ML20149K105

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Rev 0 to Jm Farley Nuclear Plant Unit 2, P/T Limits Rept
ML20149K105
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Site: Farley  Southern Nuclear icon.png
Issue date: 07/23/1997
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NUDOCS 9707290252
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Text

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ENCLOSURE 7 ,

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

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Pressure Temperature Limits Report i.

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PRESSURE TEMPERATURE LIMITS REPORT Table Of Contents ListofTables.....................................................................................................................iii ListofFigures..........................................................................................................iv 1.0 RCS Pressu re Tem peratu re Limits Report (PTLR)............. ............................................. ....... I 2.0 Operating Limits....... .....................................................................................................I 2.1 RCS Pressurefremperature (P/F) Limits (LCO - 3.4.10.1).. .I 2.2 RCP Operation Limits.. . . .1 3.0 Reactor Vessel M aterial Su rveilla nce Program .. ........................... ...................................... 6 4.0 Reactor Vessel Su n eilla nce Data Credibility............................................ . ......... ........................ 6 5.0 S u pple me n t al D ais Ta b les....... .... .... ....... ............ .......... ................ ............ .. . ...... ... I 2 6.0 References..............................................................................................................................19 1

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FARLEY UNIT 2 ii AEVISION 0 l

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l PRESSURE TEMPERATURE LIMITS REPORT List of Tables 2-1 Fa riey Unit 2 36 E FP Y Heatu p Cu n e D ata Poin (s....... ...................................... ............ . ...... 4 2-2 Fa riey Unit 2 36 E FPY Cooldown Cu n e Data Points............................. ........ ....... .... ....... .. 5 3-1 Suneillance Capsule Withdrawal Schedule........ ..... ............................................6 4-1 Sunciliance Capsule Data Calculation of Best-Fit Line as Described in Position 2.1 of Regula ory Guide 1.99, Revision 2..... ..........................................................10 4-2 Scatter of ARTar Values About a Best-Fit Line for Suncillance Plate Material....... ... .....I1 4-3 Scatter of ARTer Values About a Best Fit Line for Surveillance Weld Material ................12 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...............14 5-2 Calculation of Chemistry Factors Using Surveillance Capsule Data........ ...... .. ....................15 5-3 Reactor Vessel Tou ghness Table (U ni rradiated) ................ ................................................ . . 16 5-4 Reactor Vessel Fluence Projections tor 36 EFPY ................................. ................. ..... ....... .... I 7 5-5 Summary of Adjusted Reference Temperatures (ARTS) for Reactor Vessel ,

Beltline Materials at the 1/4-T and 3/4-T Locations for 36 EFPY ...................... ....................17 l 5-6 Calculation of Adjusted Reference Temperature at 36 EFPY for the Limiting )

Reactor Vessel Material - Intermediate Shell Plate B7212 1.................................................. I8 )

5-7 Pressu rized Thermal Shock (RTrrs) Values for 36 E FPY ............ ................................. .......19 i l l

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l FARLEY UNIT 2 iii AEVISION 0

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l PRESSURE TEMPERATURE LIMITS REPORT l

1 List of Figures j l

2-1 Farley Unit 2 Reactor Coolant System Heatup Limitations ............ ...... ....... . .......... ..... .... . 2 2-2 Farley Unit 2 Reactor Coolant System Cooldow n Limitations. ......... ....,.... ........ ....... .... ..... 3 i

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FARLEY UNIT 2 iv AEVISION O

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PRESSURE TEMPERATURE LIMITS REPORT 3

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J 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. Revisions to the PTLR shall be provided to the NRC after issuance.

This report affccts TS 3.4.10. I, RCS Pressure / Temperature (PA') Limits. All TS requirements associated with low temperature overpressure 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. These limits have been developed using the NRC-approved methodologies specified in TS 6.9.1.15 with the exception that low temperature overpressure protection (LTOP) is provided by the RHR relief valves (RHRRVs) l in lieu of the PORVs. Therefore, the operability requirements associated with LTOP will be retained in TS LCO 3.4.10.3. 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.

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2.1 RCS Pressure / Temperature (P/T) Limits (LCO - 3.4.10.1) 2.1.1 The minimum boltup temperature is 60 F.

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

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

2.1.3 The RCS PA' limits for heatup and cooldown are specified by Figures 2-1 and 242, 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 senice.

FARLEY UNIT 2 1 AEVISION O

1 PRESSURE TEMPERATURE LIMITS REPORT I

2,500 ,

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.- . . Leak Test Limit ,,

fa , _i ,. , , , . , 2. _ , . _ '

Criticahty Limit for 60 F/hr Heatup 2,250 '

'- Limiting Material.

Intermdiate Shell Plate B7212-1

  • cntica hty Lirnit 1or Limiting ART Values at 36 EFPY: 100 F/hr Heatup 1/4T.186 F 2.000 3/4T,149 F ~

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250 Criticahty Limit Based on Inservice ' !

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Hydrostatic Test Temperature (269 F) '

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for the Service Period up to 36 EFPY - ~-~ ~

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0 O 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Degree F) l j Figure 2-1 Farley Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 100 F/hr)

Applicable to 36 EFPY (Without Margins for instrument Errors). Includes vessel flange requirements of 180*F and 561 psig per 10 CFR 50, Appendix G.14 FARLEY UNIT 2 2 AEVISION 0 l

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PRESSURE TEMPERATURE LIMITS REPORT I l

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2,250 Uniting Material: '

intermedate Shell Plate B7212-1 . -

Uniting ARTValues at 36 EFPY: .

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Figure 2-2 Farley Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to l I

100*F/hr) Applicable to 36 EFPY (Without Margins for Instrument Errors). Includes vessel flange requirements of 180 F and 561 psig per 10 CFR 50. Appendix G.I'l j FARI.EY UNIT 2 3 AEVISION O i i

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PRESSURE TEMPERATURE LIMITS REPORT 60'F 60 'F Cnticality 100 *F 100 'F Cnticality Limit Leak Test Limit T l P T l P T l P T l P T P l

60 472 314 0 60 437 314 0 292 2000 65 472 314 465 65 437 314 467 314 l 2485 i 70 472 314 454 70 437 314 451 {

75 472 314 446 75 437 314 438 80 l 472 314 441 80 437 31 4 428 '

85 472 3;: 438 85 437 314 419 90 472 314 437 90 437 314 413 95 472 314 438 95 437 314 408 100 472 314 441 100 437 314 405 105 472 314 444 105 437 314 403 I 110 473 314 449 110 437 314 402 1 110 438 314 455 110 402 314 403  !

115 441 314 462 115 402 314 404 120 444 314 469

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120 402 314 407 7 125 449 314 478 125 402 314 411 l 130 455 314 488 130 403 314 416 i 135 462 314 498 135 404 314 422 140 469 314 510

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140 407 314 428 1 145 478 314 522 145 411 314 436 '

150 488 314 536 150 416 314 445 155 498 314 551 155 422 314 455 160 510 314 567 160 428 314 466 165 522 314 584 165 436 314 478 170 536 314 602 170 445 314 491 i 175 551 314 622 175 455 314 505 180 561 314 644 180 466 314 521 l 180 567 314 667 185 478 314 538 l 185 584 314 692 190 491 314 556 190 602 314 719 195 505 314 576 l 195 622 314 747 200 521 314 598 '

200 644 314 778 205 538 314 621 l 205 667 314 811 210 556 314 646 l 210 692 314 847 215 576 314 673 l 215 719 314 885 220 598 314 702 220 747 314 926 225 621 314 733 225 778 314 970 230 646 314 767 230 811 314 1018 235 673 314 803 235 847 314 1069 240 702 314 842 240 885 314 1119 245 733 314 883 245 926 314 1171 250 767 314 928 1 250 970 315 1223 255 803 315 976 l 255 1018 320 1273 260 842 320 1028 l 260 1069 325 1326 265 883 325 1083  !

265 1119 330 1383 270 928 330 1143 270 1171 335 1445 275 976 335 1206 275 1223 340 1510 280 1028 340 1275 280 1273 245 1580 285 1083 345 1348 285 1326 350 1656 290 1143 350 1426 290 1383 355 1736 295 1206 355 1510 295 1445 360 1822 300 1275 360 1599 300 1510 365 1914 305 1348 365 1695 305 1580 370 2013 310 1426 370 1798 310 1656 375 2118 315 1510 375 1908 315 1736 380 320 1599 380 2025 320 1822 385 2231 2351 l 325 1695 385 2150 325 1914 330 1798 390 2283 I 330 2013 335 1908 395 2425 l 335 2118 340 2025 340 2231 345 2150 4

345 2351 350 2283 355 2425 Table 2-1 Farley Unit 2 36 EFPY Heatup Curve Data Points (Without Margins for Instrument Errors)DI

! FARLEY UNIT 2 4 AEVISION O l

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1 PRESSURE TEMPERATURE LIMITS REPORT '

0 *F  : 20 *F 40*F 60*F 100 *F T l P T l P T l P T l P T P l

60 492 60 455 60 416 60 377 60 296 65 495 65 457 65 419 65 380 65 298 70 498 70 460 70 422 70 383 70 301 75 501 75 463 75 425 75 386 75 305 80 505 80 467 80 428 80 389 80 308 85 508 85 471 85 432 85 393 85 312 90 512 90 474 90 436 90 397 90 317 ,

95 516 95 479 95 440 95 402 95 321 100 521 100 463 100 445 100 406 100 327 105 525 105 488 105 450 105 412 105 332 110 531 110 494 110 456 110 417 110 338 i 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 12 376 155 561 155 529 155 495 155 460 155 389 160 561 160 541 160 507 160 472 160 403 165 561 165 553 165 519 165 486 165 418 170 561 170 561 170 533 170 500 170 434 175 561 175 561 175 548 175 516 175 452 180 561 180 561 180 561 180 533 180 471 180 626 180 595 .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 21 5 715 220 723 220 688 220 786 220 764 220 743 225 756 225 725 225 813 225 793 225 774 230 792 230 766 230 842 230 824 230 807 235 830 235 809 235 874 235 858 235 843 240 871 240 856 '

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 1

325 2062 '

330 2182 335 -2309 Table .2-2 Farley Unit 2 36 EFPY Cooldown Curve Data Points (Without Margins for Instrument Errors)DI J

FARLEY UNIT 2 5 AEVISION 0

l PRESSURE TEMPERATURE LIMITS REPORT 3.0 Reactor Vessel Material Surveillance Program l 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 W , 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 B7212-1 as determined by impact testing, or (2) the predicted shift in reference temperature for weld seam Il-923 as determined by Regulatory Guide 1.99, Revision 2.

l Figures 2-1 and 2-2 will be updated prior to the adjusted reference temperature (ART) for the

, limiting beltline material exceeding the ART used to generate the Pfr limits for the specified l fluence periods.

- Table 3-1 l

SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE

  • Capsule Capsule Location Lead Removal Fluence")

2 j (Degree) Factor EFPY *) (n/cm )

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! U'd' 343 3.32 1.10 5.79 x 10

W (d' 110 2.86 3.97 1.54 x 10

X) M 287 3.40 6.41 2.64 x 10

V* 107 3.09 16.7 5.57 x 10" j Z 340 2.67 Standby --

Y 290 2.67 Standby -

NOTES:

(a) WCAP-14689l 'l (b) Effective Full Power Years (EFPY) from plant startup (c) Fluence recalculated using methodology contained in WCAP-14040-NP-A, Revision 2.I'l (d) Plant-specific evaluation (c) Final capsule withdrawal required by ASTM E 185-82.

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FARLEY UNIT 2 5 REVISION 0 I 4 l

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PRESSURE TEMPERATURE LIMITS REPORT s

} 4.0 Reactor Vessel Surveillance Data Credibility Regulatory Guide 1.99, Revision 2, describes general procedures acceptable to the NRC staff for l t

q 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 I methodology for calculating the adjusted reference temperature and Charpy upper-shelf energy of r

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 rc ictor in question.

To date, there have been three surveillance capsules removed from the Farley Unit 2 reactor vessel.

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

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The purpose of this evaluation is to apply the credibility requirements of Regulatory Guide 1.99, l

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; e Lower shell plates B7210-1 and B7210-2; e Intermediate shell longitudinal weld seams19-923 A, heat number HODA;

  • Intermediate shell longitudinal weld seams19-923 B, heat number BOLA; e Lower shell longitudinal weld seams20-923 A & B, heat number 83640, Linde 0091 flux, flux lot 3490; and e Circumferential weld Il-923, heat number SP5622, Linde 0091 flux, flux lot i122.

l FARLEY UNIT 2 6 REVISION 0

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PRESSURE TEMPERATURE LIMITS REPORT PI Per WCAP-8956 , 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 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 wasjudged 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 j clectrodes were not copper-coated and do not exhibit the chemical variabihty found in copper- i 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 perfomied on the l surveillance weld are considered to be representative of the property changes expected in the I 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 appmving the selected surveillance weld material based on the NRC evaluation provided in Section 5.2.1 of NUREG-Oll7. M Although the Farley Unit 2 surveillance weld material does not meet the requirements of Cnterion l

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.

FARLEY UNIT 2 7 REVISION 0 b

I 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 permit the determination of the 30 ft-lb temperature and upper shelf energy, unambiguously.

Plots of Charpy energy versus temperature for the unirradirled condition are presented in WCAP-8956 (". 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 14. W l'l, and X 11 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 2 surveillance program meets the requirements of Criterior 2.

Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of ARTm 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 E185-82.

The least squares method, as described in Regulatory Position 2.1 will be utilized in determining a best-fit line for this data to determine if this criterion is met.

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FARLEY UNIT 2 8 REVISION 0 2

PRESSURE TEMPERATURE LIMITS REPORT I

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

Matenal Fm AR a F, Capsule ATw (xv)

U 0.579 0.847 103 87.2 0.718 Intermediate Shell Plate B7212-1 W l.54 1.12 165 184.7 1.25 (Longitudinal)

X 2.64 1.26 180 226.8 1.59 U 0.579 0.847 133. 112.7 0.718 Intermediate Shell Plate B7212-1 W l.54 1.12 165 184.7 1.25 X 2.64 1.26 190 239.3 1.59 6.45 936 1035.4 7.12 Weld Metal U 0.579 0.847 10 8.5 0.718 J

W l.54 1.12 10 11.2 1.25 X 2.64 1.26 10 12.6 1.59 3.23 30 32.3 3.56 l

l 1 NOTES: ,

(a) WCAP-146891'l l (b) F = Fluence (10 n/cm 2, E > 1.0 MeV)

(c) FF = Fluence Factor = F"' ' '*"

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FARLEY UNIT 2 9 REVISION O a

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W+ + 4 PRESSURE TEMPERATURE LIMITS REPORT Per the 27* Edition of the CRC Standard Mathematical Tables (page 497), for a straight line fit by the method ofleast squares, the values b, and b, are obtained by solving the s'ormal equations nho + b,[x, = [y, and h[x,+b,[x,2 = [ x,y, o

These equations can be re-written as follows:

N N

[y, = an + b[ x,

,.i .. i and f, x,y, = af, x, + bf, x,'

Intermediate Shell Plate B7212-1:

Based on the data provided in Table 4-1, these equations become:

936 = 6a + 6.45b and 103 5.4 = 6.45a + 7.12b Thus, b=156.8 and a=-12.5, and the equation of the straight line which provides the best fit in the sense ofleast squares is:

Y' = 156.8 (X) - 12.5 The error in predicting a value of Y corresponding to a given X value is e = Y - Y'.

TABLE 4-2 SCATTER OF ARTer VALUES ABOUT A BEST-FIT LINE FOR SURVEILLANCE PLATE MATERIAL"

Intermediate Shell ARTer Best Fit ARTe7 Scatter of ARTer Plate B72121 FF (30 ft-lb) ('F) ( F)

Orientation (*F) 0.847 103 120.3 -17.3 Longitudinal 1.12 165 163.1 1.9 1.26 180 185.1 -5.1 0.847 133 120.3 12.7 Transverse 1.12 165 163.1 1.9 1.26 190 185.1 4.9 NOTES:

(a) WCAP-146891 'l FARLEY UNIT 2 10 REVISION 0

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

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PRESSURE TEMPERATURE LIMITS REPORT i The scatter of ARTer values about a best-fit line drawn, as described in Regulatory Position 2.1, i should be less than 17 F for base metal. However, even if the fluence range is large, the scatter j should not exceed twice this value (i.e.,34 F). As shown above, the error is within 34 F of the best-  !

j fit line. Therefore, Criterion 3 is met for the Farley Unit 2 surveillance plate material.

Weld Metal ,

Based on the data provided in Table 4-1, the equation becomes:

30 = 3a + 3.23b

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! and  ;

32.3 = 3.23a + 336h Thus, b=0 and a=10, and the equation of the straight line which provides the best fit in the sense of least squares is:

Y' = 0 (X) + 10 4  !

The error in predicting a value of Y corresponding to a given X value is e = Y - Y'.  !

$ TABLE 4-3 SCATTER OF ARTer VALUES ABOUT A BEST-FIT LINE '

FOR SURVEILLANCE WELD MATERIAL"'

I ARTm7 Best Fit ARTsur Seatter of ARTm1 Material FF (30 ft-lb) (*F) (*F)

(*F) -

0847 10 10 0 Weld Metal 1.12 to 10 0 l.26 10 10 0 NOTES:

(a) WCAP-14689N J The scatter of ARTer values about a best fit line drawn, 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 cladding / base metal interface within i25 F.

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FARLEY UNIT 2 11 REVISION O

i PRESSURE TEMPERATURE LIMITS REPORT  !

The Farley Unit 2 capsule specimens are located in the reactor between the neutron shieldmg 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, Criterion 5 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 is credible.

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 ;.urveillance 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 generation of the heatup and cooldown curves and the PTS evaluation.  ;

I Table 5-5 provides a summary of the adjusted reference temperatures (ARTS) of the Farley Unit 2 i reactor vessel beltline materials at the 1/4-T and 3/4-T locations for 36 EFPY Table 5-6 shows the calculation of the ART at 36 EFPY for the limiting Farley Unit 2 reactor l vessel material (intermediate shell plate B7212-1).

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

l I

i FARLEY UNIT 2 12 REVISION O

PRESSURE TEMPERATURE LIMITS REPORT Table 5-1 COMPARISON OF SURVEILLANCE MATERIAL 30 IT-LB TRANSITION TEMPERATURE SHIFT AND UPPER SHELF ENERGY DECREASE WITH REGULATORY GUIDE 1.99, REVISION 2,

, PREDICTIONS

  • 30 fl-lb Transition Uppes Shelf Energy Fluence Temperature Shift Decrease Material Capsule (x 10"n/cm 2, Predicted Measured Predicted Measured E > 1.0 MeV) (*F) (*F) (%) (%)

U 0.579 126 103 25.5 27.7 Plate B7212-1 (Longitudinal) W l.54 167 165 32 21.5 X 2.64 188 180 37 27.7 U 0.579 126 133 25.5 27.0 Plate B7212-1 (Transverse)

W l.54 167 165 32 20.0 X 2.64 188 190 37 27.0 0 0.579 35 to 16.5 8.3 Weld Metal W 1,54 46 10 21 0 1 X 2.64 52 10 24 0 U 0.579 -- 58 -- 29.7 l I

HAZ Metal W l.54 --

109 -- 20.3 X 2.64 - 110 -

20.3 NOTES:

(a) WCAP 14689 01 FARLEY UNIT 2 13 REVISION 0 a

PRESEURE TEMPERATURE LIMITS REPORT .

{ Table 5-2

CALCULATION OF CHEMISTRY FACTORS USING SURVEILLANCE CAPSULE DATA *

, Material Capsule f* FF* ARTm FF

  • ARTmyr FF 2

I U 0.579 intermediate Shell 0.847 103 87.2 0.718 i

! Plate B72121 W l.54 1.12 165 184.7 1.25 (Longitudinal) X 2.64 1.26 180 226.8 1.59 l l . Intermediate Shell U 0.579 0.847 133 112.7 0.718 i

Plate B7212-1 W l.54 1.12 165 184.7 1.25 ,

(Trannerse) X 2.64 1.26 190 239.3 1,59

. Sum: 1035.4 7.12 Chemistry Factor = I (FF

  • ARTmyr) + I (FF 2) = 145.5 U 0.579 0.847 to 8.5 ' O.718 Weld Metal W- 1.54 1.12 10 11.2 1.25 X 2.64 1.26 10 12.6 1.59  ;

U Sum: 32.3 3.56 I i

2

, Chemistry Factor = E (FF

  • ARTmn) + I (FF ) = 9.1 .

2 i

i 4

NOTES:

(a) WCAP 146891 'l l (b) f = 11uence (x 10" n/cm2, E > 1.0 MeV)

(c) FF = fluence factor = f' 28' "

i 1

l I

FARLEY UNIT 2 14 REVISION 0 1

3 PRESSURE TEMPERATURE LIMITS REPORT Table 5-3 REACTOR VESSEL TOUGliNESS TABLE (UNIRRADI ATED)"'

Beltline Material Cu Weight % Ni Weight % IRTm (*F)

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 Surveillance Plate *' O.20 0.60 10

' Lower Shell Plate B7210-1 0.13 0.56 18 Lower Shell Plate B7210-2 0.14 0.57 10 Inter. Shell Longitudinal Weld Seam 19-923 A ") 0.02 0.96 -56 (Heat # HODA)

Inter. Shell Longitudinal Weld Seam 19-923 B ") 0.03 0.91 -60 (Heat # BOLA)

Surveillance Weld (d' O.03 0.91 -60 Circumferential Weld Scam 11-923 "'

0'14 0~07 40 (Heat # 5P5622)

Lower Shell Longitudinal Weld Seams20-923 A & B "' O.05 0.07 70 (Heat # 83640)

NOTES.

(a) WCAP-14689 03 (b) The surveillance plate is representawve ofintermediate shell plate B7212-1 (c) Generic Letter 92-01, Revision 1, Supplement I response (d) The surveillance weld is representnive of intermediate shell longitudinal weld 19-923B 1

.l FARLEY UNIT 2 15 REVISION 0 2

j

PRESSURE TEMPERATURE LIMITS REPORT Table 5-4 REACTOR VESSEL FLUENCE PROJECTIONS FOR 36 EFPY "

EFPY O' 15' 15"'" 30 30 '" 45 36 3.75 2.23 1.78 1.70 1.64 1.20 NOTES:

(a) WCAP-14689 l4 2

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

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

Table 5 5

SUMMARY

OF ADJUSTED REFERENCE TEMPERATURES (ARTS) FOR REACTOR VESSEL BELTLINE MATERIALS AT THE 1/4 T AND 3/4 T LOCATIONS FOR 36 EFPY M Material 1/4-T ( F) 3/4-T (*F)

Intermediate Shell Plate B7203 1 172 146

]

Intermediate Shell Plate B7212-1 207 169 Intermediate Shell Plate B7212-1 186 'd' 149

Using S/C Data"'

Lower Shell Plate B7210-1 162 139 Lower Shell Plate B7210-2 165 140 Intermediate Shell Longitudinal Weld Seam i g <.) 0 19-923 A (Heat # HODA)

Intermediate Shell Longitudinal Weld Seam 15 ) -6 ("19-923 B (Heat # BOLA)

Intermediate Shell Longitudinal Weld Seam 19-923 B (Heat # BOLA) -44 '" -4 7 '"

Using S/C Data Circumferential Weld Il-923 99 81 (Heat # $PS622)

Lower Shell Longitudinal Weld Seams 20- ,7 i.' -25

923 A & B (Heat # 83640) j NQTES: 1 (a) WCAP-1468914 )

(b) The ARTS presented here are based on the peak reactor vessel surface fluence of 3.75 x 10 n/cm2 (E > j 1.0 MeV) unless otherwise noted.  ;

(c) Based on surveillance capsule data contained in WCAP-12471 l21 (d) ART values used to generate heatup and cooldown curves.

2 (e) ARTS calculated using the peak vessel fluence of 1.20 x 10 n/cm (E > 1.0 MeV) at 45' FARLRY UNIT 2 16 REVISION 0 j

PRESSURE TEMPERATURE LIMITS REPORT Table 5-6 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE AT 36 EFPY FOR THE LIMITING REACTOR VESSEL MATERIAL - INTERMEDIATE SHELL PLATE B7212-1 "'

Parameter Operating Period 36 EFPY Location 1/4-T 3/4-T Chemistry Factor, CF ( F) 145.5 145.5 Fluence, f (10" n/cm2 ) *' 2.34 0.909 Fluence Factor, FF 1.23 0.973 ARTer = CF x FF ('F) 179 142 Initial RTer, I ('F) -10 -10 Margin, M (*F) "' 17 17 Adjusted Reference Temperature (ART), ( F) per Regulatory Guide 1.99, Revision 2 186 149 NOTES:

(a) WCAP-14689IU 2

(b) Fluence is based on f ,r(10" n/cm . E > 1.0 MeV) = 3 75 at 36 EFPY. The Farley Unit 2 reactor vessel wall thickness is 7,875 inches in the beltline region.

2 (c) Margin is calculated as M = 2(o,2 # ya )o s The standard deviation for the initial RTer margin term, e,, is 0 F since the initial RTer is a measured value. The standard deviation for the ARTer term, o.s, is 17"F for the plate, except that a3 need not exceed 0.5 times the mean value of ARTer.

In accordance with Regulatory Guide 1.99, Revision 2, Position 2.1, values of ca may be cut in half when based on credible surveillance data.

1 i

FARLEY UNIT 2 17 REVISION 0 2

PRESSURE TEhiPERATURE LIMITS REPORT J

i Table 5-7 PPISSURIZED THERhiAL SHOCK (RTns) VALUES FOR 36 EFPY"'

4 Surface Fluence ART" M Material CF 2 (10 n/cm . FF (CF x FF) 1 RT" E > 1.0 MeV) ( F)

  • F) (B (

Intermediate Shell 100.0 3.75 1.34 134.2 15 34 183 Plate B72031 Intermediate Shell Plate B7212-1 149.0 3.75 134 200.0 -10 34 224 P

Intermediate Shell Plate B7212-1 145.5 3.75 1.34 195.3 -10 17 202 Using S/C Data Lower Shell Plate B7210-1 89.8 3.75 1.34 120.5 18 34 173 Lower Shell Plate B7210-2 98.7 3.75 1.34 132.5 10 34 176 Intermediate Shell Longitudinal Welds19-923 A 27.0 1.20 1.05 28.4 -56 44.3- 17 (Heat # HODA) l Intermediate Shell Longitudinal Welds19-923 B 41.0 1.20 1.05 43.1 -60 43.1 26 (Heat # BOLA)

Intermediate Shell Longitudinal Welds l 19-923 B 9.1 1.20 1.05 9.6 -60 9.6 -41 (Heat # BOLA)

Using S/C Data Circumferential Weld Il 923 67.3 3.75 1.34 90.3 -40 56 106 (Heat # 5P5622)

Lower Shell Longitudinal Welds 34.05 1.20 1.05 35.8 -70 35.8 2 20-923 A & B (Heat # 83640) l i

NOTES (a) WCAP-14689 14 FARLEY UNIT 2 18 REVISION 0 2

PRESSURE TEMPERATURE LIMITS REPORT 4

6.0 References

1. WCAP-14689, Revision 1, Farley Units I and 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation, E. Terek, April 1997.

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 al., 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 al., 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 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 Heatup and Cooldown Limit Curves, January 1996.

FARLEY UNIT 2 19 REVISION 0 s

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