Text

Units 1 & 2 - Technical Specification (TS) Change No. 00-14, Pressure Temperature Limits Report (PTLR) & Request for Exemption from Requirements of 10CFR50, Appendix G, Enclosures 4 and 5
	ML022600424
Person / Time
Site:	Sequoyah ; (DPR-077, DPR-079)
Issue date:	09/06/2002
From:	James Smith; Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	GL-96-003, TS 00-14, TVA-SQN-TS-00-14
	Download: ML022600424 (175)
	v • d • e

ENCLOSURE 4 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQN)

UNITS 1 AND 2 PRESSURE TEMPERATURE LIMITS REPORTS (PTLR)

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do"On. 4*hMlt 0"d Am"ieuiop bwter No, N 10004 tOald Junie 06-2002)

TrM4CWr YM.L:LY AU,?HOPJTY see? (N)

B3Y D), L Lundy Tennessee Valley Authority, Sequoyah Unit 1 Pressure Temperature Limits Report Revision 3, May 2002 PRO JE C T-

-Sa~u-A DISC I PLINE

-N CONTRACT 9INNP~j -86305R

ýUNIT.L-J DESC. KCS PmsuwL-TMgger~lr Lim~a Rqwon DWGIDOC NNO.

PIhR-I STIfFFT---

--- O)F REV 03 RIMSI, WTC A-K

List Of figures................................................

v l.6 RCS Pressure Temperature Limits Report (PTLR)......................

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

I...

2.1 RCS Prcssure9r-emperature (RI) Limits (3.4.9ý1).....................

)...................................

3.0 Low Temperature Overpressure Protection System ([CO 3.4.12).

I 3.1 Pressuri~zr PORV Lift Setting 1nnts.....................

...... 2 3 2 Armung Temperature.................

2...........

2 4.0 Reactor Vessel Mamri2l Surveillance Program.........

2 5.0.

Supplementa Data TabIc......................................

6.0 References...............................................

..... 19

wvith Uncertainties for InSMruentlation Errors of I0O`F and 60 psig)...........................

...... 6 Table 2-2 Scquoyrah Unit I Cooldown LEtots at 32 EFPY SO

-WOO. -

-(With Uncerainties for Instmrentation Errors of 1O7 and 60 psi3)........

8 Table 3-I Selected Serpoints, Sequcyah Unit I...........................................

-10 Table 4-4 Sequoy*th Unit I Reactor Vessel Surveillance Capsule Withdrawal Schedule...............

12 Table 5-1 Comparison of the Sequoyah Unit I Surveillance Material 30 fl-lb Transition Temperature Shftts and Upper 5helf Energy Decreases with Regulatory Guide 1.99, revision 2, Predictions..........................................

13 Table 5-7 Calculation of Chemistry Factors using Sequoy-ah Unit I Surveillance Capsule Data..... 14 Table 5-3 Reactor Vessel Beltline Material Urtirradiated Toughness Properties for Sequoyah Unit I 15 Table 5-4 Neutron Fluence Projections at Key Locations on the Reactor Vessel CadBase Metal Intertiace for Sequoyah Unit 1 (101' orrm2, E 1, OMeV)..........

... 6 Table 5-5 Sequoyah Unit 1 Calculation of the ART Values for the l!4T Location @ 32 EFPY...... 17 Table 5-6 Sequoyah Unit I Calculation of the ART Values for the 3!4T Location @ 32 EFPY..... 17 Table 5-7 Sumumary of the Limiting ART Values Used in the Generation of the Sequoyah Unit!

Heatup/Cooldown Curves..............................................

I Table 5-8 RTrs Calculations for bequoyah Unit I beithne Region Materials at 32 EPFY..-. 18 iii

(Heatup Ratw of 60Ffhr) Applicable for the First 32 EFPY (w/-Margins for h, rimentation Errors of I OT and 60 psig)................

4 Figure 2-2 Sequoyah Unit I Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 32 EFPY (wvv Margrns for L nstramenation Errors of IOF and 60 psig).......................................

5 Figare S-I Sequoyah Unit I MOMS Se ts......................

i t iv

This report affect TS 3.4-9. 1 RCS Pressure/Temperature Limits (PTT)Limns. All TS requirements asociated wiLh tow Temperature Overpressure Protection Sysiem (LTOPS) are contained in TS 3.4. 12, RCS Overpressure Protection System.

2.0 RCS Pressure znd Temperature Limits The limits for TS 3.49-1 are presented in the subsection. which follow and were developed using the NRC approved methodologies specified in TS 6.9.1.15 with exception of ASME Code Case N-640-')*

(Use of KIj, WCAF- ! 53 i*[1"1 (Ellnunation of the flange RequircminO, 1996 V\\trion of Appcndix 0ý41 and the revised fluenceJsI The operability requirements associated with LTOPS are specified in TS LCO 3,4 12 and were determined to adequately protect the RCS against brittle fracture in the event of an

[TOP Transient in accordance with the methodology specified in TS 6.9.1.15.

2.1 RCS Pressure/Temperature (P/T) Limits (LCO - 3A.9.1) 2.1.1 Thc 1ufilntun boltup tcmpcraturc ia SOF 2.12 The RCS temperature rateof-change limits are:

a. A maximum heatup rate of 100IF in anyone hour period.

b. A maximum coohiown rate of 100lF in any one hour period,

c. A maximum temperature change of less than or equal to 107F in any one hour period during inc*e*vce hydrosmtie and leak testmg oprtntionns nhwe the he.zhp and cooldown limit curves, 2.A.3 The RCD P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and cnttcahty are specified by Figures 2-1 and 2-.

1-f.

lnw Temperature Ovepressure Prnneedon System (LCO 3.4.12)

The lift setpoints for the pressurizer Power Operated Relief Valves {PORVs) are presented in the subsection which follow. These lift setpoints have been developed using the NRC-approved method0ologie specified in Spmldfratiuii 3.4.12.

1

reactor midplancibelfline or for instrument inaccuracies. Th pmressure difference herween tihe pressurizer transmitter and the reactor vessel imdptaneibelthine with four reactor coolant pumps in operation is 683 psi (Ret 127.

Note:

These setpoints include allowance for the 500 F thereal transport efftet for heat iniecton transients. A demonstrated accuracy calculation (Reference 13) has been performed to confirm that the setpoints will maintain the system pressure witihin the established limits when the pressure difference between the pressure transmitter and reactor midplane and maximum temperature/pressure instrument uncertainties are applied to the setpoints.

32 Arming Temperature The LTOPS arming temerature is based upon the methodology defined in the Sequoyah Nuclear Plant Unit I Technical SpecificationsAdministrative Controls Section 6.9.1.15. The arming temperature shalt be.6 35aF.

4.0 Reactor Vessel Material Surveillance Program "The reactor vessel material irradiation surveillance specimens shall be removed and examined to rlt-ermine chhnge. in mnatenal properties. The removal schedule is provided in Table 4-1. The results of these examinations shall be used to update Figures 2-1, 2-2 and 3-1, The pressure vessel steel surveillance program (WCAP-823P 1) is in compliance with AppendixiH to 10 CFR 50. 'Reactor Vessel Material Surveillance PFvrurai Rtquhmeiwnc

"&. Thc inaterial tat requirements and the acceptunce standard utilize the reference nil-ductility temperature RTgmr, which is deterrmned in accordance with ASTM E208r'l The empirical relationship between RTNpir and the fracture toughness of the reactor vessel steel is developed in accordance with Code Case N-*4* of Section XI of the ASME Boiler and Pressure Vessel Code. Appendix Q. "Fracture Toughness Criteria for Protection Against Failuret'), The surveillance capsule removal schedule meets the requirements of ASTM E1g5 -8 2It4. The removal schedule is provided in Table 4-1, 2

Table 5-2 shows calculations of the surveillance material chemistry factors using surveillance capsule data. Note that in the calculation of the surveillance weld chemistry factor, the ratio procedure from Regulatory Guide 1.99, Revision 2 was followed. The raMio in question is equal to 0.90.

Table 5-3 provides the required Sequoyah Unit I reactor vessel toughness data.

Table 5-4 provides a summary of the fluence values used in the generation of the hearup and cooldown linmt curves and the PTS evaluation.

"Tible 5-5 and 5-6 show the calculation of the 1/4T and 3/4T adjusted reference temperature-at 32 EFPY for each belthine ma=r-al in fhf &tjuosyah Unit I rcdctor vcm&l, The limiting bcltrlic matcnit wun thu lower shell 04.

Table 5-7 provides a summary of the adjusted referene temperatur (ART) values of the Sequoyah Unit I. reactor vessel beltlme materials at the /4IT and 3/4T locations for 32 EFPY Table 5-8 provides RTyrs values for Sequoy)ah Unit I at 32 EFPY.

3

"Calculated Pressure 4PSIG) a ftt N

Lf in N

Ut..

-4 0

N) 0 O.

U0.0 1 1...

5!!!

0

Calculated Pressura (P810)1 Q

0,2 4C np V

4.

C t

T p

T P

50 0

288 272 2000 50 525 289 51-288 2485

25.

25&

523 60 525 289 525 65

ý25 288 525 70 525 288 526 75 525 2S8 52a 105 525 288 557 110 525 288 566 115 525 288 577 120 5*

299 589 125 125

'288 603 130 526 288 619 135 528 288 635 140 531 288 654 145 536 288 675 150 542 2.88 698 155 549 288 724 160 557 2,88 752 165 566 288 74, 170 577 288 819 175 589 2S8 858 180 603 288 900 185 618 288 948 190 63 5 288 1000 195 654 290 1057 200 675 295 1121 205 698 300 1191 210 724 305 1269 215 752 310 1354 270 784 315 1449 225 819 320 1531 230 858 325 1617 235 900 330 1712 240 948 335 1817 2-45 MhAO 340 1933 250 1057 345 2060 6

T T

P 260 1 t91 355 2357 265 1269 270 1354 275 1449 28o 1531 285 1617 290 1712 295 1817 300 1933 305 2060 310 2201 315 2357 7

I T

P T

p I

8 T

P

'1 P

50 50 55 60 65 70 75 80 a5 90 45 100 105 110 115 120 125 130 135 140 145 B50 155 160 165 170 175 18.0 185 159 19ý5 200 205 210 215 0

553 555 556 558 560 561 564 566 569 571 575 578 582 586' 591 596 602 608 616 623 632 642 652 064 677 691 707 724 743 764 788 314 843 5o 50 55 60 65 70 75 80 85 9,0 95 100 105 Ito 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215.

0 503 505 507 509

`510, 512 514 516 318 521 524

$27 531 535 540 545 550 556 563 571 579

.588 5,99 610 623 637 652 669 709 733 759 787 819 50 50 55 60 65 70 75 80 85 95 100 105 110 1-15 120 125 130 135 140

[45 150 155 160 165 170 175 180 185 190 195 200 205 210 215 0

457 458 459 460 462 464 465 468

.470 473 476 479 483 487 492 497 503 510 517 525 534 544 556 568 582 597 614 633 654 677 702 731 762 797 50 50

ý;55 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185

!90 195 200 205 210 215 0

408 409 410 411 412 414 416 418 420 423 426 430 434 438 443 "449 456 463 471 479 489 500 512 526 54.11 558 577 597 620

'646 674 705 740 779 50 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 190 195 200 205 210 213 0

305 306 307 308 309 315 318 321 325 329 33D 338 344 351 358 367 376 387 399 412 427 443 461 482 505 530 558 590 624 663 754 754 I

T P

1 P

S,...

-T 220 874 220 853 220 836.

220 821 220 S06 225 909 225 892 878 225 869 225 865 230

.948 230 935 230 925 230 D21 230 930 235" 99i 25 982 235 978 235 979

-240 1038 240 1034 240 1036 245 1090 230 1148 255 1212

260, 1283 265 1360 270.

1447 275 1542 280 1647 285 1763 290 1892 295 2034 300 2191 305 2364 9

S Setpoint (psig) 50 490 465 100 500 475 135 540 510 175 575 540 200 610 J70 745 685 2.80 745 685 405 745 6&5 450 2350 2350 10 Setpoint (psig)

2500 0

ce 5w0 0-MflL.

100 150 1200 250 300 350.

40w 460 Fbactor Coolant SYsten¶ Tomperature (On -in-m*I SAtpoo Figare 3-1:

Seq noyab Unit I ETOPS Seiecied Setpolnts (Plotted Data provided on Table 3-1) goo 0

so

Tj 4GO 3.39 20

-i ol(

11j 1400 3A47 3I7.96 N14V (c)

X 20

~

-1.47 17L2t'(C)

Y 320" 143.

to.03 2,19 x 10 (c~d) 401-OS tanbY (d&e) v 176" 1.08 Staadby (d~c)

W 184" 1.08 Standby Z

3560 1.8SWdby (d~e)

Updated in Capsule Y dosimettxy aoalysis (WVCAP-l S224Th).

Effective Full Power Years (EFPY) from plant startup.

Plant specific. evaluatIOM This fluence ts, not Ilees than once or greater than vvvice thie peak end of license (32 EFPY) fl uence CapsulesS~, V, W and L will reach a iluence of 2.14 X 10" (E> O )Melv, th~e 48 UPY peak vessel fluence at approximately 44 EFPY, respectivtly.

12 Notes:

(a)

(b,)

(c)

(d)

(C)

30 ft-lb Transition Temperature Shift Material Upper Shelf Energy Dcrease Predited (0/) III Measured Lower Shell T

0.261 5u3 67,,52 )16 16 Forging 04 U

0.796 J 89.3 10: 7 2035 21 (Tangentiar)

X 131 102-6 14512 23 8

Y 2-19 114.95 129,87 26.5 23 LtMWO 3T*l1 T

0.261 59,85 50-5*

16 0

Forging 04 U

0.796 89.3 67J59

ý20.5 19 (Axia1)

X 132 102-6 103.34 n

22)

Y 2.19 114.95 133.35 26,5 19 Weld Metal T

0261 111113 127.79 3530 0,76 1653`7 14402 42 26 x

1-.32 190,51 159102 45 21 Y

2.19 2134 t63.9 48 28 lIlAZ Metal T

0,261 45.48 20 U

0,796 7 8, 94 26 X

1_32

-95.-

3 Y

119 73.3 0

Notes:

(a)

Led on Regulato^' Guide 1.99, Revision 2, methodology v6ýng the menn weigh! percent values of copper and nickel of the surveillance materml.

Ib)

Calctiated using measured Charpy data plotted using CVGRAPH, Version 4.1s.

(c)

Values are based on the defmniion of upper shelf energy given in ASTM El 8-"2.

Measured Capsul Flueiice Predicted 15

5-4 0

' tIN f-S 0

5.-

U.

0 '-5.

Sf5 U.

0 5.

00 0

0

-t -

=

I t "-t

".4 U~.

0.

14 0 eN ir

0 N

0

1 4tb 0

0 4,.

00 CC ell A-1 I-'

54, 00 5.-..

Cl 00

=

I i 5 =

=

0 II IN 00

+

00 N

00 L.

I-"

U 0 9 00 IN I-IN 5.It 00 Co 0-.

(4 eN 0

5.-.

N 0'

5 LU 0'.

r-5 ci 00 IN 0

w.i i-i 0

'1

4. a:

0 0

00 IN Ii 4

5 144 V

ci 5-ii C

o 0

'.5.'

U 0

0..

E

5-U

-

,

-

i.J U

U

5

-

fi

,

i

2

ii

'

'0 Ap

(fleat A 98%,07(128 1,489) xqkte (a) The Initial RTNOT values arc mecasurcd values (b,) Thc~ czoppor andi nickel vailuat arm beist eifrniate ivaltwe for only the stu-'.-eillane owtd, rxwtA] Pnd ig th#- "Vtrflte of three data points, [0.424 (WCP-10340, Rer. 1), 0-406(WCAP-10340, Rev. 1), 0.33 (WCAP..8233) copper and 0.084 (WC.AP4lON4O. Rev. 1).'0.085 (WCAP-1 0340. Rcv. 1)1. 017 (WCAP-82~33 vikckl4j These values are treated as owe data point in the calculation of the best estimae average f'or the inter, to lower shell circ. weld shown above. Origially the 0.424/0-406, and H.84 /0.085 values were reported as single pointsp, 0.41 -0.423 and 0,08 (Per WCAP-10340, Rev. 1'6), but 0 is actually made up of two data points. Sample TW58 from capsule T was broken into two samplms TW58a and TW58b, thus providing the two, data poinrs.

(e) From N.RC Reator Ve~gel Intrtm-ty fataashe: (1ZVTD) and ultimately frorn, Rotterdam Weld Certifications.

(d) Circwimferrenfial Weld Seam W05 was fkbncated with weldxwir type SMIT 4J), Heat # 25295. flux rype MIT 89, lot 9 22V5, The siurvillance weld was fabricated with~ weld wire type SMIT 400 Heat # 25295. f lux type SMIT 99, lotr4 1103 and is representative of the mntertmdiate to. lower shell vircumfetential Weld.

(e) The surveillance weld and the three. Rotterdam tests arc averaged together for M~e 13est FE;strimate of the hitermediate to Lower Shell For*&n Circunferential Wel~d Scam.

15 Lawer Shell Forging 04 0.13-0,76 73I (Hear # 9909191281587) 1______

suvdaceed 1

0 Q397 011i RrniterdamnTesteC

=

0.30--

Ru ttrdaim Tcseý-~ '

0.46 Slest Estimate of the Irner~nediate to Lower Shell 013.3 0.11

-0 Foroig Circumlterenttal Weld Seamo W050'

I-,

'0

-t

-i r4 0

a 0

00 a

t4 O

f"1 t4

21

0 0

.r

'0 0

.L.

0

Intermnetdive Shelf Fozgmng 03 Position H Lower Shell, Forging (4 poito I I Piin2.1 intermediate to Lowcr Shell pnstoin, 1II circumierctati4 Weld Seam Posioni,7.1 (2)

(4)

Initial RT.~Tu values measured values.

ART =Initial RT,.D ART-r + -margin (oF)

&RT,

-~CF FE N4M 2 Ta-bie.5-6 Sequoyai Unit I Calculation of the AkT Valucs for ffic 14T Luuatiup @ 3Z EFTY MeilRG 199ý CF F1 lRT, 3 '# ATO j Marppn',

R2 Methw

~

(CF)

(

f 7i(F)

(T)

Intemv-diate Shell rorging 05

.Position~ 11 115,6 0.747 49 86.4 3-16

'Lower Shell Forging 04

'psiioni 1.

95 0,747 73 71O 3 4 )178 hrerwidiaieio Lower Shell Position I.;

iO.

0.74~

7~

-4 1'

34,5 5

137 CircumTerendal Weld Seamn posim I i

1 135.0

-747 40 1701,8 T 56 r

(3)

(4)

Initial RT,%v N-Alus meAsured values.

ART - Intiral PTNfT + AR'r..,Mrgn ARTNOT CF '

=

+

f

'17

Imermediste ShelL Forging.05 Posniono

1. 1 193 160 Lower shell Fbo'Sing fl.

Poirwio 1.1 205 J 178 Nwivin, 2.1 M

A 8

fmcuxae to Lwer Sbc-1 PosJiOien I.

1 8

137 Cftrmnferenfial WVeld Seam fosition 2,1 155 1

117 Table 5-8 wRT Calculartions for Scquuy-ah Unit I BIl~iu Rcgiva Miatclalb 4t -siZ 32 VVY M4aterial Fluence EF C

R

'Imargin R~T~Rr 04crWf. E>1 I D CIF (OF)

(F Intmemdiate Shell Forging -05 1L84 1,167 115ý6 134.9 3 4 40 209 Lower ghedIFotrjggin 4

1 R4 I

t7

15)

IIS 4

73 218 Lo~we ShoellForging 04 1.84

[.167 105.9 123.6 34 73 231 (Vsing, &C Da.%)______I____

Circumnfmrntal Weld MrWa 1,84 1.167 1 'j-3 1~$&2 56

-40 204 Circuwnferernial WelId Me+/-.i 1,4 1-16?

135i0 157-5 56

-40 174 (lUs' S/C Data)

Notes:,

('a) 1Nua1 R.T,,43 vailues are measured values 0b)

R-Tk~s - RTnyp-ijn + ART~r + M~n(F fe)

ARTpr F*F

2.

Code of Fcderal Regulations, IOCFR50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, ILS, Nuclear Regulatory Commission, Washington, D.C.

ASTMN E206. wundard Thsti fethod for Conducting DropWlceight Test to Determine Mil DuctiliA, Tranition Temperatur qfFerritic Stee-s, mASTM Standards, Section 3. American Society for Testing and Matenals. Philadelphia. PA_

4.

Section XI of the ASME Boiler and Pressure Vessel Code, Appendix Q Fracture Toughness Criteria for Protecuon Against Failure 51 ASTMI EI 83-82,.Annual Book ofPASTM Standards, Seetion 12, Vohune t) 0f,.Standard Prartice for Conducting Surveiltance Tests for Light-Water Cooled Nuclear Power Reactor Vessels.

6.

Regulatory Guide 1.99, Revision 2, Radiation Embrilulement of Reactor VesseltMaterials, U.S.

Nuclear Regulatory Commission. May 1988.

7.

WCAP-15224, Analysis of Capsule Y From the Tennessee Valley Audhoriry Sequoyah Unit I Roa~prrr Vmprl Radiaton Surveillance Program. TJ. Laubham. et at-Dated June 1999,

8.

CVGRAPH, Hyperbolic Tangent Curve-Fitting Progran, Version 4.1, developed by ATI Consulting, March 1999.

9.

WCAP-10340, Revision 1, "Analysis of Capsule T from the Tennessee Valley Authority Sequoyah Unit I Reactor Vessel Radiation Surveillance Programr", S.E. Yanichko. et. at.

February 1984, 10, WCAP-14040-NP-A, Revision 2, "Methodology used to Develop Cold Overpressure Mitigating System Sezlpoints and RCS Heatup and Cooldown Limit Curves", ID. Andrachek, et. al., January 1996.

11.

WCAP-15293, Revision 1. 'Sequoyah Unit I Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentationa", IH. Ledger, April 2001.

12.

Westinghouse Letter to TVA. TVA-93-105, "Cold Overpressure Mitigation System Code Case and Delta-P Calculation", May 19, 1993 13, Calculation SQN-IC-0 14, "Demonstrated Accuracy Calculation for Cold Overpressure Protection System-"

14.

ASME Code Case N-640, "Altemative Keference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1", February 26, 1999.

19

20

LeMWr o N10004 T1d*t=sE YAOLtS! MSUTOWMTY Soap (N) 0V D-., Lundy Tennessee Valley Authority Sequoyah Unit 2 Pressure Temperature Limits Report Revision 3, May 2002 PROJECT ScouI-ah DISCIPLINE N

CONTRACT 9INNP.86305B UNITL2 DESC-RCS Prre-TempeTature LunI,ort DWG/DOC NO, CLR-2 SHEET-OF REVM 03 DATE 06/10/02_ECN/DCN-FILE N2N-048 RIMS, WTC A-K

List Of Figures...............................................

............... V tO RCS Pressure Temperature Limits Report (PTLR) 1 2.0 Operating Li 1

2.I RCS Pressumremperature (PIT) Limits (3.4.91)..............................

I 3.0 Low remperature Overpressure Protection System (CO 3.4.12)..

...... 1 3.1 Pressurizer PORV Lift Setdng Limits...........................

2 3.2 Anr ing Temperamure.........................................................

2......

4.0 Reactor Vessel Material Surveillance Program............

5.0 Supplemenma Dita Tab*Ies......

6.0 References.......

1.

ii

(wit Unctrtamnes for lnsmntmrtanon Errors of lOT-and 60 psi..............

6 Table 2-2 Sequnyah Unit 2 Cooldown Limits at 32 EFPY

"" (with Unctrtainties tar lnstnuncatioa tbras o1 Ivi-and O) psig)...........................................

7 Table 3-1 Selected Setpoint., Sequoyah Unit 2 Table 4-I Sequcysh Unt 2 Reactor Vesscl Surveillance Capsule Withdrawal Schedule............ 0 Table 5-1 Compaison of the Sequoyah Unit 2 Surveillance Material 30 "-lb Transition Temperature Shift, wi]d Uppci ShclfE-Izvgy DtcrcAS.s with Rcgulatory GC ck 1.99, revision 2, Predictions..........................

I Table 5-2 Calculation of Chemistry Factors using $cquoyah Unit 2 Surveillance Capsule Dta.-.12 Tabe-5)3 Rcactor Vessel Beltline Material Uniumadiaed Toughness Properties for Sequoyah Unit 2.................

13 Table 5.4 Neutron Fluence Projections at Key Locations on the Reactor Vessel Clad/Base Metal Interface for Sequoyah Unit 2 (10 n/cmr, E> 1.0 MeV)..................... 14 Table 5-5 Sequoyah Unit 2 Calculation of the ART Values for the 1/4T Location C* 32. EFPY..... 15 Table 5-6 Sequoyah Unit 2 calculation of the ARTValucs for the 3/4T Location @ 32 EFPY...... 1I Table 5-7 Swmmany of the Limiting ART Values Used in the Generation of the Sequoyah Unit 2 lfeaup'Cooldown Curves.................................

.................... 16 Table 5-8 RTPr5 Calculations* fur -quuyah Unit 2 Bltlinue Reqivtn tvttU Lz at 32 EITh....- 16 iii

(Heatup Rare of 60*F/nr) Applicable for the First 32 EFPY (vd Margins for Inasmentation Errors of 10'F and 60 psig)

.4 Figure 2-2 Sequoyah Unit 2 Reactor Coolant System Cooldown Lrrnntauons (Cooldown Rates up to 100IF/hr) Applicable for the First 32 LEPY (wI Margins for Instrnmentation Errors of 1I0°F and 60 psig).......................................

Ftgurc 341 Sequoyah Unit 2 COM 9

iv

This report affects TS 314.9.1. RCS Pressurc)'rnmpcrature Limits (PT) Limits. All TS requiremeams associated with Low Temperaur Overpressure Protection System (LTOPS) are contained in TS 3.4.12, RCS Overpressure Protection System.

2.0 RCS Pressure and Temperature Limits The limits forTS 3.4.9-1 are presented in the subsection which follows and were developed using the NRC approved methodologies specified n TS 6.9-1.15 with excepdoa ofASME Code Case N-6401"I (Use oJf Kj, WCT-l-5315l52 (Eirmination of the FlIane Requiremnent), 1996 Version of Appendix d#'

and the revised fluencestr.

The operability requirements associated with LTOPS are specified m TS LCO 3.4.12 and were deterrmined to adequately protect the RCS against brittle fracture m the event of an LTOP Transient m' accordance with the methodology specified in TS 6.9. 1. 15.

2.A RCS Pressure/Temperature (P/I) Limits (LCO -3.4,9.1) 2.1.1 The minimum boItun tensp&rature s 5OWF 2.1,2 The RCS temperature rate-of-change limits are:

a, A maximum heatup rate oa I 0Wkf in any one hour period.

b. Amaximurn cooldown rate of 100F in any one hour period.

c. A maximum temperature change of less than or equal to 101F in any one hour period during inservice hydrostatuc and leak testing opertions above dte heanm and cooldoiwi limit curves.

2.13 The RCD P/T limits for heatup, cooldown, insencre hydrostatic and leak testing, and criticality are specified by Figures 2-4 and 2-2.

3.0 Low Temperature Overpressure Protection System (LCO 3.4.12)

The lift setpoints for the pressurizcr Power Operated ReliefValves (PORVs) are presented in the subsection which follows. These lift setpoints have been developed using the NRC-approved tuctiodologics spccificd in Spccifivation 3.4.12.

I

reactor midplunebeltdine or for instrument inacctracies. The pressure difference between the pressurizer transmitter and the reactor vessel midplane/bettline with four reactor coolant pumps in operation is 68.3 psi (Ref. 13)

Note:

These setpomts include allowance for the 50F theramal transport effect for heat injection transimts. A demtonstrated accuracy calculation (Reference 14) has been performed to confirm that the setpoints will maintain the system pressure within the established limits when the pressure difference between the pressure transmitter and reactor nidplane and maxunum temperature/pressure instrument uncertainties are applied to the setponts.

3.2 Arming Temperature The [I0PS arming temperature is based upon the methodology defined in the Sequoyah Nuclear Plant Unit 2 Technical Specifications Administrative Controls Section 6-9.1.15. The arming temerature shall be < 3500F, 4.U Reactor Vessel Material Surveillance Program The reactor vessel material irradiation surveillance specimens shall be removed and examined to determine changes in material propefties. The removal schedule is provided in Table 4-1. The results of these examinations shall be used to update Figures 2-t, 2-2 and 3-1.

The pressure vessel steel surveillance program (WCAP-85 13t 11) is in compliance with Appendix fI to 10 CFR 50, "RvRc tor Vtbzi Mltncial Survcitlancc Piouram Rcjuirmrncn& 4 1. Trh aterizal tcmt requirements and the acceptance standard utiize the reference nil-ductilit temperature RTU,0 -, which is determ-ned in accordance with ASTM E20813I The empirical relationship between RTNDT and the fracture toughness of the rcactor vessel steel is developed in accordance with Code CaOe N-640 of Section X1 of the ASME Boiler and Pressure Vessel Code, Appendix Q Ttracture Toughness Criteria for Protection Against Failure'tl, Thesurveillance capsule removal schedule meets the requirements of ASTME18582tP$.

The removal schedule isprovided in Table4-1.

2

Table 5-2 shows calculations of the surveillance material chemistry factors ustrg surveillance capsule data. Note that in the calculation of the surveillance weld chemistry factor. the ratio procedure from Regulatory Guide [.99. Revision 2 was followed. The ratio in question is equal to 0,93.

Tablc 5-3 provides the required Sequoyah UDit 2 reactor vessel toughness data Table 5-4 provides a summary of the fluence values used in the generation of the heatup and cooldown Jint curve's and the PMl evaluation.

Table 5-5 and 5-6 show the calculation of the tV4T and 314T-adjustcd rcfernce terperawure at 32 EFPY bui caci bcltlin nomtcrial in thc Scquoyah Unit 2 reactor-vetel. The, limiting beljnea nlmtenal wa the intermediate shell 05.

Table 5-7 provides a summary of the adjusted reference temperature (APT) values of the Sequoyah Unit 2 reactor vessel beltime materials at the V4T and 3/4T locations for.32 EFEY?

Table 5-8 provides RTPrs values for Sequoyah Unit 2 at 32 EFPY.

3

4.

V V

r.bC eq f Q

44 U

Q eq V

"I.

0 0

to C 0

'4) 0 to C

C r

C to 0m a

0 0

Q Q

40 in 0

)

N 0')

c 0

Calculated Pressure (PStG) o o

0 Ac 0

0*

C3, 0

C-t-

0 CA*

Z *.:

~ ~ ~~~~~~~~...

.0

-4N th

~

50 0

214 0

186 2000 50 591 214 607 203 2465 59 14 61d 60 601 214 622 65 607 214 675 70 614 214 67B 75 622 7

21

685, 105 675 214 695 110 678 214 70S 115 685 214 725 120 6os 214 745 1-5 708 214 768 130 725 214 795 135 745 214 825 140 768 214

$60 145 795.

214 898 150 825 214 942 155 860 214 990 i6o 898 2.15 1043 165

942 2129 1103 170 990 225 1168 175 1043 230 1241 180 1103 235 1322 M

' 1108 240 1*411 190 1241 245 1510 195 13,22 250 1602 200 1411 255 1697 205 t510 260 1801 210 1602 265

1916, 215 1697 270 2043 220 1801 275 2183 22$

11-01 2N0

2139, 230 2043 235 2183 240 2338 6

T p.

IT P

T P

S3teady State 20F 411 6OF OOF T

P T

P, T T P

50 50 55 60 65 70 75 80 95 90 95 100 105 110 l is 115 120 125 131) 135 140 145 1 50 155 160 165 11A 175 180 185 190 195 200 205 210 215 220 225 230 0

591 595 601 607 614 622 630 640 661 674 688 U,3 720 739 760 809 837 868 940 982 1028 IW*O 1136 1199 1268 1344 1429 1522 1625 1739 1865 20W0 2158 2328

1 50 55 60 65 105 75 80 85 i 0 911 195 140 145 115 120 125

)30 135 140 145 155 160 I 6S 0

554 558 Mf4 572 580

589, 599 610 623 636 652 6877 707 730 783 814 848 8 01 927 973 1024 59 50 65 70 75 85 99 95 10.0 105 110 115 120 125 130 135 140 145 150 155 160 165 0

103 508 514 521 5Z9 538 548 559 571 584 616 034 6q4 676 701 729ý 759 793 831 872 918 968 1025 50 30 70 75 80 85 90 95 100 105 110 115 120 125 13o 135 140 145 150 15$

160 0

461 466 470 478 486 496 506 518 546 562

-580 600

61.

647 674 704 738 "I5 816 862 913 969 5O 55 60 65 70 75 80 85 90 95 100 105 110 120 125 130 135 140 145 I50 155 0

366 i37 380 389 "399 410 423 437 453 470 490 512 536 563 593 626 663 70.4 749 800 856 918 7

~

00 00

~0 00 00 t

I.w 0

~L 0e 0

n

Peactor Coolant System Pmsesure ipsig)

(A

,.4.

260 640 4t C

C

Updated in Capsule Y dosinmtry ýiwlysis (WCAP45320()-ý Effective Full Power Yeas (EFPY) from plant startup.

Plant specific evaluation.

ThIls filuence is not less than once or greater than twice the peak end of license (32 EFPY) fluence Capsules S, V, W, and Z will reach a fluence of 2.71 x 10l1 (E > [ 0 MeV), the 48 UPY peak vessel flumnce at apmpoximately 44 EFPY.

10 (b)

(e)

(d)

(C)

Material Capsule Flueiwe Iv

I,,

,l,,....

i hleitnediate Shell T

0261 6033 63.65 1

Forging 05 u

.62 95.22 79,31 21 16 (Tangential)

X 1,22 100.23 85,7 23 8

Y 2.14 114.67 134.12 26 22 Iheimediate Shell T

0261 60.33 4.&73 17 7

Forging 05 U

(,692 85.22 66,06 21 9

(Axial)

X 1.22 10023 110.04 23 2

Y 2.14 11467 89.21 26 22 Wedd Metal T

0,261 43.12 74.56 20 2

U 0692 60-91 130,38 25 6

X

.122 71-63 44.22 29 35 Y

2.

R1*

Rt 1 6 Metal T

0261 24.58 2

U 0:692 64.03 14 X

1.22..

21929 Y

2.14 5032

-39 S.

ii ii~llIllll£,__*

30 ft-lb Transition j

Upper Shelf Energy Temperature Shift Decrease Predicted Mebasured Predicted Measured NOtes":*

(a)

Based on Regulatory Guide 1.99, Revision 2. methodology using the mean weight pment values of copper and nickel of the surveillance matenal, (b)

Calculated using measured Chpy data plotted using CVGRAPH,' Vcrsion 4.1.

(c)

Values are based on the definition of uppeT shelf energy gVen in ASIM E185-82.

11

Intesmediate Shell T

1.61E1418 0,633 63.7 40.45 0.A03 Iog 0

6.92E-+Ig 0.897 79.3 71-0 U.05 (Tangential)

X

11) +19 os9A

.1 Y2E1~i 1.205 815-34.4.11 j 214E19 120 13.1161,86 1-1457 Intermwdhate Shell T

2,019+1's 0.6ý5 48.7 360A2 GAO0 Fotging 05 U

6.92E-+18 01397 66.1 59.29, 0805 (AXWa)

X L2215+19 1,055 1 ID.0 116-0S 1.113 y

2,14E+19 1.207 j

892 107.66 L457 SU:j 677.77~F 7.5 CFM UF RT--r)+ D. FF2) = (677.77) -(7,55,6)

- 89.7~F SuIveillance weld T

216IE418 0-63.5 69.4 (74.6.)

44.07 IT.403 Maea1UL 6.92E+18 0.897 121,3 (130.4) 108.81 0.805 x

L22E+19, 1.055 41.1 (44-2) 43.36 1.113 y

214E+19 1207 s0.8 (a.)

7 1.457 SUM.

24-3.777V 3.778 CF S., W-, = Iff RTNOT)

X ')(293,77TF) + (3,779)= 77.8*F ICapsule ICapsule f "

-Nocs (a) f ý Calculated fluence ftom capsuile Y dosimetry aalwysis. resul~ts", (-k lo" nwcn, 1:> IA)meCv).

()aRTmT values ar the nmesured 30 ft-lb -shift values taken from App. B of Refl 7, roindtd to one dccizl Point.

(d)

The surveillance weld rmtal ARTthnD values have been adjusted by a rano Thctor of 0.931 12 Matcrial FF" AKINDTý"'

Vr

(Heat In, 981234 / 7814 t6)

Vessel Flan~e 2F (Heat, #980893 / 7S9N3I) hiwenmediate Shell 05 0135 0-76 L

(Ficat # 288757 1981057)

Luwci 51=U P'oig 04 0,14 07 (Hieat # 9904,69 ;1-2933231)

Iniemediate to Lower Shell Jorgmg 0-12 0.11 O

Cfrcumnferemia] Weld Scadi' Surveillaiv= Weld'ý 0113 0.11 (a) The Initial XrT4DT vadue mr measurcd values (b)

Circu~trcnftidl Weld Searm was fabricated, with weld wixc qW, SMIT $9, Heat # 4278,.Mlx rype SMIT 89, lot Ju 1211 and is representative of the intemetliae to lower shell icircurnt~rctial weld-13

I 'r

~

I2 12 1 i

,, 1-1 x

IL.

0 0:

a.

Ti e4 4

A I~

J-

-I Y -

=

T R

-t t -?

I 0

.2 0

bl

0 C

C)

I'.

4'1 C

0 0

0

a.

I'd 0

U 0

.0

'U

.0

-4 0

=-

6 6

4:'4 C

C

a.

;

-C

o i

ii 4

I-

+

o ij

ii a

I

-

r-

.'i

4I 9

b4,.

Ltz 6

-u llý I :a

1nttrnxdiate Shell.Forgilng 05 pwillon 1. L 14211 Poito ~

t 36 H

Lower Shell Forging 1,14 Posmo

[I?

9 lattrrmedite to Lower Shell Poition 1.1 11 17 99 Circumfrncrbal Weld Semm Posmoh 2, I 132 M)~j Table 5-8 RT~s Calculations for Sequo*a Unit, 2 Belfline Region Materials at 32 BFPY macilFbuence EF CF ARTr~cl Margu RTt4OTJIJ (A TP) j mcV)

-M.

?

lne~faeShell Forinmg 05 1.82 11.64 95

10.

34 Ito 155 1ntcrtnediate ShelForging 05

-4Q

-14 M97 IN4A 34 t0 048 (using &/C DatO__

Lower ShellForguig 04 1182 1.164 104 INA 34

-22 133a uuciffereniral Wcld MevcaI I.42 1-164 63 73j

-4 1125 C~irctunetentjil Wcld Nietal 1.92 1.14 77.3 90-6 56

-4 143 (a)

Initial RT%

1.t valucsý are meaured vuluts (b)

RTIvrs = RT-4D~v ARTM + Majgin (-F)

( A RTrp

= CF ýFF 1:6

2I Code of Federal Regulations. I OCFRS0. Appendix II, Reactor Vessel.Material Surveillance Program Requirements, U.S. Nuclear Regulatory Commission. Washington, D.C,

-.AST

- &L*:2ZO9, Standard Tesz Nfthodfur Conducting D,-p-Iki 0gihr Testo ct ennine XNitl-Ducgic v

Transition Temperature of Ferritie Steels, in ASTM Standards. Section 3, American Society for Testing and Materials, Philadelphia. PA.

4.

Section X1 of the ASME Boiler and Pressure Vessel Code, Appendix 4 Fracvure Toughness Criteria for Protection Against Failure

5.

ASTM Et185-82, Annual Book ofASTM Smndards. Section 12, Volume 12.02, Standard Praetice for Conducting Surveillance Tots for Light-Water Cooled Nuclear Power Reactor Vessels.

6.

Regulatory Guide 1.99. Rcvision,2, Raditation Embritnfement ofReaotor Ksel Afaterials, US.

Nuclear Regulatory Commission. May 1988.

7, WCAP-15320,Analysis of Caps We Y From Ahe Tennessee Va ley Authority Sequoyah Unit 2 A~eator tesrol Radi~nra Surveiflnnc# Prn~grnni TI L aubhamn et-aL. D~ated November 1999.

81 CVGRAPH, Hyperbolic Tangent Curve-Fitting Program Version 4.1. developed by ATI Consulting, March 1999.

9.

WCAP-1 4040-NP-A, Revision 2, "Methodology used to Develop Cold Overpressure Mitigating System Sepoints and RCS Heatup and Cookdown Limit Curves", J.D. Andrachek. et. al., January L996.

1I0 WCAP-15321, Revision 1, "Sequoyah Unit 2 Heatup and CooldowtLimit Curves for Norrmal Operation and PTLR Support Documentation", J.H-Ledger, eal... Apnl 2001.

11t.

ASME Code Case N-640, "Alternatrie Reference Fracture Toughness for Development of P-T Limit Curves for Section X1, Division I" February 26, 1999,

12.

WCAP-15315, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation For Operating PWR and BWR Plants", W Bamford, et.al.. October 1999.

13, Weatinghouse Letter to TVA, TVA 105, "Cold Overprecsure Mitigation System Code Ca'ie and Delm-P Calculation", May 19, 1993.

14.

Calculation SQN-IC-014, Demonstratd Accuracy Calculation for Cold Overpressure Protection System".

17

ENCLOSURE 5 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQN)

UNITS 1 AND 2 TOPICAL REPORTS WCAP-15293, Revision 1 (Unit 1)

WCAP-15321, Revision 1 (Unit 2)

Westhghouse Non-Proprietary Class 3 Sequoyah Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation Westinghouse Electric Company LLC

i WCAP-15293, Revision 1 Sequoyah Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation J. H. Ledger April 2001 Prepared by the Westinghouse Electric Company LLC for the Tennessee Valley Authority Approved:

,/

C. H. Boyd, Manager Engineering and Materials Technology Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355

ii PREFACE This report has been technically reviewed and verified by:

T. J. Laubham Revision 1:

An error was detected in the "OPERLIM" Computer Program that Westinghouse uses to generate pressure temperature (PT) limit curves. This error potentially effects the heatup curves when the 1996 Appendix G Methodology is used in generating the PT curves. It has been determined that WCAP-15293 Rev. 0 was impacted by this error, Thus, this revision provides corrected curves from WCAP-15293 Rev. 0.

Note that only the 60SF/hr heatup curves were affected by this error. The 1000F/hr heatup and all cooldown curves were not affected by the computer error and thus remain valid.

,f - --.

iii TABLE OF CONTENTS LIST O F TA B LE S..................................................................................................................................

iv LIST O F FIG U R E S.................................................................................................................................

v EX ECUTIV E SU M M ARY.....................................................................................................................

vi I

IN TR O D U C TIO N......................................................................................................................

1 2

FRACTURE TOUGHNESS PROPERTIES..........................................................................

2 3

CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS.............. 6 4

CALCULATION OF ADJUSTED REFERENCE TEMPERATURE....................................

10 5

HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES.....................

15 6

R E FE REN C ES.................................................................................

................................ 30 APPENDIX A: LTOPS SETPOINTS..................................................................................................

A-0 APPENDIX B: PRESSURIZED THERMAL SHOCK (PTS) RESULTS.......................................

B-0 APPENDIX C: CALCULATED FLUENCE DATA.............................................................................

C-0 APPENDIX D: UPDATED SURVEILLANCE MATERIAL 30 IT-LB TRANSITION TEMPERATURE SHIFTS AND UPPER SHELF ENERGY DECREASES................ D-0 APPENDIX E: REACTOR VESSEL BELTLINE MATERIAL PROJECTED END OF LICENSE UPPER SHELF ENERGY VALUES......................................................................

E-0 APPENDIX F: UPDATED SURVEILLANCE CAPSULE REMOVAL SCHEDULE..........................

F-0 APPENDIX G: ENABLE TEMPERATURE CALCULATIONS AND RESULTS.........................

G-0

iv LIST OF TABLES Table I Reactor Vessel Beltline Material Unirradiated Toughness Properties.............................

3 Table 2 Calculation of Chemistry Factors using Sequoyah Unit 1 Surveillance Capsule Data........ 4 Table 3 Summary of the Sequoyah Unit 1 Reactor Vessel Beltline Material Chemistry Factors...... 5 Table 4 Neutron Fluence Projections at Key Locations on the Reactor Vessel Clad/Base Metal Interface (10' n/cm2, E > 1.0 MeV)................................................................ II Table 5 Summary of the Vessel Surface, 1/4T and 3/4T Fluence Values used for the Generation of the 32 EFPY Heatup/Cooldown Curves...................................................

11 Table 6 Summary of the Calculated Fluence Factors used for the Generation of the 32 EFPY H eatup and Cooldown Curves.......................................................................................

11 Table 7 Integrated Neutron Exposure of the Sequoyah Unit I Surveillance Capsules Tested To D ate.............................................................................................................

12 Table 8 Calculation of the ART Values for the l/4T Location @ 32 EFPY............................. 13 Table 9 Calculation of the ART Values for the 3f4T Location @ 32 EFPY.............................. 13 Table 10 Summary of the Limiting ART Values Used in the Generation of the Sequoyah Unit 1 Heatup/Cooldown Curves..........................................................................................

14 Table 11 32 EFPY Heatup Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors).........................................................

22 Table 12 32 EFPY Cooldown Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors).........................................................

24 Table 13 32 EFPY Heatup Curve Data Points Using 1996 App. G (with Uncertainties for Instrumentation Errors of I0°F and 60psig)..................................

26 Table 14 32 EFPY Cooldown Curve Data Points Using 1996 App. G (with Uncertainties for Instrumentation Errors of 100F and 60psig).................................. 28

V LIST OF FIGURES Figure 1 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 60*F/hr) Applicable for the First 32 EFPY (Without Margins for Instrumentation Errors)..........................................................

16 Figure 2 Scquoyah Unit I Reactor Coolant System Heatup Limitations (Heatup Rate of 1 00°F/hr) Applicable for the First 32 EFPY (Without Margins for Instrumentation Errors)..........................................................

17 Figure 3 Sequoyah Unit I Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 1 00°F/hr) Applicable for the First 32 EFPY (Without Margins for Instrumentation Errors)............................................................

18 Figure 4 Sequoyah Unit I Reactor Coolant System Heatup Limitations (Heatupp Rate of 60°F/hr) Applicable for the First 32 EFPY (With Margins for Instrumentation Errors of 10°F and 60psig)..................................

19 Figure 5 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 100*F/hr) Applicable for the First 32 EFPY (With Margins for Instrumentation Errors of I OF and 60psig).................................

20 Figure 6 Sequoyah Unit I Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°Flhr) Applicable for the First 32 EFPY (With Margins for Instrumentation Errors of 10°F and 60psig).................................

21

vi EXECUTIVE

SUMMARY

This report provides the methodology and results of the generation of heatup and cooldown pressure temperature limit curves for normal operation of the Sequoyah Unit I reactor vessel. In addition, Pressure Temperature Limits Report (PTLR) support information, such as LTOPS Setpoint, PTS, EOL USE and Withdrawal Schedule, is documented herein in Appendices. The PT curves were generated based on the latest available reactor vessel information (Capsule Y analysis, WCAP-15224173 and the latest Pressure Temperature (P-T) Limit Curves from WCAP-12970113]). The Sequoyah Unit I heatup and cooldown pressure-temperature limit curves have been updated based on the use of the ASME Code Case N-640131, which allows the use of the Kk methodology, and a justification to lower the reactor vessel flange temperature requirement (Reference WCAP-15315['"]).

1 INTRODUCTION Heatup and cooldown limit curves are calculated using the adjusted RTTyr (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RTNDT of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ARTNDT, and adding a margin. The unirradiated RTNDT is designated as the higher of either the drop weight nil ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 600F.

RTmT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RT*NT at any time period in the reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RTrr (IRT*,T). The extent of the shift in RTmT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials."411 Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTNDT + ARTNDT + margins for uncertainties) at the 1/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region measured from the clad/base metal interface.

The heatup and cooldown curves documented in this report were generated using the most limiting ART values and the NRC approved methodology documented in WCAP-14040-NP-A, Revision 2123, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" with exception of the following: 1) The fluence values used in this report are calculated fluence values, not the best estimate fluence values (See Appendix B). 2) The Kic critical stress intensities are used in place of the Kr, critical stress intensities. This methodology is taken from approved ASME Code Case N-640 131. 3) The reactor vessel flange temperature requirement has been reduced.

Justification has been provided in WCAP-153151"1. 4) The 1996 Version ofAppendix G to Section XI141 will be used rather than the 1989 version.

WCAP-15293

2 2

FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Standard Review Planl5l. The beltlne material properties of the Sequoyah Unit I reactor vessel is presented in Table 1.

Best estimate copper (Cu) and nickel (Ni) weight percent values used to calculate chemistry factors (CF) in accordance with Regulatory Guide 1.99, Revision 2, are provided in Table 3. Additionally,, surveillance capsule data is available for four capsules (Capsules T, U, X and Y) already removed from the Sequoyah Unit I reactor vessel. This surveillance capsule data was also used to calculate CF values per Position 2.1 of Regulatory Guide 1.99, Revision 2. These CF values are presented in Table 2 and 3.

The NRC Standard Review Plan and Regulatory Guide 1.99, Revision 2 methodology used to develop the heatup and cooldown curves documented in this report is the same as that documented in WCAP-14040, Revision 2.

WCAP-15293

3 TABLE I Reactor Vessel Beltline Material Unirradiated Toughness Properties Material Description Cu (%)

Ni(%)

Initial RTNDT(a)

Intermediate Shell Forging 05 0,15 0.86 40OF (Heat # 980807/281489)

Lower Shell Forging 04 0.13 0.76 730F (Heat # 980919/281587)

Surveillance Weld"')

=

0.387 0.11 Rotterdam Testc".&

) ::

0.30 Rotterdam Test(* )

0.25 Rotterdam Testt"e 0.46 Best Estimate of the Intermediate to Lower Shell 0.35 0.11

-40OF Forging Circumferential Weld Seam W05(d-"

Notes:

(a) The Initial RT=T values are measured values (b) These copper and nickel values are best estimate values for only the surveillance weld metal and is the average of three data points [0.424 (WCAP-10340, Rev.1), 0.406 (WCAP-10340, Rev.1), 0.33 (WCAP-8233) copper and 0.084 (WCAP-10340, Rev.l), 0.085 (WCAP-10340, Rev.1), 0.17 (WCAP-8233) nickel.]. These values are treated as one data point in the calculation of the best estimate average for the inter, to lower shell circ. weld shown above. Originally the 0.424 !0.406 and 0.084 /0.085 values were reported as single points, 0.41 - 0.42 and 0.08 (Per WCAP-10340, Rev. l[6), but it is actually made up of two data points. Sample TW58 from capsule T was broken into two samples, TW58a and TW58b, thus providing the two data points.

(c) From NRC Reactor Vessel Integrity Database (RVID) and ultimately from Rotterdam Weld Certifications.

(d) Circumferential Weld Seam W05 was fabricated with weld wire type SMIT 40, Heat # 25295, Flux type SMIT

89. lot # 2275. The surveillance weld was fabricated with weld wire type SMIT 40, Heat # 25295, Flux type SMIT 89, lot # 1103 and is representative of the intermediate to lower shell circumferential weld.

(e) The surveillance weld and the three Rotterdam tests are averaged together for the Best Estimate of the Intermediate to Lower Shell Forging Circumferential Weld Scam.

The chemistry factors were calculated using Regulatory Guide 1.99 Revision 2, Positions 1.1 and 2.1.

Position 1.1 uses the Tables from the Reg. Guide along with the best estimate copper and nickel weight percents. Position 2.1 uses the surveillance capsule data from all capsules withdrawn to date. The fluence values used to determine the CFs in Table 2 are the calculated fluence values at the surveillance capsule locations. Hence, the calculated fluence values were used for all cases.

The measured ARTNDT values for the weld data were adjusted using the ratio procedure given in Position 2.1 of Regulatory Guide 1.99, Revision 2. All fluence values were obtained from the recent Sequoyah Unit 1 capsule analysis1 ir which calculated the fluences using the ENDFIB-VI scattering cross-section data set.

The fluence values used are also documented in Appendix C of this report.

WCAP-15293

4 TABLE 2 Calculation of Chemistry Factors using Sequoyah Unit I Surveillance Capsule Data Material Capsule Capsule f)

FF*'

ARTNT":)

FF*ARTNDT FF 2 Lower Shell T

2.61E+18 0.63 67.52OF 42.54OF 0.40 Forging 04 U

7.96E+18 0.94 109.70F 103.120F 0.88 (Tangential)

X 1.32E+19 1.08 145.12°F 156.73 OF 1.16 Y

2.19E+19 1.21 129.870F 157.14OF 1.47 Lower Shell T

2.61E+18 0.63 50.590F 31.87°F 0.40 Forging 04 U

7.96E+18 0.94 67.59OF 63.53OF 0.88 (Axial)

X 1.32E+19 1.08 103.34*F 11 1.61°F 1.16 Y

2.19E+19 1.21 133.35°F 161.350F 1.47 SUM:

827.89OF 7.82 CF6. = ZXF RTND)

- Z( FF2) = (827.89) + (7.82) = 105.9 0F Surveillance Weld T

2.61E+18 0.63 115.0 0F 72.50F 0.40 Matcrialjd)

U 7.96E+18 0.94 130.40F*

122.6,0 F 0.88 X

l.32E+19 1.08 143.1 0F 154.5 0F 1.16 Y

2.19E+19 1.21 147.40F 178.4oF 1.47 SUM:

528.0OF 3.91 CF &. Weld= *XFF

RTNrT) + T( FF2) = (528.0°F) + (3.91) = 135.0 0F Notes:

(a) f = Calculated fluence from capsule Y dosimetry analysis results (7), (X 1011 n/cm 2, E > 1.0 MeV).

(b)

FF = fluence factor = f&o - 0.'1-1).

(c)

ARTv,* values are the measured 30 ft-lb shift values taken from App. B of Ref. 7, rounded to one decimal point.

(d)

The surveillance weld metal ARTur values have been adjusted by a ratio factor of 0.90.

WCAP-15293

TABLE 3 Summary of the Sequoyah Unit I Reactor Vessel Beitline Material Chemistry Factors Material Reg. Guide 1.99, Rev. 2 Reg. Guide 1.99, Rev. 2 Position 1. 1 CF's Position 2.1 CF's Intermediate Shell Forging 05 115.60F Lower Shell Forging 04 95OF 105.90F Circumferential Weld W05 161.3 0F 135.0°F (Heat # 25295)

Surveillance Weld Metal 178.70F (Heat # 25295)

WCAP-15293

6 3

CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 3.1 Overall Approach The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K1, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, Kk, for the metal temperature at that time. KI, is obtained from the reference fracture toughness curve, defined in Code Case N-640, "Alternative Reference Fracture Toughness for Development of PT Limit Curves for Section XI" 3 &41 of the ASMEAppendix G to Section XI. The K10 curve is given by the following equation:

Kfr =33.2 +20.734 *"OO02(T-RT.)]

(1)

where, K1,

=

reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTND This KI, curve is based on the lower bound of static critical K, values measured as a function of temperature on specimens of SA-533 Grade B Classl, SA-508-1, SA-508-2, SA-508-3 steel.

3.2 Methodology for Pressure-Temperature Limit Curve Development The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

C* K..- + Ku < Kio (2)

where, K

=

stress intensity factor caused by membrane (pressure) stress Kr

=

stress intensity factor caused by the thermal gradients KI,

=

function of temperature relative to the RTmT of the material C

=

2.0 for Level A and Level B service limits C

1.5 for hydrostatic and leak test conditions during which the reactor core is not critical WCAP-15293

7 For membrane tension, the corresponding K, for the postulated defect is:

Kt. = M, x (pRP I t)

(3) where, Mm for an inside surface flaw is given by:

Mm

=

1.85 for 6- < 2, M,

=

0.92647r for 2541-,<3.464, M,

3.21 for f- > 3.464 Similarly, Mm for an outside surface flaw is given by:

Mm

=

1.77 for -ft < 2, M =

0.893 Ft for 2*47*-

3.464, Mm

=

3.09 for Ft > 3.464 and p = internal pressure, Ri = vessel inner radius, and t = vessel wail thickness.

For bending stress, the corresponding K, for the postulated defect is:

K0

= Mb

Maximum Stress, where Mb is two-thirds of M.

The maximum K, produced by radial thermal gradient for the postulated inside surface defect of G-2120 is Kj, = 0.953xl 0"3 x CR x te5, where CR is the cooldown rate in *F/hr., or for a postulated outside surface defect, KI, = 0.753x10l x HU x tf5, where HU is the heatup rate in TF/hr.

The through-wall temperature difference associated with the maximum thermal K, can be determined from Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from Fig.

G-2214-2 for the maximum thermal K1.

(a)

The maximum thermal K, relationship and the temperature ielationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(1) and (2).

(b)

Alternatively, the K, for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a %-thickness inside surface defect using the relationship:

/, = (1.0359C 0 + 0.6322C, + 0.4753C2 + 0.3855C 3) *.

(4)

WCAP-15293

8 or similarly, Krr during heatup for a %-thickness outside surface defect using the relationship:

Ki, = (1.043Co + 0.630C, + 0.481C2 + 0.40IC3) * *

(5) where the coefficients Co, C1, C2 and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:

cr(x) = Co + Ct(x / a) + C2(x / a)2 + C3(x / a)3 (6) and x is a variable that represents the radial distance from the appropriate (i.e., inside or outside) surface to any point on the crack front and a is the maximum crack depth.

Note, that equations 3, 4 and 5 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. No other changes were made to the OPERLIM computer code with regard to P-T calculation methodology. Therefore, the P-T curve methodology is unchanged from that described in WCAP-14040, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldwon Limit Curves"12' Section 2.6 (equations 2.6.2-4 and 2.6.3-1) with the exceptions just described above.

At any time during the heatup or cooldown transient, KI, is determined by the metal temperature at the tip of a postulated flaw at the 1/4T and 3/4T location, the appropriate value for RTmT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, Kir, fbr the reference flaw are compute& From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT (temperature) developed during cooldown results in a higher value of K1. at the 114T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in Ki. exceeds Krt, the calculated allowable pressure during cooldow,n will be greater than the steady-state value.

WCAP-15293

9 The above procedures are needed because there is no direct control on temperature at the 1/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.

Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure: The metal temperature at the crack tip lags the coolant temperature; therefore, the K1, for the 1/4T crack during heatup is lower than the Kic for the l/4T crack during steady state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower K1, values do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the I/4T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4T flaw located at the 1/4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations,.the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.

3.3 Closure Head/Vessel Flange Requirements 10 CFR Part 50, Appendix G addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RTr by at least 120°F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3107 psi), which is 621 psig for Sequoyah Unit 1 reactor vessel.

However, per WCAP-15315, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation For Operating PWR and BWR Plants"ttA1, this requirement is no longer necessary when using the methodology of Code Case N-64013). Hence, Sequoyah Unit I heatup and cooldown limit curves will be generated without flange requirements included.

WCAP-15293

10 4

CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:

ART = Initial RTmNT + ARTNDT + Margin (7)

Initial RTNT is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel CodeIR]. If measured values of initial RTNT for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.

ARTN*Tr is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:

ARTNDT

= CF

2S-O.*-o.8 )

(8)

To calculate ARTmr at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth.

=f~

e (-0 24x)

(9) where x inches (vessel beltline thickness is 8.45 inches) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then placed in Equation 9 to calculate the ARTNOT at the specific depth.

The Westinghouse Radiation Engineering and Analysis Group evaluated the vessel fluence projections as a part of WCAP-15224 and are also presented in a condensed version in Table 4 of this report. The evaluation used the ENDF/B-VI scattering cross-section data set. This is consistent with methods presented in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves"' 2l. Table 4 contains the calculated vessel surface fluences values at various azimuthal locations and Tables 5 and 6 contains the I 4T and 3/4T calculated fluences and fluence factors, per the Regulatory Guide 1.99, Revision 2, used to calculate the ART values for all beltline materials in the Sequoyah Unit I reactor vessel. Additionally, the surveillance capsule fluence values are presented in Table 7.

WCAP-15293

ii TABLE 4 Neutron Fluence Projections at Key Locations on the Reactor Vessel Clad/Base Metal Interface (10" n/cm 2, E > 1.0 MeV)

Azimuthal Location EFPY 00 150 300 450 10.03 0.205 0.321 0.409 0.637 20 0.387 0.5%

0.761 1.19 32 0.605 0.928 1.19 1.84 48 0.896 75 2.72 TABLE 5 Summary of the Vessel Surface, 1/4T and 3/4T Fluence Values used for the Generation of the 32 EFPY Heatup/Cooldown Curves Material Intermediate Shell Forging 05 Lower Shell Forging 04 Circumferential Weld Seam W05 (Heat 25295)

Note:

(a) 1/4T and 3/4T = F(s,ý) *Ce(,24X) where x is the depth into the vessel wall (i.e. 8.45*0.25 or 0.75)

TABLE 6 Summary of the Calculated Fluence Factors used for the Generation of the 32 EFPY Heatup and Cooldown Curves EFPY 1/4T FF 3/4T FF 32 1.029 0.747 WCAP-15293 II

12 TABLE 7 Integrated Neutron Exposure of the Sequoyah Unit 1 Surveillance Capsules Tested To Date Capsule Fluence T

2.61 x 10' n/cm 2, (E > 1.0 MeV)

U 7.96 x 10's n/cm2, (E > 1.0 MeV)

X 1.32 x 10' 9 n/cm2, (E > 1.0 MeV)

Y 2.19 x 10'9 n/cm2, (E > 1.0 MeV)

Margin is calculated as, M = 2 ý, 2 -+a o

. The standard deviation for the initial RTNT margin term, is oi 0°F when the initial RTNDT is a measured value, and 17'F when a generic value is available. The standard deviation for the ARTNDT margin term, aA, is 17°F for plates or forgings, and 8.5°F for plates or forgings when surveillance data is used. For welds, aa is equal to 28*F when surveillance capsule data is not used, and is 14'F (half the value) when credible surveillance capsule data is used. ca need not exceed 0.5 times the mean value of ARTs.rr.

Based on the surveillance program credibility evaluation presented in Appendix D to WCAP-15224, the Sequoyah Unit 1 surveillance program data is non-credible. In addition, following the guidance provided by the NRC in recent industry meeting, Table Chemistry Factor for the lower shell forging 04 was determined to be non-conservative. Hence, the adjusted reference temperature (ART) must be calculated using Position 2.1 along with the full margin term. Both Regulatory Guide 1.99, Revision 2, Position 1.1 and 2.1 have been shown herein. Contained in Tables 8 and 9 are the calculations of the 32 EFPY ART values used for generation of the heatup and cooldown curves.

WCAP-15293

13 TABLE 8 C a~cu la-ion eA T V h,

'*,. *C

T r A..I

^.,4 I

L

,W *.) _,*r r Material RG 1.99 CF FF (1RT-)

ARTmTr)

Margin(4 )

ART<2 "

R2 Method (0F)

I (OF)

I (0$

(0F-)

(0F)

T

. I.

-I

'IC I

J I

.termediate She)) Forging 05 Lower Shell Forging 04 Intermediate to Lower Shell Circumferential Weld Seam Notes:

Position 1.1 Position 1. 1 Position 2.1 Position 1.1 Position 2.1 115.6 95 105.9 161.3 135.0 1.029 1.029 1.029 1.029 1.029 40 73 73

-40

-40 119.0 97.8 109.0 166.0 138.9 Initial RTMT values measured values.

ART = Initial RTNDT + ARTN-T + Margin (F)

ARTNDr = CF

FF M = 2 *(ai2 + a) 112 TABLE 9 Calculation of the ART Values for the 3/4T Location @ 32 EFPY Circumferential Weld Seam N~otes-34 34 34 56 56 193 205 216 I92 155 Initial RT,,,nr values measured values.

ART = Initial RTNDT + ARTNDT + Margin (0F)

ARTNDT = CF

FF M-2* (a*2 + C,2)12 WCAP-15293 N

(1)

(2)

(3)

(4)

(1)

(2)

(3)

(4)

14 The lower shell forging 04 is the limiting beltline material for the 1/4T and 314T case (See Tables 8 and 9).

Contained in Table 10 is a summary ofthe limiting ARTs to be used in the generation of the Sequoyah Unit I reactor vessel heatup and cooldown curves.

TABLE 10 Summary of the Limiting ART Values Used in the Generation of the Sequoyah Unit I Heatup/Cooldown Curves WCAP-15293

15 5

HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel beltline region using the methods discussed in Sections 3.0 and 4.0 of this report. This approved methodology is also presented in WCAP-14040-NP-A, Revision 2 with exception to those items discussed in Section 1 of this report Figures 1, 2, 4 and 5 present the heatup curves with (10*F and 60 psig) and without margins for possible instrumentation errors using heatup rates of 60 and 100°Ffhr applicable for the first 32 EFPY. Figures 3 and 6 present the cooliown curves with (10*F and 60 psig) and without margins for possible instrumentation errors using cooldown rates of 0. 20,40, 60 and 100°F/hr applicable for 32 EFPY.

Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 1 through 6. This is in addition to other criteria which must be met before the reactor is made critical, as discussed below in the following paragraphs.

The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figures 1, 2, 4 and 5. The straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic test is defined in Code Case N-64031 (approved in February 1999) as follows:

1.5 Ki < Kl,

where, KI is the stress intensity factor covered by membrane (pressure) stress, K,. = 33.2 + 20.734 e 0'02 (T'- RrNDT)I T is the minimum permissible metal temperature, and RTmw is the metal reference nil-ductility temperature.

The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Reference 10. The pressure-temperature limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40'F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section 3.0 of this report. For the heatup and cooldown curves without margins for instrumentation errors, the minimum temperature for the in service hydrostatic leak tests for the Sequoyah Unit 1 reactor vessel at 32 EFPY is 277°F. The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40'F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.

Figures I through 6 define all of the above limits for ensuring prevention of nonductile failure for the Sequoyah Unit I reactor vessel. The data points used for the heatup and cooldown pressure-temperature limit curves shown in Figures 1 through 6 are presented in Tables II and 14.

WCAP-15293

16 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 32 EFPY:

1/T, 216-F 34T. 186°F 2500o OpedimVersimZ.1 Ruri;116421 2250 Lek TetLit-2000 opabl Acceptable Operation Operation 1750o--

S1~

~ ~ ~~60 0 O e g S I F r i 1 -

0 D g. / r 41-50

-0 DogMi

,1500 2E 1250 750 Limit based an 500 L. -inservice hydrostatic test Soo temperature (277 F)p for the service period up to 32 EFPY 250-Tm 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Dog. F)

FIGURE I Sequoyah Unit I Reactor Coolant System Heatup Limitations (Heatup Rate of 60°F/hr)

Applicable for the First 32 EFPY (Without Margins for Instrumentation Errors)

WCAP-15293

MATERIAL PROPERTY BASIS LIMiTING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 32 EFPY:

1/4/4T, 2160F 3/4T, 186-F 2500 lop-Vemlm.1 R=11542 L~ea k Test Limi 2250 Unacceptable 2000 Opemrton L

1750 0.1500-IHeatup Rate 1100 Deg. Wairk-..

E 1250

.1o000 0

Si

/i 50D A:.

t to Sol 250 Te-,

Boltup 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

FIGURE 2 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 100PF/hr)

Applicable for the First 32 EFPY (Without Margins for Instrumentation Errors)

WCAP-15293

-N 17

18 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 32 EFPY:

1/4T, 216°F 3/4T, 186-F 2500 2

1 01 1214 2250 C,,

S2000 1750 1500 1250 1000 750 500 250 0-ju

uIu I:u D

U Z 2bO 300 350 400 450 560 Moderator Temperature (Deg.F)

Figure 3 Sequoyah Unit I Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 1000F/hr) Applicable for the First 32 EFPY (Without Margins for Instrumentation Errors)

WCA -15293 C,,

-o C.,)

19 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 32 EFPY:

1/T, 2160F 3/4T, 1860F 2500 2250 2000 1750 to I. 1500 CL 71250 0

750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

FIGURE 4 Sequoyah Unit I Reactor Coolant System Heatup Limitations (Heatup Rate of 600F/hr)

Applicable for the First 32 EFPY (With Margins for Instrumentation Errors of 10"F and 60 psig)

WCAP-15293 J.

20 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 32 EFPY:

1/T, 216OF 3/4T, 186°F 2500 2250 2000 1750 Ca 1500 0

2 1250 CL

0.

1 750 250-0 0

50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

FIGURE 5 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of I00°F/hr)

Applicable for the First 32 EFPY (With Margins for Instrumentation Errors of 100F and 60 psig)

WCAP-15293

MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 32 EFPY:

1/4T, 216-F 3/41, 186-F

22500 S2000 -

a, L.

Co Co 0,

1750 1500 1250 1000 750 500 250 0

Moderator Temperature (Deg.F)

Figure 6 Sequoyah Unit I Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 009F/hr) Applicable for the First 32 EFPY (With Margins for Instrumentation Errors of 10F and 60 psig)

WCAP-15293 LI

22 TABLE I1 32 EFPY Heatup Curve Data Points Using 1996 App.G (without Uncertainties for Instrumentation Errors)

WCAP-15293

23 TABLE 11 (Continued) 32 EFPY Heatup Curve Data Points Using 1996 App.G (without Uncertainties for Instrumentation Errors)

Heatup Curves WCAP-15293 60 Heatup 60 Limit Critical 100 Heatup 100 Limit Critical Leak Test Limit T

P T

P 255 1329 255 985 260 1414 260 1041 265 1509 265 1104 270 1591 270 1172 275 1677 275 1248 280 1772 280 1332 285 1877 285 1424 290 1993 290 1526 295 2120 295 1638 300 2261 300 1762 305 2417 305 1898 310 2048 315 2214 320 2397

24 TABLE 12 32 EFPY Cooldown Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors)

WCAP-15293 Cooldown Curves Steady State 20F 40F 6OF 100F T

P T

P 50 0

50 0

50 615 50 567 50 519 50 470 50 367 55 616 55 569 55 520 55 471 55 368 60 618 60 570 60 522 60 472 60 369 65 620 65 572 65 524 65 474 65 371 70 621 70 574 70 525 70 476 70 373 75 624 75 576 75 528 75 478 75 375 80 626 80 579 80 530 80 480 80 378 85 629 85 581 85 533 85 483 85 381 90 631 90 584 90 536 90 486 90 385 95 635 95 587 95 539 95 490 95 389 100 638 100 591 100 543 100 494 100 393 105 642 105 595 105 547 105 498 105 398 110 646 110 600 110 552 110 503 110 404 115 651 115 605 115 557 115 509 115 411 120 656 120 610 120 563 120 516 120 418 125 662 125 616 125 570 125 523 125 427 130 668 130 623 130 577 130 531 130 436 135 676 135 631 135 585 135 539 135 447 140 683 140 639 140 594 140 549 140 459 145 692 145 648 145 604 145 560 145 472 150 702 150 659 150 616 150 572 150 487 155 712 155 670 155 628 155 586 155 503 160 724 160 683 160 642 160 601 160 521 165 737 165 697 165 657 165 618 165 542 170 751 170 712 170 674 170 637 170 565 175 767 175 729 175 693 175 657 175 590 180 784 180 748 180 714 180 680 180 618 185 803 185 769 185 737 185 706 185 650 190 824 190 793 190 762 190 734 190 684 195 848 195 819 195 791 195 765 195 723 200 874 200 847 200 822 200 800 200 766

25 TABLE 12 - (Continued) 32 EFPY Cooldown Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors)

Cooldown Curves WCAP-15293 Steady State 20F 0F 60F 10OF T

P T

P 205 903 205 879 205 857 205 839 205 814 210 934 210 913 210 896 210 881 210 866 215 969 215 952 215 938 215 929 215 925 220 1008 220 995 220 985 220 981 220 990 225 1051 225 1042 225 1038 225 1039 230 1098 230 1094 230 1096 235 1150 240 1208 245 1272 250 1343 255 1420 260 1507 265 1602 270 1707 275 1823 280 1952 285 2094 290 2251 295 2424

26 TABLE 13 32 EFPY Heatup Curve Data Points Using 1996 App.G (with Uncertainties for Instrumentation Errors of 10*F and 60 psig)

WCAP-15293

27 TABLE 13 (Continued) 32 EFPY Heatup Curve Data Points Using 1996 App.G (with Uncertainties for Instrumentation Errors of 10F and 60 psig)

Heatut, Curves WCAP-15293 N

60 Heatup 60 Limit Critical 100 Heatup 100 Limit Critical Leak Test Limit T

P I

T P

255 1121 350 2201 255 828 370 2337 260 1191 355 2357 260 874 265 1269 265 925 270 1354 270 981 275 1449 275 1044 280 1531 280 1112 285 1617 285 1188 290 1712 290 1272 295 1817 295 1364 300 1933 300 1466 305 2060 305 1578 310 2201 310 1702 315 2357 315 1838 320 1988 325 2154 330 2337

28 TABLE 14 32 EFPY Coold)*wn Curve Data Points Using 1996 App. G (with Uncertainties for Instrumentation Errors of 10OF and 60 psig)

Cooldown Curves Steady State 20F 40F 60F IOOF T

P T

p T

p 50 0

50 0

50 552 50 503 50 457 50 408 50 305 55 553 55 505 55 458 55 409 55 306 60 555 60 507 60 459 60 410 60 307 65 556 65 509 65 460 65 411 65 308 70 558 70 510 70 462 70 412 70 309 75 560 75 512 75 464 75 414 75 311 80 561 80 514 80 465 80 416 80 313 85 564 85 516 85 468 85 418 85 315 90 566 90 518 90 470 90 420 90 318 95 569 95 521 95 473 95 423 95 321

]00 571 100 524 100 476 100 426 100 325 105 575 105 527 105 479 105 430 105 329 110 578 110 531 110 483 110 434 110 333 115 582 115 535 115 487 115 438 115 338 120 586 120 540 120 492 120 443 120 344 125 591 125 545 125 497 125 449 125 351 130 596 130 550 130 503 130 456 130 358 135 602 135 556 135 510 135 463 135 367 140 608 140 563 140 517 140 471 140 376 145 616 145 571 145 525 145 479 145 387 150 623 150 579 150 534 150 489 150 399 155 632 155 588 155 544 155 500 155 412 160 642 160 599 160 556 160 512 160 427 165 652 165 610 165 568 165 526 165 443 170 664 170 623 170 582 170 541 170 461 175 677 175 637 175 597 175 558 175 482 180 691 180 652 180 614 180 577 180 505 185 707 185 669 185 633 185 597 185 530 190 724 190 688 190 654 190 620 190 558 195 743 195 709 195 677 195 646 195 590 200 764 200 733 200 702 200 674 200 624 205 788 205 759 205 731 205 705 205 663 210 814 210 787 210 762 210 740 210 706 215 843 215 819 215 797 215 779 215 754 WCAP-15293

29 TABLE 14 - (Continued) 32 EFPY Cooldown Curve Data Points Using 1996 App. G (with Uncertainties for Instrumentation Errors of 100F and 60 psig)

WCAP-15293 Cooldown Curves Steady State 20F 40F 60F lOOF T

P T

P 220 874 220 853 220 836 220 821 220 806 225 909 225 892 225 878 225 869 225 865 230 948 230 935 230 925 230 921 230 930 235 991 235 982 235 978 235 979 240 1038 240 1034 240 1036 245 1090 250 1148 255 1212 260 1283 265 1360 270 1447 275 1542 280 1647 285 1763 290 1892 295 2034 300 2191 305 2364

30 6

REFERENCES

1.

Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S.

Nuclear Regulatory Commission, May 1988.

2.

WCAP-14040-NP-A, Revision 2, "Methodology used to Develop Cold Overpressure Mitigating system Setpoints and RCS Heatup and Cooldown Limit Curves", 13.D. Andrachek, et. al., January 1996.

3.

ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1", February 26, 1999.

4.

Section XI of the ASME Boiler and Pressure Vessel Code, Appendix Q, "Fracture Toughness Criteria for Protection Against Failure.", Dated 1989 & December 1995.

5.

"Fracture Toughness Requirements", Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981.

6.

WCAP-10340, Revision 1, "Analysis of Capsule T from the Tennessee Valley Authority Sequoyah Unit 1 Reactor Vessel Radiation Surveillance Program", S.E. Yanichko, et. al., February 1984.

7.

WCAP-15224, "Analysis of Capsule Y From the Tennessee Valley Authority Sequoyah Unit I Reactor Vessel Radiation Surveillance Program", T.J. Laubham, et. al., Dated June 1999.

8.

1989 Section III, Division I of the ASME Boiler and Pressure Vessel Code, Paragraph NB-233 1, "Material for Vessels."

9.

CVýGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATI Consulting, March 1999.

10.

Code of Federal Regulations, 10 CFR Part 50, Appendix G "Fracture Toughness Requirements,"

U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.

11.

WCAP-7924-A, "Basis for Heatup and Cooldown Limit Curves," W. S. Hazelton, et al., April 1975.

12.

Calc. No.92-016, "WOG USE Program - Onset of Upper Shelf Energy Calculations". J. M.

Chicots, dated 11/12/92. File # WOG-108/4-18 (MUHP-5080).

13.

WCAP-12970, "Heatup and Cooldown Limit Curves for Normal Operation Sequoyah Unit I",

J.M. Chicots, et. al., Dated June, 1991.

14.

Westinghouse Letter TVA-91-242 WCAP-15293

31

15.

Westinghouse Letter TVA-91-243

16.

TVA letter No. N9664, "TASK N99-017 -Reactor Coolant System Pressure and Temperature Limit Report Development -N2N-048," W. M. Justice, August 17, 1999.

17.

TVA letter No. N9667, "TASK N99-017 - Reactor Coolant System Pressure and Temperature Limit Report Development -N2N-048," W. M. Justice, August 20, 1999.

18.

WCAP-153 15, "Reactor Vessel Closure Head / Vessel Flange Requirements Evaluation for Operating PWR and BWR Plants", W Bamford, et. al, October 1999.

19.

Westinghouse Letter TVA-93-105.

WCAP-15293

A-0 N

APPENDIX A LTOPS SETPOLNTS WCAP-15293

A-1

1.1 INTRODUCTION

Westinghouse has been requested to develop Low Temperature Overpressure Protection System (LTOPS) setpoints for Sequoyah Unit 1, for a vessel exposure of 32 EFPY. The LTOPS setpoints for the Sequoyah, Units I and 2. were last revised by Westinghouse in June of 1991 using pressure-temperature limits supplied by Tennessee Valley Authority for a vessel exposure of 16 EFPY. The results of this analysis were reported via References 14 and 15.

The June 1991 analysis was based on the September 1989 analysis which addressed: Eagle-21 implementation and Appendix G limits based on Regulatory Guide 1.99. This Section documents the development of new Sequoyah Unit I COMS setpoints for 32 EFPY.

The Low Temperature Overpressure Protection System (LTOPS) is designed to provide the capability, during relatively low temperatures Reactor Coolant System (RCS) operation (typically less than 3500F), to protect the reactor vessel from being exposed to conditions of fast propagating brittle fracture. The LTOPS is provided in addition to the administrative controls, to prevent overpressure transients and as a supplement to the RCS overpressure mitigation function of the Residual Heat Removal System (RHRS) relief valves. LTOP consists of pressurizer PORVs and actuation logic from the wide range pressure channels. Once the system is enabled, no operator action is involved for the LTOPS to perform its intended pressure mitigation function.

LTOPS setpoints are conservatively selected to prevent exceeding the pressure/temperature limits established by 10 CFR Part 50 Appendix G requirements.

Two specific transients have been defined as the design basis for LTOPS. Each of these transient scenarios assume that the RCS is in a water-solid condition and that the RHRS is isolated from the RCS. The first transient consists of a heat injection scenario in which a reactor coolant pump in a single loop is started with the RCS temperature as much as S0OF lower than the steam generator secondary side temperature and the RHRS has been inadvertently isolated. This results in a sudden heat input to the RCS from the steam generators, creating an increasing pressure transient. The second transient has been defined as a mass injection scenario into the RCS caused by the simultaneous isolation of the RHRS. isolation of letdown and failure of the normal charging flow controls to the full flow condition. The resulting mass injectiom'letdown mismatch causes an increasing pressure transient.

1.2 LTOPS SETPOINT DETERMINATION Westinghouse has developed new LTOPS setpoints for Sequoyah Unit I, based on a vessel exposure of 32 EFPY using the methodology established in WCAP-14040 (Ref. 2). This methodology maximizes the available operating margin for setpoint selection while maintaining an appropriate level of protection in support of reactor vessel integrity. Note, Appendix G pressure limit relaxation allowed by ASME Code Case N-514 was not applied.

Plant design characteristics are unchanged (i.e., heat injection and mass injection transients characteristics and related plant responses have not been altered). Therefore, a complete reanalysis is not required. The new LTOPS setpoints were developed using results of the previous heat and mass injection transient analyses.

WCAP-15293

A-2 1.2.1 Pressure Limits Selection The function of the LTOPS is to protect the reactor vessel from fast propagating brittle fracture. This has been implemented by choosing LTOPS setpoints, which prevent exceeding the limits prescribed by the applicable pressure/temperature characteristic for the specific reactor vessel material in accordance with rules given in Appendix G to 10CFR50.

The LTOPS design basis takes credit for the fact that overpressure events most likely occur during isothermal conditions in the RCS. Therefore, it is appropriate to utilize the steady-state Appendix G limit. The Sequoyah Unit 1, IOCFR50 Appendix G curve for 32 EFPY is shown by Figure A-1. This curve sets the nominal upper limit on the pressure which should not be exceeded during RCS increasing pressure transients based on reactor vessel material properties.

When a relief valve is actuated to mitigate an increasing pressure transient, the system pressure then decreases, as the relief valve releases coolant, until a reset pressure is reached where the valve is signaled to close. Note that the pressure continues to decrease below the reset pressure as the valve re-closes. The nominal lower limit on the pressure during the transient is typically established based solely on an operational consideration for the RCP #1 seal to maintain a nominal differential pressure across the seal faces for proper film-riding performance. The RCP #1 seal limit is shown in Figure 1.

-The nominal upper limit (based on the minimum of the steady-state IOCFR50 Appendix G requirement) and the nominal lower limit (based on RCP #1 seal performance criteria) create a pressure range from which the setpoints for both PORVs may be selected.

1.2.2 Mass Input Consideration For a particular mass input transient to the RCS, the relief valve will be signaled to open at a specific pressure setpoint. However, there will be a pressure overshoot during the delay time before the valve starts to move and during the time the valve is moving to the full open position. This overshoot is dependent on the dynamics of the system and the input parameters, and results in a maximum system pressure somewhat higher than the set pressure. Similarly there will be a pressure undershoot, while the valve is relieving, both due to the reset pressure being below the setpoint and to the delay in stroking the valve closed.

The previous Sequoyah analyses of multiple mass input cases were used to determine the relationship between setpressures and resulting overshootslundershoots.

1.2.3 Heat Input Consideration The heat input case is done similarly to the mass input case except that the locus of transient pressure values versus selected setpoints may be determined for several values of the initial RCS temperature. This heat input evaluation provides a range of acceptable setpoints dependent on the reactor coolant temperature, whereas the mass input case is limited to the most restrictive low temperature condition only (i.e., the mass injection transient is not sensitive to temperature).

WCAP-15293

A-3 The previous Sequoyah analyses of multiple heat input cases were used to determine the relationship between setpressures and resulting overshoots/undershoots.

1.2.4 Final Setpoint Selection Appendix 0 limits described in Section 1.2.1, were conservatively adjusted accounting for the pressure difference (AfP) between the wide range pressure transmitter and the reactor vessel limiting beltline region of 68.3 psi for 4 RCPs in operation (See Reference 19).

The results of the analyses described in Section 1.2.2 & 1.2.3, and the adjusted Appendix G limit were used to define the maximum allowable setlxpints for which the overpressure will not exceed the pressure limit applicable at a specific reactor vessel temperature. The maximum allowable setpoints are shown in Figure A-I.

Per Ref. 17. Sequoyah Demonstrated Accuracy Calculation SQN-IC014 establishes the instrument loop inaccuracy of the Sequoyah temperature and pressure instrument channels associated with the LTOPS. Previously, the LTOPS setpoints have been provided to TVA without application of instrument uncertainties.

The TVA calculation quantified the instrument channel uncertainties, applied them to the nominal Westinghouse setpoints and evaluated the result against the safety limits established by the IOCFRSO, Appendix G steady state heatup curve. The TVA demonstrated accuracy calculation will be revised to reflect the LTOPS setpoints calculated by Westinghouse under the subject task.

As such, it is not necessary for Westinghouse to include instrument uncertainties in the nominal LTOPS setpoint calculation.

Note, the heat injection results were adjusted to include 50'F thermal transport effect (difference in temperature between the RCS and steam generator at transient initiation).

The maximum allowable setpoints, adjusted to produce a smoother curve and reduced to nine data points, becomes the setpoints for PORV#2. A setpoint at a minimum temperature of 500F was selected, as requested by TVA (Ref. 16). Each of the two PORVs may have a different pressure setpoint such that only one valve will open at a time and mitigate the transient (i.e., staggered setpoints). The second valve operates only if the first fails to open on command. This design supports a single failure assumption as well as minimizing the potential for both PORVs to open simultaneously, a condition which may create excessive pressure undershoot and challenge the RCP

1 seal performance criteria.

The PORV#I selpoints were selected by adjusting the setpoint PORV#2 in relationship to the overshoots discussed in Sections 1.2.2 and 1.2.3.

The selected setpoints for PORV #1 and PORV #2 are shown in Table A-I and Figure A-2. These setpoints were evaluated using the undershoots discussed in Sections 1.2.2 and 1.2.3 to ensure that they protect against the RCP # I seal limit.

In summary, the selection of the setpoints for LTOPS considered the use of nominal upper and lower pressure limits. The upper limits are specified by the minimum of the steady-state cooldown curve as calculated in accordance with Appendix G to IOCFR50 (adjusted to account for four RCPs in operation). The lower pressure extreme is specified by the reactor coolant pump #1 seal minimum differential pressure performance criteria. The selected setpoints, shown in Table A-I WCAP-1 5293

A-4 and Figure A-2, provide protection against Appendix G, and RCP #1 seal limit violations. Note, these setpoints do not address instrumentation uncertainties.

1.3 ARMING AND ENABLE TEMPERATURES FOR LTOPS The LTOPS arming temperature is traditionally based on the temperature corresponding to when Appendix G pressure equals 2500 psia.

Based on this methodology the LTOPS arming temperature for Sequoyah conservatively continues to be 350 0F.

The enable temperature is the temperature below which the LTOPS system is required to be operable, based on vessel materials concerns. ASME Code Case N-514 requires the LTOPS to be in operation at coolant temperatures less than 200°F or at coolant temperatures less than a temperature corresponding to a reactor vessel metal temperature less than RTNDT + 500F, whichever is greater. RTmr is the highest Adjusted Reference Temperature (ART) for the limiting belt-line material at a distance one fourth of the vessel section thickness from the vessel inside surface (i.e., clad/base metal interface), as determined by Regulatory Guide 1.99, Revision 2. The minimum required enable temperature for the Sequoyah Unit 1 Reactor Vessel is 295°F at 32 EFPY of operation.

WCAP-15293

Table A-I Selected Setpoints, Sequoyah Unit I "T1rcs (Deg.F)

PORV#2 PORV#1 Setpoint (psig) Setpoint (psig) 50 490 465 100 500 475 135 540 510 175 575 540

.200 610 570 250 745 685 280 745 685 405 745 685 450 2350 2350 A-5 WCAP-15293

A-6 Figure A-i:

Sequoyah Unit I LTOPS Setpoint Selection WCAP-15293

Reactor Coolant System Pressure (pslg) 0 I

I K

I I_

I I

-j----------

I_

j -

- I...

I I

1 I

I I

r ir-I I

H--

I I

°0

.4-----_

qj1

'0 N

I qI 01 2.

I I

I a 0

S S

-4 0

S a

0 U,

a U)

.0 C

C I-4 0

-U U)

C,)

C, 0 0.

U) 0 0

0

-4 I

f I

B-O APPENDIX B PRESSURIZED THERMAL SHOCK (PTS) RESULTS WCAP-15293

N B-1 PTS Calculations:

The PTS Rule requires that for each pressurized water nuclear power reactor for which an operating license has been issued, the licensee shall have projected values of RTrs, accepted by the NRC, for each reactor vessel beltline material for the EOL fluence of the material. This assessment must specify the basis for the projected value of RTpr for each vessel beltline material, including the assumptions regarding core loading patterns, and must specify the copper and nickel contents and the fluence value used in the calculation. This assessment must be updated whenever there is a significant clange in projected values of Umzs, or upon request for a change in the expiration date for operation of the facility. (Changes to RTPrs values are considered significant if either the previous value or the current value, or both values, exceed the screening criterion prior to the expiration of the operating license, including any renewed term, if applicable, for the plant.

To verify that RTNDr, for each vessel beitline material is a bounding value for the specific reactor vessel, licensees shall consider plant-specific information that could affect the level of embrittlement. This information includes but is not limited to the reactor vessel operating temperature and any related surveillance program results. (Surveillance program results mean any data that demonstrates the embrittlement trends for the limiting beltline material, including but not limited to data from test reactors or from surveillance programs at other plants with or without surveillance program integrated per 10 CFR Part 50, Appendix H.)

Calculations:

Tables B-I and B-2 contain the results of the calculations for each of the beltline region materials in the Sequoyah Unit I Reactor Vessel. Per TVA, the EOL is 32 EFPY and the Life Extension EOL is 48 EFPY.

WCAP-15293

B-2 TABLE B-I RTmrs Calculations for Sequoyah Unit I Beltline Region Materials at 32 EFPY Malerial Fluence FF CF ARTv 3sc' Margin RTMIT(Uý RTpTs0 'l (n/c'2, E>l.0 (O0$

(0F (0F (OF)

(OF)

MeV)

Intermediate Shell Forging 05 1.84 1.167 115.6 134.9 34 40 209 Lower Shell Forging 04 1.84 1.167 95.0 110.9 34 73 218 Lower Shell Forging 04 1.84 1.167 105.9 123.6 34 73 231 (Using S/C Data)

Circumferential Weld Metal 1.84 1.167 161.3 188.2 56

-40 204 Circumferential Weld Metal 1.84 1.167 135.0 157.5 56

-40 174 (Using S/C Data)

Notes:

(a)

Initial RTmnTr values are measured values (b)

RTpes = RTorTu) + ARTrrs + Margin (OF)

(c)

ARTprs = CF

FF TABLE B-2 RTpTs Calculations for Sequoyah Unit 1 Beltline Region Materials at 48 EFPY Material Fluence FF CF ARTpr(O)

Margin RTk,,DT(L)()

RTr.s')

(ncimn' E>l.0 (0$

(0F(0F)

(OF)

MeV)

Intermediate Shell Forging 05 2.72 1.267 115.6 146.5 34 40 221 Lower Shell Forging 04 2.72 1.267 95.0 120.4 34 73 227 Lower Shell Forging 04 2.72 1.267 105.9 134.2 34 73 241 (Using S/C Data)

Circumferential Weld Metal 2.72 1.267 161.3 204.4 56

-40 220 Circumferential Weld Metal 2.72 1.267 135.0 171.0 56

-40 187 (Using S/C Data)

Notes:

(a)

Initial RTNDT values are measured values (b)

RTprs = RTNDT7 + ARTvrs + Margin (0F)

(c)

ARTp's = CF

FF WCAP-15293

B-3 All of the beltline materials in the Sequoyah Unit 1 reactor vessel are below the screening criteria values of 270TF and 300TF at 32 and 48 EFPY.

WCAP-15293

C-O APPENDIX C CALCULATED FLUENCE DATA WCAP-15293

C-1 The best estimate exposure of the Sequoyah Unit I reactor vessel presented in WCAP-15224[I was developed using a combination of absolute plant specific transport calculations and alt available plant specific measurement data. The evaluation is consistent with the methodology accepted by the NRC and documented in WCAP-14040-NP-A1 21.

Combining this measurement data base with the plant-specific calculations, the best estimate vessel exposure is obtained from the following relationship:

(Detras.

= K '$c,.

where:

Bd EI. =

The best estimate fast neutron exposure at the location of interest.

K

=

The plant specific best estimate/calculation (BE/C) bias factor derived from the surveillance capsule dosimetry data.

(Dc*

= The absolute calculated fast neutron exposure at the location of interest.

For Sequoyah Unit 1, the derived plant specific bias factors were 1.14, 1.14, 1.14 for Cl(E > 1.0 MeV),

D(E > 0.1 MeV), and dpa, respectively. Bias factors of this magnitude developed with BUGLE-96 are within expected tolerances for fluence calculated using the ENDF/B-VI based cross-section library.

Table C-I presents the reactor vessel fast neutron (E > 1.0 MeV) exposure projections using the absolute plant specific calculations. Table C-2 presents the calculated and measured fluences at the capsules.

WCAP-15293

C-2 N

Table C-I Azimuthal Variations Of The Neutron Exposure Projections On The Reactor Vessel Clad/Base Metal Interface At Core Midplane Calculated 00 150 300 4501o1 10.03 EFPY EI>1.0 MeV 2-05E+18 3.21E+18 4.09E+18 6.37E+18 E>0.1 MeV" 4.07E+18 6.41E+18 8.41E+18 1.34E+19 dpa 2.60E-03 4.06E-03 5.22E-03 8.08E-03 20 EFPY E>1.0 MeV 3.97E+18 5.96E+18 7.61E+18 1.18E+19 E>0.I MeV 8.68E+18 1.34E+19 1.76E+19 2.80E+19 dpa 5.55E-03 8.48E-03 1.09E-02 1.69E-02 32 EFPY E>1.0 MeV 6.05E+ 18 9.28E+ 18 1.19E+19 1.84E+19 E>0.1 MeV 1.42E+19 2.18E+19 2.86E+19 4.56E+19 dpa 9.09E-03 1.38E-02 1.78E-02 2.76E-02 48 EFPY E>1.0 MeV 8.96E+18 1.37E+19 1.75E+19 2.72E+19 E>0.1 MeV 2.16E+19 3.30E+19 4.33E+19 6.91E+19 dpa 1.38E-02 2.09E-02 2.69E-02 4.18E-02 Note:

a) Maximum neutron exposure projection WCAP-15293

C-3 N

Table C-2 Comparison Of Calculated And Best Estimate Integrated Neutron Exposure Of Sequoyah Unit I Surveillance Capsules T, U, X, and Y CAPSULE T Calculated Best Estimate RFIC (1(E > 1.0 MeV) [n/cm2]

2.61E+18 2.89E+18 1.10

((E > 0.1 MeV) [n/cmr]

8.74E+I18 9.62E+18 1.10 dpa 4.34E-03 4.80E-03 1.11 CAPSULE U Calculated Best Estimate BE/C D(E > 1.0 MeV) [n/cm 2]

7.96E+18 9.69E+18 1.22 (D(E > 0.1 MeV) [n/cm 2]

2.66E+19 3.16E+19 1.19 dpa 1.32E-02 1.59E-02 1.21 CAPSULE X Calculated Best Estimate BE/C (D(E > I.0 MeV) In/cm2]

1.32E+19 1.50E+19 1.14 (D(E > 0. 1 MeV) [.n/cm2]

4.42E+ 19 5.09E+19

]. 15 dpa 2.20E-02 2.5 1IE-02 1.15 CAPSUtLE Y Calculated Best Estimate BE/C

( (E > 1. 0 MeV) [n/era2 2.19E+19 2.43E+ 19 1.11

[ D(E > 0. 1 MeV) [n/cm2]

7.31E+19 S. 15E+ 19 1.12

[dpa.

3.63E-02 4.0513-02 1.12 AVERAGE BE/C RATIOS SBEIC 7

(D(E > 1.0 MeV) [n/cm 2]

'2(E > 0.1 MeV) [ncmo2]

1.14 1.14 dpa 1.14 WCAP-1 5293 C-3

D-O APPENDIX D UPDATED SURVEILLANCE MATERIAL 30 FT-LB TRANSITION TEMPERATURE SHIFTS AND UPPER SHELF ENERGY DECREASES WCAP-15293

D-1 TABLE D-1 Measured 30 ft-lb Transition Temperature Shifts of all Available Surveillance Data 30 ft-lb Transition Upper Shelf Energy Temperature Shift Decrease Material Capsule Fluence Predicted Measured Predicted Measured (I 1019 n/cm2)

(OF) (*)

(OF) (1()

Lower Shell T

0.261 59.85 67.52 16 16 Forging 04 U

0.796 89.3 109.7 20.5 21 (Tangential)

X 1.32 102.6 145.12 23 8

Y 2.19 114.95 129.87 26.5 23 Lower Shell T

0.261 59.85 50.59 16 0

Forging 04 U

0.796 89.3 67.59 20.5 19 (Axial)

X 1.32 102.6 103.34 23 22 Y

2.19 114.95 133.35 26.5 19 Weld Metal T

0.261 111.13 127.79 35 30 U

0.796 165.82 144.92 42 26 X

1.32 190.51 159.02 45 21 Y

2.19 213.44 163.8 48 28 HAZ Metal T

0.261 45.48 20 U

0.796 78.94 26 X

1.32 95.89 3

Y 2.19 73.3 10 Notes:

(a)

(b)

(C)

Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values of copper and nickel of the surveillance material.

Calculated using measured Charpy data plotted using CVGRAPH, Version 4.1 (Reference 9)

Values are based on the definition of upper shelf energy given in ASTM E 185-82.

WCAP-15293 S...N

E-0 APPENDIX E REACTOR VESSEL BELTLINE MATERIAL PROJECTED END OF LICENSE UPPER SHELF ENERGY VALUES WCAP-1 5293

TABLE E-1 Predicted End-of-License (32 EFPY) USE Calculations for all the Beltline Region Materials Material Weight %

1/4T EOL Unirradiated Projected USE Projected of Cu Fluence USE(")

Decrease EOL USE

_(10.n/rcm2)

(ft-lb)

(%)

(ft-lb)

Intermediate Shell Forging 05 0.15 1.1I 79 24 60 Lower Shell Forging 04 0.13 1.11 72 22 56 Using S/C Data Intermediate to Lower Shell 0.35 1.11 113 42 66 Circumferential Weld Seam WO5 Using S/C Data Notes:

(a)

These values were obtained from Reference 12.

N WCAP-15293

F-0 APPENDIX F UPDATED SURVEILLANCE CAPSULE REMOVAL SCHEDULE WCAP-15293

F-I The following surveillance capsule removal schedule meets the requirements ofASTM El 85-82 and is recommended for future capsules to be removed from the Sequoyah Unit I reactor -vessel. This recommended removal schedule is applicable to 32 EFPY of operation.

Table 7-1 Sequoyah Unit I Reactor Vessel Surveillance Capsule Withdrawal Schedule Removal Time Fluencc Capsule Location Lead Factor()

(EFPY)b)

(nIcm1,E>l.O MeV)(a)

T 400 3.39 1.03 2.61 x 10'3 (c)

U 1400 3.47 3.00 7.96 x 10'8 (c)

X 2200 3.47 5.27 1.32 x 10'9 (c)

Y 3200 3.43 10.03 2.19 x 1019 (cd)

S 40 LIOS Standby (d,e)

V 1760 1.08 Standby (de)

W 1840 1.08 Standby (de)

Z 3560 1.08 Standby (d,e)

Notes:

(a)

Updated in Capsule Y dosimetry analysis (Reference 7).

(b)

Effective Full Power Years (EFPY) from plant startup.

(c)

Plant specific evaluation.

(d)

This fluence is not less than once or greater than twice the peak end of license (32 EFPY) fluence (e)

Capsules S, V, W and Z will reach a fluence of 2.74 x 10'9 (E > 1.0 MeV), the 48 EFPY peak vessel fluence at approximately 44 EFPY, respectively.

WCAP-15293 N

G-0 APPENDIX G ENABLE TEMPERATURE CALCULATIONS AND RESULTS WCAP-15293

G-1 Enable Temperature Calculation:

ASME Code Case N-514 requires the low temperature overpressure (LTOP or COMS) system to be in operation at coolant temperatures less than 200'F or at coolant temperatures less than a temperature corresponding to a reactor vessel metal temperature less than RTmT + 50'F, whichever is greater. RT.MT is the highest adjusted reference temperature (ART) for the limiting beltline material at a distance one fourth of the vessel section thickness from the vessel inside surface (io. clad/base metal interface), as determined by Regulatory Guide 1.99, Revision 2.

32 EFPY The highest calculated 1/4T ART for the Sequoyah Unit 1 reactor vessel beitline region at 32 EFPY is 2160F.

From the OPERLIM computer code output for the Sequoyah Unit 1 32 EFPY P-T limit curves without margins (Configuration # 1389796830, operim film File) the maximum AT,,,a is:

Cooldown Rate (Steady-State Cooldown):

max (AT.)

at 4T-= 07F Heatup Rate of 100°F/Hr:

max (ATm,)

at l/4T = 28.9247F Enable Temperature (ENBT) =

RTN*T + 50 + max (ATmeit.), 'F

= (216 + 50 + 28.924) *F

= 294.924oF The minimum required enable temperature for the Sequoyah Unit 1 Reactor Vessel is 295*F at 32 EFPY of operation.

WCAP-15293

i i

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15321, Revision 1 Sequoyah Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation J. H. Ledger April 2001 Prepared by the Westinghouse Electric Company LLC for the Tennessee Valley Authority Approved:

'ft_.

4 C. H. Boyd, Manager T

Engineering and Materials Technology Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355

Westinghouse Non-Proprietary Class 3 Sequoyah Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation Westinghouse Electric Company LLC A

ii PREFACE This report has been technically reviewed and verified by:

T. J. Laubhamn Revision 1:

An error was detected in the "OPERLIM" Computer Program that Westinghouse uses to generate pressure temperature (PT) limit curves. This error potentially effects the heatup curves when the 1996 Appendix G Methodology is used in generating the PT curves. It has been determined that WCAP-15321 Rev. 0 was impacted by this error. Thus, this revision provides corrected curves from WCAP-15321 Rev. 0.

Note that only the 600F/hr heatup curves were affected by this error. The 100'F/hr heatup and all cooldown curves were not affected by the computer error and thus remain valid.

TABLE OF CONTENTS LIST O F TA B L E S..................................................................................................................................

iv LIST OF FIGURES............

EXECUTIVE SUM M ARY.............

vi I

INTRODUCTION 2

FRACTURE TOUGHNESS PROPERTIES 2

3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS............. 6 4

CALCULATION OF ADJUSTED REFERENCE TEMPERATURE...................................

10 5

HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES................ 15 6

R EFEREN CES.......................................

.................................................... 28 APPENDIX A: LTOPS SETPOINTS.........................

................ A-0 APPENDIX B: PRESSURIZED THERMAL SHOCK (PTS) RESULTS.............................. B-0 A PPEN D IX C-0 APPENDIX D: UPDATED SURVEILLANCE MATERIAL 30 FT-LB TRANSITION TEMPERATURE SHIFTS AND UPPER SHELF ENERGY DECREASES................ D-0 APPENDIX E: REACTOR VESSEL BELTLINE MATERIAL PROJECTED END OF LICENSE UPPER SHELF ENERGY VALUES...........................................................................

E-0 APPENDIX F: UPDATED SURVEILLANCE CAPSULE REMOVAL SCHEDULE..........................

F-0 APPENDIX G: ENABLE TEMPERATURE CALCULATIONS AND RESULTS.........................

G-0

iv LIST OF TABLES Table I Reactor Vessel Beitline Material Unirradiated Toughness Properties...........................

3 Table 2 Calculation of Chemistry Factors using Sequoyab Unit 2 Surveillance Capsule Data........ 4 Table 3 Summary of the Sequoyah Unit 2 Reactor Vessel Beltline Material Chemistry Factors..... 5 Table 4 Neutron Fluence Projections at Key Locations on the Reactor Vessel Clad/Base Metal Interface (10' n/cm2, E > 1.0 MeV).............................

1......

1 Table 5 Summary of the Vessel Surface, I/4T and 3/4T Fluence Values used for the Generation of the 32 EFPY Heatup/Cooldown Curves...................................................

1 Table 6 Summary of the Calculated Fluence Factors used for the Generation of the 32 EFPY Heatup and Cooldown Curves..................................................................................

11 Table 7 Integrated Neutron Exposure of the Sequoyah Unit 2 Surveillance Capsules Tested To D ate.........................................................................................................

12 Table 8 Calculation of the ART Values for the l/4T Location @ 32 EFPY.............................

13 Table 9 Calculation of the ART Values for the 3/4T Location @ 32 EFPY.............................

13 Table 10 Summary of the Limiting ART Values Used in the Generation of the Sequoyah Unit 2 Heatup/Cooldown Curves.........................................................................................

14 Table 11 32 EFPY Heatup Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors)......................................................

22 Table 12 32 EFPY Cooldown Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors).........................................................

24 Table 13 32 EFPY Heatup Curve Data Points Using 1996 App. G (with Uncertainties for Instrumentation Errors of 109F and 60 psig)................................

25 Table 14 32 EFPY Cooldown Curve Data Points Using 1996 App-G (with Uncertainties for Instrumentation Errors of 1O0F and 60 psig)................................

27

v LIST OF FIGURES Figure 1 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 60*F/hr) Applicable for the First 32 EFPY (Without Margins for Instrumentation Errors)..........................................................

16 Figure 2 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100°F/hr) Applicable for the First 32 EFPY (Without Margins for Instrumentation Errors)..........................................................

17 Figure 3 Sequoyah Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rites up to 100°F/hr) Applicable for the First 32 EFPY (Without Margins for Instrumentation Errors)..........................................................

18 Figure 4 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 60*F/hr) Applicable for the First 32 EFPY (With Margins for Instrumentation Errors of 10°F and 60 psig)...............................

19 Figure 5 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100*F/hr) Applicable for the First 32 EFPY (With Margins for Instrumentation Errors of I 0°F and 60 psig)................................

20 Figure 6 Sequoyah Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 32 EFPY (With Margins for Instrumentation Errors of 10°F and 60 psig)................................

21

vi EXECUTIVE

SUMMARY

This report provides the methodology and results of the generation of heatup and cooldown pressure temperature limit curves for normal operation of the Sequoyah Unit 2 reactor vessel. In addition, Pressure Temperature Limits Report (PTLR) support information, such as LTOPS Setpoint, PTS, EOL USE and Withdrawal Schedule, is documented herein under the Appendices. The PT curves were generated based on the latest available reactor vessel information (CapsuleY analysis, WCAP-1532 1W and the latest Pressure Temperature (P-T) Limit Curves from WCAP-12971I 31). The Sequoyah Unit 2 heatup and cooldown pressure-temperature limit curves have been updated based on the use of the ASME Code Case N-640131, which allows the use of the Ki. methodology, and the elimination of the reactor vessel flange temperature requirement (Ref, WCAP-153 15 '6').

1 INTRODUCTION Heatup and cooldown limit curves are calculated using the adjusted RTNDT (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RTmr of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ART Tr, and adding a margin. The unirradiated RTNDT is designated as the higher of either the drop weight nil ductility transition temperature (NDTI) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60'F.

RTmTr increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTDTr at any time period in the reactor's life, ARTm-due to the radiation exposure associated with that time period must be added to the unirradiated RTNr (IRTNDT). The extent of the shift in RTr4DT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials."I'l Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTNT + tRTNDT + margins for uncertainties) at the 1/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region measured from the clad/base metal interface.

The heatup and cooldown curves documented in this report were generated using the most limiting ART values and the NRC approved methodology documented in WCAP-14040-NP-A, Revision 212],

"Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" with exception of the following: 1) The fluence values used in this report are calculated fluence values, not the best estimate fluence values (See Appendix B). 2) The Ki, critical stress intensities are used in place of the K1. critical stress intensities. This methodology is taken from approved ASME Code Case N-640133. 3) The reactor vessel flange temperature requirement has been eliminated.

Justification has been provided in WCAP-15315161. 4) The 1996 Version of Appendix G to Section X 141 will be used rather than the 1989 version.

WCAP-15321

2 2

FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic materials in the reactor coolant pressure boundary are determined in accordance with the NRC Standard Review PlanE51. The beltline material properties of the Sequoyah Unit 2 reactor vessel is presented in Table 1.

Best estimate copper (Cu) and nickel (Ni) weight percent values used to calculate chemistry factors (CF) in accordance with Regulatory Guide 1.99, Revision 2, are provided in Table 1. Additionally, surveillance capsule data is available for four capsules (Capsules T, U, X and Y) already removed from the Sequoyah Unit 2 reactor vessel. This surveillance capsule data was also used to calculate CF values per Position 2.1 of Regulatory Guide 1.99, Revision 2 in Table 2 These CF values are summarized in Table 3.

The Regulatory Guide 1.99, Revision 2 methodology used to develop the heatup and cooldown curves documented in this report is the same as that documented in WCAP-14040, Revision 2.

WCAP-15321

3 TABLE I Reactor Vessel Beitline Material Unirradiated Toughness Properties Material Description Cu (%)

Ni(%)

Initial RT.,rya(r Intermediate Shell Forging 05 0.13 0.76 10OF (Heat #288757 / 981057)

Lower Shell Forging 04 0.14 0.76

-220F (Heat # 990469 /293323)

Intermediate to Lower SheUl Forging 0.12 0.11

.40F Circumferential Weld Seam(b Surveillance Weldt&i 0.13 0.11 (a) The Initial RTNDT values are measured vwdues (b) Circumferential Weld Seam was fabricated with weld wire type SMIT 89, Heal # 4278, Flux type SMIT 89. lot # 1211 and is representative of the intermediate to lower shell circumferential weld.

The chemistry factors were calculated using Regulatory Guide 1.99 Revision 2, Positions 1.1 and 2.1.

Position 1.1 uses the Tables from the Reg. Guide along with the best estimate copper and nickel weight percents. Position 2.1 uses the surveillance capsule data from all capsules withdrawn to date. The fluence values used to determine the CFs in Table 2 are the calculated fluence values at the surveillance capsule locations. Hence, the calculated fluence values were used for all cases.

The measured ARTNDt.

values for the weld data were adjusted using the ratio procedure given in Position 2.1 of Regulatory Guide 1.99, Revision 2. All fluence values were obtained from the recent Sequoyah Unit 2 capsule analysist~l which calculated the fluences using the ENDF/B-VI scattering cross-section data set. The fluence values used are also documented in Appendix C of this report.

WCAP-15321

TABLE 2 Calculation of Chemistry Factors using Sequoyah Unit 2 Surveillance Capsule Data Material Capsule Capsule Ps)

F-oND Tr(C)

FF*ARTNDT FF2 Intermediate Shell T

2.61E+18 0.635 63.7 40.45 0.403 Forging 05 U

6.92E+18 0.897 79.3 71.13 0.805 (Tangential)

X 1.22E+19 1.055 85.7 90.41 1.113 F

Y 2.14E+19 1.207 134.1 161.86 1.457 Intermediate Shell T

2.61E+18 0.635 48.7 30.92 0.403 Forging 05 U

6.92E+18 0.897 66.1 59.29 0.805 (Axial)

X 1.22E+19 1.055 110.0 116.05 1.113 Y

2.14E+19 1.207 89.2 107.66 1.457 SUM:

677.770F 7.556 CFos Z I(FF

RTmn,.) + Y2( FF2) = (677.77) + (7.556) = 89.70F I

II Surveillance Weld T

2.61E+18 0.635 69.4 (74.6) 44.07 0.403 Material(d)

U 6.92E+18 0.897 121.3 (130.4) 108.81 0.805 X

1.22E+19 1.055 41.1 (44.2) 43.36 1.113 Y

2.14E+19 1.207 80.8 (86.9) 97.53 1.457 SUM:

293.770F 3.778 CF s*,. wla = Z(FF

RTNryr) + X( FF2) = (293.77°F) + (3.778) = 77.89F Notes:

(a) f= Calculated fluence from capsule Y dosimetry analysis results"'>, (x 10'9 n/cm 2 a, E > 1.0 MeV).

(b)

FF = fluence factor = e28 -0..0.,,og f)

(c)

ARTNt~

values are the measured 30 ft-lb shift values taken from App. B of Ref. 7, rounded to one decimal point.

(d)

The surveillance weld metal ARTNDT values have been adjusted by a ratio factor of 0.93.

W LAt-I3j21 4

5 TABLE 3 Summary of the Sequoyah Unit 2 Reactor Vessel Beitline Material Chemistry Factors Material Reg. Guide 1.99, Rev. 2 Reg. Guide 1.99, Rev. 2 Position 1.1 CF's Position 2.1 CF's Intermediate Shell Forging 05 95OF 89.70F Lower Shell Forging 04 1040F Circumferential Weld W05 630F 77.80F (Heat # 4278)

Surveillance Weld Metal (Heat # 4278) 67.90F WCAP-15321

6 3

CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 3.1 Overall Approach The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K1, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K10, for the metal temperature at that time. K1, is obtained from the reference fracture toughness curve, defined in Code Case N-640, "Alternative Reference Fracture Toughness for Development of PT Limit Curves for Section x)'&" k 4] of the ASMEAppendix G to Section X[. The Kk curve is given by the following equation:

K10= 33.2 + 20.734

e1002(T-RT~r)]

(1)

where, Ki,

=

reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This K1. curve is based on the lower bound of static critical KI values measured as a function of temperature on specimens of SA-533 Grade B Class 1, SA-508-1, SA-508-2, SA-508-3 steel.

3.2 Methodology for Pressure-Temperature Limit Curve Development The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

C* K. + K1, < K1o (2)

where, Kim

=

stress intensity factor caused by membrane (pressure) stress K1 t

=

stress intensity factor caused by the thermal gradients K1.

=

function of temperature relative to the RTmr of the material C

=

2.0 for Level A and Level B service limits C

1.5 for hydrostatic and leak test conditions during which the reactor core is not critical WCAP-15321

7 For membrane tension, the corresponding K1 for the postulated defect is:

Ki= M. x (pR /t)

(3) where, Mm for an inside surface flaw is given by:

M.

=

1.85 for F1 < 2, Mm

=

0.926f/t" for 2<-,5ft-3.464, Mm 3.21 for l > 3.464 Similarly, Mm for an outside surface flaw is given by:

Mm

=

1.77for f <2, M,

=

0.893-N" for 2<Virt-_<3.464, Mm

=

3.09 for it- > 3.464 and p = internal pressure, Ri = vessel inner radius, and t vessel wall thickness.

"For bending stress, the corresponding K, for the postulated defect is:

KIb = Mb

Maximum Stress, where Mb is two-thirds of Mm The maximum K, produced by radial thermal gradient for the postulated inside surface defect of G-2120 is K1t= 0.953x10 3 x CR x te5, where CR is the cooldown rate in OF/hr., or for a postulated outside surface defect, KY, 0.753x10"3 x HU x t2 s, where HU is the heatup rate in IF/hr.

The through-wall temperature difference associated with the maximum thermal K, can be determined from Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from Fig.

G-2214-2 for the maximum thermal K1.

(a)

The maximum thermal K, relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(l) and (2).

(b)

Alternatively, the K, for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a YA-thickness inside surface defect using the relationship:

Kit = (1.0359C 0+.6322C, + 0.4753C2 + 0.3855C3) * *

(4)

WCAP-15321

8 or similarly, Krr during heatup for a 'A-thickness outside surface defect using the relationship:

Kit = (1.043Co + 0.630Ci + 0A8 1C2 + 0.401C3)

N (5) where the coefficients C0, C), C2 and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:

cr(x) = Co + Ci(x / a) + C2(x / a)2 + C3(x / a)3 (6) and x is a variable that. represents the radial distance from the appropriate (i.e., inside or outside) surface to any point on the crack front and a is the maximum crack depth.

Note, that equations 3, 4 and 5 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. No other changes were made to the OPERLIM computer code with regard to P-T calculation methodology. Therefore, the P-T curve methodology is unchanged from that described in WCAP-14040, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldwon Limit Curves"[2 ] Section 2.6 (equations 2.6.2-4 and 2.6.3-1) with the exceptions just described above.

At any time during the heatup or cooldown transient, K1, is determined by the metal temperature at the tip of a postulated flaw at the l/4T and 3/4T location, the appropriate value for RTmT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, K%

1, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT (temperature) developed during cooldown results in a higher value of Krc at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in Kk. exceeds K1,, the calculated allowable pressure during cooldown will be greater than the steady-state value.

WCAP-15321

9 The above procedures are needed because there is no direct control on temperature at the 1/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.

Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the KI, for the 1/4T crack during heatup is lower than the K,. for the I/4T crack during steady state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower K1, values do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the I4T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a I/4T flaw located at the 1/4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.

3.3 Closure Head/Vessel Flange Requirements 10 CFR Part 50, Appendix G addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RTNDT by at least 120°F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3107 psi), which is 621 psig for Sequoyah Unit 2. However, per WCAP-1 5315, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation For Operating PWR and BWR Plants"' 61, this requirement is no longer necessary when using the methodology of Code Case N-640)81. Hence, Sequoyah Unit 2 heatup and cooldown limit curves will be generated without flange requirements included.

WCAP-15321

10 4

CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:

ART = Initial RTm, + ARTNr + Margin (7)

Initial RTrr is the reference temperature for the unirradiated material as defined in paragraph NB-233 1 of Section III of the ASME Boiler and Pressure Vessel CodetS1. If measured values of initial RTmnrr for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.

ARTDr is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:

ARTND = CF

f(o.28-0.1o0%)

(8)

To calculate ARTND at any depth (e.g., at 1/4T or 3/47), the following formula must first be used to attenuate the fluence at the specific depth.

t*.p

) = f,*f.*

e (-0.249)

(9) where x inches (vessel beltline thickness is 8.45 inches) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then placed in Equation 8 to calculate the ARTNT at the specific depth.

The Westinghouse Radiation Engineering and Analysis Group evaluated the vessel fluence projections as a part ofWCAP-1532Orl and are also presented in a condensed version in Table 4 of this report. The evaluation used the ENDF/B-VI scattering cross-section data set. This is consistent with methods presented in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves'421. Table 4 contains the calculated vessel surface fluences values at various azimuthal locations and Tables 5 and 6 contains the 114T and 3/4T calculated fluences and fluence factors, per the Regulatory Guide 1.99, Revision 2, used to calculate the ART values for all beitline materials in the Sequoyah Unit 2 reactor vessel. Additionally, the surveillance capsule fluence values are presented in Table 7.

WCAP-15321

11 TABLE 4 Neutron Fluence Projections at Key Locations on the Reactor Vessel Clad/Base Metal Interface (10'9 nIcm2, E > 1.0 MeV)

Azimuthal Location 150 300 450 0.336 0.426 0.637 0.60 0.773 1.16 0.934 1.21 1.82 1.38 1.80 2.71 TABLE 5 Summary of the Vessel Surface, 1/4T and 3/4T Fluence Values used for the Generation of the 32 EFPY Heatup/Cooldown Curves Material Surface

/, TN'

% If,,)

Intermediate Shell Forging 05 1.82 x 10'9 1.1ox 1019 3.98 x 10" Lower Shell Forging 04 1.82 x 101' 1.10 x 1019 3.98 x l0'8 Circumferential Weld Seam 1.82 x 10"'

1. lox 1019 3.98 x 10"M (Heat 42-78)

Note:

(a) 1/4T and 3/4T = F(s~f..) *e(-24"), where x is the depth into the vessel wall (i.e. 8.45*0.25 or 0.75)

TABLE 6 Summary of the Calculated Fluence Factors used for the Generation of the 32 EFPY Heatup and Cooldown Curves EFPY I/4T FF 3/4T FF 32 1.027 0,745 WCAP-15321 11

12 TABLE 7 Integrated Neutron Exposure of the Sequoyah Unit 2 Surveillance Capsules Tested To Date Capsule Fluence T

2.61 x lO1 n/cm2, (E > 1.0 MeV)

U 6.92 x 1018 n/cm2, (E > 1.0 MeV)

X 1.22 x 1019 n/cm 2, (E > 1.0 MeV)

Y 2.14 x 10'n/cm 2, (E> 1.0 MeV)

Margin is calculated as, M =2 UA

+ u*. The standard deviation for the initial RT-rn margin term, is ai 00F when the initial RTNDT is a measured value, and 17°F when a generic value is available. The standard deviation for the ARTNDT margin term, c;, is 17'F for plates or forgings, and 8.5°F for plates or forgings when surveillance data is used. For welds, a'r is equal to 28"F when surveillance capsule data is not used, and is 14*F (half the value) when credible surveillance capsule data is used. ac need not exceed 0.5 times the mean value of ARTmT.

Based on the surveillance program credibility evaluation presented in Appendix D to WCAP-15320, the Sequoyah Unit 2 surveillance program data is non-credible. In addition, following the guidance provided by the NRC in recent industry meeting, Table Chemistry Factor for the intermediate shell forging 05 was determined to be conservative. Hence, the adjusted reference temperature (ART) must be calculated using Position 1.1 along with the full margin term. Both Regulatory Guide 1.99, Revision 2, Position 1. 1 and 2.1 have been shown herein for completeness. Contained in Tables 8 and 9 are the calculations of the 32 EFPY ART values used for generation of the heatup and cooldown curves.

WCAP-15321

13 TABLE 8 Calculation of the ART Values for the 1/4T Location ( 32 EFPY Material RG 1.99 CF FF IRTý=')

ARTINDT0()

Margin(4)

ART<2)

R2 Method (OF)

(OF)

(-F)

Intermediate Shell Forging 05 Position 1.1 95 1.027 10 97.6 34 142 Position 2.1 89.7 1.027 10 92.1 34 136 Lower Shell Forging 04 Position 1.1 104 1.027

-22 106.8 34 119 Intermediate to Lower Shell Position 1.1 63 1.027

-4 64.7 56 117 Circumferential Weld Seam Position 2.1 77.8 1.027

-4 79.9 56 132 Notes:

(1)

Initial RT=r values measured values.

(2)

ART = Initial RTmnT + ARTmyr + Margin (fF)

(3)

ARTNDT = CF

FF (4)

M = 2 *(a.2 +

cFA2)112 TABLE 9 Calculation of the ART Values for the 3/4T 32 EFPY Material RG 1.99 CF FF IRTN0) ART,,T 31 Margin(4)

ART'7)

R2 Method (OF)

(OF)

Intermediate Shell Forging 05 Position 1.1 95 0.745 10 70.8 34 115 Position 2.1 89.7 0.745 10 66.8 34 111 Lower Shell Forging 04 Position 1. 1 104 0.745

-22 77.5 34 90 Intermediate to Lower Shell Position 1.1 63 0.745

-4 46.9 56 99 Circumferential Weld Seam Position 2.1 77.8 0.745

-4 58.0 56 110 Notes:

(1)

Initial RTNDT values measured values.

(2)

ART = Initial RTmNT + ARTN= + Margin (0F)

(3)

ARTNDT = CF

FF (4)

M=2 *(CsZ2 + a,2)112 WCAP-15321

14 The intermediate shell forging 05 is the limiting betline material for the 1/4T and 3/4T case (See Tables 8 and 9). Contained in Table 10 is a summary of the limiting ARTs to be used in the generation of the Sequoyah Unit 2 reactor vessel heatup and cooldown curves.

TABLE 10 Summary of the Limiting ART Values Used in the Generation of the Sequoyah Unit 2 Heatup/Cooldown Curves WCAP-15321 EFPY 1/4T Limiting ART 3/4T Limiting ART 32 1420F 1150F

15 5

HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel beltline region using the methods discussed in Sections 3.0 and 4.0 of this report. This approved methodology is also presented in WCAP-14040-NP-A, Revision 2 with exception to those items discussed in Section I of this report.

Figures 1, 2, 4 and 5 present the heatup curves with (I0 F and 60 psig) and without margins for possible instrumentation errors using heatup rates of 60 and 100*F/hr applicable for the first 32 EFPY. Figures 3and 6 presents the cooldown curves with (10*F and 60 psig) and without margins for possible instrumentation errors using cooldown rates of 0, 20, 40, 60 and 100°F/hr applicable for 32 EFPY.

Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 1 through 6. This is in addition to other criteria which must be met before the reactor is made critical, as discussed below in the following paragraphs.

The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figures 1,2, 4 and 5. The straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic test is defined in Code Case N-640131 (approved in February 1999) as follows:

1.5 K. < Ki,

where, Km is the stress intensity factor covered by membrane (pressure) stress, K,, = 33.2 + 20.734 e0.02r'RTNwr>*

T is the minimum permissible metal temperature, and RTNTDT is the metal reference nil-ductility temperature.

The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Reference 10. The pressure-temperature limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40°F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section 3.0 of this report. For the heatup and cooldown curves without margins for instrumentation errors, the minimum temperature for the in service hydrostatic leak tests for the Sequoyah Unit 2 reactor vessel at 32 EFPY is 214*F. The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40*F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.

Figures I through 6 define all of the above limits for ensuring prevention of nonductile failure for the Sequoyah Unit 2 reactor vessel. The data points used for the heatup and cooldown pressure-temperature limit curves shown in Figures 1 through 6 are presented in Tables 11 through 14.

WCAP-15321

16 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING 05 LIMITING ART VALUES AT 32 EFPY:

2500 2250 2000 1750 S1500 S

'A L 1250 IL 3 1000 U

750 500 250 0

'1/T, 1420F 3T, 115*F Leak Test LI II Heatup Rate Criti Limit 60 Dog. F/Hr 600eg. F/Hr Criticality Limit based on

__________Inservice hydrostatic test temperature (203 F) for the Bolt-service period up to 32 EPPY Bol*up Tern p 0

50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

FIGURE 1 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 60°F/hr)

Applicable for the First 32 EPPY (Without Margins for Instrumentation Errors)

WCAP-15321 Unacceptable Operation I /

Acceptable Opermtion FI S:

I WCAP-15321

MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING 05 LIMITING ART VALUES AT 32 EFPY:

1AT, 142°F 3/44T, 115°F 2500 2250 2000 1750 0

S 1500 0 1250 S1000 U

750 5oo 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

FIGURE 2 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100*F/hr)

Applicable for the First 32 EFPY (Without Margins of for Instrumentation Errors)

WCAP-1 5321 K ->

17 N

18 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHBELL FORGING 05 LIMITING ART VALUES AT 32 EFPY:

1/4T, 142-F 3/4T, 115°F 2500 o 2250 S2000 1750 Cl)

CA.)

1500 1250 1000 750 500 250 0

illr

[ i Ii I

I i 1

1 1 Ii:

il II FT Ii i

UNACCEPTABLE OPERATION OPE___

I,;

COOLON CýLDOTN RATES F/Hr.

0 40 so 1o.1 Io Blup Tern.

J

!T CC 0

50 100

!II; II

] AI Il l lI i'1-1 I

I I

I II II I:

III' II iIt

.1 i

i Ill ACCEPTABLE LATION I ii Li_

!i i

i :

I,

11:1.1 777+-4'+:;K

-I I '

.1'

.1 III I.

Th+iIH4+L'

200 250 300 I--

Ii.

I 3.

350 Moderator Temperature

'iI I.

400 450 (Deg.F)

Figure 3 Sequoyah Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100'F/hr) Applicable for the First 32 EFPY (Without Margins for Instrumentation Errors)

WCAP-15321 500 r

b I

L I

i 1

Iltl!'

=li[

I J I i ii ii

-T-7

'I' j,

i I I I

J///l*.,

I ]

i I i J i

I I

I 50 1

19 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING 05 LIMITING ART VALUES AT 32 EFPY:

1AT, 142°F 3T, 115 0F 2500 2250 2000 Operation n.* ~6 150

-~ Deg.'F/Hr 60 Deg.FI.

IL 1500 6-g./~

. 1250 I.

S1000 750 j

Criticality Limit bas Inservlce hydrosate 500 temperature (214 F service period up tc 250 T+/-.

0 7..

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

FIGURE 4 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 60°F/hr)

Applicable for the First 32 EFPY (With Margins for Instrumentation Errors of I 0F and 60 psig)

WCAP-15321 7>

20 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING 05 LIMITING ART VALUES AT 32 EFPY:

1/4T, 142-F 3/4T, 115-F 2500 2250 2000.

1750

a. 1500 C,.

up 1 1250

0.

S1000 750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

FIGURE 5 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100 0F/hr)

Applicable for the First 32 EFPY (With Margins of for Instrumentation Errors of 10°F and 60 psig)

WCAP-15321

21

,'

N MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING 05 LIMITING ART VALUES AT 32 EFPY:

1/4T, 142-F 3/4T, 115-F 5

,7 J4-1

!II1

-I-I I I-I I--

_LL WIJ I I I

-I 750 500 250-

' l l

I lI II I

UNACCEPTABLE I

I OPERATION tI ! I I Il I

I I [

I I I i i l i I i

T 1~9I I

LLL I

I li

,i Iii iii

-T-r I I I I

7Th I I JfI!LLI II

-COOLDOWN R RAT ESF F/Br.

0 20 S40

, 0 100 7-Ii I I I 0

-f I

I I

I I t liii Ij-

r r I-L1 I I Boltup I

Imp 1-50 III l

l II I

ACCEPTABLE I OPERATION i-H--F III I I I I

Ii L

lid III III I

! I [

I "

100 150 l-- -L_

200 II 1 --I I

Moderator Temperature (Deg.F)

Figure 6 Sequoyah Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100*F/hr) Applicable for the First 32 EFPY (Without Margins for Instrumentation Errors of 10*F and 60 psig)

WCAP-15321 I

I ii Li III

_I I_

i-i-2500 tw 2250 2000

, f f-T-F

-i --'

i F -* i 450 500

- I.....i..

- -I -

j 1-I

_ _.L.:

Li Ii Il-1

-F--II!

I,

-H-I JELLr I

I i~I i

1750 1500 1250 1000 w) cr2 C->

0 F-i-F-i--,/

I

%:*:llt!'

lli*

I

-H

-I I

i Ii11':

I I

I

! I JI I,

L,JT//ii Ii Ii

] I Iill 3600

,350 400

22 TABLE 11 32 EFPY Heatup Curve Data Points Using 1996 App.G (without Uncertainties for Instrumentation Errors)

Heatup Curves 60 Heamup 60 Limit Critical 100 Heatup 100 Limit Critical Leak Test Limit T

P T

P 50 0

203 50 203 0

18 200 50 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 0

661 667 674 682 690 700 710 721 734 735 738 745 755 768 785 805 828 855 885 920 958 1002 1050 1103 1163 1228 1301 1382 1471 1570 1662 1757 200 1662 200 205 1757 205 203 203 203 203 203 203 203 203 203 203 203 203 203 203 203 205 210 215 220 225 230 235 240 245 250 255 260 265 270 WCAP-15321 0

735 738 745 755 768 785 805 828 855 885 920 958 1002 1050 1103 1163 1228 1301 1382 1471 1570 1662 1757 1861 1976 2103 2243 2398 50 50 55 60 65 70 75 80 85 90 95 too 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 2*

929 971 1019 1071 1129 1194 1266 1346 265 270 275 280 285 1758 1889 2034 2194 2371 0

661 667 674 682 690 700 706 706 706 706 706 706 706 708 713 721 731 745 761 780 803 829 858 892 203 203 203 203 203 203 203 203 203 203 203 203 203 205 210 215 220 225 230 235 240 245 250 255 260 0

706 707 708 713 721 731 745 761 780 803 829 858 892 929 971 1019 1071 1129 1194 1266 1346 1434 1531 1639 186 2000 203 2485

23 TABLE 11 (Continued) 32 EFPY Heatup Curve Data Points Using 1996 App.G (without Uncertainties for Instrumentation Errors)

Heatup Curves 60 Heatup 60 Limit Critical 100 Heatup 100 Limit Critical Leak Test Limit T

P T

P 210 1861 210 1434 215 1976 215 1531 220 2103 220 1639 225 2243 225 1758 230 2398 230 1889 235 2034 240 2194 245 2371 WCAP-15321

24 TABLE 12 32 EFPY Cooldown Curve Data Points Using 1996 App. G (without Uncertainties for Instrunentation Errors)

Cooldown Curves Steady State 20F 40F 60F 1lOF T

P T

P 50 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 0

661 667 674 682 690 700 710 721 734 748 763 780 799 820 843 869 897 928 962 1000 1042 1088 1140 1196 1259 1328 1404 1489 1582 1685 1799 1925 2064 2218 2388 50 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 0

618 624 632 640 649 659 670 683 696 712 728 747 767 790 815 843 874 908 945 987 1033 1084 50 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 0

574 581 589 598 608 619 631 644 659 676 694 714 736 761 789 819 853 891 932 978 1028 1085 50 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 0

530 538 546 556 566 578 591 606 622 640 660 682 707 734 764 798 835 876 922 973 1029 50 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 0

440 449 459 470 483 497 513 530 550 572 596 623 653 686 723 764 809 860 916 978

.1 _______________

WCAP-15321 WCAP-15321 24

TABLE 13 32 EFPY Heatup Curve Data Points Using 1996 App.G (with Uncertainties for Instrumentation Errors of 10F and 60 psig)

Heatun Curves WCAP-15321 25 60 Heatup 60 Limit Critical 100 Heatup 100 Limit Critical Leak Test Limit T

P T

P 50 0

214 0

50 0

214 0

198 2000 50 591 214 607 50 591 214 607 214 2485 55 595 214 614 55 595 214 614 60 601 214 622 60 601 214 622 65 607 214 675 65 607 214 657 70 614 214 678 70 614 214 650 75 622 214 685 75 622 214 647 80 630 214 695 80 630 214 646 85 640 214 708 85 640 214 648 90 650 214 725 90 646 214 653 95 661 214 745 95 646 214 661 100 674 214 768 100 646 214 671 105 675 214 795 105 646 214 685 110 678 214 825 110 646 214 701 115 685 214 860 115 646 214 720 120 695 214 898 120 646 214 743 125 708 214 942 125 648 214 769 130 725 214 990 130 653 214 798 135 745 215 1043 135 661 215 832 140 768 220 1103 140 671 220 869 145 795 225 1168 145 685 225 911 150 825 230 1241 150 701 230 959 155 860 235 1322 155 720 235 1011 160 898 240 1411 160 743 240 1069 165 942 245 1510 165 769 245 1134 170 990 250 1602 170 798 250 1206 175 1043 255 1697 175 832 255 1286 180 1103 260 1801 180 869 260 1374 185 1168 265 1916 185 911 265 1471 190 1241 270 2043 190 959 270 1579 195 1322 275 2183 195 1011

.275 1698 200 1411 280 2338 200 1069 280 1829 205 1510 205 1134 285 1974 210 1602 210 1206 290 2134 215 1697 215 1286 295 2311 220 1801 220 1374

26 TABLE 13 (Continued) 32 EFPY Heatup Curve Data Points Using 1996 App.G (with Uncertainties for Instrumentation Errors of 1 OF and 60 psig)

Heatup Curves 60 Heatup 60 Limit Critical 100 Heatup 100 Limit Critical Leak Test Limit T

P T

P 225 1916 225 1471 230 2043 230 1579 235 2183 235 1698 240 2338 240 1829 245 1974 250 2134 255 2311 WCAP-15321

27 TABLE 14 32 EFPY Cooldown Curve Data Points Using 1996 App. G (with Uncertainties for Instrumentation Errors of 100F and 60 psig)

Cooldown Curves Steady State 20F 40F 60F 1OOF T

P T

P 50 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 50 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 0

552 554 558 564 572 580 589 599 610 623 636 652 668 687 707 730 755 783 814 848 885 927 973 1024 50 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 0

503 508 514 521 529 538 548 559 571 584 599 616 634 654 676 701 729 759 793 831 872 918 968 1025 50 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 0

461 466 470 478 486 496 506 518 531 546 562 580 600 622 647 674 704 738 775 816 862 913 969 50 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 0

366 372 380 389 399 410 423 437 453 470 490 512 536 563 593 626 663 704 749 9o0 856 918 0

591 595 601 607 614 622 630 640 650 661 674 688 703 720 739 760 783 809 837 868 902 940 982 1028 1080 1136 1199 1268 1344 1429 1522 1625 1739 1865 2004 2158 2328 WCAP-1 5321 WCAP-1532l

28 6

REFERENCES

1.

Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S.

Nuclear Regulatory Commission, May 1988.

2.

WCAP-I 4040-NP-A, Revision 2, "Methodology used to Develop Cold Overpressure Mitigating system Setpoints and RCS Heatup and Cooldown Limit Curves", J.D. Andrachek, et. al., January 1996.

3.

ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1'" February 26, 1999.

4.

Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G; "Fracture Toughness Criteria for Protection Against Failure.", Dated 1989 & December 1995.

5.

"Fracture Toughness Requirements", Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981.

6.

WCAP-15315, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation For Operating PWR and BWR Plants", W Bamford, et.al., October 1999.

7.

WCAP-15320, "Analysis of Capsule Y From the Tennessee Valley Authority Sequoyah Unit 2 Reactor Vessel Radiation Surveillance Program", T.J. Laubham, et. al., Dated December 1999.

8.

1989 Section III, Division I of the ASME Boiler and Pressure Vessel Code, Paragraph NB-233 1, "Material for Vessels."

9.

CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATI Consulting, March 1999.

10.

Code of Federal Regulations, 10 CFR Part 50, Appendix GQ "Fracture Toughness Requirements,"

U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.

11.

WCAP-7924-A, "Basis for Heatup and Cooldown Limit Curves," W. S. Hazelton, et al., April 1975.

12.

Calc. No.92-016, "WOG USE Program - Onset of Upper Shelf Energy Calculations", I. M.

Chicots, dated l1/12J92. File# WOG-)08/4-1% (MUHP-5080).

13.

WCAP-12971, "Heatup and Cooldown Limit Curves for Normal Operation Sequoyah Unit 2".

J.M. Chicots, et. al., Dated June, 1991.

14.

Westinghouse Letter TVA-91-242 WCAP-15321

29

15.

Westinghouse Letter TVA-91-243

16.

TVA letter Number N9664, "TASK N99-017 - Reactor Coolant System Pressure and Temperature Limit Report Development - N2N-048," W. M. Justice, August 17, 1999.

17.

TVA letter Number N9667, "TASK N99-017 - Reactor Coolant System Pressure and Temperature Limit Report Development - N2N-048," W. M. Justice, August 20, 1999.

18.

Westinghouse Letter TVA-93-105.

WCAP-15321

A-0 APPENDIX A LTOPS SETPOINTS WCAP-15321

A-1

1.1 INTRODUCTION

Westinghouse has been requested to develop Low Temperature Overpressure Protection System (LTOPS) setpoints for Sequoyah Unit 2, for a vessel exposure of 32 EFPY. The LTOPS setpoints for the Sequoyah, Units I and 2, were last revised by Westinghouse in June of 1991 using pressure-temperature limits supplied by Tennessee Valley Authority for a vessel exposure of 16 EFPY. The results of this analysis were reported via References 14 and 15. The June 1991 analysis was based on the September 1989 analysis which addressed Eagle-21 implementation and Appendix G limits based on Regulatory Guide 1.99. This Section documents the development of new Sequoyah Unit 2 COMS setpoints for 32 EFPY.

The Low Temperature Overpressure Protection System (LTOPS) is designed to provide the capability, during relatively low temperatures Reactor Coolant System (RCS) operation (typically less than 3500F), to protect the reactor vessel from being exposed to conditions of fast propagating brittle fracture. The LTOPS is provided in addition to the administrative controls, to prevent overpressure transients and as a supplement to the RCS overpressure mitigation function of the Residual Heat Removal System (RHRS) relief valves. LTOP consists of pressurizer PORVs and actuation logic from the wide range pressure channels. Once the system is enabled, no operator action is involved for the LTOPS to perform its intended pressure mitigation function.

LTOPS setpoints are conservatively selected to prevent exceeding the pressure/temperature limits established by 10 CFR Part 50 Appendix G requirements.

Two specific transients have been defined as the design basis for LTOPS. Each of these transient scenarios assume that the RCS is in a water-solid condition and that the RHRS is isolated from the RCS. The first transient consists of a heat injection scenario in which a reactor coolant pump in a single loop is started with the RCS temperature as much as 50 0F lower than the steam generator secondary side temperature and the RHRS has been inadvertently isolated. This results in a sudden heat input to the RCS from the steam generators, creating an increasing pressure transient. The second transient has been defined as a mass injection scenario into the RCS caused by the simultaneous isolation of the RHRS, isolation of letdown and failure of the normal charging flow controls to the full flow condition. The resulting mass injection/letdown mismatch causes an increasing pressure transient.

WCAP-15321

A-2 1.2 LTOPS SETPOINT DETERMINATION Westinghouse has developed new LTOPS setpoints for Sequoyah Unit 2, based on a vessel exposure of 32 EFPY using the methodology established in WCAP-14040 (Ref. 2).

This methodology maximizes the available operating margin for setpoint selection while maintaining an appropriate level of protection in support of reactor vessel integrity. Note, Appendix G pressure limit relaxation allowed by ASME Code Case N-514 was not applied.

Plant design characteristics are unchanged (i.e., heat injection and mass injection transients characteristics and related plant responses have not been altered). Therefore, a complete reanalysis is not required. The new LTOPS setpoints were developed using results of the previous heat and mass injection transierit analyses.

1.2.1 Pressure Limits Selection The function of the LTOPS is to protect the reactor vessel from fast propagating brittle fracture.

This has been implemented by choosing LTOPS setpoints, which prevent exceeding the limits prescribed by the applicable pressure/temperature characteristic for the specific reactor vessel material in accordance with rules given in Appendix G to 10CFR50. The LTOPS design basis takes credit for the fact that overpressure events most likely occur during isothermal conditions in the RCS. Therefore, it is appropriate to utilize the steady-state Appendix G limit. The Sequoyah Unit 2, IOCFR50 Appendix G curve for 32 EFPY is shown by Figure A-1. This curve sets the nominal upper limit on the pressure which should not te exceeded during RCS increasing pressure transients based on reactor vessel material properties.

When a relief valve is actuated to mitigate an increasing pressure transient, the system pressure then decreases, as the relief valve releases coolant, until a reset pressure is reached where the valve is signaled to close. Note that the pressure continues to decrease below the reset pressure as the valve re-closes. The nominal lower limit on the pressure during the transient is typically established based solely on an operational consideration for the RCP #1 seal to maintain a nominal differential pressure across the seal faces for proper film-riding performance. The RCP #1 seal limit is shown in Figure A-1.

The nominal upper limit (based on the minimum of the steady-state IOCFR50 Appendix G requirement) and the nominal lower limit (based on RCP #1 seal performance criteria) create a pressure range from which the setlpoints for both PORVs may be selected.

1.2.2 Mass Input Consideration For a particular mass input transient to the RCS, the relief valve will be signaled to open at a specific pressure setpoint. However, there will be a pressure overshoot during the delay time before the valve starts to move and during the time the valve is moving to the full open position. This overshoot is dependent on the dynamics of the system and the input parameters, and results in a maximum system pressure somewhat higher than the set pressure.

Similarly there will be a WCAP-15321

A-3 pressure undershoot, while the valve is relieving, both due to the reset pressure being below the setpoint and to the delay in stroking the valve closed.

The previous Sequoyah analyses of multiple mass input cases were used to determine the relationship between setpressures and resulting overshoots/undershoots.

1.2.3 Heat Input Consideration The heat input case is done similarly to the mass input case except that the locus of transient pressure values versus selected setpoints may be determined for several values of the initial RCS temperature. This heat input evaluation provides a range of acceptable setpoints dependent on the reactor coolant temperature, whereas the mass input case is limited to the most restrictive low temperature condition only (i.e., the mass injection transient is not sensitive to temperature).

The previous Sequoyah. analyses of multiple heat input cases were used to determine the relationship between setpressures and resulting overshoots/undershoots.

1.2.4 Fimal Setpoint Selection Appendix G limits described in Secti(o 1.2.1, were conservatively adjusted accounting for the pressure difference (AhP) between the wide range pressure transmitter and the reactor vessel limiting beltline region of 68.3 psi for 4 RCPs in operation (See Reference 18).

The results of the analyses described in Section 1.2.2 & 1.2.3. and the adjusted Appendix G limit were used to define the maximum allowable setpoints for which the overpressure will not exceed the pressure limit applicable at a specific reactor vessel temperature. The maximum allowable setpoints are shown in Figure A-1.

Per Ref. 17, Sequoyah Demonstrated Accuracy Calculation SQN-IC014 establishes the instrument loop inaccuracy of the Sequoyah temperature and pressure instrument channels associated with the LTOPS. Previously, the LTOPS setpoints have been provided to TVA without application of instrument uncertainties.

The TVA calculation quantified the instrument channel uncertainties, applied them to the nominal Westinghouse setpoints and evaluated the result against the safety limits established by the IOCFR5O, Appendix G steady state heatup curve. The TVA demonstrated accuracy calculation will be revised to reflect the LTOPS setpoints calculated by Westinghouse under the subject task.

As such, it is not necessary for Westinghouse to include instrument uncertainties in the nominal LTOPS setpoint calculation.

Note, the heat injection results were adjusted to include 50*F thermal transport effect (difference in temperature between the RCS and steam generator at transient initiation).

The maximum allowable setpoints, adjusted to produce a smoother curve and reduced to nine data points, becomes the setpoints for PORV#2. A setpoint at a minimum temperature of 50°F was selected, as requested by TVA (Ref. 16). Each of the two PORVs may have a different pressure setpoint such that only one valve will open at a time and mitigate the transient (i.e., staggered WCAP-15321

A-4 setpoints). The second valve operates only if the first fails to open on command. This design supports a single failure assumption as well as minimizing the potential for both PORVs to open simultaneously, a condition which may create excessive pressure undershoot and challenge the RCP

1 seal performance criteria.

The PORV#1 setpoints were selected by adjusting the setpoint PORV#2 in relationship to the overshoots discussed in Sections 1.2.2 and 1.2.3.

The selected setpoints for PORV #1 and PORV #2 are shown in Table A-1 and Figure A-2. These setpoints were evaluated using the undershot's discussed in Sections 1.2.2 and 1.2.3 to ensure that they protect against the RCP # I seal limit.

In summary, the selection of the setpoints for LTOPS considered the use of nominal upper and lower pressure limits. The upper limits are specified by the minimum of the steady-state cooldown curve as calculated in" accordance with Appendix G to IOCFR50 (adjusted to account for four RCPs in operation). The lower pressure extreme is specified by the reactor coolant pump #1 seal minimum differential pressure performance criteria. The selected setpoints, shown in Table A-I and Figure A-2, provide protection against Appendix G and RCP #1 seal limit violations, Note, these setpoints do not address instrumentation uncertainties.

1.3 ARMING AND ENABLE TEMPERATURES FOR LTOPS The LTOPS arming temperature is traditionally based on the temperature corresponding to when Appendix G pressure equals 2500 psia.

Based on this methodology the LTOPS arming temperature for Sequoyah conservatively continues to be 350 0F.

The enable temperature is the temperature below which the LTOPS system is required to be operable, based on vessel materials concerns. ASME Code Case N-514 requires the LTOPS to be in operation at coolant temperatures less than 200°F or at coolant temperatures less than a temperature corresponding to a reactor vessel metal temperature less than RTar + 500F.

whichever is greater. RTm-is the highest Adjusted Reference Temperature (ART) for the limiting belt-line material at a distance one fourth of the vessel section thickness from the vessel inside surface (i.e., clad/base metal interface), as determined by Regulatory Guide 1.99, Revision 2. The minimum required enable temperature for the Sequoyah Unit 2 Reactor Vessel is 2250F at 32 EFPY of operation.

WCAP-15321

Table A-I Selected Setpoints, Sequoyah Unit 2 Trcs (Deg.F)

PORV#2 PORV#1 Setpoint (psig) Setpoint (psig) 50 510 485 100 580 555 135 640 610 174 745 682 200 745 685 250 745 685 278 745 685 400 745 685 450 2350 2350 A-5 WCAP-15321

)

Reactor Coolant System Pressure (pslg) 0 0

0 N

I-A 0

H 0

C,,

0 rD 0

4Il C7i U'

I

-4 0 I

'U S

S E

0 t4 0

W 0 0'

.Jr

Figure A Sequoyah Unit 2 LTOPS Selected Setpoints WCAP-15321 Sequoyah Unit 2 LTOPS Selected Setpoints 2500I" I

I I

I B

I I

t I

I I

I s

I I

2000-- - - +- -*--

- --I ---

I---- I-

,*I I

I I

I SI I

I I

1-500-0 I

I I

I SI I

I I

I 1

SP I

I I

I !

0 I

I I

]

I I

S1oo-0 50 100 150 200 250 300 M5 400 450 500 Reactor Cooiant System Temperature (*F)

PORV#2-Fwial -

FORV#1 A-7

B-0 APPENDIX 'B PRESSURIZED THERMAL SHOCK (PTS) RESULTS WCAP-15321

N PTS Calculations:

The PTS Rule requires that for each pressurized water nuclear power reactor for which an operating license has been issued, the licensee shall have projected values of RTrs, accepted by the NRC, for each reactor vessel beltline material for the EOL fluence of the material. This assessment must specify the basis for the projected value of RTyrs for each vessel beltline material, including the assumptions regarding core loading patterns, and must specify the copper and nickel contents and the fluence value used in the calculation. This assessment must be updated whenever there is a significant change in projected values of RTprs, or upon request for a change in the expiration date for operation of the facility. (Changes to RTprs values are considered significant if either the previous value or the current value, or both values, exceed the screening criterion prior to the expiration of the operating license, including any renewed term, if applicable, for the plant.

To verify that RTNDT, for each vessel beltline material is a bounding value for the specific reactor vessel, licensees shall consider plant-specific information that could affect the level of embrittlement. This information includes but is not limited to the reactor vessel operating temperature and any related surveillance program results. (Surveillance program results mean any data that demonstrates the embrittlement trends for the limiting beltline material, including but not limited to data from test reactors or from surveillance programs at other plants with or without surveillance program integrated per 10 CFR Part 50, Appendix H.)

Calculations:

Tables B-I and B-2 contain the results of the calculations for each of the beltline region materials in the Sequoyah Unit 2 Reactor Vessel. Per TVA, the EOL is 32 EFPY and the Life Extension EOL is 48 EFPY.

WCAP-15321

TABLE B-1 RTmrs Calculations for Sequoyah Unit 2 Beltline Region Materials at 32 EFPY Material Fluence FF CF ARTnrs(c Margin

RTNUr, RTyrs,)

(nlkm, E>1.0 (OF)

(OF)

MeV)

Intermediate Shell Forging 05 1.82 1.164 95 110.6 34 10 155 Intermediate Shell Forging 05 1.82 1.164 89.7 104.4 34 10 148 (Using SiC Data) 1._2 1.164 89.7 104.4

_34118 Lower Shell Forging 04 1.82 1.164 104 121.1 34

-22 133 Circumferential Weld Metal 1.82 1.164 63 73.3 56

-4 125 Circumferential Weld Metal 1.82 1.164 77.8 90.6 56

-4 143 (Using S/C Data)

Notes:

(a)

Initial RT)nrr values are measured values (b)

RT7rs = RTmrum + ARTprs + Margin (OF)

(c)

ARTrts = CF

FF TABLE B-2 RTMrs Calculations for Sequoyah Unit 2 Beltline Region Materials at 48 EFPY Material Fluence FF CF ARTy(*c Margin RT (a,ý') RT7rs*)

(nVlc, E>1.0 (OF)

(OF)

MeV)

Intermediate Shell Forging 05 2.71 1.266 95 120.3 34 10 164 Intermediate Shell Forging 05 2.71 1.266 89.7 113,6 34 10 158 (Using S/C Data)

I I

Lower Shell Forging 04 2.71 1.266 104 131.7 34

-22 144 Circumferential Weld Metal 2.71 1.266 63 79.8 56

-4 132 Circumferential Weld Metal 2.71 1.266 77.8 98.5 56

-4 151 (Using SIC Data)

Notes:

(a)

Initial RTrm values are measured values (b)

RTrrs = RTmTr

+ ARTprs + Margin (OF)

(c)

ARTprs = CF

FF WCAP-15321

-V

B-3 All of the beltline materials in the Sequoyah Unit 2 reactor vessel are below the screening criteria values of 270OF and 300OF at 32 and 49 EFPY.

WCAP-15321

C-0 APPENDIX C CALCULATED FLUENCE DATA WCAP-15321

C-1 The best estimate exposure of the Sequoyah Unit 2 reactor vessel presented in WCAP-15320171 was developed using a combination of absolute plant specific transport calculations and all available plant specific measurement data. The evaluation is consistent with the methodology accepted by the NRC and documented in WCAP-14040-NP-A[21.

Combining this measurement data base with the plant-specific calculations, the best estimate vessel exposure is obtained from the following relationship:

4Dpest.

- K C)O1ca where:

=

The best estimate fast neutron exposure at the location of interest.

K

=

The plant specific best estimate/calculation (BE/C) bias factor derived from the surveillance capsule dosimetry data.

c*.

=

The absolute calculated fast neutron exposure at the location of interest.

For Sequoyah Unit 2, the derived plant specific bias factors were 0.93, 0.98, 0.96 for (Z(E > 1.0 MeV),

(b(E > 0.1 MeV), and dpa, respectively. Bias factors of this magnitude developed with BUGLE-96 are within expected tolerances for fluence calculated using the ENDF/B-VI based cross-section library.

Table C-1 presents the reactor vessel fast neutron (E > 1.0 MeV) exposure projections using the absolute plant specific calculations. Table C-2 presents the calculated and measured fluences at the capsules.

WCAP-15321

Table C-1 Azimuthal Variations Of The Neutron Exposure Projections On The Reactor Vessel CladfBase Metal Interface At Core Midplane Calculated 00 150 300 4501o1 10.54 EFPY E>I.0 MeV 2.11E+18 3.36E+18 4.26E1+18 6.37E+18 E>O.I MeV" 5.37E+18 8.50E+18 1.lIE+19 1.70E+19 dpa 3.43E-03 5.39E-03 6.88E-03 1.03E-02 20 EFPY E>1.0 MeV 3.g0E+18 6.00E+18 7.73E+ 18 1.16E+19 E>0.I MeV 9.65E+18 1.52E+19 2.O1E+19 3.1OE+19 dpa 6.16E-03 9.61E-03 1.25E-02 1.88E-02 32EFPY E>I.0 MeV 5.93E+18 9.34E+18 1.21E+19 1.82E+19 E>0.1 MeV 1.511E+19 2.36E+19 3.16E+19 4.88E+19 dpa 9.63E-03 1.50E-02 1.96E-02 2.95E-02 48 EFPY E>1.0 MeV 8.78E+18 1.38E+19 1.80E+19 2.71E+19 E>0.1 MeV 2.23E+19 3.49E+19 4.68E+19 7.24E+19 dpa 1.42E-02 2.21E-02 2.91E-02 4.38E-02 Note:

a) Maximum neutron exposure projection WCAP-15321 C-2

C-3 Table C-2 Comparison Of Calculated And Best Estimate Integrated Neutron Exposure Of Sequoyah Unit 2 Surveillance Capsules T, U, X, andY CAPSULE T Calculated Best Estimate BE/C (D(E > 1.0 McV) [n/crr2]

2.61E+18 2.57E+18 0.98 (D(E> 0.1 MeV) [n/cm2]

8.74E+18 8.98E+18 1.03 dpa 4.34E-03 4.36E-03 1.01 CAPSULE U Calculated Best Estimate BE/C I(E> 1.0 MeV) tnrcm2j 6.92E+18 6.03E+18 0.87

<D(E > 0.1 MeV) [n/cm 2]

2.31E+19 2.11E+19 0.91 dpa 1.15E-02 1.03E-02 0.90 CAPSULE X Calculated Best Estimate BE/C 4D(E > 1.0 MeV) [n/cm2J 1.22E+19 1.04E+19 0.85 4(E > 0.1 MeV) [n/cm'I]

4.09E+19 3.63E+19 0.89 dpa 2.03E-02 1.77E-02 0.87 CAPSULEY Calculated Best Estimate BE/C

((E > 1.0 McV) [n/cm2]

2.14E+19 2.18E+19 1.02 a)(E > 0.1 MeV) [n/cm2]

7.14E+19 7.72E+19 1.08 dpa 3.54E-02 3.70E-02 1.05 AVERAGE BE/C RATIOS BE/C 4)(E > 1.0 MeV) [n/cm 2]

0.93 D(E > 0,1 MeV) [n/cr 2]

0.98 L dpa 0.96 WCAP-15321

D-O APPENDIX D UPDATED SURVEILLANCE MATERIAL 30 FT-LB TRANSITION TEMPERATURE SHIFTS AND UPPER SHELF ENERGY DECREASES WCAP-15321

D-1 TABLE D)-I Measured 30 ft-lb Transition Temperature Shifts of all Available Surveillance Data 30 ft-lb Transition Upper Shelf Energy Temperature Shift Decrease Material Capsule Fluence Predicted Measured Predicted Measu red (I liol u/cm1)

(0F) (a)

(OF) (")

(%) (a,

(%)(-)

Intermediate Shell T

0.261 60.33 63.65 17 12 Forging 05 U

0.692 85.22 79.31 21 16 (rangential)

X 1.22 100.23 85.7 23 8

Y 2.14 114.67 134.12 26 22 Intermediate Shell T

0.261 60.33 49.73 17 7

Forging 05 U

0.692 85.22 66.06 21 9

(Axial)

X 1.22 100.23 110.04 23 2

Y 2.14 114.67 89.21 26 22 Weld Metal T

0.261 43.12 74.56 20 2

U 0.692 60.91 130.38 25 6

X 1.22 71.63 44.22 29 35 Y

2.14 81.96 86.91 33 3

HAZ Metal T

0.261 24.58 2

U 0.692 64.03 14 X

1.22 28.29 19 Y

2.14 50.32 39 Notes:

(a)

Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values of copper and nickel of the surveillance material.

(b)

Calculated using measured Charpy data plotted using CVGRAPH, Version 4.1 (Reference 9)

(c)

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

WCAP-15321

E-0 APPENDIX E REACTOR VESSEL BELTLINE MATERIAL PROJECTED END OF LICENSE UPPER SHELF ENERGY VALUES N

WCAP-15321

E-1 TABLE E Predicted End-of-License (32 EFPY) USE Calculations for all the BeItline Region Materials Material Weight %

1/4T EOL Unirradiated Projected USE Projected of Cu Fluence USEa)

Decrease EOL USE (1019 n/cm2)

(ft-Nb)

(%)

(ft-lb)

Intermediate Shell Forging 05 0.13 1.10 93 18.5 76 Using SIC Data Lower Shell Forging 04 0.14 1.10 100 23 77 Intermediate to Lower Shell 0.12 1.10 102 35 66 Circumferential Weld Seam Using S/C Data Notes:

(a)

These values were obtained from Reference 12.

WCAP-15321

F-O APPENDIX F UPDATED SURVEILLANCE CAPSULE REMOVAL SCHEDULE WCAP-15321

F-I The following surveillance capsule removal schedule meets the requirements of ASTM El 85-82 and is recommended for future capsules to be removed from the Sequoyah Unit 2 reactor vessel. This recommended removal schedule is applicable to 32 EFPY of operation.

Table F-I Sequoyah Unit 2 Reactor Vessel Surveillance Capsule Withdrawal Schedule Removal Time Fluence Capsule Location Lead Factorsk (EFPY)W (n/cm2,E>l.0 MeV)(')

T 400 3.33 1.04 2.61 x 10"s (c)

U 1400 3.40 2.93 6.92 x 10'2 (c)

X 2200 3.39 5.36 1.22 x 10"9 (c)

Y 3200 3.35 10.54 2.14 x 10'9 (cd)

S 40 1.09 Standby (de)

V 1760 1.09 Standby (de)

W 1840 1.09 Standby (d,e)

Z 3560 1.09 Standby (de)

Notes:

(a)

Updated in Capsule Y dosimetry analysis (Reference 7).

(b)

Effective Full Power Years (EFPY) from plant startup.

(c)

Plant specific evaluation.

(d)

This fluence is not less than once or greater than twice the peak end of license (32 EFPY) fluence (e)

Capsules S, V, W and Z will reach a fluence of 2.71 x 10"9 (E > 1.0 MeV), the 48 EFPY peak vessel fluence at approximately 44 EFPY, respectively. If vessel fluence data is needed at the EOL for Life Extension, it is recommended that one or more of the Standby Capsules be moved to a higher flux location within the next few cycles of operation.

WCAP-15321

G-0 APPENDIX G ENABLE TEMPERATURE CALCULATIONS AND RESULTS WCAP-15321

G-1 Enable Temperature Calculation:

ASME Code Case N-514 requires the low temperature overpressure (LTOP or COMS) system to be in operation at coolant temperatures less than 200OF or at coolant temperatures less than a temperature corresponding to a reactor vessel metal temperature less than RTmr + 50°F, whichever is greater. RTIDT is the highest adjusted reference temperature (ART) for the limiting beltline material at a distance one fourth of the vessel section thickness from the vessel inside surface (ie. clad/base metal interface), as determined by Regulatory Guide 1.99, Revision 2.

32 EFPY The highest calculated 1/4T ART for the Sequoyah Unit 2 reactor vessel beltline region at 32 EFPY is 1420F.

From the OPERLIM computer code output for the Sequoyah Unit 2 32 EFPY P-T limit curves without margins (Configuration # 1676409813, operlimfilm File) the maximum AT*-, is:

Cooldown Rate (Steady-State Cooldown):

max(AT,) at 1/4T = 00F Heatup Rate of 100°F/Hr:

max (ATtain) at 114T = 28.924'F Enable Temperature (ENBT) =

RT"

-50+ max (ATma), OF

= (142 + 50 + 28.924) *F

= 220.924OF The minimum required enable temperature for the Sequoyah Unit 2 Reactor Vessel is 225OF at 32 EFPY of operation.

WCAP-15321

ENCLOSURE 6 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQN)

UNITS 1 AND 2 JUSTIFICATION FOR EXEMPTION ALLOWING USE OF AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME)

CODE CASE N-640 In accordance with 10 CFR 50.12, "Specific exemptions,"

TVA is requesting an exemption from the requirements of 10 CFR 50.60, "Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation."

The exemption would permit the use of the ASME Boiler and Pressure Vessel (B&PV)

Code,Section XI Code Case N-640, "Alternative Requirement Fracture Toughness for Development of P-T Limit Curves for ASME Section XI, Division 1," in lieu of 10 CFR 50, Appendix G, paragraph IV.A.2.b.

The proposed exemption meets the criteria of 10 CFR 50.12 as discussed below.

10 CFR 50.12(a) Requirements 10 CFR 50.12 states that the Commission may grant an exemption from requirements contained in 10 CFR 50 provided that the following is met:

1. The requested exemption is authorized by law.

No law exists which precludes the activities covered by this exemption request.

10 CFR 50.60(b) allows the use of alternatives to 10 CFR 50, Appendix G, when an exemption is granted by the Commission under 10 CFR 50.12.

2.

The requested exemption does not present an undue risk to the public health and safety.

Revised P-T limit curves were developed for Sequoyah Units 1 and 2.

The revised P-T limits use the K1 c, fracture toughness curve shown on ASME XI, Appendix A, Figure A-4200 1, in lieu of the KIa, fracture toughness curve of ASME XI, Appendix G, Figure G-2210-1, as the lower bound for fracture toughness.

The other margins involved with the ASME B&PV

Code,Section XI, Appendix G process of determining P-T limit curves remain unchanged.

Use of the K1 c curve in determining the lower bound fracture toughness in the development of P-T operating limits curve reduces the excess conservatism in the current Appendix G approach, the application of which could, in fact, reduce overall plant safety.

The KIc curve models the slow heat-up E6-1

and cooldown process of a reactor pressure vessel.

Use of this approach is justified given the initial conservatism built into the Kia curve when the curve was codified in 1974.

The initial conservatism was deemed necessary due to limited knowledge of reactor pressure vessel material fracture toughness.

Since 1974, additional knowledge has been gained about the fracture toughness of reactor pressure vessel materials and their fracture response to applied loads.

The additional knowledge demonstrates the lower bound fracture toughness provided by the KIa curve is well beyond the margin of safety required to protect against potential reactor pressure vessel failure.

The lower bound Kic fracture toughness provides an adequate margin of safety to protect against potential reactor pressure vessel failure and does not present an undue risk to public health and safety.

P-T limit curves that are based on the KIc fracture toughness limits enhance overall plant safety by opening the pressure temperature operating window.

Two primary safety benefits would be realized with no decrease to the margin of safety.

Challenges to the operators would be reduced since the requirements for maintaining high vessel temperature during pressure testing would be lessened.

Enhanced personnel safety would result because of the lower temperatures which would exist during the conduct of inspections in primary containment.

3.

The requested exemption is consistent with the common defense and security.

The subject of this exemption does not affect national defense or security issues.

The common defense and security are not impacted by approval of this exemption request.

4.

Special circumstances are present which necessitate the request for an exemption to the regulations of 10 CFR 50.60.

In accordance with 10 CFR 50.12(a) (2),

NRC will consider granting an exemption to the regulations if special circumstances are present.

This requested exemption meets the special circumstances of the following paragraphs of 10 CFR 50.12:

(a) (2) (ii)

- demonstrates the underlying purpose of the regulation will continue to be achieved; (a) (2) (iii)

- would result in undue hardship or other costs that are significant if the regulation is enforced and; E6-2

(a) (2) (v) - will provide only temporary relief from the applicable regulation and the licensee has made good faith efforts to comply with the regulation.

10 CFR 50.12(a) (2) (ii):

ASME B&PV Code,Section XI, Appendix G, provides procedures for determining allowable loading on the reactor pressure vessel and is approved for that purpose by 10 CFR 50, Appendix G.

Application of these procedures in the determination of P-T operating and test curves satisfy the underlying requirement that:

The reactor coolant pressure boundary be operated in a

regime having sufficient margin to ensure, when stressed, the reactor pressure vessel boundary behaves in a non-brittle manner, and the probability of a rapidly propagating fracture is minimized, and P-T operating and test limit curves provide adequate margin in consideration of uncertainties in determining the effects of irradiation on material properties.

The requirements of ASME B&PV Code,Section XI, Appendix G, were conservatively developed based on the level of knowledge existing in 1974 concerning reactor pressure vessel materials and the estimated effects of operation.

Since 1974, the level of experience and knowledge about these topics has increased.

This increased experience and knowledge permits relaxation of ASME B&PV Code,Section XI, Appendix G, requirements via application of ASME Code Case N-640, while maintaining the underlying purpose of the ASME B&PV Code and the NRC regulations to ensure an acceptable margin of safety.

10 CFR 50.12(a) (2) (iii):

The reactor pressure vessel pressure temperature operating window is defined by the P-T limit curves developed in accordance with the ASME B&PV Code,Section XI, Appendix G requirements.

Continued operation of Sequoyah Units 1 and 2 with these P-T limit curves without the relief provided by ASME Code Case N-640 would unnecessarily restrict the P-T operating window.

This restriction challenges the operations staff during required pressure tests.

The operator must maintain a high temperature within a limited operating window.

In addition, the higher temperatures result in greater physical stress on the inspection personnel working in the vicinity of the piping.

This constitutes an unnecessary burden that can be alleviated by the application of ASME Code Case N-640 in the development of the proposed P-T limit curves.

Implementation of the proposed P-T limit curves, as allowed by ASME Code Case N-640, does not reduce the margin of safety below acceptable limits.

10 CFR 50.12(a) (2) (v):

The requested exemption provides only temporary relief from the applicable regulation.

TVA has made a E6-3

good faith effort to comply with the regulation.

TVA requests the exemption be granted until such time that the NRC generically approves the application of ASME Code Case N-640 for use by the nuclear industry.

Conclusion for Exemption Acceptability:

Compliance with the specified requirement of 10 CFR 50.60 would result in hardship and unusual difficulty without a compensating increase in the level of quality and safety.

TVA' s proposed application of ASME Code Case N-640 for SQN Units 1 and 2 allows a reduction in the lower bound fracture toughness used in ASME B&PV Code,Section XI, Appendix G, in the determination of reactor coolant system P-T limits.

The proposed alternative is acceptable because the ASME code case maintains the relative margin of safety commensurate with that which existed at the time ASME B&PV Code,Section XI, Appendix G, was approved in 1974.

Therefore, application of ASME Code Case N-640 for Sequoyah Units 1 and 2 will ensure an acceptable margin of safety and does not present an undue risk to the public health and safety.

E6-4

ENCLOSURE 7 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQN)

UNITS 1 AND 2 JUSTIFICATION FOR EXEMPTION ALLOWING USE OF WCAP-15315 In accordance with 10 CFR 50.12, "Specific exemptions,"

TVA is requesting an exemption from the requirements of 10 CFR 50.60, "Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation."

The exemption would permit the use of WCAP-15313, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Operating PWR and BWR Plants," in lieu of the methodology required by 10 CFR 50, Appendix G, footnote 2 to Table 1. The WCAP demonstrates that the flange region is tolerant of assumed flaws in excess of 1/4 thickness at room temperature.

In addition, since there is no known degradation mechanism for this region and the fatigue usage in this region is less than 0.1, it is concluded that flaws are unlikely to initiate in this region.

Accordingly, justification exits for eliminating the reactor vessel head/flange region when determining the pressure-temperature (P

T) limits for the reactor vessel.

The proposed exemption meets the criteria of 10 CFR 50.12 as discussed below.

10 CFR 50.12(a) Requirements 10 CFR 50.12 states that the Commission may grant an exemption from requirements contained in 10 CFR 50 provided that the following is met:

1. The requested exemption is authorized by law.

No law exists which precludes the activities covered by this exemption request.

10 CFR 50.60(b) allows the use of alternatives to 10 CFR 50, Appendix G, when an exemption is granted by the Commission under 10 CFR 50.12.

2.

The requested exemption does not present an undue risk to the public health and safety.

The revised P-T curve limits being proposed for Sequoyah Units 1 and 2 rely on methodology described in WCAP-15313.

The WCAP uses a lower stress intensity factor, K~c instead of KIa, which results in higher allowable pressures.

The 10 CFR 50, Appendix G addresses the metal temperature of the closure head flange and vessel flange regions.

The regulation states that the metal temperature of the closure flange regions must exceed the material unirradiated RTNDT by at least 120 degrees Fahrenheit (OF) for normal operation when the pressure E7-1

exceeds 20 percent of the preservice hydrostatic test pressure.

Implementing the Kic stress intensity factor without eliminating the flange requirement of 10 CFR 50, Appendix G, would place a restricted operating window in the temperature range associated with the flange/closure head (i.e.,

flange RTNDT + 120'F).

In accordance with WCAP-15315, the KIc toughness has been shown to provide significant margin between the applied stress intensity factor and the fracture toughness of the flange/closure head.

The conclusion of the WCAP analysis is the integrity of the closure head/flange is not a concern for safe plant operation and testing.

Accordingly, the Sequoyah P-T limit curves are generated without flange requirements included.

3.

The requested exemption is consistent with the common defense and security.

The subject of this exemption does not affect national defense or security issues.

The common defense and security are not impacted by approval of this exemption request.

4.

Special circumstances are present which necessitate the request for an exemption to the regulations of 10 CFR 50.60.

In accordance with 10 CFR 50.12(a) (2),

NRC will consider granting an exemption to the regulations if special circumstances are present.

This requested exemption meets the special circumstances of 10 CFR 50.12(a) (2) (ii) that states:

"Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule;"

The underlying purpose of 10 CFR 50.60 and 10 CFR 50, Appendix G is to protect the integrity of the reactor coolant pressure boundary.

The results of the methods used in WCAP 15315 determined that significant margin exists between the applied stress intensity factor and the fracture toughness at virtually all crack depths when using the Krc toughness, which has been adopted by Section XI for developing Pressure Temperature Limit Curves.

Another objective of the requirements in 10 CFR 50, Appendix G is to assure that fracture margins are maintained to protect against service induced cracking due to environmental effects.

Since the governing flaw is on the outside surface (the inside is in compression) where there are no environmental effects, there is even greater assurance of fracture margin.

Therefore, it can be concluded that the integrity of the closure head/flange region is not a concern for SQN using the KIc toughness.

In addition, there are no known mechanisms of degradation for this region, other than fatigue.

The calculated design fatigue usage for this region is less than E7-2

ML022600424

Contents

Text

SUMMARY

1.1 INTRODUCTION

SUMMARY

1.1 INTRODUCTION

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ML022600424
Site:	Sequoyah (DPR-077, DPR-079)
Issue date:	09/06/2002
From:	James Smith Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GL-96-003, TS 00-14, TVA-SQN-TS-00-14
Download: ML022600424 (175)
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1.1 INTRODUCTION

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1.1 INTRODUCTION

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