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JAEA Reports

Irradiation tests report of the 34th cycle in "JOYO"

*

JNC-TN9440 2000-005, 164 Pages, 2000/06

JNC-TN9440-2000-005.pdf:4.51MB

This report summarizes the operating and irradiation data of the experimental reactor "JOYO" 34th cycle, and estimates the 35th cycle irradiation condition. Irradiation tests in the 34th cycle are as follows: (1)C-type irradiation rig (C4F) (a)High burnup perfomance test of advanced austenitic stainless steel cladding fuel pins (in collaboration with France) (2)C-type irradiation rig (C6D) (a)Large diameter fuel pins irradiation tests (3)Absorber Materials Irradiation Rig (AMIR-6) (a)Run to absorber pin's cladding breach (4)Core Materials Irradiation Rig (CMIR-5) (a)Cladding tube materials irradiation tests for "MONJU" (5)Structure Materials Irradiation Rigs (SMIR) (a)Decision of material design base standard of structure materials for prototype reactor and large reactor (6)Upper core structure irradiation Plug Rig (UPR-1-5) (a)Upper core neutron spectrum effect and accelerated irradiation effect (7)SurVeillance un-instrument Irradiation Rig (SVIR) (a)Confirmation of surveillance irradiation condition for "JOYO" (b)Material irradiation tests (in collaboration with universities) The maximum burnup driver assembly "PFD537" reached 68,500MWd/t(pin average).

JAEA Reports

Research on development of high-purity iron-based alloys; Manufacture, analysis of small amount of element and property tests

; *; ; ; Aoto, Kazumi;

JNC-TN9400 2000-059, 43 Pages, 2000/05

JNC-TN9400-2000-059.pdf:2.08MB

The purpose of this study is to understand the material properties of manufacturable high-purity iron and high-purity iron-based alloy in present technology and to get an applicable prospect for the structural and functional material of the frontier fast reactor. Then the about 10kg high-purity iron and iron-based alloy were melted using a cold-crucible induction melting furnace under the ultra-high vacuum. Subsequent to that, the compatibility between the melted material and the high-temperature sodium environment which is a special feature of the fast reactor and tensile property at room and elevated temperatures were investigated using the melted materials. Also, the creep test using the high-purity 50%Cr-Fe alloy at 550$$^{circ}$$C in air in order to understand the high temperature creep property. ln addition, the material properties such as thermal expansion coefficient, specific heat and electrical resistance were measured and to evaluate the outlook for the structural material for the fast reactor. The following results were obtained based on the property test and evaluation. (1)lt was possible to melt the about 10kg high-purity ingot and high-purity 50%Cr-Fe alloy ingot using a cold-crucible induction melting furnace under the ultra-high vacuum. (2)The tensile tests of the high-purity 50%Cr-Fe alloy were performed at room and elevated temperatures in order to understand the deformation behavior. From the experimental results, it was clear that the high-purity 50%Cr-Fe alloy possesses high strength and good ductility at elevated temperatures. (3)The physical properties (the thermal expansion coefficient and specific heat etc.) were measured using the high-purity 50%Cr-Fe alloy. lt was clear that the thermal expansion coefficient of high-purity 50%Cr-Fe alloy was smaller than that of SUS304. (4)From the corrosion test in liquid sodium, the ordinary-purity iron showed the weight loss after corrosion test. However the high-purity iron showed ...

JAEA Reports

The evaluation of material base standard of ODS ferritic stainless steel core component for fast breeder reactors

Mizuta, Shunji; ;

JNC-TN9400 2000-048, 28 Pages, 2000/04

JNC-TN9400-2000-048.pdf:0.64MB

ODS (Oxide Dispersion Strengthened) ferritic-martainsitic steels are one of the most prospective cladding materials for advanced fast breeder reactors, since they are expected to have excellent swelling resistance and superior high temperature strength due to the finely distributed stable oxide particles(Y$$_{2}$$O$$_{3}$$). Properties and the tentative strength equations for ODS ferritic-martainsitic were proposed on the basis of the latest data to apply to the feasibility study of the sodium coolant MOX fuel plant. The items of equations are follows. (1)creep rupture strength (2)correction factor of creep rupture strength (in Na and in reactor) (3)outer surface eorrosion (Na) (4)inner surface corrosion (in MOX fuel pin) (5)thermal conductivity

JAEA Reports

None

Koshizuka, Seiichi*; *; Okano, Yasushi; *; Yamaguchi, Akira

JNC-TY9400 2000-012, 91 Pages, 2000/03

JNC-TY9400-2000-012.pdf:2.82MB

no abstracts in English

JAEA Reports

None

*; *; ; Aoto, Kazumi

JNC-TY9400 2000-010, 138 Pages, 2000/03

JNC-TY9400-2000-010.pdf:5.15MB

None

JAEA Reports

None

*; *; *; Suzuki, Tatsuya*; *; *;

JNC-TY9400 2000-009, 41 Pages, 2000/03

JNC-TY9400-2000-009.pdf:1.22MB

no abstracts in English

JAEA Reports

None

*; *; *; *; Hasegawa, Makoto;

JNC-TY9400 2000-007, 50 Pages, 2000/03

JNC-TY9400-2000-007.pdf:1.29MB

no abstracts in English

JAEA Reports

Investigation of the properties of high temperature resistance alloys used in the helium gas cooled high temperature reactor

Uwaba, Tomoyuki

JNC-TN9420 2000-005, 28 Pages, 2000/03

JNC-TN9420-2000-005.pdf:0.94MB

In the first phase of the feasibility study, their basic objectives are presentating the feasible image and scenario of development of the FBR cycle system, which is composed of the fast reactor, spent fuel reprocessing and fuel manufacturing facility. In the development of the FBR system in this phase, various ideas of plants are to be studied, which include coolant types such as sodium, heavy metals, gases(CO$$_{2}$$, He), wator, and middle or small size of the reactor, and fuel types (MOX, metal and nitride). In this report, as a part of this study, materials used for the core of the helium gas cooled reactor and their integrity (corrosion, mechanical and irradiation property) under high temperature helium atmosphere were investigated from open literatures.

JAEA Reports

Study on sodium coolant loop-type reactor; Parametric study on maximum thermal stress depending on routing dimension of piping system

Tsukimori, Kazuyuki; Furuhashi, Ichiro*

JNC-TN9400 2000-049, 93 Pages, 2000/03

JNC-TN9400-2000-049.pdf:2.82MB

lt is one of the important key points to reduce thermal stress of the primary piping system in the design of sodium coolant loop-type FBR plants. The objectives of this study are to understand the characteristics of the thermal stresses in the simple S-shaped hot leg piping systems which run from the outlet nozzle of the reactor vessel (R/V) to the inlet nozzle of the intermediate heat exchanger (IHX), and to propose some recommendable routings of piping systems. Results are summarized as follows. (1)Generally, the thermal stresses in elbows are severer than those at nozzles. The tendency was observed that the stress in elbow decreases with the increase of the distance between the outlet nozzle of R/V and the inlet nozzle of IHX and also the distance between the outlet nozzle of R/V and the liquid surface level. (2)lt is expected to reduce thermal stresses in elbow to big extent by adopting super 90 degree elbows. Therefore, in these cases the dimension region which satisfies the allowable stress is broad compared with that in the case of the conventional 90 degree elbow. (3)The stress estimations in elbow based on 'MITl notice No.501' become excessively large compared with the results by FEA using shell elements, when the maximum stress occurs at the end of elbow. ln these cases, the estimation can be rationalized by replacing the maximum stress by the mean of stresses at the end and at the middle of the elbow. (4)Two routings with 105 degree elbows are recommended. 0ne has the advantage from the view point of reduction of length of pipe and the other does from the view point of reduction of thermal stresses, compared with the routing with 90 degree elbows.

JAEA Reports

Analysis of weld residual stresses by FINAS (1)

*;

JNC-TN9400 2000-047, 114 Pages, 2000/03

JNC-TN9400-2000-047.pdf:8.25MB

Prediction of weld residual stresses by a general finite element code is beneficial to the improvement of the accuracy of integrity assessment and residual life assessment of FBR plants. This reports develops an evaluation method of weld residual stresses using FINAS. Firstly, we suggested a basic procedure derived from parametric analyses with a simple weld joint model. The procedure can be summarized as follows: (1)For heat conduction analysis, prepare different models corresponding to the number of layers to be modeled. Hand over the analytical results to the following model. (2)Use multi-linear stress-strain curves for modeling the stress-strain response of base metal and weld metal. Use the isotropic hardening rule. (3)When metals are melt, use a user-subroutine to keep stresses from arising. (4)Put the thermal expansion coefficient as zero when heat is being input. Then, using the above procedure and TIG welding, we predicted the weld residual stresses of plate and tube. The results agreed well with the other reports, showing the suggested procedure was reasonable.

JAEA Reports

None

*; *; *; *

JNC-TJ3410 2000-021, 73 Pages, 2000/03

JNC-TJ3410-2000-021.pdf:52.78MB

no abstracts in English

JAEA Reports

None

*; *; *; *

JNC-TJ3410 2000-020, 80 Pages, 2000/03

JNC-TJ3410-2000-020.pdf:41.34MB

no abstracts in English

JAEA Reports

Sodium combustion computer code ASSCOPS Version 2.1; User's manual

Ohno, Shuji; Matsuki, Takuo*; ;

JNC-TN9520 2000-001, 196 Pages, 2000/01

JNC-TN9520-2000-001.pdf:5.13MB

ASSCOPS (Analysis of Simultaneous Sodium Combustion in Pool and Spray) has been developed for analyses of thermal consequences of sodium leak and fire accidents in LMFBRs. This report presents a description of the computational models, input and output data as the user's manual of ASSCOPS version 2.1. ASSCOPS is an integrated computational code based on the sodium pool fire code SOFIRE II developed by the Atomics International Division of Rockwell International, and on the sodium spray fire code SPRAY developed by the Hanford Engineering Development Laboratory in the U.S. The users of ASSCOPS need to specify the sodium leak conditions (leak flow rate and temperature, etc.), the cell geometries (cell volume, surface area and thickness of structures, etc.), and the atmospheric initial conditions such as gas temperature, pressure, and composition. ASSCOPS calculates the time histories of atmospheric temperature, pressure and of structural temperature.

JAEA Reports

Use results of MOX fuel in ATR Fugen nuclear power station

Ijima, Takashi; ; Matsumoto, Mitsuo; *

JNC-TN3410 2000-002, 93 Pages, 2000/01

JNC-TN3410-2000-002.pdf:2.54MB

Fugen Nuclear Power Station ("Fugen") is a prototype Advanced Thermal Reactor (ATR), it has been demonstrated the plutonium utilization by loading many Mixed Oxide Fuels (MOX) since the reactor start up March 1979, and no fuel defect had been occurred, The MOX fuel assemblies has the high reliability and has been loaded more than 700 fuel assemblies. This is the largest in the world as a thermal neutron reactor. However, "Fugen" is planning to stop its operation in the year 2003, because the role of the Fugen almost finished. Therefore, we are going to summarize the ATR project including the Plutonium utilization experience. This paper is summarized as part of the experience.

JAEA Reports

Material test data of SUS304 welded joints

; *

JNC-TN9450 2000-002, 335 Pages, 1999/10

JNC-TN9450-2000-002.pdf:21.65MB

This report summarizes the material test dala of SUS304 welded joints. Numbers of the data are as follows: [Tensile tests 71 (Post-irradiation: 39, others: 32) [Creep tests 77 (Post-irradiation: 20, others: 57) [Fatigue tests 50 (Post-irradiation: 0) [Creep-fatigue tests 14 (Post-irradiation: 0) This report consists of the printouts from "the structural material data processing system".

JAEA Reports

Material test data of SUS304

; *

JNC-TN9450 2000-001, 1370 Pages, 1999/10

JNC-TN9450-2000-001.pdf:117.18MB

This report summarizes the material test data of SUS304. Numbers of the data are as follows. (1)Tensile tests 738 (Post-irradiation: 250, others: 488) (2)Creep tests 434 (Post-irradiation: 89, others: 345) (3)Fatigue tests 612 (Post-irradiation: 60, others: 552) (4)Creep-fatigue tests 200 (Post-irradiation: 40, others: 160) This report consists of the printouts from "the structural material data processing system".

JAEA Reports

Irradiation creep equation of the advanced austenitic stainless steels

Mizuta, Shunji; ;

JNC-TN9400 99-082, 60 Pages, 1999/10

JNC-TN9400-99-082.pdf:1.52MB

The density measurement of the internal creep specimens irradiated in FFTF/MOTA (Fast Flux Test Facility / Material open Test Assembly) was conducted MMF (Materia1 Monitoring Facility) and accurate separation of swelling strain from total strain leaded in the derivation of the irradiation creep coefficients. Irradiation creep coefficients for PNC 316, 15Cr-20Ni base S.S. and 14Cr-25Ni base S.S. were systematically expressed, while thermal creep coefficients K, under irradiation were separately expressed for above three steels. The results obtained are follows, (1)The effect of stress induced swelling was recognized in the temperature range from 405 to 605$$^{circ}$$C. The swelling in high stress specimens have a tendency to increasing swelling. (2)The irradiation creep coefficients derived from PNC316 and l5Cr-20Ni are similar to that of derived from 20%CW316S.S., CW316Ti and CW15-15Ti which were reported by other authors. (3)The irradiation creep coefficient derived from gas pressurized tube irradiation using FFTF/MOTA expressed appropriately irradiation creep strain from fuel pins using FFTF/MFA-2(15Cr-2ONi base S.S.).

JAEA Reports

None

Arii, Yoshio

JNC-TN9200 99-009, 432 Pages, 1999/07

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

None

JAEA Reports

A quantitative evaluation of seismic margin of typical sodium piping

JNC-TN9400 99-041, 187 Pages, 1999/05

JNC-TN9400-99-041.pdf:4.62MB

lt is widely recognized that the current seismic design methods for piping involve a large amount of safety margin. From this viewpoint, a series of seismic analyses and evaluations with various design codes were made on typical LMFBR main sodium piping systems. Actual capability against seismic loads were also estimated on the piping systems. Margins contained in the current codes were quantified based on these results, and potential benefits and impacts to the piping seismic design were assessed on possible mitigation of the current code allowables. From the study, the following points were clarified; (1)A combination of inelastic time history analysis and true(without margin) strength capability allows several to twenty times as large seismic load compared with the allowable load with the current methods. (2)The new rule of the ASME is relatively compatible with the results of inelastic analysis evaluation. Hence, this new rule might be a goal for the mitigation of seismic design rule. (3)With this mitigation, seismic design accommodation such as equipping with a large number of seismic supports may become unnecessary.

JAEA Reports

Simulation of creep test on 316FR stainless steel in sodium environment at 550$$^{circ}C$$

Satmoko, A.*;

JNC-TN9400 99-035, 37 Pages, 1999/04

JNC-TN9400-99-035.pdf:1.54MB

In sodium environment, materia1 316FR stainless steel risks to suffer from carburization. In this study, an analysis using a Fortran program is conducted to evaluate the carbon influence on the creep behavior of 316FR based on experimental results from uni-axial creep test that had been performed at temperature 550$$^{circ}$$C in sodium environment simulating Fast Breeder Reactor condition. As performed in experiments, two parts are distinguished. At first, elastic-plastic behavior is used to simulate the fact that just before the beginning of creep test, specimen suffers from load or stress much higher than initial yield stress. In second part, creep condition occurs in which the applied load is kept constant. The plastic component should be included, since stresses increase due to section area reduction. For this reason, elastic-plastic-creep behavior is considered. Through time carbon penetration occurs and its concentration is evaluated empirically. This carburization phenomena are assumed to affect in increasing yield stress, decreasing creep strain rate, and increasing creep rupture strength of material. The model is capable of simulating creep test in sodium environment. Material near from surface risks to be carburized. Its material properties change leading to non-uniform distribution of stresses. Those layers of material suffer from stress concentration, and are subject to damage. By introducing a damage criteria, crack initialization can thus be predicted. And even, crack growth can be evaluated. For high stress levels, tensile strength criterion is more important than creep damage criterion. But in low stress levels, the latter gives more influence in fracture. Under high stress, time to rupture of a specimen in sodium environment is shorter than in air. But for stresses lower than 26 kgfmm$$^{2}$$, the time to rupture of creep in sodium environment is the same or little longer than in air. Quantitatively, the carburization effect at ...

64 (Records 1-20 displayed on this page)