ML082670287: Difference between revisions

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#REDIRECT [[NL-08-1385, Submittal of Revision 4 of the Pressure Temperature Limits Report]]
| number = ML082670287
| issue date = 09/22/2008
| title = Submittal of Revision 4 of the Pressure Temperature Limits Report
| author name = Ajluni M J
| author affiliation = Southern Nuclear Operating Co, Inc
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000364
| license number =
| contact person =
| case reference number = NL-08-1385
| document type = Letter, Report, Miscellaneous
| page count = 28
}}
 
=Text=
{{#Wiki_filter:Southern Nuclear Operating Company, Inc. Post Office Box 1295 Birmingham.
Alabama 35201-1295 Tel 205.9925000 SOUTHERN'\
COMPANY Energy to Serve Your WorldS" September, 22, 2008 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 Submittal of Revision 4 of the Pressure Temperature Limits 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 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 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 Enclosure Plant Pressure Temperature Limits Report Unit 2, Revision September SOUTHERN A COMPANY Energy 10 Serve JO"r WorM* Joseph M. Farley Nuclear Pressure Temperature Limits Unit Revision September 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 L TOP 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 L1RT NOT Values About a Best-Fit Line for Surveillance Plate Material 15 4-3 Scatter of L1RT NOT 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 (RT PTS) 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 Revision 4 Page 4 of 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 (L TOP) 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 L TOP are specified in TS LCO 3.4.12 and were determined to adequately protect the RCS against brittle fracture in the event of an L TOP 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. b. c. A maximum heatup of 100°F in anyone hour period. A maximum cooldown of 100°F in anyone hour period. A maximum temperature change of less than or equal to 10°F in anyone hour period during in service 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 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) The L TOP 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  at RCS temperatures  110°F and 27 psi  at RCS temperatures
< 110&deg;F). Includes vessel flange requirements of 180&deg;F and 561 psig per 10 CFR 50, Appendix G. [1] 2,500 2,250 2,000 1,750 Ci Q.-l! 1,250 l! Q. Cl" n:I -:s to)1,000 .5 750 500 250 o I I Leak Test Limit  I" il J  Criticality Limit for I I -L  60 F/hr Heatup f----t--limiting Malerial -1/4T at 32.8 EFPY Intenmediate Shell Plate B7212-1 I I ART=186 F I II ___I Criticality Limit for Limiting Material -3/4T at 32.8 EFPY: 100 F/hr Heatup i--Intenmediate Shell Plate B7203-1 I ART=149 F I I-/" I II I II I L 'I AI t 1/ I II I I I I I 1/ f/ /Unacceptable Operation
/ Acceptable Operation
-II I I 1/ / t / IJ II / / -I II I. V / I IJ / I m J V Heatup Rate J / I I\,/ V --(degree F/hr) V I nO -h ./ ./ , 00 rr I I I" ' I Cnticality Limit Based on Inservice  IV "I Hydrostatic Test Temperature (314 F) I I I (or the Service Period up to 32.8 EFPY Min. RCS Boltup Temperalure  75 F 1 I 1 I 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 1 OO&deg;F/hr) Applicable to 32.8 EFPY (adjusted to include 60 psi LlP at RCS temperatures  110&deg;F and 27 psi  at RCS temperatures
< 110&deg;F). Includes vessel flange requirements of 180&deg;F and 561 psig per 10 CFR 50, Appendix G. [11 2,500 2,250 2,000 1,750 -en 1,500 l/I- =' 1,250  l1. "C .! ns :s (,) 1,000 c: 750 500 250 o o 450 500 I I T I Limiting Material -1/4T at 32.8 EFPY: L t I I J i-I Intermediate Shell Plate 87212-1 ART =186 F I Limiting Material -3/4T at 32.8 EFPY: I Ij II Intermediate Shell Plate 87203-1 ART =149 F i I I I 1 I  I t l-t I L  I I , I i I I t II J +-I I i I  I II  I I 1 I ,I I I I I " I I I ! I l 1/ I I I I I I I /I I I f--Unacceptable Operation Acceptable Operation-f--I If I i I j I , I I I I 'I I I I I I /.rJ I I I I I j i I  c---< Cooldown Rate 0 f1j (degree F/ hr) 1\-1 I  ...7/ I I .... ,  - / I  4 -I 6 I ..... .... I  I J-L--1 J-;" 0 I II 100 I I I I II , I I I--Min. RCS Boltup Temperature
-75 F I !-+-I I I ! I I I i I 50 100 150 200 250 300 350 400 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  at RCS temperatures  110&deg;F and 27 psi  at RCS temperatures
< 1100F) [1] 60 of 60 of Criticality Limit 100 OF 100 OF Criticalitv Limit Leak Test T I p 75 470 80 470 85 470 90 470 95 470 100 470 105 470 110 471 110 438 115 441 120 444 125 449 130 455 135 462 140 469 145 478 150 488 155 498 160 510 165 522 170 536 175 551 180 561 180 567 185 584 190 602 195 622 200 644 205 667 210 692 215 719 220 747 225 778 230 811 235 847 240 885 245 926 250 970 255 1018 260 1069 265 1119 270 1171 275 1223 280 1273 285 1326 290 1383 295 1445 300 1510 305 1580 310 1656 315 1736 320 1822 325 1914 330 2013 335 2118 340 2231 345 2351 T I p 314 0 314 465 314 454 314 446 314 441 314 438 314 437 314 438 314 441 314 444 314 449 314 455 314 462 314 469 314 478 314 488 314 498 314 510 314 522 314 536 314 551 314 567 314 584 314 602 314 622 314 644 314 667 314 692 314 719 314 747 314 778 314 811 314 847 314 885 314 926 314 970 314 1018 314 1069 314 1119 314 1171 315 1223 320 1273 325 1326 330 1383 335 1445 340 1510 345 1580 350 1656 355 1736 360 1822 365 1914 370 2013 375 2118 380 2231 385 2351 T I p 75 435 80 435 85 435 90 435 95 435 100 435 105 435 110 435 110 402 115 402 120 402 125 402 130 403 135 404 140 407 145 411 150 416 155 422 160 428 165 436 170 445 175 455 180 466 185 478 190 491 195 505 200 521 205 538 210 556 215 576 220 598 225 621 230 646 235 673 240 702 245 733 250 767 255 803 260 842 265 883 270 928 275 976 280 1028 285 1083 290 1143 295 1206 300 1275 305 1348 310 1426 315 1510 320 1599 325 1695 330 1798 335 1908 340 2025 345 2150 350 2283 355 2425 T I p 314 0 314 467 314 451 314 438 314 428 314 419 314 413 314 408 314 405 314 403 314 402 314 403 314 404 314 407 314 411 314 416 314 422 314 428 314 436 314 445 314 455 314 466 314 478 314 491 314 505 314 521 314 538 314 556 314 576 314 598 314 621 314 646 314 673 314 702 314 733 314 767 314 803 314 842 314 883 314 928 315 976 320 1028 325 1083 330 1143 335 1206 340 1275 345 1348 350 1426 355 1510 360 1599 365 1695 370 1798 375 1908 380 2025 385 2150 390 2283 395 2425 T I P 292 2000 314 2485 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  at RCS temperatures  110&deg;F and 27 psi  at RCS temperatures
< 1100F) [1] o of 20&deg;F 40 of 60 of 100 of T I p 75 499 80 503 85 506 90 510 95 514 100 519 105 523 110 529 110 496 115 501 120 507 125 514 130 521 135 528 140 536 145 545 150 554 155 561 160 561 165 561 170 561 175 561 180 561 180 626 185 641 190 658 195 675 200 694 205 715 210 737 215 760 220 786 225 813 230 842 235 874 240 908 245 944 250 983 255 1025 260 1070 265 1119 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 T I p 75 461 80 465 85 469 90 472 95 477 100 481 105 486 110 492 110 459 115 464 120 470 125 477 130 484 135 492 140 500 145 509 150 519 155 529 160 541 165 553 170 561 175 561 180 561 180 595 185 611 190 628 195 647 200 667 205 689 210 712 215 737 220 764 225 793 230 824 235 858 240 894 245 932 250 974 255 1019 260 1067 265 1118 T I p 75 423 80 426 85 430 90 434 95 438 100 443 105 448 110 454 110 421 115 427 120 433 125 440 130 448 135 456 140 464 145 474 150 484 155 495 160 507 165 519 170 533 175 548 180 561 180 564 185 581 190 599 195 619 200 640 205 663 210 688 215 715 220 743 225 774 230 807 235 843 240 881 245 923 250 967 255 1015 260 1066 T I p 75 384 80 387 85 391 90 395 95 400 100 404 105 410 110 415 110 382 115 389 120 395 125 402 130 410 135 419 140 428 145 438 150 448 155 460 160 472 165 486 170 500 175 516 180 533 185 551 190 570 195 591 200 614 205 638 210 665 215 693 220 723 225 756 230 792 235 830 240 871 245 915 250 962 255 1013 260 1068 T I P 75 303 80 306 85 310 90 315 95 319 100 325 105 330 110 336 110 303 115 310 120 317 125 325 130 334 135 343 140 353 145 364 150 376 155 389 160 403 165 418 170 434 175 452 180 471 185 491 190 513 195 537 200 563 205 591 210 621 215 653 220 688 225 725 230 766 235 809 240 856 245 907 250 961 255 1019 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[4 I , 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,,[
14 1. 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 (Degree) lead Factor Removal EFPy(b) Fluence (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 10 19 (d) Y 290 3.12 19.01 6.79 x 10 19 (e) V 107 3.58 21.82 8.73 x 10 19 (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 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 [13 J*
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 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 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 seams 19-923 A, heat number HODA; Intermediate shell longitudinal weld seams 19-923 B, heat number BOLA; Lower shell longitudinal weld seams 20-923 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[5 1 , 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 seams 19-923 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 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 [6 1 , W [7 l , X [2 1 , Z [10 1 , Y [11 1 , and V [15 1. 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 When there are two or more sets of surveillance data from one reactor, the scatter of  NOT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28&deg;F for welds and 17&deg;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  NOT values about a best-fit line should be less than 1a (17&deg;F for base metal and 28&deg;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  NOT 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  NOT 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  NOT 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&deg;F of the best-fit line for all but 3. Statistically, since +/-1 a (17&deg;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 Regulatory Guide 1.99, Revision 2 Material Intermediate Shell Plate 67212-1 Capsule U W F (b) 0.605 1.73 FF (el (x) 0.859 1.151 (y) 105.5 167.7 FF X (xv) 90.7 193.0 FF 2 (x 2) 0.738 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 Intermediate Shell Plate 67212-1 (Transverse)
V U W X 8.73 0.605 1.73 2.98 1.496 0.859 1.151 1.289 218.3 124.0 168.5 200.1 326.6 106.5 193.9 258.0 2.238 0.738 1.324 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 = *
-;-
=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 =
* NOT) -;-
=16.2 of NOTES: (a) Data from Table 0-1, WCAP-16918-NP, Revision 1[15] (b) F =Fluence (10 19 n/cm 2 , 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 ilRT NOT Values about a Best-Fit Line for Surveillance Plate Material(a)
Intermediate Shell Plate B7212-1 Specimen Orientation Capsule CF (Best Fit Slope) FF !1RT NOT (30 ft-Ib) (OF) Best Fit !1RT NOT (OF) Scatter of !1RT NOT (OF) <17 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)
Surveillance Material Capsule CF (Best Fit Slope) FF !1RT NOT (30 ft-Ib) (OF) Best Fit !1RT NOT (OF) Scatter of !1RT NOT (OF) <28 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  NOT values exceed the +/-1 cr scatter band of 28&deg;F. Therefore, the surveillance weld data are deemed not credible per Criterion
: 3. Criterion The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/-25&deg;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&deg;F. Therefore, the Farley reactor vessel surveillance program meets the requirements of Criterion
: 4. Criterion 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 Therefore, this criterion is not applicable to Farley Unit Based on the preceding responses to the criteria of Regulatory Guide 1.99, Revision Section B, and the application of engineering judgment, the Farley Unit 2 surveillance are credible for the plate (base metal) material but not credible for the weld metal; surveillance capsule data are not used for the weld metal reference temperature 
------------------------
PTLR for FNP Unit 2 Revision 4 Page 17 of 24 5.0 Supplemental Data Tables Table 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 shows the calculation of the surveillance material chemistry factors using surveillance capsule data. Table provides the unirradiated Farley Unit 2 reactor vessel toughness data. Table provides a summary of thefluences used in the PTS evaluation.
Table 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 shows the calculation of the ART at 32.8 EFPY for the limiting Farley Unit 2 reactor vessel material.
Table provides RT pTs 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 Shelf Energy Decrease with Regulatory Guide 1.99, Revision 2, Predictions 30 ft-Ib Transition Upper Shelf Energy Temperature Shift Decrease Material Capsule Fluence(e)
Predicted Measured Predicted Measured (x 10 19 (OF) (b) (OF) (c) (%) (b) (%)(d) n/cm 2 , 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 (T ransverse) 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] Material Capsule F (b) FF (e) (X) ART NDT (y) FF X ART NDT (Xy) Intermediate Shell U 0.605 0.859 105.5 90.7 Plate 87212-1 W 1.73 1.151 167.7 193.0 (Longitudinal)
X 2.98 1.289 164.8 212.4 Z 4.92 1.399 200.1 280.0 Y 6.79 1.458 214.2 312.3 V 8.73 1.496 218.3 326.6 Intermediate Shell U 0.605 0.859 124.0 106.5 Plate 87212-1 W 1.73 1.151 168.5 193.9 (Transverse)
X 2.98 1.289 200.1 258.0 Z 4.92 1.399 195.8 274.0 y 6.79 1.458 231.0 336.8 V 8.73 1.496 215.3 322.1 SUM: 2906.13 CF = ART NDT) -;-
=144.6 of Weld Metal U 0.605 0.859 -27.3 (-28.4i d) -24.4 W 1.73 1.151 6.7 (7.0)(d) 8.1 X 2.98 1.289 -15.0 (-15.6) (d) -20.1 Z 4.92 1.399 9.8 (10.2) (d) 14.3 Y 6.79 1.458 66.3 (69.1) (d) 100.7 V 8.73 1.496 54.2 (56.5) (d) 84.5 SUM: 163.08 CF = ARTNDT) -;-
=16.2 of FF 2  0.738 1.324 1.662 1.958 2.125 2.238 10.046 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)  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 vll5s el + CF surv 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 % IRT NDT (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) (Heat # HODA) 0.027 0.947 -56 Inter. Shell Longitudinal Weld Seam 19-923 B (b) (Heat # BOLA) 0.027 0.913 -60 Surveillance Weld (e) 0.028 0.89 Circumferential Weld Seam 11-923 (b) (Heat # 5P5622) 0.153 0.077 -40 Lower Shell Longitudinal Weld Seams 20-923 A & B (b) (Heat # 83640) 0.051 0.096 -70 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&deg; 15&deg; 150 (e) 30&deg; 300 (e) 45&deg; 36 4.39 2.61 2.09 1.98 1.91 1.40 NOTES: From Table 7, WCAP-14689, Revision 4 (1) Fluence in 10 19 n/cm 2 (E > 1.0 MeV) Indicates location in octants with a 26&deg; neutron pad span. 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 Using SIC Data 183 (e) 147 Lower Shell Plate 87210-1 165 142 Lower Shell Plate 87210-2 168 143 Intermediate Shell Longitudinal Weld Seam 19-923 A (Heat # HODA) 28 (d) 12 (d) Intermediate Shell Longitudinal Weld Seam 19-923 8 (Heat # 80LA) 10 (d) -9 (d) Intermediate Shell Longitudinal Weld Seam 19-923 8 (Heat # 80LA) Using SIC Data -44 (d) -48 (d) Circumferential Weld 11-923 (Heat # 5P5622) 109 90 Lower Shell Longitudinal Weld Seams 20-923 A & 8 (Heat # 83640) o (d) -19 (d) NOTES: From Tables 13 & 14, WCAP-14689, Revision 4 (1) with service period adjusted to 32.8 EFPY per NP, Rev. 1 [15), The ARTs presented here are based on the peak reactor vessel surface f1uence of 4.127 x 10 19 n/cm 2 (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 10 19 n/cm 2). 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&deg;F and 3/4-T ART of 149&deg;F which bounds the limiting 1/4-T and 3/4-T ARTs shown above. ARTs calculated using the peak vessel f1uence of 1.32 x 10 19 n/cm 2 (E > 1.0 MeV) at 45&deg;. This f1uence remains bounding with to the updated 45&deg; f1uence projection for 32.8 EFPY interpolated from Table 6-2, 16918-NP, Rev. 1 151 (Le., 1.14 x 10 19 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 Plate B7212-1 Intermediate Shell Plate B7203-1 Operating Period 32.8 EFPY 32.8 EFPY Location  %-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  NOT = CF x FF (OF) 175.8 140.3 125.3 100.0 Initial RT NOT, I (OF) 10 15 15 Margin, M (OF) (e) 17 17 34 34 Adjusted Reference Temperature (ART), (OF) per Regulatory Guide 1.99, Revision 2 183(d) 147 174 149(d) 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[15 1. (b) Fluence is based on f surf (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[15 1 (Le., 3.559 x 10 19 n/cm 2). 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 RT NOT margin term, cri, is O&deg;F since the initial RT NOT is a measured value. The standard deviation for the  NOT term,  is 17&deg;F for the plate, except that  need not exceed 0.5 times the mean value of  NOT. In accordance with Regulatory Guide 1.99, Revision 2, Position 2.1, values of  may be cut in half when based on credible surveillance data. (d) Limiting  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) Material CF Surface Fluence (10 19 n/cm 2 , E> 1.0 MeV) FF ART NDT (CF x FF) (OF) I (OF) M (OF) RT pTs (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 Using SIC Data 140.3 4.39 1.38 193.6 -10 17 201 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 Welds 19-923 A (Heat # HODA) 36.8 1.40 1.09 40.1 -56 52.6 37 Intermediate Shell Longitudinal Welds 19-9238 (Heat # 80LA) 36.8 1.40 1.09 40.1 -60 40.1 20 Intermediate Shell Longitudinal Welds 19-9238 (Heat # 80LA) Using SIC Data 8.4 1.40 1.09 9.2 -60 9.2 -42 Circumferential Weld 11-923 (Heat # 5P5622) 74.1 4.39 1.38 102.3 -40 56 118 Lower Shell Longitudinal Welds 20-923 A & 8 (Heat # 83640) 37.3 1.40 1.09 40.7 -70 40.7 11 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 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. 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. WCAP-14687, Joseph M. Farley Units 1 and 2 Radiation Analysis and Dosimetry Evaluation, R. L. Bencini, June 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. 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. 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. 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. WCAP-14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, January 1996. 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 Re: Specimen Capsule Withdrawal Schedule Revisions, March 15, 14. NRC Regulatory Guide 1.190, "Calculational and Dosimetry Methods Determining Pressure Vessel Neutron Fluence," March, 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.}}

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