IFE/HRIE / /09/23

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1 Performing Organisation lnstitutt for Energiteknikk Document no.: Date IFE/HRIE / /09/23 Halden ProjecUContract no. and name ClienUSponsor Organisation and reference: Title and subtitle Comparison of In-Reactor Creep and Stress Relaxation of Cold Worked 316 and Solution Annealed 304L Stainless Steels in Thermal and Fast Neutron Spectrum Reactors Author(s) J. Foster, T. Karlsen Reviewed Approved Abstract KeyWords ISSN Supplementary Data ISBN (printed) (electronic) Numbers of Pages 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors, Colorado Springs, USA, 7-11 August 2011

2 IRRADIATION CREEP AND IRRADIATION STRESS RELAXATION OF 316 AND 304L STAINLESS STEELS IN THERMAL AND FAST NEUTRON SPECTRUM REACTORS John Paul Foster Westinghouse Electric Company 5801 Bluff Road Columbia, SC USA Torill M. Karlsen OECD Halden Reactor Project Halden Norway Irradiation creep and irradiation stress relaxation data have been obtained for CW 316 SS and SA 304L SS in the Halden reactor. The measurements were performed on-line during irradiation. The irradiation creep in this thermal neutron spectrum reactor was observed to be very different than values measured in fast neutron spectrum reactors. The steady state irradiation creep rate was higher in this and other thermal neutron spectrum reactors than in fast neutron spectrum reactors. On the other hand, the transient irradiation creep component was lower in thermal neutron spectrum reactors than fast neutron spectrum reactors. Therefore, fast reactor irradiation creep data are not recommended for application to light water reactors. I. INTRODUCTION Structural materials in light water reactors (Pressurized Water, PWR, and Boiling Water, BWR, Reactors) are subject to irradiation assisted stress corrosion cracking (IASCC). The initiation and propagation of IASCC cracks requires a tensile stress, a corrosive environment and a susceptible material. Generally, 316 stainless steel (SS) and 304 SS are used as structural materials and are reasonably resistant to corrosion; however, these materials are susceptible to IASCC. During reactor operation the neutron flux results in irradiation creep or irradiation stress relaxation. Hence, the initial stresses in structural components will vary during reactor operation. If the stresses decrease to zero due to irradiation stress relaxation, then additional IASCC cannot occur thereafter. The purpose of this program is to measure the irradiation creep and irradiation stress relaxation of common reactor structural materials in a thermal neutron reactor spectrum prototypic of PWRs and BWRs. This paper updates the previous paper 1 with higher dose data. II. EXPERIMENTAL The experimental details are described in reference 1 and will be briefly reviewed. The test samples consisted of cold worked 316 stainless steel N lot (CW 316 SS) and solution annealed 304L SS (SA 304L SS). The samples were fabricated from tube material, and were the double ligament type. The samples were loaded in uniaxial tension. The chemical compositions and mechanical properties are given in reference 1. The test conditions for each instrumented sample are listed in Table 1. The CW 316 SS sample was tested in stress relaxation and the two SA 304L SS samples were tested in creep. The instrumented sample test unit (Figure 1) consists of a loading device and length measurement system. The uniaxial sample is loaded by a flexible bellows connected to one end of the sample. Pressure acting on the outside of the bellows is controlled externally to keep the load constant. The accuracy of the pressure measurements lie within 2x10-3 MPa, with a corresponding accuracy in applied load of ± 0.4 MPa. The sample length is measured by a linear voltage differential transformer (LVDT). The sample temperature, which typically varies between 4 C of the target value during stable

3 reactor operation, is measured by means of a thermocouple in the test unit and is controlled on-line by varying the gas composition (He-Ar mixture) in the gas-gap surrounding the specimen. The sample length extension, temperature, pressure (stress) and fluence (dose) are recorded every 15 minutes. The test rig was located in ring 5 of the reactor core (see Figure 2). Self-powered vanadium neutron detectors were used to measure the thermal flux on-line during irradiation. In addition, thermal and fast fluence wires are included in the rig. The wires will be measured at the end of the test. The thermal fluence wire is a 1%Co-Al alloy, and will be used to confirm the self-powered vanadium neutron detector thermal flux measurements. Usually, there is good agreement between the self-powered vanadium detector and the fluence wire. The fast flux was calculated from the thermal flux measurements, the fuel burnup and a conversion factor. The conversion factor is based on values obtained from previous tests that were performed using on-line thermal neutron detectors and fast flux monitor wires. The rig includes Fe and Ni fast flux monitor wires which will be measured at the end of the test when the rig is discharged from the reactor and compared with the calculated fast flux values. Usually, there is good agreement between the calculated and measured fast flux values. The fast fluencies were converted to dpa using the conversion factor of 7x10 21 n/cm 2 (E>1 MeV) per 1 dpa reported by Gan et al. 2 III. RESULTS On-line irradiation stress relaxation and irradiation creep data have been obtained. Figure 3 presents the irradiation stress relaxation data for CW 316 SS. The test was performed at a nominal temperature of 370 o C. The initial stress level was 343 MPa. Figures 4 and 5 present the irradiation creep data for SA 304L SS. The test was performed at a nominal temperature of 290 o C. The initial stress levels were 110 and 92 MPa for test units 7 and 8, respectively. The plan was to run these samples in stress relaxation; however, after test initiation the test method was changed to creep because the measured strains were relatively small. As a result of the initial stress relaxation plan, the initial applied stress of the test unit 8 sample was reduced from 92 to 80 MPa. After this reduction, the stress was held constant at 80 MPa and the sample tested in creep. The average flux for all the test samples was about 1.15x10 13 n/cm 2 -sec (E>1 MeV). The elongation data in Figures 3, 4 and 5 were linearly normalized to the nominal temperature, average flux and average stress values. The temperature, flux and stress fluctuated over time during the test. The sample elongation directly depends on these parameters. In the case of temperature, the sample elongation or contraction is given by thermal expansion, ΔL = L o α ΔT, (1) where ΔL is the sample elongation, L o is the initial sample gauge length, α is the linear thermal expansion coefficient and ΔT is the variation in temperature from the nominal temperature. Equation 1 shows that ΔL is directly proportional to ΔT. In the case of stress and flux, consider the following empirical creep strain model, ΔL/L o = Aσ [1 e -Bφt ] + Cσ φt (2) where φ is the flux, σ is the stress, t the time and A, B and C are material coefficients. The elongation depends directly on the flux and stress. As a result, elongation values associated with temperature, flux and stress variations from the average values were linearly normalized to the average values of the temperature, flux and stress. The dose was normalized whenever there was no irradiation creep. This occurred when (1) the reactor power was very low so that the fast flux was also very low, and (2) when the stress was more than a factor of two lower than the nominal stress. Figure 3 presents the irradiation stress relaxation of CW 316 SS N lot (unit 9). The elongation data were linearly normalized to the nominal temperature of 370 o C, the average flux of 1.15x10 13 n/cm 2 -sec (E>1 MeV) and the average stress. Irradiation stress relaxation data are available at eleven stress levels of 343, 338, 325, 314, 303, 287, 266, 231, 220, 205 and 191 MPa. The solid line in Figure 3 shows the elongation value used to determine the dose associated with each stress level. Note that the slope of the data decreases with increasing dose level. This decrease in slope is characteristic of the transition from transient to steady state irradiation creep. Figure 6 presents the stress relaxation data. Figures 4 and 5 present the irradiation creep of SA 304L SS (units 7 and 8). The

4 elongation data were linearly normalized to the nominal temperature of 290 o C, the average flux of 1.15x10 13 n/cm 2 -sec (E>1 MeV) and the average stress of 110 MPa for unit 7 and 92 and 80 MPa for unit 8. As noted above, these samples were initially targeted for irradiation stress relaxation but were switched to irradiation creep because of their relatively small elongation. In the case of unit 7, the sample exhibits large initial shrinkage which is usually due to carbide-induced densification. The maximum shrinkage is reached at a dose of about 0.08 dpa. The elongation then slightly increases from 0.08 to 0.21 dpa indicating an increase in irradiation creep strain. The initial data of unit 8 are similar to unit 7. The sample exhibits large initial shrinkage followed by a slight increase in elongation up to about 0.19 dpa. The samples in units 7 and 8, after the initial shrinkage, exhibit steady state irradiation creep. In the case of unit 7, at 0.21 dpa the extensometer exhibits a large non-representative increasing elongation followed by a large nonrepresentative decreasing elongation. This behavior is then repeated several times with smaller increases and decreases. Note that the non-representative elongation increases are greater than the non-representative elongation decreases resulting in a net positive elongation. In general, the LVDT extensometers occasionally exhibit non-representative irradiation creep/irradiation stress relaxation data up to dose levels in the range of dpa. Eventually the LVDT extensometers stabilize and exhibit correct irradiation creep/irradiation stress relaxation elongation values. This behavior appears to occur at low doses and is considered to be related to stabilization of the rig and test units. Figure 1 shows that the test unit is very complicated. In the case of unit 7, the net irradiation creep elongation increase is considered to be steady state irradiation creep, which is approximated by the solid line. In the case of the sample in unit 8, the sample exhibits steady state irradiation creep from 0.20 to 0.82 dpa. At about 0.40 dpa the sample exhibits nonrepresentative decreasing irradiation creep elongation. The extensometer stabilizes at about 0.55 dpa. After the establishment of an initial steady state irradiation creep rate, the samples in units 7 and 8 subsequently exhibit an increased steady state irradiation creep rate. The creep rate increased by about a factor of 2.5 (the values are listed in Table 2). The increase for unit 7 occurs at a dose of about 0.74 dpa. The increase in unit 8 occurs at about 0.82 dpa. This increase in the steady state creep rate at low dose cannot be compared with the fast reactor data because the higher stressed capsules yielded during the first reactor cycle. 3 As a result, the initial diameters were taken to be the measured diameter after the first cycle. The first cycle dose was about 0.8 dpa, which is the dose that the increase occurs in the Halden Reactor. IV. DISCUSSION Fast reactor irradiation creep data are not consistent with the irradiation stress relaxation data of this study that was performed in a thermal neutron spectrum. The irradiation creep and irradiation stress relaxation tests were performed with the same material lot of 20% CW 316 SS (the lot is designated as N lot and is Carpenter Technology heat number V87210) at approximately the same temperature. The fast reactor irradiation creep tests (see references 4 and 5) were performed in EBR-II. In the case of the steady state irradiation creep component (that is, the C coefficient in equation 2 and Table 2), the EBR-II test was performed at a temperature of 377 o C. The irradiation stress relaxation test of this study was performed in a thermal neutron reactor spectrum (the Halden Reactor) at a temperature of 370 o C. The results are listed in Table 2. The creep rate measured in this study (i.e., in a thermal neutron reactor spectrum) is a factor of 2.6 higher than measured by EBR-II (fast reactor neutron spectrum) pressurized tubes (see reference 4). Further, similar results were observed with SA 304L SS. The irradiation creep tests in the Halden and EBR-II reactors were performed with the same material lot of SA 304L SS at different temperatures. The irradiation creep test of this study was performed at 290 o C. The irradiation creep test in EBR-II was performed at a temperature of 390 o C (see reference 6). The irradiation creep of SA 304L is not temperature dependent over the temperature range of 175 to 370 o C 7. As a result, the steady state irradiation creep rates of this study may be compared with the EBR-II results. Table 2 shows that the creep rate measured in this study (i.e., in a thermal neutron reactor spectrum) initially is a factor of about 7.3 higher than measured by EBR-II (fast reactor neutron spectrum). These results are in agreement with other reported thermal and fast neutron spectrum reactor irradiation creep data. In addition to the present study, two studies have reported

5 irradiation creep data measured in thermal and fast neutron spectrum reactors using the same material lots. The first data set was reported by Garnier et al. 8 Further, since the publication of reference 1, Garnier et al., 8 reported irradiation creep tests on CW 316 SS and SA 304L SS in BOR-60 (fast reactor) and Osiris (thermal reactor). The data exhibit an incubation plus steady state diameter strain versus dose dependence. Garnier et al., 8 used an irradiation creep equation that included both the initial incubation plus steady state. The steady state irradiation creep C coefficients were calculated using the linear ΔR/R versus φt data reported in Figures 1 and 3 of reference 8 because the equation used by Garnier et al., is slightly different. The linear ΔR/R versus φt data were fit to, ε = Cσ φt (3) where ε and σ are the equivalent stress and strain. The results are listed in Table 3. In the case of the BOR-60 SA 304L SS data, the C coefficients varied between 2.4 to 3.4x10-6 /MPa-dpa. The average value is given in Table 3. In the case of the CW 316 SS Osiris data, the data points less than 460 MPa-dpa appear to be in incubation and were excluded. The results show that the CW 316 SS and SA 304L SS steady state irradiation creep component in Osiris (thermal reactor) was faster than in BOR- 60 (fast reactor). For CW 316 SS, the C coefficient increased by a factor of 1.7, and for SA 304L SS the increase was 1.4. This result is consistent with the results of this study. Further, the steady state irradiation creep coefficients for CW 316 SS and SA 304L SS reported by this study are in agreement with Garnier et al. In the case of CW 316 SS, the steady state irradiation creep coefficient of 3.1x10-6 /MPa-dpa in this study is in good agreement with the 3.6x10-6 /MPa-dpa in Osiris. In the case of SA 304L SS, the steady state irradiation creep coefficients of 3.2x10-6 /MPa-dpa initially and 6.7x10-6 /MPadpa subsequently in this study are in agreement with the 3.6x10-6 /MPa-dpa in Osiris. The second thermal and fast reactor irradiation creep data set using the same material lots was reported by Mosedale and co-workers. Mosedale and co-workers performed irradiation creep tests using CW M316 SS, CW FV548 SS and fully heat-treated (FHT) PE16 springs fabricated with the same material lots and tested in the DFR (fast reactor with samples tested in the temperature range of o C) and DMTR (thermal reactor with samples tested at a temperature of 100 o C). An evaluation of the reported data was performed by Foster and Boltax. 9 Foster and Boltax reported that the irradiation creep strain rate was a factor of 2.5 higher in the DMTR (thermal reactor) than in the DFR (fast reactor). This result is consistent with the results of this study. The reason for the increase in the SA 304L SS irradiation creep rate in the fluence range of 0.74 to 0.82 dpa is not clear. This increase could be related to the initiation of bubble and/or void swelling. TEM examination of the samples is required for confirmation. Unfortunately, TEM examination cannot be performed until the test is discharged from irradiation. The initial densification and increase in the steady state irradiation creep rate at low dose values cannot be compared with the EBR-II data because the diameter measurements from first irradiation cycle in EBR-II were invalid. The diameter measurements after the first cycle were inconsistent with the pre-test diameter measurements. The increased irradiation creep in thermal neutron reactors in comparison with fast reactors could possibly be explained by the increase in dose rate associated with 59 Ni. Garner et al., 10 report that 59 Ni results in increased dose rates in thermal reactors relative to fast reactors. This effect is not included in the method currently used to calculate dose rates. Hence, current methods used to calculate thermal reactor dose rates underestimate the actual dose rate in thermal reactors. However, in the case of the present study, irradiation testing has only been performed up to about 1 dpa which is relatively low. It is uncertain that large dose rate increases are associated with relatively low dose levels in the Halden Reactor. Work is inprogress with Dr. Lawrence Greenwood/Pacific Northwest National Laboratory to perform 59 Ni calculations in order to evaluate this effect. In the case of the transient irradiation creep component (that is, the A coefficient in equation 2 and Table 2), data are only available for CW 316 SS. Figures 4 and 5 show that the SA 304L SS samples did not exhibit a transient irradiation creep component. The transient irradiation creep coefficient measured in this study is a factor of 13 less than the value measured in EBR-II (see reference 3). The EBR- II test temperature of 450 o C is much higher than for this study (370 o C). However, Figure 5 of reference 2 shows that the transient irradiation creep coefficient, within experimental scatter, is

6 constant over the temperature range of about 378 to 464 o C. Hence, the results for the transient irradiation creep component of this study may be directly compared with EBR-II test data even though the temperatures are different. Note that the difference in the thermal versus the fast reactor transient irradiation creep coefficients are opposite the steady state irradiation creep component. V. CONCLUSIONS The following conclusions may be drawn from the Results and Discussion: 1. The steady state irradiation creep rate is higher in thermal neutron spectrum reactors than in fast neutron spectrum reactors. On the other hand, the transient irradiation creep component is lower in thermal neutron spectrum reactors than in fast neutron spectrum reactors. 2. Irradiation creep and irradiation stress relaxation differences between thermal and fast reactors show that data from thermal reactors such as Halden should be used for LWR applications. ACKNOWLEDGEMENTS The authors gratefully acknowledge Dr. D. L. Porter (Argonne National Laboratory- West) for providing the CW 316 SS N lot and SA 304L SS tubing. The authors would also like to acknowledge, with thanks, the efforts of Halden Project staff in both the design and production phase of this test as well as in the day-to-day follow-up that is required during irradiation. In addition to the reactor control room staff particular thanks are given to Ø. Brennvall, V. Grismanovs, N-W. Høgberg, R. Suther, R. Van Nieuwenhove and B. Volkov. REFERENCES 1. John Paul Foster and Torill Karlsen, Comparison of Irradiation Creep and Irradiation Stress Relaxation of 316 and 304L Stainless Steels in Thermal and Fast Neutron Spectrum Reactors, 14 th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors, Virginia Beach, Virginia, August 23-27, J. Gan, D.J. Edwards, E.P. Simonen, S.M. Bruemmer and G.S. Was, Microstructural Evolution and Hardening in 300-Series Stainless Steels: Comparisons Between Neutron and Proton Irradiations, 10 th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors, Lake Tahoe, Nevada, August 5-9, G. L. McVay, L. C. Walters and G. D. Hudman, Neutron Irradiation-Induced Creep of Helium-Pressurized 304L Stainless Steel Capsules, J. Nucl. Mater. 79 (1979) J. P. Foster, K. Bunde, M. L. Grossbeck and E. R. Gilbert, Temperature Dependence of the 20% Cold Worked 316 Stainless Steel Steady State Irradiation Creep rate, J. Nucl. Mater. 270 (1999) E. R. Gilbert, Irradiation Embrittlement and Creep in Fuel Cladding and Core Components, Proceedings of the British Nuclear Energy Society Conference, London, November 9-10, 1972, John Paul Foster, Kermit Bunde and Douglas L. Porter, Irradiation Creep of Annealed 304L Stainless Steel at Low Dose Levels, J. Nucl. Mater. 317 (2003) J. P. Foster, E.R. Gilbert, K. Bunde and D.L. Porter, Relationship Between In- Reactor Stress Relaxation and Irradiation Creep, J. Nucl. Mater. 252 (1998) J. Garnier, P. Dubuisson, C. Pokor, E. Lemaire, N. Monteil, J. P. Massoud, Relaxation and Irradiation Creep of PWR Baffle Bolt Materials, Fontevraud 7, Contribution of Materials Investigation to the Resolution of Problems Encountered in Pressurized Water Reactors, French Nuclear Energy Society SFEN, September 26-30, J. P. Foster, and A. Boltax, Correlation of Irradiation Creep Data Obtained in Fast Reactor and Thermal Neutron Spectra with Displacement Cross- Sections, J. Nucl. Mater. 89 (1980) F.A. Garner, Malcolm Griffiths, L.R. Greenwood and E.R. Gilbert, Impact of Ni-59 (n,α) and (n,p) Reactions on

7 Dpa Rate, Heating Rate, Gas Generation and Stress Relaxation in LMR, LWR and CANDU Reactors, 14 th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors, Virginia Beach, Virginia, August 23-27, 2009.

8 Table 1. Instrumented sample test matrix. Material Unit Stress (MPa) T ( o C) Test Type SA 304L SS Creep SA 304L SS Creep CW 316 SS N lot Stress Relaxation Table 2. Irradiation creep and irradiation stress relaxation material coefficients. Material Unit T A C Study ( o C) (10-6 /MPa) (10-6 /MPa-dpa) CW 316 SS N Lot This study Ref Ref. 4 SA 304L SS Initial This study Initial This study Average = Subsequent This study Subsequent This study Average = Ref. 6 Table 3. Steady state irradiation creep coefficients calculated using the data reported by Garnier et al. 8. Material Reactor T C ( o C) (10-6 /MPa-dpa) CW 316 SS Osiris BOR SA 304L SS Osiris BOR

9 Figure 1. Schematic of the instrumented sample test unit.

10 Figure 2. Test position in the reactor core. The test rig is located in ring 5 at position S76.

11 ELONGATION, ΔL (um) DOSE (dpa) MPa Figure 3. CW 316 SS/Carpenter heat V87210 (N lot) irradiation stress relaxation (unit 9) tested at 370 o C and 345 MPa. The solid line shows the elongation value used to determine the dose associated with each stress level.

12 ELONGATION, ΔL (um) Densification DOSE (dpa) Figure 4. SA 304L SS irradiation creep (unit 7) tested at 290 o C and 110 MPa. The solid lines show the maximum densification and the approximate net irradiation creep strain rate.

13 ELONGATION, ΔL (um) Densification DOSE (dpa) MPa Figure 5. SA 304L SS irradiation creep (unit 8) tested at 290 o C and 92 MPa. The solid lines show the maximum densification and the approximate net irradiation creep strain rate.

14 STRESS/INITIAL STRESS, σ/σo (fraction) DOSE (dpa) Figure 6. CW 316 SS irradiation stress relaxation tested at 370 o C (Carpenter heat V87210).