ML20197B647

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Rev 1 to Jfnp - Unit 2 Pressure Temperature Limits Rept
ML20197B647
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Site: Farley  Southern Nuclear icon.png
Issue date: 12/18/1997
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NUDOCS 9712240016
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Joseph M. Farley Nuclear Plant - Unit 2 P..ssure Temperature Limits Report -

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' Joseph M. Farley Nuclear Plant Unit 2 4

Pressure Temperature Limits Report 4

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PRESSURE TEMPERATURE LIMITS REPORT

- Table Of Contents j ust or Tabies . . .. . . . ... . iii ust or rigures .. .. .. . . .. . . . . .. _ iv 1.0 RCS Pressure Temperature Limits Heport (PTLR) .. ... .. .... .. . ....... I 2.0 Operating Limits ... . . . . . . . . ... .... ..I 4

2.1 RCS Pressure / Temperature (P/T) Limits (LCO . 3.4.10.1).. .. ... . .... ..... .................../

- 2.2 RCP Operation Limits . . . .... . . ..... . . . . . . . . . . . . . . . . . . . ., . . . . . . . . . ........./~-

3.0 Reactor Vessel Matertal Surveillance Program ....... . . .... .. .. ..... . . 6 -

4.0 Reactor Vessel Surveillance Data Credih11ity . ....... ....- .. . .-.-..... . -.. 7 5.0 Supplementat Data Tables .. .. .. .. . . ...~ .. . . .. ..12 6.0 References . . . .. . ... .. .. . . ..... 19 a

FARLEY UNIT' 2' ii REVISION 1^

PRESSURE TEMPERA 111RE LIMITS REPORT I

List of Tables -

21 Farley Unit 2 33.8 EFPY lleatup Curve Data Points . ~ ~ .~. .~.. .~ ..-~ -~ -~... 4 22 Fariey Unit 2 33.8 EFPY Cooldown Curve Data Points . .. . .~. ~. ... 3 31 Survelllance Capsule Withdrawal Schedule. .. .~ ~.-.... -..~ - ... ~.-~ .~.- 6 41 Surveillance Capsule Data Calculation of flest-Fit Line as Described in Position 2.1 of Regutatory Guide 1.99, Revision 2 ..... ~. - .. . . ... ~. ... I0 42 Scatter of ARTh or Values About a liest. Fit Line for Surveillance Plate Material. .. . .. it 43 Scatter of ARTsor Values About a Best Fit Line for Surveillance Weld Material. .... .. ....it 51 Comparison of Surveillance Material 30 Ft-Lb Transition Temperature Shifts and Upper Shelf Energy Decrease with Regulatory Guide 1.99, Revision 2, Predictions .... ..13 5-2 Calculation of Chemistry Factors Using Surveillance Capsule Data.. . . ... .. .....14 53 Res.ctor Vessel Toughness Tsble (Unirradlated)....~ . .-. .. ..- .. ~.- . -. . ....- 15 5-4 Reaefor Vessel FIuence Projections for 36 EFPY .... ~.. ~ . . . ... ... ... ..I6 55 Summary of Adjusted Reference Temperatures (ARTS) for Reactor Vessel Beltline Materials at the 1/4.T and 3/4.T Locations for 33.8 EFPY. .. . .. .. . 16 5-6 Ca:culation of Adjusted Reference Temperature at 33.8 EFPY for the Limiting Reaetor Vessel Material . . . .... .. . . ... .. ..... . .. . .17 5-7 Pressurised Thermal Shock (RTers) Values for 36 EFPY - . . ... ... .... .. ...-.I8 FARLEY UNIT 2 iii REVISION I r

. . . _ _ . .- ~ - . ~ ~ . . . . _ . _ . . . _ _ -- . . ., . . . _. _ _ . . _ _ _ _ _ . . . _ _

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' PRESSURE TEMPERATURE LIMITS REPORT 4

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Lia of Figures e

21-l . Farwy Unk 2 Reacter Coolent System Hentup Limitstions. .. .2-1- ,

-- 2 2 ; Farley Unit 2 Reactor Coolant System Cooldewn Limitations . - 3 e

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  • PRESSURE TEMPERATURE LIMITS REPORT

,1.0 RCS. 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) 6.9.1.15. Resisions to the PTLR shall be provided to the NRC after issuance..

'Ihis repott afTects TS 3.4.10.1; RCS Pressure r remperature Limits (P/r) Limits. All TS requirements associated with low temperature ove pressure protection (LTOP) are contained in TS 3.4.10.3, RCS Overpressure Protection Systems.

2.0 Operating Limits The limits for TS 3.4.10.1 are presented in the subsection which follows and were developed using the NRC approved methodologies specified in TS 6.9.1.15. He operability requirements associated with LTOP are specified in TS LCO 3.4.10.3 and were detennmed to adequately protect the RCS against brittle fracture in the event of an LTOP transient in accordance with the methodology specified in TS 6.9.1.15. 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 3.4.10.1) 2.1.1 The minimum boltup temperature is 70*F.

2.1.2 The RCS temperature rate of change limits are:

a. A maximum hertup of 100'F in any one hour period.
b. A maximum cooldown of 100*F in any one hour period.
c. A maximum temperature change ofless than or equal to 10*F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit emves.

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 10*F with the exception that a second pump may be started for the purpose of maintaining continuous flow while taking the operating pump out ofservice.

FARLEY UNIT 2 1 REVISION I

.~ - . . - . . . - -. . - - . - - - - - . . . - - .. _ . .. . . -.

- PRESSURE TEMPERATURE LIMITS REPORT =-

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--2.500

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I Leak Test Urrut I~ 1 1 Crtocality Urrut for I l -; . 60 F/hr Heatup

.) f) 2,2% -

Umegng MaterW.114T at 33 8 EFPY: .. -

t intermeeste Shoe Plate B72121 ~ j -i

. ART *196 F . 4 Criticality Umst for Umeng Maternal.3/4T at 33.8 EFPY:- j W a up Intermodese Shen Place 87203-1 l. -I ' ,

2.000 - ARTS 149 F i

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eor the Sennce Pened up to 33 8 EFPY WWE RCS Sotup Temperature e 70 F l I' t 0

'0 50 : 100 150;- 200 250- 300- 350 400 450 500 indicated Temperature (Degree F)

- Figure 2 l'

' Farley Unit 2 Iw.ctor Coolant System licatup Limitations (Heatup Rates up to 100*F/hr) Applicable to 33.8 -

EFPY (adjusted to include 60 psi AP at RCS temperatures 2 Il0*F and 25 psi AP at RCS temperatuies <

110*F); includes vessel flange requirements of IS0*F and $61 psig per 10 CFR 50, Appendix G.;'l

~ FA RLEY UNIT 2 2 REVISION I W e- , y -: 2v- y - , s-v-. gr ,

PRESSURE TEMPERATURE 1.lMITS REPOR1.

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. Umting Mdertid 1/4T at 28 EFPt Irtermdale Shd Pime B72121 - -l  !

2.250 ART = 186 F , ,

UmtirgIMenad 3/4T a 33.8 EFPf, I-Irtermdme Shen Rae B720M  !

ART = 149 F  ; l 2.000

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0 50 100 150 200 250 300 350 .400 450 500 IrxicdedTemperature(Degree F)

- Figure 2 2 Farley Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100*F/hr)

Applicable to 33.8 EFPY (adju:ted to include 60 psi AP at RCS temperatures 2110*F and 25 psi AP at RCS temperatures < 110*F), includes vessel flange requirements of 180 F and $61 psig per 10 CFR 50, Appendix G.I" FARLEY UNIT 2 3 REVISION I

I PRESSURE TEMPERATURE LIMITS REPORT 4

60 *F 60 *F Cnticahty Limit 100*F 100 *F Cnticahty Limit Leak Test T l P T l P T l P T l P T l P 70 472 314 0 70 437 314 0 292 2000 75 472 314 465 75 437 314 467 314 2485 80 472 314 454 80 - 437 314 451 85 472 314 446 85 437 314 438 90 472 314 441 90 437 314 428 95 472 314 438 95 437 314 419 100 472 314 437 100 437 314 413 105 472 314 438 105 437 314 408 110 473 314 441 110 437 314 405 110 438 314 444 110 402 314 403 ,

115 441 314 449 115 402 314 402 120 444 314 455 120 402 314 403 125 449 314 462 125 402 314 404 130 455 314 469 130 403 314 407 135 462 314 478 135 404 Sid 411 140 469 314 488 140 407 314 416 145 478 314 498 145 411 314 422 150 488 314 510 150 416 314 428 155 498 314 522 155 422 314 436 160 510 314 536 160 428 314 445 165 522 314 551 165 436 314 455 170 536 314 567 170 445 314 4C6 175 551 314 584 175 455 314 478 180 561 314 602 180 466 314 491 180 567 314 622 185 478 314 505 185 584 314 6J4 190 491 314 521 100 602 314 66? 195 505 314 538 195 622 314 692 200 521 314 556 200 644 314 719 205 538 314 576 205 667 314 747 210 556 314 598 210 692 314 778 215 576 314 621 215 719 314 811 220 598 314 646 220 747 314 847 225 621 314 673 225 778 314 885 230 646 314 702 230 811 314 926 235 673 314 733 235 847 314 970 240 702 314 767 240 885 314 1018 245 733 314 803 245 926 314 1069 250 767 314 842 250 970 314 1119 255 803 314 883 255 1018 314 1171 260 842 314 928 260 1069 315 1223 265 883 315 976 265 1119 320 1273 270 928 320 1028 270 1171 325 1326 275 976 325 1083 275 1223 330 1383 280 1028 330 1143 280 1273 335 1445 285 1083 335 1206 285 1320 340 1510 290 1143 340 1275 290 1383 345 1580 295 1206 345 1348 295 1445 350 1656 300 1275 350 1426 300 1510 355 1736 305 1348 355 1510 305 1580 360 1822 310 1426 360 1599 310 1656 365 1914 315 1510 365 1695 315 1736 370 2013 320 1599 370 1798 320 1822 375 2118 325 1695 375 1908 325 1914 380 2231 330 1798 380 2025 330 2013 385 2351 335 1908 385 2150 335 2118 340 2025 J90 2283 340 2231 345 2150 395 2425 345 2351 350 2283 355 2425 Table 2-1 Farley Unit 2 33.8 EFPY lleatup Curve Data Points (adjusted to include 60 psi AP at RCS temperatureQ 110*F and 25 psi AP at RCS t mperatures < 110*F)N o

FARLEY UNIT 2 4 REVISION I

m- . - , . - _. . ._. __ ,_ _____.m.m~..._... . . . . . - _ - . ..m. - -.

PRESSURE TEMPERATURE LIMITS REPORT -

o *F - 20*F 40*F 60 *F 100 *F +

T l-- -P- T l P T j P- T l P T l P 70 498 70 . .460 70 ,422 70 383 70 - 301-.

~75: 501 75: 463' ' 75 ' = 425 -- 75 ~ 386- 75 . 305- r 80-- 505 80 - ~467- 80 -428 80 .389 80 - 308-  ?

- 85 508 85 471 85 .432' 85 393 85 312-90 : - $12 - '90 474 ' 90 ' 436 90 '397 90 317

-$16: -479 402- 95 321 95 954 95 440 95 1

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, 100-.- $21- ' '100 '

-483 100- '445- 100 406 :100~ 327 105 525 t 105 488 105 - 450 105 412 105 332 110- 531 z110 '494 110 456 110- 417 - 110 338 .

110 496 1110 459 110 421- :110 382- 110 303-

~115 C01 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 l

.140 536 140 500 140 464 140 - 428 140 353 145 545 145 509 145' 474 145 438 145 364 150 -554 150- $19 - 150 484 150 448 150 376 l

> 155 - 561 155 529' 155 495 155 460 155 389 4 160 561- 160 541 ISO 507 160 472 160 403 165- 561 .165- 553 165 519- 165 486- 165 418 170 561 170 561 170 533 170 80 0 170 434' 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 61d 190 599 19$ 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 i 220 786 220 764 220 743 *25

. 756 225 'T5 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 4

260 1070 260 1067- 260 1066 265 t119 255 1118

! 270 1171 .

275. 1226 280 1286 285 1351

, 290 1420 295- 1494 300 1573 305 1658 j 310 1749 315- 1846 4

320 1951-325. 2062.

330 -

2182L i 335 2309 Table 2 2 Farley Unit 2 33.8 EFPY Cooldown Curve Data Points (adjusted to include 60 psi aP at RCS temperatures 2 Il0*F and 25 psi AP at RCS temperatures < 110*F)DI 1FARLEY UNIT 2 5 REVISION 1-

PRESSURE TEMPERATURE LIMITS REPORT 3.0 Reactor Vessel Material Surveillance Program The reactor vessel material surveillance prograr.4 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 31. Consistent with specific requirements for Farley Unit 2 associated with the granting HI of an exemption to Appendix II of 10 CFR 50 documented in NUREG-0117 , Figures 21 and 2 2 are based on the greater, or limiting value, of the following: (1) the actual shift in reference temperature for plate B7212 1 an detennined by i.npact testing, or (2) the predicted shitl in reference temperature for weld seam Il 923 as determined by Regulatory Guide 1.99, Revision

2. 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 nnWtWl exceeds the ART used to generate the P/T limits shown in Figures 2-1 and 3 2 for the specifM fluence period.

Table 3-1 SURVEILLANCE CAPSULE WITilDRAWAL SCIIEDULE

  • Capsule Capsule Locadou Lead Removal Fluenec 2

(Degree) Factor EFPY*' (n/cm )

343 3.11 1.10 639 x 10

W* 110 2.69 3.97 1.85 x 10" X* 287 3.21 6.41 3,18 x 10" Z 340 2.77 12.2 4.39 x 10

V 107 3.21 17.2 6.52 x 10

Y 290 2.77 Standby -

NOTES:

(a) WCAP-14689, Revision 2 U3 (b) Effective Full' Power Years (EFPY) from plant startup (c) Plant-specific evaluation FARLEY UNIT 2 6 REVISION 1

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 suneillance data sets become available from the reactor in question.

To date, there have been three capsules 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 surveilk..e 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 sun eillance data is credible.

Criterion 1: Materials in the capsules should be thosejudged 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 w elds, 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 suf6cient 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; e Lower shell plates B7210-1 and B7210-2;
  • Lower shell longitudinal weld seams20-923 A & B, heat number 83640, Linde 0091 Oux, flux lot 3490; and e Circumferential weld Il 923, heat number SP5622, Linde 0091 flux, tiux lot 1122.

FARLEY UNIT 2 7 REVISION 1

PRESSURE TEMPERATURE LihflTS REPORT Per WC/38956 ,Wthe Parley Unit 2 surveillance program was based on ASTM El85 73, Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels. Per Section 4.1 of ASTM El85 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 ofinitial transition temperature, upper shelf energy level, and estimated increase in transition temperature considering chemical composition (copper and phosphorus) and neutron fluence.

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 are 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 are 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. W Although the Farley Unit 2 surveillance weld material does not meet the requirements of Criterion I, conservative methods of analysis are available to predict the radiation characteristics of the limiting beltline weld. The limiting beltline plate material is intennediate 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.

FARLEY UNIT 2 8 REVISION 1

PRESSURE TEMPERATURE LIMITS REPORT

Criterion 2
- Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to pennit the determination of the 30 ft lb temperature and upper shelf energy unambiguously, Plots of Charpy energy versus temperature for the unirradiated condition are presented in WCAP-8956 Dl, 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 l'1, WP3 and X Pl.

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

Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter ofARTwor 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 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 E18f-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 following page]

FARLEY UNIT 2 9 REVISION 1

PRESSURE TEMPERATURE LIMITS REPORT

. TABLE 4-1 SURVEILLANCE CAPSULE DATA CALCULATION OF f1EST FIT LINE AS DESCRIBED IN POSITION 2.1 OF REGULATORY GUIDE 1.99. REVISION 2"'

  1. "D' Material Capsule F*' )

A T or 2 U 0.639 0.874 103 90.0 0.764 Intermediate Shell Plate 1372121 W l.85 1.17 -165 193.1 1.37 ngitudnal) 234.0 1.69 X 3.18 1.30 180 L

U 0.639 0.874 133 116.2 0.764 Intermediate Shell Plate 1172121 W l.85 1.17 165 193.1 1.37

" '# "'" I 247.0 1.69 X 3.18 1.30 190 i 1073.4 7.65 2

CF = I(FF

  • ARTsor) + I(FF ) = 140.3 'F Weld Metal U 0.639 0.874 10 8.74 0.764 W l.85 1.17 10 11.7 1.37 X 3.18 1.30 10 13.0 1.69 p 33.4 3.82 2

CF = I(FF

(a) WCAP 14689, Revision 2 N (b) F = Fluence (10 Wem*, E > 1.0 MeV)

.(c) FF = Fluence Factoi - F#'""

FARLEY UNIT 2 10 REVISION 1 4

r

._ PRESSURE TEMPERATURE LIMITS REPORT TABLE 4-2 SCATTER OF ARTuor VALUES ABOUT A DEST-FIT LINE FOR SURVEILLANCE PLATE MATERIAL'"

Intenuediate Shell ARTuor Best Fit ARTwor Scatter of ARTsor Plate B72121 FF- (30 ft lb) ('F) . . (*F)

Orientation ('F) 0.874 103 122.6 19.6 Longitudinal 1.17 165 164.2 0.8 1.30 180 182.4 2.4 0.874 133 122.6 10.4 1.17 165 164.2 0.8 Transverse 1.30 190- 182.4 7.6 NOTES:

(a) WCAP 14680, Revision 2 UI The scatter of ARTuor 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 one n (17 F) for base metal. From a statistical point of view, one c (17'F) should encompass approximately 68% of the data and one would es.pect at least one data point to be outside the one a limits. As shown above, the scatter is within 17'F of the best fit line for all data points except one. The one point that is not within the scatter is 19.6'F below the best fit line (i.e., the best fit line over predicts this data point). Based on the above discussion and engineering judgment, the Farley Unit 2 surveillance program conservatively predicts the material properties changes for the beltline plate. Therefore, the intent of Criterion 3 is met for the Farley Unit 2 surveillance plate material.

TABLE 4-3 SCATTER OF ARTsor VALUES ABOUT A DEST-FIT LINE FOR SURVEILLANCE WELD MATERIAL'"

~

ARhor(30 (1 lb) Best Fit ARTsor Scatter of ARTsor Material FF

(*F) (*F) (*F) 0.874 10 7.6 2.4 Weld Metal 1.17 10 10.2 0.2 1.30 10 11.3 1.3 NOTES, (a) WCAP-14689, Revision 2 UI FARLEY UNIT 2 11 REVISION I

PRESSURE TEMPERATURE LIMITS REPORT The scatter of ARTum vclues about a best fit line drawn with the y intercept equal to zero, as described in Regulatory Position 2.1,is less than 28 F as shown above. Therefore, Criterion 3 is met for the Farley Unit 2 surveillance weld material.

Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the claddingtase metal interface within 125'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 sp&ns 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 r.ot include correlation monitor material. Therefore, this criterion is not applicable to Farley Unit 2.

CONCI USION:

Based on the preceding responses to the criteria of Regulatory Guide 1.99, Revision 2, Section B, and the application of engineeringjudgment, the Farley Unit 2 surveillance d'ta is credib!e.

5.0 Supplemental Data Tables Table 5 1 contains a comparison of measured surveillance material 30 ft 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 2 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 Parley Unit 2 reactor vessel beltline materials at the 1/4-T and 3/4 T locations for 33.8 EFPY.

Table 5-6 shows the calculation of the ART at 33.8 EFPY for the limiting Farley Unit 2 reactor vessel material.

Table 5-7 provides RT,rs values for Farley Unit 2 for 36 EFPY.

- FARLEY UNIT 2 12 REVISION 1

PRESSURE TEMPERATURE LIMITS REPORT t

Table 5.I COMPARISON OF SURVEILLANCE MATERIAL 30 FT.Lil TRANSITION TEMPERATURE SHIFT AND UPPER SilEl.F ENERGY DECREASE WITl! REGULATORY GUIDE 1.99, REVISION 2, PREDICTIONS

  • 3011-Ib Transition Upper ShellT.nergy Fluence Temperature Shift Decrease Material Capsule (x 10" rvem*, Predicted Measured Predicted Measured E > 1.0 MeV) ('F) (*F) (%) (%)

U 0.639 130.2 103 26 28 Plate 872121 (longitudinal) W l.85 174.2 165 34 22 X 3.18 194.3 180 39 28 U 0.639 130.2 133 26 27 Plate B72121 (Transverse)

W  !.85 174.2 165 34 20 X 3.18 194.3 190 39 27 0 0.639 ' 33.4 10 17 8 Weld Metal W l.85 44.7 10 22 0 X 3.18 49.8 10 25 0 U 0.639 - 58 .. 30 llAZ Metal W l.85 .. 109 .. 20 X 3.18 .. 110 .. 20 NOTES:

UI (a) WCAP.14689. Revision 2 FARLEY UNIT 2 13 REVISION 1

PRESSURE TEMPERATURE LIMITS REPORT Table 5-2 CALCULATION OF CilEMISTRY FACTORS USING l

SURVEILLANCE CAPSULE DATA 'I 2

Material l Capsule f *' FF" ARTsor FF

~

Intermediate Shell 0 0.639 0.874 103 90.0 0.764 Plate !!72121 W l.85 1.17 165 193.1 1.37 (Longitudinal.) X 1.30 180 234.0 1.69 3.18 Intermediate Shell U 0.639 0.874 133 116.2 0.764 Plate 1372121 W l.85 1.17 165 193.1 1.37 (Transverse) X 3.18 1.30 190 247.0 1.69 Sum: 1073.4 7.65 2

Chemistry Factor - E (FF ' ARTsor) + E (FF ) = 140.3 'F U 0.639 0.874 9.6 8.4 0.764 Weld Metal"' W l.85 1.17 9.6 11.2 1.37 X 3.18 1.30 9.6 12.5 1.69 Sum: 32.1 3.82 Chemistry Factor = E (FF ' ARTuor) + E (FF ) = 8.4 'F NOTES, 4 (a) WCAP 14689, Revision 2 08 (b) f = fluence (x 10" n/cm 2, E > l.0 MeV)

(c) i F = Ouence factor = f *i ' '5 "

(d) ARTsar values were multiplied by a ratio of 0.96 (CF,,,,,, + CF,,,,a = 36.8 + 38.2 = 0.96)

FARLEY UNIT 2 14 REVISION I

PRESSURE TEMPERATUPE LIMITS RFPORT 4

Table 5 3 REACTOR VESSEL TOUGilNESS TABLE (UNIRRADIATED)"'

Beltline Material . Cu Weight % Ni Weight % IRTsor (*F)

Closure Ilead Flange .- .. 60 .

Vessel Flange .. 60 Inter. Shell Plate B72031 0.14 0.60 15 Inter. Shell Plate B72121 0.20 0.60 10 Lower Shell Plate B7210-1 0.13 0.56 18 Lower Shel, Plate B7210 2 0.14 0.57 10 Inter. Shell Longitudinal Weld Seam 19-923 A O.027 0.947 56

- (IIcat # 110DA)

Inter. Shell Longitudinal Weld Seam 19 923 8 '" .0.027 0.913 -60 (lleat # BOLA)

Surveillance Weld "' O.028 0.89 Circumferential Weld Seam 11923 *' 0.077 40 O.153 (lleat # $P5622)

Lower Shell Longitudinal Weld Seams20-923 A & B '" 0.051 0.096 -70 (lleat # 83640) t NOTES:

(a) WCAP 14689, Revision 2i 'l

. (b) Dest-estimate copper and nickel from CE NPSD 1039I 'l (c) The best estimate copper and nickel value represents the average of two chemistry measurements performed on the surveillance we!d and documented in WCAP-8956 IU and WCAP.11438 (71 The surveillance weld is representative ofintermediate shell longitudinal weld 19 923B FARLEY UNIT 2 15 REVISION 1

PRESSURE TEMPERATURE LIMITS REPORT Table 5 4 RF SCTOR VES3EL FLUENCE PROJECTIONS FOR 36 EFPY '

liFPY O' 15' I5" 30* 30* '" 45' 36 4.39 2.61 2.09 1.98 1.91 1.40 NOTLS.

(a) WCAP 14689, Revision 2 I4 3

- (b) Fluence in 10" n/cm (E > 1.0 MeV)

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

Table 5-3

SUMMARY

OF ADJUSTED REFERENCE TEMPERATURES (ARTS) FOR REACTOR VESSEL BELTLINE MATERIALS AT Ti!E 1/4-T AND 3/4 T LOCATIONS FOR 33.8 EFPY'"'

Material 1/4 T ('F) 3/4 T ('F)

Intermediate Shell Plate B72031 174 149 "'

Intermediate Shell Plate B72121 211 173 Intermediate Shell Plate B72121 Ig3 ki 147 Using S/C Data 1.ower Shell Plate B7210-1 165 142 Lower Shell Plate B7210 2 168 143 Intermediate Shell Longitudinal Weld Seam 23 (* 12 (*

19 923 A (Ileat # llODA)

Intermediate Shell Longitudinal Weld Seam toia 9 (*

19 923 B (llest # BOLA)

Intermediate Shell Longitudinal Weld Seam 19-923 B (Heat # BOLA) -44 '* -4 8 '*

Using S/C Data Circumferential Weld 11-923 109 90 (lleat # SP5622)

Lower Shell Longitudinal Weld Seams20-923 0 ** 19 '*

A & B (lleat # 83640)

NOTES:

(a) WCAP 14689, Revision 2 IU 3

(b) The ARTS presented here are based on the peak reactor vessel surface fluence of 4.127 x 10" n/cm (E > 1.0 MeV) unless otherwise noted.

(c) Limiting 1/4 T and 3/4 T ART values. The P/T 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.

3 id) ARTS calculated using the peak vessel Duence of 1.32 x 10* n'em (E > 1.0 MeV) at 45'

)

FARLEY UNIT 2 16 REVISION 1

~ _ - . . -

PRESSURE TEMPERA 1URE LIMITS REPORT Table 5-6 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE AT 33.8 EFPY FOR TiiE LIMITING REACTOR VESSEL MATERIAL "'

Parameter Intermediate Shell Plate Intermediate Shell Plate B72121 B72031 Operating Period 33.8 EFPY 33.8 EFPY Location  %.T %T  %-T  %-T Chemistry Factor, CF (*F) 140.3 1403 100.0 100.0 a 1.00 Fluence. f(10* n'em ) m 2.573 1.00 2.573 Fluence Factor, FF 1.253 1.00 1.253 1.00 175.8 1403 125.3 100.0 ARTwor = CF x FF (*F)

Initial RTuor,I('F) 10 10 15 15 M argin, M ('F) "' 17 17 34 34 Adjusted Reference Temperature (ART),(*F) per 383 347 I74 I49 Reguiatory Guide 1.99, Revision 2

NOTES, (a) WCAP 14689, Revision 2I 'I 2

(b) Fluence is based on fw(10" n/cm , E > 1.0 MeV) = 4.127 at 33.8 EFPY. The Farley Unit 2 reactor vessel wall thickness is 7.875 inches in the beltline region.

(c) Margin is calculated as M = 2(o,' + o3 2)*8. The standard deviation for the initial RTsor margin term, o,, is 0*F since the initial RT8or is a measured value. The standard deviation for the ARTN otterm, o3, is 17'F for the plate, except that o3 need not exceed 0.5 times the mean value of ARTsor. In accordance with Regulatory Guide 1.99, Revision 2, Position 2.1, values of o3 may be cut in half when based on credible surveillance data.

FARLEY UNIT 2 17 REVISION 1

PREFSURE TEMPERATURE LIMITS REPORT Table 5 7 PRESSURIZED THERMAL SilOCK (RTm) VALUES FOR 36 LFPY * ,

Surface Fluence ART ND '

I M Material CF (10 Wem3, FF (CF x FF)

(*F) op) - og RT{5 E > 1.0 MeV)

Intermediate Shell 100.0 4.39'  !.38 138.0 15 34 187 Plate B72031 Intermediate Shell -

149.0 '4.39 1,38 205.6 10 34 230 Plate 872121 Intermediate Shell Plate B7212 l' 140.3 4.39 1,38 193.6 -10 17 201 Using S/C Data 09.8 4.39 1.38 123.9 18 34 176 72 0 1 Lower Shell Plate 98.7 4.39 1.38 136.2 10 34 180 B7210 2 Intermediate Shell Longitudinal Welds 36.8 1.40 1.09 40.1 56 52.6 3  :

19 923 A (Heat # llODA)

Intermediate Shell Longitudinal Weldt 36.8 1.40 1.09 40.1 60 40.1 20 19-923 B (lient # BOLA)

Intermediate Shell Longitudinal Welds 19 923 B 8.4 1.40 1.09 9.2 60 9.2 42 (IIcat # BOLA)

Using S'C Data Circumferential Weld 11-923 74.1 4.39 1.38 102.3 -40 56 118 (lleat # SP5622)

Lower Shell Longitudinal Welds 37.3 1.40 1.09 40.7 -70 40.7 11 20-923 A & B (llcat # 83640)- ,

NOTES:

(a) WCAP 14689, Revision 2 UI FARLEY UNIT 2 18 REVISION 1

PRES $URE T Eht!' ERA 1URE EthillS REPORT 6.0 References

1. WCAP 14689, Revision 2 Farley Units I and 2 lleatup and Cooldown Limit Cunes for Normal Operation and PTLR Suppon Documentation, E. Terek, December 1997.
2. WCAP 12471, Analysis of Capsule X from the Alabama Power Company Joseph ht. Farley Unit 2 Reactor Vessel Radiation Sun'eillance Program E. Terek, et al., December 1989.
3. WCAP 14687, Joseph M. Farley Units I and 2 Radiation Analysis and Neutron Dosimetry Evaluation, R. L. Ilencini, June 1996.
4. NUREG 0117, Supplement $ to the Safety Evaluation Repon (NUREO 75/034), Omce 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. $0 364., March 19, 1981.
5. WCAP 8956, Alabama Power Company Joseph M. Farley Nuclear Plant Unit No. 2 Reactor Vessel Radiation Suneillance Program, J. A. Davidson, et al., AuFust 1977.
6. WCAP 10425, Analysis of Cariute U from the Alabama Power Company Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program, S. E. Yanichko, et al., 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 al., April 1987.
8. WCAP 14040 NP A, Revision 2 Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Ileatup and Cooldown Limit Curves, January 1996.
9. CE NPSD 1039, Revision 2. Ilest Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds, Lsmbustion Engineering Owners Group, June 1997.

FARLEY UNIT 2 19 REVISION 1

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