NL-08-1385, Submittal of Revision 4 of the Pressure Temperature Limits Report

Submittal of Revision 4 of the Pressure Temperature Limits Report

ML082670287
Person / Time
Site:	Farley
Issue date:	09/22/2008
From:	Ajluni M Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-08-1385
Download: ML082670287 (28)

Text

Southern Nuclear Operating Company, Inc.

Post Office Box 1295 Birmingham. Alabama 35201-1295 Tel 205.9925000 SOUTHERN'\

COMPANY September, 22, 2008 Energy to Serve Your WorldS" Docket No.: 50-364 NL-08-1385 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant - Unit 2 Submittal of Revision 4 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 2 Technical Specifications, Southern Nuclear Operating Company (SNC) hereby submits Revision 4 of the FNP Unit 2 Pressure Temperature Limits Report (PTLR).

Revision 4 of the PTLR updates the reactor vessel radiation surveillance program capsule withdrawal schedule, the data credibility analysis and the supplemental data sections to reflect the surveillance capsule analysis report (WCAP-16918 NP Rev. 1, submitted April 18, 2008) for Capsule V, the sixth and final capsule.

In addition, as a result of the Capsule V analysis the service period for the existing pressure temperature limit curves in the PTLR is reduced and the curves are re-Iabeled from to 33.8 to 32.8 EFPY.

An ex-vessel neutron dosimetry system has been installed to enable long term monitoring of the reactor vessel following withdrawal of the last capsule.

This letter contai-ns no NRC commitments. If you have any questions, please advise.

Sincerely,

~~~

M. J. Ajluni Manager, Nuclear Licensing MJNDWD/phr

Enclosures:

Joseph M. Farley Nuclear Plant Pressure Temperature Limits Report Unit 2, Revision 4, September 2008

U. S. Nuclear Regulatory Commission NL-08-1385 Page 2 cc: Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Mr. J. R. Johnson, Vice President - Farley Mr. D. H. Jones, Vice President - Engineering RTYPE: CFA04.054; LC# 14838 U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. K. D. Feintuch, NRR Project Manager - Farley Mr. E. L. Crowe, Senior Resident Inspector - Farley

Joseph M. Farley Nuclear Plant - Unit 2 Enclosure 1 Plant Pressure Temperature Limits Report Unit 2, Revision 4, September 2008

SOUTHERN COMPANY A

Energy 10 Serve JO"r WorM*

Joseph M. Farley Nuclear Plant Pressure Temperature Limits Report Unit 2 Revision 4 September 2008

PTLR for FNP Unit 2 Revision 4 Page 1 of 24 Table of Contents List of Tables 2 List of Figures , 3 1.0 RCS Pressure Temperature Limits Report (PTLR) 4 2.0 Operating Limits 5 2.1 RCS PressurelTemperature (PIT) Limits (LCO - 3.4.3) 5 2.2 RCP Operation Limits 5 2.3 LTOP Arming Temperature (LCO - 3.4.12) 5 3.0 Reactor Vessel Material Surveillance Program 10 4.0 Reactor Vessel Surveillance Data Credibility 11 5.0 Supplemental Data Tables 17 6.0 References 24

PTLR for FNP Unit 2 Revision 4 Page 2 of 24 List of Tables 2-1 Farley Unit 232.8 EFPY Heatup Curve Data Points 8 2-2 Farley Unit 2 32.8 EFPY Cooldown Curve Data Points 9 3-1 Surveillance Capsule Withdrawal Schedule 10 4-1 Surveillance Capsule Data Calculation of Best-Fit Line as Described in Position 2.1 of Regulatory Guide 1.99, Revision 2 14 4-2 Scatter of L1RTNOT Values About a Best-Fit Line for Surveillance Plate Material 15 4-3 Scatter of L1RTNOT Values About a Best-Fit Line for Surveillance Weld Material 15 5-1 Comparison of Surveillance Material 30 Ft-Lb Transition Temperature Shifts and Upper Shelf Energy Decrease with Regulatory Guide 1.99, Revision 2, Predictions 18 5-2 Calculation of Chemistry Factors Using Surveillance Capsule Data 19 5-3 Reactor Vessel Toughness Table (Unirradiated) 20 5-4 Reactor Vessel Fluence Projections for 36 EFPY 21 5-5 Summary of Adjusted Reference Temperatures (ARTs) for Reactor Vessel Beltline Materials at the 1/4-T and 3/4-T Locations for 32.8 EFPY 21 5-6 Calculation of Adjusted Reference Temperature at 32.8 EFPY for the Limiting Reactor Vessel Material... 22 5-7 Pressurized Thermal Shock (RTPTS) Values for 36 EFPY 23

PTLR for FNP Unit 2 Revision 4 Page 3 of 24 List of Figures 2-1 Farley Unit 2 Reactor Coolant System Heatup Limitations 6 2-2 Farley Unit 2 Reactor Coolant System Cooldown Limitations 7

PTLR for FNP Unit 2 Revision 4 Page 4 of 24 This page intentionally blank.

PTLR for FNP Unit 2 Revision 4 Page 5 of 24 1.0 ReS Pressure Temperature Limits Report (PTLR)

This PTLR for Farley Nuclear Plant - Unit 2 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 PressurelTemperature (PIT) 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 NRC-approved 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 PIT limits for flow losses associated with the RCPs.

2.1 RCS PressurelTemperature (PIT) Limits (LCO - 3.4.3) 2.1.1 The minimum boltup temperature is 75°F.

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

a. A maximum heatup of 100°F in anyone hour period.

b. A maximum cooldown of 100°F in anyone hour period.

c. A maximum temperature change of less than or equal to 10°F in anyone hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

2.1.3 The RCS PIT 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 110°F 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.

2.3 LTOP Arming Temperature (LCO - 3.4.12) 2.3.1 The LTOP system arming temperature is 325°F.

PTLR for FNP Unit 2 Revision 4 Page 6 of 24 Figure 2*1 Farley Unit 2 Reactor Coolant System Heatup Limitations(a)

(Heatup Rates up to 100°F/hr) Applicable to 32.8 EFPY (adjusted to include 60 psi ~P at RCS temperatures ~ 110°F and 27 psi ~P at RCS temperatures < 110°F). Includes vessel flange requirements of 180°F and 561 psig per 10 CFR 50, Appendix G. [1]

2,500 I I Leak Test Limit ~ I" il J I 2,250 f----

I -L I\~I ~H Criticality Limit for 60 F/hr Heatup limiting Malerial - 1/4T at 32.8 EFPY t--

I I ___I Intenmediate Shell Plate B7212-1 ART=186 F Criticality Limit for i--

Limiting Material - 3/4T at 32.8 EFPY:

Intenmediate Shell Plate B7203-1 I II 100 F/hr Heatup I

2,000 I-ART=149 F I 1,750

/"

I I II II I L t 1/'I AI Ci 1,500 I II Vi

-Q.

l!

J I

Unacceptable Operation I

I I 1/ f/ //

I Acceptable Operation -

1/1 Ul 1,250 l! II I I Q.

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s 1,000

/

/

/

IJ II

.5 I V II I 750

/ I IJ / I m J V Heatup Rate J /

-- (degree F/hr)

I\,/ V I V I 500 nO - h ./ ./ ,

00 rr I 250 IV I

II I" ' I Cnticality Limit Based on Inservice "I Hydrostatic Test Temperature (314 F)

I (or the Service Period up to 32.8 EFPY ~

Min. RCS Boltup Temperalure ~ 75 F 1 I 1

I o

o 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Degree F)

NOTE: a) From Fig. 12, WCAP-14689, Rev. 4[1] with service period adjusted to 32.8 EFPY per WCAP-16918-NP. Rev. 1[15J

PTLR for FNP Unit 2 Revision 4 Page 7 of 24 Figure 2-2 Farley Unit 2 Reactor Coolant System Cooldown Limitations(a)

(Cooldown Rates up to 1OO°F/hr) Applicable to 32.8 EFPY (adjusted to include 60 psi LlP at RCS temperatures ~ 110°F and 27 psi ~P at RCS temperatures < 110°F). Includes vessel flange requirements of 180°F and 561 psig per 10 CFR 50, Appendix G. [11 2,500 I I T

L t II Limiting Material - 1/4T at 32.8 EFPY:

Intermediate Shell Plate 87212-1 J I I 2,250 i- ART = 186 F I

Limiting Material - 3/4T at 32.8 EFPY: I 2,000 i Intermediate Shell Plate 87203-1 ART = 149 F I I I Ij1 II I

~ I t l- t I L ~ I I , I 1,750 i I I t II J i ~ I II

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1,250 1,000 f--

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Unacceptable Operation I

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Acceptable Operation-

, I j

sc: I I I

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I j Cooldown Rate (degree F / hr) 0 1\-

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I- - Min. RCS Boltup Temperature - 75 F o I I I ! I I I i I

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o 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Degree F)

NOTE: a) From Fig. 13, WCAP-14689, Rev. 4[1] with service period adjusted to 32.8 EFPY per WCAP-16918-NP, Rev. 1[15]

PTLR for FNP Unit 2 Revision 4 Page 8 of 24 Table 2-1 Farley Unit 2 - 32.8 EFPY Heatup Curve Data Points(a)

(adjusted to include 60 psi ~P at RCS temperatures ~ 110°F and 27 psi ~P at RCS temperatures < 1100F) [1]

60 of 60 of Criticality Limit 100 OF 100 OF Criticalitv Limit Leak Test T I p T I p T I p T I p T I P 75 470 314 0 75 435 314 0 292 2000 80 470 314 465 80 435 314 467 314 2485 85 470 314 454 85 435 314 451 90 470 314 446 90 435 314 438 95 470 314 441 95 435 314 428 100 470 314 438 100 435 314 419 105 470 314 437 105 435 314 413 110 471 314 438 110 435 314 408 110 438 314 441 110 402 314 405 115 441 314 444 115 402 314 403 120 444 314 449 120 402 314 402 125 449 314 455 125 402 314 403 130 455 314 462 130 403 314 404 135 462 314 469 135 404 314 407 140 469 314 478 140 407 314 411 145 478 314 488 145 411 314 416 150 488 314 498 150 416 314 422 155 498 314 510 155 422 314 428 160 510 314 522 160 428 314 436 165 522 314 536 165 436 314 445 170 536 314 551 170 445 314 455 175 551 314 567 175 455 314 466 180 561 314 584 180 466 314 478 180 567 314 602 185 478 314 491 185 584 314 622 190 491 314 505 190 602 314 644 195 505 314 521 195 622 314 667 200 521 314 538 200 644 314 692 205 538 314 556 205 667 314 719 210 556 314 576 210 692 314 747 215 576 314 598 215 719 314 778 220 598 314 621 220 747 314 811 225 621 314 646 225 778 314 847 230 646 314 673 230 811 314 885 235 673 314 702 235 847 314 926 240 702 314 733 240 885 314 970 245 733 314 767 245 926 314 1018 250 767 314 803 250 970 314 1069 255 803 314 842 255 1018 314 1119 260 842 314 883 260 1069 314 1171 265 883 314 928 265 1119 315 1223 270 928 315 976 270 1171 320 1273 275 976 320 1028 275 1223 325 1326 280 1028 325 1083 280 1273 330 1383 285 1083 330 1143 285 1326 335 1445 290 1143 335 1206 290 1383 340 1510 295 1206 340 1275 295 1445 345 1580 300 1275 345 1348 300 1510 350 1656 305 1348 350 1426 305 1580 355 1736 310 1426 355 1510 310 1656 360 1822 315 1510 360 1599 315 1736 365 1914 320 1599 365 1695 320 1822 370 2013 325 1695 370 1798 325 1914 375 2118 330 1798 375 1908 330 2013 380 2231 335 1908 380 2025 335 2118 385 2351 340 2025 385 2150 340 2231 345 2150 390 2283 345 2351 350 2283 395 2425 355 2425 NOTE: a) From Table 28, WCAP-14689, Rev. 4[11 with service period adjusted to 32.8 EFPY per WCAP-16918-NP, Rev. 1[15]

PTLR for FNP Unit 2 Revision 4 Page 9 of 24 Table 2-2 Farley Unit 2 - 32.8 EFPY Cooldown Curve Data Points(a)

(adjusted to include 60 psi ~P at RCS temperatures ~ 110°F and 27 psi ~P at RCS temperatures < 1100F) [1]

oof 20°F 40 of 60 of 100 of T I p T I p T I p T I p T I P 75 499 75 461 75 423 75 384 75 303 80 503 80 465 80 426 80 387 80 306 85 506 85 469 85 430 85 391 85 310 90 510 90 472 90 434 90 395 90 315 95 514 95 477 95 438 95 400 95 319 100 519 100 481 100 443 100 404 100 325 105 523 105 486 105 448 105 410 105 330 110 529 110 492 110 454 110 415 110 336 110 496 110 459 110 421 110 382 110 303 115 501 115 464 115 427 115 389 115 310 120 507 120 470 120 433 120 395 120 317 125 514 125 477 125 440 125 402 125 325 130 521 130 484 130 448 130 410 130 334 135 528 135 492 135 456 135 419 135 343 140 536 140 500 140 464 140 428 140 353 145 545 145 509 145 474 145 438 145 364 150 554 150 519 150 484 150 448 150 376 155 561 155 529 155 495 155 460 155 389 160 561 160 541 160 507 160 472 160 403 165 561 165 553 165 519 165 486 165 418 170 561 170 561 170 533 170 500 170 434 175 561 175 561 175 548 175 516 175 452 180 561 180 561 180 561 180 533 180 471 180 626 180 595 180 564 185 551 185 491 185 641 185 611 185 581 190 570 190 513 190 658 190 628 190 599 195 591 195 537 195 675 195 647 195 619 200 614 200 563 200 694 200 667 200 640 205 638 205 591 205 715 205 689 205 663 210 665 210 621 210 737 210 712 210 688 215 693 215 653 215 760 215 737 215 715 220 723 220 688 220 786 220 764 220 743 225 756 225 725 225 813 225 793 225 774 230 792 230 766 230 842 230 824 230 807 235 830 235 809 235 874 235 858 235 843 240 871 240 856 240 908 240 894 240 881 245 915 245 907 245 944 245 932 245 923 250 962 250 961 250 983 250 974 250 967 255 1013 255 1019 255 1025 255 1019 255 1015 260 1068 260 1070 260 1067 260 1066 265 1119 265 1118 270 1171 275 1226 280 1286 285 1351 290 1420 295 1494 300 1573 305 1658 310 1749 315 1846 320 1951 325 2062 330 2182 335 2309 NOTE: a) From Table 29, WCAP-14689, Rev. 4[11 with service period adjusted to 32.8 EFPY per WCAP-16918-NP, Rev. 1[15]

PTLR for FNP Unit 2 Revision 4 Page 10 of 24 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. Consistent with specific requirements for Farley Unit 2 associated with the granting of an exemption to Appendix H of 10 CFR 50 documented in NUREG-0117[4I, Figures 2-1 and 2-2 are based on the greater, or limiting value, of the following: (1) the actual shift in reference temperature for plate 87212-1 as determined by impact testing, or (2) the predicted shift in reference temperature for weld seam 11-923 as determined by Regulatory Guide 1.99, Revision 2. The neutron transport and dosimetry evaluation methodologies used follow the guidance and meet the requirements of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence,,[141. Results from the reactor vessel surveillance program will 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 PIT limits shown in Figures 2-1 and 2-2 for the specified f1uence period.

Table 3-1 SURVEilLANCE CAPSULE WITHDRAWAL SCHEDULE (a) (9)

Capsule (e) Capsule location lead Removal Fluence (Degree) Factor EFPy(b) (n/cm 2 )

U 343 3.26 1.11 6.05 x 10 18 W 110 2.84 3.96 1.73x10 19 X 287 3.38 6.43 2.98 x 10 19 Z 340 2.98 13.85 4.92 x 1019 (d)

Y 290 3.12 19.01 6.79 x 1019 (e)

V 107 3.58 21.82 8.73 x 1019 (I)

Notes:

a) Data from Table 7-1, WCAP-16918-NP, Revision 1 [15]

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 fluence 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 fluence 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) NRC approval is required prior to changing the capsule withdrawal schedule.

Reference:

NRC Administrative letter 97-04. The schedule has been completed as submitted by SNC letter Nl-04-0372, March 5. 2004 [12J and approved by NRC letter dated March 15, 2004 [13J*

PTLR for FNP Unit 2 Revision 4 Page 11 of 24 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 Iight-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.

All six surveillance capsules have been removed from the Farley Unit 2 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 2 reactor vessel surveillance data and determine if the Farley Unit 2 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 2 reactor vessel consists of the following beltline region materials:

• Intermediate shell plates B7203-1 and B7212-1;

• Lower shell plates B7210-1 and B721 0-2;

• Intermediate shell longitudinal weld seams19-923 A, heat number HODA;

• Intermediate shell longitudinal weld seams19-923 B, heat number BOLA;

• Lower shell longitudinal weld seams20-923 A & B, heat number 83640, Linde 0091 flux, flux lot 3490; and

• Circumferential weld 11-923, heat number 5P5622, Linde 0091 flux, flux lot 1122.

PTLR for FNP Unit 2 Revision 4 Page 12 of 24 Per WCAP-8956[51, the Farley Unit 2 surveillance program was based on ASTM E185-73, Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels. Per Section 4.1 of ASTM E185-73, 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 f1uence.

At the time the Farley Unit 2 surveillance capsule program was developed, intermediate shell plate B7212-1 was judged to be most limiting and was therefore utilized in the surveillance program.

The surveillance program weld for Farley Unit 2 was fabricated using the shielded metal arc welding process and E8018 stick electrodes, in a manner similar to that used to fabricate intermediate shell longitudinal seams19-923 A (heat HODA) and B (heat BOLA). These electrodes were not copper-coated and do not exhibit the chemical variability found in copper-coated submerged arc weld wire. Although the surveillance weld material does not represent the limiting reactor vessel beltline weld, 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. The NRC explicitly approved the selection of the Farley Unit 2 surveillance weld material on the basis that the limiting beltline material (i.e., intermediate plate B7212-1) was included in the surveillance program and conservative methods of analysis contained in Regulatory Guide 1.99 were available to predict the radiation characteristics of the limiting beltline weld. The NRC incorporated an exemption to the requirements of Appendix H to 10 CFR Part 50 in the Farley Unit 2 Operating License, thereby approving the selected surveillance weld material based on the NRC evaluation provided in Section 5.2.1 of NUREG-0117. [41 Although the Farley Unit 2 surveillance weld material does not meet the requirements of Criterion 1, conservative methods of analysis are available to predict the radiation characteristics of the limiting beltline weld. The limiting beltline plate material is intermediate plate B7212-1 which is more limiting than any of the reactor vessel beltline welds and is included in the reactor vessel material surveillance program. Therefore, the Farley Unit 2 reactor vessel material surveillance program provides assurance that the radiation damage to the vessel can be adequately determined and the integrity of the Farley Unit 2 reactor vessel will be ensured during normal plant operations and anticipated operational occurrences. Therefore, the Farley Unit 2 reactor vessel surveillance program meets the intent 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 ft-Ib temperature and upper shelf energy, unambiguously.

PTLR for FNP Unit 2 Revision 4 Page 13 of 24 Plots of Charpy energy versus temperature for the unirradiated condition are presented in WCAP-8956 [51, Alabama Power Company Joseph M. Farley Nuclear Plant Unit No.2 Reactor Vessel Radiation Surveillance Program, dated AUgust 1977.

Plots of Charpy energy versus temperature for the irradiated conditions are presented in the reactor vessel surveillance capsule reports for Capsules U [61, W [7l, X [21, Z [101, Y [11 1, and V [151.

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

Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of ~RTNOT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°F for welds and 17°F for base metal. Even if the f1uence 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 of Regulatory Guide 1.99, Revision 2.1, was used to calculate a best-fit line for the base metal and weld data to determine if this criterion is met. Per Regulatory Position 2, the scatter of ~RTNOT values about a best-fit line should be less than 1a (17°F for base metal and 28°F for welds).

Plate (Base Metal) Evaluation:

The scatter values obtained for the base metal (intermediate shell plate B7212-1) are shown in Table 4-1, which indicates that two of the measured ~RTNOT values are slightly lower than 1a below the predicted value, while one value slightly exceeds the upper 1a bound. The data show that all measured ~RTNOT values are well below the upper 1a bound (i.e. the predicted value conservatively exceeds the measured value) except one (Capsule Y, transverse specimen).

That three measured ~RTNOT values fall slightly outside the +/-1 a bounds can be attributed to several factors, such as 1) the inherent uncertainty in the Charpy test data, 2) the use of a symmetric hyperbolic tangent Charpy curve fitting program vs.

an asymmetric hyperbolic tangent Charpy curve fitting program or a hand-drawn curve using engineering judgement, and 3) rounding errors.

Of the 12 data points, scatter is within 17°F of the best-fit line for all but 3.

Statistically, since +/-1 a (17°F) is expected to encompass approximately 68% of the data, for a set of 12 scatter values one could expect 3 to be outside the +/- 1a bounds, as observed, hence the data is statistically credible, and based on the arguments above, the plate data meets the intent of Criterion 3.

PTLR for FNP Unit 2 Revision 4 Page 14 of 24 Table 4*1 Surveillance Capsule Data Calculation of Best-Fit Line as Described in Position 2.1 of Regulatory Guide 1.99, Revision 2 (al FF (el FF X 2 F (b) ~RTNoT FF Material Capsule ~RTNOT 2 (x) (y) (x )

(xv)

U 0.605 0.859 105.5 90.7 0.738 Intermediate Shell Plate 67212-1 W 1.73 1.151 167.7 193.0 1.324 (Longitudinal) X 2.98 1.289 164.8 212.4 1.662 Z 4.92 1.399 200.1 280.0 1.958 Y 6.79 1.458 214.2 312.3 2.125 V 8.73 1.496 218.3 326.6 2.238 Intermediate Shell U 0.605 0.859 124.0 106.5 0.738 Plate 67212-1 W 1.73 1.151 168.5 193.9 1.324 (Transverse)

X 2.98 1.289 200.1 258.0 1.662 Z 4.92 1.399 195.8 274.0 1.958 Y 6.79 1.458 231.0 336.8 2.125 V 8.73 1.496 215.3 322.1 2.238 t

=1 2906.13 20.091 CF =~(FF * ~RTNoT) -;- ~(FF2) =144.6 of Weld Metal U 0.605 0.859 -28.4 -24.4 0.738 W 1.73 1.151 7.0 8.1 1.324 X 2.98 1.289 -15.6 -20.1 1.662 Z 4.92 1.399 10.2 14.3 1.958 Y 6.79 1.458 69.1 100.7 2.125 V 8.73 1.496 56.5 84.5 2.238 t

.. 1 163.08 10.046 CF =~(FF * ~RTNOT) -;- ~(FF2) =16.2 of NOTES

(a) Data from Table 0-1, WCAP-16918-NP, Revision 1[15]

(b) F Fluence (10 19 n/cm2, E> 1.0 MeV)

=

(c) FF = Fluence Factor = F (0.28 -0.1 log f)

PTLR for FNP Unit 2 Revision 4 Page 15 of 24 Table 4*2 Scatter of ilRTNOT Values about a Best-Fit Line for Surveillance Plate Material(a)

Intermediate Shell CF !1RTNOT Best Fit Scatter Plate B7212-1 Capsule (Best Fit FF (30 ft-Ib) !1RTNOT of <17 Specimen Slope) (OF) (OF) !1RTNOT OF Orientation (OF)

U 144.6 0.8593 105.5 124.3 18.8 No W 144.6 1.1508 167.7 166.4 -1.3 Yes Longitudinal X 144.6 1.2891 164.8 186.4 21.6 No Z 144.6 1.3992 200.1 202.3 2.2 Yes Y 144.6 1.4579 214.2 210.8 -3.4 Yes V 144.6 1.4960 218.3 216.3 -2.0 Yes U 144.6 0.8593 124.0 124.3 0.3 Yes W 144.6 1.1508 168.5 166.4 -2.1 Yes Transverse X 144.6 1.2891 200.1 186.4 -13.7 Yes Z 144.6 1.3992 195.8 202.3 6.5 Yes Y 144.6 1.4579 231.0 210.8 -20.2 No V 144.6 1.4960 215.3 216.3 1.0 Yes NOTES:

(a) Data from Table 0-2, WCAP-16918-NP, Revision 1 [15]

Table 4*3 Scatter of ilRT NOT Values about a Best-Fit Line for Surveillance Weld Material(a)

CF !1RTNOT Best Fit Scatter Surveillance Capsule (Best Fit FF (30 ft-Ib) !1RTNOT of <28 Material Slope) (OF) (OF) !1RTNOT of (OF)

U 16.2 0.8593 -28.4 13.9 42.3 No W 16.2 1.1508 7.0 18.6 11.6 Yes Weld Metal X 16.2 1.2891 -15.6 20.9 36.5 No Z 16.2 1.3992 10.2 22.7 12.5 Yes Y 16.2 1.4579 69.1 23.6 -45.5 No V 16.2 1.4960 56.5 24.2 32.3 No NOTES:

(a) Data from Table 0-2, WCAP-16918-NP, Revision 1 [15]

PTLR for FNP Unit 2 Revision 4 Page 16 of 24 Weld Evaluation:

The scatter values obtained for the weld metal are shown in Table 4-2, which indicates that 4 of the 6 measured ~RTNOT values exceed the +/-1 cr scatter band of 28°F. Therefore, the surveillance weld data are deemed not credible per Criterion 3.

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 +/-25°F.

The Farley Unit 2 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 beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions and will not differ by more than 25°F. Therefore, the Farley reactor vessel 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 2 surveillance program does not include correlation monitor material.

Therefore, this criterion is not applicable to Farley Unit 2.

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 2 surveillance data are credible for the plate (base metal) material but not credible for the weld metal; hence, surveillance capsule data are not used for the weld metal reference temperature values.

PTLR for FNP Unit 2 Revision 4 Page 17 of 24 5.0 Supplemental Data Tables Table 5-1 contains a comparison of measured surveillance material 30 ft-Ib 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 2 reactor vessel toughness data.

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

Table 5-5 provides a summary of the adjusted reference temperatures (ARTs) of the Farley Unit 2 reactor vessel beltline materials at the 1/4-T and 3/4-T locations for 32.8 EFPY.

Table 5-6 shows the calculation of the ART at 32.8 EFPY for the limiting Farley Unit 2 reactor vessel material.

Table 5-7 provides RTpTs values for Farley Unit 2 for 36 EFPY.

PTLR for FNP Unit 2 Revision 4 Page 18 of 24 Table 5*1 Comparison of Surveillance Material 30 ft-Ib Transition Temperature Shift and Upper Shelf Energy Decrease with Regulatory Guide 1.99, Revision 2, Predictions (a) 30 ft-Ib Transition Upper Shelf Energy Temperature Shift Decrease Material Capsule Fluence(e) Predicted Measured Predicted Measured (x 1019 (OF) (b) (OF) (c) (%) (b) (%)(d) 2 n/cm , E>

1.0 MeV)

Intermediate Shell U 0.605 128.0 105.5 26 27 Plate B7212-1 W 1.73 171.5 167.7 33 22 (Longitudinal) X 2.98 192.1 164.8 37 26 Z 4.92 208.5 200.1 42 28 y 6.79 217.2 214.2 45 36 V 8.73 222.9 218.3 48 34 Intermediate Shell U 0.605 128.0 124.0 26 27 Plate B7212-1 W 1.73 171.5 168.5 33 21 (Transverse) X 2.98 192.1 200.1 37 28 Z 4.92 208.5 195.8 42 29 y 6.79 217.2 231.0 45 42 V 8.73 222.9 215.3 48 27 Surveillance U 0.605 32.8 -28.4 17 8 Program W 1.73 44.0 7.0 22 0 Weld Metal X 2.98 49.2 -15.6 24 0 Z 4.92 53.4 10.2 27 8 y 6.79 55.7 69.1 30 5 V 8.73 57.1 56.5 32 14 Heat Affected Zone U 0.605 - - 219.8 -- 30 Material W 1.73 --- 268.8 - - 20 X 2.98 - - 230.5 -- 19 Z 4.92 - - 263.8 -- 20 Y 6.79 - - 269.6 -- 35 V 8.73 - - 322.4 -- 25 NOTES:

(a) Data from Table 5-10, WCAP-16918-NP, Revision 1 [15]

(b) Based on Reg. Guide 1.99, Rev. 2 methodology using the mean weight percent values of copper and nickel of the surveillance material.

(c) Calculated using measured Charpy data plotted using CVGRAPH, Version 5,3.

(d) Values are based on the definition of upper shelf energy given in ASTM E185-82.

(e) The fluence values presented here are the calculated values, not the best estimate values.

PTLR for FNP Unit 2 Revision 4 Page 19 of 24 Table 5*2 Calculation of Chemistry Factors Using Surveillance Capsule Data la]

FF (e)

FF X 2 F (b) ARTNDT FF Material Capsule ARTNDT (X) (y) (~)

(Xy)

U 0.605 0.859 105.5 90.7 0.738 Intermediate Shell Plate 87212-1 W 1.73 1.151 167.7 193.0 1.324 (Longitudinal) X 2.98 1.289 164.8 212.4 1.662 Z 4.92 1.399 200.1 280.0 1.958 Y 6.79 1.458 214.2 312.3 2.125 V 8.73 1.496 218.3 326.6 2.238 Intermediate Shell U 0.605 0.859 124.0 106.5 0.738 Plate 87212-1 W 1.73 1.151 168.5 193.9 1.324 (Transverse)

X 2.98 1.289 200.1 258.0 1.662 Z 4.92 1.399 195.8 274.0 1.958 y 6.79 1.458 231.0 336.8 2.125 V 8.73 1.496 215.3 322.1 2.238 SUM: 2906.13 20.091 CF =~(FF'" ARTNDT) -;- ~(FF2) =144.6 of Weld Metal U 0.605 0.859 -27.3 (-28.4id) -24.4 0.738 W 1.73 1.151 6.7 (7.0)(d) 8.1 1.324 X 2.98 1.289 -15.0 (-15.6) (d) -20.1 1.662 Z 4.92 1.399 9.8 (10.2) (d) 14.3 1.958 Y 6.79 1.458 66.3 (69.1) (d) 100.7 2.125 V 8.73 1.496 54.2 (56.5) (d) 84.5 2.238 SUM: 163.08 10.046 CF =~(FF'" ARTNDT) -;- ~(FF2) =16.2 of NOTES: (a) Data from Table 0-1, WCAP-16918-NP, Revision 1 [15]

(b) F =Fluence (1019 nfcm 2 , E> 1.0 MeV)

(c) FF = Fluence Factor = F (0.28 - 0.1 log f)

(d) ~RT NDT values from Table 4-1 (shown in parentheses) were multiplied by a ratio of 0.96 (from WCAP-14689, Rev. 4[11 Table 4, CF vll5sel + CFsurv weld = 36.8 + 38.2 = 0.96) to calculate the best fit chemistry factor (CF) as provided by Reg. Guide 1.99, Rev. 2, Position 2.1.

PTLR for FNP Unit 2 Revision 4 Page 20 of 24 Table 5-3 Reactor Vessel Toughness Table (Unirradiated) (a)

Beltline Material Cu Weight % Ni Weight % IRTNDT (OF)

Closure Head Flange - - 60 Vessel Flange - - 60 Inter. Shell Plate B7203-1 0.14 0.60 15 Inter. Shell Plate B7212-1 0.20 0.60 -10 Lower Shell Plate B7210-1 0.13 0.56 18 Lower Shell Plate B721 0-2 0.14 0.57 10 Inter. Shell Longitudinal Weld Seam 19-923 A (b) 0.027 0.947 -56 (Heat # HODA)

Inter. Shell Longitudinal Weld Seam 19-923 B (b) 0.027 0.913 -60 (Heat # BOLA)

Surveillance Weld (e) 0.028 0.89 -

Circumferential Weld 0.153 0.077 -40 Seam 11-923 (b)

(Heat # 5P5622)

Lower Shell Longitudinal Weld Seams20-923 A & B (b) 0.051 0.096 -70 (Heat # 83640)

NOTES:

(a) From Table 2, WCAP-14689, Revision 4 [1]

(b) Best-estimate copper and nickel 'from CE NPSD-1039 [9]

(c) The best-estimate copper and nickel value represents the average of two chemistry measurements performed on the surveillance weld and documented in WCAP-8956 [5]

and WCAP-11438 [7]. The surveillance weld is representative of intermediate shell longitudinal weld 19-923B.

PTLR for FNP Unit 2 Revision 4 Page 21 of 24 Table 5-4 Reactor Vessel Fluence Projections for 36 EFPY (a,b)

EFPY 0° 15° 150 (e) 30° 300 (e) 45° 36 4.39 2.61 2.09 1.98 1.91 1.40 NOTES:

(a) From Table 7, WCAP-14689, Revision 4 (1) 19 2 (b) Fluence in 10 n/cm (E > 1.0 MeV)

(c) Indicates location in octants with a 26° neutron pad span.

(d) These f1uence projections remain bounding with respect to the updated 36 EFPY f1uence projections in Table 6-2, WCAP-16918-NP, Revision 1(15).

Table 5*5 Summary of Adjusted Reference Temperatures (ARTs) for Reactor Vessel 8eltline Materials at the 1/4-T and 3/4-T Locations for 32.8 EFPY (a,b)

Material 1/4-T (OF) 3/4-T (OF)

Intermediate Shell Plate 87203-1 174 149 (e)

Intermediate Shell Plate 87212-1 211 173 Intermediate Shell Plate 87212-1 183 (e) 147 Using SIC Data Lower Shell Plate 87210-1 165 142 Lower Shell Plate 87210-2 168 143 Intermediate Shell Longitudinal Weld 28 (d) 12 (d)

Seam 19-923 A (Heat # HODA)

Intermediate Shell Longitudinal Weld 10 (d) -9 (d)

Seam 19-923 8 (Heat # 80LA)

Intermediate Shell Longitudinal Weld Seam 19-923 8 (Heat # 80LA) -44 (d) -48 (d)

Using SIC Data Circumferential Weld 11-923 109 90 (Heat # 5P5622)

Lower Shell Longitudinal Weld Seams o (d) -19 (d)20-923 A & 8 (Heat # 83640)

NOTES:

(a) From Tables 13 & 14, WCAP-14689, Revision 4 (1) with service period adjusted to 32.8 EFPY per WCAP-16918 NP, Rev. 1 [15),

19 (b) The ARTs presented here are based on the peak reactor vessel surface f1uence of 4.127 x 10 n/cm2 (E > 1.0 MeV) unless otherwise noted. This f1uence remains bounding with resEect to the updated peak f1uence projection for 32.8 EFPY interpolated from Table 6-2, WCAP-16918-NP, Rev. 1[1 ) (Le., 3.559 x 1019 n/cm2 ).

(c) Limiting 1/4-T and 3/4-T ART values. The PIT limit curves are those previously generated based on 1/4-T ART of 186°F and 3/4-T ART of 149°F which bounds the limiting 1/4-T and 3/4-T ARTs shown above.

19 2 (d) ARTs calculated using the peak vessel f1uence of 1.32 x 10 n/cm (E > 1.0 MeV) at 45°. This f1uence remains bounding with res~ect to the updated 45° f1uence projection for 32.8 EFPY interpolated from Table 6-2, WCAP 16918-NP, Rev. 1 151 (Le., 1.14 x 1019 n/cm 2 ).

PTLR for FNP Unit 2 Revision 4 Page 22 of 24 Table 5-6 Calculation of Adjusted Reference Temperature at 32.8 EFPY for the Limiting Reactor Vessel Material (a)

Parameter Intermediate Shell Intermediate Shell Plate B7212-1 Plate B7203-1 Operating Period 32.8 EFPY 32.8 EFPY Location ~-T  %-T ~-T  %-T Chemistry Factor, CF (OF) 140.3 140.3 100.0 100.0 Fluence, f (10 19 n/cm 2 ) (b) 2.573 1.00 2.573 1.00 Fluence Factor, FF 1.253 1.00 1.253 1.00

~RT NOT = CF x FF (OF) 175.8 140.3 125.3 100.0 Initial RTNOT, I (OF) -10 -10 15 15 Margin, M (OF) (e) 17 17 34 34 Adjusted Reference Temperature 183(d) 147 174 149(d)

(ART), (OF) per Regulatory Guide 1.99, Revision 2 NOTES:

(a) From Tables 13 & 14 (using surveillance capsule data), WCAP-14689, Revision 4 [11 with service period adjusted to 32.8 EFPY per WCAP-16918-NP, Revision 1[151.

=

(b) Fluence is based on fsurf (10 19 n/cm 2 , E> 1.0 MeV) 4.127. This f1uence remains bounding with respect to the updated peak fluence projection for 32.8 EFPY interpolated from Table 6-2, WCAP-16918-NP, Rev. 1[151 (Le., 3.559 x 1019 n/cm2). The Farley Unit 2 reactor vessel wall thickness is 7.875 inches in the beltline region.

(c) Margin is calculated as M = 2(cri2 + crl) 0.5. The standard deviation for the initial RTNOT margin term, cri, is O°F since the initial RTNOT is a measured value. The standard deviation for the ~RTNOT term, cr~, is 17°F for the plate, except that cr~ need not exceed 0.5 times the mean value of ~RTNOT. In accordance with Regulatory Guide 1.99, Revision 2, Position 2.1, values of cr~ may be cut in half when based on credible surveillance data.

(d) Limiting ~-T and %-T ART values.

PTLR for FNP Unit 2 Revision 4 Page 23 of 24 Table 5*7 Pressurized Thermal Shock (RT pis) Values for 36 EFPY (a)

Surface ARTNDT Fluence (CF x Material FF I M RTpTs CF (10 19 n/cm 2 , FF) (OF) (OF)

(OF)

E> 1.0 MeV) (OF)

Intermediate Shell Plate 87203-1 100.0 4.39 1.38 138.0 15 34 187 Intermediate Shell Plate 87212-1 149.0 4.39 1.38 205.6 -10 34 230 Intermediate Shell Plate 87212-1 140.3 4.39 1.38 193.6 -10 17 201 Using SIC Data Lower Shell Plate 87210-1 89.8 4.39 1.38 123.9 18 34 176 Lower Shell Plate 87210-2 98.7 4.39 1.38 136.2 10 34 180 Intermediate Shell Longitudinal Welds19-923 A 36.8 1.40 1.09 40.1 -56 52.6 37 (Heat # HODA)

Intermediate Shell Longitudinal Welds 19-9238 36.8 1.40 1.09 40.1 -60 40.1 20 (Heat # 80LA)

Intermediate Shell Longitudinal Welds 19-9238 (Heat # 80LA) 8.4 1.40 1.09 9.2 -60 9.2 -42 Using SIC Data Circumferential Weld 11-923 74.1 4.39 1.38 102.3 -40 56 118 (Heat # 5P5622)

Lower Shell Longitudinal Welds20-923 A & 8 37.3 1.40 1.09 40.7 -70 40.7 11 (Heat # 83640)

NOTES:

(a) From Table C-2, WCAP-14689, Revision 4 [1]

(b) These f1uence values remain bounding with respect to the updated 36 EFPY f1uence values from Table 6-2, WCAP-16918-NP, Rev. 1[15].

PTLR for FNP Unit 2 Revision 4 Page 24 of 24 6.0 References

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

2. WCAP-12471, Analysis of Capsule X from the Alabama Power Company Joseph M.

Farley Unit 2 Reactor Vessel Radiation Surveillance Program, E. Terek, et aI.,

December 1989.

3. WCAP-14687, Joseph M. Farley Units 1 and 2 Radiation Analysis and Neutron Dosimetry Evaluation, R. L. Bencini, June 1996.

4. NUREG-0117, Supplement 5 to the Safety Evaluation Report (NUREG-75/034),

Office of Nuclear Reactor Regulation, U. S. Nuclear Regulatory Commission in the matter of Alabama Power Company Joseph M. Farley Nuclear Plant Unit 2, Docket No. 50-364., March 19, 1981.

5. WCAP-8956, Alabama Power Company Joseph M. Farley Nuclear Plant Unit No.2 Reactor Vessel Radiation Surveillance Program, J. A. Davidson, et aI., August 1977.

6. WCAP-10425, Analysis of Capsule U from the Alabama Power Company Joseph M.

Farley Unit 2 Reactor Vessel Radiation Surveillance Program, S. E. Yanichko, et aI.,

October 1983.

7. WCAP-11438, Analysis of Capsule W from the Alabama Power Company Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program, S. E. Yanichko, et aI., April 1987.

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

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-15171, Revision 1, Analysis of Capsule Z from the Alabama Power Company Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program, February 2000.

11. WCAP-16351-NP, Revision 0, Analysis of Capsule Y from the Southern Nuclear Operating Company Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program, February 2005.

12. SNC letter NL-04-0372, Reactor Material Surveillance Program Specimen Capsule Withdrawal Schedule Revisions - Additional Information, March 5, 2004.

13. NRC letter (SNC LC #14001) Joseph M. Farley Nuclear Plant, Units 1 and 2 Re: Specimen Capsule Withdrawal Schedule Revisions, March 15, 2004.

14. NRC Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March, 2001.

15. WCAP-16918-NP, Revision 1, Analysis of Capsule V from the Southern Nuclear Operating Company Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program, April 2008.