ML20108D086

Surveillance Specimen Program Evaluation for Grand Gulf Nuclear Station

ML20108D086
Person / Time
Site:	Grand Gulf
Issue date:	04/30/1996
From:	Branlund B, Chu C, Frew B GENERAL ELECTRIC CO.
To:
Shared Package
ML20108D076	List: ML20108D074 ML20108D086
References
GE-NE-1301807-01-R01, GE-NE-1301807-1-R1, GE-NE-B1301807, NUDOCS 9605070269
Download: ML20108D086 (42)

Text

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GE Nuclear Energy

]

Technical Services Business GeneralElectnc Company GE-NE-B1301807-01R1 175 Curtner Avenue, San Jose, CA 95125 April 1996 l

Surveillance Specimen Program Evaluation 1

for Grand Gulf Nuclear Station l

l Prepared by:

B.D. Frew, Engineer  !

Materials Applications l Verified'by: 1M,.3 <- O.L. b b C.L. Chu, Senior Engineer  !

RPV Integrity l

Approved bM. W_

B.J. Branlund, Project Manager <

RPV Integrity 9605070269 960502 PDR ADOCK 05000416 P PDR

l GE NucIzar En:rgy GE-NE-B1301807-01R1 \

. 1 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The only undertakings of General Electric Company respecting information in  ;

1 this document are contained in the contract between Entergy Operations, Inc. '

(EOl) and General Electric Company, and nothing in this document shall be construed as changing the purchase order. The use of this information by 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 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.

I il

GE NucI:arEn:rgy GE-NE-B1301807-01R1 Table of Contents

1. ABSTRACT 1

2. INTRODUCTION 3

3. COMPARISON WITII OTIIER SURVEILLANCE DATA 10

4. PRESSURE-TEMPERATURE (P-T) CURVES 17

5. SUPPLEMENTAL SURVEILLANCE PROGRAM 20

6. REVISED SURVEILLANCE SCIIEDULE 22 t

7.0 CONCLUSION

S 27

8.0 REFERENCES

30 APPENDIX A 31

. I i

lll l

\

GE Nuct:arEnergy GE-NE-B1301807-01R1 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 8 FIGURE 3-1: . MEASURED SHIFT VS. PREDICTED SHIFT FOR BASE METAL 15 FIGURE 3-2: MEASURED SHIFT VS. PREDICTED SHIFT FOR WELD METAL 16 FIGURE 4-1: COMPARISON OF Ku AND K ic 19 FIGURE 6-1: Ki n VS. EFPY FOR GRAND GULF WELD MATERIAL 24 FIGURE 6-2: K ia VS. PREDICTED SHIFT 25 FIGURE 6-3: PREDICTED SHIFT VS. EFPY, GRAND GULF SURVEILLANCE CAPSULE 26 FIGURE A-1: ART VS. EFPY 37 Table of Tables.

TABLE 3-1: BWR SURVEILLANCE PROGRAM RESULTS FOR BASE METAL 12 TABLE 3-2: BWR SURVEILLANCE PROGRAM RESULTS FOR WELD METAL 13 TABLE 3-3: FLUX WIRE RESULTS 14 TABLE A-1: GRAND GULF RPV MATERIAL DATA 36 4

i I

iV

GE NuclearEn:rgy GE-NE-B1301807-01R1 l 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.

V

GENucIsarEn rgy GE-NE-B1301807-01R1

1. ABSTRACT l l

Grand Gulf Nuclear Station (Grand Gulf, GGNS) 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 eight (8) Effective Full Power Years (EFPY) for GGNS.

The original schedule was developed in accordance with the requirements of 10CFR50, Appendix H. This schedule did not account for GGNS specific conditions: l

. Excellent allcy chemistry (low copper of 0.02-0.0S%)

18 2 Low RPV beltline fluence (<5 X10 n/cm 32 EFPY fluence) ,

1 e Resulting low shift in the reference nil-ductility temperature (RTwor). '

If the current schedule'is used, the measured data may not be useful, as the expected shift in RTNor (ARTuor) is low. Therefore, the surveillance program's withdrawal schedule should be extended.

The extended schedule can be justified because:

. Actual BWR data shows predicted ARTwor values based on Reg. Guide 1.99 2

Revision 2 (Rev 2) to bound the measured ARTwor values- I i

. The inherent conservatism present in the pressure-temperature (P-T) curves  ;

for BWR's; '

. The derived fracture toughness values are lower bound values and are based on crack arrest (K.)

i rather than the higher crack initiation (K ic) toughness.

1

1 GE NuclearEn:rgy GE-NE-B1301807-01R1 Based on the evaluation presented in this report, the recommended withdrawal schedule for the first surveillance capsule for Grand Gulf is 24 EFPY.

1 1

1 l

I 2

GE NucI:arEnergy GE-NE-B1301807-01R1

2. INTRODUCTION l

l Vessel fracture toughness is a major concern for nuclear vessels; irradiation is known to decrease the fracture toughness of vessel materials. Therefore, measurement of the long term effects of vessel irradiation is a key component of surveillance programs. Entergy Operations, Inc. (EOI) maintains a vessel surveillance program at GGNS in accordance with 10CFR50, Appendix H' to meet the requirements of the NRC. l The Grand Gulf surveillance program meets the requirements of 10CFR50, Appendix H and ASTM E185-73 (for design) for the following reasons:

. The selected base metal and weld metal are the limiting beltline plate and weld materials;

. The materials have a similar fabrication history to the vessel;

. The number, type, and design of specimens are consistent with ASTM E185-73. j The surveillance program implemented at Grand Gulf consists of three specimen holders installed in the reactor during vessel construction. The number of holders was determined per ASTM E185-73; Grand Gulf is defined as a case 'A' l

plant since the Grand Gulf vessel has a RTuors hift less than 100'F and will be exposed to a fluence of less than 5 X10'8 n/cm2 over the design lifetime of the '

plant.

The three specimen holders were designed, built, and analyzed to ASME Section Ill,1971 Edition, with Addenda through Winter 1972. The selection of holder location was based on three criteria to duplicate as closely as possible the 3

GE NuclDarEnergy GE-NE-B1301807-01R1 temperature history, neutron flux spectrum, and maximum accumulated RPV beltline fluence:

. interference / accessibility with other reactor hardware (e.g., jet pumps).

. peak fluence as a function of height; I e peak fluence as a function of azimuth; 1

i Using these criteria, the three locations selected were the 3 ,177* and 183 1

vessel azimuths (available areas not occupied by jet pumps); in addition, a neutron dosimeter was placed at the 3* azimuth. Each holder contains twelve I (12) Charpy V-notch specimens of the weld, base metal and heat-affected zone , l for a total of 36 specimens. To provide baseline information, a set of unirradiated l specimens are kept, as well as archive material for additional testing. )

The current testing schedule, developed in accordance with 10CFR50, Appendix H, requires that the first specimen holder be iemoved at 8 EFPY and i the second to be removed at 24 EFPY; the testing and reporting is to be i performed in accordance with ASTM E185-82. For a case 'A' plant, ASTM E185-73 recommends the first and second capsules to be removed when the capsule fluence reaches 100% of the wall fluence. Since the Grand Gulf surveillance holders are unlikely to reach this fluence during the lifetime of the plant, a 25%

and 75% of design life criteria (similar to a case 'B' plant) was used to develop the 8 and 24 EFPY schedule for the first two capsules.

Based on actual ART calculations performed in accordance with Rev 2 (see 1 Appendix A), the ARTwor for Grand Gulf is expected to be low (<50 F). If the first capsule is removed at 8 EFPY, the actual shift may not be large enough to be distinguished from the data scatter, as a result of the low fluence on the capsule (2.25 X10" n/cm 2) and chemistry of the Grand Gulf vessel material. Thus, the 4 l

GE NuclearEnergy GE-NE-B1301807-01R1 data obtained may not be credible for predicting the material behavior, as it may l be indistinguishable from the unirradiated data.

I j If the requirements of ASTM E185-82 were applied to determine the schedule, Grand Gulf would be defined as a case 'A' plant (<100 F ARTuor <5 X10'8 n/cm 2 l lifetime fluence.) The schedule for a case 'A' plant indicates that the first capsule i should be withdrawn when the vessel wall fluence is 5 X 1018 n/cm2, or when the l ARTwor reaches 50 F, whichever is first. The Grand Gulf vessel wall is unlikely to reach these conditions during the design lifetime of the plant; therefore, early capsule withdrawal is not entical to continued operation. 1 Early capsule withdrawal was recommended for two reasons:

i (1) Data would be provided for revised 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 the surveillance capsule 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 anomalous conditions, i.e. a greater than expected shift in RTsor. I However, early withdrawal at 8 EFPY of the GGNS capsule is not essential for the following three reasons:

1. Data from other BWR surveillance capsules shows that the GGNS first cycle flux wire calculations fall within expected data scatter.

5

GENucI:arEn rgy GE-NE-B1301807-01R1 Therefore, the GGNS 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 RTuorshifts and measured RTN or shifts from other BWR surveillance. 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 predictec. shift plus margin conservatively bounds the actual shifts measured from 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. The GGNS surveillance program is enhanced by the BWROG's supplemental surveillance program (SSP). The SSP contains the I Grand Gulf limiting weld and plate beltline materials. This program l supplements the GGNS surveillance program by providing timely detection of unusual RTuorshifts. The fluences on the SSP capsules are comparable to the end-of-life (EOL) fluence for the GGNS vessel wall.

This report shows that the surveillance capsule testing schedule for GGNS should be extended for the following reasons:

. The fluence experienced by the GGNS vessel wall is low;

• The GGNS vessel wall and weld material in the beltline region has excellent alloy chemistry (i.e. Iow copper of 0.02-0.06%);

e The actual shift may not be distinguishable from the data scatter with early testing.

The justification for extending the sche,dule is based on the following reasons:

6

GE Nucl:arEn:rgy GE-NE-B1301807-01R1

. Predicted shifts bound the actual BWR industry surveillance results; I

. The P-T curve calculations are inherently conservative;

. The supplemental surveillance program will enhance the GGNS surveillance program by providing for timely detection of unusual RTwor  !

shifts Extension of the surveillance prograrn schedule will ensure that credible data is obtained and continued safe operation of GGNS is ensured by maintaining the GGNS P-T curves in accordance with Rev 2. i 1

1 l

l l

1 7

GE Nuclear Energy a5.yg.g339;gg79gg; Plate Predicted Shift vs. Measured Shift 100 - - -

A ,

80 *_,_

, .

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]:s a Measured g ag g 20 Predicted g A ,

A l

l l

~**,,,

a ga a A * . .

0 t s s ,__.

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-20 a *

-40 0 5 10 15 20 25 30 35 40 45 50 Predicted Shift Figure 2-1: Measured Shift vs. Predicted Shift for Base Metal 8

. e ,

GE NuclearEnergy GE-NE-B1301807-01R1 BWR Weld Predicted Shift vs. Measured Shift 140 120 ,-

100 s-+-

• • A ,

80 -

~

e 60 -

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• Predicted

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0 10 20 30 40 50 60 70 80 t Predicted Shift Figure 2-2: Measured Shift vs. Predicted Shift for Weld Metal 9

b GENuclearEn rgy GE-NE-B1301807-01R1

3. COMPARISON WITH OTHER SURVEILLANCE DATA The evaluation of the shift in the RTuor for Grand Gulf (see Appendix A) was 4

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 RTwor shift indicate that the Grand Gulf vessel will not experience a 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 1

surveillance data from other BWR's.

A significant number of surveillance capsules from BWR's have been tested.

4 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 bound by the calculated Rev 2 shift plus margin. For example, the actual 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 Grand Gulf's current schedule, and at higher fluence levels.

Similarly, Table 3-2 lists surveillance capsule data for weld material. The measured shifts are bound 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.

The predicted shift (plus margin values) versus the measured shift, plotted previously in section 2, are repeated in Figures 3-1 and 3-2 for all available BWR data; the data is from Tables 3-1 and 3-2, respectively. These graphs show that j 10

s GE Nuclear Enorgy GE-NE-B1301807-01R1 the actual shifts are bound by the predicted shift i the margin term. Based on these data, the measured shift for Grand Gulf would be conservatively bound by the Rev 2 calculation.

Since fluence has a significant effect on the Rev 2 calculation, use of an appropriate fluence value is essential to shift predictions. 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 l flux wire testing. This data indicates that, for a given BWR type and size, the fluence values fall within the expected data scatter. For example, for BWR/4 and 2

6 251" vessels, the 32 EFPY fluences range from 5.9 to 9.0 X 10'7 n/cm . Based l l

on this data, the fluence used for the ART calculations (as described in Appendix l

l A) for GGNS is considered accurate. The fluence used to evaluate the GGNS ART was determined from flux wire measurements ; the peak fluence value used was 2.5 X 1018 n/cm 2 1

1 Other than fluence, the most significant effect on the ART is the chemistry factor (CF). The CF is determined from the copper and nickellevels, copper having the more significant effect. 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 Grand Gulf, the copper level was determined to be consistent with the reported values with no significant variation.

l Based on the evaluation of previous surveillance data of actual shifts and 1

fluences, the measured fluence for GGNS, and the chemistry of the GGNS vessel material, the actual (low) shift for GGNS is expected to be conservatively bound by the calculated value of shift + margin.

11

GE NuclearEnergy GE-NE-B1301807-01R1

>l MeV I .99.R EV2 REV2 38 FT-LB RPY Capnele FLUENCE EEFPY DELTA DELTA + TEST PLANT BWR ID I D. Ce Ni CF (al8'l?) RTNDT M ARGIN SHIFT (in) (deg) (m/cm ^2)

BW R/2 AC 2 213 30 0 23 0.46 146 7 3 60 5 80 35 8 69 8 55 146.7 4 78 7 98 41.9 75 9 82 AS 2 213 210 0 17 0 11 79 5 7 46 8 15 28 7 62.7 72 BH R/3 H 3 251 215 0 20 0 45 131.0 0.52 6 23 90 43 0 23 AR 3 251 215 0.12 0 54 89 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 61 300 140 7 6 60 14 80 48 0 82 0 78 A 3 205 30 0 17 0 65 128 3 2 90 7.08 27 6 61 6 A3 3 188 to 0 10 0.72 66 0 5 70 6 90 20.7 54 7 0 190 66 0 12 60 15 85 30 7 64 7 2 AO 3 224 30 0 13 0 63 91 8 2.30 4 17 17 2 51 2 25 W 3 251 215 0.20 0 55 143 0 0.55 6 64 10 3 44 3 4 AB 3 251 215 0.10 0.54 65 0 0 66 5 63 53 39 3 -2 BW R/4 Y 4 251 30 0 14 0 55 98 0 l.52 8 20 14 2 48 2 38 Q 4 218 30 0 21 0 76 164 6 - 2.30 - 6 80 30 9 64 9 52 300 164 6 2 80 $ 1.20 34 7 68 7 $3 N 4 '183 288 0 15 0 70 112 5 4 90 5 90 32 6 66 6 42 C 4 218 30 0 12 0 63 83.5 2 60 5 98 16 9 50 9 23 K 4 218 30 0 13 0 70 93 5 2 40 5 75 IS O 52 0 58 F 4 218 30 0 08 0 63 51 0 2.30 6 58 96 43 6 3 AY 4 251 30 0.09 0 64 58 0 1 42 6 01 80 42 0 4 P 4 251 120 0 10 0 54 65 0 1 80 7.53 10.5 44 5 -5 3 4 251 30 0 13 0 63 91 8 1 60 7 58 13 7 47.7 16 AW 4 251 30 0 09 0 61 58 0 1.40 6 68 80 42 0 24 AT 4 251 30 0 12 0 63 83 0 1.30 6 20 10 8 44 8 -2 0 4 205 30 0 11 0 66 74 9 0 43 7.54 45 38 5 19 BW R/$

AX 5 251 300 0 14 0.54 97 0 0 90 6.50 99 43 9 28 AZ 5 251 300 0 10 0 48 65 0 1.15 6 98 78 41 8 N/A BW R16 R 6 218 3 0 029 06 20 84 5 67 7.7 41.7 17 AE 6 218 177 0 06 06 37 96 6 85 15.1 49 1 4 AF 6 218 3 0 09 0 58 58 11.0 6 99 25 3 $9 3 14 Table 3-t BWR Surveillance Program Results for Base Metal 12

. 4 GE Nucle:rEnemy GE-NE-B1301807-01R1 alai I.99,RE% 2 RE42 48 5 T-LB RPY Capsele FIEENCE EEFPY DELTA DELTA + TEST FL?NT BWR ID 8.D. Ce N CF (nie *I7) RTNDT M ARGIN SHIFT (int ideg) (%) (%) Im/ cam *2 n SW R/2 AC 2 213 30 0 17 0 07 81 4 78 5 to 23 I 79 I N/A 300 86 36 7 98 19 8 75 8 N/A AS 2 'I3 210 0.29 0 05 131.5 75 8 IS 47 6 103 6 BW R/3 H 3 258 215 02 0 45 137 0.52 6.23 9.5 65.5 0 AR 3 251 215 02 0 32 119 0.28 5 98 51 61.1 5 AL 3 224 210 02 1 05 228 5 39 9 00 58 3 114 3 22 300 66 14 80 77 4 133 4 76 A 3 205 30 0 05 0 92 68 29 7 08 14 6 70 6 A3 3 188 to 03 0 09 138 57 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 23 4 87 33.1 89 1 55 WV 3 251 215 0 17 03 105 5 0 55 6 64 e6 63 6 0 AB 3 254 215 0 16 0 29 1001 0 66 5 63 82 64 2 43 BWRf4 Y 4 251 30 02 0 33 128 1.52 8.20 18 5 74.5 I Q 4 218 30 0 23 0 75 194 5 24 6 80 37 4 93 4 61 300 28 1120 41 0 97 0 62 N 4 143 288 0 02 0 95 27 49 5 90 78 63 8 0 C 4 218 30 0.31 0 72 216 26 5 98 43 6 99 6 K 4 218 30 0 25 0 76 212 24 5.75 40 8 96 8 F 4 218 30 0 13 0 12 68 8 2.3 6 58 12 9 68 9 0 AY 4 251 30 0 08 0 59 105 1 42 6 01 14 5 70 5 68 P 4 251 120 01 0 32 84 2 1.8 7 53 13 6 69 6 17 J 4 251 30 0.11 0 41 102.5 I6 7 58 15 3 78.3 16 AW 4 258 30 0 02 0 95 27 14 6 68 37 59 7 21 AT 4 251 30 0 02 0 95 27 18 6 20 43 60 3 -32 0 4 205 30 0 03 0.93 41 0 43 7 54 25 58 5 5

• BW R/S AX 5 251 300 0.21 0 78 194 09 6 50 19 8 75 8 35 AZ 5 251 300 0 04 0 89 54 1.15 6 98 65 62.5 19 BW Rr6 R 6 218 3 0 072 0.76 97.5 84 5 67 37 3 93.3 28 AE 6 218 177 0 08 0 83 108 96 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 13

GE NucI:arEn:rgy GE-NE-B1301807-01R1 4

>l MeV RPV Capsule FLUENCE @EFPY PLANT BWR ID 1.D . (x 10 ^ 17)

(Is) (deg) (a/c m ^2)

BW R/2 AC 2 213 30 3.60 5.80 300 4.78 7.98 AS 2 213 210 7.46 8.15 BW R/3 11 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

AO 3 224 30 2.30 4.17 W 3 251 215 0.55 6.64 AD 3 251 215 0.66 5.63 BW R/4 Y 4 251 30 f.52 8.20 Q 4 218 10 2.30 6.80

00 2.80 11.20 N 4 lai 2s8 4.90 5.90 C 4 21F 30 2.60 5.98 K 4 2 ! 's 30 2.40 5.75 I

, F 4 216 30 2.30 6.58 i y AY 4 251 30 0.20 1.02 l AY 4 251 30 1.42 6.01 )

• 4 251 120 1.80 7.53 e 4 251 30 1.60 7.58 l AW 4 251 30 1.40 6.6R

. AT 4 251 30 1.30 l O 4 205 30 0.43 7.54 i

BW R/5 AX 5 251 30 0.20 1.38

- AX 5 25] 300 0.90 6 50 I

AZ 5 251 30 0.21 1.36 AZ 5 251 300 N/A N/A AK 5 251 30 0.14 0.90 BW R/6 R 6 218 3 8.4 5.67 i

', AP 6 251 3 0.26 0.93

, A6 6 218 177 96 6.85 AF 6 218 3 11.0 6.99 i X 6 218 3 1.39 1.00 Table 3-3: Flux Wire Results 14

. i GE Nucle:rEnemy GE-NE-B1301807-01R1 Plate Predicted Shift vs. Measured Shift 100 80

,.*w=*

A A

60 **. A A * ~ . * ,- - A

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g j, A A g

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0 5 10 15 20 25 30 35 40 45 50 Predicted shift Figure 3-1: Measured Shift vs. Predicted Shift for Base Metal 15

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, GE Nucl:arEn:rgy GE-NE-B1301807-01R1

4. PRESSURE-TEMPERATURE (P-T) CURVES i The shift in RTwor obtained from surveillance testing is used to evaluate the long term effects of irradiation on the fracture toughness of the vessel. The reference fracture toughness (K a) i is determined using the shift in RTwor ; K ia is part of the cafculations of the P-T curves performed in accordance with ASME Section Ill, I 5

Appendix G. The current GGNS P-T curves were calculated with the 10 EFPY shift in RTuor.

l The K ai correlation was developed from several sets of material data on pressure vessel steel.5 The K ai curve was drawn to bound the available data. Thus, the correlation has inherent conservatism.

I In addition, operation of GGNS follows the steam saturation curve, therefore, the operating temperatures are expected to be well in excess of the minimum I required temperature. During normal and accident conditions, the GGNS )

maintains more than adequate margins. The operationalissues of Pressurized )

i Thermal Shock (PTS) and Low Temperature Over Pressurization (LTOP) are not i applicable to GGNS. The limiting case for GGNS is the pressure test.

l The P-T curve associated with the pressure test is calculated using the crack arrest fracture toughness, Ki n (K.).i The static crack initiation fracture toughness, Ki c, is significantly higher than K i a in the temperature range of interest S.

Therefore use of K ni conservatively bounds the fracture toughness of the vessel.

7 Figure 4-1 is a plot of K,i and K ci as a function of T-RTwo7. The K,i curve is shown to be lower than the K ci curve, conservatively bounding the fracture toughness. For example, at a pressure test temperature of 150 F and a vessel ART of 41 F (corresponding to 20 EFPY for Grand Gulf), the fracture toughness for initiation and arrest are estimated to be:

1 17

~

GE NucbcrEn:rgy GE-NE-B1301807-01R1 Ki e = 216.62 ksiVin i K i a = 87.23 ksiVin  !

1 I

i Thus the K ei value is approximately 2.4 times the Ki , value, clearly showing Ki , to l conservatively bound the calculations.  !

The combination of lower bound fracture toughness, the GGNS operating characteristics and the conservative fracture toughness values indicate that the GGNS vessel fracture toughness is not a significant concern over the life of the i plant.

I l

I l

l 18

GE NuclearEnergy GE-NE-B1301807-01R1 Kia versus KIC 240.00 .

e s

,y . KIC=216.6 su @20 EFPY 200.00 2, e

a e 160.00 - ,' s e a 5 a g '

- e e 120.00 q s a  !

w ,  !

g a E a Kla=87.2 .

80.00

-s 4'~ @20 EFPY

,# s e

1. 8 Kla i

- a - - - KlC  !

40.00 .- **

' i

..........,.- s ,

150-41 F

,4 =109 8

@20 EFPY 0.00 ,

i

-250 -200 -150 -100 -50 0 50 100 150 200 Temperature Relative to RTndt (T-RTndt), 'F Figure 4-1: Comparison of K,i and K i c 19

\

GE NuclearEn:rgy GE-NE-B1301807-01R1 i

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 BWRs than PWRs, there is not much BWR surveillance data at higher J fluences available to analyze, nor would there be for many years.

1

. Rev 2 imposed some hardships on pressure testing for BWRs, some of which might be relieved if a better understanding of the BWR 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 I capsules were designed to systematically evaluate embrittlement trends in BWRs. For example:

20

f GE Nuctsar En:'rgy GE-NE-B1301807-01R1 i

l*

• 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 i matching flux but with different fluence, while some have matching fluence and at a differing flux level.

l . The materials used were selected to bound the range of chemistries in i

BWR beltline materials, and in most cases am BWR beltline materials.

)

. Irradiations are being done in BWRs to correctly simulate conditions like  !

I i

temperature, neutron spectrum and transient operation. '

Relationship to Grand Gulf l

, 1 The SSP has the GGNS surveillance plate material, the GGNS surveillance i

weld material among the materials in the capsules. At least one of these

)

materials is in each of the 7 capsules in the SSP holders. Thus, the SSP results will be applicable to Grand Gulf for two reasons:  !

l l

. Generically, the SSP results will be from representative environmental conditions on materials representative of all BWRs, including GGNS; 1

. Specifically, results will be developed which will provide information on all the GGNS plate and weld surveillance materials, and will be directly applicable to the Grand Gulf surveillance program.

The SSP capsules, when tested, will have collected between 5x10'7 n/cm 2 and 2x10'8 n/cm2fluence, which bounds the end-of-life fluence (EOL) for the l GGNS vessel. Thus, the results of the SSP are complementary to the GGNS

surveillance program such that postponement of the capsule withdrawals will have minimal impact on the understanding of irradiation effects on the GGNS I

vessel.

i 21 4

x

GE Nuclear Energy GE-NE-B1301807-01R1

6. REVISED SURVEILLANCE SCHEDULE The surveillance program is intended to characterize the vessel properties as a function of irradiation over the life of GGNS. The Charpy impact energy obtained from the prescribed testing is used to evaluate the reference fracture toughness of the GGNS vessel (Ki a)in accordance with ASME Section Ill, Appendix G. The 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 GGNS 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 Ki a is for a 1025 psig pressure test. The pressure test temperature was modified on eight year intervals for illustration purposes; the six year intental noted between 18 and 24 years, together with the final interval of eight years, was selected to reach 32 EFPY at EOL. This figure shows that significant margin remains between the limiting K i and the K ai used to i calculate the P-T curves. Thus the K a i is expected to conservatively bound the required vessel fracture toughness. l l

Since the K i a is considered a conservative prediction, and the expected shift in RTuo7i s low, the first surveillance program testing should be at the time at which a majority of the shift in the vessel RTuo7 has been achieved. Early testing of the surveillance specimens may result in the measured shift being less than the data scatter (sometimes resulting in negative shifts in RTuo7). Correct selection of the removal time will ensure credible data. 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. Also,if an anomalous shift were to occur, the SSP willidentify a greater than expected shift.

22

l GE NucI:arEnstyy GE-NE-B1301807-01R1 l The surveillance program schedule should be developed to measure a significant portion of the fracture toughness shift. For GGNS, the limiting weld i material was used to determine this fracture toughness change. To illustrate i

. this, Figure 6-2 is a plot of the fracture toughness as a function of the predicted i

shift in RTsor. As is clearly shown, the fracture toughness decreases as a 4

i function of the shift. The fracture toughness at the beginning of plant life is

. 200 ksiVin and at 32 EFPY is 80.1 ksiVin. Therefore, the change in fracture toughness over the design life of the plant is 119.9 ksiVin.

i .

To determine the schedule for first capsule withdrawal, a value of 75% of the 1

l predicted fracture toughness change ((0.75)(119.1 ksiVin)= 89.9 ksiVin) over the 4

design life of GGNS 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% criterion has been met. This criterion is met at 200 ksiVin - 89.9 ksiVin, or 110.1 ksiVin.- This change in fracture toughness is expected to be achieved when the shift, reading from Figure 6-2, has reached a value of 28.5'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 RTuor as a function of the capsule EFPY. Using the shift value of 28.5 F, the capsule will experience a similar shift at approximately 24 EFPY.

Using a criterion of 75% of the expected change in fracture toughness as the appropriate measurement of vessel embrittlement for GGNS, the first surveillance capsule should be removed at 24 EFPY. The combination of the low expected shift and the inherent margin in the Km calculations will result in a credible set of surveillance data, while maintaining safety.

23

GE NuclearEnergy GE-NE-B1301807-01R1 Grand Gulf il4T RPV Weld K1 vs EFPY for Pressure of 1025 psig 200.0 180.0 -

KIR 160.0

- - - Required KI 140.0 120.0 100 0 -

3  %

I 80.0

~

10 EFPY P-T Curves 18 EFPY P-T Curves 24 EFPY P-T Curves 32 EFPY P-T Curves _

60'0 -

Pressure Test 134'F Pressure Test 140*F Pressure Test 150*F Pressure Test 160*F 40.0 20.0 0.0 0 4 8 12 16 20 24 28 32 EFPY Figure 6-1: Km vs. EFPY for Grand Gulf Weld Material 24

GE NuclearEnergy GE-NE-B1301807-01R1 Grand Gulf 1/4T RPV Weld KI vs EFPY 200.0 '

- 1 I

\ '

32 EFPY P-T Curves 180.0 '

Pressure Test 160'F

=, at 1025 psig 160.0 , ,

l- - KIR R 32 EFPY l N

=

140.0

'N '

=

_ I20.0 -

.E **=,,

g - - - . - - - - - - - - . - - - -- . - - -g _

m 100 0 '

'N

. 75% of shift =% '

I.

110.1 ksi sqrt lin) s 80.0 - ,

a 60_0 ,

a 40.0 ,

Corresponding shift ,28.5 F 20.0  ;

ad 0.0 0.0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 ARTndt Figure 6-2: K i a vs. Predicted Shift 25

= _ . . .

GE NuclearEnergy GE-NE-B1301807-01R1 GRAND GULF Shift vs. EFPY 28.5 F shift from Figure 6-2 N ,, **

30 0 x ,a+

4 . - .. .-

~

25.0 -

, . - - gpp

• withdrawal

, schedule g 20 0 ,-#

2 '

r 15 0 -

9

1. ~

10 0 e 9

9 50 #

/

00 0 4 8 12 16 20 24 28 32 EFPY l Base = = = Weld l Figure 6-3: Predicted shift vs. EFPY, Grand Gulf Surveillance Capsule 26

. GE NucI:arEnstyy GE-NE-B1301807-01R1

. I

7.0 CONCLUSION

S i-l The purpose of the vessel surveillance program is to characterize the vessel

properties as a function of irradiation. The original schedule for Grand Gulf was i

determined according to 10CFR50, Appendix H, resulting in a withdrawal j

schedule of 8 EFPY for the first surveillance capsule.

i Schedules developed according to 10CFR50, Appendix H, however, are intended to apply.to all nuclear power plants. The schedules do not take into account some specific characteristics of Grand Gulf, a low fluence and excellent chemistry (0.02-0.06% copper); the combination of these factors results in a low shift in RTNor. If the first capsule is removed and tested according to the current schedule (8 EFPY), the data obtained is likely not to be useful.

Since the data is unlikely to be useful, the surveillance schedule should be extended. The schedule can be extended for the following reasons:

1. Evaluation of similar data obtained from actual surveillance programs has shown the predictions of fluence, shift and chemistry are bound by expected values. In particular, the BWR/6 data has shown small RTwors hifts for capsules removed at EFPY similar to the current GGNS withdrawal schedule.

Therefore, the surveillance capsule withdrawal schedule should be extended based on the conservatism in calculated shift in RTuo7

2. In addition, the P-T curves contain inherent conservatism, as noted in Section 4. The fracture toughness values used for these calculations are considered to be lower bound values and are significantly less than the crack initiation fracture toughness in the temperature range of interest. At '

operating temperatures, GGNS maintains more than adequate margins; the -

27 i

l r

GENucl:arEn rgy GE-NE-B1301807-01R1 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 specimens and also identify any anomalous information in the predicted values. This characterization will enhance the understanding of vessel embrittlement issues and provide specific data for GGNS. Hence the change in schedule for the GGNS surveillance specimens will not have a significant effec *. on the understanding of vessel irradiation issues.

These reasons justify extending the withdrawal schedule while maintaining reactor safety margins, and provide for more accurate measured data near EOL.

Therefore, the surveillance schedule should be modified.

The material property of most concern is the fracture toughness of the vessel; the surveillance schedule should be based on evaluation of this property. Since the fracture toughness (K i n), is dependent on the shift in RTuor, the optimum EFPY for removal of the capsule ensures credible data (measuring significant shift), while identifying any anomalous conditions. If such an anomalous shift were to occur, the margin between Kim and Ki n, as well as the inherent conservatism of the calculations, can provide a sufficient safety margin for extending the surveillance schedule. In addition the operation of GGNS follows the steam saturation curve; the operating temperatures are expected to be well in excess of the minimum required temperature.

As shown in section 6, the appropriate K ni value selected was 75% of the predicted change in Km. Using this value to determine the appropriate shift in the capsule (hence the appropriate EFPY), the recommended withdrawal 1

i schedule for the first surveillance Grand Gulf capsule is 24 EFPY Removal of 1

i j 28

. GENuclearEn rgy GE-NE-B1301807-01R1 i

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 GGNS limiting weld and plate materials) will soon be available. The i

combination of the data from the first capsule and the SSP would be used to develop the appropriate schedule for the second capsule.

l i

i I

l i

l l 4 I

l 1

29 l .

1 1

GE Nucbar Energy GE-NE-B1301807-01R1

8.0 REFERENCES

l

1. " Reactor Vessel Material Surveillance Program Requirements," Appendix H I 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. " Flux Wire Dosimeter Evaluation for Grand Gulf Nuclear Power Station," GE Report EAS-35-0387, April 1987.

4. " Bounding Assessment of BWR/2-6 Reactor Pressure Vessel Integrity 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 Ill, Appendix A,1992 Edition through Summer 1993 Addenda.

30

i

. GE NuclearEnergy GE-NE-B1301807-01R1 APPENDIX A l l

ADJUSTED REFERENCE TEMPERATURE (ART) CALCULATION  :

1 I

l i

l l

31

l .

. GE NuclearEnergy GE-NE-B1301807-01R1

l. ,

l The ART is, according to Rev 2, a function of the initial RTuor, the shift, sid a margin term. The shift in RTuoris dependent on the chemistry (specifically i copper and nickel) and fluence. The methods of Rev 2 are used to determine l the ART; the method used depends on whether or not surveillance specimen l data is available. l i l In order to re-evaluate the surveillance specimen program schedule, the ART for l

both the vesselitself and the specimens must be calculated. For Grand Gulf, 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 = Initial RTuor + ARTuor + Margin (1)

Initial RTuor is the reference temperature determined according to ASME i Section Ill, 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)

ARTyy = CF *fe2s-omosf) (2) l 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),

a f = f.,f * (e42" ) (3)

'i 32 s

i

. GE NucharEnergy GE-NE-B1301807-01R1 where f,og is the calculated neutron fluence at the vessel ID and x is the depth into the vessel measured from the inner (wetted) surface. - For theca calculations, 18 2 the value of f,yg used was 2.5 X10 n/cm , cbtained from the flux wire analysis .

l 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 l

CF and fluence), it is unnecessary to add extra conservatism by using the upper bound fluence. Any uncertainty in the t!uence is captured by the Margin term.

The Margin term is given by equation (4):

Margin = 2do2 #y 2 (4) where ci = standard deviation of the initial RTuor o3 = standard deviation for ARTwor The standard deviation for ARTwor, c3, is assumed to be 28 F for welds and 17*F for base metal, except that o3 need not exceed 0.50 times the mean 2

ARTwor . The conservative nature of the RTNor determination results in ci being equal to zero. l

~

l Using equations (1) to (4), the ART can be calculated for plants with no surveillance data, including Grand Gulf.

l EXAMPLE CALCULATION l

To better illustrate the ART methodology, the following calculation was i performed for the Grand Gulf base material (Heat #C2594-2); this material was i

i l

l 33 i

W

< GE Nucl:arEn:rgy GE-NE-B1301807-01R1 used in the surveillance capsule. The data was obtained from the Grand Gulf UFSAR:

Initial RTsor: 0F Nicket 0.63 % ,

Copper: 0.04 %

2 Peak Fluence: 2.5 X10'8 n/cm (32 EFPY at vessel wall)

Wall Thickness: 6.19 inches From Table 2 of Rev 2, the chemistry factor for this heat of material is 26. The fluence at the 1/4T depth, determined from equation (3), is equal to:

18 f = (2.5 X10 /10'8) *e<a.2ci.ss) f = .25*0.690 f = 0.172 The change in reference temperature, ARTuor, is calculated according to equation (2):

l ARTuor = 26 *0.172( 284.10 iog o.172)

ARTuor = 26*0.534 = 13.9 For the margin term, the standard deviation of the initial RTuor, oi, is assumed to be zero. The standard deviation for ARTwor, o,, is 13.9 F, as it is base metal and less than the 17 F maximum standard deviation.

Therefore, the ART at 32 EFPY for plate C2594-2 is:

34

s GE NucI:arEnergy GE-NE-B1301807-01R1 ART = 0 + 13.9 + 13.9 = 27.8 F This calculation was repeated for all of the vessel be!tline materials. The results of the calculations for all the beltline materials are shown in Table A-1. Figures 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.

I i

I d

I 35

O T ' t GE Nuclear Energy GE-NE-B1301807-01R1 i

Base Base I Thickness = 6.19 inches 32 EFPY Peak I.D. fluence = 2.50E+18 n/cm' 32 EFPY Peak 1/4 T fluence = 1.72E+18 n/cm* I W eld W eld Thickness = 6.19 inches 32 EFPY Peak 1.D. fluence = 2.50E+18 n/cm' i 32 EFPY Peak I/4 T fluence = 1.72E+18 n/cm*

Initial 32 EFPY 32 EFPY 32 EFPY COMPONENT llEAT OR llEAT/ LOT %Cu %Ni CF RTndt Delta RTndt M argin S hift ART l

'F *F 'F 'F 'F t

BASE:

B E LT LIN E <

C2593-2 0.04 0.59 26 -30.0 13.9 13.9 27.8 -2.2 C2594-1 0.04 0.63 26 -10.0 13.9 13.9 27.8 17.8  ;

C2594-2* 0.04 0.63 26 0.0 13.9 13.9 27.8 27.8 A 122 4-l

• 0.04 0.65 26 0.0 13.9 13.9 27.8 27.8 l i

V ERTIC A L W E LDS:  !

627260/B 322 A 27A E

• 0.06 1.08 82 -30.0 43.8 43.8 87.7 57.7 626677/C301 A27AF' O.03 1.04 41 -20 21.9 21.9 43.8 23.8 i S P6214 B/0331

• 0.02 0.82 27 -50 14.4 14.4 28.9 -21.1 I

• IIEAT FROM WillCll SURVEILLANCE SPECIMENS WERE TAKEN Table A-1: Grand Gulf RPV Material Data 36

. i GE NuclearEnergY GE-NE-B1301807-01R1 GRAND GULF RPV 1/4T ART vs. EFPY 60.0

~~

50.0 40.0 30.0 ,_ '

- 20.0 t

s s V.

E 10.0 -

a' 0.0 f

l

-10.0 ,

9

-20.0 ,'

-30.0 0 4 8 12 16 20 24 28 32 EFPY l BASE - - - WELD l Figure A-1: ART vs. EFPY 37

. _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _