B130180, Surveillance Specimen Program Evaluation for River Bend Station
| ML20117K115 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 04/30/1996 |
| From: | Frew B GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20117K092 | List: |
| References | |
| GE-NE-B1301807, NUDOCS 9609110135 | |
| Download: ML20117K115 (44) | |
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i GE Nuclear Energy
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TechnicalServices Business GE-NE-B1301807-02 175 CurtnerAvenue, San Jose, CA 95125
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Surveillance Specimen Program Evaluation
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for
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River Bend Station
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Prepared by:
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B.D. Frew, Engineer Materials Ap lications
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Verified by:
C.L. Chu, Senior Engineer RPV Integrity Approved by:
- 7. M #d4 4'4.r/9g B.J. Branlund, Project Manager
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RPV integrity
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GENucI: rEn:rgy GE-NE-B1301807-02 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully
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The only undertakings of General Electric Company respecting information in this document are contained in the contract between Entergy Operations, Inc.
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(EOI) and General Electric Company, and nothing in this document shall be construed as changing the purchase order. The use of this information by
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anyone other than EOi, or for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized use, General
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Electric Company makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.
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GE Nucl:arEn:rgy GE-NE-B1301807-02 Table of Contents
- 1. ABSTRACT 1
- 2. INTRODUCTION 3
- 3. COMPARISON WITH OTHER SURVEILLANCE DATA 10
- 4. PRESSURE-TEMPERATURE (P-T) CURVES 18 S. SUPPLEMENTAL SURVEILLANCE PROGRAM 21
- 6. REVISED SURVEILLANCE SCHEDULE 23
7.0 CONCLUSION
S 28
8.0 REFERENCES
31 APPENDIX A 32 iii
GE NuclearEn:rgy GE-NE-B1301807-02 Table of Figures FIGURE 2-1: MEASURED SHIFT VS. PREDICTED SHIFT FOR BASE METAL 8
FIGURE 2-2: MEASURED SHIFT VS. PREDICTED SHIFT FOR WELD METAL 9
FIGURE 3-1: MEASURED SHIFT VS. PREDICTED SHIFT FOR BASE METAL 16 FIGURE 3-2: MEASURED SHIFT VS. PREDICTED SHIFT FOR WELD METAL 17 FIGURE 4-1: COMPARISON OF Ku AND K c 20 i
FIGURE 6-I: K m VS. EFPY FOR RIVER BEND WELD MATERIAL 25 i
FIGURE 6-2: Kra VS. PREDICTED SHIFT 26 FIGURE 6-3: PREDICTED SHIFT VS. EFPY, RIVER BEND SURVEILLANCE CAPSULE 27 FIGURE A-1: ART VS. EFPY 38 Table of Tables TABLE 3-1: BWR SURVEILLANCE PROGRAM RESULTS FOR BASE METAL 13 TABLE 3-2: BWR SURVEILLANCE PROGRAM RESULTS FOR WELD METAL 14 TABLE 3-3: FLUX WIRE RESULTS 15 f
TABLE A-1: RIVER BEND RPV MATERIAL DATA 37
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GE Nucl:arEn:rgy GE-NE-B1301807-02 ACKNOWLEDGMENTS The assistance of several people in preparation of this document is greatly appreciated.
The author would like to thank Sam Ranganath, Har Mehta, Ericka Sleight and Tom Caine for valuable technical input to this document, and Sylvia Van Diemen for data compilation, i
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V
GENucinarEn rgy GE-NE-B1301807-02 l
- 1. ABSTRACT i
l River Bend Station (River Bend, RBS) has maintained vessel surveillance programs to meet the requirements of 10CFR50, Appendix H'. The current surveillance program schedule requires that the first surveillance capsule be removed at six (6) Effective Full Power Years (EFPY) for RBS.
I The original schedule was developed in accordance with the requirements of l
10CFR50, Appendix H. This schedule did not account for RBS specific conditions:
Excellent alloy chemistry (Iow copper-0.03-0.09%);
Low RPV 1/4T beltline fluence (<5 X10'8 n/cm 32 EFPY fluence);
2 e
Resulting low shift in the reference nil-ductility temperature (RTsor),
e especially for the RBS plate material.
If the current schedule is used, the measured data for the plate material may not be useful, as the expected shift in RTuor (ARTwor) is low. In addition, the data normally provided by early testing can be obtained from the BWR Owner's Group Supplemental Surveillance Program (SSP). Therefore, the surveillance program's withdrawal schedule should be extended.
The extended schedule can be justified because:
Actual BWR data shows predicted ARTuo7 values based on Reg. Guide 1.99 2
Revision 2 (Rev 2) to bound the measured ARTuor values; The inherent conservatism present in the pressure-temperature (P-T) curves e
for BWR's; 1
l GE NucI:arEri.ngy GE-NE-B1301807-02 I
The derived fracture toughness values are lower bound values and are based on crack arrest (Ka) rather than the higher crack initiation (K c) toughness; i
The SSP will provide early detection of anomalous (i.e., greater than expected) shifts.
I Based on the evaluation presented in this report, the recommended withdrawal l
schedule for the first surveillance capsule for River Bend is 10.4 EFPY.
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l GE Nucle:rEn:rgy GE-NE-B1301807-02
- 2. INTRODUCTION i
Vessel fracture toughness is a major consideration for nuclear vessels; j
irradiation is known to decrease the fracture toughness of vessel materials.
Therefore, measurement of the long term effects of vessel irradiation is a key l
i component of surveillance programs. Entergy Operations, Inc. (EOI) maintains a vessel surveillance program at RBS in accordance with 10CFR50, Appendix H' to monitor for changes in fracture toughness of vessel beltline materials as required by the NRC.
The River Bend surveillance program meets the requirements of 10CFR50, Appendix H and ASTM E185-73 (for design) for the following reasons:
4 i
l The selected base and weld metals are representative of the vessel beltline materials; The materials have a similar fabrication history to the vessel; The number, type, and design of specimens are consistent with ASTM E185-73.
a The surveillance program implemented at River Bend consists of three specimen holders installed in the reactor during vessel construction. The number of holders was determined per ASTM E185-73. River Bend was defined as a case i
'A' plant, since the River Bend vessel had a RTwor hift less than 100 F and the s
peak fluence was less than 5 X10'8 n/cm (at 1/4T) over the design lifetime of the 2
plant.
The three specimen holders were designed, built, and analyzed to ASME 4
Section lil,1971 Edition, with Addenda through Summer 1973. The selection of holder location was based on three criteria to duplicate as closely as possible the 2
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GE NuclearEn:rgy GE-NE-31301807-02 temperature history, neutron flux spectrum, and maximum accumulated RPV beltline fluence:
interference / accessibility with other reactor hardware (e.g., jet pumps);
e peak fluence as a function of height; e
peak fluence as a function of azimuth.
e Using these criteria, the three locations selected were the 3*,177* and 183*
vessel azimuths (available areas not occupied byjet pumps); in addition, a neutron dosimeter was placed at the 3* azimuth. Each holder contains twelve (12) Charpy V-notch specimens of the weld, base metal and heat-affected zone, for a total of 36 specimens. To provide baseline information, archive material is available for additional testing.
At the time the capsules were designed, ASTM E185-73 recommended the first and second capsules to be removed when the capsule fluence reaches 100% of the wall fluence at 32 EFPY. The current testing schedule, developed in accordance with 10CFR50, Appendix H and ASTM E185-82, requires that the first specimen holder be removed at 6 EFPY and the second to be removed at 15 EFPY. The testing and reporting is also to be performed in accordance with ASTM E185-82.
Early capsule withdrawal was recommended for two reasons:
- 1. Data would be provided for future pressure-temperature (P-T) curve calculations. The data would be used to remove conservatism present in the (P-T) calculations. The P-T curves would be recalculated after the first capsule had been removed, using the measured fluence from I
4
GE Nucle:rEnstgy GE-NE-B1301807-02 the flux wire results instead of the fluence calculated from the first cycle flux wire measurements.
- 2. The data obtained from the first capsule would be used to identify any anomalauc, conditions, i.e. a greater than expected shift in RTuor.
However, early withdrawal at 6 EFPY of the RBS capsule is not essential for continued safe operation for the following four reasons:
- 1. Data from other BWR surveillance capsules shows that the RBS first cycle flux wire calculations fall within expected data scatter.
Therefore, the RBS fluence values calculated from first cycle flux wire measurements are appropriate for use in Rev 2 predictions.
- 2. Predicted shifts bound the measured results based on review of predicted RTwor hifts and measured RTuor hifts from other BWR s
s surveillance capsules. Figure 2-1 is a plot of actual shift measurements versus predicted shifts (calculated per Rev 2) for base material. This figure shows that the predicted shift plus margin conservatively bounds the actual shifts measured from BWR surveillance specimen data. The same plot for weld material (Figure 2-2) again shows the predicted shift plus margin term bounds the measured shift.
- 3. Based on actual ART calculations performed in accordance with Rev 2 (see Appendix A), the shift for the River Bend plate is calculated to be 80'F at 32 EFPY. If the first capsule is removed at 6 EFPY, the actual shift may not be large enough to be differentiated from the data scatter, since the predicted fluence on the capsule (8.3 X 10'7 n/cm )
2 5
GE NucIxrEn:rgy GE-NE-B1301807-02 B
is low, and the chemistry of the River Bend vessel plate materialis excellent (0.08-0.09% copper). Thus, the data obtained may not be useful for predicting the material behavior, as it may be indistinguishable from the unirradiated data.
- 4. SSP specimens will provide early tesi data about the limiting RBS weld materials (which are the limiting materials). This program supplements l
the RBS surveillance program by providing timely detection of unusual RT or hifts. The fluences on the SSP capsules are comparable to N
s the fluence for the RBS vessel wall in the time frame of interest.
Although the RBS plate is not in the SSP program, some of the materials used have chemistries which are similar to River Bend, as well as others with poorer chemistries (i.e. higher copper) which result in a larger predicted shift.
I This report shows that the surveillance capsule testing schedule for RBS should be extended for the following reasons:
The fluence experienced by the RBS vessel wall is low; The RBS vessel wall and weld materialin the beltline region has l
excellent alloy chemistry (i.e. Iow copper-0.03-0.09%);
The actual shift in the RBS plate material may not be distinguishable l
from the data scatter with early testing.
The justification for extending the schedule is based on the following reasons:
Predicted shifts bound the actual BWR industry surveillance results; The P-T curve calculations are inherently conservative; I
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GE Nucl::rEn:rgy GE-NE-B1301807-02 The supplemental surveillance program will supplement the RBS surveillance program by providing for the timely detection of unusual RTuor hifts.
s Extension of the surveillance program schedule will ensure that credible data is obtained and continued safe operation of RBS is ensured by using the SSP data and maintaining the RBS P-T curves in accordance with Rev 2.
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GE NuclearEnergy GE-NE-B1301807-02 BWR Weld Predicted Shift vs. Measured Shift 140 120
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10 20 30 40 50 60 70 80 Predicted Shift Figure 2-2: Measured Shift vs. Predicted Shift for Weld Metal 9
l GE NucI::rEn:rgy GE-NE-B1301807-02
- 3. COMPARISON WITH OTHER SURVEILLANCE DATA The evaluation of the shift in the RTuor for River Bend (see Appendix A) was performed using the techniques of Rev 2 for vessel material and the flux wire data from the first cycle (i.e., no additional surveillance data.) These predicted values of RTuor hift indicate that the River Bend vessel will not experience a s
large shift over vessel life. To confirm the conservative predicted shift plus margin values (used to modify the surveillance program schedule), a comparison has been made between calculated shift and fluence values and actual measured surveillance data from other BWR's.
A significant number of surveillance capsules from BWR's have been tested.
Table 3-1 is a tabulation of the base metal results from these surveillance programs. The most significant feature is, for a range of material chemistries and fluences, the expected shift is bounded by the calculated Rev 2 shift plus margin. For example, the measured BWR/6 shifts are less than the predicted Rev 2 shift plus margin values by an average of 38*F. The results for BWR/6 show a small shift (17 F max.) for capsules removed at an EFPY similar to River Bend's current schedule, and at similar fluence levels. This data indicates that l
the River Bend shift (measured at 6 EFPY) will be small and may not be distinguishable from data scatter.
Similarly, Table 3-2 lists surveillance capsule data for weld material. The measured shifts are bounded by the predicted shift plus margin values. BWR/6 weld data shows the predicted shift plus margin to exceed the measured values by an average of 64 F. The maximum shift observed was 35*F, while the predicted shift plus margin was 86 F.
I 10
GENucle:rEn:my GE-NE-B1301807-02 The predicted shift plus margin values are plotted against the measured shifts in Figures 3-1 and 3-2 for all BWR data available; the data is from Tables 3-1 and 3-2, respectively. These graphs show that the measured shifts are bounded by 2
the predicted shift i the margin term. Based on these data, the measured shift for River Bend would be conservatively bounded by the Rev 2 calculation.
Since fluence has a significant effect on the Rev 2 calculation, use of an appropriate fluence value is essential for accurate shift prediction. The shift +
margin predictions in Tables 3-1 and 3-2 utilize fluence values determined from flux wires removed early in plant life; Table 3-3 contains the results of actual BWR flux wire testing. As shown in Table 3-3, for a given BWR type and size, the fluence values fall within the data scatter. For example, for BWR/6 218" vessels (as is River Bend), the 32 EFPY capsule fluences range from 4.44 to 8
2 5.04 X 10 n/cm. Based on this data, the fluence used for the ART calculations (as described in Appendix A) for RBS is considered accurate.
The fluerce used to evaluate the RBS ART was determined from flux wire measurements. The peak ID fluence value of 6.6 X 10'8 n/cm is equal to the 2
value in Paragraph 5.3.1.6.2 of the RBS USAR. Therefore, the actual RBS fluence is conservatively bounded by the calculated value.
Other than fluence, the most significant effect on the ART is the chemistry factor (CF). The CF is determined from the copper and nickel levels, copper having the more significant effect.
4 A study has been performed on the copper levels present in BWR beltline materials, in response to NRC letter 92-01, Supplement 1. The intent was to identify the plants with significant variation in the reported copper levels. For 11
1 GEiv'ucI:arEn:rgy GE-NE-B1301807-02 River Bend, the copper level was determined to be consistent with the reported values with no significant variation.
Based on the evaluation of previous surveillance data of actual shifts and fluences, the measured fluence for RBS, and the chemistry of the RBS vessel material, the actual shift for RBS is expected to be conservatively bounded by the calculated value of shift + margin.
12
GE NuclearEnergy GE-NE-B1301807-02
>lMeV l.99,R EV 2 REV2 39 FT-LB RPV Capseie FLUENCE
@ EFPY DELTA DELTA +
TEST PLANT BWR ID LD.
Ce Ni CF (steal?)
RTNDT MARGIN SHIFT (la)
(4et)
(ekse *2)
BW R/2 AC 2
213 30 0 23 0 46 146 7 3 60
$ $0 35 8 69 5 55 146.7 4 78 7 98 41 9 75 9 32 AS 2
213 110 0 17 0 11 79.5 746 8 15 28 7 62.7 72 B W R/3 H
3 251 215 0 20 0 45 138 0 0 52 6 23 90 41 0 23 AR 3
251 215 0 12 0 54 39 5 0 71 5 98
- 7. 7 41.7 12 AL 3
224 210 0 21 0 49 140 7 3 90 9 00 35.9 69 9 el 300 140.7 6 60 14 to 48 0 82 0 78 A
3 205 30 0 17 0 65 128 3 2 90 7 08 27 6 61 6 Al 3
ISS 10 0.30 0 72 66 0 5.70 6 90 20 7 54.7 0
190 66 0 12 60 15 55 30 7 64.7 2
AG 3
224 30 0 13 0 63 91 8 2 30 4 17 17.2 51.2 25 W
3 253 215 0 20 0.55 143 0 0.55 6 64 10.3 44.3 4
AB 3
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4 251 30 0 14 0 55 98 0 1 52 3 20 14 2 43.2 38 Q
4 2't 30 0.21 0.76 164 6 2.30 6 80 30 9 64 9 52 300 164 6 2 to il 20 34 7 68.7 53 N
4 183 288 O IS 0.70 112.5 4 90 5.90 32 6 66 6 42 C
4 218 30 0 12 0 63 33 5 2 60
$ 98 16 9 50 9 23 K
4 218 30 0 13 0.70 93 5 2 40 5.75 38 0 52 0 58 F
4 218 30 O ct 0 61 58 0 2 30 6 58 96 43 6 3
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251 30 0 09 0 61 58 0 1 40 6 68 80 42.0 24 AT 4
251 30 0 12 0 63 83 0 1 30 6 20 10 8 44 8 2
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205 30 0 11 0 66 74.9 0 43 7.54 4.5 38.5 19 BW RI5 AX 5
251 300 0 14 0 54 97.0 0 90 6 50 99 43.9 28 AZ 251 300 0 10 0 48 65.0 1.15 6 98 78 41.#
N/A BW R16 R
6 218 3
0.029 06 20 84 5 67 7.7 43.7 17 AE 6
218 177 0 06 0.6 37 96 6 85 15 1 49.1 4
AF 6
238 3
0 09 0.58 58 13.0 6 99 25.3 59.3 14 Table 3-1: BWR Surveillance Program Results for Base Metal L
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GE Nuclear Energy GE-NE-B1301807-02 l
>l MeV l.99,REv2 REV1 30 FT.LB RPV Capsule FLUENCE
@EFPY DELTA DELTA +
TEST PLANT BWR ID LD.
Cu Ni CF (alo^t?)
RTNDT MARGIN SHIFT 6e)
(de8)
(%)
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BW R/2 AC 2
213 30 0 17 0.07 81 4.78 5.80 23.1 79.1 N/A 300 81 3.6 7.98 19.8 75.8 N/A AS 2
213 210 0.29 0.05 131.5 7.5 8.15 47.6 103.6 l
BWR/3 Il 3
251 215 0.2 0 45 137 0.52 6.23 9.5 65.5 0
AR 3
251 215 0.2 0.32 II9 0.28 5.98 5.1 61.1 5
AL 3
224 210 0.2 1.05 228.5 39 9 00 58.3 114.3 22 300 66 I4.80 77.4 133.4 76 A
3 205 30 0.05 0.92 68 2.9 7.08 14.6 70.6 A3 3
188 10 0.3 0.09 138 5.7 6.90 43 3 99.3 77 190 12.6 15.85 64.1 120.1 95 AG 3
224 30 0.16 0.79 176.5 2.3 4.17 33.1 89.1 55
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251 215 0.17 0.3 105.5 0.55 6.64 7.6 63 6 0
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251 215 0.16 0.29 100.1 0 66 5 63 8.2 64 2 43 1
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251 30 0.2 0.33 128 1.52 8.20 18.5 74.5 I
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218 30 0.23 0.75 194.5 2.4 6.80 37.4 93.4 61 300 2.8 11.20 41.0 97.0 62 N
4 183 288 0.02 0 95 27 4.9 5.90 7.8 63 8 0
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218 30 0.31 0.72 216 2.6 5.98 43.6 99.6 K
4 218 30 0.28 0.76 212 2.4 5.75 40.8 96.8 F
4 218 30 0.I3 0.I2 68 8 2.3 6 58 12.9 68 9 0
AY 4
251 30 0 08 0.59 105 I.42 6.0I 14.5 70.5 61 P
4 251 I20 0.1 0.32 84.2 I8 7.53 13.6 69.6 17 3
4 251 30 0.11 0 48 102.5 16 7.58 15.3 71.3 16 AW 4
251 30 0.02 0.95 27 1.4 6.68 3.7 59.7 21 AT 4
251 30 0.02 0 95 27 8.8 6.20 4.3 60.3
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4 205 30 0.03 0.93 41 0.43 7.54 2.5 58.5 5
BWR/S AX 5
251 300 0.21 0.78 194 0.9 6.50 19.8 75.8 35 AZ 5
251 300 0 04 0.89 54 1.15 6.98 6.5 62.5 19 BWR/6 R
6 218 3
0.072 0.76 97.5 8.4 5.67 37.3 93.3 28 AE 6
218 177 0.08 0.88 108 9.6 6.85 44.1 100.1 23 AF 6
218 3
0.05 0.87 68 11.0 6.99 29.7 85.7 35 Table 3-2: BWR Surveillance Program Results for Weld Metal l
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GE NuclerrEnugy GE-NE-81301807-02
>l MeV RPV Capsule FLUENCE
@EFPY PLANT BWR ID
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213 30 3.60 5.80 300 4.78 7.98 AS 2
213 210 7.46 8.15 l
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3 251 215 0.52 6.23 AR 3
251 215 0.71 5.98 AL 3
224 210 3.90 9.00 300 6.60 14.80 A
3 205 30 2.90 AJ 3
188 10 5.70 6.90 190 12.60 15.85 AG 3
224 30 2.30 4.17 W
3 251 215 0.55 6.64 AB 3
251 215 0.66 5.63 BWR/4 Y
4 251 30 1.52 8.20 Q
4 218 30 2.30 6.80 300 2.80 11.20 N
4 183 288 4.90 5.90 C
4 218 30 2.60 5.98 K
4 218 30 2.40 5.75 F
4 218 30 2.30 6.58 AY 4
251 30 0.20 1.02 AY 4
251 30 1.42 6.01 P
4 251 120 1.80 7.53 J
4 251 30 1.60 7.58 AW 4
251 30 1.40 6.68 AT 4
251 30 1.30 0
4 205 30 0.43 7.54 BWR/5 AX 5
251 30 0.20 1.38 AX 5
251 300 0.90 6.50 AZ 5
251 30 0.21 1.36 AZ 5
251 300 N/A N/A AK 5
251 30 0.14 0.90 BWR/6 R
6 218 3
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218 177 9.6 6.85 AF 6
218 3
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1.39 1.00 Table 3-3: Flux Wire Results 15
i GE NuclearEnergy GE-NE-B1301807-02 i
Plate Predicted Shift vs. Measured Shift 100 1
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GE NuclearEnergy GE-NE-B1301607-02
- 4. PRESSURE-TEMPERATURE (P-T) CURVES The shift in RTuor btained from surveillance testing is used to evaluate the long o
term effects ofirradiation on the fracture toughness of the vessel. The reference fracture toughness (K a) is determined using the shift in RTuor ; K a is part of the i
i calculations of the P-T curves performed in accordance with ASME Section lil, Appendix G. The current RBS P-T curves were calculated with the 8 EFPY shift in RTuo7 The K ai correlation was developed from several sets of material data on pressure vessel steel.5 The Kai curve was drawn to bound the available data. Thus the correlation has inherent conservatism.
In addition, operation of RBS follows the steam saturation curve, therefore, the operating temperatures are expected to be well in excess of the minimum required temperature. During normal and accident conditions, the RBS maintains more than adequate margins. The operationalissues of Pressurized Thermal Shock (PTS) and Low Temperature Over Pressurization (LTOP) are not applicable to RBS. The limiting case for RBS is the pressure test.
The P-T curve associated with the pressure test is calculated using the crack arrest fracture toughness, K a (K.). The static crack initiation fracture toughness, i
i K c, is significantly higher than K a in the temperature range of interest.
8 i
i Therefore use of K ai conservatively bounds the fracture toughness of the vessel.
7 Figure 4-1 is a plot of K, and K ci as a function of T-RTno7 The K, curve is i
i shown to be lower than the K ci curve, conservatively bounding the fracture toughness. For example, at a pressure test temperature of 169 F and a vessel ART of 65 F (corresponding to 12 EFPY for River Bend), the fracture toughness for initiation and arrest are estimated to be:
18
GE NuclearEn:rgy GE-NE-B1301807-02 K c = 199.2 ksiVin i
K a =83.0 ksiVin i
Thus the K ci value is approximately 2.4 times the Kr value, clearly showing K, to i
conservatively bound the calculations.
The combination of lower bound fracture toughness, the RBS operating characteristics and the conservative fracture toughness values indicate that the RBS vessel fracture toughness is not a significant concem over the life of the l
plant.
i 19
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GE NuclearEn:rgy GE-NE-81301807-02
- 5. SUPPLEMENTAL SURVEILLANCE PROGRAM The BWR Owner's Group (BWROG) is in the midst of a supplemental test program designed to significantly increase the amount of BWR surveillance data in a systematic manner which should permit the development of a BWR-specific equivalent to Rev 2.
Description The BWROG Supplemental Surveillance Program (SSP) was begun in the late 1980s when the BWROG concluded from their review of BWR surveillance data the following:
Due to the smaller number of capsules per plant and the relatively fewer number of BWRs than PWRs, there is limited BWR surveillance data at higher fluences available to analyze; Rev 2 imposed some hardships on pressure testing for BWRs, some of e
which might be relieved if better predictive models of the BVAR embrittlement phenomenon were obtained.
In light of these issues, the BWROG prepared supplemental capsules which were installed in Cooper and Oyster Creek. Specimen withdrawals are planned for 1996,2000, and 2002.
The results of the SSP will be the equivalent of 84 additional surveillance capsules, compared to about 25 which have been tested to date. These capsules were designed to systematically evaluate embrittlement trends in BWRs. For example:
21
GE NuclearEn:rgy GE-NE-B1301807-02 The capsules are positioned so that flux differs by a factor of 2. Also, irradiation times differ by a factor of 2. In this way, some capsules have matching flux but with different fluence, while some have matching fluence and at a differing flux level; i
The materials used were selected to bound the range of chemistries in BWR beltline materials, and in most cases am BWR beltline materials; Irradiations are being done in BWRs to correctly simulate conditions like e
temperature, neutron spectrum and transient operation.
Relationship to River Bend The SSP has the RBS limiting weld material among the materials in the capsules. This materialis in four of the 7 capsules in the SSP holders. The SSP results will be applicable to River Bend for two reasons:
Generically, the SSP results will be from representative environmental conditions on materials representative of all BWRs, including RBS; Specifically, results will be developed which will provide information on the limiting RBS weld material, and will be directly applicable to the River Bend surveillance program.
The SSP capsules, when tested, will have collected between 5x10'7 n/cm 2 18 2
(3.6 EFPY for RBS at 1/4T) and 2x10 n/cm (14.3 EFPY for RBS at 1/4T) fluence. Thus, the results of the SSP are complementary to the RBS surveillance program such that postponement of the capsule withdrawals will have minimal impact on the understanding of irradiation effects on the RBS vessel.
22
GE NucbcrEnergy GE-NE-B1301807-02
- 6. REVISED SURVElLLANCE SCHEDULE The surveillance program is intended to characterize the vessel properties as a function of irradiation over the life of RBS. The Charpy impact energy obtained from the prescribed testing is used to evaluate the referenca fracture toughness of the RBS vessel (Ka)in accordance with ASME Section Ill, Appendix G. The i
schedule for the surveillance program testing should be designed for the expected shift in vessel fracture toughness.
The expected shift in fracture toughness of the RBS weld material (the limiting material) as a function of EFPY is plotted in Figure 6-1. Since the pressure test is the limiting case, the calculated K a is for a 1025 psig pressure test. The i
pressure test temperature was modified on eight year intervals for illustration purposes. This figure shows that significant margin remains between the required K and the Kai used to calculate the P-T curves. Thus the K a is i
i expected to conservatively bound the required vessel fracture toughness.
Since the K a is considered a conservative prediction, and the SSP will identify a i
greater than expected shift, the first surveillance capsule testing should be at the time at which a majority of the shift in the vessel RTwor has been achieved.
Early testing of the surveillance plate specimens may result in the measured shift being less than the data scatter (sometimes resulting in negative shifts in RTwo7).
Correct selection of the removal time will ensure credible data from all specimens. If the shift is greater than expected, then the margin present in the P-T calculations together with the limiting fracture toughness represent an added margin of safety.
Since the SSP can be used to measure anomalous shifts, the first surveillance capsule testing schedule should be developed to measure a significant portion of 23
l GE NucitarEnsmy GE-NE-B1301807-02 the fracture toughness change. To illustrate the change, Figure 6-2 is a plot of the fracture toughness as a function of the predicted shift in RTwo7 or the RBS f
weld material. As is clearly shown, the fracture toughness decreases as a function of the shift. The fracture toughness at the beginning of plant life is 200
-[
kslVin and at 32 EFPY is 72.1 ksiVin. Therefore, the change in fracture toughness over the design life of the plant is 127.9 ksiVin.
f To determine the schedule for first capsule withdrawal, a value of 75% of the h
predicted fracture toughr' ass change ((0.75)(127.9 ksivin)= 95.9 ksiVin) over the design life of RBS was selected as an appropriate criteria. If a significant shift is to occur, this value is large enough to ensure its detectability. Therefore, the first surveillance capsule should be removed when the 75% criteria has been met.
This criteria is met at 200 ksiVin - 95.9 ksiVin, or 104.1 ksiVin. This change in fracture toughness is expected to be achieved when the shift, reading from Figure 6-2, has reached a value of 60'F.
f Since the capsule is intended to measure this shift, the value obtained from
(
Figure 6-2 can be used to determine when the capsule has achieved a similar value. Figure 6-3 is a plot of the shift in RTwor as a function of the capsule
[
EFPY. Using the shift value of 60*F, the capsule will expe'ience a similar shift for the limiting weld material at approximately 10.4 EFPY.
[
Using a criteria of 75% of the change in predicted fracture toughness as the appropriate measurement of vessel embrittlement for RBS, the first surveillance capsule should be removed at 10.4 EFPY. The combination of the low expected
(
shift for the plate material, SSP data on the weld material, and the inherent margin in the K ai calculations will result in a credible set of surveillance data,
(
while ensuring continued safe operation of RBS.
[
r 24
___________a
GE NuclearEnergy GE-NE-B1301807-02 River Bend 114T RPV Weld l
Ki vs EFPY for Pressure of 1025 psig 200.0 180.0 160.0 KIR
\\
= = = Required K1 l
140.0 120.0
( 100.0 80'0
..h
_2D...
...._[
60.0 8 EFPY P-T Curves 16 EFPY P-T Curves 24 EFPY P-T Curves 32 EFPY P-T Curves Pressure Test 149'F Pretsure Test 169'F Pressure Test 182*F Pressure Test 192*F 40.0 l
20.0 0.0 0.0 4.0 8.0 12.0 16.0 20.0 24.0 28.0 32.0 EFPY Figure 6-1: K a vs. EFPY for River Bend Weld Material i
25
GE NuclearEnztgy GE-NE-B1301807-02 River Bend 1/4T Weld KIR @ 32 EFPY 200.0
\\
JL A
180.0
\\
gg g, g
fracture totghness.
Fracture 95.9 kai sq@)
T@s 160 0
\\
Change 127.9 140.0 kai sqrt (in)
N 120 0 sv
......5 g
g_
j 75% of Chay@/
104.1 ksi scrt (in) j QP-~
^
60 0
- y Corresponcing shift 60 F 40.0 20.0 0.0 0
10 20 30 40 50 60 70 80 90 100 Shift (F)
Figure 6-2: K n vs. Predicted Shift i
26
GE NuclearEntigy GE-NE-B1301807-02 River Bend Surveillance Capsule Shift vs. EFPY 100.000 90.000 X
80.000 60 F shift
/
fm Figm 6 2
/
70.000 60.000 -
==- --
C E
50.000
/
l E
d) 40.000 7
~ ~ ~ "
=
30.003
~~
...... Base weld a
20.000
~
10.4 EFPY withdrawal schedule 10.000 a
0.000 0
4 8
12 16 20 24 28 32 EFPY Figure 6-3: Predicted shift vs. EFPY, River Bend Surveillance Capsule 27
GE NucI:arEn:rgy GE-NE-81301807-02
\\
i
7.0 CONCLUSION
S The purpose of the vessel surveillance program is to characterize the vessel properties as a function of irradiation. The original schedule for River Bend was determined according to 10CFR50, Appendix H, resulting in a withdrawal schedule of six (6) EFPY for the first surveillance capsule.
Schedules developed according to 10CFR50, Appendix H, however, are general guidelines for all reactor pressure vessels. The schedules do not take into account some specific characteristics of River Bend, a low fluence and excellent alloy chemistry (0.03-0.09% copper), which results in a low shift in RTuor i
(especially for the RBS plate material). If the first capsule is removed and tested according to the current schedule (6 EFPY), the data obtained for the plate specimens is likely not to be useful.
l I
If the data is unlikely to be useful, and early information on the limiting RBS weld material can be obtained from the SSP to identify anomalous shifts, the
]
surveillance schedule should be extended. The schedule can be extended for the following reasons:
1 1
l
- 1. Evaluation of similar data obtained from actual surveillance programs has shown the predictions of fluence, shift and chemistry are bouno by expected values. In particular, the BWR/6 data has shown small RTuor hifts for s
capsules removed at EFPY similar to the current RBS withdrawal schedule.
Therefore, the suiveillance capsule withdrawal schedule should be extended based on the conservatism in the calculated shift of RTuo7 t
- 2. In addition, the P-T curves contain inherent conservatism, as noted in Section 4. The fracture toughness values used for these calculations are i
28
GE NucI:arEn:rgy GE-NE-B1301807-02 considered to be lower bound values and are significantly less than the crack f
initiation fracture toughness in the temperature range of interest. At j
operating temperatures, RBS maintains more than adequate margins; the I
limiting condition is the pressure test. This conservatism provides an added margin of safety; therefore, the capsule withdrawal schedule can be modified.
- 3. In addition, the SSP data will complement the available data on surveillance l
specimens and also identify any anomalous information in the predicted j
values. This characterization will enhance the understanding of vessel I
embrittlement issues and provide specific data for RBS on the limiting weld j
material. Hence the change in schedule for the RBS surveillance specimens a
j will not have a significant effect on the understanding of vessel irradiation issues.
.i l
These reasons justify extending the withdrawal schedule while maintaining reactor safety margins, and provide for more accurate measured data near EOL.
l Therefore, the surveillance schedule should be modified.
The material property of most concem is the fracture toughness of the vessel; the surveillance schedule should be based on evaluation of this property. Since j
the fracture toughness (K g), is dependent on the shift in RTwo7, the optimum i
j EFPY for removal of the capsule ensures credible data (measuring significant I
shift), while identifying any anomalous conditions. If such an anomalous shift were to occur, the margin between K mi and K n, as well as the inherent i
I conservatism of the calculations, can provide a sufficient safety margin for extending the surveillance schedule. In addition, operation of RBS follows the steam saturation curve; the operating temperatures are expected to be well in excess of the minimum required temperature.
4 29
[
{
GE NuclearEnergy GE-NE-B1301807-02 As shown in section 6, the appropriate K ai value selected was 75% of the predicted change in K n. Using this value to determine the appropriate shift in i
the capsule (hence the appropriate EFPY), the recommended withdrawal schedule for the first River Bend surveillance capsule is 10.4 EFPY. Removal of
(
the capsule at the specified EFPY will obtain the most credible data for fracture toughness predictions.
At this time, a recommended extended schedule for the second surveillance
[
capsule has not been determined. Additional data from the SSP capsules (using the RBS limiting weld material) will soon be available. The first capsule would also provide data on the limiting materials. The combination of these two data sources would be used to develop the appropriate schedule for the second capsule.
{
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30
GE NuclearEnergy GE-NE-B1301807-02
8.0 REFERENCES
- 1. " Reactor Vessel Material Surveillance Program Requirements," Appendix H to Part 50 of Title 10 of the Code of Federal Regulations, December 1995.
- 2. " Radiation Embrittlement of Reactor Vessel Materials," U.S. NRC Regulatory Guide 1.99, Revision 2, May 1988.
- 3. "Implernentation of Regulatory Guide 1.99 Revision 2 for River Bend Station Unit 1," GE Report SASR 89-20, Revision 1, March 1990.
- 4. " Bounding Assessment of BWR/2-6 Reactor Pressure VesselIntegrity Issues," GENE Report #523-A106-1195, BWR VIP-08, November 1995.
- 5. S.T. Rolfe and J.D. Barsom, Fracture and Fatigue Controlin Structures, Prentice-Hall, Inc., New Jersey,1977, p. 447.
- 6. Ibid., p. 455.
- 7. ASME Section XI, Appendix A,1992 Edition through Summer 1993 Addenda.
31 l
GE NuclearEnergy GE-NE-B1301807-02 APPENDIX A ADJUSTED REFERENCE TEMPERATURE (ART) CALCULATION 32
The ART is, according to Rev 2, a function of the initial RTuor, the shift, and a margin term. The shift in RTuoris dependent on the chemistry (specifically copper and nickel) and fluence. The methods of Rev 2 are used to determine the ART; the method used depends on whether or not surveillance specimen data is available.
In order to re-evaluate the surveillance specimen program schedule, the ART for both the vesselitself and the specimens must be calculated. For River Bend, surveillance specimens have not been tested, which requires the method of evaluating ART without surveillance specimens, as described below.
The ART for each beltline material is given by the following equation:
ART = InitialRTuor + ARTuor + Margin (1)
Initial RTuo7 s the reference temperature determined according to ASME i
Section lil, Paragraph NB-2331 for the unirradiated material.
The shift in the reference temperature, ARTuor, is determined by a combination of the chemistry and fluence as shown by equation (2):
ARTw7 = CF
- f @'***** A (2)
The CF is the chemistry factor (dependent on the copper and nickel content) and is determined from the tables for weld and base materialin Rev 2. The fluence, f, at any depth in the vessel wall, is determined by equation (3),
f = f,, * (e" 4')
(3) 33
GE NuclearEnemy GE-NE-B1301807-02 where /,,f s the calculated neutron fluence at the vessel ID and x is the depth i
into the vessel measured from the inner (wetted) surface. For these calculations, 18 2
the value of f,,f used was 6.6 X10 n/cm, obtait:u from the flux wire analysis.
The Margin term is included to obtain the upper bound values of the ART. Since the Margin term provides upper bound values of the ART (which is a function of CF and fluence), it is unnecessary to add extra conservatism by using the upper bound fluence. Any uncertainty in the fluence is captured by the Margin term.
The Margin term is given by equation (4):
Mamin = 24o +o' (4) where ci = standard deviation of the initial RTuor o, = standard deviation for ARTuor The standard deviation for ARTwor, o,, is assumed to be 28 F for welds and 17 F for base metal, except that a, need not exceed 0.50 times the mean ARTuor,2 The conservative nature of the RTuor determination results in o, being equal to zero.
Using equations (1) to (4), the ART can be calculated for plants with no surveillance data, including River Bend.
E< AMPLE CALCULATION To better illustrate the ART methodology, the following calculation was performed for the River Bend base material (Heat #C3054-2); this material was 34
GENucIxrEn:my GE-NE-B1301807-02 used in the surveillance capsule. The data was obtained from Table 5.3-1 of the River Bend USAR:
Initial RTuo7:
2*F Nickel:
0.70%
Copper:
0.09 %
Peak Fluence:
6.6 X 10'8 n/cm (32 EFPY at vessel wall) 2 Wall Thickness:
5.41 inches From Table 2 of Rev 2, the chemistry factor for this heat of materialis 58. The fluence at the 1/4T depth, determined from equation (3), is equal to:
f = (6.6 X10 s/10") *ew241.ss) f =.66*0.723 f = 0.48 The change in reference temperature, ARTuo7, is calculated according to equation (2):
ARTuor = 58 *0.48<.2mio % o.48)
ARTuor = 58*0.796 = 46 For the margin term, the standard deviation of the initial RTuor, oi, is assumed to be zero. The standard deviation for ARTuor, o,, is 17 F, as it is base metal.
Therefore, the ART at 32. EFPY for plate C3054-2 is:
ART = 2 + 46 + 34 = B2 E 35
GE NucI:arEn:tgy GE-NE-B1301807-02 This calculation was repeated for all of the vessel beltline materials. The results of the calculations for all the beltline materials are shown in Table A-1. Figure A-1 is a plot of the ART against EFPY for the expected plant lifetime for the limiting materials, which are the materials with the highest ART after 32 EFPY.
36
g GE NuclearEnergy GE-NE-81301807-02 Base Base Thickness =
5.41 inches 32 EFPY Penk I.D. fluence - 6.60E+18 32 EFPY Peak I/4 T fluence = 4.77E+18 W eld W eld Thickness =
5.41 inches 32 EFPY Peak 1.D. fluence = 6.60E +18 32 EFPY Peak I/4 T fluence = 4.77E+18 Initial 32 EFPY 32 EFPY 32 EFPY COMPONENT ID llEAT OR llEAT/ LOT
%Cu
%Ni CF RTndt Delta RTndt M argin S:iift ART
- F
'F
'F "F
'F BASE:
BELTLINE C313 8-2 0.08 0.63 51 9.0 40.5 34.0 74.5 83.5 C-3054-1 0.09 0.7 58
-20.0 46.0 34.0 80.0 60.0 C3054-2*
0.09 0.7 58 2.0 46.0 34.0 80.0 82.0 VERTICAL WELDS:
492L8471/A421 B27AE 0.04 0.95 54
-60.0 42.9 42.9 85.7 25.7 492L8471/A421 B27AF 0.03 0.98 41
-50 32.5 32.5 65.1 15.1 SP6756/0342(1)*
0.09 0.92 122
-50 96.8 56.0 152.8 102.8 SP6756/0342(2)*
0.09 0.93 122
-60.0 96.8 56.0 152.8 92.8 i
l l
- IlEAT FROM WillCil SURVEILLANCE SPECIMENS WERE TAKEN Table A-1: River Bend RPV Material Data 37 i
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