Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
;
JNC TN9400 2000-035, 164 Pages, 2000/03
JNC-TN9400-2000-035.pdf:3.67MB
High Strength Ferritic/Martensitic Steel (PNC-FMS : 0.12C-11Cr-0,5Mo-2W-0.2V-0.05Nb), developed by JNC, is one of the candidate materials for the long-life core of large-scale fast breeder reactor. Ductile brittle transition temperature (DBTT) was tentatively determined in 1992 in material design base standard of PNC-FMS. Howevcr, specimen size effect on impact property and upper shelf energy (USE) have not been evaluated. ln this study, effects of specimen size, thermal aging and neutron irradiation on the charpy impact property of PNC-FMS were evaluated, using together with recently obtained data. The design value of USE and DBTT as fabricated and each correlation of aging and irradiation effects were determined. The results are summarized as follows. (1)lt was found that USE is related to (Bb) as USE=m(Bb)
, where B is specimen width, b is ligament size and both m and n are constant. For PNC-FMS, n value is equal to 1.4. It's possible to determine n value from USE (J) for full size specimen using the correlation: n=1.38
10
USE + 1.20. (2)lt was clarified that DBTT is correlated with (BKt) as DBTT=p(log
BKt)+q, where Kt is elastic stress concentration factor and both p and q are constant. For PNC-FMS, the correlation is as follows: DBTT=119(logBKt)-160. (3)DBTT estimated at the irradiation temperature from 350 to 650 
C for sub size specimen (width and height are 3 and 10 mm, respectively), was below 180 C, based on the design value of DBTT as fabricated and each correlation of aging and irradiation effects.
Satmoko, A.*;
JNC TN9400 99-035, 37 Pages, 1999/04
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 ...
; Yoshida, Eiichi;
PNC TN9410 94-262, 120 Pages, 1994/09
This study clarified the tensie properties of Mod.9Cr-1Mo, 2.25Cr-1Mo and SUS321 steels at ultra-high temperature up to 1,200C which will be used in analysys and evaluation of the tube burst in steam generators of fast breeder reaetors. (1)Tensile strength of Mod.9Cr-1Mo, 2.25Cr-1Mo and SUS321 steels at 1,200C were 2.5, 2,and 2.5kg/mm, respectively. (2)The difference for tensile strength and 0.2% yeild strength between specimen heat rate and heat holding time could not be found in the present. (3)The temperatures of the tube burst at the maximum internal pressure of 150kg/cm corresponding to the practical use condition were expected to be approximately 960C for Mod.9Cr-1M0, 860C for 2.25Cr-1Mo and 1040C for SUS321, respectively. These tests result will be reflected to evaluation of tube burst by sodium water reaction.
; *; ; Yoshida, Eiichi;
PNC TN9410 94-261, 143 Pages, 1994/06
In this study, tensile, creep and relaxation test in air were performed in order to examine the mechanical properties of Mod.9Cr-1Mo steel which is a candidate material for once throuth type steam generator of large scale fast breeder reactor. Tested materials were plate(12mmt) simulating heat exchenger tube and heat exchenger tube of Mod.9Cr-1Mo steel and 9Cr-2Mo steel was also tested as reference material. Results obtained are summarized as follows. (1)Tensile properties (a)Ultimate tensile strength and 0.2% yield strength of Mod.9Cr-1Mo steels were higher than the tentative Su and Sy values of the design allowable stress in the test temperature below 600C. (b)Ultimate tensile strength of Mod.9Cr-1Mo steels plate and tube were higher than that of 9Cr-2Mo Steels. (3)The difference in ultimate tensile strength and 0.2% yield strength between steel plate and tube could not be found in these tests. (2)Creep properties (a)Creep rupture strength of Mod.9Cr-1Mo steel plate and tube was higher than the tentative S
value of the design creep-rupture stress intensity at 500
600C, and this tendency is significant in the range of longer rupture time. (b)For the relation between steady creep rate and creep rupture time, steady creep rates obtained in this study coincided well with the 
of tentative creep strain equation. (c)Creep rupture strength of Mod.9Cr-1Mo steel plate and tube was higher than that of 9Cr-2Mo steel. (3)Relaxation properties (a)In the strain range of 0.10.5%, stress rapidly relaxed during the short hold time, and stress relaxation tended to be saturate beyond 50hours. These relaxation stresses became large in higher temperature and higher strain level. (b)Stress relaxation behavior was predicted approximately by tentative creep equation of Mod.9Cr-1Mo steel. The analysis of these test results is continued to develop of evaluation method of material strength.
Shimoyama, Kazuhito; Hamada, Hirotsugu; Tanabe, Hiromi; Usami, Masayuki
PNC TN9410 93-212, 134 Pages, 1993/09
A one-through unit type steam generator (SG) having the Mod.9Cr-1MO Steel for its heat transfer tube is considered to be promising for the development of large FBR SGs. Wastage data of the tube material was already obtained for the micro-/small leak region as formerly reported. Therefore, intermediate leak wastage tests were conducted in the range from 10 g/s to around 200 g/s by using the SWAT-1 test facility and the test results are summarized as follows: (1)The wastage resistivity of the Mod.9Cr-1Mo steel is between that of 2.25Cr-1Mo steel and austenitic stainless steel; namely, the Mod.9Cr-1Mo steel has about half the of wastage rate of the 2.25Cr-1Mo steel. An experimental wastage formula in the intermediate leak region was derived from the test data. (2)Almost all of the wastage profile of target tubes was toroidal type and it became about half the cross section area of the 2.25Cr-1Mo steel. An experimental formula on initial leak diameters versus equivalent secondary failure diameters was derived in the intermediate leak region. These test results would be applied to failure propagation analysis code LBAP which is to be used for the design of a one-through unit type SG.
*; Watanabe, Ryuzo*
PNC TJ9601 93-004, 68 Pages, 1993/03
Molybdenum/stainless steel functionally gradient material (FGM). which will be used as long life fuel cladding tubes in the fast breeder reactor, has been fabricated by slurry dipping and sintering process emphasizing the increase in unti-corrosion against liquid sodium and fission products. Slurries of different compositions were prepared by mixing the appropriate amount of molybdenum and stainless steel powders in the ethanol. Green compacts giving cylindrical shape, substrates. were formed by die pressing of stainless steel powders. The substrates were dipped in the slurry, dried in the air and CIP'ed stepwizely: they were encapsulated in Pyrex glass tubes and then HIP'ed 2h at 1573K at the pressure of 150MPa. The microstructural observation in the cross section of the sintered compacts revealed that the uniform dipped layer was formed and there was no defect such as large residual pores or small cracks. A defect free Mo/stainless steel FGM was successfully fabricated by the slurry dipping and sintering process, however, in the case of single phase coating of Mo layer on the stainless steel substrate, the serious delamination was observed. Some oxides and compounds were detected in the FGM layer by the use of SEM-EDX and EPMA analysis.
; ;
PNC TN9410 92-148, 65 Pages, 1992/02
Mod.9Cr-1Mo steel is the material whose future utilization is expected as an advanced material of the steam generater of the Fast Breeder Reactors. A procedure for evaluation of weldment is being developed as one of the main concerns for the utilization of this material. The purpose of this report is to propose a fatigue strength evaluation method of Mod.9Cr-1Mo weldment which incorporates the effect of the heat affected zone which forms the softest portion of the weldment on the strength of the weldment. The mechanisms of strain concentration in the weldments of Mod.9Cr-1Mo steel was analysed and fatigue fracture was evaluated in terms of strain concentration. It was shown that the maximum strain concentration occured in the heat affeted zone at the begining of cyclic strain loading but that as a result of cyclic softning which was evident with the base metal but negligible with the heat affected zone the maximun strain concentration moved on to the base metal after half-life under the assumption that the hardness of the base metal and the heat affected zone coincide with each other at about half-life. Furthermore, based on this result, fatigue damage was evaluated. It was shown that the accumulated fatigue damage was not the maximum in the heat affected zone and that fatigue failure did not occur in the heat affected zone as far as the present analyses concerns. It was also made clear that fatigue failure occur in the base metal due to the strain concentration at strain ranges higher than elastic reagion and that it occur in the weld metal due to its inferior fatigue strength compared with the base metal. As a result, it was clarified that the fatigue strength of Mod.9Cr-1Mo weldment is reasonably evaluated by FE-analysis, which observation is able to be applied to the analysis of structures with weldments to be evaluated.
; ; *; *; *; Yoshida, Eiichi;
PNC TN9450 91-010, 259 Pages, 1991/10
In order to advancement in materials strength standard on elevated temperature design guide of the FBRs and evaluation method of materials strength behavior, this report are presented about the creep properties of Mod.9Cr-1Mo steels for steam generator, based on the R&D results obtained through the activities of material tests. Contents of the data sheet are as follows; Material : Mod.9Cr-1Mo steels (Base Metal) Plate 7 Heats (F2, F6, F7, F9, F10, NSC1, NSC2) Forging 8 Heats (F4, F5, F8, F11, VIM, ESR, F520, F550) Tube 1 Heats (F3) Test temperature : 450650C Test method : According to JIS and FBR Metallic Materials Test Method Test environment : In Air and in Sodium Number of deta : 314 points
; Hirakawa, Yasushi; Furukawa, Tomohiro; *; *; *; *
PNC TN9450 91-004, 71 Pages, 1991/07
In order to advancement in materials strength standard on elevated temperature design guide of the FBRs and evaluation method of materials strength behavior, this report are presented about the low-cycle fatigue properties of Mod.9Cr-1Mo steel, based on the R&D results obtained through the sctivities of material tests. Contents of the data sheet are as follows; [Material ; Mod.9Cr-1Mo steel(SR)] F2 Heat 1,0001,00012mm
(Plate) F4 Heat 1,0001,000250mm(Forging) F6 Heat 1,0001,00025mm(Plate) [Environment; In Air and in Sodium] [Test temperature ; 450, 500, 550, 600 and 650C] [Strain rate ; 0.1%/sec (10mm/mm/sec)] [Strain range ; 0.38% 1.86%] [Number of deta ; 83 points]
280mmt).*
PNC TN9410 90-122, 58 Pages, 1990/06
This study was performed to examine the tensile properties of 9Cr-Mo steel forgings, which are promising as the candidate materials for steam generator of large scale fast breeder reactor the influence of thermal aging and sampling location/direction. These results are to be reflected on fundation of materials strength standard. Test materials are three kinds of 9Cr-Mo steel forgings (thichiness:250280mmt) such as Mod.9Cr-1Mo (F4,F8 heats), 9Cr-1Mo-Nb-V(G3 hrats), 9Cr-2Mo (H6 heats) steels. The heat treatment on the thermal aging were carried out at 500, 550C for 3000hours. Results obtained are summarized as follows. (1)Tensile properties were much about the same between Mod.9Cr-1Mo and 9Cr-1Mo-Nb-V Steel forgings. (2)Tensile strength of four materials at direction Z was slightly lower than that of direction L or C and there are no significant diffrence on the sampling location/direction. (3)Tensile strength of Mod.9Cr-1Mo steel and 9Cr-1Mo-Nb-V steel forgings after thermal aging for 3000hours was same as that of as-received ones. However, the strength of 9Cr-2Mo steel forging at high temperature dicreased because of aging effect. (4)The 0.2 percent yield stress and tensile strength of forged steels satisfied the PNC preliminary design standard values of Sy and Su, but inferior to plating and tubing steels. However, tensile strengh of Mod.9Cr-1Mo steel forging (F8 heat) at the lower temperature than 400C and 9Cr-2Mo steel forging after thermal aging were lower than that of preliminary values. These results were reflected to advance the specification of Mod.9Cr-1Mo steel thick forgings.
*; *; Koakutsu, Toru; *; *
PNC TN9410 89-148, 158 Pages, 1989/10
For the purpose of the verification of the evaluation method on the sodium environmental effect on the mechanical properties of the structural materials used for the prototype LMFBR "MONJU" and the rationalization of the evaluation method for large scale LMFBRs, SUS 304 and 316 austenitic stainless steels and 2.25Cr-1Mo steel (NT) were carried out. Test specimens were exposed to a sodium loop for 10,000 hours at 400 600C simulating the primary and secondary coolant systems of the prototype LMFBR "MONJU". After the exposure, fatigue tests were performed in sodium environment. Fatigue tests were also performed on the thermal aged material for 10.000 hours in inert gas. The results obtaind were as follows. (1)The difference between the fatigue lives of sodium exposed materials and thermal aged materials was very small for both kinds of steels and these lives were almost the same as these of as - received materials. (2)Caburization was recognized on the surface of SUS 304 and SUS 316 austenitic stainless steels in the cage tested in sodium after sodium exposed. In the case of 2.25Cr-1Mo steel (NT), some decarburization was observed at 500 C. (3)The fatigue lives didnot depend on the exposure history such as sodium exposed materials and thermal aged materials for 10,000 hours. The carburization and decarburization effects were very small on fatigue life. The fatigue lives were affected by the environment in which fatigue tests are conducted.
Seshimo, Ichiro; Ukai, Shigeharu; Nomura, Shigeo; Shikakura, Sakae
PNC TN9410 89-122, 47 Pages, 1989/08
Fuel cladding integrity must be confirmed even at the anticipated transient event in LMFBR. In case of loss of coolant accident, cladding highest temperature is limited to 830 for Monju design from a view point of preventing the cladding creep rupture failure by increasing internal gas pressure. In this study, using the recent thermal transient test results of modified SUS316 stainless steel, Larson-Miller Parameter Life Fraction method was applied for predicting the optimum failure temperature. With this method, different thermal transient data can be evaluated systematically, and the effect of irradiation on cladding failure temperature was analyzed. Through this analysis, the Monju design limit temperature of 830 for cladding failure can be changed to 966, and alternative limiting temperature of 920 defined for preventing the coolant sodium boiling becomes a critical factor. This results shows the possibility of improvement for the design limit of this event.
*; *
PNC TN9410 88-105, 206 Pages, 1988/04
KOM-MSS W/G studies to rationalize and revise "Elevated Temperature Structural Design Standard - Material Strength Standard - for Fast Breeder Reactor" based on new engineering techniques, knowledges of new structural materials and new evaluation method, furthermore, based on rationalized safety argorithm and design margines which have been investigated at the economical point of view. In 1987, five sub-groups organized in KOM-MSS W/G began to study on each main theme as followed. (1)1st S/G : To study material properties of Mod.9Cr-1Mo steel and its weldment at elevated temperature and to revise "Material Strength Standard (tentative in 1986)". (2)2nd S/G : To study material properties of Mod. austenitic stainless steels at elevated temperature. (3)3rd S/G : To study material properties of weldment of SUS304 at elevated temperature. (4)4th S/G : To extend the application of the usual elevated temperature structural design guide for class 1 components of prototype LMFBR. (5)5th S/G : To rationalize the evaluation method for creep-fatigue damage.