Proceedings of the ASME 2009 Pressure Vessels and Piping Division Conference PVP2009 July 26-30, 2009, Prague, Czech Republic Proceedings of PVP2009 2009 ASME Pressure Vessels and Piping Division Conference July 26-30, 2009, Prague, Czech Republic PVP2009-77279 PVP2009-77279 IASCC EVALUATION METHOD OF IRRADIATED COLD WORKED 316SS BAFFLE FORMER BOLT IN PWR PRIMARY WATER Kiyotomo NAKATA Japan Nuclear Energy Safety Organization (JNES), Japan [email protected] Kenichi TAKAKURA Japan Nuclear Energy Safety Organization (JNES), Japan [email protected] Noboru KUBO Mitsubishi Heavy Industries (MHI), Ltd., Japan [email protected] Koji FUJIMOTO Mitsubishi Heavy Industries (MHI), Ltd., Japan [email protected] ABSTRACT Irradiation Assisted Stress Corrosion Cracking (IASCC) is a matter of great concern as degradation of core internal components in light water nuclear reactor. To clarify the IASCC initiation conditions of baffle former bolt (BFB), constant load stress corrosion cracking (SCC) tests were carried out in simulated PWR primary water (290, 320, 340°C) using C-ring type specimens. Based on the SCC test results, IASCC initiation time becomes shorter with increasing fluence and increasing applied stress, IASCC initiation threshold stress becomes lower with increasing fluence. A test temperature effect was observed in SCC initiation time, but it was not clear the effect of test temperature for SCC initiation threshold stress. These results suggest that IASCC initiation threshold criteria can be described with stress in specimen and fluence. This paper describes the whole evaluation procedure to secure structural integrity of irradiated baffle structure in PWR primary environments, including the threshold stress diagram of IASCC initiation and the irradiation creep formula. . INTRODUCTION Austenitic stainless steels are widely used as structural alloy in reactor pressure vessel internal components because of their high strength, ductility, and fracture toughness. However, exposure to neutron irradiation results in changes in microstructure, mechanical properties and microchemistry of the material. Kimihisa SAKIMA Mitsubishi Heavy Industries (MHI), Ltd., Japan [email protected] IASCC caused by the effect of neutron irradiation during long term plant operation in high temperature water environments is considered to be one of the major concerns of in-core structural materials for Light Water Reactors (LWR). IASCC is recognized as the most important degradation phenomena affecting the integrity of LWRs It takes the form of intergranular stress corrosion cracking (IGSCC). Actually, the baffle former bolts cracking have been caused occasionally in the overseas plants, e.g. BUGEY-2 in France, since 1988. Although such events are not found in the Japanese plants so far, it cannot be denied that the same events might occur in future. A number of studies on IASCC have been performed over the last two decades however the mechanism of IASCC is not fully understood. Furthermore, practical engineering databases about IASCC to evaluate the core shroud’s internal integrity are of an insufficient amount. Japan Nuclear Energy Safety organization (JNES) has been conducting a project related to IASCC as part of safety research & development study for the aging management & maintenance of the nuclear power plant. The objective of IASCC project is to prepare an IASCC evaluation guide for the regulatory side, which uses it to evaluate utility’s ageing management technical evaluation report. The project was begun in JFY 2000, completed in JFY 2008. The target component is the BFB, as can be seen from Figure 1, the BFB is the parts forming the baffle structure, witch composes the core region and primary coolant passage. The SCC initiation test in simulated PWR primary water had been conducted, on 1 Copyright © 2009 by ASME Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 10/26/2017 Terms of Use: http://www.asme.org/about-asme/terms-of-use the bases of SCC test results, the structural integrity evaluation guide for BFB is proposed as JNES document. TEST PROCEDURE Materials IASCC lifetime evaluation should preferably be carried out for BFBs from actual plants, however, these BFBs have today a fluence of only 20dpa at highest (corresponding to operating period of 20 years). BFB material irradiated up to fluence of 60dpa (corresponding to operating period of 60 years) are not available. Therefore, flux thimble tubes (FTTs), which are a tube in which a neutron flux measurer, and located in the center of a reactor, is used as sample instead of BFBs. The FTTs are made of type 316CW (cold worked) stainless steel similar to BFBs, but it has a higher fluence than BFBs due to different location in a reactor. For this reason, much research work on IASCC, using FTTs have been performed and reported [1-10]. In this study, constant load SCC tests have been conducted for both irradiated FTTs (maximum 70dpa) and BFBs (maximum 20dpa) to evaluate IASCC lifetime of BFBs at high irradiation level. Chemical compositions for test materials are shown in Table 1, and condition of cold working and grain size are shown in Table 2. The degree of cold work is 20% for BFBs, and 12% for FTTs. Additionally, there is a remarkable difference in the grain size of the two materials. The FTTs have a finer grain size, which can be attributed to recrystallization caused by repeated cold working and heat treatments. SCC Test Pieces Constant load SCC test is required for assessment of IASCC lifetime. In this study, C-ring type specimens were adopted for following three reasons: (1) SCC initiation tests require relatively high applied stress. It has been reported that stress above 0.2% yield strength of material is required for SCC initiation. (2) Specimen and test jig have to be as small as possible considering handling of the materials in a hot cell laboratory. (3) Bending stress at neck is dominant in an actual BFBs, and test under bending stress is possible in the case of C-ring test specimen in accordance with the actual case. The shape of the C-ring type specimen is shown in Figure 2. The outer diameter is 7.65mm and the wall thickness is 1.27mm for all specimens. The length of the specimen is 10mm for the FTTs, but 7 to 10mm for the BFBs due to limitations of available material. C-ring type specimens from the BFBs were machined from head, neck under head, straight portions, and threaded rejoin of the BFBs, and finished removing surface by mill scale. But Cring type specimens from the FTTs were prepared by simply cutting the tube into prescribed length (10mm), and not finished removing surface. Two pin holes with diameter of 1.2mm are made in the upper and lower parts of the specimen to fix the specimens firmly into the test jig. SCC Test Condition Constant load SCC tests were conducted in simulated PWR water environment. Test conditions are shown in Table 3. The tests were conducted at 290, 320, and 340°C. The maximum total test period was 2,000 hours. Figure 3 is a schematic illustration of the test jig structure with specimens installed. Three C-ring type specimens can be tested simultaneously. SCC initiation time (rupture of C-ring type specimen) was detected by a laser displacement sensor and by a change in the indication of the load cell. The applied stress in the C-ring type specimen was evaluated by finite element method (FEM) analysis using stress-strain curves obtained using tensile tests for each fluence level in high temperature. Based on the FEM analysis result, the applied load and stress was determined assuming that specimen cracks initiation occurs on the outer surface of C-ring type specimens. RESULTS AND DISCUSSION Constant load SCC tests were conducted using the C-ring type specimens (Total 81 specimens; 27 specimens from the BFBs and 54 specimens from the FTTs) to investigate the effect of fluence, temperature, and stress to determine SCC initiation time and SCC threshold stress. Observation of fracture surfaces after constant load SCC test Figure 4 through 5 show results of stereo microscope observations of outer surface cracks and SEM observations on fracture surfaces near the outer surface of the C-ring type specimens after the constant load SCC test. The surface cracks in the BFBs specimen were observed a more zigzag appearance while the surface cracks in the FTT specimen were straighter. This difference is considered to be caused by difference in grain size. The fracture surfaces of all the specimens show typical “intergranular” features. Dependence of SCC initiation time on applied stress Figure 6 shows the relationship between the SCC initiation time and applied stress. The top, middle, and bottom sides in Figure 6 corresponds to constant load SCC test results at 290, 320, and 340°C, respectively. As the results, the SCC initiation time decreases with increasing applied stress (σ) and fluence. Dependence of SCC initiation stress on fluence and applied stress Figure 7 shows the results from SCC initiation tests and describes the relationship between presence of SCC crack and test conditions, i.e. applied stress and fluence. The top, middle, and bottom sides in Figure 7 correspond to results at 290, 320, and 340°C, respectively. The result clearly indicates that the applied stress required for SCC initiation decreases with increasing fluence. Additionally, it was observed that SCC initiates even at a stress of approximately 600MPa (0.6σy) in the case of a fluence of 40dpa, and only 400MPa (0.4σy) at a fluence of 70dpa. At 2 Copyright © 2009 by ASME Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 10/26/2017 Terms of Use: http://www.asme.org/about-asme/terms-of-use 70dpa, SCC occurs at 340°C but does not occur at 320°C, see Figure 8. Effect of test temperature on SCC initiation tests The effect of test temperature on SCC initiation time and SCC initiation threshold stress was analyzed based on the result from the constant load SCC tests to show it in Figure 7. Significantly difference in the “SCC initiation threshold stress” between 290, 320, and 340°C was not observed. But the effect of temperature on “SCC initiation time” was observed. Therefore, I show the result that activation energy with a representative SCC test results (FTTs) in Figure 9. As a result, the activation energy was 26.3-34.7kcal/mol (an average of 30.6kcal/mol). Effect of material difference on SCC initiation Effects of differences in material on SCC initiation threshold stress were discussed in the following. SCC threshold stress for BFBs and FTTs show significant difference from Figure 8. To evaluate the validity of lifetime evaluation for actual plants BFBs using data obtained by testing FTTs, it is important to investigate the effect of minor differences in materials on IASCC characteristics in detail. As a result, a difference was not recognized to tensile strength, hardness with the bulk, RIS, quantity of irradiation generation gas (helium and hydrogen). However, irradiation hardenings of outer surface layer of the C-ring type specimens were observed with the increase of fluence to show it in Figure 10. In other words, the tendency that the hardening of the outer surface layer before the irradiation seemed to just increase was recognized in FTTs in comparison with BFBs. Therefore, we show the result that relationship between the SCC initiation threshold stress and Vickers hardness of the outer surface layer (about 2µm) for C-ring type specimens in Figure 11. As a result, we understood that we could correlate between SCC initiation threshold stresses and outer surface layer hardness of the C-ring type specimens. Therefore, for the difference of the SCC initiation threshold stress of BFBs and FTTs, we confirmed what the hardening on the outer surface layer of the C-ring type specimens influenced. IASCC EVALUATION METHOD Life time evaluation concept In the baffle structure, the core baffle is fixed to the core baffle plate by many baffle former bolts, and the baffle plate is installed to the core barrel by some barrel former bolts, i.e. the functions of the baffle structure are maintained by many baffle former bolts (Figure 1). Therefore, even if a crack is generated in one of the baffle former bolts, it does not cause the loss of the baffle structure functions immediately. However, when the number of cracked bolts is increased, the function to retain the baffle plate during seismic event is adversely affected. In other words, the increase in the number of cracked bolts deteriorates the plant safety and functions. Thus, for the plant maintenance, the baffle structural integrity has to be evaluated taking into account the number and the locations of the bolts which are found to have the potential of crack initiation in the evaluation. Figure 13 illustrates the essential features of procedure to specify the appropriate inspection period for the retention of baffle structure functions. For aging plant integrity evaluation, the inspection period is determined by comparing the number of bolts required for the retention of the baffle structure functions with the number of the bolts with cracking which is predicted by the evaluation method in this paper. Considering SCC test results, IASCC initiation of BFB can be described by material degradation index under neutron irradiation and stress values in BFB. The unit of dpa (displacement per atom) is adopted for the indicator of the material degradation by neutron irradiation. The threshold stress value of IASCC initiation decreases according to the neutron irradiation increasing, therefore, for the life time evaluation for aging PWR component, it is needed to consider the BFB behaviors appropriately which vary from hour to hour. Threshold stress diagram of IASCC initiation According to the result of constant load SCC test, the threshold stress for IASCC initiation of the BFBs and FTTs were observed slightly different. We think that the surface hardening layer influences it. However, the mechanism of such difference has not yet been clarified enough. Therefore, for conservative and practical evaluation, based on the neutron irradiation index (dpa) and the stress in the BFB at the evaluation time point, three regions according to the level of crack initiation possibility are allocated to the diagram in Figure 14. The each region is defined as follows: (1) Case that BFB stress and fluence are included in Region 1 The BFB is considered to have no possibility of crack initiation at the evaluation time point (2) Case that BFB stress and fluence are included in Region 2 The BFB is considered to have a possibility of crack initiation at the evaluation time point (3) Case that BFB stress and fluence are included in Region 3 The BFB is considered to have a high possibility of crack initiation at the evaluation time point Stress evaluation The stress in the BFB for the evaluation of the crack initiation possibility is calculated as the cumulative changes during the time from the start of operation to the evaluation time point. The reason is that, the stress at the specific evaluation time point is affected by the history of changes in configuration and mechanical property of material according to the increase in the neutron irradiation. The cause of the stress is considered to be the following factors: (1) The load factors before plant operation (the tightening stress, etc.) 3 Copyright © 2009 by ASME Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 10/26/2017 Terms of Use: http://www.asme.org/about-asme/terms-of-use (2) The thermal load during plant operation (the stresses due to the structural deformation caused by the temperature distribution, etc.) (3) The load factors due to aging (the stresses due to the structural deformation, e.g. swelling, caused by the increase in the neutron irradiation level, the stresses due to the changes in mechanical property caused by irradiation creep/neutron irradiation, etc.) For the irradiation creep estimation, JNES had been conducted in-pile creep tests in Halden reactor since 2005 to 2006, and examined the irradiation creep rate formula (Figure 12). When the neutron irradiation is proportional to operation time, the stress change in the BFB according to the time history is similar to that according to the neutron irradiation. Evaluation procedure To evaluate the crack initiation time of the evaluating BFB, the following procedure is applicable: (1) Conduct an evaluation to predict the stress history and the neutron irradiation history of the BFB, then, based on the results, estimate the stress history as a function of the fluence. (2) Compare the stress history obtained in above step with the threshold stress diagram of IASCC initiation (Figure 14), then, specify to which region evaluating BFB corresponds. (3) Adopt above two step evaluation to all the managing BFBs, these are indispensable BFBs to keep baffles structure function, then, the structural integrity of whole baffles structure should be judged from the number of damage BFB. SUMMARY AND CONCLUSIONS We examined the IASCC initiation behavior in the BFB, by the constant load SCC tests in the hot lab facilities. Based on the SCC test, the SCC initiation time was very short, around several hundred of hours, compared to plant operation time. Therefore, we focused on the relation between the material degradation index due to neutron irradiation, dpa , and the BFB stress. We present the evaluation method to secure structural integrity of the baffle structure, including the threshold stress diagram (fluence (dpa) vs. stress) and irradiation creep rate formula. ACKNOWLEDGMENTS The JNES IASCC project “Evaluation Technology for Irradiation Assisted Stress Corrosion Cracking” is supported by the Ministry of Economy, Trade and Industry (METI). REFERENCES [1] JAPEIC Report: Development of technology for long life plant (PLEX), “ Irradiated stainless steel SCC tests (PWR)”, (1997/3) [in Japanese]. [2] I. Suzuki, M. Koyama, H. Kanasaki, H. Mimaki, M. Akiyama, T. Okubo, Y. Mishima, T. R. Mager, “Stress Corrosion Cracking of Irradiated Stainless Steels in Simulated PWR Primary Water”, International Conference on Nuclear Engineering, ASME, 5 (1996), p.205-213 [3] R. P. Shogan, T. R. Mager, “Susceptibility of Type 316 Stainless Steel to Irradiation Assisted Stress Corrosion Cracking in a PWR Environment”, Proceedings of the 10th International Conference Environmental Degradation of Materials in Nuclear Power Systems- Water Reactors, NACE, (2001). [4] K. Fukuya, K. Fuji, M. Nakano, N. Nakajima, M.Kodama, “Stress Corrosion Cracking on Cold-Worked 316 Stainless Steels Irradiated to High Fluence”, Proceedings of the 10th International Conference Environmental Degradation of Materials in Nuclear Power Systems- Water Reactors, NACE, (2001). [5] J. Conermann, R. Shogan, K. Fujimoto, T. Yonezawa, Y. Yamaguchi, “Irradiation Effects in a Highly Irradiated Cold Worked Stainless Steel Removed from a Commercial PWR”, Proceedings of 12th International Conference Environmental Degradation of Materials in Nuclear Power Systems- Water Reactors, TMS, (2005). [6] K. Fujimoto, T. Yonezawa, E. Wachi, Y. Yamaguchi, M. Nakano, R. P. Shogan, J. P. Massoud, T. R. Mager, “Effect of the Accelerated Irradiation and Hydrogen/helium Gas on IASCC Characteristics for Highly Irradiated Austenitic Stainless Steels”, Proceedings of 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems- Water Reactors, NACE, (2005). [7] K. Fukuya, K. Fuji, H. Nishioka, Y. Kitunai, “Evaluation of Microstructures and Microchemistry in Cold-work 316 Stainless Steels under PWR Irradiation”, J. Nuclear Science and Technology, Vol.43, No.2, pp.159-173, (2006). [8] P. Freyer, T. R. Mager, M. A. Burke, “Hot Cell Crack Initiation Testing of Various Heats of Highly Irradiated 316 Stainless Steel Components Obtained from three Commercial PWRs”, Proceedings of 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Whistler, British Columbia August 19 - 23, 2007. [9] H. Nishioka, K. Fukuya, K. Fujii, T. Torimaru, “IASCC Properties and Mechanical Behavior of Stainless Steels Irradiated up to 73dpa”, Proceedings of 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Whistler, British Columbia August 19 - 23, 2007. [10] K. Takakura, K. Nakata, M. Ando, K. Fujimoto, E. Wachi, “Lifetime Evaluation for IASCC Initiation of Cold Worked 316 Stainless Steel’s BFB in PWR Primary Water” Proceedings of 13th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, August 19-23, 2007, Whistler, Canada. 4 Copyright © 2009 by ASME Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 10/26/2017 Terms of Use: http://www.asme.org/about-asme/terms-of-use Former Plate Core Barrel Baffle Former Bolt Former Plate Baffle Plate Core Barrel Baffle Plate Former Plate Baffle Former Bolt Core Barrel Barrel Former Bolt Former Plate Figure 1. Core Internals of PWR Plant 5 Copyright © 2009 by ASME Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 10/26/2017 Terms of Use: http://www.asme.org/about-asme/terms-of-use Table 1. Material C 0.05 0.06 0.05 BFB FTT1 FTT2 Si 0.55 0.63 0.63 Chemical composition of BFB&FTT materials Chemical Composition (wt%) Mn P S Ni Cr 1.55 0.021 0.025 12.45 17.71 1.68 0.020 0.005 12.25 16.25 1.70 0.019 0.005 12.32 16.59 Mo 2.26 2.36 2.31 Fe Bal. Bal. Bal. Table 2. Final cold work condition and grain size number Material Final Cold Work Grain Size No. BFB 20% 4.8 FTT 12% 9.2 1.27 φ7.65 Table 3. Constant load SCC test condition Items Condition H3BO3 1,200 ppm as B LiOH 2.0 ppm as Li Dissolved Oxygen (DO) < 5 ppb Dissolved Hydrogen (DH) 30 cc/kg H2O・STP Temparture 320, 340°C Duration Time Max. 2,000hr 7~10 Figure 2. Constant load SCC test specimen (C-ring type: BFB&FTT) Figure 3. Constant load SCC test equipment and specimen holder 6 Copyright © 2009 by ASME Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 10/26/2017 Terms of Use: http://www.asme.org/about-asme/terms-of-use BFB FTT Figure 4. Photograph of C-ring BFB and FTT specimen after SCC test BFB FTT Figure 5. SEM images of C-ring BFB and FTT specimen after SCC test 7 Copyright © 2009 by ASME Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 10/26/2017 Terms of Use: http://www.asme.org/about-asme/terms-of-use 1400 1400 Test Temparature:290℃ 1200 1000 1000 Stress, σ (MPa) Stress, σ (MPa) Test Temparature:290℃ 1200 800 600 400 FTT Failures 800 600 400 20dpa FTT Failures 200 200 40dpa FTT Failures 70dpa FTT Failures 0 0.1 1 10 100 Time to Failure, t (hr) 1000 0 10000 0- 1400 20 40 60 Fluence, φ(dpa) 80 100 1400 Test Temparature:340℃ Test Temparature:320℃ 1200 1200 1000 1000 BFB Failures FTT Non-Failures 10dpa 10dpa 20dpa 20dpa 10dpa 20dpa 20dpa 30dpa 40dpa 40dpa 70dpa 70dpa 800 600 400 200 0 Stress, σ (MPa) Stress, σ (MPa) FTT Failures BFB Failures BFB Non-Failures BFB Failures BFB Non-Failures FTT Failures FTT Failures Non-FTT Failures Non-FTT Failures FTT Failures Non-FTT Failures FTT Failures Non-FTT Failures 0.1 800 600 400 200 0 1 10 100 Time to Failure, t (hr) 1000 10000 -0 1400 Test Temparature:340℃ 1200 1200 1000 1000 Stress, σ (MPa) Stress, σ (MPa) 40 60 Fluence, φ(dpa) 80 100 1400 Test Temparature:320℃ 800 600 20dpa 20dpa 20dpa 40dpa 40dpa 70dpa 70dpa 400 200 0 0.1 BFB Failures BFB Non-Failures FTT Failures FTT Failures Non-FTT Failures FTT Failures Non-FTT Failures 1 10 100 Time to Failure, t (hr) Test Temparature:290, 320, 340℃ 600 400 1000 0 10000 0- BFB Non-Failures (340℃) BFB Non-Failures (320℃) FTT Non-Failures (340℃) FTT Non-Failures (320℃) FTT Non-Failures (290℃) INSS Non-Failures 〔9 〕 WH Non-Failures 〔8〕 20 40 60 Fluence, φ(dpa) 80 100 Figure 7. Constant load SCC test results at 290/320/340°C as stress (σ) versus fluence BFB Failures (340℃) BFB Failures (320℃) FTT Failures (340℃) FTT Failures (320℃) FTT Failures (290℃) INSS Failures 〔9〕 WH Failures 〔8 〕 -2 -3 Threshold stress of BFBs 1000 BFB Failures FTT Failures 200 1400 1200 BFB Non-Failures FTT Non-Failures 800 Figure 6. Constant load SCC test results at 290/320/340°C as stress (σ) versus failure time 800 ln(1/t) Stress, σ (MPa) 20 600 400 ○75dpa,800MPa:75dpa, 30.7kcal/mol 800MPa ▲40dpa,800MPa: 4.7kcal/mol 40dpa, 800MPa ■40dpa, 600MPa: 26.3kcal/mol 40dpa, 600MPa Total: 30.6kcal/mol -4 -5 -6 y = -15384x + 21.199 200 Threshold Stress of FTTs -7 1.6E-03 0 -0 Figure 8. 20 40 60 Fluence, φ(dpa) 80 100 Figure 9. Total constant load SCC test results as stress versus fluence 8 1.7E-03 1/T 1.8E-03 Test temperature (activation energy) effect of constant load SCC tests Copyright © 2009 by ASME Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 10/26/2017 Terms of Use: http://www.asme.org/about-asme/terms-of-use 600 600 BFB(1) FTT(1) B3-13(BFB 11.4dpa) B4-11(FTT 10.5dpa) B3-14(BFB 10.4dpa) 20dpa 500 B3-31(BFB 19.8dpa) B4-28(FTT 19dpa) Vikers hardness (HV0.1) Vikers hardness (HV0.1) B4-27(FTT 19.5dpa) 500 B3-32(BFB 19.8dpa) 400 300 B4-12(FTT 9.5dpa) 20dpa 10dpa 200 FTT 0dpa 400 300 10dpa 200 0dpa 100 0 50 100 150 Depth profile from outer surface of C-ring type specimens (µm) Figure 10. 100 200 0 50 100 150 Depth profile from outer surface of C-ring type specimens (µm) 200 Hardness profile from outer surface of C-ring type specimen for BFB&FTT SCC initiation threshold stress(MPa) 1000 900 800 700 600 y = -3.3088x + 2036.6 R2 = 0.7204 500 400 300 200 BFB 100 FTT 0 300 350 400 450 500 Vickers hardness of the outer surface layer (HV0.1) Figure 11. Relationship between SCC initiation threshold stress and surface hardness of C-ring type specimen Creep Rate, mm/mm/dpa 0.001 Irradiation creep rate formula 340℃ ε=A・σ・[ 1-exp(-A’・F)] + B0・σ(1+B1・σ3)・F 290℃ ε : Creep strain B0 : 0.98x10-6 /(dpa・MPa) B1 : 0.33x10-8 MPa -3 σ : Stress (MPa) A : 0.73x10-6 /MPa F : fluence (dpa) A’ : 4.3 /dpa Temperature : 290C-340C Formula Bσ(1+B''σ^3) 0.0005 0 0 200 400 600 800 Stress, MPa Comparison of Halden experimental data and this formula Figure 12. Formula of irradiation creep rate 9 Copyright © 2009 by ASME Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 10/26/2017 Terms of Use: http://www.asme.org/about-asme/terms-of-use :Achievements in the Project Stress Evaluation of Bolt (Detailed Analysis) -Tightening Stress -Thermal Deformation -Irradiation creep -Swelling -Irradiation Hardening Evaluation of Threshold Stress due to IASCC (The SCC Test applying constant load with various temperatures, irradiation levels and stresses) Evaluation of Plant Operating Time and Irradiation Level (considering the difference in operating characteristics) Prediction of Timing of When BFB Is Damaged Defining the allowable number of defect bolts for the retention of the baffle structure functions. Inspection:Determine whether the number of defect bolts can be within the allowable number of defect bolts(i.e. the continuing use is possible or not) by the next time of inspection Repair Work Is Necessary. Repair Work:Replace part or all of the defect bolts. Figure 13. Essential Features of Procedure to Specify Appropriate Check Timing for Retention of Baffle Structure Functions バッフルフォーマボルト割れ無し 1400 試験温度:290、320、340℃ Test Temperatures:290 deg.C., 320 deg.C., and 340 deg.C. Stress σ(MPa) 1200 No Crack in Baffle Former Bolt test specimen バッフルフォーマボルト割れ有り Crack Generation in Baffle Former Bolt test specimen No Crackシンブルチューブ割れ無し in Thimble Tube test specimen Crack Generation in Thimble Tube test specimen シンブルチューブ割れ有り 1000 Region 3 800 600 Region 2 400 200 0 Region 1 Curve A:Threshold line of SCC Initiation in Thimble Tube test specimen -0 20 40 60 Neutron Irradiation Level(dpa) 80 100 Figure 14. Threshold Stress Diagram of IASCC Initiation 10 Copyright © 2009 by ASME Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 10/26/2017 Terms of Use: http://www.asme.org/about-asme/terms-of-use
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