ML20197B639

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Rev 1 of Jfnp - Unit 1 Pressure Temperature Limits Rept
ML20197B639
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Site: Farley Southern Nuclear icon.png
Issue date: 12/18/1997
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SOUTHERN NUCLEAR OPERATING CO.
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ML20197B622 List:
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NUDOCS 9712240015
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Enclosure 2 Joseph M. Farley Nuclear Plant - Unit 1 Pressure Temperature Limits Report 4

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i PRESSURE TEMPERATURE LIMITS REPORT

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Table Of Contents ust or Tables . _ . . _ . .. __ .. iii ust or ngurn .~ . ._.. .~. . .. iv l l

e 1.0 R(.S Presnre Temperature Limits Report (PTLR) ~ .~.-~.~.-~~~~~~~~~~~~~~~l  ;

2.0 Operating Limits. . . . ..... .. . .. . ... . . . ...I .

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.- 2.1 RCS Pressure / Temperature (P/T) Limits (LCO . 3.4.10.8)......... . .. . ..... .. ~ . . . . .. /

2.2 RCP Operation Limits.. . .. .. .. . ... .. . . .. .........................................._..../

3.0 Heactor Vessel Maiertal Surtelilanee Program .. . ..~. ..-~.--.. m. . .... . 6 ,

4.0 Reactor Vessel Surveillanee Data CredIbllity . . . . .... . . .. . . . . ..7

$.0 Suppiemental Data Tables ..... . . . .. . . ..... .!2

- 6.0 Re te re ne es . . ... ..... ..... .. . .. ...... . ......... . . . . . 19  ;

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- FARLEY UNIT I - il . REVISION 1 g , - e w ,p ,r y.--% - -yavm-

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PRESSURE TEMPERATURE LIMITS REPORT List of Tables -

t 21 Fariey Unit i 21.9 EFPY 11eaiup Curve Data Polnt: .~... - ~ ~ -... .. ~ .-. 4 22 Farley Unit 121.9 EFPY Cooldown Cun e Data Polnts.~ .~. ~. .~. ... . .. .~5 ,

t 3.I Surveillance Capsule W1ihdrmwal Schedule . .. . ..... . ... .. 6 41 - Surveillance Capsule Data Calculation of Best. Fit Line as Described in Position 2.1 of Regulatory Gulde 1.99, Res islon 2..... . ~ .-. ~ ---..~ ~.. ... .. 10 42 Scatter of ARTsor Values About a Best. Fit Line for Suneillance Plate Material ... . ....i t 43 Seatter of ART,ot Values About a Best. Fit Line for Surveillance Weld Material.... . it 51 Comparison of Surveillance Material 30 FT.LB Transiflon Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Resision 2. Predictions.. .13 52 Calculation of Chemistry Factors Using Surveillance Capsule Data... . ............ . . .14  !

53 Hencter Yessel Toughness Table (Unirradlated).. ... .. ... . . ..... ... .... .... ..I5 j 54 Heaetor Yessel FIuence Projections ... .. .......... .. . .......... .. . ....... . ..............I 5 5-$ Summary of Adjusted Reference Temperatures (ARTO for Reactor Vessel Beltline Materials at the 1/4.T and 3/4.T Locations for 21.9 EFPY .. .. .. . .. . ... 16 1

56 Calculation of Adjusted Reference Temperature at 21.9 EFPY for the 1.imiting Reactor Vessel Matertal . Lower Shell Plate B69l91..... . -~.. . .~o.. o- ..~.... ..~o..!7 57' Pressurired Thermal Shock (RTets) Values for 36 EFPY.. . ... . .... .'.. . ... . .. I8

.FARLEY UNIT I iii REVISION 1

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t PRESSURE TEMPERATURE LIMITS REPCGT .i i

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21 ~ . Farley Unit i Reacter Coolani Sptem Heatup Limitatlens ... m .. . 2

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.22 Farley Unit i Reactor Coolant Sptem Cooldown Limitations . .. J r

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PRESStJRE TEMPERATURE LIMITS REPORT =

1.0 RCS Pressure Temperature Limits Report (PTLR)

This PTLR for Farley Nuclear Plant Unit I has been prepared in accordance with the requiremed of Technical Specification (TS) 6.9.1.15. Revisions to the PTLR shall be provided to the NRC aller issuance.

This report affects TS 3.4.10.I, RCS Pressurefremperature Limits (P/T) Limits. All TS 1 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 and were developed using the methodologies specified in TS 6.9.1.15 . The operability requirements associated with LTOP are specilled in TS LCO 3.4.10.3 and were detennined 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/r 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 heatup 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 insenice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

2.1.3_ The RCS P/T limits for heatup and cooldown are sper!fied 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 bss 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.

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- FARLEY UNIT 1 i REVISION 1 -

PRESSURE 1LMPERA1URE LIMITS REPORT 2.500 Leak Test Linut r i i Cntecality unut for l

60 F/rv Heatup l l 2.250 Umtary Matonal. < i Lower Shen Plate B6919-1 -

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Umtsg ART Values at 219 EFPY:

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250 \ Hyorost49c Test Tomgerature (269 0 j for the Eerwce Paraf up to 219 EFPY

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

Figure 21 Farley Unit i Reactor Coolant System lleatup Limitations (IIcatup Rates up to 100*F/hr#

Applicable to 21.9 EFPY (adjusted to include 60 psi AP at RCS temperatures 2 Il0'F and 25 psi AP for RCS temperatures < llo'T). Includes vessel flange requirements of 180*F and 56i psig per 10 CFR 50, Appendis G.DI TARLEY UNIT I 2 REVislON I

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- Farley Unit i Reactor Coolant System Cooldow n Limitations (Cooldown Rates up to 100'F/hr)

Applicable to 21.9 EFPY (adjusted to include 60 psi AP at RCS temperatures 2 Il0*F and 25 psi AP for
  • RCS temperatures < 110'F), includes vessel flange requirements of 180*F and 561 psig per 10 CFR 50, Appendis O.l4 FARLEY UNIT I ')

. REVISION 1

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PRESSURE TEMPERA 1URE 1.lMITS REPORT 60 *t 60 *f Cnticahty Limit 100 *t 100 *f Cntwahty Limit Leak lest [

1 P 1 P 1 P 1 l P T l P l l l 70 00 289 0 70 457 269 0 267 2000 .

289 2485  !

75 490 289 480 75 457 289 481 289 466  ;

00 490 289 469 60 457 85 490 289 462 85 457 289 454 90 490 289 457 90 457 289 444 ,

95 490 289 455 95 457 289 436 100 493 289 456 100 457 289 430 105 491 289 458 105 457 289 426 .

110 493 289 461 110 457 289 423.

110 458 289 466 110 422 289 422 i

115 461 289 473 115 422 289 423 120 466 289 480 120 422 289 424 125 473 289 488 125 423 289 427 130 480 289 498 130 424 289 432 135 488 289 509 135 427 289 437 140 498 289 $21 140 432 289 443 145 509 289 534 145 437 289 451 150 $21 289 547 150 443 209 460 155 534 289 $63 155 451 289 469 160 547 289 579 160 460 289 481 .

165 561 289 597 165 469 28D 493 170 561 299 617 170 481 289 506 175 561 289 637 175 493 289 521 180 561 289 660 180 506 289 537 180 617 289 684 185 021 289 555 185 637 289 710 190 $37 289 574 190 660 289 738 195 555 289 595

  • l 195 664 289 768 200 574 289 617 200 710 289 800 205 505 289 641 205 734 289 835 21C 617 289 668 210 768 289 873 215 641 289 696 215 800 289 913 220 668 289 726 -

220 835 289 956 225 696 289 759 225 873 289 1002 230 726 289 794 230 913 289 1051 235 759 289 832 235 956 290 1105 240 794 290 872 240 1002 295 1162 245 832 295 916 245 1051 300 1223 250 872 300 963 250 1105 305 1288 255 916 305 1013 255 1162 310 1348 260 963 310 1064 260 1223 315 1434 265 1013 315 1125 265 1288 320 1515 270 1068 320 1188 270 1358 325 1601 275 1125 325 1254 275 1434 330 1694 280 1188 330 1326 280 1515 335 1793 285 1254 335 1402 285 1601 340 1899 290 1326 340 1484 290 1604 345 2012 295 1402 345 1572 295 1793 350 2120 300 1484 350 1666 300 1899 355 2233 305 1572 355 1767 305- 2012 360 2354 310 1666 360 1874 310 2120 315 1767 365 1988 315 2233 320 1874 370 2111 320 2354 325 1988 375 2241 330 2111 380 2380 335 2241 340 2380 Table 21 Farley Unit i 21.9 EFPY IIcatup Curs e Data Points pdjusted to include 60 psi AP at RCS temperatures 2 Il0'F and 25 psi AP for RCS temperatures < 110'F)l4 FARLEY UNIT I 4 REVISION I

PRESSURE TEMPERATURE LIMITS REPORT 0 'l' 20 *t 40*F 60 *f 100 't 1 P T P T P 1 l P 1 l P l l l 70 516 70 479 TO 442 TO 404 70 326 75 521 75 4 54 75 447 -75 409 75 331 80 525 80 489 80 452 80 414 80 337 85 531 85 4 94 85 457 85 420 85 343 90 536. 90 500 90 463 90 426 90 349 95 542 95 506 95 470 95 433 95 357 100 549 100 513 100 477 100 440 100 365 105 556 105 $20 105 464 105 448 105 373 110 561 110 528 110 492 110 456 110 382 110 528 110 493 110 457 110 421 110 347 115 536 115 501 115 466 115 t" 115 357 120 545 120 510 120 475 120 120 368 125 554 125 520 125 485 125 da. 125 380 130 561 130 530 130 496 130 462 130 393 135 561 135 541 135 508 135 475 13f 407 140 561 140 $$3 140 521 140 488 140 422-145 561 145 561 145 534 145 502 145 438 150 561 150 561 150 549 150 518 150 455 155 561 155 561 155 561 155 535 155 474 160 561 100 561 160 561 160 553 100 494 165 561 165 561 165 561 165 561 165 516 170 561 170 561 170 561 170 561 170 540 175 561 175 561 175 561 175 561 175 561 180 561 180 $61 180 561 180 561 180 $61 180 715 iB0 689 180 664 180 640 180 593 185 737 185 713 185 689 185 666 185 623 190 760 190 738 190 716 190 694 190 655 195 786 195 765 195 744 195 725 195 690 200 813 200 794 200 775 200 758 200 727 205 842 205 825 205 808 205 793 205 767 210 874 210 858 210 844 210 831 210 811 215 908 215 894 215 882 215 872 215 857 220 944 220 933 220 923 220 916 220 908 225 983 225 974 225 968 225 963 225 962 230 1025 230 1019 230 #015 230 1014 230 1020 235 1070 235 1067 235 1066 235 1069 240 1119 240 1118

-245 1171 250 1226 255 1206 260 1351 265 1420 270 1494 275 1573 280 1658 285 1749 290 1846 295 1951 300 2062 305 2182 310 2309 Table 2 2

- Earley Unit i 21.9 EFPY Cooldown Curve Data Points (adjusted to include 60 psi AP at RCS temperatures 2 Il0'F and 25 psi AP for RCS temperatures < 110*F)l4 FARLEY UNIT I 5 REVISION I

PRESSURE 1EMPERA1URE LIMITS REPORT 3.0 Reactor Vessel Material Surveillance Program The reactor vessel material surveillance program is in compliance with 10 CFR 50, Appendix II, and is described in Section 5.4.3.6 of the Farley FSAR.1he removal schedule is provided in Table 31. The results of these examinations shall be ased to update Figures 21 and 2 2 if the results indicate that the adjusted reference temperature (ART) for the limiting beltline material exceeds the ART used to generate the P/T limits shown in Figures 21 and 2 2 for the specified fluence period.

Table 31 SURVEILLANCE CAPSULE W1111DRAWAL SCilEDULE "'

Capsule Capsule Location Lead Removal Fluence (Degree) ractor EfPY*' (n'em')

Y"' 343 3.11 1.13 6.42 x 10

U "' 107 3.18 3.02 1.81 x 10" X "' 287 3.30 6.12 3.24 x 10

W "' i10 3.02 12.43 5. ! 7 x 10

V 290 3.02 Standby -

340 3.02 Standby --

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bESl (a) WCAP.14689, Revision 2 I4 (b) Effective Full Pow er Years (ErPY) from plant stanup (c) Plant-specifie eva'2ation i

l FARLEY UNIT 1 6 REVISION 1

PRES $URE 10MPERATURE 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 raoiation embrittlement of the low alloy steels cunently 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 suiteillance capsule data. The methods of Position C.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.

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

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

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

Criterion 1: hfaterials 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 0 to 10 CFR 50, Fracture Toughness Requirements, December 19,1995, to be:

the reactor vessel (shell material including w elds, heat affected rones, 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 matenal with regard to radiation damage.

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

e Intennediate shell plates B6903 2 and 116903 3; e Lower shell plates B69191 and 116919 2; e Intermediate shell longitudinal weld seams 19 894 A & D, heat number 33A277, Linde 1092 flux, ilux lot 3889; e Lower shell longitudinal weld seams 20 894 A & D, heat number 90099, Linde 0091 flux, Oux lot 3977; and e Circumferential weld I l 894, heat number 632% /, Linde 0091 flux, aux lot 3999.

FARLEY UNIT I 7 REVISION I

PRESSURE TEMPERA 11)RE LIMITS REPORT Per WCAP 88105, the Unit I suneillance 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.

Therefore, at the time the Farley Unit I surveillance capsule program was developed, lower shell plate H6919 1 was judged to be most limiting based on the above recommendations and was utilf red in the sun *eillance program.

The surveillance prograrn weld for Farley Unit I was fabricated using the same heat of weld wire used to fabricate the middle shell axial seams 19 894 A & B (heat 33A277). The results of mechanical propeny tests performed on the surveillance weld are considered to be representative of the propeny changes expected in the reactor vessel beltline seams.

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

Criterion 2: Scatter in the plots of Charpy energy versus temperr*n for the irradiated and unirradiated conditions should be amall enough to permit the determination of the 30 fi lb temperature and upper shelf e.e.,;y, unambiguously.

Plots of Charpy energy versus temperature for the unirradiated condition are presented in the Unit I reactor vessel surveillance program description contained in WCAP 8810 W.

Plots of Charpy energy versus temperature for the irradiated conditions are presented in the reactor vessel surveillance capsule repons for capsules Y W, UW, XW, and WW.

Based on engineering judgment, the scatter in the data presented in these plots is small enough to detennine the 30 f1 lb temperature and upper shelf energy of the Farley Unit I surveillance materials unambiguously. Therefore, the Farley Unit I surveillance program meets the requirements of Criterion 2.

FARLEY UNIT l' 8 REVISION 1

I PRESSURE 10MPERA1URE LIMITS REPORT f

Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter ,

ofARTm 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 i

the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fall this criterion for use in shift >

calculations, they may be credible for determining decrease in upper shelf energy I if the upper shelf can be clearly determined, following the definition given in l 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. l (Continued on the following page) i l

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w FARLEY UNIT I 9 REVISION I t

l PRESSURE TEMPERATURE LIMITS REPORT Table 4 ;

1 SURVEILLANCE CAPSULE DATA C ALCULATION OF BEST FIT LINE AS DESCRIBED IN l POSITION 2.1 OF REGULATORY GUIDE I.99, REVISION 2 '"  !

Material Capsule F*'

Fi AR or 3 sot { )

Y 0.642 0.876 85 74.5 0 767 Lower Shall Plate B69191 U l.81 1.16 105 121.8 1.35 I " '" "*'I X 3.24 1.31 135 176.9 1,72 W 5.17 1.41 155 218.6 1.99 Y OM2 0.876 $$ 48.2 0.767 Lower Shell Plate 1169191 U l.81 1.16 90 104.4 1.35

"'*"' * #* I X 3.24 1.31 105 137.6 1.72 W 5.17 1.41 145 204.5 1.99 1086.5 11.65 CF = I(FF

  • ARTuot) + I(FF ) = 93.3'F Weld Metal Y 0.642 0.876 80 70.1 0.767 U l.81 1.16 80 92.8 1.35 X 3.24 1.31 100 131.0 1.72 W 5,17 1.41 95 134,0 1.99

$ 427.9 5.83 CF = I(FF

(a) -WCAP 14689, Revision 2 M (b) F = Fluence (10" tvem#, E > l.0 MeV)

  1. "N (c) FF = Fluence Factor = F FARLEY UN!T 1 10 REVISION I 1

PRES $URE TEMPERATURE LIMITS REPORT i

l 18ble 4 2 SCATTEROF ARTsut VALUES ABOUT A DEST FIT LINE FOR SURVEILLANCE PLATE MATERIAL"'

Lower Shell Plate ARl'up; Best fit ARTsor Scatter of ARTN ot B6919 1 FF (30 ft-lb) ('F) (*F)'

Orientation (*F) 0.876 85 81.7 3.3 Longitudinal I.16 10$ 108.2 3.2 1.31 13$ 122.2 12.8 l.41 15$ 131.6 23.4

) 0.876 $$ 81.7 26.7 1.16 90 108.2 18.2 Transverse 1 31 10$ 122.2 17.2 1.41 145 131.6 13.4 NOTES:

(a) WCAP.14689, Revision .'IU

'the scatter of ARTuor values about a best fit line dr ;xn with the y. intercept equal to zero, as described in Regulatory Position 2.1, should be less than 17'F for base metal. As shown above, the scatter of four of the data points are not within 17'F of the best fit line. Therefore, this criteria is not met for the Parley Unit I surveillance plate material. Since all of the data is not within 17'F ,

of the best fit line, SNC has chosen to use the CF from this surveillance data along with a a3 of 17'F when predicting the Farley Unit i vessel ptoperties.

Table 4 3 SCATTER OF ARTunt VALUES ABOUT A BEST FIT LINE FOR SURVEILLANCE WELD MATERIAL"'

ARTsor (30 f t Ib') Best Fit ARTwog Scatter of ARTsor Material FF

(*F) ('F) (*F) 0.876 80 64.3 15.7 Weld Metal 1.16 80 85.1 5.1 .

1.31 100 96.2 3.8 1.41 95 103.5 8.5 NOTES:

(a) WCAP.!4689, Revision 2(4 FARLEY UNIT I 11 REVISION 1 -

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PRESSURE TEMPERAWRE LIMITS REPORT lhe scatter of ARTsnt values about a best fit line drawn with the y intercept equal to zert, as described in Regulatory Position 2.1, is less than 28'F as shown above. Therefbre Criterion 3 is met for the Farley Unit I surveillance weld material.

Criterion 4: 'Ihe irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding / base metal interface within 25'F.

The Farley Unit I capsule specimens are located in the reactor between the neutron shielding pads and the vessel wall and are positioned opposite the center of the core. The test capsules are in guide tubes attached to the neutron shielding pads. The location of the specimens with respect to the reactor vessel 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 surveillance program meets the requirements of Criterion 4.

Criterion 5: The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the data base for that material.

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

COliCLUS10ft flased on the preceding responses to the criteria of Regulatory Guide 1.99, Revision 2, Section D, and the applicatien of engineering judgment, the Farley Unit I surveillance plate material data is not credible and the Farley Unit I surveillance weld data is credible.

5.0 Supplemental Data Tables Table 51 contains a comparison of measured surveillance material 30 n lb transition temperature shins 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 unitradiated Farley Unit I reactor vessel toughness data.

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

Table 5 5 provides a summary of the adjusted reference temperatures ( ARTS) of the Farley Unit I reactor vessel beltline materials at the 1/4 T and 3/4 T locations for 21.9 EFPY.

Table 5 6 show s the calculation of the ART at 21.9 EFPY for the limiting Farley Unit I reactor vessel material (lower shell plate B6919 1).

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

FARLEY UNIT I 12 REVISION 1

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PRESSURE T0htPERA1URE LlhilTS REPORT i

r Table 31 l COMPARISON OF SURVEILLANCE htATERIAL 30 iT-Lil TRANSITION TEhtPERATURE SillFTS AND UPPER SilELF ENEROY DECREASES WITil REGULATORY GUIDE 1.99, REVISION 2, i

PREDICTIONS

t Upper S$cif Energy i 30 ft lb 1ransition Fluence Temperature Shift Decrease i Material Capsule (* 10n/cm'. Predicted Measured Predicted hicasured E > l.0 hieV)_ ('F) (*F) (%) (%) y Y 0.642 85.7 83 21 9 U l.81 113.7 105 27 21

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X 3.24 128.0 13$ 31 19 W 3.17 137.8 IS$ 34 22 0

Y 0,642 85.7 $$ 21 U l.81 113.7 90 27 9 p , 9,9,,

(Transverse)

X 3.24 128.0 10$ 31 11 W $.17 137.8 145 34 16 Y 0.642 68.4 80 25 13 [

Weld hietal U l.81 90.8 80 33 28 X 3.24 102.2 100 38 23 W $.17 110.0 95 42 26 Y 0.642 - 60 .. 1I IIAZ hietal U l.81 .. 120 - 26 X 2.24 125 .. 19 W $.17 - .. I10 14 ,

NO7 ES:

.(a) WCAP 14689, Revision 2 til  ;

FARLEY UNIT I 13 REVISION 1 4

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PRESSURE TEMPERATURE LIMITS REPORT .

Table $ i cal.CULATION OF CllEMISTP,Y FACTORS USING SURVEILLANCE CAPSULE DATA

  • Material Capsule f "" Fl"" ARTuoi FF
  • ARTso, FF Y 0.642 0.876 85 74.5 0.767 Lower Shell Plate 116919 1 O l 1.81 1.16 10$ 121.8 1.35 (Longitudinal) X 3.24 1.31 135 176.9 1,72 i

W $.17 1.41 155 218.6 1.99 Lower Shell Plate Y 0.642 0.876 $$ 48.2 0.767 B6919.l~ U l.81 1.16 90 104.4 1.35 (Transverse) X 3.24 1.31 10$ 137.6 1.72 W $.17 1.41 145 204.5 1.99 Sum: 1086.5 11.65 d Chemistry Factor (CF)-I(FF

  • ARTsor) + I(FF') = 93.3*F Y 0.642 0.876 l 29.6 113.5 0.767 U l.81 1.16 129.6 150.3 1.35 X 162,0 212.2 1,72 Weld Metal
  • 3.24 1.31 W 5.17 1.41 153.9 217.0 1.99 Sum: 693.0 $.83 Chemistry factor (CF) = I(FF
  • AR1soi) + I(FF ) = 118.9'F 4

NOTES; I

(a) WCAP 14689 Revision 2 "

(b) f = Duence (s 10 rt'em 3, E > 1.0 MeV)

(c) FF = Ouence factor = f *2"*

(d) ARTsorvalues were multiplied by a ratio factor of 1.62 (CF,a + CF n .,a = 126.2 + 78.1 = 1.62)

FARLEY UNIT I 14 REVISION 1

k PRESSURE TEMPERATURE LIMITS REPORT Table 5 3 REAClOR VESSEL TOUGHNESS TABLE (UNIRRADIATED)"'

Beltline Mad.al Cu Weight % Ni Weight % IRTsw ('F) r=

Closure llead Flange -- -- 60-Vessel Fuge - - 60 Intermediate Shell Plate B6903 2 0.13 0.60 0 Intermediate Shell Plate B6903 3 0.12 0.56 10 Lower Sheti Plate B69191 0.14 0.55 15 Lower Shell Plate B6919-2 0.14 0.56 5 Intermediate Shell Longitudinal Weld Scams19-894 A & B O.258 0.165 56 (IIcat # 33A277)

Surveillance Weld"' O.14 0.19 -

Circumferential Weld Seam Il 894 '*'

0.205 0.105 56 (licat # 6329637)

Lower Shell Longitudinal Weld Seams 20 R94 A & B O.197 0.060 -56 (Ileat # 90099)

NOTES, (a) WCAP 14689, Revision 2 11 (b) Best-estimate copp r and nickel from CE NPSD-1039 M (c) The surveillance weld is representative ofinterme'diate shelllongitudinal welds 19 894 A & B. Best-estimate copper and nickel values represent a single chemical analysis documented in WCAP-8810 W

Table 5 4 REACTOR VESSEL FLUENCE PROJECTIONS FOR 36 EFPY "

EFPY 0* 15* 15*

  • 30* 30* "' 45* l 1 36 4.34 2.68 2.14 2.01 1.93 1.35 l NOiES:

(a) WCAP 14689, Resision 2l 'I 2

(b) Fluence in 10" nhm (E > l.0 Men -

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

FARLEY UNIT I 15 REVISION I

PRESSURE TEMPERATURE LIMITS REPORT

- Table 5 5

SUMMARY

OF ADJUSTED REFERENCE TEMPERATURES (ARTS) FOR REACTOR VESSEL BELTLINE MATERIALS AT TIIE 1/4 T AND 3/4 T LOCATIONS FOR 21.9 EFPY *

Material 1/4-T - 3/4 T

(*F) ('F)

Intermediate Shell Plate B6903 2 138 114 Intermediate Shell Plate B6903 3 138 115

- Lower Shell Plate B6919-1 161 "' 13.5

  • Lower Shell Plate B6919-1 156 131 U>ing S/C Data Lower Shell Plate B6919 2 151- 126 Intermediate Shell Longitudina! Weld 82,e II3 g3 Seams19-894 A & B (lleat # 33A277)

Intermediate Shell Longitudinal Weld Seams 19 894 A & B 85 56l4*

(lleat # 33A277)

Using S/C Data 4 Circumferential Weld Il 894 123 97 (lleat # 6329637)

Lower Shell Longitudinal Weld Seams 20- 6., g3 8 5 ,,

894 A & B (lleat # 90099)

NOTES:

(a) WCAP 14689, Revision 2 DI 2

(b) The ARTS presented here are based on the peak reactor vessel surface fluence of 2.718 y 10" n'cm (E >

1.0 MeV) unless otherwise noted.

(c) Limi.ing 1/4 T and 3/4 T ART values. The P/T limit curves are those previously generated based on a 1/4T ART of 161*F and a 3/4 T ART value of 136'F which bounds the limiting 1/4 T and 3/4-T ARTS shown above.

2 (d) ARTS calcolated using the peak vessel fluence of 0.8307 x 10" n/cm (E > 1.0 MeV) at 45*

- FARLEY UNIT 1 16 REVISION I

PRESSURE TEMPERATURE LIMITS REPORT

- Table 5 6 CALCULATION OF ADJUSTED REFERENCE TEMPERA'IURE AT 21.9 EFPY FOR THE LIMITING REACTOR VESSEL MATERIAL . LOWER SilELL PLATE B6919 1 "'

Parameter bperating Period 21.9 EFPY Location 1/4 T 3/4.T Chemistry Factor, CF (*F) 97.8 91.8 Fluence, f(10* n/cm') *' l.695 0.659 Fluence Factor, FF 1.145 0.883 i12.0 86.4 ARTsor - CF x FF (.'F) initial RTsnt, I('F) 15 15 Margin, M (*F) 34 34 Adjusted Reference Temperature (ART), ('F) rer 161 135

. Regulatory Guide 1.99, Revision 2 NOTES:

(a) WCAP 14689, Revision 2 til (b) Fluence is' . sed on fu(10 n'em', E > 1.0 MeV) = 2.718 at 21.9 EFPY. The Farley Unit I reactor vessel war .ickness is 7.875 inches in the beltline region.

FARLEY UNIT I 17 AEVISION 1

PRESSURE TEMPERATURE LIMITS REPORT Table 5 7-PRESSURIZED TilERMAL SilOCK (RTm) VALUES FOR 36 EFPY '"

Surface Fluence ARTsnr I M RT 2

Material CF (10" n!cm . FF (CF x FF)

E > 1.0 MeV) ('F)

'F) . (*O (*h Intermediate Shell 91.0 4.34 1374 125.0 0 34 159 Plate B6903 2 Intermediate Shell 82.2 4.34 1.374 112.9 10 34 157 Plate B6903-3 Lower Shell Plate 97.8 4.34 1.374 134.4 15 34 183 B69191 Lower Shell Plate B6919-1 93.3 4.34 1.374 128.2 15 34

  • 177 Using S/C Data Lower Shell Plate 98.2 4.34 1.374 134.9 5 34 174 B6919 2 Intermediate Shell Longitudinal Welds 126.2 . 35 1.083 136.7 56 66 147 19 894 A & B (lleat # 33A277)

Intermediate Shell Longitudinal Welds 19 894 A & B I18.9 1.35 1.083 128.8 56 44 117 (fleat # 33A277)

Using S/C Data Circumferential Weld 11894- 98.4 4.34 1.374 135.2 -56 66 145 (lleat # 6329637)

Lower Shell Longitudinal Welds 91.4 1.35 1.083 99.0 56 66 109 20-894 A & B (lleat # 90099)

EQlfd:

(a) WCAP 14689, Revision 2"I (b) c3 = 17'F since the plate surveillance data did not meet credibility criteria FARLEY UNIT 1 18 REVISION 1

-- y .. - 4 f_- -

PRESSURE TEMPERATURE LIMITS REPORT a

q

?

+

  • L 6.0E References; ..

- . ., i

g. .

y

_ i

-i,

1. J WCAP.14689, Revision 2, Farlhy Units 1 and 2 Heatup and Cooldown Limit Curves for

~ Normal Operation and PTLR Support Documentation, E. Terek, December 1997. i

. - 2, iWCAP 14196, Analysis of Capsule W from the Alabama Power Company Farley Unit .1

. . Reactor Vessel Radiation Surveillance Program, P. A. Peters, et al., February 1995.

~'

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

L4, WCAP 14040-NP A; Revision 2, Methodology Used to Develop Cold Overpressure N

- Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, January 1996.

5

- 5.T WCAP-8810, Southern Alabama Power Company Joseph M. Farley Nuclear Plant Unit -

No, i Reactor Vessel Radiation Surveillance Program, J. A. Davidson, et al., December 1976 '

l 6. WCAP 9717, Analysis of Capsule Y from the Alabama Power Company Farley Unit No.~'l

- Reactor Vessel Radiation Surveillance Program, S. E. Yanichko, et al., June 1980. j 7 iWCAP 10474,- Analysis of Capsule U from the Alabama Power Company Joseph M. Farley

- Unit 1 Reactor Vessel Radiation Surveillance Program, R. S. Boggs, et al., February 1984.  ;

8. pWCAP-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. i

^

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

g l

5 5

I c_-

b

.._a., - .

1FARLEY UNIT I- 19 - REVISION 1 Y $ WM e h -.h i r; gp-4., w ..~,._%y = . .%.g , s,wa -my,.4 3 g,  % .y. gyg 9.y.p ,.w e -c g,

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