INFLUENCE OF SPECIMEN SIZE/TYPE ON THE FRACTURE TOUGHNESS ...envdeg2015.org/final-proceedings/ENVDEG/papers/ENVDEG224.pdf · Five different RPV materials were selected for the irradiation

17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors August 9-13, 2015, Ottawa, Ontario, Canada

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INFLUENCE OF SPECIMEN SIZE/TYPE ON THE FRACTURE TOUGHNESS OF FIVE IRRADIATED RPV MATERIALS

Randy K. Nanstad1, Mikhail A. Sokolov1, Enrico Lucon2 1Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN, 37831, USA

2National Institute for Standards and Technology, 325 Broadway, Boulder, CO, 80303, USA; previously at SCK•CEN, Belgian Nuclear Research Center, Mol (Belgium)

ABSTRACT The Heavy-Section Steel Irradiation (HSSI) Program had previously irradiated five reactor pressure vessel (RPV) steels/welds at fast neutron fluxes of about 4 to 8×1011 n/cm2/s (>1 MeV) to fluences from 0.5 to 3.4×1019 n/cm2 and at 288 °C. The unirradiated fracture toughness tests were performed by Oak Ridge National Laboratory with 12.7-mm and 25.4-mm thick (0.5T and 1T) compact specimens, while the HSSI Program provided tensile and 5×10-mm three-point bend specimens to SCK•CEN for irradiation in the in-pile section of the Belgian Reactor BR2 at fluxes >1013 n/cm2/s and subsequent testing by SCK•CEN. The BR2 irradiations were conducted at about 2 and 4×1013 n/cm2/s with irradiation temperature between 295 °C and 300 °C (water temperature), and to fluences between 6 and10×1019 n/cm2. The irradiation-induced shifts of the Master Curve reference temperatures, ∆T0, for most of the materials deviated from the embrittlement correlations much more than expected, motivating the testing of 5×10-mm three-point bend specimens of all five materials in the unirradiated condition to eliminate specimen size and geometry as a variable. Tests of the unirradiated small bend specimens resulted in Master Curve reference temperatures, T0, 25 °C to 53 °C lower than those from the larger compact specimens, meaning that the irradiation-induced reference temperature shifts, ∆T0, were larger than the initial measurements, resulting in much improved agreement between the measured and predicted fracture toughness shifts.

Keywords: bend, compact, dpa, fracture toughness, irradiation, Master Curve, neutron fluence, reactor pressure vessel, specimen size

1. INTRODUCTION The Heavy-Section Steel Irradiation (HSSI) Program [1], sponsored by the U.S. Nuclear Regulatory Commission, had previously irradiated five reactor pressure vessel (RPV) steels and welds at fast neutron fluxes of about 4 to 8×1011 n/cm2/s (>1 MeV) to fluences from 0.5 to 3.4×1019 n/cm2 and at 288 °C. Of specific interest to the HSSI Program was the effect of neutron flux on the irradiation-induced change in fracture toughness. The BR2 research reactor in Belgium, operated by the Belgian Nuclear Research Center SCK•CEN, was well suited for this purpose as it could accommodate the irradiation of mechanical test specimens at the temperatures and fluxes of interest to both HSSI and SCK•CEN programs that investigate effects of irradiation in materials used for commercial fission reactor power plants. To enable a determination of flux effects, the proposed project was that most of the specimens would be irradiated in a position where fast flux would allow for a neutron fluence of about 3 to 4×1019 n/cm2, values close to the fluences from previous results at lower fluxes in other reactors.

2. DESCRIPTION OF MATERIALS Five different RPV materials were selected for the irradiation experiment, with the criteria for selection based on the existence of fracture toughness data at relatively high neutron fluence, an irradiation temperature of about 290 °C, and sufficient material for the experiment and for additional testing in the unirradiated condition if needed. The five materials were the beltline weld for the Midland Reactor Unit 1 (MBW) [2], the Palisades weld (PW) [3], HSST Plate 02 (HSST-02) [4, 5], HSSI Weld 73W (73W) [6],


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and an IAEA Reference Plate JRQ (JRQ) [7]. All three welds were fabricated with the submerged-arc welding process in ASTM A533 grade B class 1steel, while HSST Plate 02 is a 305-mm (12-in.) thick A533 grade B class 1 plate and the JRQ Plate is of the same ASTM specification but is 203-mm (8-in.) thick. All three welds were post-weld heat treated according to the requirements of the ASME Boiler and Pressure Vessel Code. The chemical compositions for those five materials are provided in Table 1. The two chemical elements of most interest in this group of materials are copper and nickel due to their significant influence on irradiation sensitivity and to their variation within the group. The copper contents vary from 0.14 wt % to 0.31 wt %, while the nickel contents vary from 0.57 wt % to 1.2 wt %.

Tensile testing of the unirradiated materials was performed at ORNL in accordance with ASTM E8 [8]; the tensile properties for each material are provided in Table 2, showing yield and ultimate tensile strengths at room temperature from 466 MPa to 512 MPa and from 580 MPa to 627 MPa, respectively.

The initial unirradiated fracture toughness tests were performed by ORNL with 12.7 mm and 25.4 mm thick (0.5 and 1.0 in.) compact specimens [designated 0.5TC(T) and 1TC(T)] machined and fatigue precracked in accordance with ASTM E1921 [9].

For every material, the following specimens were provided to SCK•CEN for inclusion in the irradiation experiment:

Subsize tensile, 2.4 mm diameter, 24 mm long (reduced section = 12 mm long) – 6 each

Three-point bend, 5 mm thick, 10 mm wide, 27.5 mm long, 0.2TSE(B) – 12 each

Atom probe blank, 0.5×0.5 mm, 10 mm long – 1 each

Small-angle neutron scattering blank, 10×10×0.25 mm thick – 1 each

For HSST Plate 02 only, 12 miniature C(T) specimens, 4.2 mm thick, 0.16TC(T), were also provided. About half of the 0.2TSE(B) fracture toughness specimens were fatigue precracked by ORNL prior to irradiation, while the remaining bend specimens were fatigue precracked by SCK•CEN following irradiation. Following irradiation, all the subsize tensile, three-point bend, and compact specimens were tested by SCK•CEN in collaboration with ORNL to determine the ASTM E1921 Master Curve reference temperature, T0.

3. IRRADIATION CONDITIONS To enable a determination of flux effects, the proposed project was that four of the five materials be irradiated in the In-Pile Section 3 (IPS-3) “backside position” of the BR2 because the fast flux of about 2.5×1013 n/cm2/s and one reactor cycle would allow for a neutron fluence of about 3 to 4×1019 n/cm2, close to the fluences from previous results obtained at lower fluxes in other reactors. For the higher flux irradiation of the Palisades weld, it was proposed that the same complement of specimens be inserted for one cycle in the In-Pile Section 2 (IPS-2) location, such that the somewhat higher flux would provide a fluence of about 5×1019 n/cm2. Unfortunately, the irradiation experiment was inadvertently kept in the reactor twice as long as planned resulting in fluences twice as high as the targets, rendering evaluations of neutron flux effects problematic. The irradiation experiment was designated Fusion and Reactor Materials Irradiation SCK•CEN /ORNL – RPV Steels (FRISCO-R) by SCK•CEN. The resulting average values of fast fluence, flux and displacements per atom (dpa) for the irradiated specimens are shown in Table 3. The lower and higher flux irradiations were performed at average fluxes of about 2.4 and 4.8 ×1013 n/cm2/s (E>1 MeV), respectively, with irradiation temperatures between 295 °C and 300 °C (water temp), and to average fluences of 6.45×1019 and 1.05×1020 n/cm2, respectively. The corresponding values of displacements per atom (dpa) are 0.097 and 0.158 dpa, respectively. For the case of the HSST-02 mini-CT specimens, the values of flux, fluence, and dpa are 2.16×1013, 7.20×1019, and 0.108, respectively. All flux, fluence, and dpa values were calculated bySCK•CEN.


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4. TEST RESULTS 4.1 Tensile Tests of Irradiated Specimens

Tensile tests of the irradiated specimens were performed from -150 °C to 300 °C, with the results shown in Figure 1(a) and the room temperature results with change in yield strength with respect to the unirradiated condition shown in Table 4 and Figure 1(b). For fluences of ~ 6 and 10×1019 n/cm2, the yield strength increases (∆σy) averaged 152 ± 13.6 MPa. Figure 2(a) shows that the Palisades weld exhibited less than a 10 % greater increase in yield strength when irradiated to 1.0×1020 compared with that at 6.2×1019 (the flux associated with the higher fluence was also higher but was likely not a significant factor in this flux range), while Figure 2(b) shows some indication of saturation in irradiation-induced yield strength increase vs fluence. Note that the data in Fig. 2(b) are from both tensile specimen tests and from microhardness-to-yield-strength correlations with the equation shown in the figure [10]. As shown in Figure 1(b), there are somewhat disparate results of yield strength changes for the Plate JRQ exhibiting some tendency for saturation followed by increasing yield strength change with fluence.

4.2 Fracture Toughness Tests As stated earlier, the initial unirradiated fracture toughness tests were performed with either 0.5TC(T) or 1TC(T) specimens. The Master Curve reference temperature, T0, for the five materials ranged from -90 °C for the Palisades Weld to -26 °C for HSST Plate 02 based on those specimens. Following testing of the irradiated 0.2TSE(B) specimens, examination of the irradiation-induced ∆T0 values indicated very disparate results relative to those from various predictive models and poor correspondence with typical results in the literature for the correlation between ∆T0 and ∆σy. Such results compelled the machining of additional 0.2TSE(B) specimens for testing in the unirradiated condition so that the same size/type of specimen was used in both unirradiated and irradiated conditions to eliminate specimen size/type as a variable. One example of the results is shown in Figure 3 for HSSI Weld 73W, where Fig. 3(a) shows the ∆T0 based on the unirradiated reference T0 from tests of 1TC(T) specimens, while Figure 3(b) shows that the ∆T0 shift at the high fluence of 6.5×1019 obtained with irradiated 0.2TSE(B) specimens is 53 °C greater when the ∆T0 is based on the reference T0 from the unirradiated 0.2TSE(B) specimen tests. The label with specimen size/type above each bar in the graph indicates the specimen size/type used to obtain the irradiated data. Note that all bar graphs show the flux as the top number (n/cm/s, e.g., 2.0E13), while the fluence is the bottom number (×10-19 n/cm2). Figures 4 (a) and 4 (b) show a similar example for the Midland Beltline Weld, in which case the ∆T0 is 44 °C greater when the same size/type of specimen [i.e., 0.2TSE(B)] was used for both unirradiated and irradiated conditions.

With various predictive embrittlement models applied to each of the five materials, Figures 5(a) and 5(b) show predicted ∆T0 vs measured ∆T0 for the cases with the T0 for the unirradiated condition from the compact specimens and the one with T0 for the unirradiated condition from the 0.2TSE(B) specimens, respectively. Accounting for specimen size/type effects resulted in much better agreement between the measured and predicted fracture toughness shifts. The “predicted-measured” average for the five materials was -33 °C in the case of the unirradiated T0 values from the larger specimens, Fig. 5(a), and -6 °C in the case of the unirradiated T0 values from the small bend specimens, Fig. 5 (b), indicating much better agreement with the various predictive models when specimen size/geometry differences do not create a bias relative to measured irradiation-induced T0 shifts.

Figure 6 shows that the Palisades Weld, in contrast to the tensile yield strength results, Fig. 1(b), exhibited increasing ∆T0 with increasing fluence for both the low and high flux conditions. The results shown for a flux of 6.0×1011 are from [3].

For HSSI Weld 73W, Figure 3(b) shows increasing ∆T0 (~55 °C) with fluence, but the corresponding yield strength change was only 22 MPa. The Midland Beltline Weld, Figure 4(b), showed a similar


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increase in ∆T0 (~45 °C) with fluence and with a corresponding yield strength change of only 13 MPa. Figures 7 and 8 show increasing ∆T0 results with fluence for HSST Plate 02 and Plate JRQ, respectively.

Figure 9 shows the relationship between irradiation-induced T0 increase and yield strength increase for the five irradiated materials with the unirradiated T0 based on the 0.2TSE(B) specimen tests. Although the results show quite large variations from the mean, the slope of the linear fit to the data follows a relationship similar to that reported in [11], with the coefficient also similar to correlations between yield strength increase and Charpy impact 41-J shift in [12, 13].

Table 5 and Figures 10 (a) and (b) show summaries of the results indicating the difference in unirradiated T0 between the original specimens and the 0.2TSE(B) specimens. The comparisons reveal differences in T0 between the small bend and large compact specimens from 25 °C to 53 °C, with the 0.2TSEB specimen providing lower T0 values in every case but one. Similar results of specimen size and geometry effects have been reported in the literature [14, 15, 16]. It is noteworthy that there are quite large differences for the five different materials and, moreover, that the difference for HSST Plate 02 shows the small bend specimen to give a slightly higher T0 value than the larger compact specimen, indicating that there is no specimen size/geometry effect for that material.

There are many papers in the literature that discuss potential reasons of specimen size and geometry effects in the determination of T0, with most of them focusing on the issue of constraint. Figure 11 provides one example of a series of tests with one specific RPV weld that demonstrates significant effects of specimen size/geometry, with results from more than 200 10×10-mm bend specimens [the precracked Charpy V-notch (PCVN) specimen] giving a T0 value 22 °C lower than that from the 1TC(T) specimens (36 data) [16]. References 14 and 15 provide results for an RPV plate with a bias of -36 °C for the PCVN specimen relative to the 1TC(T) specimen. An analysis of the results presented in this paper has not yet been performed relative to loss of constraint, elastic crack-tip T-stress, the Q parameter, etc., but the reader is referred to [17, 18] for detailed discussions of constraint effects and to [19, 20] for examples of specific discussions relative to RPV steel experimental results for bend and compact specimens.

4.3 Microstructural Investigations Atom probe tomography (APT) characterizations were performed on the microstructures of the Midland weld, Palisades weld, Weld 73W, and HSST Plate 02. As previously reported for other neutron-irradiated RPV steels, high number densities of Cu-, Ni-, Mn-, and Si-enriched precipitates were observed [3]. Figure 12 shows an example of an APT examination of one of the Midland Beltline Weld 0.2TSE(B) specimens irradiated to 6.45×1019 n/cm2 in the BR2 reactor. Those precipitates are largely responsible for radiation hardening and embrittlement of the Midland Beltline Weld as well as all other similar RPV steels. Additionally, small-angle neutron scattering experiments have been performed with similar observations obtained with the APT experiments. The detailed results of these microstructural investigations will be reported separately.

5. CONCLUSIONS

Five RPV materials were irradiated at high fluxes [>1013 n/cm2/s (>1 MeV)] in the BR2 reactor to compare with results from previous irradiations at lower fluxes, but comparisons are confounded by significant differences in fluences. The use of 0.2TSEB specimens for irradiation and relatively large specimens (e.g., 0.5T and 1TCT) for determining the unirradiated reference T0 resulted in significantly non-conservative determinations of T0 for most of the materials. These results led to further testing of all five materials in the unirradiated condition with 0.2TSEB specimens to eliminate the potential issue of specimen size/type in the evaluation. The additional testing resulted in the T0 values for four of the five materials to decrease between 25 °C and 53 °C, while the remaining material, HSST Plate 02, exhibited an 8 °C increase in T0 with the smaller specimens. The decreases in T0 resulted in increases of the irradiation-induced shifts in the Master Curve reference temperature, ∆T0, for the four materials such that


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agreement between predicted shifts and measured shifts was significantly improved. Microstructural characterization with atom probe tomography revealed irradiation-induced embrittlement caused largely by high number densities of Cu-, Ni-, Mn-, and Si-enriched precipitates.

6. ACKNOWLEDGEMENTS The authors acknowledge the assistance of Eric Manneschmidt and Ronald Swain for mechanical testing at ORNL, and the collaboration of Roel Vanuytven, Jan Knaeps and Paul Wouters for execution of the mechanical tests at SCK•CEN. We further acknowledge the U.S. Department of Energy Light-Water Reactor Sustainability Program and Dr. Jeremy Busby, program manager, for support in preparation of this document. The experiments reported in this document resulted from research under the Heavy-Section Steel Irradiation Program sponsored by the Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission under Interagency Agreement 1886-T778-08 with the U.S. Department of Energy. This paper was prepared by Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, Tennessee 37831-6285, managed by UT-Battelle, LLC, for the U.S. Department of Energy, under contract DE-AC05-00OR22725. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes.

REFERENCES

[1] R. K. Nanstad, B. R. Bass, J. G. Merkle, C. E. Pugh, T. M. Rosseel, and M. A. Sokolov, “ Heavy-section Steel Technology and Irradiation Programs-Retrospective and Prospective Views,” J. Pressure Vessel Technology, December 2010, Vol. 132/064001-1.

[2] R. K. Nanstad, D. E. McCabe, and R. L. Swain, “Evaluation of Variability in Material Properties and Chemical Composition for Midland Reactor Weld WF-70,” pp. 125-156 in Effects of Radiation on Materials: 18th International Symposium, ASTM STP 1325, R. K. Nanstad, M. L. Hamilton, F. A. Garner, and A. S. Kumar, Eds., American Society for Testing and Materials, West Conshohocken, PA, 1999.

[3] M. K. Miller, M. A. Sokolov, R. K. Nanstad and K. F. Russell, “APT Characterization of High Nickel RPV Steels,” J. Nucl. Mater. 351 (2006) 187-196.

[4] Canonico, D. A., and Berggren, R. G., 1971, “Tensile and Impact Properties of Thick-Section Plate and Weldments,” Nucl. Eng. Des., 17(1), pp. 4-15.

[5] Stelzman, W. J., Berggren, R. G., and Jones, Jr., T. N., “ORNL Characterization of Heavy-Section Steel Technology Program Plates 01, 02, and 03,” NUREG/CR-4092 (ORNL/TM-9491), Oak Ridge National Laboratory, Oak Ridge, TN, April 1985.

[6] R. K. Nanstad, D. E. McCabe, B. H. Menke, S. K. Iskander, and F. M. Haggag, "Effects of Irradiation on KIc Curves for High Copper Welds," pp. 214-33 in Effects of Radiation on Materials: 14th International Symposium (Volume II), ASTM STP 1046, proceedings of symposium held at Andover, Massachusetts, June 27-29, 1988, N. H. Packan, R. E. Stoller, and A. S. Kumar, Eds., American Society for Testing and Materials, Philadelphia, 1990.

[7] R. K. Nanstad, Ph. Tipping, R. D. Kalkhof, and M. A. Sokolov, “Irradiation and Post-Annealing Reirradiation Effects on Fracture Toughness of RPV Steel Heat JRQ,” in The Effects of Radiation on Materials: 21st International symposium, ASTM STP 1447, M. L. Grossbeck, T. R. Allen, R. G. Lott, and A. S. Kumar, Eds., ASTM International, West Conshohocken, PA, December, 2004.

[8] Standard Test Methods for Tension Testing of Metallic Materials, ASTM E8-01, ASTM International, West Conshohocken, PA, United States, 2002


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[9] Standard Test Method for Determination of Reference Temperature, T0, for Ferritic Steels in the Transition range, ASTM E 1921-02, ASTM International, West Conshohocken, PA, United States, 2002.

[10] Mancuso, J. F., Spitznagel., J. A., Shogan, R. P. and Holland, J. R., “CorrelationBetween Microhardness, Tensile Properties, and Notch Ductility of Irradiated Ferritic Steels,” Effects of Radiation on Materials: Tenth International Symposium, ASTM STP 725, D. Kramer, H. R. Brager and J. S. Perrin, Eds., American Society for Testing and Materials, Philadelphia, PA, United States, 1981.

[11] Sokolov, M. A., and Nanstad, R. K., “Comparison of Irradiation-Induced Shifts of KJc and Charpy Impact Toughness For Reactor Pressure Vessel Steels,” Effects of Radiation on Materials: 18th International Symposium, ASTM STP 1325, R. K. Nanstad, M. L. Hamilton, F. A. Garner, and A. S. Kumar, Eds., American Society for Testing and Materials, West Conshohocken, PA, 1999.

[12] R. K. Nanstad and R. G. Berggren, "Effects of Irradiation Temperature on Charpy and Tensile Properties of High Copper, Low Upper Shelf, Submerged Arc Welds," Effects of Radiation on Materials: 16th International Symposium, ASTM STP 1175, A. S. Kumar, D. S. Gelles, R. K. Nanstad, and E. A. Little, Eds., American Society for Testing and Materials, Philadelphia, 1993.

[13] G. R. Odette, P. M. Lombrozo, and R. A. Wullaert, “Relationship Between Irradiation Hardening and Embrittlement of Pressure Vessel Steels,” pp. 840-860 in Effects of Radiation on Materials: 12th International Symposium, ASTM STP 870, F. A. Garner, J. S. Perrin, Eds., American Society for Testing and Materials, 1985.

[14] R. K. Nanstad, M. A. Sokolov, J. G. Merkle, and D. E. McCabe, “Experimental Evaluation of Deformation and Constraint Characteristics in Precracked Charpy and Other Three-Point Bend Specimens,” Proc. ASME 2007 Pressure Vessel and Piping Conference and Eighth International Conference on CREEP and Fatigue at Elevated Temperatures, San Antonio, Texas, July 22-26, 2007.

[15] Randy K. Nanstad, et. al., “IAEA Coordinated Research Project on Master Curve Approach to Monitor Fracture Toughness of RPV Steels: Final Results of the Experimental Exercise to Support Constraint Effects,” Proceedings of the ASME PVP2009, PVP2009-78022, Prague, Czech Republic, July 2009.

[16] Van Der sluys, W. A., Merkle, J. G., Young, B., and Nanstad, R. K., “Part I: Results from the MPC Cooperative Test Program on the Use of Precracked Charpy Specimens for T0 Determination,” Indexing Fracture Toughness Data, Welding Research Council Bulletin 486, November 2003.

[17] Anderson T. L., Fracture Mechanics – Fundamentals and Applications, Third Edition, CRC Press, Taylor and Francis Group, Boca Raton, 2005.

[18] Wallin, Kim, Fracture Toughness of Engineering Materials - Estimation and Application, EMAS Publishing – a FESI Subsidiary, Whittle House, Warrington, WA3, 6FW, UK, 2011

[19] Wasiluk, B., Petti, J., and Dodds Jr, R. H., “Constraint Differences Between C(T) and SE(B) for the Euro-material,” Presentation at ASTM E08.08.03 Meeting in Salt Lake city, Utah, May 2004.

[20] Joyce, J. A. and Tregoning, R. L., “Quantification of Specimen Geometry Effects on the Master Curve and T0 Reference Temperature,” ECF13, San Sebastian, Spain, 2000.


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Table 1. Chemical composition of the RPV steels irradiated in the FRISCO-R experiment.

Material C Mn Si S P Cr V Cu Mo Ni W Al

PW 0.11 1.25 0.18 0.017 0.014 0.04 0.003 0.20 0.55 1.2 - -

MBW 0.09 1.607 0.622 0.009 0.017 0.12 0.005 0.256 0.41 0.574 <0.01 0.015

HSST-02 0.23 1.55 0.20 0.014 0.009 0.04 0.003 0.14 0.53 0.67 <0.01 0.019

73W 0.10 1.56 0.45 0.005 0.005 0.25 0.003 0.31 0.58 0.60 <0.01 0.005

JRQ 0.18 1.42 0.24 0.004 0.017 0.12 0.002 0.14 0.51 0.84 <0.01 0.014

Table 2. Room temperature tensile properties of RPV steels irradiated in the FRISCO-R experiment.

Material Yield Strength, MPa

Tensile Strength, MPa

Total Elongation, %

Reduction of Area, %

MBW 512 586 Not determined

Not determined

PW 470 580 Not determined

Not determined

HSST-02 466 614 21 66

73W 490 599 22 68

JRQ 487 627 25 74


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Table 3. Agerage values of fast fluence, flux and dpa for the irradiated specimens.

Material Specimen

Type

Fluence

(n/cm2)

Flux

(n/cm2)

dpa

PW

(High Fluence)

Tensile

0.2TSE(B)

0.97 x 1020

1.08 x 1020

4.45 x 1013

4.93 x 1013

0.146

0.161

PW

(Low Fluence)

Tensile

0.2TSE(B)

6.18 x 1019

6.51 x 1019

1.85 x 1013

1.95 x 1013

0.093

0.098

MBW Tensile

0.2TSE(B)

6.18 x 1019

6.51 x 1019

1.85 x 1013

1.95 x 1013

0.093

0.098

HSST-02 Tensile

0.2TSE(B)

MC(T)

6.04 x 1019

6.49 x 1019

7.84 x 1019

1.81 x 1013

1.95 x 1013

2.35 x 1013

0.091

0.097

0.118

73W Tensile

0.2TSE(B)

6.18 x 1019

6.51 x 1019

1.85 x 1013

1.95 x 1013

0.093

0.098

JRQ Tensile

0.2TSE(B)

6.18 x 1019

6.51 x 1019

1.85 x 1013

1.95 x 1013

0.093

0.098


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Table 4. Average tensile yield strengths (YS) at room temperature (R.T.) and increase (∆YS) due to irradiation at the irradiation flux and fluence for each material.

Material Flux n/cm2/s

Fluence n/cm2

Unirr YS MPa

Irr YS MPa

∆YS at R.T., MPa

Palisades Weld

6.0E+11 1.40E+19 470 620 150

Palisades Weld

6.0E+11 3.40E+19 470 620 150

Palisades Weld

2.0E+13 6.00E+19 470 618 148

Palisades Weld

4.40E+13 1.00E+20 470 623 153

Weld 73W 3.0E+12 1.5E+19 494 648 154

Weld 73W 1.8E+13 6.2E+19 494 670 176

Midland Beltline

Weld

6.0E+11 5.0E+18 512 632 120

Midland Beltline

Weld

6.0E+11 1.0E+19 512 644 132

Midland Beltline

Weld

6.0E+11 3.4E+19 512 575 63

Midland Beltline

Weld

1.8E+13 6.2E+19 512 657 145

Plate 02 6.0E+11 2.1E+19 460 592 132

Plate 02 1.8E+13 6.0E+19 460 604 144

JRQ Plate (Serrano)

3.6E+18 477 563 86

JRQ Plate (IAEA CRP5)

3.8E+19 477 570 93

JRQ Plate (PSI/ORNL)

5.0E+12 5.0E+19 477 581 104

JRQ Plate 1.8E+13 6.2E+19 477 623 146


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Table 5. Master Curve T0 results for five RPV materials determined with different size/type specimens, with the small specimen bias indicated, and ∆T0 for each material with the 0.2TSEB specimens irradiated in the BR2 reactor to ~ 6.5x1019 n/cm2 (>1 MeV).

Material Unirradiated

T0, °C

Bias, °C

0.2TSEB

Irradiated

∆T0, °C

0.2TSEB 0.5T&1TCT 0.2TSEB

Palisades Weld -90 -115 -25 135

Weld 73W -63 -116 -53 155

Midland Weld -54 -98 -44 127

Plate 02 -26 -18 8 112

JRQ Plate -54 -79 -25 134

(a) (b)

Figure 1. Tensile test results for five RPV materials irradiated in the BR2 reactor, showing (a) yield strength in the irradiated condition, and (b) the irradiation-induced increases in room temperature yield strength with fluence.


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(a) (b)

Figure 2. Tensile results for Palisades weld at two flux/fluence conditions show (a) higher yield strength at higher fluence as expected, and (b) indications of “saturation” at very high fluence.

(a) (b)

Figure 3. Example with HSSI Weld 73W showing (a) the ∆T0 at 1.5 and 6.5×1019 based on T0 from unirradiated 1TC(T) specimens, and (b) the ∆T0 at 6.5×1019 based on T0 from unirradiated 0.2TSEB specimens. The T0 from 0.2TSEB specimens was 53 °C lower than that determined with 1TC(T) specimens.


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(a) (b)

Figure 4. Example with Midland Beltline Weld showing (a) the ∆T0 at 0.5, 1.0, 3.4 and 6.5×1019 based on T0 from unirradiated 0.5TC(T) and 1TC(T) specimens, and (b) the ∆T0 at 6.5×1019 based on T0 from unirradiated 0.2TSEB specimens. The T0 from 0.2TSEB specimens was 44 °C lower than that determined with 1TC(T) specimens.

(a) (b)

Figure 5. Comparisons of predicted ∆T0 and measured ∆T0 at ~ 6.5×1019 n/cm2 with (a) unirradiated T0 based on 0.5TC(T) and 1TC(T) specimens, and (b) unirradiated T0 based on 0.2TSEB specimens. Average “predicted-measured” values were significantly improved with unirradiated T0 from the small bend specimens.


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Figure 6. In contrast to tensile results, fracture toughness for Palisades weld, irradiated at low and high flux, exhibits increasing ∆T0 with increasing fluence.

Figure 7. Fracture toughness for HSST Plate 02 irradiated at low and high flux shows increasing ∆T0 with fluence.


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Figure 8. Fracture toughness for Plate JRQ irradiated at low and high flux shows increasing ∆T0 with fluence.

Figure 9. Relationship of ∆T0 to ∆σy for the five irradiated

materials is similar to other published data.


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(a) (b)

Figure 10. Bar graphs showing (a) reference temperature, T0, for five RPV materials based on different specimen sizes, and (b) the small specimen bias in T0 for each material.

Figure 11. Fracture toughness results from ORNL tests with 1TC(T) specimens of unirradiated HSSI Weld 72W. The master curves are shown for Weld 72W 1TC(T) and the MPC PCVN round robin test program with more than 200 PCVN specimens, indicating a PCVN bias of -21 °C.


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(a)

(b)

Figure 12. Atom probe tomography of Midland Beltline Weld irradiated to 6.45×1019 n/cm2 revealed (a) irradiation-induced precipitates consisting of Cu, Ni, Mn, Si and with small amounts of P, and (b) Si, P, and Mn segregation to grain boundaries.

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INFLUENCE OF SPECIMEN SIZE/TYPE ON THE FRACTURE TOUGHNESS ...envdeg2015.org/final-proceedings/ENVDEG/papers/ENVDEG224.pdf · Five different RPV materials were selected for the irradiation