ML022960411

Letter Regards Response to Request for Supplemental Information License Renewal Applications

ML022960411
Person / Time
Site:	Surry, North Anna
Issue date:	10/15/2002
From:	Hartz L Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
02-601, LR/DWL R0
Download: ML022960411 (58)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMONI) VIRGINIA 23261 October 15, 2002 United States Nuclear Regulatory Commission Serial No.:

02-601 Attention: Document Control Desk LR/DWL RO Washington, DC 20555-0001 Docket Nos.:

50-280/281 50-338/339 License Nos.: DPR-32/37 NPF-4/7 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

SURRY AND NORTH ANNA POWER STATIONS UNITS 1 AND 2 RESPONSE TO REQUEST FOR SUPPLEMENTAL INFORMATION LICENSE RENEWAL APPLICATIONS Dominion and the NRC staff have engaged in a series of discussions regarding the Surry and North Anna reactor vessel integrity evaluations applicable to the license renewal period. During these discussions, the NRC staff requested various details of the Dominion evaluations for the Surry and North Anna reactor vessels to facilitate staff confirmatory evaluations.

Preliminary responses were provided via e-mail and subsequent phone conversations. The formal response to the staff's request for North Anna Units 1 and 2 is presented in Attachment 1. The formal response to the staff's request for Surry Units 1 and 2 is presented in Attachment 2. The attached information provides reasonable assurance that the Surry and North Anna reactor vessels can comply with 1 0CFR50.61, 1 OCFR50 Appendix G, 1 OCFR50 Appendix H, and can safely operate during the period of the renewed license.

Should you have any questions regarding this submittal, please contact Mr. J. E.

Wroniewicz at (804) 273-2186.

Very truly yours, Leslie N. Hartz Vice President - Nuclear Engineering Attachment Commitments made in this letter: None

Serial No.: 02-601 SPS/NAPS LR - Supplemental Information cc page 1 of 4 cc:

U. S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, GA 30303-8931 Mr. M. J. Morgan NRC Senior Resident Inspector North Anna Power Station Mr. R. A. Musser NRC Senior Resident Inspector Surry Power Station Mr. J. E. Reasor, Jr.

Old Dominion Electric Cooperative Innsbrook Corporate Center 4201 Dominion Blvd.

Suite 300 Glen Allen, VA 23060 Ms. Ellie Irons, EIR Program Manager Virginia Dept. of Environmental Quality 629 East Main St., 6 th Fl Richmond, VA 23219 Mr. David Paylor, Program Coordinator Virginia Dept. of Environmental Quality P.O. Box 10009 Richmond, VA 23240-0009 Mr. Joe Hassell, Environmental Manager Virginia Dept. of Environmental Quality Water Division P.O. Box 10009 Richmond, VA 23240-0009 Mr. Frank Daniel, Regional Director Virginia Dept. of Environmental Quality Tidewater Regional Office 5636 Southern Blvd.

Virginia Beach, VA 23462

Serial No.: 02-601 SPS/NAPS LR - Supplemental Information cc page 2 of 4 Mr. Gregory Clayton, Regional Director Virginia Dept. of Environmental Quality Northern Virginia Regional Office 13901 Crown Ct.

Woodbridge, VA 22193 Mr. Frank Fulgham, Program Manager Virginia Dept. of Agriculture & Consumer Services Office of Plant & Pest Services 1100 Bank St.

Richmond, VA 23219 Mr. David Brickley, Agency Director Virginia Dept. of Conservation & Recreation 203 Governor St.

Richmond, VA 23219 Mr. William Woodfin, Director Virginia Dept. of Game & Inland Fisheries 4010 West Broad St.

Richmond, VA 23230 Mr. Robert Hicks, Director Virginia Dept. of Health Office of Environmental Health Services 1500 East Main St., Room 115 Richmond, VA 23219 Ms. Kathleen S. Kilpatrick, Director Virginia Dept. of Historic Resources State Historic Preservation Office 2801 Kensington Ave.

Richmond, VA 23221 Dr. Ethel Eaton, Archeologist Senior Virginia Dept. of Historic Resources State Historic Preservation Office 2801 Kensington Ave.

Richmond, VA 23221

Serial No.: 02-601 SPS/NAPS LR - Supplemental Information cc page 3 of 4 Mr. Robert W. Grabb, Assistant Commissioner Virginia Marine Resources Commission 2600 Washington Ave.

Newport News, VA 23607 Dr. John Olney, Associate Professor Virginia Institute of Marine Science School of Marine Science Gloucester Point, VA 23062 Mr. John Simkins Virginia Dept. of Transportation Environmental Division 1401 East Broad St.

Richmond, VA 23219 Mr. Robert Burnley Virginia Economic Development Partnership 901 East Byrd St.

Richmond, VA 23219 Mr. William F. Stephens, Director Virginia State Corporation Commission Division of Energy Regulation 1300 East Main St., 4t Fl., Tyler Bldg.

Richmond, VA 23219 Mr. Michael Cline, State Coordinator Commonwealth of Virginia Department of Emergency Management 10501 Trade Rd.

Richmond, VA 23236-3713 Mr. Terry Lewis, County Administrator P.O. Box 65 Surry, VA 23883 Mr. Lee Lintecum Louisa County Administrator P.O. Box 160 Louisa, VA 23093

Serial No.: 02-601 SPS/NAPS LR - Supplemental Information cc page 4 of 4 Mr. Douglas C. Walker Acting Spotsylvania County Administrator P.O. Box 99 Spotsylvania, VA 22553 Ms. Brenda G. Bailey, County Administrator P.O. Box 11 Orange, VA 22960 Chairman Reeva Tilley Virginia Council on Indians P.O. Box 1475 Richmond, VA 23218 Mr. Don Lillywhite, Director Economics Information Services Virginia Employment Commission State Data Center 703 East Main St., Room 213 Richmond, VA 23219 Mr. Alan Zoellner Government Information Department Swem Library College of William and Mary Landrum Dr.

P.O. Box 8794 Williamsburg, VA 23187-8794 Mr. Walter Newsome Government Information Resources Alderman Library University of Virginia 160 McCormick Rd.

P.O. Box 400154 Charlottesville, VA 22904-4154

SN: 02-601 Docket Nos.: 50-280/281 50-338/339

Subject:

License Renewal Supplemental Information COMMONWEALTH OF VIRGINIA

) )

COUNTY OF HENRICO

)

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Leslie N. Hartz, who is Vice President - Nuclear Engineering, of Virginia Electric and Power Company. She has affirmed before me that she is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of her knowledge and belief.

Acknowledged before me this 15th day of October, 2002.

My Commission Expires: March 31, 2004.

Notary Pub Ic (SEAL)

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 1 of 13 License Renewal - Supplemental Information Serial No.02-601 North Anna Power Station, Units 1 and 2 License Renewal Applications Virginia Electric and Power Company (Dominion)

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 2 of 13 REACTOR VESSEL NEUTRON EMBRITTLEMENT - NORTH ANNA The following information concerning Reactor Vessel Beltline Neutron Fluence, Pressurized Thermal Shock, Charpy Upper Shelf Energy, and Limits for Heatup and Cooldown was prepared in support of North Anna license renewal application. This information demonstrates an ability to comply with applicable regulations governing reactor vessel integrity including 10 CFR 50 Appendix G, 10 CFR 50 Appendix H, and 10 CFR 50.61 during a postulated 20-year license renewal period.

1.

Calculated Beltline Fluence The reactor vessel beltline neutron fluence values applicable to a postulated 20 year license renewal period were calculated using the Virginia Power Reactor Vessel Fluence Methodology Topical Report [Ref. 1].

The methodology described in that report was developed in accordance with Draft Regulatory Guide DG 1053 [Ref. 2]. The reactor vessel fluence calculational methodology was benchmarked using a

combination of Virginia Power surveillance capsules, pressure vessel simulator measurements, and Surry Unit 1 ex-vessel cavity dosimetry measurements.

The underlying requirement of DG-1053 is that the fluence determination should be made on a plant-specific, best-estimate basis rather than on a generic conservative basis.

The methodology used to determine the best-estimate fluence must be demonstrated to have an associated uncertainty of +/-20 percent at the 1-sigma level.

This level of uncertainty is consistent with the assumptions made in the development of the Pressurized Thermal Shock (PTS) screening criteria for vessel welds and plates.

The fluence analyses performed in accordance with the approved Topical Report VEP NAF-3-A used ENDF/B-VI cross sections.

Specifically, the discrete-ordinates calculations used the BUGLE-93 cross section library and the Monte Carlo calculations used the MCNPDAT6 cross section library. These cross sections do not introduce the biases in calculated fluences that have been seen when using earlier cross section libraries (libraries based on END/B-IV and early ENDF/B-V cross sections).

More generally, Dominion's methodology as documented in VEP-NAF-3-A was benchmarked against the PCA experimental results, measured in-vessel dosimetry results, and measured ex-vessel dosimetry results. These benchmark results, along with analytical uncertainty estimates, demonstrate that the methodology has an uncertainty of less than 20% (1 sigma).

Table 1-A presents calculated peak neutron fluence values for the North Anna Unit 1 and Unit 2 reactor pressure vessels at the beginning and end of the license renewal period (BOLRP and EOLRP).

Tables 1-B and 1-C present calculated peak neutron

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 3 of 13 fluence values for individual reactor vessel beltline materials at EOLRP for Units 1 and 2, respectively.

Table 1 -A Calculated Peak Fluence Values North Anna Unit Number EFPY at BOLRP EFPY at EOLRP Fluence* at Clad/Base Metal Interface BOLRP EOLRP Fluence* at 1/4of wall thickness BOLRP EOLRP Fluence* at 3/4 of wall thickness BOLRP EOLRP 1

32.3 50.3 3.920 5.900 2.446 3.681 0.952 1.433 2

34.3 52.3 3.960 5.910 2.471 3.687 0.962 1.435

• Note: All fluence values are in units of 10'9n/cm 2 (E > 1.0 Mev)

Table 1-B North Anna Unit 1 EOLRP Fluence Values RPV Weld Wire Heat Location Inner Surface 1/,T Fluence*

%-T Fluence*

or Material ID Fluence*

990286/295213 Nozzle Shell Forging 0.211 0.132 0.051 990311/298244 Intermediate Shell Forging 5.900 3.681 1.433 990400/292332 Lower Shell Forging 5.900 3.681 1.433 25295 Nozzle to Int. Shell Circ Weld (0D 94%)

0.211 0.132 0.051 4278 Nozzle to Int. Shell Circ Weld (ID 6%)

0.211 0.132 0.051 25531 Int. to Lower Shell Circ Weld 5.900 3.681 1.433

• Note: All fluence values are in units of 10"9n/cm 2 (E > 1.0 Mev)

Table 1-C North Anna Unit 2 EOLRP Fluence Values RPV Weld Wire Heat Location Inner Surface 1/,T Fluence*

%-T Fluence*

or Material ID Fluence*

990598/291396 Nozzle Shell Forging 0.225 0.140 0.055 990496/292424 Intermediate Shell Forging 5.910 3.687 1.435 990533/297355 Lower Shell Forging 5.910 3.687 1.435 4278 Nozzle to Int. Shell Circ Weld (0D 94%)

0.225 0.140 0.055 801 Nozzle to Int. Shell Circ Weld (ID 6%)

0.225 0.140 0.055 716126 Int. to Lower Shell Circ Weld 5,910 3.687 1.435

• Note: All fluence values are in units of 1019n/cm2 (E > 1.0 Mev)

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 4 of 13

2.

Pressurized Thermal Shock The following values were calculated in accordance with 10 CFR 50.61.

Table 2 North Anna Unit 1 Values of RTPTS at 50.3 EFPY Inter. Shell Lower Shell Circ.

Limiting Materials 990311/

990400/

Weld 298244 292332 25531 Initial Ref. NDT Temp. (IF) 17 38 19 Copper Content(%)

0.12 0.16 0.11 Nickel Content (%)

0.82 0.83 0.13 Table* Chemistry Factor (IF) 86.0 Table* Margin (IF) 34 Table* Ref. PTS Temp. (OF) 174.3 S/C** Chemistry Factor (IF) 88.9 93.1 S/C** Margin (IF) 17 28 S/C** Ref. PTS Temp. (OF) 182.5 180.4 Table 3 North Anna Unit 2 Values of RTPTS at 52.3 EFPY Inter. Shell Lower Shell Circ.

Limiting Materials 990496/

990533/

Weld 292424 297355 716126 Initial Ref. NDT Temp. (IF) 75 56

-48 Copper Content (%)

0.10 0.13 0.07 Nickel Content (%)

0.85 0.83 0.05 Table* Chemistry Factor (IF) 67.0 96.0 Table* Margin (IF) 34 34 Table* Ref. PTS Temp. (°F) 205.1 227.7 S/C** Chemistry Factor (IF) 10.4 S/C** Margin (OF) 14.9 S/C** Ref. PTS Temp. (OF)

-18.2

• Table refers to Tables 1 and 2 in Reg. Guide 1.99 Rev. 2 [Ref. 3]

"**Note: Chemistry factor determined using credible surveillance capsule (S/C) data.

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 5 of 13

3. Upper Shelf Energy The requirements on upper shelf energy are included in 10 CFR 50, Appendix G.

10 CFR 50, Appendix G requires utilities to submit an analysis at least 3 years prior to the time that the upper shelf energy of any of the RPV material is predicted to drop below 50 ft-lb, as measured by Charpy V-notch specimen testing.

There are two methods that can be used to estimate the change in upper shelf energy (USE) with irradiation, depending on the availability of credible surveillance capsule data as defined in Revision 2 of Regulatory Guide 1.99. For vessel beltline materials that are not in the surveillance program or not credible, the Charpy upper shelf energy is assumed to decrease as a function of fluence and copper content, as indicated in Regulatory Guide 1.99, Revision 2 [Ref. 3].

When two or more credible surveillance data sets become available from the reactor, they may be used to determine the Charpy USE of the surveillance material.

The surveillance data are then used in conjunction with the Regulatory Guide data to predict the change in USE of the RPV due to irradiation.

Using the 1/ thickness fluence of Section 1, the values of upper shelf energy (USE) in Tables 4 and 5 were calculated for the North Anna Unit 1 and Unit 2 reactor pressure vessels at the end of the license renewal period being evaluated.

Table 4 North Anna Unit 1 USE Values at 50.3 EFPY Inter. Shell Lower Shell Circ.

Limiting Materials 990311/

990400/

Weld 298244 292332 25531 Initial USE Value (ft-lbs)*

92 85 102 Decrease (%)

28 34 34 USE Value (ft-lbs) 65.9 56.5 67.8

• Note: Initial values are measured.

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 6 of 13 Table 5 North Anna Unit 2 USE Values at 52.3 EFPY Inter. Shell Lower Shell Circ.

Limiting Materials 990496/

990533/

Weld 292424 297355 716126 Initial USE Value (ft-lbs)*

74 80 107 Decrease (%)

26 30 28 USE Value (ft-lbs) 54.8 56.3 76.6

• Note: Initial values are measured.

As shown by these results, the upper shelf energy (USE) values at the end of the license renewal period are greater than the NRC (10CFR50) Appendix G requirement of 50 foot-pounds for the limiting materials.

4. Limits for Heatup and Cooldown Figure 1 presents the heatup curves, without margin for instrumentation errors, for a maximum rate of 60°F/hour for the limiting material in the North Anna Units 1 and 2 reactor pressure vessel beltline. Note on these curves, that moderator temperature is Reactor Coolant System water temperature. Likewise, Figure 2 presents the cooldown curves, without margin for instrumentation errors, for a maximum rate of 1001F/hour for the limiting material in the North Anna Units 1 and 2 reactor pressure vessel beltline.

The heatup curves of Figure 1 and the cooldown curves of Figure 2 are based upon the limiting adjusted reference temperature (ART) values from Tables 6 and 7; and are valid for up to 50.3 EFPY in Unit 1 and for up to 52.3 EFPY in Unit 2. Since these curves provide sufficient margin on the operating window relative to the pump seal requirements, no additional actions are required for the license renewal periods of North Anna Unit 1 and Unit 2.

Maximum allowable low temperature over-pressure protection system (LTOPS) power operated relief valve (PORV) setpoints have been developed which bound both North Anna Units 1 and 2. They were developed based on end of license renewal heatup and cooldown curves using the current Westinghouse methodology [Ref. 5]. The setpoints conservatively account for instrument uncertainties and the pressure difference between the wide range pressure transmitter and the reactor vessel limiting beltline region.

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 7 of 13 The following PORV setpoints, which depend upon the reactor coolant system (RCS) temperature, will provide adequate margin to the North Anna Units 1 and 2 Appendix G limits throughout a 20 year license renewal period with no restrictions on the number of RCPs running:

RCS Temperature PORV Setpoint TRCS < 130OF 395 psig 130OF < TRCS < 305°F 450 psig Table 6 North Anna Unit 1 ART Values at 50.3 EFPY Beltline Materials ART at 1/4 T ART at / T Intermed. Shell Forging 990311/298244 Table* Chemistry Factor 166.1 OF 145.6 OF Lower Shell Forging 990400/292332 S/C** Chemistry Factor 174.0 OF 152.8 OF Circumferential Weld 25531 S/C** Chemistry Factor 171.5 OF 149.4 OF

• Table refers to Tables 1 and 2 in Reg. Guide 1.99 Rev. 2 [Ref. 3]

• • Note: Chemistry factor determined using credible surveillance capsule (S/C) data.

Table 7 North Anna Unit 2 ART Values at 52.3 EFPY Beltline Materials ART at 1/4 T ART at 3/4 T Intermed. Shell Forging 990496/292424 Table* Chemistry Factor 198.7 OF 182.7 OF Lower Shell Forging 990533/297355 S/C** Chemistry Factor 218.5 OF 195.6 OF Circumferential Weld 716126 S/C** Chemistry Factor

-19.2 OF

-21.7 OF

• Table refers to Tables 1 and 2 in Reg. Guide 1.99 Rev. 2 [Ref. 3]

"**Note: Chemistry factor determined using credible surveillance capsule (S/C) data.

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 8 of 13

5. Reactor Vessel Surveillance Program The revised North Anna Unit 1 and 2 surveillance capsule withdrawal schedules [Ref.

4], which include provisions for license renewal, in the form of footnotes, are provided in Tables 8

and 9,

respectively.

Dominion anticipates implementation of the recommendation of GALL report for the withdrawal of the final plant-specific surveillance capsules.

==

Conclusion:==

The aforementioned information demonstrates an ability to comply with applicable regulations during a postulated 20-year license renewal period. Required analysis will be performed and implemented in accordance with the requirements of the applicable regulations, and in anticipation of the expiration of affected plant Technical Specifications.

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 9 of 13 Table 8 SURVEILLANCE CAPSULE WITHDRAWAL FOR NORTH ANNA UNIT I SCHEDULE' Capsule Capsule Lead Capsule Status 4 Estimated Insert Est. Capsule Ident.

Location 2 Factor3 Withdrawal EFPY/Year Fluence (xl019)5 EFPY/Year V

1650 1.6 Active 1.1/1979 NA 0.30 U

650 1.0 Active 5.90/1987 NA 0.88 W

2450 1.03 Active 14.7/1998 NA 2.04 Z

3050 0.69 Active 6 16.1/2000 NA 1.48 Z

1650 1.6 NA 16.1/2000 1.48 Z

1650 1.6 EOL/2018 NA 4.64 T

550 0.69 Standby7 16.1/2000 NA 1.48 T

2450 1.03 NA 16.1/2000 1.48 T

2450 1.03 NA NA 3.52 (EOL)

Y 295" 1.03 Standby7 NA NA 4.24 (EOL)

S 450 0.55 Standby7 NA NA 2.27 (EOL)

X 2850 1.6 Stand by 7 EOL/2018 NA 6.59 Withdrawal schedule meets requirements of ASTM E-185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels,"

dated July 1, 1982.

2 See North Anna UFSAR Figure 5.4-4 for original capsule installation locations.

3 Lead Factor is defined in ASTM E-1 85-82 as the ratio of the neutron flux density at the location of the specimens in a surveillance capsule to the neutron flux density at the reactor pressure vessel inside surface at the peak fluence location.

4 Capsules required to satisfy the requirements of ASTM E-185-82 during the current license period are designated Active. Capsules not required by ASTM E-185-82, but which are maintained for contingencies, are designated Standby.

Surveillance capsule neutron fluence estimates based on fluence analysis presented in WCAP-1 1777, "Analysis of Capsule U from the Virginia Electric and Power Company North Anna Unit 1 Reactor Vessel Radiation Surveillance Program," dated February 1988.

Capsule X may be withdrawn at EOL in lieu of Capsule Z to satisfy ASTM E-185-82 fourth capsule requirement for the current license period.

Capsules T, Y, S, and X are available to satisfy potential fluence monitoring requirements during a postulated 20 year license renewal period.

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 10 of 13 Table 9 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE' FOR NORTH ANNA UNIT 2 Capsule Capsule Lead Estimated Insert Est. Capsule Cdent.

Location9 Factor Capsule Status WithdEFPY/Year EFPY/Year Fluence (x1019)12 V

1650 1.66 Active 1.0/1982 NA 0.25 U

650 1.19 Active 6.3/1989 NA 1.07 W

2450 1.19 Active 15.3/1999 NA 2.58 Z

3050 0.81 Standby13 15.3/1999 NA 1.76 Z

1650 1.66 NA 15.3/1999 1.76 Z

1650 1.66 NA NA 5.82 (EOL)

T 55c 0.81 Standby12 15.3/1999 NA 1.76 T

65o 1.19 NA 15.3/1999 1.76 T

650 1.19 NA NA 4.67 (EOL)

Y 295c 1.19 Standby13 NA NA 5.50 (EOL)

S 450 0.65 Standby'3 NA NA 3.00 (EOL)

X 2850 1.72 Active14 EOL/2020 NA 7.95 Withdrawal schedule meets requirements of ASTM E-1 85-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels,"

dated July 1, 1982.

9 See North Anna UFSAR Figure 5.4-4 for original capsule installation locations.

'o Lead Factor is defined in ASTM E-1 85-82 as the ratio of the neutron flux density at the location of the specimens in a surveillance capsule to the neutron flux density at the reactor pressure vessel inside surface at the peak fluence location.

Capsules required to satisfy the requirements of ASTM E-1 85-82 during the current license period are designated Active. Capsules not required by ASTM E-185-82, but which are maintained for contingencies, are designated Standby.

12 Surveillance capsule neutron fluence estimates based on fluence analysis presented in WCAP-12497, "Analysis of Capsule U from the Virginia Electric and Power Company North Anna Unit 2 Reactor Vessel Radiation Surveillance Program," dated January 1990.

Capsules Z, T, Y, and S are available to satisfy potential fluence monitoring requirements during a postulated 20 year license renewal period. Capsule Y may be withdrawn in lieu of capsule X to satisfy ASTM E-185-82 fourth capsule requirement for the current license period.

'4 Withdrawal of Capsule X at EOL satisfies ASTM E-185-82 requirement for EOL capsule, and provide material properties data at a fluence which exceeds that expected to be achieved at the end of a postulated 20 year license renewal period.

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 11 of 13 MATERIAL PROPERTY BASIS LIMITING MATERIAL:

Lower Shell Forging LIMITING ART VALUES AT EOLR:

1/4T, 218.50F 3/4T, 195.60F 2 5 0 0 i,

i...

2250 S2000 w 1750-LEAK TZBT LIMIT NACCEPTABLE R0PERAT ION 1500 1250

-W CA 1000-IHATUP RATE UP TO 60 F/Hr.

-iI(

750 500 250 0 0 CHIT. LIMIT F0R 60 F/Hr.

DOLT-UP TEMP.

CNI7ICALjTT LIM;7 AjABD ON IN aI TICc D1SRO T TIC TTBT TEMPiRATURI (176 I4 FoE THI BERMVcI PCR3OR UP

.O ZOL REEWAL i

50 100 150 200 250 300 350 400 450 Moderator Temperature (Deg.F)

Figure - 1 North Anna Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 601F/hour With Margins of 0 OF and 0 psi for Instrumentation Errors) Applicable to the End of License Renewal Period IACCEPTABLE [f+/-I" OPERAT[ I O

] I I

=

t 500 e..--..--

IIIIl*]f t-t-i f

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 12 of 13 MATERIAL PROPERTY BASIS LIMITING MATERIAL:

Lower Shell Forging LIMITING ART VALUES AT EOLR:

1/4T, 218.5°F 3/4T, 195.60F 2500 11 11,

-11a 3 I-S2250 2000 11111 1750 -

UNACCEPTABLEý OPERATION 1500 1250 1000i ACCEPTABLE OPERATION I

IiIi II I

I I

Hli+

I I 50 I

100 I

150 Mode r a to r I I I I I I I i 1

1 WFI 200 250 Temper I

I "T-

-F 300 350 400 450 ature (Deg.F)

North Anna Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 1001F/hour With Margins of 0 OF and 0 psi for Instrumentation Errors) Applicable to the End of License Renewal Period H RATESV I-COO."'K 750 500 250-

--I 200

-H401 so0 0

BOLT-U P TElP.

S I I I I I I I I II 0

Figure - 2 I !

1 500

.l i

rl a

.i.

.I I I I I

I

i.

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 13 of 13

References:

1.

Virginia Power Topical Report VEP-NAF-3A, "Reactor Vessel Fluence Analysis Methodology," dated November, 1997.

2.

Draft Regulatory Guide DG-1053, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," June 1996 previous draft was DG-1025, September 1993.

3.

NRC Reg. Guide 1.99 Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May, 1988.

4.

Letter from J. P. O'Hanlon to USNRC, "Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Reactor Vessel Surveillance Capsule Withdrawal Schedules," Serial No.98-646, dated December 17, 1998.

5.

WCAP-14040-NP-A, Rev. 2, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, January 1996.

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 1 of 39 License Renewal - Supplemental Information Serial No.02-601 Surry Power Station, Units 1 and 2 License Renewal Applications Virginia Electric and Power Company (Dominion)

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 2 of 39 REACTOR VESSEL NEUTRON EMBRITTLEMENT - SURRY The following information concerning Reactor Vessel Beltline Neutron Fluence, Pressurized Thermal Shock, Charpy Upper Shelf Energy, and Limits for Heatup and Cooldown was prepared in support of Surry license renewal application. This information demonstrates an ability to comply with applicable regulations governing reactor vessel integrity including 10 CFR 50 Appendix G, 10 CFR 50 Appendix H, and 10 CFR 50.61 during a postulated 20-year license renewal period.

1. Calculated Beitline Fluence The reactor vessel beltline neutron fluence values applicable to a postulated 20 year license renewal period were calculated using the Virginia Power Reactor Vessel Fluence Methodology Topical Report [Ref. 1].

The methodology described in that report was developed in accordance with Draft Regulatory Guide DG 1053 [Ref. 2]. The reactor vessel fluence calculational methodology was benchmarked using a

combination of Virginia Power surveillance capsules, pressure vessel simulator measurements, and Surry Unit 1 ex-vessel cavity dosimetry measurements.

The underlying requirement of DG-1053 is that the fluence determination should be made on a plant-specific, best-estimate basis rather than on a generic conservative basis.

The methodology used to determine the best-estimate fluence must be demonstrated to have an associated uncertainty of +/-20 percent at the 1-sigma level.

This level of uncertainty is consistent with the assumptions made in the development of the Pressurized Thermal Shock (PTS) screening criteria for vessel welds and plates.

The fluence analyses performed in accordance with the approved Topical Report VEP NAF-3-A used ENDF/B-VI cross sections.

Specifically, the discrete-ordinates calculations used the BUGLE-93 cross section library and the Monte Carlo calculations used the MCNPDAT6 cross section library. These cross sections do not introduce the biases in calculated fluences that have been seen when using earlier cross section libraries (libraries based on END/B-IV and early ENDF/B-V cross sections).

More generally, Dominion's methodology as documented in VEP-NAF-3-A was benchmarked against the PCA experimental results, measured in-vessel dosimetry results, and measured ex-vessel dosimetry results. These benchmark results, along with analytical uncertainty estimates, demonstrate that the methodology has an uncertainty of less than 20% (1 sigma).

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 3 of 39 Table 1-A presents calculated peak neutron fluence values for the Surry Unit 1 and Unit 2 reactor pressure vessels at the beginning and end of the license renewal period (BOLRP and EOLRP). Tables 1-B and 1-C present calculated peak neutron fluence values for individual reactor vessel beltline materials at EOLRP for Units 1 and 2, respectively.

Table 1-A Calculated Peak Fluence Values Surry Unit Number 1

2 EFPY at BOLRP 29.6 30.1 EFPY at EOLRP 47.6 48.1 Fluence* at Clad/Base Metal Interface BOLRP 3.530 3.520 EOLRP 5.400 5.340 Fluence* at 1/4 of wall thickness BOLRP 2.154 2.147 EOLRP 3.294 3.258 Fluence* at 3/4 of wall thickness BOLRP 0.802 0.799 EOLRP 1.226 1.213

• Note: All fluence values are in units of 1019n/cm2 (E > 1.0 MeV)

Table 1-B Surry Unit 1 EOLRP Fluence Values RPV Weld Wire Heat or Inner Surface Location Flece

/-T Fluence*

3/4/-T Fluence*

Material ID Fluence*

122V109VA1 Nozzle Shell Forging 0.496 0.303 0.113 C4326-1 Intermediate Shell 5.400 3.294 1.226 C4326-2 Intermediate Shell 5.400 3.294 1.226 4415-1 Lower Shell 5.400 3.294 1.226 4415-2 Lower Shell 5.400 3.294 1.226 J726/25017 Nozzle to Int Shell Circ Weld 0.496 0.303 0.113 SA-1585/72445 Int. to Low Sh. Circ (ID 40%)

4.700 2.867 1.067 SA-1650/72445 Int. to Low Sh. Circ (OD 60%)

4.700 2.867 1.067 SA-1494/8T1554 Int Shell Long. Welds L3 & L4 0.914 0.558 0.208 SA-1494/8T1554 Lower Shell Long. Weld Li 0.790 0.482 0.179 SA-1526/299L44 Lower Shell Long. Weld L2 0.790 0.482 0.179

• Note: All fluence values are in units of 1019n/cm 2 (E > 1.0 Mev)

Note: Results reflect presence of part-length hafnium flux suppression inserts installed in Surry Unit 1 reload cores.

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 4 of 39 Table 1-C Surry Unit 2 EOLRP Fluence Values RPV Weld Wire Heat or Location Inner Surface

'-T Fluence*

%-T Fluence*

Material ID Fluence*

123V303VA1 Nozzle Shell Forging 0.471 0.287 0.107 C4331-2 Intermediate Shell 5.340 3.258 1.213 C4339-2 Intermediate Shell 5.340 3.258 1.213 C4208-2 Lower Shell 5.340 3.258 1.213 C4339-1 Lower Shell 5.340 3.258 1.213 L737/4275 Nozzle to Int Shell Circ Weld 0.471 0.287 0.107 R3008/0227 Int. to Lower Shell Circ Weld 5.340 3.258 1.213 WF-4/8T1762 Int. Shell Long. L4 (ID 50%)

1.080 0.659 0.245 SA-1585/72445 Int. Sh. L3 (100%), L4 (OD 1.080 0.659 0.245 50%)

WF-4/8T1762 LS L2 (ID 63%), Li (100%)

1.080 0.659 0.245 WF-8/8T1762 LS Long. Weld L2 (OD 37%)

1.080 0.659 0.245

• Note: All fluence values are in units of 1019n/cm2 (E > 1.0 MeV)

2. Pressurized Thermal Shock The following values were calculated in accordance with 10 CFR 50.61.

Table 2 Surry Unit I Values of RTpTs at 47.6 EFPY Limiting Materials Initial Ref. NDT Temp. (IF)

Copper Content (%)

Nickel Content (%)

Surface Fluence (1019n/cm 2)

Table* Chemistry Factor (OF)

Table* Margin (°F)

Table* Ref. PTS Temp. (OF)

S/C** Chemistry Factor (IF)

S/C** Margin (OF)

S/C** Ref. PTS Temp. (OF)

Lower Shell 4415-1 20 0.11 0.50 5.40 85.0 17.0 157.4 Circ.

Weld 72445

-5 0.22 0.54 4.70 Long.

Weld 299L44

-7 0.34 0.68 0.79 220.6 69.5 268.5 138.0 48.3 235.2

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 5 of 39 Table 3 Surry Unit 2 Values of RTPTs at 48.1 EFPY Lower Circ.

Long.

Limiting Materials Shell Weld Weld C4208-2 0227 8T1 762 Initial Ref. NDT Temp. (IF)

-30 0

-5 Copper Content (%)

0.15 0.19 0.19 Nickel Content (%)

0.55 0.55 0.57 Surface Fluence (1019n/cm 2) 5.34 5.34 1.08 Table* Chemistry Factor (IF) 107.3 152.4 Table* Margin (IF) 34.0 68.5 Table* Ref. PTS Temp. (OF) 155.8 219.1 S/C** Chemistry Factor (IF) 128.0 S/C** Margin (°F) 48.8 S/C** Ref. PTS Temp. (IF) 230.0

• Table refers to Tables 1 and 2 in Reg. Guide 1.99 Rev. 2 [Ref. 3]

• • Note: Chemistry factor determined using credible surveillance capsule (S/C) data [Ref. 4].

Because the value of the RTPTS for weld material SA-1526 (i.e., 299L44) is only slightly lower than the PTS screening criterion of 2701F, NRC staff requested confirmation of the basis for selecting the RG 1.99 Revision 2 [Ref. 3] Position 1.1 chemistry factor in lieu of the chemistry factor determined in accordance with Position 2.1 based on available surveillance data.

In addition, the NRC requested confirmation that the neutron fluence values for surveillance capsules containing weld material fabricated with the same weld wire heat as SA-1526 were developed with methods that meet the requirements of RG 1.190, "Calculational and Dosimetry Methods For Determining Pressure Vessel Neutron Fluence." An evaluation of the surveillance capsule neutron fluence values, and the basis for the selected Chemistry Factor are provided below.

Dominion contracted Framatome to confirm that the neutron fluence values associated with surveillance capsules containing weld material fabricated with weld wire heat number 299L44 were developed in a manner consistent with their approved topical methodology, BAW-2241P-A [Ref. 7], which complies with the requirements of RG 1.190.

If such confirmation could not be provided, Framatome was requested to provide RG 1.190-compliant fluence values using the methodology of BAW-2241P-A

[Ref. 7]. [Ref. 22] presents the results of the Framatome surveillance capsule fluence analyses and evaluations. The surveillance capsule neutron fluence values presented

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 6 of 39 in [Ref. 22] were verified as compliant with the requirements of BAW-2241 P-A and RG 1.190, and are used in the RG 1.99 Revision 2 Position 2.1 calculations presented in Appendices A through E.

Table 10 presents a comparison between the previously submitted fluence values and the revised RG 1.190-compliant fluences.

In addition to the fluence calculations, Framatome performed a fabrication records search to confirm the nature and origin of both the Surry 1 longitudinal weld L2 and the Surry 1 surveillance weld material. The results of their records search are presented in

[Ref. 23]. Their record search confirmed that the Surry Unit 1 reactor vessel lower shell longitudinal weld seam (designated "L2") was fabricated with weld wire heat 299L44 and flux lot 8596. Fabrication records for surveillance weld material indicate that the same weld wire heat and flux lot as the second longitudinal weld seam were used in the construction of the surveillance weld. It also appears that the surveillance weld seam was fabricated along with the shell longitudinal seam, as Weld Control Records show that the same welder did the welding process for both parts on the same days and at the same times, using the same machine. The surveillance material received the same heat treatment as the shell, and shipping records indicate that the surveillance material was sent to Rotterdam.

This is evidence suggests that measured values of ARTNDT obtained from Surry Unit 1 surveillance material are unbiased estimators of the irradiated behavior of the Surry Unit 1 reactor vessel beltline material fabricated from the same heat of weld wire, and that no adjustments for irradiation temperature or chemical composition variability are required for determination of surveillance material credibility or for application of the surveillance data to evaluate the beltline material.

When evaluated in this manner, and with the revised RG-1.190-compliant surveillance capsule fluence values presented in [Ref. 22], the Surry Unit 1 surveillance data is determined to be credible (so the ARTNDT margin term may be divided by 2), and a RG 1.99 Revision 2 Position 2.1 Chemistry Factor of 215.80 F is calculated. (See Appendix D.) The resulting RTFIS value would be conservatively bounded by that presented in Table 2 for the SA-1526/299L44 beltline material. Because the surveillance weld is representative of the reactor vessel beltline weld, and the surveillance data is credible, a Chemistry Factor of 215.8 0F is applicable to the Surry Unit 1 weld fabricated with weld wire heat number 299L44.

The NRC requested that Dominion perform several additional RG 1.99 Revision 2 Position 2.1 calculations using all available surveillance data applicable to weld wire heat number 299L44 in order to verify the conservatism of the Chemistry Factor used to evaluate the Surry Unit 1 beltline material fabricated with weld wire heat number 299L44. The results of the requested calculations are presented in Appendices A, B, and C. Calculations were performed under the following conditions:

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 7 of 39 A. The data credibility evaluation was performed with corrections for differences in irradiation temperature and chemical composition.

The evaluation of the conservatism of using the Position 1.1 Chemistry Factor to evaluate the beltline weld material (i.e., verification that the difference between measured and predicted RTNOT shift values is less than 2oY, or 56°F) was performed using a trend curve based on the RG 1.99 Revision 2 Position 1.1 Chemistry Factor for the mean surveillance material chemical composition.

B. Calculations were performed without corrections for differences in irradiation temperature and chemical composition. The evaluation of the conservatism of using the Position 1.1 Chemistry Factor to evaluate the beltline weld material was, again, performed using a trend curve based on the RG 1.99 Revision 2 Position 1.1 Chemistry Factor for the mean surveillance material chemical composition.

C. Calculations were performed without corrections for differences in irradiation temperature and chemical composition. The evaluation of the conservatism of using the Position 1.1 Chemistry Factor to evaluate the beltline weld material was performed using trend curves based on RG 1.99 Revision 2 Position 1.1 Chemistry Factors for individual surveillance capsule chemical compositions.

For each of these conditions, the surveillance data were determined to be non-credible (i.e., one or more measured values of RTNT shift was determined to differ from the best estimate ARTNDT trend curve for the surveillance data by more than lGA, or 28°F).

In Case A, use of the RG 1.99 Rev. 2 Position 1.1 chemistry factor was determined to be conservative for all surveillance data points.

In Case B, use of the RG 1.99 Rev. 2 Position 1.1 chemistry factor was determined to be non-conservative for only one surveillance data point.

Since there are nine surveillance data points in the population and only one falls outside of 2 TA, it is concluded that use of the RG 1.99 Revision 2 Position 1.1 Chemistry Factor to assess the beltline material is conservative.

In Case C, use of the RG 1.99 Rev. 2 Position 1.1 chemistry factor was determined to be non-conservative for only one surveillance data point. Again, because there are nine surveillance data points in the population and only one falls outside of 2GA, it is concluded that use of the RG 1.99 Revision 2 Position 1.1 Chemistry Factor to assess the beltline material is conservative.

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 8 of 39 A conservative means of assessing the possible effects of chemical variability is to assume that all surveillance specimens fabricated from weld wire heat 299L44 are unbiased estimators of the irradiated behavior of the Surry Unit 1 reactor vessel beltline material fabricated from the same heat of weld wire.

Therefore, in addition to the evaluations requested by the NRC (i.e., Appendices A, B, and C) and the evaluation of Surry Unit 1 surveillance data only (Appendix D), Dominion performed an additional evaluation (Appendix E) under this assumption.

Consistent with NRC guidance, a correction for differences in irradiation temperature was applied in the RG 1.99 Revision 2 Position 2.1 data credibility determination since all surveillance specimens in the data set were not irradiated in the same reactor vessel. Further, a correction for differences in irradiation temperature was applied for the application of surveillance data to the beltline material since the surveillance specimens were not irradiated in Surry Unit 1.

Standard NRC practice requires the reactor vessel beltline material chemistry to be assumed to be the "mean for the heat" chemical composition. The calculation of the "mean for the heat" chemical composition is based on available material chemistry data for materials fabricated from the same heat of weld wire as the reactor vessel beltline.

It follows that, if Surry Unit 1 SA-1526 beltline material is represented by the "mean for the heat" chemical composition of weld materials fabricated from weld wire heat 299L44, measured values of ARTNDT obtained from surveillance specimens fabricated from weld wire heat 299L44 are unbiased estimators of the mean ARTNDT trend for the Surry Unit 1 SA-1526 beltline material. This relationship remains valid regardless of how the measured chemical composition of surveillance materials compares to the best-estimate chemical composition of the reactor vessel beltline material. NUREG/CR 6551 [Ref. 20] provides further confirmation that individual 299L44 surveillance data points should be considered unbiased estimators of the mean irradiated behavior of the Surry Unit 1 SA-1526 beltline material. The discussion on Page 89 and Figure D.12 of Ref. 20 demonstrates that, for Linde 80 materials, Cu concentrations above 0.26% do not affect ARTNOT' The calculations summarized in Appendix E demonstrate that the surveillance data were determined to be non-credible and, again, only one of nine data points was determined to be non-conservative with respect to RG 1.99 Revision 2 Position 1.1 predictions.

The chemistry factor based on the surveillance data (although non credible) was determined to be 229.3 IF. This Chemistry Factor is not significantly different from either the RG 1.99 Revision 2 Position 2.1 Chemistry Factor based only on Surry Unit 1 surveillance data (i.e., 215.8 0F), or the RG 1.99 Revision 2 Position 1.1 Chemistry Factor based on the "mean for the heat" chemical composition for welds fabricated with weld wire heat 299L44 (i.e., 220.6°F). Thus, the Appendix E evaluation

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 9 of 39 provides further confirmation of the conservatism of RG 1.99 Revision 2 Position 1.1 calculations for this material, and of the assertion that measured values of ARTNDT obtained from weld materials fabricated weld wire heat 299L44 may be considered unbiased estimators of the irradiated behavior of reactor vessel beltline material fabricated from the same heat of weld wire.

The RG 1.99 Revision 2 Position 2.1 ratio procedure requires the Chemistry Factor determined based on surveillance data to be increased by the ratio of the RG 1.99 Revision 2 Position 1.1 Chemistry Factor (based on the "mean for the heat" chemical composition) and the Position 1.1 Chemistry Factor for the surveillance material (based on the surveillance material mean chemical composition).

The foregoing data and evaluations confirm that application of the RG 1.99 Revision 2 Position 2.1 ratio procedure to 299L44 surveillance data for its application to the Surry Unit 1 SA-1526 beltline material would be excessively conservative and unnecessary. Use of multiple (i.e., 9) surveillance data points in the Appendix E evaluation ensures that the possible effects of chemical composition variability have been addressed.

Framatome Topical Report BAW-2308 [Ref. 21] represents an additional source of analytical margin for the Surry 1 weld material fabricated with weld wire heat 299L44.

This report provides a technical basis for reducing the Initial RTNoT value for 299L44 materials from -70F with a standard deviation of 20.6°F to -81.8°F with a standard deviation of 11.6°F.

The RTPTS margin provided by this topical, once approved, will support acceptable RTPTS screening calculations well beyond the end of the proposed license renewal period.

The foregoing evaluations provide reasonable assurance that the Surry 1 reactor vessel can comply with 10CFR50.61 and can safely operate during the period of the renewed license.

3. Upper Shelf Energy The requirements on upper shelf energy are included in 10 CFR 50, Appendix G.

10 CFR 50, Appendix G requires utilities to submit an analysis at least 3 years prior to the time that the upper shelf energy of any of the reactor pressure vessel (RPV) material is predicted to drop below 50 ft-lb, as measured by Charpy V-notch specimen testing.

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 10 of 39 There are two methods that can be used to estimate the change in upper shelf energy (USE) with irradiation, depending on the availability of credible surveillance capsule data as defined in Revision 2 of Regulatory Guide 1.99. For vessel beltline materials that are not in the surveillance program or not credible, the Charpy upper shelf energy is assumed to decrease as a function of fluence and copper content, as indicated in Regulatory Guide 1.99, Revision 2.

When two or more credible surveillance data sets become available from the reactor, they may be used to determine the Charpy USE of the surveillance material.

The surveillance data are then used in conjunction with the Regulatory Guide data to predict the change in USE of the RPV due to irradiation.

Using the 1/4 thickness (1/4 T) fluence values per Section 1, the values of upper shelf energy (USE) in Tables 4 and 5 were calculated for the Surry Unit 1 and Unit 2 reactor pressure vessels at the end of the license renewal period being evaluated.

Table 4 Surry Unit 1 USE Values at 47.6 EFPY Lower Circ.

Long.

Limiting Materials Shell Weld Weld 4415-2 72445 299L44 Initial USE Value (ft-lbs)*

83 77 70 1/4/T Fluence (1019n/cm 2) 3.29 2.87 0.482 Decrease (%)

26 45 38 USE Value (ft-lbs) 61.6 42.1 43.6 Table 5 Surry Unit 2 USE Values at 48.1 EFPY Intermed.

Circ.

Long.

Limiting Materials Shell Weld Weld C4331-2 0227 8T1762 Initial USE Value (ft-lbs)*

84 90 70 1/4 T Fluence (1019n/cm2) 3.26 3.26 0.659 Decrease (%)

27 43 30 USE Value (ft-lbs) 60.9 51.2 49.2

• Note: Initial values are measured.

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 11 of 39 As shown by these results, the upper shelf energy (USE) values at the end of the license renewal period are greater than the NRC (10CFR50) Appendix G requirement of 50 foot-pounds for some of the limiting materials.

For the other limiting materials (welds), an equivalent margins analysis (EMA) was used to justify the acceptability of values below the 50 foot-pound requirement. This EMA [Ref. 17] uses the methodology of the approved USE Reports [Refs. 18 and 19]. Four service levels - A, B, C and D are evaluated for an equivalent margins analysis. There are two conditions for each loading case to be met. They are:

1. The applied J-integral shall be less than J-integral of the material at a ductile flaw extension of 0.10 inches.

2. Flow extensions shall be ductile and stable.

In addition to these two conditions, one more requirement has to be met for service level D, which is as follows:

3. The extent of stable flaw extension shall be less than or equal to 75% of the vessel wall thickness, and the remaining ligament shall not be subject to tensile instability.

In addition, the NRC has requested further information regarding the history of the USE submittals and NRC approvals.

In Generic Letter 92-01, Rev. 1, "Reactor Vessel Structural Integrity, 10CFR50.54(f)," dated March 6, 1992 [Ref. 8], the NRC requested information regarding the USE values for beltline materials used in the Surry reactor pressure vessels.

Dominion transmitted a B&W report (BAW-2166, [Ref. 10] dated June 1992) via June 29, 1992 letter [Ref. 9] entitled, "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Response to Generic Letter 92-01, Reactor Vessel Structural Integrity."

The NRC responded on July 21, 1993 with a letter [Ref. 11], "Surry Power Station, Units 1 and 2 - Request for Additional Information Regarding Generic Letter 92-01, Revision 3," requesting specific information on the confirmation of topicals that would be used for licensing basis for USE values.

Dominion responded to the RAIs via September 23, 1993 letter [Ref. 12], "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Response to Request for Additional Information Regarding Response to Generic Letter 92-01, Rev. 1."

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 12 of 39 The NRC then requested verification of the USE data for the then newly-developed Reactor Vessel Integrity Database (RVID) in a letter [Ref. 13] dated May 24, 1994, "Generic Letter 92-01, Revision 1, Reactor Vessel Structural Integrity, Surry Power Station, Units 1 and 2, (TAC NOS. M83739 and M83740)." The requested verification from Dominion was provided via the transmittal of B&W report BAW-2222 [Ref. 15],

dated June 1994.

Dominion's letter [Ref. 14] transmitting BAW-2222 was "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Generic Letter 92-01, Revision 1, Reactor Vessel Structural Integrity Request for Additional Information,"

dated June 30, 1994.

Following the June 30, 1994 Dominion letter, the NRC issued "GL 92-01, Rev. 1 Reactor Vessel Structural Integrity, Supplement 1," on May 19, 1995 [Ref. 16]. In this letter, the NRC stated that all plants were able to demonstrate compliance with the requirements of Appendix G with regards to the issue of USE.

4. Limits for Heatup and Cooldown Figure 1 presents the heatup curves, without margin for instrumentation errors, for a maximum rate of 60°F/hour for the limiting material in the Surry Units 1 and 2 reactor pressure vessel beltline.

Note on these curves, that moderator temperature is the Reactor Coolant System water temperature. Likewise, Figure 2 presents the cooldown curves, without margin for instrumentation errors, for a maximum rate of 100 0F/hour for the limiting material in the Surry Units 1 and 2 reactor pressure vessel beltline. The heatup curves of Figure 1 and the cooldown curves of Figure - 2 are based upon the limiting adjusted reference temperature (ART) values from Tables 6 and 7, and are valid for up to 47.6 EFPY in Unit 1 and for up to 48.1 EFPY in Unit 2. Since these curves provide sufficient margin on the operating window relative to the pump seal requirements, no additional actions are required for the license renewal periods of Surry Unit 1 and Unit 2.

Maximum allowable low temperature over-pressure protection system (LTOPS) power operated relief valve (PORV) setpoints have been developed which bound both Surry Units 1 and 2.

They were developed based on end of license renewal heatup and cooldown curves using the current Westinghouse methodology (Ref. 5). The setpoints conservatively account for instrument uncertainties and the pressure difference between the wide range pressure transmitter and the reactor vessel limiting beltline region.

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 13 of 39 The following PORV setpoints will provide adequate margin to the Surry Units 1 and 2 Appendix G limits throughout a 20 year license renewal period with no restrictions on the number of RCPs running:

RCS Temperature PORV Setpoint TRCS < 3250F 399 psig Table 6 Surry Unit 1 ART Values at 47.6 EFPY Beltline Materials ART at 1/4 T ART at 3/4 T Lower Shell 4415-1 S/C** Chemistry Factor 148.6 OF 126.8 OF Circumferential Weld 72445 S/C** Chemistry Factor 220.0 OF 183.9 OF Longitudinal Weld 299L44 Table* Chemistry Factor 238.2 OF 182.5 OF

• Table refers to Tables 1 and 2 in Reg. Guide 1.99 Rev. 1 [Ref. 3]

"**Note: Chemistry factor determined using credible surveillance capsule (S/C) data [Ref. 4].

Table 7 Surry Unit 2 ART Values at 48.1 EFPY Beltline Materials ART at 'A T ART at 3/4 T Lower Shell C4208-2 Table* Chemistry Factor 144.5 OF 117.0 OF Circumferential Weld 0227 S/C** Chemistry Factor 216.5 OF 183.7 OF Longitudinal Weld 8T1 762 Table* Chemistry Factor 198.0 OF 157.8 OF

• Table refers to Tables 1 and 2 in Reg. Guide 1.99 Rev. 1 [Ref. 3]

"**Note: Chemistry factor determined using credible surveillance capsule (S/C) data [Ref. 41.

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 14 of 39

5. Reactor Vessel Surveillance Program The Surry Unit 1 and 2 surveillance capsule withdrawal schedules [Ref. 6], which include provisions for license renewal are provided in Tables 8 and 9, respectively.

Specifically, Dominion has already acquired surveillance capsule data for Surry Units 1 and 2 that bounds, in terms of accumulated fluence, the predicted end-of-license renewal inner surface fluence at the limiting beltline weld material (i.e. 0.79x1019 n/cm 2 for the Surry Unit 1 lower shell longitudinal weld, SA-1526); surveillance data from the Surry Unit 1 surveillance program has been collected at fluences as high as 1.992 x11019 n/cm 2 [Ref. 22].

Additional Surry Units 1 and 2 standby surveillance capsules are available to provide additional material properties data and fluence monitoring during the license renewal period.

Dominion anticipates implementation of the recommendation of Generic Aging Lessons Learned Report (GALL) [Ref. 24] report for the withdrawal of the final plant-specific surveillance capsules.

==

Conclusion:==

The aforementioned information demonstrates an ability to comply with applicable regulations during a postulated 20-year license renewal period. Required analysis will be performed and implemented in accordance with the requirements of the applicable regulations, and in anticipation of the expiration of affected plant Technical Specifications.

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 15 of 39 Table 8 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE SURRY UNIT 1 Withdrawal Schedulea Capsule Identification T

W V

X X

X z

z z

Y Y

Y U

U U

Sc Capsule Loc.

(Deg.)

285 55 165 65 165 165 245 285 285 305 165 165 45 65 65 295 Withdraw EFPY (Year) 1.5/1974 3.0/1978 6.0/1986 13.3/1994 NA 15.8/1997 13.3/1994 NA 28.8/EOL 15.8/1997 NA NA 13.3/1994 NA NA 21.0/2002 Insert EFPY (Year)

NA NA NA NA 13.3/1994 NA NA 13.3/1994 NA NA 15.8/1997 NA NA 13.3/1994 NA NA Est. Caps. Fluenceb (X 1019) 0.289 0.431 1.94 1.88 NA 2.42 1.88 NA 5.18 1.48 NA 4.24 0.99 NA 3.04 2.90 (EOL)

(EOL)

NOTES:

(1) The capsule fluence estimates may be compared to the Surry I calculated end-of-license (EOL) 0' vessel inner surface fluence of 3.96 x 1019 n/cm2.

(2) All Surry 1 surveillance capsules need only be evaluated for dosimetry to obtain data for use in fluence analysis. Irradiated materials properties need not be determined for any Surry 1 capsules unless the data is desired for use in the development of revised heatup and cooldown curves.

a. Surry Unit I is a participant in B&WOG Master Integrated Reactor Vessel Materials Surveil lance Program (B&W Report BAW-1543)

b. Fluence estimates based on WCAP-11015, Revision I (Surry Units 1 and 2 Reactor Vessel Fluence and RTPTS Evaluations, dated April, 1987) extrapolations beyond Surry 1 Cycle 7, with 80% capacity factor assumed, and core thermal power uprating to 2,546 MWt assumed at beginning of Cycle 11.

c. Capsule S to be withdrawn and evaluated for dosimetry if no data from reactor cavity dosim etry is available.

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 16 of 39 Table 9 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE SURRY UNIT 2 Withdrawal Schedulea Capsule Capsule Loc.

Withdraw Insert EFPY Est. Caps. Fluence1 )

Identification (Deg.)

EFPY (Year)

(Year)

(x 1019)

X 285 1.2/1975 NA 0.301 W

245 3.8/1979 NA 0.654 V

165 8.5/1986 NA 2.02 Y

295 13.7/1994 NA 1.97 Y

165 NA 13.7/1994 NA Y

165 19.7/2002 NA 3.14 U

65 25.6/2008 NA 3.54 T

55 19.7/2002 NA 1.93 T

165 NA 19.7/2002 NA T

165 NA NA 3.83 (EOL)

Z 305 13.7/1994 NA 1.38 Z

245 NA 13.7/1994 NA Z

245 NA NA 3.45 (EOL)

Sc 45 15.0/1996 NA 2.31 (EOL)

W1 285 NA 10.9/1991 NA M71 d 285 16.4/1997 NA 1.53 NOTES:

(1) The capsule fluence estimates may be compared to the Surry 2 calculated end-of-license (EOL) 00 vessel timer surface fluence of 3,43 x 10i9orcm 2.

(2) With the exception of Capsule Y and W1, all Sunsy 2 surveillance capsules need only be evaluated for dosimetry to obtain data for use in fluence analysis. Irradiated materials properties need not be determnined for any Sun-y 2 capsules (except Y and W 1) umless the data is desired for use in the development of revised heatup and cooldown curves,

a. Surry Unit 2 is a participant in B&WQG Master Integrated Reactor Vessel Materials Surveillance Program (B&W Report BAW-1543)

b. Fluence estimates based on WCAP-I 1015, Revision ] (Surry Units ] and 2 Reactor Vessel Fluence and RTPTS Evaluations, dated April, 1987) extrapolations beyond Surry 2 Cycle 7, with 80% capacity factor assumed, and core thermal power uprating to 2546 MWt assumed at beginning of Cycle 11.

c. Capsule S to be withdrawn and evaluated for dosimetry if no data from reactor cavity dosimetry is available.

d. Master Integrated Reactor Vessel Materials Surveillance Program capsule.

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 17 of 39 Table 10: Fluence Value Comparison Based on [Ref. 22]

Previously RG 1.190-RG 1.190 Surveillance Capsule ID Recorded Fluence Compliant Fluence Compliant Fluence (x 1019)

(x 1019)

Reference Capsule TMI2-LG1 Casl M2LI0.83 0.83

[7]

(BWOG CR-3 Irrad.)

Capsule CR3-LG1 (BWOG CR-3 Irrad.)

Capsule TMI2-LG1 0.968 0.968

[7]

(BWOG CR-3 Irrad.)

Three Mile Island Unit 1 0.866 0.882

[22]

Capsule C Three Mile Island Unit 1 0.107 0.097

[7]

Capsule E Capsule W-1 (CR-3 0.669 0.669

[25]

NBD)

Surry Unit 1 Capsule T 0.281 0.292

[22]

Surry Unit 1 Capsule V 1.940 1.992

[22]

Surry Unit 1 Capsule X 1.599 1.599

[26]

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 18 of 39 MATERIAL PROPERTY BASIS LIMITING MATERIAL:

LOWER SHELL LONGITUDINAL WELD (1/4T)

INTERMEDIATE TO LOWER SHELL CIRC. WELDS (3/4T)

LIMITING ART VALUES AT EOL:

1/4T, 238.20F 3/4T, 183.90F 2500 2250 LZAR TEST LIMIT S

2000 1750 -A U NACCEPTABLE OPERAT ION SAC CEPTABLE1 I

II I

I ll F

CR1.LIMIT/Hr I

CRITICALITY LIN T

DAMOD OI INSZRVICU DYDNOBTATIC TEST T?

PRNATDR0 (3B4 N)

I0I TIE 33EICX PERIOD OF TO 3OLR 2

1

(

0 -

260 250 300 350 Moderator Temperature 400 450 (Deg.F)

Surry Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 600F/hr) Applicable to End of License Renewal (With Margins of 00F and 0 psi for Instrumentation Errors)

CJ2 C-)

I HEATUP RATE J

HUP TO 6

/i 1500 1250 1000 750 500 250 0

BOLT-UP TEMP.

5o0

!60 150 Figure - 1 56O 4

I

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 19 of 39 MATERIAL PROPERTY BASIS LIMITING MATERIAL:

LOWER SHELL LONGITUDINAL WELD (1/4T)

INTERMEDIATE TO LOWER SHELL CIRC. WELDS (3/4T)

LIMITING ART VALUES AT EOLR:

1/4T, 238.20F 3/4T, 183.9°F 2500

= 2250 S2000 1750 1500 1250-UNACCEPTABLEJ OPERATION ACCEPTABLEH OPERATION H 1000 CO OLDOWNH IRATE S F/Hr 0a2VI i i

,0

'CO I i.i

i. I I I10 i5 2

2 200 250 300 I.

350 Moderator Temperature

(

400 450 De g F)

Surry Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0,20,40, 60 and 1 001F/hr) Applicable to End of License Renewal (With Margins of 0°F and 0 psi for Instrumentation Errors) 750t 500 250 0-BiLT-P E.

SII

[ i !

0 50 Figure - 2 500 5

0 7 19 7

-T= I I4 - -

i I

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 20 of 39

References:

1.

Virginia Power Topical Report VEP-NAF-3A, "Reactor Vessel Fluence Analysis Methodology," dated November, 1997.

2.

Draft Regulatory Guide DG-1053, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," June 1996 previous draft was DG-1 025, September 1993.

3.

NRC Reg. Guide 1.99 Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May, 1988

4.

Letter from J. P. O'Hanlon to USNRC, "Virginia Electric and Power Company, Surry and North Anna Power Stations Units 1 and 2, Surry 1 Reactor Vessel Surveillance Capsule X Analysis Report, GL 92-01, Revision 1, Supplement 1, Response to Request for Additional Information and Topical Report on Reactor Vessel Fluence Analysis Methodology," Serial No.98-252, dated June 18, 1998.

5.

WCAP-14040-NP-A, Rev. 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves",

January 1996.

6.

Tables 4.1-12 and 4.1-13, Surry Units 1 and 2 UFSAR

7.

Framatome Topical Report BAW-2241P-A, "Fluence and Uncertainty Methodologies," Rev. 01, December 1999.

8.

NRC Generic Letter 92-01, Rev. 1, "Reactor Vessel Structural Integrity, 10CFR50.54(f)," dated March 6, 1992.

9.

Dominion Letter, "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Response to Generic Letter 92-01, Reactor Vessel Structural Integrity," dated June 29, 1992.

10.

B&W Report BAW-2166, "B&W Owners Group Response to Generic Letter 92 01," dated June 1992

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 21 of 39

11.

NRC Letter, "Surry Power Station, Units 1 and 2 - Request for Additional Information Regarding Generic Letter 92-01, Revision 3," dated July 21, 1993.

12.

Dominion Letter, "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Response to Request for Additional Information Regarding Response to Generic Letter 92-01, Rev. 1," dated September 23, 1993.

13.

NRC Letter, "Generic Letter 92-01, Revision 1, Reactor Vessel Structural Integrity, Surry Power Station, Units 1 and 2, (TAC NOS. M83739 and M83740),"

dated May 24, 1994.

14.

Dominion Letter, "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Generic Letter 92-01, Revision 1, Reactor Vessel Structural Integrity Request for Additional Information," dated June 30, 1994.

15.

B&W Report BAW-2222, "Response to Closure Letters to Generic Letter 92-01, Revision 1," dated June 1994.

16.

NRC Letter, "NRC GL 92-01, Rev. 1 Reactor Vessel Structural Integrity, Supplement 1," dated May 19, 1995.

17.

Framatome Report BAW-2323, "Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of Surry Units 1 and 2 For Extended Life Through 48 Effective Full Power Years," June 1998.

18.

Framatome Report BAW-2178P-A, "Low Upper-Shelf Toughness Fracture Mechanics Analyses of Reactor Vessels of B&W Owners Reactor Vessel Working Group for Level C & D Service Loads," July 1994.

19.

Framatome Report BAW-2192P-A, "Low Upper-Shelf Toughness Fracture Analysis of Reactor Vessels B&W Owners Reactor Vessel Working Group for Level A & B Conditions," July 1994.

20.

NUREG/CR-6551, "Improved Embrittlement Correlations for Reactor Vessel Steels," prepared by Eason, E. D., Wright, J. E., and Odette, G. R., November 1998.

21.

Framatome Report BAW-2308, "Initial RTNDT of Linde 80 Weld Materials," July 2002.

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 22 of 39

22.

Framatome Report 51-5021094-00, "Consistent and Benchmarked Neutron Fluence Values for Reactor Vessel Materials Surveillance Capsules Containing Weld Material Fabricated with Weld Wire Heat Number "299L44" Applicable to Surry Unit 1 Weld Material SA-1 525 (Designated Weld "L2"), October 2002.

23.

Framatome Report 51-5021199-00, "Surry 1 Lower Shell Longitudinal welds and RVSP Material,", October 7, 2002.

24.

NUREG-1 801, "Generic Aging Lessons Learned (GALL) Report," dated July 2001.

25.

Framatome Report BAW-2350P, "Test Results of W1 Capsule, Master Integrated Reactor Vessel Surveillance Program," April 1999.

26.

Framatome Report BAW-2324, "Analysis of Capsule X, Virginia Power Surry Unit No. 1," April 1998.

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 23 of 39 Appendix A RG 1.99 Revision 2 Position 2.1 Calculations All Available Surveillance Data for Weld Wire Heat 299L44 Including Corrections for Irradiation Temperature and Chemistry Variation Effects

caIc98h (SM-1008 Rev. 0 Add. G).xls Table 2:

Surry Unit 1 Weld Material SA-1526 (All Data, Irrad. Temp and Chemistry Corrections)

Capsule ID (Including Source)

Capsule TM12-LGI (BWOG CR-3 tIrad.)

Capsule CR3-LG1 (EWOG CR-3 tread.)

Capsule ITMI2-Lýil (EIWOG CR-3 Iread.)

Three M Island Unit 1 Capu Capsule W-1 (CR-3 NBD)

M 1 Capsule T burryunit 1CapsuleV Sur Unt Capsule X Table 3:

Copper (wt%)

0,~o360 0360 0.230 0230 0.23_.00 Nickel (wt%)

0.700 0.700 0.670 0.670 0.700 0.640 0.640 0.640 Irradiation Temperature (F) 556.0 556.0 556.0 556.0 546.3 533.9 Fluence (x1E19) 0.130 0.755 0007 0.097 0.669 0.292 1,99 Measured Delta RT(NDT) (F) 216 202 226 166 74 262 171 250 234 Data Used In Assessing Vessel?

(Yes or No)

Yes Yes Yes Yes Yes Yes Yes Yes Yes Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 24 of 39 Surry Unit 1 Weld Material SA-1526 (All Data, Irrad. Temp and Chemistry Corrections)

Irradiation Measured Delta.

Adjusted Delta Capsule ID (Including Source)

Copper w(t%

Nickel (Wro)

Temperature (F)

Fluence Factor RT(NDT) (F)

RT(NDT) (F)

Capsule TMI2-LGI (BWOG CR-3 Iread 0.370 0.700 556.0 0.9477 2t6 198 Capsule CR3-L1 (BWDO CR-3 tread.

0.360 0.700 556.0 0.9212 202 1te Capsle MI-C3

• ead.1 0330 0.670 556.0 0.9909 226 225 Three Mile Islandn C

33 0670 5060 0.9648 166 167 Three Mie a

I E

0.330 0.670 5560 0.4108 74 78 Ca ule W-1 CR-3 NBD) 0.360 0.700 546.3 0.8673 262 234 Suyn3 0.640s 533.9 0.6633 171 184 Surry Unit 1 Capsule V 0.230 0.640 5398.

1.1881 250 283 SU Unit 1 Capsule 140230 0.640

04.

tt0 234 2683

• a* leX

.60 542.0 1.1296 234 2-68

• For credibility check. measured shift values are adjusted to average surveillance material chemistry and Irradiation temperature as required. See Table 4.

"- Predicted Delta-RT(NDT) Is based upot the RG 1.99 Revision 2 Position 2.1 Chemistry Factor determined with the Adjusted De0ta-RT(NDT) values. (225.5 degrees F)

Predicted Delta.

RT(NDT) (F)"

214 208 223 218 93 200 150 268 255 Adjusted - Predicted Dolta-RT(NDT) (F)

.6

-19

-14 34 35 15 13 SurveIllance Data Credible or Non Credible?

Credible Credible Credible Non-Credible Credible Non-Credible Non-Credible Credible Credible

Docket Nos. 50-280/281 50-338/339 caOC98h (SM-1-O08 Rev. 0 Add. G).xls Serial No.: 02-601 Table 4:

Surry Unit I Weld Material SA-1 526 (All Data, Irrad. Temp and Chemistry Corrections)

Beltline Material CF Determination It SUr.. Data Non.

Surveillance Data Credible, Verify Chemistry Factor Beltline Material Beittine Material Irradiation Position 1.1 Chemistry Position 2.1 Chemistry Credible or Non.

Conservatism of Applied to BeItflne Beltine Material ID Copper (wt%)

Nicket (M-%)

Temperature (F)

Factor Factor Credible?

Poslion 1.1 CF Materrial SA-520/299L4r 4

0.340 0.600 542.0 220.6 249.0 Non-Credible Conservative 2206 Measured shift values are adjusted to the average surveillance material chemistry and irradiation temperature, and are veifted to be within 2 sigma of the trend curve based on RG 1t99 Rev. 2 Position 1. I.

For a large poputatior of surveillance data (e.g., 9 data points for SA-t526), nue or two slightly non covsorvative dale points do not Invalidate the conclusion that use of RG 1.99 Rev. 2 Postlon 1. t Is conservative.

Ifsurveillance data are non-credible but the Pos. t. 1 CFI3 shown to ho coosrrvatlvo. the lower of tth Pos. t. t and Pos. 2. 1 thomlstry tactors Is applied to the bhirine rmathral with a full margin term.

If surveiilance data are non-c(rediblo andthe POa. t. t CFis shown to he nonconservative, the greater of rho Pos. t. t and Pos. 2.1 chemistry factors Is applied to the beftline material with a fullmargin tern".

Credibility and Conservatism Assessment Summary Conservatism Check forPos. t.1 CFrsfen SUm. Data Nun-Credible (1)

(2)

(3)

(4)

Are adjusted Temperature Chemistry Temperature Conrrection Chemistry Correction surveillance data Correction Correction Iruteillance Data Applied to Surv. Data for Applied to Suv. Data withil 2 sigma of Applied for Applied for Credible or Non-Application to Beitline for Application to Adjusted Delta.

Predicted Delta.

Adjusted - Predicted the applied CF Capsule ID (Including Source)

Credibility?

Credibility?

Credible?

Material?

Beltltne Material?

RT(NDT) (F)

RT(NDT) (F)

Delta-RT(NDT) (F) trend curve?

• Capsule TMI2-LGI (BWCG CR-3 Irrad.)

Yes Yes Credible Yes Yes 198 197 1

Conservative Capsute CR3-LG1 (BWDG CR-3 luad.)

Yes Yes Credible Yes Yes 188 191

-3 Conservative Capsule TMI2-LGI (BWOG CR-3 trrad.

Yes Yes Credibe Yes Y225 206 t 9 Conserva tive

....Three Mile Istand Unit 1 Capsule C Yes Yes N on-Cle dibte Yes Yes 567 200

.33 Conservative Three Mile Istand Unit 1 Capsute E Yes Yes Credible Yes Yes 70 05

-7 Conservative Capsule w-t (CR-3 N oD )

Yes Yes Non-Credible Yes Yes 234 104 49 Conservative Sure i Unit 1 Capsule T Yes Yes Non-Credible Yes Yes 167 130 46 Conservative S

lrr d Unit 1 Capsule V Yes Yes Credible Yes Yes23 24?

37 Conservative Surry Unit 1 Capsule V Yes Yes Credible Yes Yes 208 235 34 Conservative (I) For the credibility determination, a temperature correction is not applied to measured values Of transition temperature shill it applicable surveillance data were irradiated in a single reactor (i.e., were Irradiated at a similar temperature).

(2) For the credibility determination. a chemistry correction Is not applied to measured values of transition temperature shift it applicable surveillance data were obtained ftrom a single source (i.e., were machined from the same block of material).

(3) For determination of the tiline material chemistry factor, a temperature correction is not applied to measured values of transition temperature shift If applicable surveillance data were Irradiated In the reactor vessel which Is being evaluated (i.e., were irradiated at a similar temperature). A temperature correction Is applied only in the conservative direction.

(4) For determination of the ptetline materiat chemistry factor. a chemlsty correction (i.e., ratio procedure) is not applied to measured values of transition temperature shift It the chemical cormposition of applicable surveillance data Is essentialy Identical to the best-estimate chemical composition of the beliline material being evaluated.

-Predicted Defta-RT(NDT) Is based upon the RG 1.99 Revision 2 Position 1. 1 Chemistry Factor for the average chemical composition of the surveillance materials. (207.7 degrees F)

Page 25 of 39

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 26 of 39 Appendix B RG 1.99 Revision 2 Position 2.1 Calculations All Available Surveillance Data for Weld Wire Heat 299L44 Without Corrections for Irradiation Temperature and Chemistry Variation Effects

Copper (wt%)

0 370 0.360 0.330 0.330 Nickel (wM%)

0.700 0.700 0.670 0.670 Irradiation Temperature (F) 556.0 556.0 556.0 556.00.882 calc98h (SM-1008 Rev. 0 Add, G).xls Table 2:

Capsule ID (Including Source)

Capsule TMI2-LG1 (IWOG CA-3 Irrad.)

Capsule CR3-LG1 (BWOG CR-3 Irred.)

Capsule TMI2-LG1 (BWOG CR-3 Irrad.)

Three Mile Island Unit 1 Capsule C Three Mile Island Unit 1 Capsule E Capsule W-1 (CR-3 NBD)

Surry Unit 1 Capsule T Suny Unit 1 Capsule V Surry Unit 1 Capsule X Surry Unit 1 Weld Material SA-1526 (All Data, No Corrections for Irrad. Temp and Chemistry)

I 1

1 r

Data Used In Assessing Vessel?

(Yes or No)

Yes Yes Yes Yes Yes Yes Yes Yes Yes I

1-I I

I L ________

I ____________

I __________

I Fluence (xtE19) 0.830 0.755 0.968 0.3W0 0230 0.230 0.230 01600 0.700 0.640 0.640 06840 5513 0

74 Measured Delta RT(NDT) (F) 216 202 226 166 166 T548.3 533.9 538.8 542.0 30 689 0.292 1.992 1.599 262 171 250 234 Table 3:

Surry Unit 1 Weld Material SA-1526 (All Data, No Corrections for Irrad. Temp and Chemistry)

Capsule ID (Including Source)

Capsule TMI2-LG1 (BWOG CR-3 Irrad.)

Capsule CT3-LG1 (BWOG CR-3 Irrad.1 Capsule TM12-LG1 (BWOG CR-3 Irrad.1 r

1 r

r Coeoer twt %/

u.370 556

.9477

)

2:

F4 Irradiation 0700 556 202 0

0 Measured Delta-Adjusted Delta RT(NDT) (F)

RT(NDT) (F) 216 21-6 216 216 202 Predicted Delta RT(NDT) (F)"

210 204 T M a

i.

0 56.L 0,999 2226 226 219 The illlndU*1Casl n**

n*n*

Adjusted - Predicted Delta-RT(NDT) (F) 6

-2 7

Thr....

... e.e MC i6.i u.9b64 166 168 213

-47 Non-Credible Three Mile Island Unit 1 Capsule E 0.330 0,670 556.0 0.4108 74 74 91

.17 Credible Capsule W-1 (CR-3 NBD) 0.360 0.700 546.3 0.8873 262 262 196 66 Non-Credible Surry Unit I Capsule T 0.230 0.640 533.9 0.6633 171 171 147 24 Credible U

Capsule V 0.230 0.640 538.8 1.1881 250 250 263

-13 Credible Surry Unit 1 Capsule X 0.230 0.640 542.0 11.296 234 234 250

-16 Credible Surveillance Data Credible or Non Credible?

Credible Credible Credible A ______

I.

j ______

"For credibility check, measured shift values are adjusted to average surveillance material chemistty and irradiation temperature as required. See Table 4. This sensitivity case Includes no corrections for Irrad. temp. or chemistry.

"" Predicted Delta-RT(NDT) is based upon the RG 1.99 Revision 2 Position 2. 1 Chemistry Factor determined with the Adjusted Delta.RT(NDT) values. (221.2 degrees F)

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 27 of 39 U..*.*J

() I¢. I[ }

T 5 N Copper (vA%)

Temp uenceactor U.,0O E516.3 0.669 0.292 1.992 1.599 NickA1 IwtOA

• 0.700 55fi 0 O QA?7 0360 U./*J n Q*t*

Table 4:

calc98h (SM-1 008 Rev. 0 Add, G).xs Surry Unit 1 Weld Material SA-1526 (All Data, No Corrections for Irrad. Temp and Chemistry)

Beltline Material CF Determination Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 28 of 39

'Measured shift values are adjusted to the average surveillance material chemistry and irradiation temperature, and are verified to be within 2 sigma of the trend curve based on RG 1.99 Rev. 2 Position I. 7.

For a large population of surveillance data (e.g., 9 data points for SA. 1526), one or two slightly non-conservative data points do not invalidate the conclusion that use of RG 1.99 Rev. 2 Postion I. t is conservative.

If surveillance data are non-credible but the Pos. 1.1 CF is shown to be conservative, the lower of the Pos. 1.1 and POs. 2.1 chemistry factors is applied to the beltline material with a full margin term.

If surveillance data are non-credible and the Pos. 1. 1 CF is shown to be non-conservative, the greater of the Pos. 1. t and Pos. 2. t chemistry factors is applied to the beitline material with a full margin term.

Credibility and Conservatism Assessment Summary

______Conservatism Check for Pos. 1.1 CF when Surv Data Non-Credible (1)

(2)

(3)

(4)

Are adjusted Temperature Chemistry Temperature Correction Chemistry Correction surveillance data Correction Correction Surveillance Data Applied to Surv. Data for Applied to Surv. Data for within 2 sigma of Applied for Applied for Credible or Non-Application to Beltline Application to Beltline Adjusted Delta-Predicted Delta-Adjusted - Predicted the applied CF Capsule ID (Including Source)

Credibility?

Credibility?

Credible?

Material?

Material?

RT(NDT) (F)° RT(NDT) (F)...

Delta-RT(NDT) (F) trend curve? '

"Capsule TMI2-LG1 (BWOG CR-3 Irrad.)

No No Credible No No 216 197 19 Conservative Capsule CR3-LG 1 (BWOG CR-3 Irrad.)

No No Credible No No 202 191 11 Conservative Capsule TMI2-LG1 )BWOG CR-3 Irrad.)

No No Credible No No 226 206 20 Conservative Three Mile Island Unit I Capsule C No No Non-Credible No No 166 200

-34 Conservative Three Mile Island Unit I Capsule E No No Credible No No 74 85

-11 Conservative Capsule W-1 (CR-3 NBD)

No No Non-Credible No No 262 184 78 Non-Conservative Surry Unit I Capsule T No No Credible No No 171 138 33 Conservative Surry Unit 1 Capsule V No No Credible No No 250 247 3

Conservative Surry Unit I Capsule X No No Credible No No 234 235

-1 Conservative (1) For the credibility deterrmination, a temperature correction Is not applied to measured values of transition temperature shift If applicable surveillance data were irradiated in a single reactor (i.e.. were irradiated at a similar temperature). This sensitivity case assumes that all surveillance data were Irradiated at similar conditions.

(2) For the credibility determrlnation, a chemistry correction is not applied to measured values of transition temperature shift if applicable surveillance data were obtained from a single source (i.e., were machined from the same block of material). This sensitivity case assumes that all surveillance data were obtained from a single source.

(3) For determination of the beltilne material chemistry factor, a temperature correction is not applied to measured values of transition temperature shift if applicable surveillance data were Irradiated In the reactor vessel which is being evaluated (i.e., were irradiated at a slrimlar temperature). A temperature correction is applied only in the conservative direction.

This sensitivity case assumes that all surveillance data were Irradiated at conditions similar to those of the beltlilne material being evaluated.

(4) For determination of the beftline material chemistry factor, a chemistry correction (i.e., ratio procedure) is not applied to measured values of transition temperature shift if the chemical composition of applicable surveillance data Is essentially Identical to the best-estimate chemical composition of the belltine material being evaluated.

This sensitivity case assumes that the chemical composition of surveillance data Is essentially Identical to that of the beltllne material being evaluated.

- Predicted Delta-RT(NDT) is based upon the RG 1.99 Revision 2 Position 1.1 Chemistry Factor for the average chemical composition of the surveillance materials. (207. 7 degrees F)

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 29 of 39 Appendix C RG 1.99 Revision 2 Position 2.1 Calculations All Available Surveillance Data for Weld Wire Heat 299L44 Without Corrections for Irradiation Temperature and Chemistry Variation Effects and Using RG 1.99 Revision 2 Position 1.1 Chemistry Factors Based on Individual Surveillance Capsule Chemistries

Docket Nos. 50-280/281 50-338/339 calc98h (SM-1008 Rev. 0 Add. G).xs Serial No.: 02-601 Surry Unit 1 Weld Material SA-1526 (All Data, No Corrections, Conservatism Check Based on Individual Capsule CF's)

Page 30 of 39 Table 2:

Table 3:

Surry Unit 1 Weld Material SA-1526 (All Data, No Corrections, Conservatism Check Based on Individual Capsule CF's)

Surveillance Data Irradiation Measured Delta-Adjusted Delta-Predicted Della-Adjusted - Predicted Credible or Non Capsule ID (Including Source)

Copper (wt%)

Nickel (wt%)

Temperature iF)

Fluence Factor RT(NDT) (F)

RT(NDT) iF)"

RT(NDT) iF)*"

Delta-RT(NDT) iF)

Credible?

Capsule TMI2-LG1 (BWOG CR-3 Irrad.)

0.370 0.700 5560 0.9477 216 216 210 6

Credible Capsule CR3-LG1 (BWOG CR-3 Irrad.)

0.360 0.700 556.0 0.9212 202 202 204

-2 Credible Capsule TMI2-LG1 (BWOG CR-3 Irrad.)

0.330 0.670 556.0 0.9909 226 226 219 7

Credible Three Mile Island Unit 1 Capsule C 0.330 0.670 556.0 0.9648 166 166 213

-47 Non-Credible Three Mile Island Unit I Capsule E 0.330 0.670 556.0 0.4108 74 74 91

-17 Credible Capsule W-1 (CR-3 NBD) 0.360 0.700 546.3 0.8873 262 262 196 66 Non-Credible Su.y Unit 1 Capsule T 0.230 0.640 533.9 0.6633 171 171 147 24 Credible Surry Unit I Capsule V 0.230 0.640 538.8 1.1881 250 250 263

-13 Credible Sure/Unit 1 Capsule X 0.230 0.640 542.0 1.1296 234 234 250

-16 Credible

'For credibility check, measured shift values are adjusted to average surveillance material chemistry and irradiation temperature as required. See Table 4. This sensitivity case Includes no corrections for Irrad. temp. or chemistry.

-Predicted Delta-RT(NDT) is based upon the RG 1.99 Revision 2 Position 2. I Chemistry Factor determined with the Adjusted Delta-RT(NDT) values. (221.2 degrees F)

Data Used In Irradiation Measured Delta-Assessing Vessel?

Capsule ID (Including Source)

Copper (wt%)

Nickel (wt%)

Temperature iF)

Fluence (x1E19)

RT(NDT) (F)

(Yes or No)

Capsule TMI2-LG1 (BWOG CR-3 Irrad.)

0.370 0.700 556.0 0.830 216 Yes Capsule CR3-LG I (SWOG CR-3 Irrad.)

0.360 0.700 556.0 0.755 202 Yes Capsule TMI2-LG1 (BWOG CR-3 Irrad.)

0.330 0.670 556.0 0.968 226 Yes Three Mile Island Unit 1 Capsule C 0.330 0.670 556.0 0.882 166 Yes Three Mile Island Unit I Capsule E 0.330 0.670 556.0 0.097 74 Yes Capsule W-1 (CR-3 NBD) 0.360 0,700 546.3 0.669 262 Yes Surry Unit 1 Capsule T 0.230 0.640 533.9 0.292 171 Yes Surry Unit 1 Capsule V 0.230 0.640 538.8 1.992 250 Yes Surry Unit 1 Capsule X 0.230 0.640 542.0 1.599 234 Yes

calc98h (SM-1008 Rev. 0 Add. G).xls Surry Unit 1 Weld Material SA-1526 (All Data, No Corrections, Conservatism Check Based on Individual Capsule CF's)

Page 31 of 39 Beltline Material CF Determination II Surv. Data Non Surveillance Data Credible, Verify Chemistry Factor Beltilne Material Betlline Material Irradiation Position 2.1 Chemistry Credible or Non-Conservatism of Applied to Belttline Bettlino Material ID Copper (wt%)

Nickel (wt%)

Temperature (F)

Position 1. 1 Chemistry Factor Factor Credible?

Position 1t1 CF Material SA-15261299L44 0.340 0.680 542.0 220.6 221,2 Non-Credible Conservative 220.6

• Measured shift values are adjusted to the average surveillance material chemistry and irradiation temperature, and are verified to be within 2 sigma of the trend curve based on RG 1.99 Rev. 2 Position 1I.t.

For a large population of surveillance data (e.g., 9 data points for SA-1526), one or two slightly non-conservative data points do not invalidate the conclusion that use of RG 1.99 Rev. 2Postion t.1 is conservative.

"If surveillance data are non-credible but the Pos. 1. I CF is shown to be conservative, the lower of the Pos. 1t. and Pos. 2.1 chemistry factors is applied to the beltline material with a full margin term.

If surveillance data are non-credible and the Pos. I. 1 CF is shown to be non-conservative, the greater of the Pos. 1. 1 and Pos, 2. 1 chemistry factors is applied to the beitline material with a full margin term.

Credibility and Conservatism Assessment Summary Conservatism Check for Pos. 1.1 CF when Surv. Data Non-Credible (1)

(2)

(3)

(4)

Are adjusted Temperature Chemistry Temperature Correction Chemistry Correction surveillance data Correction Correction Surveillance Data Applied to Surv. Data for Applied to Surv. Data for within 2 sigma of Applied for Applied for Credible or Non-Application to Beltline Application to Beitline Adjusted Delta-Predicted Delta-Adjusted - Predicted the applied CF Capsule ID (Including Source)

Credibility?

Credibility?

Credible?

Material?

Material?

RT(NDT) (F)*

RT(NDT) (F)

Delta-RT(NDT) (F) trend curve?

• Capsule TMt2-LG1 (BWOG CR-3 Irrad.)

No No Credible No No 216 222

-6 Conservative Capsule CR3-LG1 (BWOG CR-3 Irrad.)

No No Credible No No 202 212

.10 Conservative Capsule TMI2-LG1 (BWOG CR-3 Irrad,)

No No Credible No No 226 213 13 Conservative Three Mile Island Unit 1 Capsule C No No Non-Credible No No 166 208

-42 Conservative Three Mile Island Unit I Capsule E No No Credible No No 74 88

-14 Conservative Capsule W-1 (CR-3 NeD)

No No Non-Credible No No 262 205 57 Non-Conservative Surry Unit 1 Capsule T No No Credible No No 171 117 54 Conservative Surty Unit 1 Capsule V No No Credible No No 250 209 41 Conservative Surry Unit 1 Capsule X No No Credible No No 234 199 35 Conservative (1) For the credibility determination. a temperature correction Is not applied to measured values of transition temperature shift if applicable surveillance data were irradiated in a single reactor (I.e., were Irradiated at a similar temperature). This sensitivity case assumes that all surveillance data were Irradiated at similar conditions.

(2) For the credibility determination, a chemistry correction Is not applied to measured values of transition temperature shift if applicable surveillance data were obtained from a single source (i.e., were machined from the same block of material). This sensitivity case assumes that all surveillance data were obtained from a single source.

(3) For determination of the behtline material chemistry factor, a temperature correction is not applied to measured values of transition temperature shift if applicable surveillance data were Irradiated In the reactor vessel which is being evaluated (i.e., were irradiated at a sirmilar temperature). A temperature correction is applied only in the conservative direction.

This sensitivity case assumes that all survetllance data were Irradiated at conditions similar to those of the belttlne materiat being evaluated.

(4) FOr determination of the beftline material chemistry factor, a chemistry correction (i.e., ratio procedure) is not applied to measured values of transition temperature shift if the chemical composition of applicable surveillance data Is essentially identical to the best-estimate chemical composition of the beltline material being evaluated.

This sensitivity case assumes that the chemical composition of surveillance data Is essentially Identical to that of the betltlne material being evaluated.

Table 4:

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 32 of 39 Appendix D RG 1.99 Revision 2 Position 2.1 Calculations Surry Unit 1 Surveillance Data for Weld Wire Heat 299L44 (No Temperature Correction, No Chemistry Correction)

Docket Nos. 50-280/281 50-338/339 calc98h (SM-1 008 Rev. 0 Add. G),x(s Serial No.: 02-601 Surry Unit 1 Weld Material SA-1526 (Surry 1 Data Only, No Temp Correction, No Chem Correction)

Page 33 of 39 Table 3:

Surry Unit 1 Weld Material SA-1 526 (Surry 1 Data Only, No Temp Correction, No Chem Correction)

Surveillance Data Irradiation Measured Delta-Adjusted Delta-Predicted Delta-Adjusted - Predicted Credible or Non Capsule ID (Including Source)

Copper (wt%)

Nickel (wt%)

Temperature (F)

Fluence Factor RT(NDT) (F)

RT(NDT) (F)

RT(NDT) (F)

Delta-RT(NDT) (F)

Credible?

Surry Unit I Capsule T 0.230 0,640 533.9 0.6633 171 171 143 28 Credible Surry Unit 1 Capsule V 0.230 0.640 538.8 1.1881 250 250 256

-6 Credible Surry Unit 1 Capsule X 0.230 0.640 542.0 1.1296 234 234 244

-10 Credible

• For credibiiity check, measured shift values are adjusted to average surveillance material chemistry and irradiation temperature as required. See Table 4. This sensitivity case Includes no corrections for Irrad. temp. or chemistry.

°Predicted Delta-RT(NDT) is based upon the RG 1.99 Revision 2 Position 2.1 Chemistry Factor determined with the Adjusted Delta-RT(NDT) values. (215.8 degrees F)

Table 2:

Data Used In Irradiation Measured Delta-Assessing Vessel?

Capsule ID (Including Source)

Copper (wI%)

Nickel (wt%)

Temperature (F)

Fluence (x1E19)

RT(NDT) (F)

(Yes or No)

Capsule TMI2-LG1 (BWOG CR-3 Irrad.)

0.370 0.700 556.0 0.830 216 No Capsule CR3-LG1 (BWOG CR-3 Irrad.)

0.360 0.700 556.0 0.755 202 No Capsule TMI2-LG1 (BWOG CR-3 Irrad.)

0.330 0.670 556.0 0,968 226 No Three Mile Island Unit 1 Capsule C 0.330 0.670 556.0 0.882 166 No Three Mile Island Unit 1 Capsule E 0.330 0.670 556.0 0.097 74 No Capsule W-1 (CR-3 NBD) 0.360 0,700 546.3 0.669 262 No Surry Unit 1 Capsule T 0.230 0.640 533.9 0.292 171 Yes Sury Unit 1 Capsule V 0.230 0.640 538.8 1.992 250 Yes Surry Unit 1 Capsule X 0.230 0.640 542.0 1.599 234 Yes

calc98h (SM-1008 Rev. 0 Add. G).xls Surry Unit 1 Weld Material SA-1526 (Surry 1 Data Only, No Temp Correction, No Chem Correction)

Table 4:

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 34 of 39 Beltline Material CF Determination If Surv.' Data Non Surveillance Data Credible, Veriry Chemistry Factor Bettline Material Beltline Material Irradiation Position 2.1 Chemistry Credible or Non-Conservatism of Applied to Beltline Beitline Material ID Copper (wt%)

Nickel (wt%)

Temperature (F) Position 1.1 Chemistry Factor Factor Credible?

Position 1.1 CF Material*"

SA-1526/299L44 0.340 0.680 542.0 220.6 215.8 Credible 215.8 Measured shift values am adjusted to the average surveillance material chemistry and irradiation temperature, and are verified to be within 2 sigma of the trend curve based on RG 1.99 Rev. 2 Position 1.1.

For a large population of surveillance data (e.g., 9 data points for SA-1526), one or two slightly non-conservative data points do not invalidate the conclusion that use of RG 1.99 Rev. 2 Postion 1. 1 is conservative, "If surveillance data are non-credible but the Pos. f.1 CF is shown to be conservative, the lower of the Pos t. 1 and Pos. 2.1 chemistry factors is applied to the belline material with a full margin term.

If surveillance data are non-credible and the Pos. 1. 1 CF is shown to be non-conservative, the greater of the Pos. 1. 1 and Pos. 2.1 chemistry factors is applied to the bettline material with a full margin term.

Credibility and Conservatism Assessment Summary Conservatism Check for Pos. 1.1 CF when Surv. Data Non-Credible (1)

(2)

(3)

(4)

Are adjusted Temperature Chemistry Temperature Correction Chemistry Correction surveillance data Correction Correction Surveillance Data Applied to Surv, Data for Applied to Surv. Data for within 2 sigma of Applied for Applied for Credible or Non-Application to Beltline Application to Beitline Adjusted Delta.

Predicted Delta-Adjusted - Predicted the applied CF Capsule ID (Including Source)

Credibility?

Credibility?

Credible?

Material?

Material?

RT(NDT) (F)

RT(NDT) (F)...

Deita-RT(NDT) (F) trend curve?

Capsule TMI2-LG1 (BWOG CR-3 Irrad.)

Capsule CR3-LG1 (BWOG CR-3 Irrad.)

Capsule TMI2-LG1 (BWOG CR-3 Irrad.)

Three Mile Island Unit 1 Capsule C Three Mile Island Unit 1 Capsule E Capsule W-t (CR-3 NBD)

Surry Unit 1 Capsule T No No Credible No No Surry Unit I Capsule V No No Credible No No Surry Unit 1 Capsule X No No Credible No No (1) For the credibility determination, a temperature correction is not applied to measured values of transition temperature shift if applicable surveillance data were irradiated In a single reactor (i.e., were irradiated at a similar temperature). This sensitivity case assumes that all surveillance data were Irradiated at similar conditions.

(2) For the credibility determination, a chemistry correction is not applied to measured values of transition temperature shift if applicable surveillance data were obtained from a single source (i.e.. were machined from the same block of material). This sensitivity case assumes that all surveillance data were obtained from a single source.

(3) For determination of the belilne material chemistry factor, a temperature correction is not applied to measured values of transition temperature shift if applicable surveillance data were Irradiated In the reactor vessel which is being evaluated (i.e., were irradiated at a similar temperature). A temperature correction is applied only in the conservative direction.

This sensitivity case assumes that all surveillance data were Irradiated at conditions similar to those of the beltltne material being evaluated.

(4) For determination of the beltline material chemistry factor, a chemistry correction (i.e.. ratio procedure) is not applied to measured values of transition temperature shift if the chemical composition of applicable surveillance data is essentially identical to the best-estimate chemical composition of the beitline material being evaluated.

This sensitivity case assumes that the chemical composition of surveillance data Is essentially Identical to that of the beftline material being evaluated.

- Predicted Delta-RT(NDT) Is based upon the RG 1.99 Revision 2 Position 1. 1 Chemistry Factor for the average chemical composition of the surveillance materials. (175.8 degrees F)

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 35 of 39 3

0 0.5 1

1.5 2

2.5 Fluence (x 1 E19 n/sq.cm.)

SA-1 526 (Case D) 300 250

£UU L

(0 150 S100 50 0

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 36 of 39 Appendix E RG 1.99 Revision 2 Position 2.1 Calculations All Surveillance Data for Weld Wire Heat 299L44 (Temperature Correction, No Chemistry Correction)

calc98h (SM-1008 Rev. 0 Add. G).xls Surry Unit 1 Weld Material SA-1 526 (All Data, Temp Correction, No Chem Correction)

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 37 of 39 Data Used In Irradiation Measured Delta-Assessing Vessel?

Capsule ID (Including Source)

Copper (wt%)

Nickel (wt%)

Temperature (F)

Fluence (xtE1g)

RT(NDT) (F)

(Yes or No)

Capsule TMI2-LG1 (BWOG CR-3 Irrad.)

0.370 0.700 556.0 0.830 216 Yes Capsule CR3-LG1 (BWOG CR-3 Irrad.)

0.360 0.700 556.0 0.755 202 Yes Capsule TMI2-LG1 (BWOG CR-3 Irrad.)

0.330 0.670 556.0 0.968 226 Yes Three Mile Island Unit 1 Capsule C 0.330 0.670 556.0 0.882 166 Yes Three Mile Island Unit 1 Capsule E 0.330 0.670 556.0 0.097 74 Yes Capsule W-1 (CR-3 NBD) 0.360 0.700 546.3 0.669 262 Yes Surry Unit 1 Capsule T 0.230 0,640 533.9 0.292 171 Yes Surry Unit 1 Capsule V 0.230 0.640 538.8 1.992 250 Yes Surry Unit 1 Capsule X 0.230 0.640 542.0 1.599 234 Yes Surry Unit 1 Weld Material SA-1526 (All Data, Temp Correction, No Chem Correction)

Surveillance Data Irradiation Measured Delta-Adjusted Delta-Predicted Delta-Adjusted - Predicted Credible or Non Capsule ID (Including Source)

Copper (wt%)

Nickel (wt%)

Temperature (F)

Fluence Factor RT(NDT) (F)

RT(NDT) (F)*

RT(NDT) (F)

Delta-RT(NDT) (F)

Credible?

Capsule TMI2-LG1 (BWOG CR-3 Irrad.)

0.370 0.700 556.0 0.9477 216 223 209 14 Credible Capsule CR3-LGI (BWOG CR-3 Irrad.)

0.360 0.700 556.0 0.9212 202 209 203 6

Credible Capsule TMI2-LG1 (BWOG CR-3 Irrad.)

0.330 0.670 556.0 0.9909 226 233 219 14 Credible Three Mile Island Unit 1 Capsule C 0.330 0.670 556.0 0.9648 166 173 213

-40 Non-Credible Three Mile Island Unit 1 Capsule E 0.330 0.670 556.0 0.4108 74 81 91

-10 Credible CapsuleW-1 (CR-3 N8D) 0.360 0.700 546.3 0.8873 262 259 196 63 Non-Credible Surr Unit 1 Capsule T 0.230 0.640 533.9 0.6633 171 156 146 9

Credible Surry Unit 1 Capsule V 0.230 0.640 538.8 1.1881 250 240 262

-23 Credible Surry Unit 1 Capsule X 0.230 0.640 542.0 1.1296 234 227 249

.22 Credible For credibility check, measured shift values are adjusted to average surveillance material chemistry and irradiation temperature as required. See Table 4. This sensitivity case Includes no corrections for chemistry.

Predicted Delta-RT(NDT) Is based upon the RG 1.99 Revision 2 Position 2.1 Chemistry Factor determined with the Adjusted Delta-RT(NDT) values. (220.8 degrees F)

Table 2:

Table 3:

calc98h (SM-1008 Rev. 0 Add. G).xls Surry Unit 1 Weld Material SA-1526 (All Data, Temp Correction, No Chem Correction)

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 38 of 39 Beltline Material CF Determination If Surv. Data Non Surveillance Data Credible, Verify Chemistry Factor Bettline Material Beltline Material Irradiation Position 2.1 Chemistry Credible or Non-Conservatism of Applied to Beltline Bettline Material ID Copper (wt%)

Nickel (wt%)

Temperature (F) Position 1.1 Chemistry Factor Factor Credible?

Position 1.1 CF Material SA-15261299L44 0.340 0.680 542.0 220.6 229.3 Non-Credible Conservative 220.6

'Measured shift values are adjusted to the average surveillance material chemistry and irradiation temperature, and are verified to be within 2 sigma of the trend curve based on RG 1.99 Rev. 2 Position 1. 1.

For a large population of surveillance data (e.g.. 9 data points for SA-1526), one or two slightly non-conservative data points do not invalidate the conclusion that use of RG 1.99 Rev. 2 Postion 1. 1 is conservative.

If surveillance data are non-credible but the Pos. 1,1 CF is shown to be conservative, the lower of the Pos. 1. t and Pos. 2.1 chemistry factors is applied to the bettline material with a full margin term.

If surveillance data are non-credible and the Pos. 1. 1 CF is shown to be non-conservative, the greater of the Pos. 1.1 and Pos. 2.1 chemistry factors is applied to the bettline material with a full margin term.

Credibility and Conservatism Assessment Summary Conservatism Check for Pos. 1.1 CF when Surv. Data Non-Credible (1)

(2)

(3)

(4)

Are adjusted Temperature Chemistry Temperature Correction Chemistry Correction surveillance data Correction Correction Surveillance Data Applied to Surv. Data for Applied to Surv. Data for within 2 sigma of Applied for Applied for Credible or Non-Application to Beltline Application to Beltline Adjusted Delta-Predicted Delta-Adjusted - Predicted the applied CF Capsule ID (Including Source)

Credibility?

Credibility?

Credible?

Material?

Material?

RT(NDT) (F)

RT(NDT) (F)...

Delta-RT(NDT) (F)'

trend curve?

• Capsule TMI2-LG1 (BWOG CR-3 Irrad.)

Yes No Credible Yes No 223 197 26 Conservative Capsule CR3-LG1 (BWOG CR-3 Irrad.)

Yes No Credible Yes No 209 191 18 Conservative Capsule TMI2-LG1 (BWOG CR-3 Irrad.)

Yes No Credible Yes No 233 206 27 Conservative Three Mile Island Unit 1 Capsule C Yes No Non-Credible Yes No 173 200

-27 Conservative Three Mile Island Unit 1 Capsule E Yes No Credible Yes No 81 85

-4 Conservative Capsule W-1 (CR-3 NBD)

Yes No Non-Credible Yes No 259 184 75 1 Non-Conservative Surry Unit 1 Capsule T Yes No Credible Yes No 156 138 18 Conservative um,/Unit 1 Capsule V Yes No Credible Yes No 240 247

-7 Conservative Surry Unit 1 Capsule X Yes No Credible Yes No 227 235

-8 Conservative (1) For the credibility determination, a temperature correction is not applied to measured values of transition temperature shift if applicable surveillance data were Irradiated In a single reactor (i.e., were irradiated at a similar temperature).

(2) For the credibility determination, a chemistry correction is not applied to measured values of transition temperature shift if applicable surveillance data were obtained from a single source (i.e., were machined from the same block of material). This sensitivity case assumes that all surveillance data were obtained from a single source.

(3) For determination of the bettline material chemistry factor, a temperature correction is not applied to measured values of transition temperature shift if applicable surveillance data were Irradiated In the reactor vessel which is being evaluated (i.e., were irradiated at a similar temperature). A temperature correction is applied only in the conservative direction.

(4) For determination of the bettline material chemistry factor, a chemistry correction (i.e., ratio procedure) is not applied to measured values of transition temperature shift it the chemical composition of applicable surveillance data is essentially Identical to the best-estimate chemical composition of the bettline material being evaluated.

This sensitivity case assumes that the chemical composition of surveillance data Is essentially Identical to that of the bettllne material being evaluated.

"Predicted Delta-RT(NDT) Is based upon the RG 1.99 Revision 2 Position 1. 1 Chemistry Factor for the average chemical composition of the surveillance materials. (207.7 degrees F)

Table 4:

Docket Nos. 50-280/281 50-338/339 Serial No.: 02-601 Page 39 of 39 0

0.5 1

1.5 2

2.5 3

Fluence (x 1E19 n/sq.cm.)

SA-1 526 (Case E) 300 250 200 LI.L

• -150 z I-100 50 0