Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 21

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

JAEA Reports

None

JNC-TN1400 2000-012, 250 Pages, 2000/11

JNC-TN1400-2000-012.pdf:10.18MB

no abstracts in English

JAEA Reports

None

JNC-TN1400 2000-010, 70 Pages, 2000/10

JNC-TN1400-2000-010.pdf:2.87MB

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

Development and validation of Multi-DimensionaI sodium combustion analysis code AQUA-SF

Takata, Takashi; Yamaguchi, Akira

JNC-TN9400 2000-065, 152 Pages, 2000/06

JNC-TN9400-2000-065.pdf:6.26MB
JNC-TN9400-2000-065(errata).pdf:0.12MB

ln the liquid metal fast reactor (LMFR) using liquid sodium as a coolant, it is important to evaluate the effect of the sodium combustion on the structure, etc. Most of the previous analytical works are based on a zone model, in which the principal variables are treated as volume-average quantities. Therefore spatial distribution of gas and structure temperatures, chemical species concentration are neglected. Therefore, a multi-dimensional sodium combustion analysis code AQUA-SF (Advanced simulation using Quadratic Upstream differencing Algorithm - Sodium Fire version) has been developed for the purpose of analyzing the sodium combustion phenomenon considering the multi-dimensional effect. This code is based on a multi-dimensional thermal hydraulics code AQUA that employs SIMPLEST-ANL method. Sodium combustion models are coupled with AQUA; one is a liquid droplet model for spray combustion, and the other is a flame sheet model for pool combustion. A gas radiation model is added for radiation heat transfer. Some other models necessary for the sodium combustion analysis, such as a chemical species transfer, a compressibility, are also added. ln AQUA-SF code, bounded QUICK method in space scheme and bounded three-point implicit method in time scheme are implemented. Verification analyses of sodium combustion tests shown in the following have been carried out. (1)pool combustion test (RUN-D1) (2)spray combustion test (RUN-E1) (3)sodium leakage combustion test (Sodium Fire Test-II) (4)smaII-scale leakage combustion test (RUN,F7-1) ln each verification analysis, good agreements are obtained and the validity of AQUA-SF code is confirmed.

JAEA Reports

lnvestigation of thermal-hydraulic issues resulting in the use of various coolants

; Yamaguchi, Akira

JNC-TN9400 2000-056, 150 Pages, 2000/05

JNC-TN9400-2000-056.pdf:6.67MB

[Purpose] The work was performed to make clear thermal-hydraulic issues resulting in the use of various coolants for fast reactors. [Method] Plant design features due to a use of working fluid other than sodium and design concepts relating a simplification of safety related systems were investigated. And based on the results, quantitative evaluation was made on the topical themal-hydraulic issues. Then both thermal stratification and striping phenomena were evaluated by the used of thermo-hydraulics computer programs. [Results] (1)Thermal-hydraulic issues Topical thermal-hydraulic issues of gaseous and heavy metal cooled reactors were extracted. (a)Gas cooled reactors : natural circulation,flow-induced vibration, depressurization accident (b)Heavy metal cooled reactors : thermal stratification, flow-induced vibration, sloshing And also the thermal-hydraulic issues relating compact reactor assembly and RVACS were extracted resulting from a simplification of safety related systems. (2)Evaluation of thermal stratification and striping phenomena. The following order of affects for the phenomena was obtained: (a) Thermal stratification: CO$$_{2}$$ $$<$$ Sodium $$<$$ Lead, (b) Thermal Striping: CO$$_{2}$$ $$<$$ Lead $$<$$ Sodium

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

Examination of safety design guideline; Safety objective and elimination of re-criticality issues

; ; *;

JNC-TN9400 2000-043, 23 Pages, 2000/03

JNC-TN9400-2000-043.pdf:1.1MB

ln the feasibility study on commercialized fast breeder reactor (FBR) cycle systems conducted in JNC, it is required for candidate FBR plants that the level of safety should be enhanced so as to assure: (1)Comparative or superior safety level to that of light water reactors (LWRs), and (2)releaf of the public from anxiety about potential nuclear hazard. Adopting Passive safety characteristics is one of the measures. To attain the above safety objective, we considered implication of the basic safety principles for nuclear power plants that were created by the international nuclear safety advisory group of IAEA. The way to relieve from the anxiety was also taken into account. Then a definite safety objective was set from the standpoint of prevention of core disruptive accident (CDA). Furthermore, as a definite safety goal relating to reactor coresafety, elimination of re-criticality issues under CDA was set by considering characteristics of FBR in comparison with those of LWR. To examine measures for elimination of re-criticality issues, we developed a quick method to estimate possibility of re-criticality under CDA, by drawing a map about criticality characteristics under CDA in various degraded cores. Then hopeful measures were proposed for elimination of re-criticality issues in sodium-cooled FBR with mixed-oxide fuel. Molten fuel discharge behavior of their measures was preliminarily analyzed. We concluded that discharge capability of "a subassembly with an internal duct" was effective, and that "partial removal of axial blanket" was also effective as one of the measures though it has small effect on core performance.

JAEA Reports

Comparative study for minor actinide transmutation in various fast reactor core concepts (1)

JNC-TN9400 2000-007, 77 Pages, 1999/12

JNC-TN9400-2000-007.pdf:2.17MB

Comparative study for various core concepts is being carried out in a frame work of the study for minor actinide (MA) transmutation using a fast reactor. Different fuel types (Oxide, Nitride, Metal) and coolants (Sodium, Lead) were investigated. It is found that neither nitride nor metal-fueled core has significantly more excellent efficiency for MA transmutation comparing with an oxide-fueled core when the basic performance of these cores as a power reactor are fixed. The MA transmutation Properties of lead-cooled fast reaetor (BREST-300) and sodium-cooled fast reactor (3800MWth large core) were compared. The sodium-cooled reactor surpasses BREST-300 on the MA transmutation rate. Meanwhile, it is found that BREST-300 is excellent from the viewpoint of loading much more MA in the core to attain larger MA transmutation amount. The effect of MA to coolant void reactivity is considered by the sensitivity analysis. It is found that the lead void reactivity has different sensible energy regions on MA nuclides from those for the sodium void reactivity.

JAEA Reports

None

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

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

no abstracts in English

JAEA Reports

None

Arii, Yoshio

JNC-TN9200 99-009, 432 Pages, 1999/07

JNC-TN9200-99-009.pdf:17.27MB

None

JAEA Reports

Thermal-Hydraulic investigation on severaI fast reactor design concepts

Ohshima, Hiroyuki; ; ; Yamaguchi, Akira; Nishi, Yoshihisa*; Ueda, Nobuyuki*; *

JNC-TN9400 2000-077, 223 Pages, 1999/05

JNC-TN9400-2000-077.pdf:6.24MB

The feasibility study (Phase l) is being carried out at JNC to build up new design concepts of practical fast reactors (FRs) from the viewpoint of economy, safety, effective use of resources, reduction of environmental burden and non-proliferation. This report describes the results of the investigation, related to decay heat removal, core/fuel-assembly thermal-hydraulics and thermal-hydraulic correlations, that was performed in fiscal l999 as a part of the feasibility study. ln the study of the decay heat removal, the effects of several design parameters on the performance of the reactor vessel auxiliary cooling system (RVACS) in a middle-scale sodium-cooled FR were clarified by using a plant dynamic analysis code. The upper limit of RVACS performance was preliminarily estimated at approximately 0.5$$sim$$0.6 MWe. Numerical methods for the plant dynamic analysis of gas-and heavy-metal-cooled FRs were also developed. They were applied to the preliminary calculations of the transition from scram to natural circulation and the transient characteristics in tentative plant design concepts were clarified. ln addition, a dimensionless number indicating natural circulation performance was deduced for the comparison of several plant design concepts. With respect to the core/fuel-assembly thermal-hydraulics, numerical analysis methods were improved for the pin-type fuel assembly of gas-and heavy-metal-cooled FRs, the coated-particle- type fuel assembly of helium-gas-cooled FR, and the ductless core of sodium-and heavy-metal-cooled FRs. As preliminary evaluations, thermal-hydraulics in the heavy-metal-cooled FR fuel assembly was compared with sodium-cooled one and thermal-hydraulic analyses of carbon-dioxide- and helium-gas-cooled FR fuel assemblies were performed. The analysis for the fuel assembly with inside duct of sodium-cooled FR was also carried out. The correlations of pressure loss and heat transfer coefficient were investigated for the thermal-hydraulic ...

JAEA Reports

PNC Technical Review No.107

PNC-TN1340 98-003, 126 Pages, 1998/09

PNC-TN1340-98-003.pdf:17.88MB

no abstracts in English

JAEA Reports

PNC Technical Review No.106

PNC-TN1340 98-002, 122 Pages, 1998/06

PNC-TN1340-98-002.pdf:18.1MB

no abstracts in English

JAEA Reports

None

PNC-TN1410 97-043, 167 Pages, 1997/11

PNC-TN1410-97-043.pdf:7.22MB

no abstracts in English

JAEA Reports

PNC Technical Review No.103

PNC-TN1340 97-003, 101 Pages, 1997/09

PNC-TN1340-97-003.pdf:12.06MB

no abstracts in English

JAEA Reports

PNC Technical Review No.101

PNC-TN1340 97-001, 154 Pages, 1997/03

PNC-TN1340-97-001.pdf:21.55MB

no abstracts in English

JAEA Reports

Study of thermohydraulic behavior within the fuel bundle under a loss of flow condition

M.E.Kab*;

PNC-TN9410 92-018, 58 Pages, 1992/01

PNC-TN9410-92-018.pdf:1.31MB

This report describes the result of the analysis of unprotected Loss of Flow (LOF) ansient experiment conducted at the PLANt Dynamics Test Loop (PLANDTL) experimentalfility by Super System Code (SSC) and SubAssembly Boiling EvolutioN Analysis (SABENA)ode. This report also describes the effect of the modification we made in SSC with t recent void fraction and two-phase friction multiplier models during the analysis othe experiment. After the analysis, it was found that the two-fluid two-phase flow mel of SABENA 1-D is better than the homogeneous model of SSC in predictiong the therhydraulic behavior within the simulated fuel bundle test section of thePLANDTL facily in case of high quality sodium boiling experiment. Moreover, it wasalso revealed tt the two-fluid one dimensional model is not accurate enough in predicting the onsetf boiling and axial evolution of boiling region inside the heatedchannel.

JAEA Reports

DEVELOPMENT OF CERAMIC LINER FOR FBR BUILDING

; *; Kawada, Koji

PNC-TN9410 91-092, 11 Pages, 1991/01

PNC-TN9410-91-092.pdf:1.53MB

TO DEVELOP A CERAMIC LINER,A SELECTION TEST OF MATERIALS,AN IMPROVEMENT TEST OFSELECTED MATERIAL,AND A FEASIBILITY TEST OF THE LINER HAVE BEEN CONDUCTED.IN THE SELECTION TEST,FIFTY COMMERCIALLY AVAILABLE HIGH TEMPERATURE CEMENT ANDCERAMICS WERE SUBJECTED TO THERMAL SHOCK TEST(TST),SODIUM EXPOSURE TEST(SET),ANDSODIUM FLAME EXPOSURE TEST(SFET). FROM TEST RESULTS,ALUMINA/SILICON-CARBIDE(AL2O3-SIC)MIXTURE BASE CASTABLE REFRACTORY WAS SELECTED IN CONSIDERATION OF MATERIAL COST,AND MATERIAL AVAILABILITY FOR A SIMPLER LINER CONSTRUCTION IN THE BUILDINGS.THE SELECTED MATERIAL WAS SUBJECTED TO THE IMPROVEMENT TEST. FROM THE TEST,PROPERWEIGHT FRACTIONS OF ADDITIVES SUCH AS ALUMINA CEMENT AND SILICA WERE DETERMINED.DRYING CONDITIONS WERE ALSO DETERMINED.FINALLY,A SODIUM BURNING PAN MADE OF CONCRETE WHOSE INNER SURFACES WERECOVERED WITH THE IMPROVED AL2O3-SIC BASE CASTABLE REFRACTORY WAS FABRICATED AND WASUSED FOR A SODIUM BURNING TEST.

JAEA Reports

Experimental Study on Equilibrium Partition Coefficient of Volatile Fission Products between Liquid Sodium and the Gas Phase

; ; ; ;

PNC-TN9410 91-091, 13 Pages, 1991/01

PNC-TN9410-91-091.pdf:0.31MB

A series of tests has been conducted to obtain gas-liquid equilibrium partition coefficient Kd of volatile fission products such as cesium,iodine,and tellurium in sodium. In the test a sodium pool mixed with an FP simulant was heated by an electric furnace and the solvent of trapped vapors by filters was quantitatively analyzed. The results are,(1)Cs shows the highest Kd (20-100), (2)Kd of iodine scatters as wide as 0.02-0.5at 450$$^{circ}C$$and 0.3-0.8 at 650$$^{circ}C$$,(3)the Kd values of Cs and I agree well with the theoretical ones reported by Castleman et al., and (4)if a sodium-telluride which is hard to vaporize than pure Te is assumed, measured Kd of Te agrees with that theoretical.

JAEA Reports

ANALYSIS OF LARGE LEAK SODIUM-WATER REACTION IN LARGE FBR

;

PNC-TN9410 91-028, 14 Pages, 1991/01

PNC-TN9410-91-028.pdf:0.36MB

A COMPUTER CODE,SWACS,WAS DEVELOPED TO ANALYZE A LARGE LEAK SODIUM-WATER REACTION EVENT IN AN LMFBR STEAM GENERATOR. THE JAPANESE PROTOTYPE REACTOR,MONJU,HAS A COVER GAS SPACE IN ITS STEAM GENERATOR BUT DIFFERENT DESIGNS ARE ALSOCONSIDERED FOR A FUTURE LARGER PLANT. THEREFORE,SWACS WAS MODIFIED TO ANALYZE THESODIUM-WATER REACTION EVENT UNDER SUCH VARIOUS DESIGNS. SO FAR THE CALCULATIONALMODULE OF AN INITIAL SPIKE PRESSURE AND ITS PROPAGATION TO IHTS WAS IMPROVED AND THERESULTS WERE COMPARED WITH THE DATA FROM LLTR AT ETEC, U.S.A. AND WATER-EXPLOSIVE SIMULATION TESTS AT PNC,JAPAN. THE COMPARISON REVEALED A FAIRLY GOODAGREEMENT BETWEEN THE TESTS AND THE ANALYSES. FOLLOWING THE VALIDATION STUDY,SWACS WAS USED FOR THE APPLICATION ANALYSIS TO COMPARE THE PRESSURE BEHAVIORBETWEEN THE COVER-GAS TYPE AND THE NO-COVER-GAS TYPE STEAM GENERATOR OF A FUTURELARGER PLANT. THE ANALYSIS CLARIFIED THE APPLICABILITY OF SWACS TO SUCH A DESIGN STUDYFROM A VIEWPOINT OF SUPPRESSING THE SWR PRESSURE.

JAEA Reports

Failure propagation analysis of LMFBR steam generator tube; Analysis of SWAT-3 runs 14 and 15 by LEAP II code

*; *; *; *

PNC-TN941 82-100, 48 Pages, 1982/04

PNC-TN941-82-100.pdf:0.81MB

The Computer code LEAP II had been developed in order to analize failure propagation phenomena by the sodium-water reaction in the steam generator of LMFBR. Here reported is verification analysis of the LEAP code by using Runs 14 and 15 test results of Steam Generator Safety Test Facility (SWAT-3). The main results are as follows: (1)As the results of parametric survey, the effects of the significant parameters such as time mesh, jet division number, etc. to the code were understood. (2)In comparison with the test results of Runs 14 and 15 of SWAT-3, the LEAP code can estimate the phenomena conservatively enough.

21 (Records 1-20 displayed on this page)