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JAEA Reports

Report for the Participation in GLOBAL2001

; ; Shigetome, Yoshiaki

JNC-TN8200 2001-006, 19 Pages, 2001/12

JNC-TN8200-2001-006.pdf:0.92MB

None

JAEA Reports

None

; Inagaki, Tatsutoshi*

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

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

None

JAEA Reports

Feasibility studies on commercialized fast breeder reactor cycle system (Phase I) interim report

; Inagaki, Tatsutoshi*

JNC-TY1400 2000-003, 92 Pages, 2000/08

JNC-TY1400-2000-003.pdf:3.9MB

Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power company (JAPCO, that is the representative of the electric utilities in Japan) have established a new organization to develop a commercialized fast breeder reactor (FBR) cycle system since July 1, 1999 and feasibility studies (F/S) have been undertaken in order to determine the promising concepts and to define the necessary R&D tasks. In the first two-year phase, a number of candidate concepts will be selected from various options, featuring innovative technologies. In the F/S, the options are evaluated and conceptual designs are examined considering the attainable perspectives for following: (1) ensuring safety, (2) economic competitiveness to future LWRs, (3) efficient utilization of resources, (4) reduction of environmental burden and (5) enhancement of nuclear non-proliferation. The F/S should also guide the necessary R&D to commercialize FBR cycle system.

JAEA Reports

None

*

JNC-TN1440 2000-007, 115 Pages, 2000/08

JNC-TN1440-2000-007.pdf:4.45MB

no abstracts in English

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

lrradiation behavior and performance model of nitride fuel

; ;

JNC-TN9400 2000-041, 29 Pages, 2000/03

JNC-TN9400-2000-041.pdf:1.18MB

Irradiation behavior and performance models were investigated in order to apply for nitride fuel options in feasibility study on fast breeder reactor and related recycle systems. (1)MechanicaI design of nitride fuel pin: The behaviors of fission gas release (increase of internal Pressure) and fuel-to-cladding chemical interaction (decrease of cladding thickness) are needed to evaluate cumulative damage fraction in case of fuel pin mechanical design. The behaviors of fission gas release and fuel-to-cladding chemical interaction were investigated from the past studies up to high burnuP, since the lower fission gas release in nitride fuel than in oxide fuel could contribute to reduce the plenum volume and result in the shortening of fuel Pin length. (2)Fuel pin smear density: The higher fuel smear density is preferred for the higher fissile density to improve the core characteristic. The behaviors of fuel pellet swelling were investigated from the past studies up to higher burnup, since the larger fuel pellet swelling in nitride fuel than in oxide fuel would restrict high burunp capability due to fuel-cladding mechanical interaction. (3)Compatibility of nitride fuel with high Temperature water: Compatibility of nitride fuel with high temperature water were investigated from the past studies to contribute water cooled fast breeder reactor options.

JAEA Reports

lnvestigation for corrosion behavior of ferritic core materials in C0$$_{2}$$ gas cooled reactor

; ; Mizuta, Shunji

JNC-TN9400 2000-040, 41 Pages, 2000/03

JNC-TN9400-2000-040.pdf:0.85MB

The corrosion behavior of ferritic stainless steels applied to core components under C0$$_{2}$$ gas environment was investigated in order to be helpful to fuel design in C0$$_{2}$$ gas cooled reactor as the feasibility study for fast breeder reactor. The dependence of the corrosion behavior, before a breakaway occurs, on C0$$_{2}$$ gas temperature, Si and Cr contents of ferritic steels was determined quantitatively. The following correlations to calculate the metal loss thickness was established. X = 4.4w w = √(k$$times$$t) k = $$alpha$$ $$times$$ exp( - 5.45[Si]) $$times$$ exp( - 1.09[Cr]) $$times$$ exp( - 11253/T) $$alpha$$ = 1.65 $$times$$ 10$$^{8}$$$$sim$$4.40 $$times$$ 10$$^{9}$$ X : metal loss thickness[$$mu$$ml, w : corrosion weight gain [mg/cm$$^{2}$$] k : parabola constant [(mg/cm$$^{2}$$)$$^{2}$$/hr], t : time [hr], $$alpha$$ : constant [Si] : Si content[wt.%], [Cr] : Cr content [wt.%], T : temperature [K]

JAEA Reports

Investigations on the evaluation methods of the irradiation performance of FBR metallic fuel for the design study

;

JNC-TN9400 2000-031, 15 Pages, 2000/03

JNC-TN9400-2000-031.pdf:0.53MB

For the irradiation performance of metallic fuel, many of the analyses were conducted in USA using EBR-l and EBR-II. ln this study, based on the published data and papers on the above results, the appropriate methods to the evaluation of the irradiation performance of FBR metallic fuel for the design study were considered, as the fbasibility study for FBR. The followings are the targets in this work; (1)deformation of cladding (2)deformation of fuel slug (3)FP gas release (4)fluctuation of the bonding Na level in the fuel pin (5)FCCI

JAEA Reports

None

JNC-TN1440 2000-003, 88 Pages, 1999/08

JNC-TN1440-2000-003.pdf:5.11MB

no abstracts in English

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