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Journal Articles

Research of Takasaki Advanced Radiation Research Institute

Tamada, Masao

Genshiryoku Nenkan 2016, p.148 - 152, 2015/10

Takasaki Advanced Radiation Research Institute (TARRI) was established in April 1963. There are irradiation facilities of quantum beam such as ion, electron, Co-60 $$gamma$$-rays. TARRI has researched as complementary utilization together with other quantum beams such as neutron, synchrotron radiation, etc. as a part of the quantum beam platform. Recent technology transfer are outcomes of mutation-breeding sake yeast for quality sake brewed from the finest rice and cesium removal adsorbent for cartridge-type filters. Radiation-induced crosslinked resin was commercialized recently as a school teaching material. Dissemination of radiation technology has been continued through outreach activities.

Journal Articles

Novel breeding technique using ion beams

Watanabe, Hiroshi; Tanaka, Atsushi

Genshiryoku eye, 49(5), p.22 - 25, 2003/05

no abstracts in English

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

JNC-TN1400 2001-014, 437 Pages, 2001/10

JNC-TN1400-2001-014.pdf:23.1MB

no abstracts in English

Journal Articles

Electron-beam decomposition of PCDD/F in flue gas from municipal waste incinerator

Hirota, Koichi

Isotope News, (566), p.9 - 11, 2001/07

Japan Atomic Energy Research Institute has started the research on Electron-beam technology for decomposing more than 90 % of PCDD/D from MSWI in October of 2000. This test facility is operated till March of 2002. Then, this technology will be transferred to private companies after the feasibility study is completed.

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

Studies on sodium cooled fast breeder reactor

Nibe, Nobuaki; Shimakawa, Yoshio; ; ; ; ;

JNC-TN9400 2000-074, 388 Pages, 2000/06

JNC-TN9400-2000-074.pdf:13.32MB

Large sized sodium-cooled fast breeder reactors of large-size are being studied and have been operated in Japan and many countries. ln this feasibility study, evaluation was made on technical feasibinty for design concepts or 1 loop type and 3 pool types, specially from the viewpoint of improvement of economical competence. The design concepts include the ideas of cost reduction measures such as large-scaled components, reduction of loop number and integration of components on the basic of utilization of sodium characteristics. From the results of the evaluation, it may be possible for all the concepts to attain the economical target of 200 thousands yen per kilowatt, though further confirmation should be made for technical feasibility of those concepts. ln addition, the following items were listed up as further cost-reduction measures. (1)Higher temperature cooling system and steam cycle efficiency (2)Shortening of construction term (3)Reduction of safety systems by using measuring instruments with high performmce (4)Adoption of SG-ACS

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

An Evaluation study of measures for prevention of Re-criticality in sodium-cooled large FBR with MOX fuel

JNC-TN9400 2000-038, 98 Pages, 2000/04

JNC-TN9400-2000-038.pdf:7.49MB

As an effort in the feasibility study on commercialized Fast Breeder Reactor cycle systems, an evaluation of the measures to prevent the energetic re-criticality in sodium-cooled large MOX core, which is one of the candidates for the commercialized reactor, has been performed. The core disruptive accident analysis of Demonstration FBR showed that the fuel compaction of the molten fuel by radial motion in a large molten core pool had a potential to drive the severe super-prompt re-criticality phenomena in ULOF sequence. ln order to prevent occurrence of the energetic re-criticality, a subassembly with an inner duct and the removal of a part of LAB are suggested based on CMR (Controlled Material Relocation) concept. The objective of this study is the comparison of the effectiveness of CMR among these measures by the analysis using SIMMER-III. The molten fuel in the subassembly with inner duct flows out faster than that from other measures. The subassembly with inner duct will work effectively in preventing energetic re-criticality. Though the molten fuel in the subassembly without a part of LAB flows out a little slower, it is still one of the promising measures. However, the UAB should be also removed from the same pin to prevent the fuel re-entries into the core region due to the pressurization by FCl below the core, unless it disturbs the core performance. The effect of the axial fuel length of the center pin to CMR behavior is small, compared to the effect of the existence of UAB.

Journal Articles

Present state and future view of significant metals-recovery from seawater

Tamada, Masao; Seko, Noriaki

Isotope News, p.2 - 6, 2000/04

no abstracts in English

JAEA Reports

MA transmutation in various fast reactor core concepts

; Iwai, Takehiko*; Jin, Tomoyuki*

JNC-TN9400 2000-080, 532 Pages, 2000/03

JNC-TN9400-2000-080.pdf:14.98MB

Transmutation Property of minor actinide nuclides (MA) was analyzed for fast reactor cores having different coolant and fuel types in order to obtain basic data for evaluating an ability of the efficient use of resource and reducing the effect on the environment. The investigated fast reactor cores were (a) sodium cooled and oxide fueled core, (b) lead cooled and nitride fueled core (BREST-300), (c) carbon dioxide gas cooled and oxide fueled core (ETGBR), (d)lead cooled and oxide fueled core, (e) sodium cooled and nitride fueled core (both He-bond and sodium-bond), and (f) sodium cooled and metallic fueled core. Followings were observed for the relation between MA transmutation property and the different types of coolant and fuel of fast reactor core. (1) For the MA transmutation rate, the relation "Oxide $$<$$ Metal $$<$$ Nitride" was found out for difference of fuel type. A main reason of the increment of MA transmutation rate is that the neutron flux level rises on nitride and metallic fueled cores in comparison with oxide core. (2) The relation "Lead $$<$$ Sodium and Carbon dioxide" can be seen for the MA transmutation rate in the difference of coolant, but it is not clear whether the cause is driven from the difference of coolant itself on the difference of core design. (3) The changes of MA transmutation property mentioned above are comparatively small.

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

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($$mu$$m), using the following correlation: C = (D$$times$$t)/$$rho$$$$times$$10$$^{-3}$$, where t is exposure time(hr) and $$rho$$ 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 60$$mu$$m corrosion thickness after 30000 hr) for austenitic steels, and below 500$$^{circ}$$C (about 80 $$mu$$m after 30000 hr) for ferritic steels.

JAEA Reports

The development of mass balance estimation code; The development and the analyzed example with object type code(I)

;

JNC-TN9400 2000-034, 48 Pages, 2000/03

JNC-TN9400-2000-034.pdf:1.56MB

The study and the development to put FBR (Fast Breeder Reactor) to practical use have been doing. So many kinds of technologies are investigated to construct nuclear fuel recycle received to the society. The most important aim of reprocessing has been to extract U and Pu from spent fuels effectively, but, now, the demands for reprocessing are many kinds on nuclear fuel recycle system's construction. These need to be accepted sufficiently. The system that consists of electrolysis, extraction, with molten salt and melting metal, volatilization and condensation using the difference of vapor pressure is suggested, because, differently from LWR (Light Water Reactor), FBR can use the low decontamination factor's fuel. When the engineering scale plant is designed, the dry reprocessing has unsolved problems(ex. process flow) because of less demonstrative scale plants of the dry reprocessing than ones of the wet reprocessing. So the analysis and the estimation of mass balance that is most fundamental in the dry reprocessing system's design need to keep up with the system's alteration (to add new processes etc.) flexibly. This study aim is to develop the mass balance estimation code of dry reprocessing that satisfies the demand mentioned above.

49 (Records 1-20 displayed on this page)