NL-04-2138, Submittal of Revision 2 of the Pressure Temperature Limits Report

Submittal of Revision 2 of the Pressure Temperature Limits Report

ML043140288
Person / Time
Site:	Farley
Issue date:	11/05/2004
From:	Stinson L Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-04-2138
Download: ML043140288 (24)

Text

L M. Stinson (Mike)

Vice President Southern Nuclear Operating Company. Inc.

40 Inverness Center Parkway Post Office Box 1295 Birmingham. Alabama 35201 Tel 205.992.5181 Fax 205.992.0341 4

SOUTHERIN ANA COMPAW Eneriv toServe Yourl World' NlT-04-2 138 November 5, 2004 Docket No.:

50-348 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant - Unit I Submittal of Revision 2 of the Pressure Temperature Limits Report Ladies and Gentlemen:

In accordance with Section 5.6.6 of the Joseph M. Farley Nuclear Plant (FNP) Unit I Technical Specifications, Southern Nuclear Operating Company (SNC) hereby submits Revision 2 of the FNP Unit I Pressure Temperature Limits Report (PTLR). Changes include revision of the surveillance capsule withdrawal schedule (as approved by NRC letter dated March 15, 2004), replacement of the heatup and cooldown limit curves with new curves extending the service period to 29.0 effective full power years (EFPY), and updating of the surveillance data credibility analysis and supplemental data sections to reflect the results of the latest surveillance capsule analysis report (WCAP-16221 -NP, Rev. 0, submitted April 16, 2004).

This letter contains no NRC commitments. If you have any questions, please advise.

Sincerely, L. M. Stinson LMS/DXVD/sdl

Enclosure:

Joseph M. Farley Nuclear Plant Unit I Pressure Temperature Limits Report, Revision 2 cc:

Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Mr. J. R. Johnson, General Manager - Plant Farley RTYPE: CFA04.054; LC# 14170 U. S. Nuclear Regulatorv Commission Dr. W. D. Travers, Regional Administrator Mr. S. E. Peters, NRR Project Manager - Farley Mr. C. A. Patterson, Senior Resident Inspector - Farley A I U(,

Q )A

Joseph M. Farley Nuclear Plant Unit I Pressure Temperature Limits Report REVISION 2

/O-Z?-o (.'

DATE

PRESSURE TEMPERATURE LIMITS REPORT Table Of Contents List of Tables................................................................................................................................................

iii List of Figures............................................................

1.0 RCS Pressure Temperature Limits Report (PT is LR)..

I 2.0 Operating Limits..

1 2.1 RCS Prcssure/Tcmpcrature (P/T) Limits (LCO - 3.4.3).

2.2 RCP Operation Limits.

3.0 Reactor Vessel Material Surveillance Program....................................................................................6 4.0 Reactor Vessel Surveillance Data Credibility 7

5.0 Supplemental Data Tables

.12 6.0 References.

1l FARLEY UNIT I ii REVISION 2

PRESSURE TEMPERATURE LIMITS REPORT List of Tables 2-1 Farley Unit 1 29.0 EFPY Ileatup Curve Data Points.................................................................

4 2-2 Farcye Unit 1 29.0 EFPY Cooldown Curve Data Points................................................................. 5 3-1 Surveillance Capsule Withdrawal Schedule.................................................................

6 4-1 Surveillance Capsule Data Calculation of Best-Fit Line as Described in Position 2.1 of Regulatory Guide 1.99, Revision 2................................................................

10 4-2 Scatter of ARTNDT 'alues About a Best-Fit Line for Surveillance Plate Material...................... 11 4-3 Scatter of ARTNDT Values About a Best-Fit Line for Surveillance Weld Material..................... 11 5-1 Comparison of Surveillance Material 30 FT-LB Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions................ 13 5-2 Calculation of Chemistry Factors Using Surveillance Capsule Data........................................... 14 5-3 Reactor Vessel Toughness Table (Unirradiated)................................................................

15 5-4 Reactor Vessel Fluence Projections................................................................

15 5-5 Summary of Adjusted Reference Temperatures (ARTs) for Reactor Vessel Beltline Materials at the 1/4-T and 3/4-T Locations for 29.0 EFP1.............................................. 16 5-6 Calculation of Adjusted Reference Temperature at 29.0 EFPV for the Limiting Reactor Vessel Material - Lower Shell Plate B6919-1................................................................

17 5-7 Pressurized Thermal Shock (RTp-rs) Values for 36 EFPY............................................................ 18 FARLEY UNIT I ii REVISION 2

PRESSURE TEMPERATURE LIMITS REPORT List of Figures 2-1 Farley Unit I Reactor Coolant System fleatup Limitations....................................................

2 2-2 Farley Unit I Reactor Coolant System Cooldown Limitations....................................................

3 FARLEY UNIT I iv REVISION 2

PRESSURE TEMPERATURE LIMITS REPORT 1.0 RCS Pressure Temperature Limits Report (PTLR)

This PTLR for Farley Nuclear Plant - Unit I has been prepared in accordance with the requirement of Technical Specification (TS) 5.6.6. Revisions to the PTLR shall be provided to the NRC after issuance.

This report affects TS 3.4.3, RCS Pressure/Temperature Limits (P/T) Limits. All TS requirements associated with low temperature overpressure protection (LTOP) are contained in TS 3.4.12, RCS Overpressure Protection Systems.

2.0 Operating Limits The limits for TS 3.4.3 are presented in the subsection which follows and were developed using the methodologies specified in TS 5.6.6. The operability requirements associated with LTOP are specified in TS LCO 3.4.12 and were determined to adequately protect the RCS against brittle fracture in the event of an LTOP transient in accordance with the methodology specified in TS 5.6.6. The limitation on the number of operating reactor coolant pumps (RCPs) is necessary to assure operation consistent with the pressure corrections incorporated in the P/T limits for flow losses associated with the RCPs.

2.1 RCS Pressure/Temperature (P/T) Limits (LCO - 3.4.3) 2.1.1 The minimum boltup temperature is 750F.

2.1.2 The RCS temperature rate-of-change limits are:

a. A maximum heatup of 1007F in any one hour period.

b. A maximum cooldowrn of 1000F in any one hour period.

c. A maximum temperature change of less than or equal to 100F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

2.1.3 The RCS P/T limits for heatup and cooldown are specified by Figures 2-1 and 2-2, respectively.

2.2 RCP Operation Limits 2.2.1 The number of operating RCPs is limited to one at RCS temperatures less than I 107F with the exception that a second pump may be started for the purpose of maintaining continuous flow while taking the operating pump out of service.

FARLEY UNIT I I

REVISION 2

a.

-CL C)

IL C

2500 175 125001 2000 ~

1750 1500 1250 100 70 PRESSURE TEMPERATURE LIMITS REPORT LEAK TEST LIMIT

• Limiting Material:

1

~

,Lnimiing ART Values at29.0OEFPY:

1 1 4 1. 1 7 lI F

,I F

I Critcaity Ulr for 3141. 4S F ]

~ : : : -1:

F /tv H e atup UNACCEPT BLE_:ACCEPTABLE OPEERATION Heallt Ra ACCEPTABLE:

OPR6D~0 50 10 0 jI 200 50 30 30 40 45 00 Indicated Temperature (Degrees F)

Figure 2-1 Farley Unit I Reactor Coolant System Heatup Limitations (Hclatup Rates up to IOO0F/hr) Applicable to 29.0 EFPY (adjusted to include 60 psi AP at RCS temperatures Ž I I 00F and 27 psi AP for RCS temperatures <I I 100F). Includes vessel flange requirements of lO0'F and 561 psig per 10 CFR 50, Appendix G. 11I FARLEY UNIT I 2RVSO 2

REVISION 2

PRESSURE TEMPERATURE LIMITS REPORT C) in In2 Ca

.V C

2500 2250 2000 1750 i1500 1250

.1000 I OO 750 500 250 0

0

.I.'I

'I l'

j IlI I 'I I 1' TI T

1-Ltiti" Material:

Lower Shell Plate B6919-1 Lknhnig ART Values at 29.0OEFPY:

I*

1/4.. 170 F a

a

.3T, 14F

, a,:f a.-,

OPERATION' C Iodow

.Rate I',,

'I-r' I '

(DegreeF *hr)

I I*

60

• *Min.

RCS o*tup Temperature 75 Lrwrhl~be61-

.f

.l 50 100 150 200 250 300

'I

.1

.I I

I I

I

.I I

7 I

I I

I I

i I

I I

l l l

• • z I I

l l

I I

I I

I I

• I

• l, I

I I

I i

I j

j

• l I

I I

I I

I I

I I

l l

l

.,. I

.. I l l l 500

• I

, II l;ACCEPTABLE

'OPERATION I

a a

a a

a.

35 40

,Aj 150 Indicated Temperature (Degrees F)

Figure 2-2 Farlcy Unit I Reactor Coolant System Cooldown Limitations (Cooldowvn Rates up to 1 00°F/hr)

Applicable to 29.0 EFPY (adjusted to include 60 psi AP at RCS temperatures Ž I 10°F and 27 psi AP for RCS temperatures < II OWF). Includes vessel flange requirements of I 80°F and 561 psig per 10 CFR 50, Appendix G. 1Il FARLEY UNIT I 3

REV'ISION 2

PRESSURE TEMPERATURE LIMITS REPORT 60OF I 60F Criticality LimitI 100F

° I00OFCriticalityLimitI Leak Test Limit T

I P

I T

I P

I T

I P

T I

I T

I P

75 80 85 90 95 100 105 110 110 115 120 125 130 135 140 145 1so 155 160 165 170 175 180 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 260 265 270 275 280 285 290 295 300 305 310 315 320 325 330 473 473 473 473 473 473 473 474 441 444 447 452 458 465 473 482 492 503 515 528 542 557 561 574 591 611 631 653 677 703 730 760 791 826 862 902 944 989 1038 1091 1147 1207 1271 1341 1415 1494 1579 1670 1768 1872 1983 2102 2229 2364 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 300 305 310 315 320 325 330 335 340 345 350 355 360 365 370 0

467 456 448 443 441 440 441 444 447 452 458 465 473 482 492 503 515 528 542 557 574 591 611 631 653 677 703 730 760 791 826 862 902 944 989 1038 1091 1147 1207 1271 1341 1415 1494 1579 1670 1768 1872 1983 2102 2229 2364 75 80 85 90 95 100 105 110 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 260 265 270 275 280 285 290 295 300 305 310 315 320 325 330 335 340 345 350 438 438 438 438 438 438 438 438 405 405 405 405 406 408 411 415 420 426 433 441 450 460 472 484 497 512 528 546 565 585 608 632 657 685 715 747 782 819 859 902 948 998 1051 1108 1169 1234 1305 1380 1460 1546 1639 1737 1843 1956 2076 2204 2341 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 300 305 310 315 320 325 330 335 340 345 350 355 360 365 370 375 380 385 390 0

469 454 441 430 422 415 411 407 406 405 406 408 411 415 420 426 433 441 450 460 472 484 497 512 528 546 565 585 608 632 657 685 715 747 782 819 859 902 948 998 1051 1108 1169 1234 1305 1380 1460 1546 1639 1737 1843 1956 2076 2204 2341 276 2000 298 2485 Table 2-1 Farlcy Unit 1 29.0 EFPY Heatup Curve Data Points (adjusted to include 60 psi AP at RCS temperatures 11 0IF and 27 psi AP for RCS temperatures < II O1F) Il FARLEY UNIT I 4

REVISION 2

PRESSURE TEMPERATURE LIMITS REPORT 0OF 20 OF T

40 0 F 1

66 0

OF l

100 0 F T

-T -

I T

I _

I.

I P

I T

I P

I T--F 75 80 85 90 95 100 105 110 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 260 265 270 275 280 285 290 295 300 305 310 315 320 511 515 520 524 530 535 541 548 515 522 530 538 546 556 561 561 561 561 561 561 561 561 561 676 698 719 741 765 791 819 849 880 915 952 991 1034 1080 1129 1182 1238 1299 1364 1434 1509 1589 1675 1768 1867 1972 2086 2206 2336 75 80 85 90 95 100 105 110 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 180 185 190 195 200 205 210 215 220 225 230 235 240 245 474 478 483 488 493 499 505 512 479 486 494 503 512 522 532 543 556 561 561 561 561 561 561 651 672 694 717 743 770 799 831 865 902 941 983 1028 1077 75 80 85 90 95 100 105 110 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 180 185 190 195 200 205 210 215 220 225 230 235 240 245 436 440 445 450 456 462 469 476 443 450 458 467 477 487 498 510 523 537 552 561 561 561 561 624 645 669 694 721 750 781 815 851 890 932 977 1025 1077 75 80 85 90 95 100 105 110 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 180 185 190 195 200 205 210 215 220 225 230 235 240 398 402 407 412 418 424 431 439 406 413 422 431 441 452 464 476 490 505 520 537 556 561 561 597 620 644 671 700 731 764 800 838 880 925 973 1024 75 80 85 90 95 100 105 110 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 318 323 329 334 341 347 355 363 330 338 348 358 369 381 394 408 423 440 458 477 497 520 544 570 598 629 661 696 734 775 819 867 918 973 1024 Table 2-2 Farlcy Unit 1 29.0 EFPY Cooldown Curve Data Points (adjusted to include 60 psi AP at RCS temperatures : I 100F and 27 psi AP for RCS temperatures < I 1 01F) 1 ']

FARLEY UNIT I 5s REVISION 2

PRESSURE TEMPERATURE LIMITS REPORT 3.0 Reactor Vessel Material Surveillance Program The reactor vessel material surveillance program is in compliance with 10 CFR 50, Appendix H, and is described in Section 5.4.3.6 of the Farley FSAR. The removal schedule is provided in Table 3-1. The results of these examinations shall be used to update Figures 2-1 and 2-2 if the results indicate that the adjusted reference temperature (ART) for the limiting beltline material exceeds the ART used to generate the P/T limits shown in Figures 2-1 and 2-2 for the specified fluence period.

Table 3-1 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE (a)

Capsule Capsule Location Lead Removal Fluence (Degree)

Factor EFPY (n/cm2) y(cl 343 3.24 1.15 6.12x IO8 U(0 107 3.34 3.08 1.73 x 10'9 X i" 287 3.35 6.11 3.06 x 10'9 W(0 110 3.01 12.43 4.75 x 10 (d}

VIC 290 3.04 20.16 7.14 x 10'l9" Z

340 3.04

-24 8.44 x 10'9<0¢g)

NOTES:

(a) WCAP-16221-NP, Revision 0 110l (b) Effective Full Power Years (EFPY) from plant startup (c) Plant-specific evaluation (d) This fluence is not less than once or greater than twice the peak EOL fluence for the initial 40-year license term.

(e) This flucnce is not less than once or greater than twice the peak EOL fluence for a 20-year license renewal term to 60 years.

(f) This projected fluencc is not less than once or greater than twice the peak EOL fluence for an additional 20-year license renewal term to 80 years.

(g) This projected fluencc was obtained using WCAP-16221-NP, Rev. 00 l, Table 6-2, "Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Basc Metal Interface." This table projects neutron fluences at various EFPY. The tabulated fluence values for neutrons with E>1.0 McV were found at 00 azimuth (the peak fluencc position) for EFPY values above and below the desired withdrawal EFPY. The fluence value corresponding to the desired withdrawal EFPY was then determined by interpolation from the tabulated values and multiplied by the capsule lead factor to yield the projected capsule fluence value listed above FARLEY UNIT I 6

REVISION 2

PRESSURE TEMPERATURE LIMITS REPORT 4.0 Reactor Vessel Surveillance Data Credibility Regulatory Guide 1.99, Revision 2, describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position C.2 of Regulatory Guide 1.99, Revision 2, describes the methodology for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Position C.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.

To date, there have been five surveillance capsules removed from the Farley Unit I reactor vessel.

In accordance with the discussion of Regulatory Guide 1.99, Revision 2, there are five requirements that must be met for the surveillance data to be judged credible.

The purpose of this evaluation is to apply the credibility requirements of Regulatory Guide 1.99, Revision 2, to the Farley Unit I reactor vessel surveillance data and determine if the Farley Unit I surveillance data is credible.

Criterion 1:

Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.

The beltline region of the reactor vessel is defined in Appendix G to 10 CFR 50, Fracture Toughness Requirements, December 19, 1995, to be:

the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage.

The Farley Unit I reactor vessel consists of the following beltline region materials:

Intermediate shell plates B6903-2 and B6903-3; Lower shell plates B6919-1 and B6919-2; Intermediate shell longitudinal weld seams19-894 A & B, heat number 33A277, Linde 1092 flux, flux lot 3889; Lower shell longitudinal weld seams20-894 A & B, heat number 90099, Linde 0091 flux, flux lot 3977; and Circumferential weld 11-894, heat number 6329637, Linde 0091 flux, flux lot 3999.

FARLEY UNIT I 7

REVISION 2

PRESSURE TEMPERATURE LIMITS REPORT Per WCAP-8810rO', the Unit I surveillance program was based on ASTM E185-73, Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels. Per Section 4.1 of ASTM El 85-73, the base metal and weld metal to be included in the program should represent the material that may limit the operation of the reactor during its lifetime. The test material should be selected on the basis of initial transition temperature, upper shelf energy level, and estimated increase in transition temperature considering chemical composition (copper and phosphorus) and neutron fluence.

Therefore, at the time the Farley Unit I surveillance capsule program was developed, lower shell plate B6919-1 was judged to be most limiting based on the above recommendations and was utilized in the surveillance program.

The surveillance program weld for Farley Unit I was fabricated using the same heat of weld wire used to fabricate the middle shell axial seams19-894 A & B (heat 33A277). The results of mechanical property tests performed on the surveillance weld are considered to be representative of the property changes expected in the reactor vessel beltline seams.

Therefore, the materials selected for use in the Farley Unit I surveillance program were those judged to be most likely controlling with regard to radiation embrittlement according to the accepted methodology at the time the surveillance program was developed. Based on the above, the Farley Unit I surveillance program meets the requirements of Criterion 1.

Criterion 2:

Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 fl-lb temperature and upper shelf energy, unambiguously.

Plots of Charpy energy versus temperature for the unirradiated condition are presented in the Unit I reactor vessel surveillance program description contained in WCAP-88101').

Plots of Charpy energy versus temperature for the irradiated conditions are presented in the reactor vessel surveillance capsule reports for capsules y [61, U III, X I'], W 12'land V1'01.

Based on engineering judgment, the scatter in the data presented in these plots is small enough to determine the 30 fl-lb temperature and upper shelf energy of the Farley Unit I surveillance materials unambiguously. Therefore, the Farley Unit I surveillance program meets the requirements of Criterion 2.

FARLEY UNIT I 8

REVISION 2

PRESSURE TEMPERATURE LIMITS REPORT Criterion 3:

When there are two or more sets of surveillance data from one reactor, the scatter of ARTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 280F for welds and 17'F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82.

The least squares method, as described in Regulatory Position 2. 1, will be utilized in determining a best-fit line for this data to determine if this criterion is met.

[Continued on the following page]

FARLEY UNIT I 9

REVISION 2

PRESSURE TEMPERATURE LIMITS REPORT Table 4-1 SURVEILLANCE CAPSULE DATA CALCULATION OF BEST-FIT LINE AS DESCRIBED IN POSITION 2.1 OF REGULATORY GUIDE 1.99, REVISION 2 (d)

-ibr FF W ARTNDT FFx I FF 2 Material Capsule F (b NOT ARTNDT 2X)

I I I I

(xy)

()

Lower Shell Y

0.612 0.862 64.83 55.883 0.744 Plate B6919-1 U

1.73 1.151 110.34 127.001 1.324 (Longitudinal)

X 3.06 1.295 129.71 167.974 1.678 W

4.75 1.392 145.57 202.633 1.938 V

7.14 1.466 178.01 260.963 2.149 Lower Shell Y

0.612 0.862 70.45 60.728 0.744 Plate B6919-1 U

1.73 1.151 100.51 115.687 1.324 (Transverse)

X 3.06 1.295 110.72 143.382 1.678 W

4.75 1.392 150.54 209.552 1.938 V

7.14 1.466 161.87 237.301 2.149 SUM:

1581.104 15.666 CFB6 919.1 = X(FF

• ARTNDT) + 7:(FF2) = (1 581.104) + (15.666) =1 00.9°F Wcld Metal Y

0.612 0.862 72.92 62.857 0.744 U

1.73 1.151 81.13 93.381 1.324 X

3.06 1.295 93.19 120.681 1.678 W

4.75 1.392 104.17 145.005 1.938 V

7.14 1.466 123.29 180.743 2.149 SUM:

602.667 7.833 CFsuw, wld = X(FF

• ARTNDT) + E(FF2) = (602.667) + (7.833) = 76.9°F NOTES:

(a) F = Fluence (10'9 n/cm2, E > 1.0 MeV)

(b) FF = Fluencc Factor = F4028 -0 Iogf)

(c) ARTNDT values are the measured 30 ft-lb shift values from Appendix C of WCAP-16221, Revision 0 flu (d) Data from Table D-l, WCAP-16221, Revision 0110]

FARLEY UNIT I 10 REVISION 2

PRESSURE TEMPERATURE LIMITS REPORT Table 4-2 SCATTER OF ARTNDT VALUES ABOUT A BEST-FIT LINE FOR SURVEILLANCE PLATE MATERIAL(a)

[

Lower Shell l

CF ARTNDT Best Fit Scatter of ARTNDT Plate B6919-1 Capsule (Best Fit FF (30 fi-lb)

ARTNDT (OF)

Orientation Slope)

(OF)

(OF)

Y 100.9 0.862 64.83 86.98

-22.2 U

100.9 1.151 110.34 116.14

-5.8 Longitudinal X

100.9 1.295 129.71 130.67

-1.0 W

100.9 1.392 145.57 140.45 5.1 V

100.9 1.466 178.01 147.92 30.1 Y

100.9 0.862 70.45 86.98

-16.5 U

100.9 1.151 100.51 116.14

-15.6 Transverse X

100.9 1.295 110.72 130.67

-20.0 X

100.9 1.392 150.54 140.45 10.1 V

100.9 1.466 161.87 147.92 14.0 NOTES:

(a) Data from Table D-2, WCAP-16221-NP, Revision 0['° The scatter of ARTNDT values about a best-fit line drawn with the y-intercept equal to zero, as described in Regulatory Position 2.1, should be less than 17'F for base metal. As shown above, the scatter of three of the data points are not within 17'F of the best-fit line. Therefore, Criterion 3 is not met for the Farley Unit I surveillance plate material. Since not all of the data is within 17'F of the best fit line, SNC has chosen to use the CF from this surveillance data along with a G, of 17'F when predicting the Farley Unit I vessel properties.

Table 4-3 SCATTER OF ARTNDT VALUES ABOUT A BEST-FIT LINE FOR SURVEILLANCE WELD MATERIAL(a)

CF ARTNDT Best Fit Scatter of ARTNDT Material Capsule (Best Fit FF (30 fl-b) j ARTNDT (OF) l Slope)

(OF) l (OF)

Y 76.9 0.862 72.92 66.29 6.6 U

76.9 1.151 81.13 88.51

-7.4 Weld Metal X

76.9 1.295 93.19 99.59

-6.4 W

76.9 1.392 104.17 107.04

-2.9 V

76.9 1.466 123.29 112.74 10.6 NOTES:

(a) Data from Tablc D-2, WCAP-16221-NP, Revision 0101 FARLEY UNIT I I1I REVISION 2

PRESSURE TEMPERATURE LIMITS REPORT The scatter of ARTNDT values about a best-fit line drawn with the y-intercept equal to zero, as described in Regulatory Position 2.1, is less than 280F as shown above. Therefore, Criterion 3 is met for the Farley Unit I surveillance weld material.

Criterion 4:

The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +250F.

The Farley Unit I capsule specimens are located in the reactor between the neutron shielding pads and the vessel wall and are positioned opposite the center of the core. The test capsules are in guide tubes attached to the neutron shielding pads. The location of the specimens with respect to the reactor vessel beiline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions and will not differ by more than 250F. Therefore, the Farley surveillance program meets the requirements of Criterion 4.

Criterion 5:

The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the data base for that material.

The Farley Unit I surveillance program does not include correlation monitor material. Therefore, Criterion 5 is not applicable to Farley Unit 1.

CONCLUSION:

Based on the preceding responses to the criteria of Regulatory Guide 1.99, Revision 2, Section B, and the application of engineering judgment, the Farley Unit I surveillance plate material data is not credible and the Farley Unit I surveillance weld data is credible.

5.0 Supplemental Data Tables Table 5-1 contains a comparison of measured surveillance material 30 fl-lb transition temperature shifts and upper shelf energy decreases with Regulatory Guide 1.99, Revision 2, predictions.

Table 5-2 shows the calculation of the surveillance material chemistry factors using surveillance capsule data.

Table 5-3 provides the unirradiated Farley Unit I reactor vessel toughness data.

Table 5-4 provides a summary of the fluences used in the PTS evaluation.

Table 5-5 provides a summary of the adjusted reference temperatures (ARTs) of the Farley Unit I reactor vessel beltline materials at the l/4-T and 3/4-T locations for 29.0 EFPY.

Table 5-6 shows the calculation of the ART at 29.0 EFPY for the limiting Farley Unit I reactor vessel material (lower shell plate B6919-1).

Table 5-7 provides RTPTS values for Farley Unit I for 36 EFPY.

FARLEY UNIT I 12 REVISION 2

PRESSURE TEMPERATURE LIMITS REPORT Table 5-1 COMPARISON OF SURVEILLANCE MATERIAL 30 FT-LB TRANSITION TEMPERATURE SHIFTS AND UPPER SHELF ENERGY DECREASES WITH REGULATORY GUIDE 1.99, REVISION 2, PREDICTIONS (a) 30 fl-lb Transition Upper Shelf Energy Temperature Shift Decrease Material Capsule Fluence(d)

Predicted Measured Predicted Measured (x Io0' n/cm2, (OF) (a)

(OF) (b)

(%) (a)

(%)(C)

E > 1.0 MeV)

Lower Shell Y

0.612 84.30 64.83 20 8.6 Plate B6919-1 U

1.73 112.57 110.34 25.5 22.9 (Longitudinal)

X 3.06 126.65 129.71 29 18.6 W

4.75 136.14 145.57 33 22.1 V

7.14 143.37 178.01 35 22.1 Lower Shell Y

0.612 84.30 70.45 20 0

PlateB6919-1 U

1.73 112.57 100.51 25.5 8.9 (Transverse)

X 3.06 126.65 110.72 29 11.1 W

4.75 136.14 150.54 33 15.6 V

7.14 143.37 161.87 35 20 Surveillance Y

0.612 67.32 72.92 25 12.1 Program U

1.73 89.89 81.13 32 29.5 Weld Metal X

3.06 101.14 93.19 36 22.8 W

4.75 108.72 104.17 40 26.2 V

7.14 114.49 123.29 42.5 26.2 Heat Affected Zone Y

0.612 32.55 10.3 Material U

1.73 160.37 25.8 X

3.06 136.91 21.9 WV 4.75 126.04 14.2 V

7.14 174.39 16.8 NOTES:

(a) Data fromTable 5-10, WCAP-16221-NP, Revision Oll0 FARLEY UNIT I 1 3 REVISION 2

PRESSURE TEMPERATURE LIMITS REPORT Table 5-2 CALCULATION OF CHEMISTRY FACTORS USING SURVEILLANCE CAPSULE DATA(d)

Material Capsule Capsule e",

FF(b)

ARTNDT(C)

FF*ARTNDT FF2 Lower Shell Y

0.612 0.862 64.83 55.883 0.744 Plate B6919-1 U

1.73 1.151 110.34 127.001 1.324 (Longitudinal)

X 3.06 1.295 129.71 167.974 1.678 W

4.75 1.392 145.57 202.633 1.938 V

7.14 1.466 178.01 260.963 2.149 Lower Shell Y

0.612 0.862 70.45 60.728 0.744 Plate B6919-1 U

1.73 1.151 100.51 115.687 1.324 (Transverse) x 3.06 1.295 110.72 143.382 1.678 W

4.75 1.392 150.54 209.552 1.938 V

7.14 1.466 161.87 237.301 2.149 SUM:

1581.104 15.666 CFB69 19.1 = X(FF

• RTNDT) + X(FF2) = (1581.104) + (15.666) = 100.9°F Surveillance Weld Y

0.612 0.862 118.13 (72.92)(c) 101.828 0.744 Material U

1.73 1.151 131.43 (81.13)(c' 151.276 1.324 X

3.06 1.295 150.97 (93.19)('c 195.506 1.678 W

4.75 1.392 168.76 (104.17)(c) 234.914 1.938 V

7.14 1.466 199.73 (123.29)(c) 292.804 2.149 SUM:

976.328 7.833 CF Su.r,. Weld = X(FF

• RTNDrT) + X( FF2) = (976.328) + (7.833) = 124.6°F NOTES:

(a) f= fluence (x 10'9 n/cm2, E > 1.0 McV) from WCAP-16221-NP 1101 (b) FF = fluence factor= f102s8 -01 logr)

(c) ARTNDT values from Table 4-1 (shown in parentheses) were multiplied by a ratio factor of 1.62 (CF,,,.l + CF,..td = 126.2 + 78.1 = 1.62), to calculate the best fit chemistry factor (CF) as provided by Reg.

Guide 1.99 Rev. 2, Position 2.1.

(d) Based on Table D-l, WCAP-1 622 1 -NP, Revision 01"]

FARLEY UNIT I 14 RE-VISION 2

PRESSURE TEMPERATURE LIMITS REPORT Table 5-3 REACTOR VESSEL TOUGHNESS TABLE (UNIRRADIATED) (a)

Beltline Material Cu Weight %

Ni Weight %

IRTNDT (OF)

Closurc Head Flange 60 Vessel Flange 60 Intermediate Shell Plate B6903-2 0.13 0.60 0

Intermcdiate Shell Plate B6903-3 0.12 0.56 10 Lower Shell Plate B6919-1 0.14 0.55 15 Lower Shell Plate B6919-2 0.14 0.56 5

Intermediate Shell Longitudinal Weld Seams19-894 A & B lb) 0.258 0.165

-56 (Heat # 33A277)

Surveillance Wcld(c) 0.14 0.19 Circumferential Weld Scam 11-894 lb)

(Heat # 6329637) 0.205 0.105

-56 Lower Shell Longitudinal Weld Scams20-894 A & B lb, 0.197 0.060

-56 (Heat # 90099)

NOTES:

(a) WCAP-14689, Revision 411 (b) Best-estimate copper and nickel from CE NPSD-1039191 (c) The surveillance weld is representativc of intcrmediatc shell longitudinal.velds19-894 A & B. Best-estimate copper and nickel values represent a single chemical analysis documented in WCAP-88 1015]

Table 5-4 REACTOR VESSEL FLUENCE PROJECTIONS FOR 36 EFPY (a. bi EFPY 00 15O 30° 450 36 4.14 2.41 1.73 1.24 NOTES:

(a) From Tabic 6-2 of WCAP-16221 -NP Rcv. 0 [1 (b) Fluencc in 1019 n'/cm 2 (E > 1.0 McV)

FARLEY UNIT I 15 REVISION 2

PRESSURE TEMPERATURE LIMITS REPORT Table 5-5

SUMMARY

OF ADJUSTED REFERENCE TEMPERATURES (ARTs) FOR REACTOR VESSEL BELTLINE MATERIALS AT THE 1/4-T AND 3/4-T LOCATIONS FOR 29.0 EFPY (a b)

Material 1/4-T 3/4-T (0F)

(OF)

Intermediate Shell Plate B6903-2 143 120 Intermediate Shell Plate B6903-3 143 121 Lower Shell Plate B6919-1 166 141 Lower Shell Plate B6919-1 Using S/C Data 170 (

144(c)

Lower Shell Plate B6919-2 157 131 Intermediate Shell Longitudinal Weld Id) 88 (d)

Seams19-894 A & B 120 (Heat # 33A277)

Intermediate Shell Longitudinal WeId Seams19-894 A & B 9

d) 6d(

(Heat # 33A277) 97 65 Using S/C Data Circumferential Weld 11-894 (Heat # 6329637) 128 103 Lower Shell Longitudinal Weld Seams 20-9 0 (d) 67 d) 894 A & B 9

(Heat # 90099)

NOTES:

(a) From Westinghouse letter ALA-04-170t 11 1.

(b) The ARTs presented here are based on the peak reactor vessel surface fluence of 3.34 x 10 19 n/cm2 (E >

1.0 MeV) from WCAP-1622 1-NP, Revision 01'01, Table 6-2 unless otherwise noted.

(c) Limiting 1/4T and 3/4T ART values. The P/T limit curves are based on ART values of 1700F and 1450F, originally documented in WCAP-14689, Revision 4 1'] for 31.9 EFPY using a peak fluence of 3.868 x 1 0'9 n/cm2 (E > 1.0 MeV) and a chemistry factor (CF) of 97.80F (per Reg. Guide 1.99 Rev. 2, Position 1.1). However, subsequent to the analysis of Capsule V (ref. WCAP-16221-NP, Revision 0 llo) the CF using Reg. Guide 1.99 Rev. 2, Position 2.1 increased from 93.30F to 1 00.90F, making the Position 2.1 ART higher than the Position 1.1 ART used by WCAP-14689. Thus, in order to maintain the I/4T ART value at 1701F (the value used in generating the P/T limit curves, Figs. 2-1 and 2-2), the service period of the P/T curves was reduced to 29.0 EFPY. This reduction was calculated using the updated fluence value cited in (b) above. Since maintaining a 3/4T ART of 1450F would cause a lesser reduction in the service period (to 30 EFPY), the I/4T location is more limiting. Therefore, while the 1450F 3/4T ART value used in generation of the P/T limit curves is still cited in Figs. 2-1 and 2-2, the updated 3/4T ART value of 1441F at 29.0 EFPY is tabulated.

(d) ARTs calculated using the peak vessel fluence of 1.01 x IO"9 n/cm2 (E > 1.0 MeV) at 450, interpolated from WCAP-16221-NP, Revision 0110], Table 6-2.

FARLEY UNIT I 16 REVISION 2

PRESSURE TEMPERATURE LIMITS REPORT Table 5-6 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE AT 29.0 EFPY FOR THE LIMITING REACTOR VESSEL MATERIAL - LOWER SHELL PLATE B6919-1 (a)

Parameter Value Operating Period 29.0 EFPY Location I/4-T 314-T Chemistry Factor, CF (°F) 100.9 100.9 Fluencc, f (I ol9 n/cm2) (b) 2.080 0.809 Fluence Factor, FF 1.199 0.941 ARTNDT = CF x FF (°F) 121.0 94.9 Initial RTNDT, I (°F) 15 15 Margin, M (°F) 34 34 Adjusted Reference Temperature (ART), (°F) per 170'c 144(c)

Regulatory Guide 1.99, Revision 2 NOTES:

(a) From Wcstinghous letter ALA-04-170111(.

(b) Fluence is based on fsurI (1019 n/cm2, E> I.0 McV) =3.34, from WCAP-16221 -NP, Revision oll, Table 6-2. The Farley Unit I reactor vessel wall thickness is 7.875 inches in the beltlinc region.

(c) Refer to note (c) of Table 5-5.

FARLEY UNIT I 17 REVISION 2

PRESSURE TEMPERATURE LIMITS REPORT Table 5-7 PRESSURIZED THERMAL SHOCK (RTprs) VALUES FOR 36 EFPY '3' Surface Fluence ARTNDT I

M RTp-s MaeralCF (10'9 n/cm2, j FF (CF x FF)

Material l E> 1.0 MeV) L F

(FF)

(0F)

(OF)

(OF)

Intermediate Shell Plate B6903-2 91.0 4.14 1.36 123.8 0

34 158 Intermediate Shell Plate B6903-3 82.2 4.14 1.36 111.8 10 34 156 Lower Shell Plate B6919-1 97.8 4.14 1.36 133.0 15 34 182 Lower Shell Plate 100.9 4.14 1.36 137.2 15 34 18 6(b B6919-1 Using S/C Data Lower Shell Plate B6919-2 98.2 4.14 1.36 133.6 5

34 173 Intermediate Shell Longitudinal Welds19-894 A & B 126.3 1.24 1.06 133.9

-56 66 144 (Heat # 33A277)

Intermediate Shell 124.6 1.24 1.06 132.1

-56 44 120 Longitudinal Welds19-894 A & B (Heat # 33A277)

Using S/C Data Circumferential Weld 98.4 4.14 1.36 133.8

-56 66 144 11-894 (Heat # 6329637)

Lower Shell Longitudinal Welds20-894 A & B 91.4 1.24 1.06 96.9

-56 66 107 (Heat # 90099)

NOTES:

(a) From Westinghouse letter ALA-04-170t111.

(b) This RTpT5 value was calculated using the CF from the surveillance data and a full qa margin of 1 7°F, since this surveillance data is not credible.

FARLEY UNIT I 18 REVISION 2

PRESSURE TEMPERATURE LIMITS REPORT 6.0 References

1. WCAP-14689, Revision 4, Farley Units I and 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation, E. Terek, April 1998.

2. WCAP-14196, Analysis of Capsule W from the Alabama Power Company Farley Unit I Reactor Vessel Radiation Surveillance Program, P. A. Peters, et al., February 1995.

3. WCAP-14687, Joseph M. Farley Units I and 2 Radiation Analysis and Neutron Dosimetry Evaluation, R. L. Bencini, June 1996.

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

5. WCAP-88 10, Southern Alabama Power Company Joseph M. Farley Nuclear Plant Unit No. I Reactor Vessel Radiation Surveillance Program, J. A. Davidson, et al., December 1976

6. WCAP-9717, Analysis of Capsule Y from the Alabama Power Company Farley Unit No. I Reactor Vessel Radiation Surveillance Program, S. E. Yanichko, et al., June 1980.

7. WCAP-10474, Analysis of Capsule U from the Alabama Power Company Joseph M. Farley Unit I Reactor Vessel Radiation Surveillance Program, R. S. Boggs, et al., February 1984.

8. WCAP-12471, Analysis of Capsule X from the Alabama Power Company Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program, E. Terek, et al., December 1989.

9. CE NPSD-1039, Revision 2, Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds, Combustion Engineering Owners Group, June 1997.

10. WCAP-16221 -NP, Revision 0, Analysis of Capsule V from the Southern Nuclear Operating Company, Joseph M. Farley Unit I Reactor Vessel Radiation Surveillance Program, K.G.

Knight, et. al., March 2004.

11. Westinghouse letter ALA-04-170, Review of Revision 2 Draft of Pressure-Temperature Limits Report, E.C. Arnold to L. M. Stinson, October 25, 2004.

FARLEY UNIT I 19 REVISION 2