast breeder reactor

; ; ; ;

JNC TN9400 2000-066, 52 Pages, 2000/06

JNC-TN9400-2000-066.pdf:1.82MB

Phase I of feasibility studies on commercialized fast reactor system is being peformed for two years from Japanese Fiscal Year 1999. In this report, results of the study on fluid fuel reactors (especialiy a molten salt fast breeder reactor concept) are described from the viewpoint of technical and economical concerns of the plant system design. ln JFY1999, we have started to investigate the fluid fuel reactors as alternative concepts of sodium cooled FBR systems with MOX fuel, and selected the unique concept of a molten chloride fast, breeder reactor, whose U-Pu fuel cycle can be related to both light water reactors and fast breeder reactors on the basis of present technical data and design experiences. We selected a preliminary composition of molten fuel and conceptual plant design through evaluation of technical and economical issues essential for the molten salt reactors and then compared them with reference design concepts of sodium cooled FBR systems under limited information on the molten chloride fast breeder reactors. The following results were obtained. (1)The molten chloride fast breeder reactors have inherent safety features in the core and plant performances, ad the fluid fuel is quite promising for cost reduction of the fuel fabrication and reprocessing. (2)On the other hand, the inventory of the molten chloride fuel becomes high and thermal conductivity of the coolant is inferior compared to those of sodium cooled FBR systems, then, the size of main components such as lHX's becomes larger and the amount of construction materials is seems to be increased. (3)Furthermore economical vessel and piping materials which contact with the molten chloride salts are required to be developed. From the results, it is concluded that further steps to investigate the molten chloride fast breeder reactor concepts are too early to be conducted.

JAEA Reports

Expansion of material balance analysis function on nuclear fuel cycle

Ohtaki, Akira; ; ; *; *;

JNC TN9410 2000-006, 74 Pages, 2000/04

JNC-TN9410-2000-006.pdf:3.01MB

To evaluate materials balance in nuclear fuel cycle quickly and quantitatively the fuel cycle requirement code "FAMILY" was improved. And an accumulated TRU&LLFP quantity analysis code was developed. The contents are as follows: (1)A calculation ability of minor actinide production and expenditure was added to the "FAMILY" code. (2)An output program for the "FAMILY" calculation results was developed. (3)A simple version of "FAMILY" code was developed. (4)An analysis code for accumulated TRU&LLFP quantity in nuclear fuel cycle was developed.

JAEA Reports

Evaluation of cost reduction method for manufacturing ODS Ferritic claddings

Fujiwara, Masayuki; Mizuta, Shunji;

JNC TN9400 2000-050, 19 Pages, 2000/04

JNC-TN9400-2000-050.pdf:0.82MB

For evaluating the fast reactor system technology, it is important to evaluate the practical feasibility of ODS ferritic cdaddings, which is the most promising matelials to attain the goal of high coolant temperature and more than 150 GWd/t. Based on the results of their technology development, mass production process with highly economically benefit as well as manufacturing cost estimation of ODS ferritic claddings were preliminarily conducted. From the view point of future utility scale, the cost for manufacturig mother tubes has a dominant factor in the total manufacturing cost. The method to reduce the cost of mother tube manufacturing was also preliminarily investigated.

JAEA Reports

lnvestigation for corrosion behavior of core materials in lead cooled reactor

Kaito, Takeji

JNC TN9400 2000-039, 19 Pages, 2000/03

JNC-TN9400-2000-039.pdf:0.66MB

The corrosion behavior of core materials in lead cooled reactor was investigated as the feasibility study for fast breeder reactor. The results are summarized as follows. (1)The corrosion of stainless steels under lead and lithium occurs mainly due to the dissolution of nickel. Consequently ferritic stainless steels have better resistance to corrosion under lead and lithium than austenitic stainless steels, and the corrosion resistance of high nickel steels is worst. (2)The dissolution rate, D(mg/m $^{2}$ /h), is correlated with lead and lithium temperature, T(K), as log $_{10}$ Da = 10.7873 - 6459.3/ T and log $_{10}$ Df = 7.6185 - 4848.4/T, where D a is the dissolution rate for austenitic steels and D f is for ferritic steels. lt's possible to calculate the corrosion thickness, C(m), using the following correlation: C = (Dt)/10 $^{-3}$ , where t is exposure time(hr) and is density of the core matelial (g/cm $^{3}$ ). (3)The corrosion thickness estimated for austenitic steels using above correlations was extremely larger than ferritic steels, about 6 times at 400 $^{circ}$ C and more than 20 times at above 600 $^{circ}$ C. lt's considered that applicable temperature in lead cooled reactor core is below 400 $^{circ}$ C (about 60m corrosion thickness after 30000 hr) for austenitic steels, and below 500 $^{circ}$ C (about 80 m after 30000 hr) for ferritic steels.