JNC-TN9440 2000-005, 164 Pages, 2000/06
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).
Fujiwara, Masayuki; Mizuta, Shunji;
JNC-TN9400 2000-050, 19 Pages, 2000/04
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.
Mizuta, Shunji; ;
JNC-TN9400 2000-048, 28 Pages, 2000/04
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(YO). 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
*; Miyamoto, Shuji*; *; *; *; Hasegawa, Makoto; Yamazaki, Yoshio
JNC-TY9400 2000-008, 20 Pages, 2000/03
no abstracts in English
*; *; *; *; Hasegawa, Makoto;
JNC-TY9400 2000-007, 50 Pages, 2000/03
no abstracts in English
JNC-TN9400 2000-047, 114 Pages, 2000/03
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.
Momma, Yoshio*; *; ; ; ; Aoto, Kazumi
JNC-TN9400 2000-044, 22 Pages, 2000/03
ln the past the microstructural observation was mostly applied to understand the materials behavior qualitatively in R&D of the new materials and the life prediction for the fast breeder reactor components. However, the correlation between the changes in properties and microstrutures must be clarified to ensure the structural integrity. Particularly we are interested in the method to correlate the long-term properties and microstructural changes at high temperatures. The current research is to quantify the changes in microstructure of the weld metal for the welded structure of the reactor vessel. ln this research we have conducted creep testing of the weld metals at 823 and 873K up to 37,000h. Two types of the weld metals (16Cr-8Ni-2Mo and 18Cr-12Ni-Mo) were subjected to the creep testing. Based on the areas of the precipitates, the microstructural characterization with time and creep damage was attempted. The creep strength of the 16Cr-8Ni-2Mo weld metal is lower than that of the 18Cr-12Ni-Mo one at higher stresses, shorter times. But there is a trend toward to become similar strength with lower stresses and increasing times. The creep-rupture ductility of the 16Cr-8Ni-2Mo weld metal is superior to that of the 18Cr-12Ni-Mo one. The creep-rupture takes place at the interface of the sigma () phases precipitated in the delta () ferrites at 823K lower stresses and 873K. The amount of precipitates in the 16Cr-8Ni-2Mo weld metal is smaller than that in the 18Cr-12Ni-Mo one at each temperature and stress. Also it is apparent that the amount of the precipitates is primarily responsible to the decomposition of the phase, because the amount of the residual ferrites measured by the Magne-Gauge reduces with times. Using the Larson-Miller parameter it was possible to correlate the amount of the precipitates linearly with the LMP values.
; Yoshida, Eiichi; Aoto, Kazumi
JNC-TN9400 2000-042, 112 Pages, 2000/03
A sodium-water reaction drove from the single tube break in steam generator of FBR might overheat labor tubes rapidly under internal pressure loadings. lf the temperature of tube wall becomes too high, it has to be evaluated that the stress of tube does not exceed the material strength limit to prevent the propagation of tube rupture. This study clarified the tensile and creep properties of Mod.9Cr-1Mo steel at ultra-high temperature which will be used in evaluation of the tube burst by sodium-water reaction. The strain rates for tensile test are from 10%/min to 10%/sec, and creep-rupture time is maximum 277sec. The range of test temperature is 700C to 1300C. The main results obtained were as follows; (1)The evaluation data on the relationship between tensile strength and strain rate and creep-rupture strength in shorter time on Mod.9Cr-1Mo steel were acquired. (2)Short-term mechanical properties of Mod.9Cr-1Mo steel were evaluated based on the results of tensile and creep-rupture tests up to 1300C. As a result of the evaluation, recommended equation of creep-rupture strength in the short-term was proposed. (3)Tensile and creep-rupture strength of Mod.9Cr-1Mo steel tube showed the value which was higher than the 2 1/4Cr-1Mo steel, and it was proven to have the superior properties.
; ; Mizuta, Shunji
JNC-TN9400 2000-040, 41 Pages, 2000/03
The corrosion behavior of ferritic stainless steels applied to core components under C0 gas environment was investigated in order to be helpful to fuel design in C0 gas cooled reactor as the feasibility study for fast breeder reactor. The dependence of the corrosion behavior, before a breakaway occurs, on C0 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 = √(kt) k = exp( - 5.45[Si]) exp( - 1.09[Cr]) exp( - 11253/T) = 1.65 104.40 10 X : metal loss thickness[ml, w : corrosion weight gain [mg/cm] k : parabola constant [(mg/cm)/hr], t : time [hr], : constant [Si] : Si content[wt.%], [Cr] : Cr content [wt.%], T : temperature [K]
; Aoto, Kazumi; ;
JNC-TN9400 2000-022, 46 Pages, 2000/03
ln this report, a study on the behaviors of the magnetization induced by mechanical damages is carried out. By introducing mechanical damages to a test-piece with a tension or/and a zero, tension fatigue testing and measuring the corresponding leakage flux signal, natural magnetization change is proved and found increasing with the mechanical damages (viz. plastic deformation or fatigue damages) though a saturation occurs when damage gets too large. From the experimental results of fatigue testing utilizing test-pieces with a central slit, it was verified that observing the natural leakage flux density (leakage flux without applying external magnetic field) is a reasonable way to identify fatigue cracks. A feature parameter (area of the B hysteresis curve) of the in-situ magnetic field signal measured during the fatigue testing is proposed for predicting the fatigue damages, which is found depending on the cyclic number of the applied loading. At last, residual magnetic fields of a magnetized test-piece are also measured and found depending on the applied plastic deformation in case that the plastic strain is not too small. From these experimental results, it is found that the approach detecting natural magnetization is applicable for monitoring the damage status though it may be not efficient for a scanning inspection concerning its small signal magnitude. On the other hand, the method employing permanent magnet is robust against the environment noise but possibly not valid for the ISl of a structural component with a relative low damage level. For practical application, efforts to evaluate the feasibility of the proposed method are necessary for more testing conditions especially its suitability in a practical environment.
JNC-TN9440 2000-002, 157 Pages, 2000/02
This report summarizes the operating and irradiation data of the experimental reactor "JOYO" 33rd cycle, and estimates the 34th cycle irradiation condition. Irradiation tests in the 33rd cycle are as follows: (1)B-type irradiation rig (B9) (a)High burn up performance tests of "MONJU" fuel pins, advanced austenitic steel cladding fuel pins, large diameter fuel pins, ferrite steel cladding fuel pins and large diameter annular pellet fuel pins (b)Mixed carbide and nitride fuel pins irradiation tests (in collaboration with JAERI) (2)C-type irradiation rig (C4F) (a)High burn up performance test of advanced austenitic stainless steel cladding fuel pins (in collaboration with France) (3)C-type irradiation rig (C6D) (a)Large diameter fuel pins irradiation tests (4)Absorber Materials Irradiation Rig (AMIR-6) (a)Run to absorber pin's cladding breach (5)Core Materials Irradiation Rig (CMIR-5) (a)Cladding tube materials irradiation tests for "MONJU" (6)Core Materials Irradiation Rig (CMIR-5-1) (a)Core materials irradiation tests (7)Structure Materials Irradiation Rigs(SMIR) (a)Material irradiation tests (in collaboration with universities) (b)Surveillance back up tests for "MONJU" (8)Upper core structure Irradiation Plug Rig (UPR-1-5) (a)Upper core neutron spectrum effect and accelerated irradiation effect. The maximum burnup driver assembly "PFD516" reached 64,300MWd/t (pin average).
; Yamagata, Ichiro; ; Akasaka, Naoaki
JNC-TN9400 2000-046, 24 Pages, 2000/02
lt is well known that solute atoms are segregated on surface, grain boundary, etc. and composition changed partially in irradiated austenitic stainless steel. For understanding radiation induced segregation (RIS), we adopt a Fe-15Cr-20Ni-x (x: Si, Mo) which is basically alloy system in PNC1520, and size of Si, Mo are different from matrix atoms to investigate RIS behaviors. The specimens were irradiated by "Joyo" fast reactor that irradiation condition is 3.5 10 n/m (E>0.1Mev) at 476C. After irradiation, the specimen were observed and analyzed with EDS (Energy Dispersive X-ray Spectroscope) of 400kV TEM (Transmission Electron Microscope). The behavior of RIS depends on size of solute atoms of alloy. For example, oversized atoms are decreased and undersized atoms are increased in sink. RIS of voids are as same as or more than grain boundaries and smaller than precipitates. The void denuded zone was existed nearby G.B. in case of combinations between the grains from G.B.0ne of the reasons in this, the voids swepted by moving G.B. in radiation induced G.B. migration.
; ; Mizuta, Shunji
JNC-TN9400 2000-023, 126 Pages, 2000/02
Modified 316 and 15Cr-20Ni base austenitic stainless steels had been developed by Japan Nuclear Cycle Development lnstitute as the candidate materials for Monju and Demonstration fast breeder reactor. Previously, irradiation creep correlation of modified 316 and 15Cr-20Ni had been evaluated using pressurized tubes irradiated in FFTF/MOTA. 0n the other hand, for other austenitic S.S. developed abroad, it was reported that irradiation creep behavior of fuel pin could not be sufficiently described using results of pressurized tube experiments. ln this study, irradiation creep properties of modified 316 and 15Cr-20Ni fuel pins (MFA-I, 2) irradiated in FFTF were evaluated. And irradiation deformation of MFA-1, 2 fuel pins were estimated using the irradiation creep correlation based on MOTA data. The results are summarized as follows : (1)Irradiation creep compliance B calculated from MFA-I, 2 data are 5.6 15.010 [(I0n/m, E>0.1Mev)(MPa)], Which are larger than B based on MOTA data of 2.26.410 and are within the range of B of other austenitic S.S. abroad. (2)Creep-swelling coupling coefficient D derived from MFA-1, 2 data tend to decrease with increasing swelling rate. And the range of D based on MFA-1, 2 data include values calculated from MOTA data of 3.88.210 [(MPa)] and for other austenitic S.S. abroad. (3)As the result that irradiation creep deformation of MFA-1, 2 fuel pins could be appropriately estimated using the irradiation creep correlation derived from MOTA data, it is considered that the creep, correlation based on MOTA data can be applied to estimation of fuel pin deformation.
Shibata, Toshio*; *; *; Tsuru, Toru*; Inoue, Hiroyuki*
JNC-TJ8400 2000-013, 38 Pages, 2000/02
*; *; Tanai, Kenji
JNC-TN8400 99-049, 94 Pages, 1999/11
This paper reports the results of design analysis and trial manufacturing of copper-carbon steel composite overpacks. The overpack is one of the key components of the engineered barrier system, hence, it is necessary to confirm the applicability of current technique in their manufacture. The Copper-Carbon steel composite overpack consists of a double container, an outer vessel made of oxygen-free, high-purity copper as the corrosion allowance material, and an inner vessel made of carbon steel as the pressure-resistant material. The trial manufacturing in this time, only the copper outer vessel has been fabricated. Both oxygen-free copper and oxygen-free phosphorus copper were used as materials for the outer vessel. For the shell and bottom portion, these materials were formed integrally by a backward extrusion method. For sealing the top cover plate to the main body, an electron-beam welding method was applied. After manufacturing, mechanical testing of specimens from the copper vessels were carried out. It was confirmed that current technique has sufficient feasibility to manufacture outer vessel. In addition, potential for irradiation embrittlement of the inner carbon-steel vessel by irradiation from vitrified waste over the life time of the overpack has been analyzed. It was shown that the small degree of irradiation embrittlement gives no significant impact on the pressure resistance of the carbon-steel vessel. Future research and development items regarding copper-carbon steel composite overpacks are also discussed.
*; *; Tanai, Kenji
JNC-TN8400 99-047, 54 Pages, 1999/11
This paper reports on the design process for a carbon-steel overpack as a key component in the engineered barrier system of a deep geological repository described in the 2nd progress report. The results of the research and development regarding design requirements, configuration, manufacturing and inspection of overpack are also described. The concept of a composite overpack composed of two different materials is also considered. First, the design requirements for an overpack and presume environmental and design conditions for a repository are provided. For a candidate material of carbon steel overpack, forging material is selected considering enough experience of using this material in nuclear power boilers and other components. Second, loading conditions after emplacement in a repository are set and the pressure-resistant thickness of overpack is calculated. The corrosion thickness to achieve an assigned 1000 year life time and the required thickness to prevent radiolysis of ground water which might enhance corrosion rate are also determined. As aresult, the total required thickness of a carbon-steel overpack is conservatively estimated to 190 mm. This is a reduction of about 30% from the previous estimate provided in the 1st Progress Report. Additional items that must be considered in manufacturring and operating overpacks (i.e. sealing of vitrified waste, examination of main body and sealing welding, mechanism of handling) are evaluated on the basis of current technology, specific future data needs are identified. With respect to the concept of composite overpack (i.e., an outer vessel to provide corrosion-allowance or corrosion-resistant performance and an inner vessel to provide pressure-resistance), the differences in design concepts between the carbon-steel overpack and such composite overpacks are analyzed. Future data needs and analytical capabilities with respect to overpacks are also summarized.
JNC-TN9450 2000-002, 335 Pages, 1999/10
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".
JNC-TN9450 2000-001, 1370 Pages, 1999/10
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".
Mizuta, Shunji; ;
JNC-TN9400 99-082, 60 Pages, 1999/10
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 605C. 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.).
; Aoto, Kazumi;
JNC-TN9400 99-061, 32 Pages, 1999/07
In this report, reconstruction of magnetic charges induced by mechanical damages in a test piece of SUS304 stainless steel is performed as a part of eforts to establish a passive nondestructive testing method on the basis of the inspection of leakage magnetic field. The approach for solving this typical ill-posed inverse problem is selected as a way in the least square method category. Concerning the ill-poseness of the system of equations, an iteration algorithm is adopted to its solving in which the designations of initial profile, the weight coefficients and the total number of iterations are taken as means of reqularization. From examples using simulated input data, it is verified that the approach gives good reconstruction results in case of signals with a relative high S/N ratio. For improving the robustness of the proposed method, a Galerkin procedure with base functions chosen as the Daubechies' wavelet is also introduced for discretizing the governing equation. By comparing the reconstruction results of the least square method and those using wavelet discretization, it is found that the wavelet used approach is more feasible in the inversion of noise polluted signals. Reconstruction of 1-D and 2-D magnetic charges with the least square strategy and reconstruction of an 1-D problem with the wavelet used method are carried out from both simulated and measured magnetic field signals which are used as the validation of the proposed inversion strategy.