Swelling of cold-worked austenitic stainless steels irradiated in HFIR under spectrally tailored conditions

Wakai, Eiichi; Hashimoto, Naoyuki*; Robertson, J. P.*; Sawai, Tomotsugu; Hishinuma, Akimichi

Journal of Nuclear Materials, 307-311(Part.1), p.352 - 356, 2002/12

https://doi.org/10.1016/S0022-3115(02)01189-3

no abstracts in English

None

*

JNC TN1400 2000-006, 68 Pages, 2000/07

JNC-TN1400-2000-006.pdf:2.18MB

no abstracts in English

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/h), is correlated with lead and lithium temperature, T(K), as log Da = 10.7873 - 6459.3/ T and logDf = 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(m), using the following correlation: C = (Dt)/10, where t is exposure time(hr) and is density of the core matelial (g/cm). (3)The corrosion thickness estimated for austenitic steels using above correlations was extremely larger than ferritic steels, about 6 times at 400C and more than 20 times at above 600C. lt's considered that applicable temperature in lead cooled reactor core is below 400C (about 60m corrosion thickness after 30000 hr) for austenitic steels, and below 500C (about 80 m after 30000 hr) for ferritic steels.

Development of optimized advanced austenic steels (II); Evaluation of out-of-pile testing results of the fabricated fuel claddings

Uwaba, Tomoyuki; Mizuta, Shunji;

JNC TN9400 2000-028, 41 Pages, 2000/03

JNC-TN9400-2000-028.pdf:2.52MB

14Cr-25Ni optimized advanced austenic steels have been developed to improve the swelling resistance of 15Cr-20Ni austenic stainless steels used for FBR fuel cladding. ln this improvement, Ti,Nb,V and P were dissolved into 14Cr-25Ni marix by means of the high-temperature solution treatment to make finely distributed and stabilized precipitates in the operation. Furthermore, at the final stage of cold-working, cold-working level increased and residual stress was reduced. ln this study, as fablicated microstructure observation, solubility of alloying elements and grain size test in the manufacturing process were evaluated. Following results were obtained. (1)Spherical precipitates were observed in the grain. Most of them were identified as conjugated carbo-nitride [Ti,Nb(C,N)] by EDX analysis. (2)The dissolved percentages of Ti and Ni in the matrix were about 70% and 30% respectively. Undissoved Ti and Nb may react with undissolved carbon and precipitate as MC carbides. (3)High-temperature solution treatment is effective for the sufficient solubility of alloying elements, but it is likely to induce very large grains, which is the cause of defective signal in the ultrasonic alloy testing. The results of the grain size test showed that the large grain size is reduced in low Nb (0.1wt%) alloy compared with the standard alloy (0.2wt%Nb), and the effectiveness for the grain size control by reducing the Nb content was confirmed. Also, it was suggested that the intermediate heat treatment and cold work conditions would possibly avoid the occurrence of the large grain at the final heat treatment.

Post-lrradiation examination on Fe-15Cr-20Ni series model alloy irradiated by CMIR-2(1); Effect of defect sink and size of solute atom on radiation induced segregation(1)

; Yamagata, Ichiro; Donomae, Takako; Akasaka, Naoaki

JNC TN9400 2000-046, 24 Pages, 2000/02

JNC-TN9400-2000-046.pdf:1.1MB

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.

Simulation of creep test on 316FR stainless steel in sodium environment at 550

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 550C 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, the time to rupture of creep in sodium environment is the same or little longer than in air. Quantitatively, the carburization effect at ...

Microstructural evolution of welded austenitic stainless steel irradiated in HFIR target experiments

Sawai, Tomotsugu; Shiba, Kiyoyuki; Hishinuma, Akimichi

Journal of Nuclear Materials, 258-263, p.1997 - 2001, 1998/00

https://doi.org/10.1016/S0022-3115(98)00076-2

no abstracts in English

Study of irradiation damage structures in austenitic stainless steels

Hamada, Shozo

JAERI-Tech 97-041, 180 Pages, 1997/08

JAERI-Tech-97-041.pdf:13.34MB

no abstracts in English

Development of analytical model for evaluating temperature fluctuation in coolant (X); Investigation of thermally response characteristics of in-vessel structures using the BEMSET code

PNC TN9410 96-136, 92 Pages, 1996/05

PNC-TN9410-96-136.pdf:2.53MB

Thermal striping phenomena characterized by stationary random temperature fluctuations are observed in the region immediately above the core exit of liquid-metal-cooled fast breeder reactors (LMFBRs) due to the interactions of cold sodium flowing out of a control rod (C/R) assembly and hot sodium flowing out of adjacent fuel assemblies (F/As). Therefore the in-vessel components located in the core outlet region, such as upper core structure (UCS), flow guide tube, C/R upper guide tube, etc, must be protected against the stationary random thermal process which might induce high-cycle fatigue. In this study, thermally response characteristics of the flow guide tube made by SUS 316 stainless steels were investigated using a boundary element method code BEMSET under the temperature transient conditions of Sine wave, quasi-random wave, and Sine wave with quasi-random components. From the numerical investigations, it was concluded that the detailed handling on turbulence phenomena in coolant is very important in the evaluation of actual LMFBRs, because of the thermally response of the structures are influenced significantly on random fluctuating components.

Development of advanced materials for reactors

Hishinuma, Akimichi

Materia, 34(3), p.328 - 331, 1995/00

https://ci.nii.ac.jp/naid/10001648693

no abstracts in English