ML032521478
| ML032521478 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 09/03/2003 |
| From: | Salas P Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC MB6436, TAC MB6437, TAC MB9484 WCAP-15293, Rev 2, WCAP-15321, Rev 2 | |
| Download: ML032521478 (170) | |
Text
Tennessee Valley Authority, Post Office Box 2000, Soddy-Daisy, Tennessee 37384-2000 September 3, 2003 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Gentlemen:
In the Matter of ) Docket Nos. 50-327 Tennessee Valley Authority 50-328 SEQUOYAH NUCLEAR PLANT (SQN) - UNITS 1 AND 2 - UPDATED PRESSURE TEMPERATURE LIMITS REPORTS (PTLRs) AND TOPICAL REPORTS FOR SQN TECHNICAL SPECIFICATION (TS) CHANGE NO. 00-14
References:
- 1. TVA letter to NRC dated March 28, 2003, "Sequoyah Nuclear Plant (SQN) - Response to Request for Additional Information (RAI)
Regarding Technical Specification Change 00-14, 'Pressure Temperature Limits Report (PTLR) and Request for Exemption from the Requirements of 10 CFR 50, Appendix G' (TAC Nos. MB6436 and MB6437)"
- 2. NRC letter to TVA dated July 31, 2003, "Sequoyah Nuclear Plant, Unit 2 - Issuance of Amendment Regarding Reactor Coolant System Heatup and Cooldown Curves (TAC No.
MB9484) (TS 03-08)"
This letter provides updated information for NRC review of SQN TS Change 00-14. This information is provided in response to NRC questions discussed in Reference 1. TVA' s responses to the NRC questions discuss the need for updating the SQN PTLRs and the associated topical reports. provides the updated PTLRs for each unit (PTLR Revision 4, Unit 1, and PTLR, Revision 5, Unit 2). Enclosure 2 provides the Westinghouse Topical Reports that support the respective PTLRs, (i.e., WCAP-15293, Revision 2 [Unit 1] and WCAP-15321, Revision 2 [Unit 2]).
Pmrted recycedpaper
U.S. Nuclear Regulatory Commission Page 2 September 3, 2003 It may be noted that SQN TS Change 03-08 was recently approved by NRC (Reference 2) to update the Unit 2 heatup and cooldown limits. This TS change included limits based on analysis for the reactor head flange area and extended the applicability of SQN's Unit 2 TS heatup and cooldown limits beyond August 2003. The updated limits provided in the enclosed PTLRs supersede SQ s current Unit 2 TS limits.
Accordingly, TVA plans to submit a supplement to TS change 00-14 to reflect the updated limits for Unit 2. In addition, the TS change supplement will revise the TS administrative section for each unit to update the reference documents that support the enclosed PTLRs.
The next milestone associated with SQN' s heatup and cooldown limits is based on SQN Unit 1 (the current Unit 1 TS limits expire after 16 effective full-power years). TVA' s projection for this milestone is January 2005. NRC review and approval of TS change 00-14 is requested to support continued operation of SQN Unit 1 beyond this milestone.
There are no regulatory commitments being made by this submittal. This letter is being sent in accordance with NRC RIS 2001-05. If you have any questions about this change, please telephone me at (423) 843-7170 or J. D. Smith at (423) 843-6672.
and Industry Affairs Manager I declare under penalty of perjury that the forqoi is true and correct. Executed on this 3 day off Enclosures cc (Enclosures):
Mr. Michael J. Marshall, Jr., Senior Project Manager U.S. Nuclear Regulatory Commission Mail Stop 0-8G9A 11555 Rockville Pike Rockville, Maryland 20852-2739
ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQN)
UNITS 1 AND 2 PRESSURE TEMPERATURE LIMITS REPORTS (PTLR)
, SURE TEMPERATURE LIMITS REPORT QA.Recoyd 888 080825 80
- -. APPROVED Th &as aot feieve NW Coacrer from any partdof Oa -
SponsIbrIy for thO sorrectaess of deslgn. drwos and dimsorns.
Lettr No. N10073 pate: Auust 25. 2003 TENNESSEE YALLW AVPIOJ -
ScEP. (N) BY D.L ndY Tennessee Valley Authority Sequoyah Unit I Pressure Temperature Limits Report Revision 4, July 2003 PROJECT Seqiuovah DISCIPLINE N CONTACT 99NAN-251 787 UNT1 DESC._ RCS Pressur-Temnerature Limt Report DWG/DOC NO. PTLR-1 SHE - OF - REV. 04 DATE 08t25/03 ECN/DCN J FIE N2N4058 RIMS, WTC A-K
PRESSURE TEMPERATURE LEMMIS REPORr Table Of Contents List Of Tables .......... iv List Of Figures........... v 1.0 RCS Pressure Temperature Limits Report (PTLR).1 2.0 Operating Limits .1 2.1 RCS Pressure/Temperature (PM Limits (3.4.9.1) .
3.0 Low Temperature Overpressure Protection System (LCO 3.4.12) .1 3.1 Pressurizer PORV Lift Setting Limits .2 3.2 Aming Temperature .2 4.0 Reactor Vessel Material Surveillance Program .2 5.0 Supplemental Data Tables .3 6.0 References .19 ii
PRESSURE TEMPERATURE LIMiS REPORT List Of Tables Table 2-1 Sequoyah Unit 1 Heatup Limits at 32 EFPY (with Uncertainties for Instrumentation Errors of I00 F and 60 psig) .......................................... 6 Table 2-2 Sequoyah Unit 1 Cooldomn Limits at 32 EFPY (with Uncertainties forInstnnentationErrors of 10TF and 60 psig) 8 Table 3-1 Selected Setpoints, Sequoyah Unit 1.10 Table 4-1 Sequoyah Unit 1 Reactor Vessel Surveillance Capsule Withdrawal Schedule .12 Table 5-1 Comparison of the Sequoyah Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifcs and Upper Shelf Energy Decreases with Regulatory Guide 1.99, revision 2, Predictions .13 Table 5-2 Calculation of Chemistry Factors using Sequoyah Unit 1 Surveillance Capsule Data . 14 Table 5-3 Reactor Vessel Beltline Material Unirradiated Toughness Properties for Sequoyah Unit ......... 15 Table 5-4 Peak Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base Metal Interface for-Sequoyah Unit 1 (1019 ncm2 , E > 1.0 MeV).................. 16 Table 5-5 Sequoyah Unit 1 Calculation of the ART Values for the 1/4T Location @ 32 EFPY . 17 Table 5-6 Sequoyah Unit 1 Calculation of the ART Values for the 3/4T Location @32 EFPY . 17 Table 5-7 Summary of the Limiting ART Values Used in the Generation of the Sequoyah Unit 1 Heatup/Cooldown Curves .18 Table 5-8 RTPTS Calculations for Sequoyah Unit 1 Beltline Region Materials at 32 EFPY . 18 iii
PRESSURE TEMPERATURE LEMIS REPORT List Of Figures Figure 2-1 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 1000 F/hr) Applicable for the First 32 EFPY (w/ Margins for Instrumentation Errors of 100 F and 60 psig)..........................................4 Figure 2-2 Sequoyah Unit 1 Reactor Coolant System Cooldovm Limitations (Cooldown Rates up to 100'F/br) Applicable for the First 32 EFPY (w/ Margins for Instrumentation Errors of 100 F and 60 psig) .5 Figure 3-1 Sequoyah Unit I COMS Setpoints .11 iv
PRESSURE TEMPRATURELMrS REPOEr 1.0 RCS Pressure Temperature Limits Report (PTLR)
This PTLR for Sequoyah Unit has been prepared in accordance with the requirements of Technical Specification (TS) 6.9.1.15. Revisions to the PTLR shall be provided to the NRC after issuance.
This report affects TS 3.4.9.1, RCS Pressure/Temperature Limits (P/T) Limits. All TS requirements associated with Low Temperature Overpressure Protection System (LTOPS) are contained in TS 3.4.12, RCS Overpressure Protection System.
2.0 RCS Pressure and Temperature Limits The limits for TS 3.4.9.1 are presented in the subsections which follow and were developed using the NRC approved methodologies specified in TS 6.9.1.15 with exception of ASME Code Case N-64O[3]
(Use of 4), WCAP-15984-PI 41 (Elimination of the Flange Requirement), 1996 Version of Appendix 4]
and the revised fluencesm. 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 LTOP Transient in accordance with the methodology specified in TS 6.9.1.15.
2.1 RCS Pressure/Temperature (P/T) Limits (LCO - 3.4.9.1) 2.1.1 The minimum boltup temperature is 500 F 2.1.2 The RCS temperature rate-of-change limits are:
- a. A maximum heatup rate of 100F in any one hour period.
- b. A maximum cooldown rate of 100F in any one hour period.
- c. A maximum temperature change of less than or equal to 10F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
2.1.3 The RCD P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticality are specified by Figures 2-1 and 2-2.
3.0 Low Temperature Overpressure Protection System (LCO 3A.12)
The lift setpoints for the pressurizer Power Operated Relief Valves (PORVs) are presented in the subsections which follow. These lift setpoints have been developed using the NRC-approved methodologies specified in Specification 6.9.1.15.
1
PRESSURE TEMPERATURE LIMiTS REPORT 3.1 Pressurizer PORV Lift Setting Limits The pressurizer PORV lift setpoints are specified by Figure 3.1 and Table 3.1 (Ref. 10). The limits for the LTOPS setpoints are contained in the 32 EFPY steady-state curves (Table 2-2), which are beitline conditions and are not compensated for pressure differences between the pressurizer transmitter and the reactor midplane/beltline or for instrument inaccuracies. The pressure difference between the pressurizer transmitter and the reactor vessel midplane/beltline with four reactor coolant pumps in operation is 68.3 psi (Ref. 11).
Note: These setpoints include allowance for the 500 F thermal transport effect for heat injection transients. A demonstrated accuracy calculation (Reference 12) 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 midplane and maximum temperature/pressure instrument uncertainties are applied to the setpoints.
3.2 Arming Temperature The LTOPS anning temperature is based upon the methodology defined in the Sequoyah Nuclear Plant Unit 1 Technical Specifications Administrative Controls Section 6.9.1.15. The arming temperature shall be 5 350 0F.
4.0 Reactor Vessel Material Surveillance Program The reactor vessel material irradiation surveillance specimens shall be removed and examined to determine changes in material 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-823311]) is in compliance with Appendix H to 10 CFR 50, "Reactor Vessel Material Surveillance Program Requirements"'21 . The material test requirements and the acceptance standard utilize the reference nil-ductility temperature RTNDT, which is determined in accordance with ASTM E23p]. The empirical relationship between RTNDT and the fracture toughness of the reactor vessel steel is developed in accordance with Code Case N-640 of Section Xl of the ASME Boiler and Pressure Vessel Code, Appendix G, "Fracture Toughness Criteria for Protection Against Failure 41. The surveillance capsule removal schedule meets the requirements of ASTM E185-821'1. The removal schedule is provided in Table 4-1.
2
PRESSURE TEMPERATURE LMrIS REPORT 5.0 Supplemental Data Tables Table 5-1 contains a comparison of measured surveillance material 30 f$-lb transition temperature shifts and upper shelf energy decreases with Regulatory Guide 1.99, Revision 216, predictions.
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 ratio in question is equal to 0.90.
Table 5-3 provides the required Sequoyah Unit 1 reactor vessel toughness data.
Table 5-4 provides a summary of the fluence values used in the generation of the heatup and cooldown limit curves and the PTS evaluation.
Table 5-5 and 5-6 show the calculation of the I/4T and 314T adjusted reference temperature at 32 EFPY for each beltline material in the Sequoyah Unit 1 reactor vesseL The limiting beltline material was the lower shell 04.
Table 5-7 provides a summary of the adjusted reference temperature (ART) values of the Sequoyah Unit 1 reactor vessel beltline materials at the 1/4T and 3/4T locations for 32 EFPY.
Table 5-8 provides RTPTS values for Sequoyah Unit 1 at 32 EFPY.
3
PRESSURE TEMPERATURE LMMS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 32 EFPY: 1/4T, 216°F 3/4T, 186°F 2500250 lperlim Verslon:5.1 Run:15680 I-2250 ........................ . Leak Test L..it . . .
Unacceptable l l Acceptable 2000 . I.....-Operation 0 rat Heatup Rate Critical Limit o 15/ l100 Dog FlHrl100Dm FIHr efi1250 . 1. <; ot 7 1000 - - ---- --- ............. ... ...... ............... ..-........
750j Criticality Limit based on inservice hydrostatic test temperature (288 F) for the 500 - _ f + period srvice ; up to 32 EFPY 2 5 0- . ..............
250.~~~Boltup .Botp....
Temp = 0F 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 2-1 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 100Fhr) Applicable for the First 32 EFPY (wlfiargins for Instrumentation Error of 10°F and 60 psig) (PlottedDataprovided on Table 2-1).
4
PRESSURE TEMPERATURE LAMS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 32 EFPY: 1/4T, 216 0F 314T, 1860 F 2500 2250 2000 1750 CL 1500 CD 0
U O 1250 A.
M la IS
,, 1000 750 500 250 0
0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 2-2 Sequoyah Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 32 EFPY (w/Margins for Instramentation Error of 10F and 60 psig) (PlottedDataprovidedon Table 2-2) 5
PRESSURE TE RATURE LIMIS REPORr Table 2-1 Sequoyah Unit 1 Heatup Limits at 32 EFPY (with Uncertainties for Instrumentation Errors of IOF and 60 psig) 100 Heatup 100 Critical Limit eakTest Limit T P I T P T P 50 0 288 0 272 2000 50 477 288 -477 288 2485 55 477 288 477 60 477 288 477 65 477 288 477 70 477 288 478 75 477 288 478 80 477 288 480 85 477 288 481 90 477 288 483 95 477 288 485 100 477 288 487 105 477 288 490 110 477 288 493 115 477 288 497
. 120 477 288 500 125 477 288 505 130 477 288 508 135 477 288 515 140 477 288 517 145 477 288 527 150 477 288 528 155 478 288 541 160 480 288 541 165 483 288 555 170 487 288 557 175 493 288 571 180 500 288 575 185 508 288 589 190 517 288 609 195 528 288 631 200 541 288 656 205 555 288 684 210 571 288 714 215 589 288 748 220 609 290 786 usC 421 IYOC SIR7 6
PRESSURE ThMPERATURE LIS REPORr Table 2 (Continued)
Sequoyah Unit 1 Heatup Limits at 32 EFPY (with Uncertainties for Instrumentation Errors of 1OF and 60 psig) 100 Heatup 100 Critical Limit T P T P .
230 656 300 874 235 684 305 925 240 714 310 981 245 748 315 1044 250 786 320 1112 255 828 325 1188 260 874 330 1272 265 925 335 1364 270 981 340 1466 275 1044 345 1578 280 1112 350 1702 285 1188 355 1838 290 1272 360 1988 295 1364 365 2154 300 1466 370 2337 305 1578 310 1702 315 1838 320 1988 325 2154 330 2337 7
PRESSURE TEMPERATURE UMITS REPORT Table 2-2 Sequoyah Unit 1 Cooldown Limits at 32 EFPY (with Uncertainties for Instrumentation Errors of 10 0F and 60 psig)
Steady State 20F Pt
[ 40F T P I 60F T
POF P I T P T P I 50 0 50 0 50 0 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 95 321 100 571 100 524 100 476 100 426 100 325 105 575 105 527 105 479 105 430 105 329 110 578 110 531 110 483 110 434 110 333 115 582 115 535 115 487 115 438 115 338 120 586 120 540 120 492 120 443 120 344 125 591 125 545 125 497 125 449 125 351 130 596 130 550 130 503 130 456 130 358 135 602 135 556 135 510 135 463 135 367 140 608 140 563 140 517 140 471 140 376 145 616 145 571 145 525 145 479 145 387 150 623 150 579 150 534 150 489 150 399 155 632 155 588 155 544 155 500 155 412 160 642 160 599 160 556 160 512 160 427 165 652 165 610 165 568 165 526 165 443 170 664 170 623 170 582 170 541 170 461 175 677 175 637 175 597 175 558 175 482 180 691 180 652 180 614 180 577 180 505 185 707 185 669 185 633 185 597 185 530 190 724 190 688 190 654 190 620 190 558 195 743 195 709 195 677 195 646 195 590 200 764 200 733 200 702 200 674 200 624 205 788 205 759 205 731 205 705 205 663 210 814 210 787 210 762 210 740 210 706 215 843 215 819 215 797 215 779 215 754 8
PRESSURE TEMPERATURE LIMTS REPORr Table 2 (Continued)
Sequoyah Unit 1 Cooldown Limits at 32 EFPY (without Uncertainties for Instrumentation Errors)
Steady State 20F 40F 60F 1OOF T P T P T P T P T P 220 874 220 853 220 836 220 821 220 806 225 909 225 892 225 878 225 869 225 865 230 948 230 935 230 925 230 921 235 991 235 982 235 978 240 1038 240 1034 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 9
PRESSURE TEMPERATURE LIMITS REPORr Table 3-1 Selected Setpoints, Sequoyah Unit 1 Trcs (Deg.F) PORV#2 PORV#1 Setpoint Setpoint (psig)
(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 10
PRESSURE ElNMPERATURE LIMS REPORT Sequoyah Unit I LTOPS Selected Setpolnts 2500
_ 2000
.-c a
E 1500 CL 0
E 2
t 1000
.5 a
0: 00 0 50 100 150 200 250 300 350 400 450 500 Reactor Coolant System Temperature (F) l PORM2Sepoint - PORV#1 Setpot 11
PRESSURE 1EMPPERATU L1MNIS REPORTr Figure 3-1: Sequoyah Unit 1 LTOPS Selected Setpoints (PlottedDataprovidedon Table 3-1)
Table 4-1 Sequoyah Unit 1 Reactor Vessel Surveillance Capsule Withdrawal Schedule Removal Ine Fluence Capsule Location Lead Factor(') (EFP) (n/cm2 ,E>1.0 MeV)(a)
T 40° 339 1.03 2.61 x 10' (c)
U 140° 3.47 3.00 7.96 x 10'8 (c)
X 2200 3.47 5.27 132 x109 (c)
Y 3200 3.43 10.03 2.19 x 1019 (c,d)
S 40 1.08 Standby (de)
V 1760 1.08 Standby (de)
W 184° 1.08 Standby (dwe)
Z 3560 1.08 Standby (de)
Notes:
(a) Updated in Capsule Y dosimetry analysis (WCAP-15224[M).
(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 1019 (E > 1.0 MeV), the 48 EFPY peak vessel fluence at approximately 44 EFPY, respectively.
12
PRESSURE TEMPERATURE LIMIS REPORr Table 5-1 Companson of the Sequoyah Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 30 ft-lb Transition Upper Shelf Energy Temperature Shift Decrease Material Capsule Fluence Predicted Measured Predicted Measured (x 101" n/cm2) (.F) )(a) z b) (%)(a) ()
Lower Shell T 0.261 59.85 67.52 16 16 Forging 04 U 0.796 893 109.7 20.5 21 (Tangential) X 1.32 102.6 145.12 23 8 (Heat#980919 Y 2.19 114.95 129.87 26.5 23 2815 87 _ _ _ _ ___ _ _ _ _
Lower Shell T 0.261 59.85 5059 16 0 Forging 04 U 0.796 893 6759 20.5 19 (Axia) X 1.32 102.6 10334 23 22 (Heat #980919/ Y 2.19 114.95 13335 265 19 281587) .
Weld Metal T 0.261 111.13 127.79 35 30 (Heat #25295) (d) U 0.796 165.82 144.92 42 26 X 132 190.51 159.02 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 132 95.89 3 Y 2.19 733 10 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['].
(c) Values are based on the definition of upper shelf energy given in ASTM El 85-82.
(d) Surveillance Weld was fabricated from weld wire type SMIT 40, heat # 25295, Flux Type SMIT 89, Lot # 1103.
13
PRESSURE TEMPERATURE LIMITS REPORT Table 5-2 Calculation of Chemistry Factors using Sequoyah Unit 1 Surveillance Capsule Data Material Capsule Capsule f FPO) ARTNDTc) FF*ARTNDT FE Lower Shell T 2.61E+18 0.63 67.52°F 4254°F 0.40 Forging 04 U 7.96E+18 0.94 l09.7 0F 103.12°F 0.88 0
(Tangential) X 132E+19 1.08 145.12 0F 156.73 F 1.16 (Heat #980919 / y 2.19E+19 1.21 129.87°F 157.14°F 1.47 281587)
Lower Shell T 2.61E+18 0.63 50.59°F 31.87°F 0.40 Forging 04 U 7.96E+18 0.94 6759°F 63.53°F 0.88 (Axial) X 1.32E+19 1.08 103.34°F 111.61°F 1.16 (Heat #980919 Y 2.19E+19 1.21 133.35°F 161350 F 1.47 281587)
SUM: 827.89°F 7.82 CFo4= XFF
- RTm) + (FF 2 ) = (827.89)+ (7.82) = 105.9.F Surveillance Weld T 2.61E+18 0.63 115.0°F 725°F 0.40 Materia fd) U 7.96E+18 0.94 130AF 122.60F 0.88 (Heat #25295) (e) X 132E+19 1.08 143.10F 1545 0F 1.16 Y 2.19E+19 1.21 147.4°F 178.40F 1.47 SUM: 528.0F 3.91 CSF Sr.Wild = (FF
- RTND) 7 ( FF 2 ) = (528.0'F) + (3.91) =135.0°F Notes:
(a) f = Calculated fluence from capsule Y dosimetry analysis results (7), (x 10"9 n/cm2 , E > 1.0 MeV).
(b) FF = fluence factor = f20.OJ-o (c) ARTmT values are the measured 30 ft-lb shift values taken from App. B ofRef. 7, rounded to one decimal point (d) The surveillance weld metal ARTNDT values have been adjusted by a ratio factor of 0.90.
(e) Surveillance Weld was fabricated from weld wire type SMIT 40, heat # 25295, Flux Type SMIT 89, lot# 1103 14
PRESSURE TEMPERATURE LIMMIS REPORr Table 5-3 Reactor Vessel Beline Material Unirradiated Toughness Properties for Sequoyah Unit I Material Description Cu (/) Ni(%) Initial RTNDT~a Intermediate Shell Forging 05 0.15 0.86 400 F (Heat # 980807/281489)
Lower Shell Forging 04 0.13 0.76 730F (Heat# 980919/21587"_______
Surveillance Weld (Heat # 25295f" "d) 0387 0.11 ---
Rotterdam Test' *= 030 --- ---
Rotterdam Testf" => 0.25 --- ---
Rotterdam Tset ) =0.46-Best Estimate of the Intermediate to Lower Shell 0.35 0.11 -400F Forging Circumferential Weld Seam W05 Heat #
25295) (d,.)
Notes:
(a) The Initial RTmT values are measured values (b) These copper and nickel values are best estimate values for only the surveillance weld metal and is the average oftree data points [0.424 (WCAP-10340, Rev.l), 0.406 (WCAP-10340, Rev.1), 0.33 (WCAP-8233) copper and 0.084 (WCAP-10340, Rev.1), 0.085 (WCAP-10340, Rev.l), 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. 174), 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 fabricatedwithweldwiretype SMIT 40, Heat # 25295, Fluxtype SMT 89, lot # 2275. The surveillance weld was fabricated with weld wire type SMIT 40, Heat # 25295, Flx 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 Seam.
15
PRESSURE TEPERATURE LMTS REPORT Table 5-4 Peak Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base Metal Interface for Sequoyah Unit 1 (1019 n/cu?, E > 1.0 MeV)
Azimuthal Location EFPY 00 150 300 450 10.03 0.205 0.321 0.409 0.637 20 0387 0.596 0.761 1.18 32 0.605 0.928 1.19 1.84 48 0.896 137 1.75 2.72 16
PRESSURE TEMPERATURE LMTS REPORr Table 5-5 Sequoyah Unit 1 Calculation of the ART Values for the 1/4T Location @ 32 EFPY()
Material RG 1.99 R2 CF FF IRTNor(l) ARTNr 3( ) Margin 4 ) ART()
Method (OF) (OF) (OF) (0F) (OF)
Intermediate ShellForging 05 Position 1.1 115.6 1.029 40 119.0 34 193 Lower Shell Forging 04 Position 1.1 95 1.029 73 97.8 34 205 Position 2.1 1059 1.029 73 109.0 34(5) 216 Intermediate to Lower Shell Position 1.1 1613 1.029 -40 166.0 56 182 Circumferential Weld Seam Position 2.1 135.0 1.029 -40 138.9 5 155 Notes:
(1) Initial RTNm values measured values.
(2) ART = Inial RTvr + ARITNE + Margin (OF)
(3) ARTmT = CF
- FF (4) M = 2 *((e + a,,)
(5) Data deemed not-credible (See Reference 7a), thus the fAll A will be used to determine margin.
(6) Neutron Fluence value used for all material is the highest value from Table 54 for 32 EFPY.
Table 5-6 Sequoyah Unit 1 Calculation of the ART Values for the 314T Location @ 32 EFPY(6 Material RG 1.99 R2 CF FF IRTNMr1 ) &RTNM(3) Margin(4) ART(2)
Method (OF) (OF) (OF) (OF) (OF)
Intermediate Shell Forging 05 Position 1.1 115.6 0.747 40 86.4 34 160 Lower Shell Forging 04 Position 1.1 95 0.747 73 71.0 34 178 Position 2.1 105.9 0.747 73 79.1 34) 186 Intermediate to Lower Shell Position 1.1 161.3 0.747 -40 1205 56 137 Circumferential Weld Seam Position 2.1 135.0 0.747 -40 100.8 56s 117 No (1) Initial RTmn values measured values.
(2) ART = Initial RTNDr + ARTN + Margin (0F)
(3) ARTm = CF
- F (4) M = 2 *(q + A2)
(5) Data deemed not-credible (See Reference 7a), thus the fll ar will be used to determine margin.
(6) Neutron Fluence value used for all material is the highest value from Table 5-4 for 32 EFPY.
17
PRFSSURE TEMPERATURE LIMTS REPORr Table 5-7 Summary of the Sequoyah Unit 1 Reactor Vessel Beltline Matenal ART Values Material RG 1.99 R2 1/4 ART 3/4 ART Method (OF) (0F)
Intermediate Shell Forging 05 Position 1.1 193 160 Lower Shell Forging 04 Position 1.1 205 178 Position 2.1 216 186 Intermediate to Lower Shell Position 1.1 182 137 Circumferential Weld Seam Position 2.1 155 117 Table 5-8 RTPTS Calculations for Sequoyah Unit 1 Beltine Region Materials at 32 EFPY(e)
...~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~b Material Fluence FF CF ARTpns) Margin RTNry(') RTmps (n/CM2 , E 1.0 (OF) (OF)(r (OXF) (OF)
_ __ __ _ __ _ _ _ _ _ _ _ _ MeV)_ _ _ __ _ _
Intermediate Shell Forging 05 1.84 1.167 115.6 134.9 34 40 209 Lower ShellForging 04 1.84 1.167 95.0 110.9 34 73 218 Lower ShellForging 04 1.84 1.167 105.9 123.6 34(d) 73 231 (Using S/C Data)
Circumferential Weld Metal 1.84 1.167 1613 188.2 56 -40 204 Circumferential Weld Metal 1.84 1.167 135.0 157.5 56d) -40 174 (Using S/C Data)
Notes:
(a) 1fital RTNDT values are measured values (b) RTpTs = RTNDU + ARTps + Margin (F)
(c) ARTpTs = CF
- FF (d) Data deemed not-credible (See Reference 7a), thus the fll cA, will be used to determine margin.
(e) Neutron Fluence value used for all material is the highest value from Table 5-4 for 32 EFPY.
18
PRESSURE TEMPERATURE IMTS REPORT 6.0 References
- 1. WCAP-8233, Tennessee Valley Authority Sequoyah Unit No. 1 Reactor Vessel Radiation Surveillance Program,S. E. Yanichko, et. al, December, 1973.
- 2. Code of Federal Regulations, 10CFR50, Appendix H.,Reactor Vessel MaterialSurveillance ProgramRequirements, U.S. Nuclear Regulatory Commission, Washington, D.C.
- 3. ASTM E23 StandardTest Method Notched Bar Impact Testing of Metallic Materials,in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.
- 4. Section XM of the ASME Boiler and Pressure Vessel Code, Appendix G, FractureToughness Criteriafor ProtectionAgainst Failure
- 5. ASTM E185-82, Annual Book ofASTM Standards, Section 12, Volume 12.02, StandardPractice for Conducting Surveillance Testsfor Light-Water CooledNuclearPower Reactor Vessels.
- 6. Regulatory Guide 1.99, Revision 2, RadiationEmbrittlement ofReactor Vessel Materials,U.S.
Nuclear Regulatory Commission, May 1988.
7a. WCAP-1 5224, Analysis of Capsule Y From the Tennessee Valley Authority Sequoyah Unit Reactor Vessel Radiation Surveillance Program, T.J. Laubham, et. al., Dated June 1999.
7b. WCAP-13333,Analysis of CapsuleXFrom the Tennessee ValleyAuthority Sequoyah Unit 1 Reactor Vessel Radiation Surveillance Program, MA. Ramirez, S. L. Anderson, L. Albertin, Dated June 1992.
7c. SwRI Project 06-8851, Reactor Vessel Material Surveillance Program for Sequoyah Unit No. 1:
Analysis of Capsule U, P. K. Nair, et al., October 1986.
7d. 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.
- 8. CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by Anl Consulting March 1999.
- 9. WCAP-14040-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.
- 10. WCAP-15293, Revision 2, "Sequoyah Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation", J.H. Ledger, July 2003.
19
i PRESSURE TEMPERATURE LIMMS REPOir
- 11. Westinghouse Letter to TVA, TVA-93-105, "Cold Overpressure Mitigation System Code Case and Delta-P Calculation", May 19, 1993.
- 12. Calculation SQN-IC-014, "Demonstrated Accuracy Calculation for Cold Overpressure Protection System."
- 13. ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section A, Division 1", February 26, 1999.
- 14. WCAP-15984-P, Revision 01, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation For Sequoyah Units 1 and 2", W. Bamford, et.al, April 2003.
20
-ESSURE TEMPERATURE LIMITS REPOR11 ll4Reord B 88 080825 &01
- APPROVED SponrubEy lor Ihe correctuess of deuign. d.Iils onE dimensins.
LActc No. N1W073 Pao:.: 7MEt2, 2003 TeVESSEE YiLLU* AUT O -
sEP (N) BY 0. L Umdy Tennessee Valley Authority Sequoyah Unit 2 Pressure Temperature Limits Report Revision 5, July 2003 PROJECT SecUoavah DISCIPLINE N CONTRACT 99NAN-251787 UNIT 2 DESC. RCS Pressure-Tanemrature Lmit Report DWG/DOC NO. PTLR-2 SHEET -OF - REV. 05 DATE 08/25/03 ECN/DCN - FILE N2N-058 RIMS, WTC A-K
PRESSURE TEMPERATURE L1MHIS REPO~r Table Of Contents List Of Tables ........... iv List Of Figures ........... v 1.0 RCS Pressure Temperature Limits Report (PTLR) .1 2.0 Operating Limits .1 2.1 RCS Pressure/Temperature (Pi[) Limits (3.4.9.1) .
3.0 Low Temperature Overpressure Protection System (LCO 3.4.12) .1 3.1 Pressurizer PORV Lift Setting Limits .2 3.2 Arming Temperature .2 4.0 Reactor Vessel Material Surveillance Program .2 5.0 Supplemental Data Tables .3 6.0 References .17 i
PRESSURE TEMPERATURE LIMIS REPORr List Of Tables Table 2-1 Sequoyah Unit 2 Heatup Limits at 32 EFPY (withUncertainties for InstumentationErrors of 10F and 60 pig)........................................... 6 Table 2-2 Sequoyah Unit 2 Cooldown Limits at 32 EFPY (with Uncertainties for Instrumentation Errors of 100F and 60 psig) .......................................... 7 Table 3-1 Selected Setpoints, Sequoyah Unit 2 ..................................................... 8 Table 4-1 Sequoyah Unit 2 Reactor Vessel Surveillance Capsule Withdrawal Schedule ................. 10 Table 5-1 Comparison of the Sequoyah Unit 2 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, revision 2, Predictions ......................................................... 11 Table 5-2 Calculation of Chemistry Factors using Sequoyah Unit 2 Surveillance Capsule Data ...... 12 Table 5-3 Reactor Vessel Beltline Material Unirradiated Toughness Properties for Sequoyah Unit 2 .13 Table 5-4 Peak Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base Metal Interface for Sequoyah Unit 2 (1019 n/cm2, E > 1.0 MeV) .................. 14 Table 5-5 Sequoyah Unit 2 Calculation of the ART Values for the 1/4T Location @ 32 EFPY ...... 15 Table 5-6 Sequoyah Unit 2 Calculation of the ART Values for the 3/4T Location @ 32 EFPY ...... 15 Table 5-7 Summary of the Limiting ART Values Used in the Generation of the Sequoyah Unit 2 Heatup/Cooldown Curves ...................................................... 16 Table 5-8 RTPTS Calculations for Sequoyah Unit 2 Betline Region Materials at 32 EFPY ............ 16 m..
PRESSURE TEPERATURE LIMITS REPORr List Of Figures Figure 2-1 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 1000F/hr) Applicable for the First 32 EFPY (w/ Margins for Instrumentation Errors of 100 F and 60 psig)........................................ 4 Figure 2-2 Sequoyah Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F~r) Applicable for the First 32 EFPY (w/ Margins for Instrumentation Errors of I 0F and 60 psig) .5 Figure 3-1 Sequoyah Unit 2 COMS Setpoints................................................................................9 iv
PRESSURE TIMERATURE LIMMIS REPORr 1.0 RCS Pressure Temperature Limits Report (PTLR)
This PTLR for Sequoyah Unit 2 has been prepared in accordance with the requirements of Technical Specification (TS) 6.9.1.15. Revisions to the PTLR shall be provided to the NRC after issuance.
This report affects TS 3.4.9.1, RCS Pressure/Temperature Limits (P/1) Limits. All TS requirements associated with Low Temperature Overpressure Protection System (LTOPS) are contained in TS 3.4.12, RCS Overpressure Protection System.
2.0 RCS Pressure and Temperature Limits The limits for TS 3.4.9.1 are presented in the subsections which follow and were developed using the NRC approved methodologies specified in TS 6.9.1.15 with exception of ASME Code Case N-640 111 ]
(Use of K1,), WCAP-15984-P' 21 (Elimination of the Flange Requirement), 1996 Version of Appendix $4]
and the revised fluencesM. 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 LTOP Transient in accordance with the methodology specified in TS 6.9.1.15.
2.1 RCS Pressure/Temperature (P/T) Limits (LCO - 3.4.9.1) 2.1.1 The minimum boltup temperature is 500 F 2.1.2 The RCS temperature rate-of-change limits are:
- a. A maximum heatup rate of 1000 F in any one hour period.
- b. A maximum cooldown rate of 100F in any one hour period.
- c. A maximum temperature change of less than or equal to 10F in any one hour period duing inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
2.1.3 The RCD P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticality are specified by Figures 2-1 and 2-2.
3.0 Low Tenperature Overpressure Protection System (LCO 3.4.12)
The lift setpoints for the pressurizer Power Operated Relief Valves (PORVs) are presented in the subsections which follow. These lift setpoints have been developed using the NRC-approved methodologies specified in Specification 6.9.1.15.
1
PRESSURE TEMPERATURE LMrIS REPORT 3.1 Pressurizer PORV Lift Setting Limits The pressurizer PORV lift setpoints are specified by Figure 3.1 and Table 3.1 (Ref. 10). The limits for the LTOPS setpoints are contained in the 32 EFPY steady-state curves (Table 2-2), which are beitline conditions and are not compensated for pressure differences between the pressurizer transmitter and the reactor midplaneAbeltline or for instrument inaccuracies. The pressure difference between the pressurizer transmitter and the reactor vessel midplanelbeltline with four reactor coolant pumps in operation is 68.3 psi (Ref. 13).
Note: These setpoints include allowance for the 50'F thermal transport effect for heat injection transients. A demonstrated 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 midplane and maximum temperature/pressure instrument uncertainties are applied to the setpoints.
3.2 Arming Temperature The LTOPS 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 temperature shall be g 3500F.
4.0 Reactor Vessel Material Surveillance Program The reactor vessel material irradiation surveillance specimens shall be removed and examined to determine changes in material 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-8513[' 1 ) is in compliance with Appendix H to 10 CFR 50, "Reactor Vessel Material Surveillance Program Requirements" 2 ]. The material test requirements and the acceptance standard utilize the reference nil-ductility temperature RTNDT, which is determined in accordance with ASTM E23 3. The empirical relationship between RTNDT and the fracture toughness of the reactor vessel steel is developed in accordance with Code Case N-640 of Section Xl of the ASME Boiler and Pressure Vessel Code, Appendix G, "Fracture Toughness Criteria for Protection Against Failhre!441. The surveillance capsule removal schedule meets the requirements of ASTM E185-82t5l. The removal schedule is provided in Table 4-1.
2
PRESSURE TEMPERATURE LEMiTS REPORr 5.0 Supplemental Data Tables Table 5-1 contains a comparison of measured surveillance material 30 ft-lb transition temperature shifts and upper shelf energy decreases with Regulatory Guide 1.99, Revision 2[2, predictions.
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 ratio in question is equal to 0.93.
Table 5-3 provides the required Sequoyah Unit 2 reactor vessel toughness data.
Table 5-4 provides a summary of the fluence values used in the generation of the heatup and cooldown limit curves and the PTS evaluation.
Table 5-5 and 5-6 show the calculation of the 1/4T and 3/4T adjusted reference temperature at 32 EFPY for each beldine material in the Sequoyah Unit 2 reactor vessel. The limiting beltline material was the intermediate shell 05.
Table 5-7 provides a summary of the adjusted reference temperature (ART) values of the Sequoyah Unit 2 reactor vessel beltline materials at the 1/4T and 3/4T locations for 32 EFPY.
Table 5-8 provides RTprs values for Sequoyah Unit 2 at 32 EFPY.
3
PRESSURE TEMPERATURE IMIS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING 05 LIMlTING ART VALUES AT 32 EFPY: 1/4T, 142°F 3/4T, 115°F 2500 Operlim Verslon:5.1 Run:6694 lLeak Test Llmit 2250 ......................... ........... . ...... ........................ ....... ...... ...........
Unacepable M I I Acceptable 2000 Operation Operation 1750 0
2- 1500 Tieitup Rate~
.I110 Dg. F/HF/Hr-0 E 1250 0.
0 750 0, 1000
., // . '1' _
750 CriticalIlty Limit based on
_ ~~~~~~~~~nservics hydrostatic test 500 _ ~~+ i1, w- temperature214 Ffor the Mi nu mum . , service period up to 32 EFPY Boltup . . .
. _ ...... = 60F /.lTemp - ............... ................. ................
250
. . ... . ... .......... . .. i tu F) for th e . .i ... i 0
0 S0 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 2-1 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100°Flhr) Applicable for the First 32 EFPY (w/ Margins for Instrumentation Error of 10°F and 60 psig) (PlottedDataprovided on Table 2-1) 4
PRESSURE TEMPTURE LIMTS REPORr MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING 05 LIMITING ART VALUES AT 32 EFPY: 14T, I42F 3/4T, 115 0F 2500 2250 2000 1750 F
to
- e. 1500 2-0 1250 a-
.5 1 000 0
U 750 Soo 250 0 0,.... 50 . I. 1 .. ,. .. ... 2 , 20i .. ,..i.....,
3 i . 400
, . I4 T*I 5 550 0 50 100 '150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 2-2 Sequoyah Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 32 EFPY (wi Margins for Instrumentation Error of 107F and 60 psig) (Ploffed Dataprovided on Table 2-2) 5
PRESSUR TEMPERATURE LIMMS REPORT Table 2-1 Sequoyah Unit 2 Heatup Limits at 32 EFPY (with Uncertainties for Instrumentation Errors of 1OF and 60 psig) 100 Heatup 100 Critical Limit ILeak Te Limit T P I T P fT P 50 0 214 0 198 2000 50 591 214 607 214 2485 55 595 214 614 60 601 214 622 65 607 214 657 70 614 214 650 75 622 214 647 80 630 214 646 85 640 214 648 90 646 214 653 95 646 214 661 100 646 214 671 105 646 214 680 110 646 214 685 115 646 214 701 120 646 214 720 125 648 214 743 130 653 214 769 135 661 214 798 140 671 215 832 145 685 220 -869 150 701 225 911 155 720 230 959 160 743 235 1011 165 769 240 1069 170 798 245 1134 175 832 250 1206 180 869 255 1286 185 911 260 1374 190 959 265 1471 195 1011 270 1579 200 1069 275 1698 205 1134 280 1829 210 1206 285 1974 215 1286 290 2134 220 1374 295 2311 225 1471 230 1579 235 1698 240 1829 245 1974 250 2134 2SS 2311 6
PRESSURE TEMPERATURE LIMIS REPORr Table 2-2 Sequoyah Unit 2 Cooldown Limits at 32 EFPY (with Uncertainties for Instrumentation Errors of 100F and 60 psig)
Steady State 20F 40F 60F lOOF T P T P.I T P T P T P 50 0 50 0 50 0 50 0 50 0 50 591 50 552 50 503 50 461 50 366 55 595 55 554 55 508 55 466 55 372 60 601 60 558 60 514 60 470 60 380 65 607 65 564 65 521 65 478 65 389 70 614 70 572 70 529 70 486 70 399 75 622 75 580 75 538 75 496 75 410 80 630 80 589 80 548 80 506 80 423 85 640 85 599 85 559 85 518 85 437 90 650 90 610 90 571 90 531 90 453 95 661 95 623 95 584 95 546 95 470 100 674 100 636 100 599 100 562 100 490 105 688 105 652 105 616 105 580 105 512 110 703 110 668 110 634 110 600 110 536 115 720 115 687 115 654 115 622 115 563 120 739 120 707 120 676 120 647 120 593 125 760 125 730 125 701 125 674 125 626 130 783 130 755 130 729 130 704 130 663 135 809 135 783 135 759 135 738 135 704 140 837 140 814 140 793 140 775 140 749 145 868 145 848 145 831 145 816 145 800 150 902 150 885 150 872 150 862 150 856 155 940 155 927 155 918 155 913 160 982 160 973 160 968 165 1028 165 1024 170 1080 175 1136 180 1199 185 1268 190 1344 195 1429 200 1522 205 1625 210 1739 215 1865 220 2004 225 2158 230 2328 7
PRESSURE T LMS IEPOR UMBERATURE Table 3-1 Selected Setpoints, Sequoyah Unit 2 Trcs (Deg.F) PORV#2 PORV#1 Setpoint Setpoint (psig)
(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 8
PRESSURE TEMPERATURE LEMTS REPORr Sequoyah Unit 2 LTOPS Selected Setpolnts S--
Q a:
f 2
0 1500 CL E
1 1000 (0
C5 o 500 0 50 100 150 200 250 300 350 400 450 500 Reactor Coolant System Temperature OF)
- PORV#249nal a PORV#1 Figure 3-1: Sequoyah Unit 2 LTOP Selected Setpoints (PloftedDataprovided on Table 3-1) 9
PRESSURE TEMPERATURE LIMMIS REPO~r Table 4-1 Sequoyah Unit 2 Reactor Vessel Surveillance Capsule Withdrawal Schedule Fluence Removal Removal Time me Fluence Lead Factor(^) gXMff) (n/cm2,F>1.0 MeV)tw 333 1.04 261 x 10"(c) 3.40 2.93 6.92 x 10's (c) 339 536 1.22x10 19 (c) 335 10.54 2.14 x 1019 (cd 1.09 Standby (e) 1.09 Standby (e) 1.09 Standby (e) 1.09 Standby (e)
Notes:
(a) Updated in Capsule Y dosimetry analysis (WCAP-15320).
(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 0'9 (E > .0 MeV), the 48 EFPY peak vessel fluence at approximately 44 EFPY.
10
PRESSURE TE:MPERATURE IMITS REPORT Table 5-1 Comparison of the Sequoyah Unit 2 Surveillance Material 30 t-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 30 ft-lb Transition Upper Shelf Energy Temperature Shift Decrease Material Capsule Fluence Predicted Measured Predicted Measured (tnlX 101 ncm) I)(") (O]F) () t') %
Intermediate Shell T 0.261 6033 63.65 17 12 Forging 05 U 0.692 85.22 7931 21 16 (Tangential) X 1.22 100.23 85.7 23 8 (Heat #288757 Y 2.14 114.67 134.12 26 22 981057) _ _ _ _ _ _
Intermediate Shell T 0.261 60.33 48.73 17 7 Forging 05 U 0.692 85.22 66.06 21 9 (Axial) X 1.22 100.23 110.04 23 2 (Heat # 288757/
981057) Y 2.14 114.67 89.21 26 22 Weld Metal T 0.261 43.12 74.56 20 2 (Heat # 4278)d) 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.l181.
(c) Values are based on the definition of upper self energy given in ASTM El 85-82.
(d) Surveillance Weld was fabricated from weld wire type SSIT 89, heat # 4278, Flux Type SMIT 89, lot # 1211.
11
PRESSURE TEPERATURE LIMMIS REPORT Table 5-2 Calculation of Chemisty Factors using Sequoyah Unit 2 Surveillance Capsule Data Matenal Capsule Capsule 0 FFO ARTNDT() FF*ARTNDT FF2 Intermediate Shell T 2.61E+18 0.635 63.7 40.45 0.403 Forging 05 U 6.92E+18 0.897 793 71.13 0.805 (Tangential) X 22E+19 1.055 85.7 90.41 1.113 (Heat # 288757/ Y 2.14E+19 1.207 134.1 161.86 1.457 981057) _
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 1M2E+19 1.055 110.0 116.05 1.113 (Heat # 288757 / Y 2.14E+19' 1.207' 89.2 107.66 1.457 981057)
SUM: 677.77°F 7.556 CF5 = (FF
- RTT) + ( FF) = (677.77) + (7.556) = 89.7°F Surveillance Weld T 2.61E+18 0.635 69.4 (74.6) 44.07 0.403 Material U 6-92E+18 0.897 1213 (130.4) 108.81 0.805 (Heat # 4278)(e) X 1.22E+19 1.055 41.1 (44.2) 4336 1.113 Y 2.14E+19 1.207 80.8 (86.9) 97.53 1.457 SUM: 293.77°F 3.778 2
CFS..wdd= IF*RT)nr,) + (WFF )=(293.77F)- (3.778)=77.8°F Notes:
(a) f = Calculated fluence from capsule Y dosimetry analysis results (7), (x lo"0n/cm2 , E > 1.0 MeV).
(b) FF = fluence factor = f(.28- O.1og (c) ARTNur values are the measured 30 ft-lb shift values taken from App. B ofRef. 7, rounded to one decimal point.
(d) The surveillance weld metal ARTNm values have been adjusted by a ratio factor of 0.93.
(e) Surveillance Weld was fabricated from weld wire type SMIT 89, heat # 4278, Flux Type SMIT 89, lot # 1211.
12
PRESSURE TEMPERTURE LEMMS REPORT Table 5-3 Reactor Vessel Beitline Material Unirradiated Toughness Properties for Sequoyah Unit 2 Material Description Cu (%) Ni(%) Initial RTNDT(a)
Intermediate Shell 05 0.13 0.76 10F (Heat # 288757/ 981057)
Lower Shell Forging 04 0.14 0.76 -220F (Heat # 990469/ 293323)
Intermediate to Lower Shell Forging 0.12 0.11 -40F Circumferential Weld Sean')
Surveillance Weldd° 0.13 0.11 ---
Notes:
(a) The Initial RTmT values are measured values (b) Circumferential Weld Seam was fabricated with weld wire type SM[T 89, Heat #4278, Flux type SM[T 89, lot
- 1211 and is representative ofthe intermediate to lower shell circumferential weld.
13
PRESSURE TEMPERATURE LIMTS REPORT Table 5-4 Peak Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base Metal Interface for Sequoyah Unit 2 (10'9 n/cm 2 , E > 1.0 MeV)
Azimuthal Location EFPY O0 l5 300 45 10.54 0.211 0.336 0.426 0.637 20 0.38 0.60 0.773 1.16 32 0593 0.934 1.21 1.82 48 0.878 1.38 1.80 2.71 14
PRESSURE TEMPERATURE LiMITS REPORr Table 5-5 Sequoyah Unit 2 Calculation of the ART Values for the /4T Location @ 32 EFPY(
Material RG 1.99 R2 CF FF IRTNm,') ARTNwr2 Margin 4) ART Method (F) (OF) (OF) (OF) (F)
Intemediate 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 5) 132 Notes:
(1) Inial RTm values measured values.
(2) ART = Initial RTNrr + ARTNm + Margin (f)
(3) ARTmT = CF
- E (4) M = 2 *(q + 2)P (5) Data deemed not-credible (See Reference 7a), thus the full cr& will be used to determine margin.
(6) Neutron Fluence value used for all material is the highest value from Table 5-4 for 32 EFPY.
Table 5-6 Sequoyah Unit 2 Calculation of the ART Values for the 3/4T Location @ 32 EFPY(6 Material RG 1.99 R2 CF FF IRTxw(') &RTWD?) Margin(4) ART 2 Method (OF) (OF) ( (0F) (Ok)
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 77.5 340) 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 RTNvr values measured values.
(2) ART = Iitial RTmn + ARTIr + Margin (f)
(3) ARTmm = CF
- FF (4) M = 2 *(di2 + X 2)1 (5) Data deemed not-credible (See Reference 7a), thus the full at will be used to determine margin.
(6) Neutron Fluence value used for all material is the highest value from Table 5-4 for 32 EFPY.
15
PRESSURE TMPERATURE LBUTS REPORT Table 5-7 Summary of the Sequoyah Unit 2 Reactor Vessel Belfine Material ART Values Material RG 199 R2 1/4 ART 3/4 ART Method (OF) (OF)
Intermediate Shell Forging 05 Position 1.1 142 115 Position 2.1 136 111 Lower Shell Forging 04 Position 1.1 119 90 Intermediate to Lower Shell Position 1.1 117 99 Circtunferential Weld Seam Position 2.1 132 110 Table 5-8 RTPTS Calculations for Sequoyah Unit 2 Beldine Region Materials at 32 EFPY(e)
Material Fluence FF CF &RTPTs° Margin RTNDt(M) RTpn (n/cm2, E>1.0 (0F) (0F) (OF) (OF)
_ __ _ _ _ _ _ _ _ _ _ _ _ _ M eV)_ _ _ _ _ _ _ _ _ _ _ _ _
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 S/C Data)
Lower Shell Forging 04 1.82 1.164 104 121.1 34(d) -22 133 Circumferential Weld Metal 1.82 1.164 63 733 56 -4 125 Circumferential Weld Metal 1.82 1.164 77.8 90.6 56 (d) -4 143 (Using S/C Data) .
Notes:
(a) Initial RTf, values are measured values (b) RTns = RTNmu) + ARTm + Margin (F)
(c) ARTm = CF
- FF (d) Data deemed not-credible (See Reference 7a), thusthe full c will be used to determine margin.
(e) Neuiron Fluence value used for all material is the highest value from Table 5-4 for 32 EFPY.
16
PRESSURE TEMPERATURE LIMS REPORr 6.0 References
- 1. WCAP-8513, Tennessee Valley Authority Sequoyah UnitNo. 2 Reactor Vessel Radiation Surveillance Program,J. A. Davidson, et. aL, November, 1975.
- 2. Code of Federal Regulations, 10CFR50, Appendix H, Reactor Vessel MaterialSurveillance ProgramRequirements, U.S. Nuclear Regulatory Commission, Washington, D.C.
- 3. ASTM E23 StandardTest Method NotchedBar Impact Testing ofMetallic Materials,in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.
- 4. Section X1 of the ASME Boiler and Pressure Vessel Code, Appendix G, FractureToughness Criteriafor ProtectionAgainstFailure
- 5. ASTM E185-82, Annual Book ofASTM Standards, Section 12, Volume 12.02, StandardPractice for Conducting Surveillance Testsfor Light-Water Cooled Nuclear Power Reactor Vessels.
- 6. Regulatory Guide 1.99, Revision 2, RadiationEmbrittlement ofReactor Vessel Materials,U.S.
Nuclear Regulatory Commission, May 1988.
7a. WCAP-15320, Analysis of Capsule Y From the Tennessee Valley Authority Sequoyah Unit 2 Reactor Vessel Radiation SurveillanceProgram, T.J. Laubham, et. al, Dated November 1999.
7b. WCAP-10509, Analysis of Capsule TFrom the Tennessee ValleyAuthority Sequoyah Unit 2 Reactor Vessel Radiation SurveillanceProgram,R. S. Boggs, et a, Dated April 1984.
7c. Southwest Research Institute Nondestructive Evaluation Science and Technology Division, "Reactor Vessel Material Suveillance Program and Technology Division, "Reactor Vessel Material Surveillance Program for Sequoyah Unit 2: Analysis of Capsule U", Final Report SwRI Project 17-8851 TVA Contact 85PJH-964430, January 1990.
7d. WCAP-13545, Analysis of CapsuleXFrom the Tennessee Valley Authority Sequoyah Unit 2 Reactor Vessel RadiationSurveillanceProgram, M. A. Ranirez, S. L. Anderson, A. Madeyski, Dated November 1992.
- 8. CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATI Consulting, March 1999.
- 9. WCAP-14040-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.
- 10. WCAP-15321, Revision 2, "Sequoyah Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation", J.H. Ledger, et.al, July 2003.
17
PRESSURE TEMPERU LIMS REPORT
- 11. ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1", Febny 26, 1999.
- 12. WCAP-15984-P, Revision 01, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation For Sequoyah Units 1 and 2', W. Bamford, et.al, April 2003.
- 13. Westinghouse Letter to TVA, TVA-93-105, "Cold Overpressure Mitigation System Code Case and Delta-P Calculation", May 19, 1993.
- 14. Calculation SQN-IC-014, Demonstrated Accuracy Calculation for Cold Overpressure Protection System".
18
ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQN)
UNITS 1 AND 2 TOPICAL REPORTS .
WCAP-15293, Revision 2 (Unit 1)
WCAP-15321, Revision 2 (Unit 2)
-Weing houeNnPrnprietaty-Class 3.
WCA 15293 iIJuly 00 ftc-Ason 2 Sequoyah, UniI Keatup and Cooldown Limit'.Curves for'..
I: ormal Opration and, PT,1LRSSport P Documentatio`n'.
O~~Wesnn-Ouse:
Westinghouse Non-Proprietary aass 3 i
WCAP-15293, Revision 2 Sequoyah Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation J. H. Ledger July 2003 Prepared by the Westinghouse Electric Company LIC for the Tennessee Valley Authority ApprovedK'
( J. A. dresham, Manager g/ Engineering and Materials Technology Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-055 0)2003 Westinghouse Electric Company LLC All Rights Reserved
12 PREFACE This report has been technically reviewed and verified by:
T. J. Laubham 7 /11/63 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 PTcurves. It has been determined that WCAP-1 5293 Rev. 0 was impacted by this error. Thus, this revision provides corrected curves from WCAP- 15293 Rev.
O.
Note that only the 60F/hr heatup curves were affected by this error. The 1 0F/hr heatup and all cooldown curves were not affected by the computer error and thus remain valid.
Revision 2:
This report was revised to incorporate comments regarding the use of full cq in calculating the margin term for the ART values on Surveillance Materials. Notations were made in the Introduction as well as in Tables 8 and 9. In addition, the referenced WCAP for flange requirement removal was changed from WCAP-15315 toWCAP-15984-P.
iii TABLE OF CONTENTS LIST OF TABLES ........ iv LIST OF FIGURES v EXECUTIVE
SUMMARY
.................. vi 1 INTRODUCTION ................................................ ;. 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 REFERENCES .30 APPENDIXA: 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 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 .......................... iE-0 APPENDIX F: UPDATED SURVEILLANCE CAPSULE REMOVAL SCHEDULE ........................... F-0 APPENDIX G: ENABLE TEMPERATURE CALCULATIONS AND RESULTS ................................. G-0
iv LIST OF TABLES Table 1 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 Betline Material Chenstry Factors ....5 Table 4 Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base Metal Interface (1019.nkm , E > 1.0 MeV) . 1 Table 5 Summary of the Vessel Surface, 114T 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 Heatup and Cooldown Curves.1 Table 7 Integrated Neutron Exposure of the Sequoyah Unit 1 Surveillance Capsules Tested To Date ............................................. 12 Table 8 Calculation of the ART Values for the /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 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 104F 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 6 0 °F/hr) Applicable for the First 32 EFPY (Without Margins for Instrumentation Errors) ........................................... 16 Figure 2 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 1000F/hr) Applicable for the First 32 EFPY (Without Margins for Instrumentation Errors)..................................................................17 Figure 3 Sequoyah Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to I00°F/hr) Applicable for the First 32 EFPY (Without Margins for Instrumentation Errors) ............................... 18 Figure 4 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 6 0°F/br) Applicable for the First 32 EFPY (With Margins for Instrumentation Errors of 10F and 60psig) ....................................... 19 Figure 5 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 100 0 F/hr) Applicable for the First 32 EFPY (With Margins for Instrumentation Errors of 10F and 60psig) ....................................... 20 Figure 6 Sequoyah Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 32 EFPY (With Margins for Instrumentation Errors of 100 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 1 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-15224M and the latest Pressure-Temperature (P-T) Limit Curves from WCAP-12970' 1 3b. The Sequoyah Unit 1 heatup and cooldown pressure-temperature limit curves have been updated based on the use of the ASME Code Case N-64013 1, which allows the use of the Kk4 methodology, and a justification to lower the reactor vessel flange temperature requirement (Reference WCAP-15984-Prlh]).
1 1 INTRODUCTION Heatup and cooldown limit curves are calculated using the adjusted RTNDT (reference nil-ductility temperature) corresponding to the limiting beitline region material of the reactor vessel. The adjusted RTNDT of the limiting material in the core region ofthe reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ARTNwD, and adding a margin. The unirradiated RT 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 600 F.
RTNDT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTNDT 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 RTNDT (RTNur). The extent of the shift in RTNm 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."t111 Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTNIT + ARTNDT + margins for uncertainties) at the 14T 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 2, "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 K1 . critical stress intensities are used in place of the K18 critical stress intensities. This methodology is taken from approved ASME Code Case N-640131. 3) The reactor vessel flange tenperature requirement has been reduced.
Justification has been provided in WCAP-15984-P 1 8 ). 4) The 1996 Version of Appendix G to Section XI will be used rather than the 1989 version. 5) The full q&was used to calculate the margin term for the surveillance materials (base and weld metals) since they were deemed not-credible (See Reference 7). Per NRCprocedural guidance from the November 12, 1997 and February l2!/13'h, 1998 meetings, the full ct should be used when surveillance data is not credible and the Reg Guide Table Chemistry Factor (CF) is non-conservative.
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 Plan 51. The beltline material properties of the Sequoyah Unit 1reactor 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.
3 TABLE 1 Reactor Vessel Beltline Material Uniiradiated Toughness Properties Material Description Cu (%/,) Ni(/o) Initial RTNDTt Intermediate Shell Forging 05 0.15 0.86 400F (Heat # 980807/281489)
Lower Shell Forging 04 0.13 0.76 730F (Heat # 980919281587)
Surveillance Weld (Heat # 25295)N 4 e 0.387 0.11 ---
Rotterdam Test4 ) 0.30 --- ---
Rotterdam Test(% ) 0.25 Rotterdam Test(` e 0.46 ---
Best Estimate of the Inter. to Lower Shell Forging 0.35 0.11 -400F Circ. Weld Seam W05 (Heat # 25295)(de)
Notes:
(a) The Initial RTNDT 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-1 0340, Rev. ), 0.406 (WCAP-1 0340, Rev. l), 0.33 (WCAP-8233) copper and 0.084 (WCAP-10340, Rev.1), 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. 1l6) 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 ofthe intennediate 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 Seam.
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 ARTmT 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 analysism 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.
4 TABLE 2 Calculation of Chemistry Factors using Sequoyah Unit I Surveillance Capsule Data Material Capsule Capsule f° FF° ATNDTc FF*ARTNDT FF2 Lower Shell T 2.61E+18 0.63 67.52°F 42.54°F 0.40 Forging 04 U 7.96E+18 0.94 109.7 0F 103.12°F 0.88 (Tangential) X 1.32E+19 1.08 145.12°F 156.73°F 1.16 (Heat Y 2.19E+19 1.21 129.87 0F 157.14°F 1.47 980919/281587)
Lower Shell T 2.61E+18 0.63 50.59°F 31.87°F 0.40 Forging 04 U 7.96E+18 0.94 67.59°F 63.53°F 0.88 (Axial) X 1.32E+19 1.08 103.34°F 111.61°F 1.16 Y 2.19E+19 1.21 133.35°F 161.35°F 1.47 (Heat # 980919/ SUM: 827.89°F 7.82 281587) CFo4 = XFF
- RTNwT) + 7- FF2) = (827.89) * (7.82) = 105.9*F Surveillance Weld T 2.61E+18 0.63 115.0°F 72.50F 0.40 Material(4 U 7.96E+18 0.94 130.4°F 122.60F 0.88 X 1.32E+19 1.08 143.1 0F 154.5 0F 1.16 (Heat # 25295)(0 Y 2.19E+19 1.21 147.4°F 178.4 0F 1.47 SUM: 528.0°F 3.91 CF s.wem = £F* RTNDT) - £(FF2 ) = (528.0 0 F)+ (3.91) = 135.0°F Notes:
(a) f = Calculated fluence from capsule Y dosimetry analysis results ', (x 1019 n/cr2, E > 1.0 MeV).
(b) FF = fluence factor =f0.21-.l1og .
(c) ARTmT 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 ARTND values have been adjusted by a ratio factor of 0.90.
(e) Surveillance Weld was fabricated from weld wire type SMIT 40, Heat # 25295, Flux Type SMIT 89, Lot #
1103.
5 TABLE 3 Summary of the Sequoyah Unit 1 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 (Heat # 980807/281489)
Lower Shell Forging 04 950 F 105.90 F (Heat # 980919/281587)
Circumferential Weld W05 161.3 0F 135.0F (Heat # 25295)
Surveillance Weld Metal 178.7 0F ---
(Heat # 25295)
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, 4 for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K.1 , for the metal temperature at that time. K10 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 a 4] of the ASME Appendix G to Section XI. The K 0 curve is given by the following equation:
K10 =33.2+20.734*et(.2(T-T= (1)
- where, K1 = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT 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, + Kf<tKc (2)
- where, Khn = stress intensity factor caused by membrane (pressure) stress Kt = stress intensity factor caused by the thermal gradients Kir = function of temperature relative to the RTNDT 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:
Kim = M. x (pRidt) (3) where, Mm for an inside surface flaw is given by:
Mm = 1.85for ft <2, Mm = 0.9264t for 2< t< 3.464, Mm = 3.21 for i > 3.464 Similarly, Mm for an outside surface flaw is given by:
Mm = 1.77for li <2, Mm = 0.8934t for 2< < 3.464, Mm = 3.09for i- >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:
KR = Mb
- Maximum Stress, where Mb is two-thirds of Mm The maximum K1 produced by radial thermal gradient for the postulated inside surface defect of G-2120 is 4 = 0.953x10- x CR x e, where CR is the cooldown rate in OF/hr., or for a postulated outside surface defect, Kt = 0.753xl O x HU x t5, where HU is the heatup rate in 0F/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 KY.
(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)(1) and (2).
(b) Altematively, the K for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a l/4-thickness inside surface defect using the relationship:
Kt = (1.0359Co + 0.6322C1 + 0.4753C2+ 03855C3) *4= (4)
8 or similarly, Kr during heatup for a l4-thickness outside surface defect using the relationship:
i = (1.043Co + 0.630C, + 0.48 1C2 + 0.401C3) * (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+ Ci(x a) + C2(x/a)2 + C3(xl 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"' 21 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, Kk is determined by the metal temperature at the tip of a postulated flaw at the I/4T and 3/4T location, the appropriate value for RTNDT, 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, KIt, 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 KI, at the /4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in KI, exceeds ,, the calculated allowable pressure during cooldown will be greater than the steady-state value.
9 The above procedures are needed because there is no direct control on temperature at the 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 /4T crack during heatup is lower than the Ku for the 1/4T 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 /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 RTNDT by at least 120 0F 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-15984-NP, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation For Sequoyah Units I and 2T'[18], this requirement is no longer necessary when using the methodology of Code Case N-640 31. Hence, Sequoyah Unit 1 heatup and cooldown limit curves will be generated without flange requirements included.
10 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the belthlne region is given by the following expression:
ART = Initial RTmyr + ARTNDT + Margin (7)
Initial RTmT is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel CodelS'. If measured values of initial RTm'T 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.
ARTNm is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:
ARTWDT= CF
- f8-O.logJ (8)
To calculate ARTNDT 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(dx) = fr0ace
- 24x(9) where x inches (vessel beltlhne 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 ARTNw 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"tl. Table 4 contains the calculated vessel surface fluences values at various azimuthal locations and Tables 5 and 6 contains the 1/4T 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 1 reactor vessel. Additionally, the surveillance capsule fluence values are presented in Table 7.
11 TABLE 4 Neutron Fluence Projections at Key Azimutal Locations on the Reactor Vessel Clad[Base Metal Interface (1019 n'cm2 , E > 1.0 MeV)
Azimuthal Locaton EFPY 00 150 300 450 10.03 0.205 0.321 0.409 0.637 20 0.387 0.596 0.761 1.18 32 0.605 0.928 1.19 1.84 48 0.896 1.37 1.75 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 Surface V. T(a) l T(a)
Intermediate Shell Forging 05 1.84 x 1019 1.11 x 1019 4.02 x 10 (Heat# 980807/281489)
Lower Shell Forging 04 1.84 x 109- 1.11 x lo," 4.02 x 1018 (Heat # 980919t281587)
Circumferential Weld Seam W05 1.84 x 1018 1.11 x 109 4.02 x 10'7 (Heat # 25295)
Note.
24 (a) 14T and 314T = Fk) *e( "'),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
.32 1.029 0.747 WCAP-15293
12 TABLE 7 Integrated Neutron Exposure of the Sequoyah Unit 1 Surveillance Capsules Tested To Date Capsule Fluence T 2.61 x 10kg n/cm2 , (E > 1.0 MeV)
U 7.96 x 1018 ncm2 , (E > 1.0 MeV)
X 1.32 x 10'9 n/cm2 , (E > 1.0 MeV)
Y 2.19 x i0'9 n/cm2 , (E > 1.0MeV)
Margin is calculated as, M = 2 CF1i + as The standard deviation for the initial RTNDT margin term, is oi 0 0F 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, cA, is 17'F for plates or forgings, and 8.5 0F for plates or forgings when surveillance data is used. For welds, 0A is equal to 28 0F when surveillance capsule data is not used, and is 14'F (half the value) when credible surveillance capsule data is used. tA need not exceed 0.5 times the mean value of ARTNDT.
Based on the surveillance program credibility evaluation presented in Appendix D to WCAP-1 5224, 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.
13 TABLE 8 Calculation of the ART Values for the 1/4T Location ( 32 EFPY(6)
Material RG 1.99 CF FF IRTNDT(1) &RTN°) Margind ARTM R2 Method (0F) (OF) (OF) (OF) (0F)
Intermediate Shell Forging 05 Position 1.1 115.6 1.029 40 119.0 34 193 Lower Shell Forging 04 Position 1.1 95 1.029 73 97.8 34(- 205 (Heat# 980919/281587) Position 2.1 105.9 1.029 73 109.0 34 216 Intermediate to Lower Shell Position 1.1 161.3 1.029 -40 166.0 56 182 Circumferential Weld Seam Position 2.1 135.0 1.029 -40 1389 56(s 155 Notes.
(1) Initial RTNDr values measured values.
(2) ART = Initial RTNDT + RTNDr + Margin (0F)
(3) ARTNW = CF
- FF (4) M=2*(qj + TA2)m (5) Data Deemed not-credible (See Reference 7), thus the full yAwill be used to determine margin.
(6) Neutron Fluence values used for all material is the highest value from Table 4 for 32 EFPY.
TABLE 9 Calculation of the ART Values for the 3/4T Location @ 32 EFPY(6)
Material RG 1.99 CF FF IRTNDT) ARTNDTM Margin ART( 2 R2 Method (0F) (OF) (OF) (IF) (OF)
Intermediate Shell Forging 05 Position 1.1 115.6 0.747 40 86.4 34 160 Lower Shell Forging 04 Position 1.1 95 0.747 73 71.0 34(s 178 (Heat# 980919/281587) Position 2.1 105.9 0.747 73 79.1 34 186 Intermediate to Lower Shell Position 1.1 161.3 0.747 -40 120.5 56 137 Circumferential Weld Seam Position 2.1 135.0 0.747 -40 100.8 56() 117 Notes:
(1) Initial RTNDy values measured values.
(2) ART = Initial RTNDT + ARTmT + Margin (OF)
(3) ARTNDT = CF
- FF (4) M=2*(ai2 + a,&2)m (5) Data Deemed not-credible (See Reference 7), thus the full q,& will be used to determine margin.
(6) Neutron Fluence values used for all material is the highest value from Table 4 for 32 EFPY.
14 The lower shell forging 04 is the limiting beltline material for the I/4T and 3/4T case (See Tables 8 and 9).
Contained in Table 10 is a summary ofthe limiting ARTs to be used in the generation ofthe Sequoyah Unit 1reactor vessel heatup and cooldown curves.
TABLE 10 Summary of the. Limiting ART Values Used in the Generation of the Sequoyah Unit 1 Heatup/Cooldown Curves Material RG 1.99 R2 32 EFPY 32 EFPY Method 1/4 ART 3/4 ART (OF) (OF)
Intennediate Shell Forging 05 Position 1.1 193 160 Lower Shell Forging 04 Position 1.1 205 178 Position 2.1 216 186 Intermediate to Lower Shell Position 1.1 182 137 Circumferential Weld Seam Position 2.1 155 117 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 beitline 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 0 F and 60 psig) and without margins for possible instrumentation errors using heatup rates of 60 and 100 0 F/hr applicable for the first 32 EFPY. Figures 3 and 6 present the cooldown curves with (10 0F and 60 psig) and without margins for possible instrumentation errors using cooldown rates of 0, 20, 40,60 and 100*F/br 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-640r3 (approved in February 1999) as follows:
1.5 Kim< K10
- where, Ki is the stress intensity factor covered by membrane (pressure) stress, Kle = 33.2 + 20.734 e[0.2 C(-RTNWr)]
T is the minimum permissible metal temperature, and RTNDT 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 1 through 6 define all of the above limits for ensuring prevention of nonductile failure for the Sequoyah Unit 1 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 and 14.
16 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 32 EFPY: 1/4T, 216°F 3/4T, 186°F 2500 [
2250 ...... LeakTestLlmit / . J.
2000 ~~Unacceptable.ccepabler 2 0 00 . .... . Operation l .... ........... .................. ... .................. ................ Se. rtonJ Critical Lim t o Hea tup Rate I /rSiia l 60 Dog. Ftr - _L2Ž_ 1r60 Dg LF/FHr 1 0 . ...... .. ........ ..
750-- - -
,< / ~~~~~~~CriticalityLm It based on Ins a rvice hyd rostat c ts 500 ...... +11~ - --------- tam praturc 277 F) for the c ---
servic priod 250 0 so 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
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)
17 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 32 EFPY: 1/4T, 2160 F 3/4T, 1860 F 2500 2250 2000 1750 a
-1500 2
2 1250 I-S
- w1000 10 750 500 0
0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 2 Sequoyah Unit I Reactor Coolant System Heatup Limitations (Heatup Rate of 100 0F/hr)
Applicable for the First 32 EFPY (Without Margins for Instrumentation Errors)
18 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 32 EFPY: 1/4T, 216*F 3/4T, 186 0 F W.-
.:W 4A-Figure 3 Sequoyah Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 32 EFPY (Without Margins for Instrumentation Errors)
19 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 32 EFPY: 114T, 216 0 F 3/4T, 1860 F 2500 in a
0 C
'4 Lu n
'U U
0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Dog. F)
Figure 4 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 600 F/hr)
Applicable for the First 32 EFPY (With Margins for Instrumentation Errors of 10F and 60 psig)
20 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 32 EFPY: 1/4T, 216°F 3/4T, 186°F 2500-- -------
_ Oprli Vion51Run1680 l / -
2250~~~~~~~~~~~~~~~~
2250--~~~~~~~~~Hrh~eaTest La
.. _ .......... ...~..... i. l............... ................ l_
.1......
... I.
2000 .. Unacceptable ......... .......... ............. Acceptable Operation Operation 1750 _ _
He-tup Rate Critical Limit 100 Dog. FlHr 100 Deg. FIHrl
&1500 -......... ..
L. 12 50 --- ----------
7 5 0 . ................ ................. ....... ........... ....... . ..... . ............ ......... .. . ................ ..... ..... ................. ................
/ 1 l gritice y mit based onI
/t l Z~~Inservice hydrostatic tstl
, z{ i ~~temp-rature 288 F) for thel 500 .......... .................. ........ ................. .............. service period up to 32 EFPY .............
250 - ------------------- ................................ . ........ ...... ......... ......... ........
0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure S Sequoyah Unit 1 Reactor Coolant System Heatup Limitations (~eatup Rate of 100°Flhr)
Applicable for the First 32 EFPY (With Margins for Instrumentation Errors of 10°F and 60 psig)
21 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 32 EFPY: 1/4T, 216 0 F 3/4T, 186 0F 4..~~A. ~~ ~ ~': ', A 4 -%.
7"~~~~~~~~~~~~~~~U Figure 6 Sequoyah Unit 1 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 0F and 60 psig)
22 TABLE 11 32 EFPY Heatup Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors)
Heatup Curves 60 Heatup 60 Limit 100 Heatup 100 Limit Leak Test Limit Trtical Critcal T P T P T P. T P T P 50 0 277 0 50 0 277 0 260 2000 50 585 277 585 50 537 277 537 277 2485 55 585 277 585 55 537 277 537 60 585 277 585 60 537 277 537 65 585 277 585 65 537 277 537 95 585 277 585 95 537 277 537 100 585 277 585 100 537 277 537 105 585 277 585 105 537 277 537 110 585 277 585 110 537 277 537 115 585 277 586 115 537 277 537 120 586 277 588 120 537 277 537 125 588 277 591 125 537 277 537 130 591 277 596 130 537 277 537 135 596 277 602 135 537 277 537 140 602 277 609 140 537 277 538 145 609 277 617 145 538 277 540 150 617 277 626 150 540 277 543 155 626 277 637 155 543 277 547 160 637 277 649 160 547 277 553 165 649 277 663 165 553 277 560 170 663 277 678 170 560 277 568 175 678 277 695 175 568 277 577 180 695 277 714 180 577 277 588 185 714 277 735 185 588 277 601 190 735 277 758 190 601 277 615 195 758 277 784 195 615 277 631 200 784 277 812 200 631 277 649 205 812 277 844 205 649 277 669 210 844 277 879 210 669 277 691 215 879 277 918 215 691 277 716 220 918 277 960 220 716 277 744 225 960 277 1008 225 744 277 774 230 1008 277 1060 230 774 277 808 235 1060 280 1117 235 808 280 846 240 1117 285 1181 240 846 285 888 245 1181 290 1251 245 888 290 934 250 1251 295 1329 250 934 295 985 WCAP-15293
23 TABLE 11 - (Continued) 32 EFPY Heatup Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors) eatup Curves _
60 Heatup 60 Limit 100 Heatup 100 Cdtical Limit Critical T P T P T P T P 255 1329 300 1414 255 985 300 1041 260 1414 305 1509 260 1041 305 1104 265 1509 310 1613 265 1104 310 1172 270 1613 315 1728 270 1172 315 1248 275 1728 320 1838 275 1248 320 1332 280 1838 325 1946 280 1332 325 1424 285 1946 330 2066 285 1424 330 1526 290 2066 335 2199 290 1526 335 1638 295 2199 340 2345 -295 1638 340 1762 300 2345 300 1762 345 1898 305 1898 350 2048 310 2048 355 2214 315 2214 360 2397 320 2397 WCAP-15293
24 TABLE 12 32 EFPY Cooldown Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors)
Cooldown Curves Steady State 20F 40F 60F lOOF T P T P T P T P T P 50 0 50 0 50 0 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 80 578 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 WCAP-15293
(
25 TABLE 12- (Continued) 32 EFPY Cooldown Curve Data Points Using 1996 App. G (without Uncertainties for histrumentation Errors)
Cooldown Curves -
Steady State 20F 40F 60F lOOF T P T P T P 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 220 995 220 985 220 981 225 1051 225 1042 225 1038 230 1098 230 1094 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 WCAP-15293
26 TABLE 13 32 EFPY Heatup Curve Data Points Using 1996 App. G (with Uncertainties for Insirumentation Errors of 100F and 60 psig)
Heatup Curves 60 Heatup T 60 Limit 100 ;imit LeatuTet mit Critical Critical T P T P T P T P T P 50 0 288 0 50 0 288 0 272 2000 50 525 288 525 50 477 288 477 288 2485 55 525 288 525 55 477 288 477 60 525 288 525 60 477 288 477 65 525 288 525 65 477 288 477 70 525 288 525 70 477 288 477 75 525 288 525 75 477 288 477 105 525 288 525 105 477 288 477 110 525 288 525 110 477 288 477 115 525 288 526 115 477 288 477 120 525 288 528 120 477 288 477 125 525 288 531 125 477 288 477 130 526 288 536 130 477 288 477 135 528 288 542 135 477 288 477 140 531 288 549 140 477 288 478 145 536 288 557 145 477 288 480 150 542 288 566 150 477 288 483 155 549 288 577 155 478 288 487 160 557 288 589 160 480 288 493 165 566 288 603 165 483 288 500 170 577 288 618 170 487 288 508 175 589 288 635 175 493 288 517 180 603 288 654 180 500 288 528 185 618 288 675 185 508 288 541 190 635 288 698 190 517 288 555 195 654 288 724 195 528 288 571 200 675 288 752 200 541 288 589 205 698 288 784 205 555 288 609 210 724 288 819 210 571 288 631 215 752 288 858 215 589 288 656 220 784 288 900 220 609 288 684 225 819 288 948 225 631 288 714 230 858 288 1000 230 656 288 748 235 900 290 - 1057 235 684 290 786 240 948 295 1121 240 714 295 828 245 1000 300 1191 245 748 300 874 250 1057 305 1269 250 786 e 305 925 WCAP-15293
27
- TABLE 13 - (Continued) 32 EFPY Heatup Curve Data Points Using 1996 App. G (with Uncertainties for Instrumentation Errors of 100 F and 60 psig)
Heatup Curves 60 Heatup 60 Limit 100 Heatup 100 Critical Limit Critical T P T P T P T P 255 1121 310 1354 255 828 310 981 260 1191 315 1449 260 874 315 1044 265 1269 320 1553 265 925 320 1112 270 1354 325 1668 270 981 325 1188 275 1449 330 1778 275 1044 330 1272 280 1553 335 1886 280 1112 335 1364 285 1668 340 2006 285 1188 340 1466 290 1778 345 2139 290 1272 345 1578 295 1886 350 2285 295 1364 350 1702 300 2006 355 2446 300 1466 355 1838 305 2139 305 1578 360 1988 310 2285 310 1702 365 2154 315 2446 315 1838 370 2337 320 1988 325 2154
. ____________ .____________ 330 2337 WCAP-15293
28 TABLE 14 32 EFPY Cooldown Curve Data Points Using 1996 App. G (with Uncerinties for Intrumentation Errors of 10OF and 60 psig)
Cooldown Curves Steady State 20F 40F 60F lOOF T P T P T P T P T P 50 0 50 0 50 0 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 95 321 100 571 100 524 100 476 100 426 100 325 105 575 105 527 105 479 105 430 105 329 110 578 110 531 110 483 110 434 110 333 115 582 115 535 115 487 115 438 115 338 120 586 120 540 120 492 120 443 120. 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 tnmentation Errors of 101F and 60 psig) o wn C urves Cool _ _ _ _ _ _ _ _ _ _ _ _ _ _
Steady State 20F 40F 60F lOOF T P T P T P T P T P 220 874 220 853 220 836 220 821 220 806 225 909 225 892 225 878 225 869 225 865 230 948 230 935 230 925 230 921 235 991 235 982 235 978 240 1038 240 1034 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 WCAP-15293
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", 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 A "Fracture Toughness Criteria for ProtectionAgainst 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 ValleyAuthority Sequoyah Unit 1 Reactor Vessel Radiation Surveillance Program", S.E. Yanichko, et al., February 1984.
- 7. WCAP-1 5224, "Analysis of Capsule Y From the Tennessee Valley Authority Sequoyah Unit 1 Reactor Vessel Radiation Surveillance Program", T.J. Laubham, et. al., Dated June 1999.
- 8. 1989 Section III, Division 1 of the ASME Boiler and Pressure Vessel Code, Paragraph NB-2331, "Material for Vessels."
- 9. CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATI Consulting, March 1999.
- 10. a Code of Federal Regulations, 10 CFR Part 50, Appendix '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.
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 1", 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. TVAletter No. N9667, "TASKN99-017 - Reactor Coolant System Pressure and Temperature Limit Report Development - N2N-048," W. M. Justice, August 20, 1999.
- 18. WCAP-15984-P, Revision 01, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Sequoyah Units 1 and 2", W. Bamford, et. al, April 2003.
- 19. Westinghouse Letter TVA-93-105.
A-O APPENDIX A LTOPS SETPOINTS 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 1 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 1 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 350TF),
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°F 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.
1.2 LTOPS SETPOINT DETERMINATION Westinghouse has developed new LTOPS setpoints for Sequoyah Unit 1, 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.
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 IOCFR50. 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, 10CFR50 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 minimmn of the steady-state 10CFR50 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).
The previous Sequoyah analyses of multiple heat input cases were used to determine the relationship between setpressures and resulting overshoots/undershoots.
A-3 1.2.4 Final Setpoint Selection Appendix G limits described in Section 1.2.1, were conservatively adjusted accounting for the pressure difference (AP) between the wide range pressure transmitter and the reactor vessel limiting beitline 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 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-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 10CFR50, 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 500 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 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 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 undershoots discussed in Sections 1.2.2 and 1.2.3 to ensure that they protect against the RCP # 1 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-1 and Figure A-2, provide protection against Appendix G, and RCP #1 seal limit violations. Note, these setpoints do not address instrumentation uncertainties.
A-4 1.3 ARMING AND ENABLE TEMPERATURES FOI 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 amng temperature for Sequoyah conservatively continues to be 350F.
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 2000 F or at coolant temperatures less than a temperature corresponding to a reactor vessel metal temperature less than RTNDT + 501F, whichever is greater. RTNTDT 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 I Reactor Vessel is 2950 F at 32 EFPY of operation.
A-5 Table A-1 Selected Setpoints, Sequoyah Unit 1 Trcs (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 WCAP-15293
A-6 Sequoyah Unit I LTOPS Setpolnt Selection 2500 000 la 1500 E
e 2DOO I1000 a L U
0 a
& 500 A
v 0 50 100 150 200 250 300 350 400 450 500 Reactor Coolant System Temperature (F)
- Appendix G -. -- AppendixG Urrtw 4 RCPs Runnir
- - - -RCP Seal Uit Maxhiun Allowable SP ..-..-.-
4 PORV#2 Setpoint a PORV#1 Setpoint Figure A-1: Sequoyah Unit 1 LTOPS Setpoint Selection WCAP-15293
A-7 Sequoyah Unit I LTOPS Selected Setpoints 2000 e 1500 a
0, IL E 1000 Ih.l 50
- 500 0 50 100 150 200 250 300 350 400 450 500 Reactor Coolant System Temperature (n
- -- PORV#2 Setpoint 1; PORV#1 Setpoint Figure A Sequoyah Unit 1 LTOPS Selected Setpoints WCAP-15293
B-0 APPENDIX B PRESSURIZED THERMAL SHOCK (PTS) RESULTS WCAP-15293
B-i 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 RTpT , accepted by the NRC, for each reactor vessel beitline material for the EOL fluence of the material. This assessment must specify the basis for the projected value of RTpTs 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 RTyps, 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 belthlne 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-1 and B-2 contain the results of the calculations for each of the beltline region materials in the Sequoyah Unit 1 Reactor Vessel. Per TVA, the EOL is 32 EFPY and the Life Extension EOL is 48 EFPY.
B-2 TABLE B-1 RTprs Calculations for Sequoyah Unit I Beitline Region Materials at 32 EFPY (
Material Fluence FF CF ARTps(- Margin RTNDrTu)(a RTpTsn (nicm, E 1.0 (OF) (0 F) (OF) (OF)
Mev) (OF)
Intermediate Shell Forging 05 1.84 1.167 115.6 134.9 34 40 209 Lower ShellForging 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)
(a) Initial RTNDr values are measured values (b) RTprs = RTNDT(U) + ARTprs + Margin (OF)
(c) ARTprs = CF
- FF (d) Data deemed not-credible (See Reference 7), thus the full qA will be used to determine margin.
(e) Neutron fluence value used for all material is the highest value from Table 4 for 32 EFPY.
TABLE B-2 RTm Calculations for Sequoyah Unit 1 Beltline Region Materials at 48 EFPY (
Material Fluence FF CF ARTrs(c) Margin RTNnm(') RTPrs)
(fcm%2 E>1.0 (0F) (OF) (OF) (OF) (OF)
MeV)
Intermediate ShellForgingO5 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) t L 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) I
'Notes:
(a) Initial RTD values are measured values (b) RTm = RTNDTMu) + ARTpTs + Margin (F)
(c) ARTpr = CF
- F (d) Data deemed not-credible (See Reference 7), thus the full q will be used to determine margin (e) Neutron fluence value used for all material is the highest value from Table 4 for 48 EFPY.
All of the beitline materials in the Sequoyah Unit 1 reactor vessel are below the screening criteria values of 2700 F and 3000 F at 32 and 48 EFPY WCAP-15293
c-0 APPENDIX C CALCULATED FLUENCE DATA WCAP-15293
c-l The best estimate exposure of the Sequoyah Unit 1 reactor vessel presented in WCAP-15224M 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 1 21.
Combining this measurement data base with the plant-specific calculations, the best estimate vessel exposure is obtained from the following relationship:
= K Icrc where:
4XWEt. = 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 Ocsac. = 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 O(E > 1.0 MeV),
4I(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.
C-2 Table C-1 Azimuthal Variations Of The Neutron Exposure Projections On The Reactor Vessel Clad/Base Metal Interface At Core Midplane Calculated 00 150 30° 4501a]
10.03 EFPY E>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 3.87E+18 5.96E+18 7.61E+18 1.18E+19 E>O.1 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 Table C-2 Comparison Of Calculated And Best Estimate Integrated Neutron Exposure Of Sequoyah Unit 1 Surveillance Capsules T, U, X, and Y CAPSULE T Calculated Best Estimate BE/C
.1(E> 1.0 MeV) n/cm 2 ] 2.61E+18 2.89E+18 1.10 O(E > 0. IMeV) n/cm 2] 8.74E+18 9.62E+18 1.10 dpa 4.34E-03 4.80E-03 1.11 CAPSULE U Calculated Best Estimate BE/C
-(E> 1.0 MeV) [n/cm 2] 7.96E+18 9.69E+18 1.22 45(E > 0.IMeV) n/cm2 ] 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>1.0MeV) [n/cm2 ] 1.32E+19 1.50E+19 1.14 4(E>0.1MeV) [n/cm2 ] 4.42E+19 5.09E+19 1.15 dpa 2.20E-02 2.51E-02 1.15 CAPSULE Y Calculated Best Estimate BE/C O(E> 1.0 MeV) [nlcm 2] 2.19E+19 2.43E+19 1.11 4(E>0.lMeV) [n/cm 2] 7.31E+19 8.15E+19 1.12 dpa 3.63E-02 4.05E-02 1.12 AVERAGE BE/C RATIOS BE/C O(E>1.0MeV) [n/cm2 ] 1.14 O(E>0.IMeV) [n/cm 2 ] 1.14 dpa 1.14 WCAP-15293
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 (X 109 n/cm2) (F)(a) O) (b) (%) (a) (%)()
Lower Shell T 0261 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 (Heat# Y 2.19 114.95 129.87 26.5 23 980919/281587)
Lower Shell T 0.261 59.85 50.59 16 0 Forging 4 U 0.796 89.3 67.59 20.5 19 (Axial) X 1.32 102.6 103.34 23 22 (Heat # Y 2.19 114.95 133.35 26.5 19 9809191281587)
Weld Metal T 0.261 111.13 127.79 35 30 (Heat # 25295)m' 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) 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.
(d) Surveillance Weld was fabricated from weld wire type SMIT 40, Heat # 25295, Flux Type SMrr 89, Lot #
1103.
E-0 APPENDIX E
- REACTOR VESSEL BELTLINE MATERIAL PROJECTED END OF LICENSE UPPER SHELF ENERGY VALUES WCAP-15293
E-1 TABLE E-1 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 USE(& Decrease EOL USE (1019 ncm2) (ft-lb) () (ft-lb) hnermediate Shell Forging 05 0.15 1.11 79 24 60 (Heat# 980807t281489)
Lower Shell Forging 04 0.13 1.11 72 22 56 Using S/C Data (Heat # 980919/281587)
Intennediate to Lower Shell 0.35 1.11 113 42 66 Circumferential Weld Seam W05 Using SIC Data (Heat # 25295)
(a) These values were obtained from Reference 12.
F-0 APPENDIX F
-UPDATED SURVEILLANCE CAPSULE REMOVAL SCHEDULE WCAP-15293
- F-1 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 1 reactor vessel. This recommended removal schedule is applicable to 32 EFPY of operation-Table 7-1 Sequoyah Unit 1 Reactor Vessel Surveillance Capsule Withdrawal Schedule Removal Time Fluence Capsule Location Lead Factor() (EFPY(") (n/cm,E>1.O MeV)W T 400 3.39 1.03 2.61 x 1"' (c)
U 1400 3.47 3.00 7.96 x 101" (c)
X 2200 3.47 5.27 1.32 x 109 (c)
Y 3200 3.43 10.03 2.19 x 109 (c,d)
S 40 1.08 Standby (de)
V 1760 1.08 Standby (de)
W 1840 1.08 Standby (de)
Z 3560 1.08 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 2.74 x 1019 (E > 1.0 MeV), the 48 EFPY peak vessel fluence at approximately 44 EFPY,-respectively.
G-O 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 2000 F or at coolant temperatures less than a temperature corresponding to a reactor vessel metal temperature less than RTNDT + 500 F, whichever is greater. RTNDT is the highest adjusted reference temperature (ART) for the limiting beitline 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 1 reactor vessel beltline region at 32 EFPY is 216F.
From the OPERLIM computer code output for the Sequoyah Unit 1 32 EFPY P-T limit curves without margins (Configuration # 1389796830, operlim film File) the maximum AT.,W is:
Cooldown Rate (Steady-State Cooldown):
max (ATd) at I/4T = 00F Heatup Rate of IO00FIt:
max (AT,.w) at 1/4T = 28.924 0 F Enable Temperature (ENBT) = RTNDT + 50 + max (ATeta), IF
= (216 + 50 + 28.924) IF
= 294.924IF The minimum required enable temperature for the Sequoyah Unit 1 Reactor Vessel is 295IF at 32 EFPY of operation.
Wes,:nhouse No preayCass s WCA-15321 . July2003
- - 2 Revii . ,.
Sequoya'h Uit2 Heatup' and Cooown Limit 'Cures for
... ,Normal:: Operation and PTLR Support Docu mentation . . .. . . ... .
WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15321, Revision 2 Sequoyah Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation J. H. Ledger July 2003 Prepared by the Westinghouse Electric Company LLC for the Tennessee Valley Authority Approv4
//1. shamManager t/Engineering and Materials Teclmology Wastingbowe Electric Company LLC Eneay Systems P.O. Box 355 Pittsburgh. PA 15230-0355
°2003 Westinghouse Electric Company LLC All Rights Reserved
Ii PREFACE This repot has been tcnically reviewed and verifed by:
T J. Laubbarn __ __ __ _ __ __ _ __ __ _
Revision 1:
An error was detected in the OPERLM Computer Program that Westinghouse uses to generate pressure-temperature (PT) limit curves. This error potentially cffects the heatup curves when the 1996 Appendix G Methodology is used in gerating the PT curves. It has been determined that WCAP-15321 Rev. 0 was impacted by this error Thus, tdus revision provides corrected curves fiom WCAP-1532 1 Rev. 0.
Note that only the 6OFlhrheatup curves were affected by this error. The lOOFfir heatup aind all cooldown curves were not affected by the computer error and thus remain valid.
Revision 2:
This report was rcvised to inorporate commnts regarding the use of fill Ain calculating the margin tenn for the ART values on Surveillance Materials Notations were made in the Itroduction as well as in Tables 8 and 9. In addition, the referenced WCAP for flange requirement removal was Chandged fiom WCAP-15315 toWCAP-15984-P.
iii TABLE OF CONTENTS LIST OF TABLES ................................................... . iv LIST OF FIGURES ................................................... .v EXECUTIVE
SUMMARY
.................................................... vi INTRODUCTION..........................................................................................................................1 2 FRACTURE TOUGHNESS PROPERTIES .2 3 CRITERIA FORALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS . 6 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE .10 S HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES . 15 6 REFERENCES .28 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 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 1 Reactor Vessel Beltline Material Unirradiated Toughness Properties ................................ 3 Table 2 Calculation of Chemistry Factors using Sequoyah 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 Azimuthal Locations on the Reactor Vessel Clad/Base Metal Interface (1019 n/cm2 , E > 1.0 MeV) .11 Table 5 Summary of the Vessel Surface, l/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 Heatup and Cooldown Curves .11 Table 7 Integrated Neutron Exposure of the Sequoyah Unit 2 Surveillance Capsules Tested To Date .12 Table 8 Calculation of the ART Values for the /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 Instmentation Errors of 100F and 60 psig) ........... 25 Table 14 32 EFPY Cooldown Curve Data Points Using 1996 App. G (with Uncertainties for Instrumientation Errors of 100F and 60 psig) .27
v LIST OF FIGURES Figure 1 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 60°Ftbr) 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°Fr) Applicable for the First 32 EFPY (Without Margins for Instrumentation Errors) .............................. 17 Figure 3 Sequoyah Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 1OF/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 10F and 60 psig) ...................................... 19 Figure 5 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100F/br) Applicable for the First 32 EFPY (With Margins for Instrumentation Errors of I 10F and 60 psig) ...................................... 20 Figure 6 Sequoyah Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to OOFhr) Applicable for the First 32 EFPY (With Margins for Instrumentation Errors of IOOP 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 (Capsule Y analysis, WCAP-15321M and the latest Pressure-Temperature (P-T) Limit Curves from WCAP-1297l' 3 1 ). The Sequoyah Unit 2 heatup and cooldown pressure-temperature limit curves have been updated based on the use of the ASME Code Case N640 3 3,which allows the use of the Kk, methodology, and the elimination of the reactor vessel flange temperature requirement (Ref, WCAP-15984-Pt).
1 I INTRODUCTION Heatup and cooldown limit curves are calculated using the adjusted RTDT (reference nil-ductility temperature) corresponding to the limiting beitline region material of the reactor vessel. The adjusted RTm-1 of the limiting material in the core region of the reactor vessel is determined by using the uniradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ARTNDT, and adding a margin. The unirradiated RTNry 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 60 0F.
RTNDT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTNDT 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 RTm (RTN<DT). The extent of the shift in RTNm 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."111 Regulatory Guide 1.99, Revision 2, is used for the calculation ofAdjusted 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 2[21, "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 Ky. critical stress intensities are used in place of the K, critical stress intensities. This methodology is taken from approved ASME Code Case N-640 131 . 3) The reactor vessel flange temperature requirement has been eliminated. Justification has been provided in WCAP-15984-P[61. 4) The 1996 Version of Appendix G to Section XIY4 will be used rather than the 1989 version. 5) The full (A was used to calculate the margin term for the surveillance materials (base and weld metals) since they were deemed not-credible (See Reference 7). Per NRC procedural guidance from the November 12, 1997 and February 12hll3%, 1998 meetings, the full GA should be used when surveillance data is not credible and the Reg Guide Table Chemistry Factor (CF) is non-conservative.
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 Plan 5 ]. The beitline 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.
3 TABLE 1 Reactor Vessel Beltline Material Unirradiated Toughness Properties Material Description Cu (%) Ni(%) Initial RTNDT(a)
Intennediate Shell Forging 05 0.13 0.76 10F (Heat #288757 / 981057)
Lower Shell Forging 04 0.14 0.76 -220F (Heat # 990469 / 293323)
Intermediate to Lower Shell Forging 0.12 0.11 4 0F Circumferential Weld Seam' Surveillance Weld') 0.13 0.11 ---
Notes:
(a) The Initial RTNDT values are measured values (b) Circumferential Weld Seam was fabricated with weld wire type SMIT 89, Heat #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 ART 1m 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 analysisl7 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.
4 TABLE 2 Calculation of Chemistry Factors using Sequoyah Unit 2 Surveillance Capsule Data
_~_ - _
Material Capsule Capsule j<) FF°'b ARTNDTO FF*ARTmT 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 (Heat # Y 2.14E+19 1.207 134.1 161.86 1.457 288757/981057)
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 1.207 892 107.66 1.457 (Heat # 288757/ SUM: 677.77°F 7.556 981057) CF5 = X(F
- RTNDT) + £( FF2) = (677.77) + (7.556) = 89.7°F 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 (442) 43.36 1.113 (Heat # 4278)(") Y 2.14E+19 1.207 80.8 (869) 97.53 1.457 SUM: 293.77°F 3.778 CF sw.Wd = Y(FF
- RTmT) + .( FF2) = (293.77 0F) + (3.778) = 77.8°F NoteS:
(a) f = Calculated flUence from capsule Y dosimetry analysis results °, (x 1019 n/cn?, E > 1.0 MeV).
(b) FF = fuence factor = f2s *-O f)
(c) ARTNDT 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 ARTmT values have been adjusted by a ratio factor of 0.93.
(e) Surveillance Weld was fabricated from weld wire type SMT 89, Heat #4278, Flux Type SMIT 89, Lot #
1211.
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 950F 89.70F (Heat # 288757/981057)
Lower Shell Forging 04 104TF .
(Heat # 990469/293323)
Circumferential Weld W05 630F 77.8 0F (Heat # 4278)
Surveillance Weld Metal 67.90F (Heat # 4278)
6 3 CRITERIA FORALLOWABLE 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, Kb for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KI., for the metal temperature at that time. Kk, 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 &4] of the ASME Appendix G to Section XI. The K1, curve is given by the following equation:
Ki= 33.2+ 20.734* et002(r-RitiI)] (1)
- where, Kle = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This Ku 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* Kim + Kt <cKk (2)
- where, Kh,, = stress intensity factor caused by membrane (pressure) stress K = stress intensity factor caused by the thermal gradients Kc,= function of temperature relative to the R1TNm 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 K% for the postulated defect is:
Kim = Mm (pR/I t) (3) where, Mm for an inside surface flaw is given by:
Mm = 1.85 for 1< 2, Mm = 0.926 ft for 2*J-t3.464, Mm = 3.21 for t > 3.464 Similarly, Mm for an outside surface flaw is given by:
Mm = 1.77 for f < 2, Mm = 0.893Vr for 2SU* 3464, Mm = 3.09 for t > 3.464 and p = internal pressure, Ri = vessel inner radius, and t = vessel wall thickness.
For bending stress, the corresponding K1 for the postulated defect is:
Kb = M;
- 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 Y. = 0.953xlO- 3 x CR x t5, where CR is the cooldown rate in F/hr., or for a postulated outside surface defect, Kt = 0.753x1 0-3 x HU x t2 5, where HU is the heatup rate in F/hr.
The through-wall temperature difference associated with the maximum thermal K1 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 .
(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)(1) and (2).
(b) Alternatively, the Ki for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a %/-thicknessinside surface defect using the relationship:
Kit = (1.0359 Co + 0.6322Ci + 0.4753C2 + 03855C3) *fj / (4)
8 or similarly, KT during heatup for a /4-thickness outside surface defect using the relationship:
Kft = (1.043Co+ 0.63OCi + 0.481C2 + 0.401C3)
- f (5) where the coefficients C0, Cl, C2 and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:
a(x) = Co+ Ci(x/a) + C2(xla)2 + C3(xla) 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"' 21 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, K. is determined by the metal temperature at the tip of a postulated flaw at the /4T and 3/4T location, the appropriate value for RTNDT, 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, Y1 t, 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 K1 , at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in KI, exceeds Kht, the calculated allowable pressure during cooldown will be greater than the steady-state value.
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 K& for the 1/4T crack during heatup is lower than the EKk for the 1/4T 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 1/4T 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 114T 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 RTmT by at least 1201F 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 5984-NP, 'Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation For Sequoyah Units 1 and 2"', this requirement is no longer necessary when using the methodology of Code Case N-640P8 ]. Hence, Sequoyah Unit 2 heatup and cooldown limit curves will be generated without flange requirements included.
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 RTNDT + 4RTNDT + Margin (7)
Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Codel8 ]. If measured values of initial RTNwy for the material in question are not available, generic mean values for that class of material may be used ifthere are sufficient test results to establish amean and standard deviation for the class.
bRTNDT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:
ARTryr CF
- fO.l0.OIg (
To calculate ARTNDT at any depth (e.g., at /4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth.
f(&p) = f *
- e (24X))
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 ARTmT at the specific depth.
The Westinghouse Radiation Engineering and Analysis Group evaluated the vessel fluence projections as a part of WCAP-15320M 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" 21 . Table 4 contains the calculated vessel surface fluences values at various azimuthal locations and Tables 5 and 6 contains the 1/4T 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 2 reactor vessel. Additionally, the surveillance capsule fluence values are presented in Table 7.
11 TABLE 4 Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base Metal Interface (1019 n/cm2 , E > 1.0 MeV)
Azimuthal Location EFPY 00 150 300 450 10.54 0.211 0.336 0.426 0.637 20 0.38 0.60 0.773 1.16 32 0.593 0.934 1.21 1.82 48 0.878 1.38 1.80 2.71 TABLE 5 Summary of the Vessel Surface, 1/4T and 3/4T Fluence Values used for the Generation ofthe 32 EFPY Heatup/Cooldown Curves Material Surface l KA l T(1)
Intermediate Shell Forging 05 1.82 x 1019 1.10 x 1019 3.98 x 10s (Heat # 288757/981057)
Lower Shell Forging 04 1.82 x 101 9 1.10 x 1ol 9
3.98 x 10" (Heat # 990469/293323)
Circumferential Weld Seam 1.82 x 10" 1.10 x 1019 3.98 x 1018 (Heat 4278)
Note:.
(a) 1/4T and 3/4T = F sd.) *e4 2 4"), 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 l EFPY 1/4T FF 3/4T FF 32 1.027 0.745 WCAP-15321
12 TABLE 7 Integrated Neutron Exposure of the Sequoyah Unit 2 Surveillance Capsules Tested To Date Capsule Fluence T 2.61 x 10" n/cm2 , (E > 1.0 MeV)
U 6.92 x 10" n/cm2 (E > 1.0 MeV)
X 1.22 x 1019 n/cm 2 , (E > 1.0 MeV)
Y 2.14 x 10' 9 n/cm2 , (E > 1.0 MeV)
Margin is calculated as, M = 2 47T j. The standard deviation for the initial RTNDT margin term, is ai 0 F when the initial RTNJT is a measured value, and 17 0F when a generic value is available. The 0
standard deviation for the ARTNDT margin term, cTA,is 170F for plates or forgings, and 8.5 0 F for plates or forgings when surveillance data is used. For welds, qA is equal to 28 0F when surveillance capsule data is not used, and is 140F (half the value) when credible surveillance capsule data is used. crA need not exceed 0.5 times the mean value ofARTNDT-Based on the surveillance program credibility evahation 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.
13 TABLE 8 Calculation of the ART Values for the /4T Loaio ~32 ~~~ EFP - -
Material RG 1.99 CF FF IRTDr(') ARTD(3) Margin4 ART R2 Method (0F) (OF) (OF) (CF) (°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(5) 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 RTmT values measured values.
(2) ART = Initial RTmT + ARTNDT + Margin (F)
(3) ARTNIT = CF
- FF (4) M = 2 (o2 + 2):
(5) Data Deemed not-credible (See Reference 7), thus the fall ,&will be used to detennine margin.
(6) Neutron Fluence value used for all material is the highest value from Table 4 for 32 EFPY.
TABLE 9 Calculation of the ART Values for the 314T Location @ 32 EFPY)
Material RG 1.99 CF FF IRTND'S) ARTNDT0) Margi 43 ARTO R2 Method (0F) (OF) (OF) (0F) (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(5) 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 S6(S 110 Notes:
(1) Initial RTNM values measured values.
(2) ART = Initial RTNDT + ARTNDT + Margin (F)
(3) ARTNDT = CF *
(4) M = 2 *(ci 2 + rA2)"2 (5) Data Deemed not-credible (See Reference 7), thus the full e, will be used to determine margin.
(6) Neutron Fhlence value used for all material is the highest value from Table 4 for 32 EFPY.
14 The intermediate shell forging 05 is the limiting beltline 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 Material RG 1.99 R2 32 FPY 32 EFPY Method % ART 3/4 ART (OF) (OF)
Intmediate Shell Forging 05 Position 1.1 142 115 Position 2.1 136 111 Lower Shell Forging 04 Position 1.1 119 90 Intermediate to Lower Shell Position 1.1 117 99 Circumferential Weld Seam Position 2.1 132 110 WCAP-15321
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 (lO0 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 0 F and 60 psig) and without margins for possible instrumentation errors using cooldown rates of 0, 20, 40, 60 and 100 0F/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-64013 1 (approved in February 1999) as follows:
1.5 Kim< K,
- where, Khn is the stress intensity factor covered by membrane (pressure) stress, Kle = 33.2 + 20134 el0o2 (T-RTNDT),
T is the minimum permissible metal temperature, and RTNDT 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 0 F. The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.
Figures 1 through 6 define all of the above limits for ensuring prevention of nonductile failure for the Sequoyah Unit 2reactorvessel. The datapoints used fortheheatup and cooldown pressure-temperature limit curves shown in Figures 1 through 6 are presented in Tables 11 through 14.
16 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING 05 LIMITING ART VALUES AT 32 EFPY: 114T, 142-F 3/4T, 115F 2500 2250 2000 1750 (0
1500 a
10 1250 a
1000 a
0 750 500 250 0
0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Dog. F)
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)
17 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING 05 LIMITING ART VALUES AT 32 EFPY: 1/4T, 142°F 3/4T, 115F 2500 Operlim Verslon:5.1 Run:27850 Test .Leak Lim t l .
2250 . .............. I......... ....... .. .... .............. . ............ ......... .......... .-........
2000 Unacceptable Operation J t. Acceptable Operation l
1750 Heatup Rate
~~~~~~~~~~.....
.............. ... ... ....... .. ........... .......... ................. ....... .... ...f............ .
0 100 Deg F/FI00rD LimitH co Q: 150 3~ 1 Soo 250
. .............. ..... .. i.
M ................ ................. .. . ............. .. ..... ......... ......... ......... -.T.---.--.i.-...-
'5 1 000 i
150 75 750 Crit cality Limit based on inservlce hydrostatic test
................. ......... ............... ................ temperature (203 F) forthe ...... .................-fteprue 23Ffoth service period up to 32 EFPY 250 .Tem p
tu....
............... ........- r ............ ...........-..... .... v r...... 4 .......
-...........rr r ..... .r.... ............ . ....
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 100F/hr) Applicable for the First 32 EFPY (Without Margins for Instrumentation Errors)
18 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING 05 LIMITING ART VALUES AT 32 EFPY: 1/4T, 142°F 3/4T, 115°F 2':wo: - -- -- ~~~~~~~~~~~~~~.. . ... .... . .........................
A~~
~~~~~~~ 4~~~~~~~~44 Fgr 3 e t2 tor C tS m Coo Li (o Rs tol r)Applicabe for te First 32 .F. (WithoutMarginsfr.-Istm i Er ors)
- ~ -DeMdeao Temerawr 2X 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)
19 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING 05 LIMITINGARTVALUESAT 32 EFPY: 1/4T, 142 0 F 3/4T, 115F 2500 2250 2000 1750 5n a
t0 1500 0.
co a- 1250 0-a 4-SW
= 1000 4,
'U U
750 t
500 250 0
0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Dog. F)
Figure 4 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 601F/hr) Applicable for the First 32 EFPY (With Margins for Instrumentation Errors of 10 0F and 60 psig)
20 MATERIAL PROPERTY BASIS LIMrING MATERIAL: INTERMEDIATE SHELL FORGING 05 LIMITING ART VALUES AT 32 EFPY: 1/4T, 142-F 3/4T, 115F 2500 2250 2000 1750 5.
1500 t
FE 1250 la a
.5 a 1000 C
750 I
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 IO00F/hr) Applicable for the First 32 EFFY (With Margins for Instrumentation Errors of 10F and 60 psig)
21 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING 05 LIMITING ART VALUES AT 32 EFPY: 1/4T, 142°F
- 3/4T, 115°F 2500 4 ss ,4 Co 2250 C 2000 -
-1 1750 VXACCEPTARLZ OIIA T I 0K s-0 co 1500 -
a-1 250 -
ACC EPTAB LX OPIRAT I ON 1 000 c o o ED aTir 01 RAT CD 750 -
,) 30 Edi 500 - z a G
aS 0 a Faj Tb L. LI _
Baltup -I 250 - I i _L 1.-
IiI H T.
0*
50 I l0 l50 10 200 250 300 r 350 400 450 500D 0
Moderator Temper a t u r e ( Deg . F Figure 6 Sequoyah Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°Flhr) Applicable for the First 32 EFPY (With Margins for Instrumentation Errors of 10 0F and 60 psig)
22 TABLE 11 32 EFPY Heatup Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors)
Eleatup Curves_
60 Heatup 60 Limit 100 Heatup 100 Limit Leak Test Limit Critical Critical T P T P T P T P T P 50 0 203 0 50 0 203 0 186 2000 50 661 203 667 50 661 203 667 203 2485 55 667 203 674 55 667 203 674 60 674 203 682 60 674 203 682 65 682 203 735 65 682 203 717 70 690 203 738 70 690 203 710 75 700 203 745 75 700 203 707 80 710 203 755 80 706 203 706 85 721 203 768 85 706 203 708 90 734 203 785 90 706 203 713 95 735 203 805 95 706 203 721 100 738 203 828 100 706 203 731 105 745 203 855 105 706 203 745 110 755 203 885 110 706 203 761 115 768 203 920 115 708 203 780 120 785 203 958 120 713 203 803 125 805 203 1002 125 721 203 829 130 828 203 1050 130 731 203 858 135 855 205 1103 135 745 205 892 140 885 210 1163 140 761 210 929 145 920 215 1228 145 780 215 971 150 958 220 1301 150 803 220 1019 155 1002 225 1382 155 829 225 1071 160 1050 230 1471 160 858 230 1129 165 1103 235 1570 165 892 235 1194 170 1163 240 1678 170 929 240 1266 175 1228 245 1799 175 971 245 1346 180 1301 250 1930 180 1019 250 1434 185 1382 255 2049 185 1071 255 1531 190 1471 260 2181 190 1129 260 1639 195 1570 265 2326 195 1194 265 1758 200 1678 200 1266 270 1889 205 1799 205 1346 275 2034 WCAP-15321
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 100 Heatup 100 Critical Limit Critical T P T P T P T P 210 1930 210 1434 280 2194 215 2049 215 1531 285 2371 220 2181 220 1639 225 2326 225 1758 230 1889 235 2034 240 2194
__ __ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ 245 2371 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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 1 OOF T P T P T P T P T P 50 0 50 0 50 0 50 0 50 0 50 661 50 618 50 574 50 530 50 440 55 667 55 624 55 581 55 538 55 449 60 674 60 632 60 589 60 546 60 459 65 682 65 640 65 598 65 556 65 470 70 690 70 649 70 608 70 566 70 483 75 700 75 659 75 619 75 578 75 497 80 710 80 670 80 631 80 591 80 513 85 721 85 683 85 644 85 606 85 530 90 734 90 696 90 659 90 622 90 550 95 748 95 712 95 676 95 640 95 572 100 763 100 728 100 694 100 660 100 596 105 780 105 747 105 714 105 682 105 623 110 799 110 767 110 736 110 707 110 653 115 820 115 790 115 761 115 734 115 686 120 843 120 815 120 789 120 764 120 723 125 869 125 843 125 819 125 798 125 764 130 897 130 874 130 853 130 835 130 809 135 928 135 908 135 891 135 876 135 860 140 962 140 945 140 932 140 922 140 916 145 1000 145 987 145 978 145 973 150 1042 150 1033 150 1028 155 1088 155 1084 160 1140 165 1196 170 1259 175 1328 180 1404 185 1489 190 1582 195 1685 200 1799 205 1925 210 2064 215 2218 220 2388 WCAP-15321
25 TABLE 13 32 EFPY Heatup Curve Data Points Using 1996 App. G (with Uncertainties for Instumentation Errors of 100F and 60 psig)
Heatup Curves 60 Heatup 60 Limit 100 Heatup 1 00 Limit Leak Test Limit Crical J Critical T P T P T P 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 75 622 214 647 105 675 214 695 105 646 214 646 110 678 214 708 110 646 214 648 115 685 214 725 115 646 214 653 120 695 214 745 120 646 214 661 125 708 214 768 125 648 214 671 130 725 214 795 130 653 214 685 135 745 214 825 135 661 214 701 140 768 214 860 140 671 214 720 145 795 214 898 145 685 214 743 150 825 214 942. 150 701 214 769 155 860 214 990 155 720 214 798 160 898 215 1043 160 743 215 832 165 942 220 1103 165 769 220 869 170 990 225 1168 170 798 225 911 175 1043 230 1241 175 832 230 959 180 1103 235 1322 180 869 235 1011 185 1168 240 1411 185 911 240 1069 190 1241 245 1510 190 959 245 1134 195 1322 250 1618 195 1011 250 1206 200 1411 255 1739 200 1069 255 1286 205 1510 260 1870 205 1134 260 1374 210 1618 265 1989 210 1206 265 1471 215 1739 270 2121 215 1286 270 1579 220 1870 275 2266 220 1374 275 1698 WCAP-15321
26 TABLE 13 - (Continued) 32 EFPY Heatup Curve Data Points Using 1996 App. G (with Uncertainties for Inrtrumentation Errors of IOF and 60 psig)
Heatup Curves _
60 Heatup 60 Limit 100 Heatup 100 Critical Limit Critical T P T P T P T P 225 1989 280 2426 225 1471 280 1829 230 2121 230 1579 285 1974 235 2266 235 1698 290 2134 240 2426 240 1829 295 2311 245 1974 250 2134 255 2311 WCAP-15321
27 TABIE 14 32 EFPY Cooldown Curve Data Points Using 1996 App. G (withUncertaines for Instrumentation Errors of 107F and 60 psig)
Cooldown Curves Steady State 1 20F 40F 1 60F lOOF T P T P T T P I T P 50 0 50 0 50 0 50 0 .50 0 50 591 50 552 50 503 50 461 50 366 55 595 55 554 55 508 55 466 55 372 60 601 60 558 60 514 60 470 60 380 65 607 65 564 65 521 65 478 65 389 70 614 70 572 70 529 70 486 70 399 75 622 75 580 75 538 75 496 75 410 80 630 80 589 80 548 80 506 80 423 85 640 85 599 85 559 85 518 85 437 90 650 90 610 90 571 90 531 90 453 95 661 95 623 95 584 95 546 95 470 100 674 100 636 100 599 100 562 .100 490 105 688 105 652 105 616 105 580 105 512 110 703 110 668 110 634 110 600 110 536 115 720 115 687 115 654 115 622 115 563 120 739 120 707 120 676 120 647 120 593 125 760 125 730 125 701 125 674 125 626 130 783 130 755 130 729 130 704 130 663 135 809 135 783 135 759 135 738 135 704 140 837 140 814 140 793 140 775 140 749 145 868 145 848 145 831 145 816 145 800 150 902 150 885 150 872 150 862 150 856 155 940 155 927 155 918 155 913 160 982 160 973 160 968 165 1028 165 1024 170 1080 175 1136 180 1199 185 1268 190 1344 195 1429 200 1522 205 1625 210 1739 215 1865 220 2004 225 2158 230 2328 WCAP-15321
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-14040-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 Cy "Fracture Toughness Criteria for ProtectionAgainst 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-15984-P, Revision 01, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation For Sequoyah Units 1 and 2", W. Bamford, et.al., April 2003.
- 7. WCAP-15320, "Analysis of Capsule Y From the Tennessee Valley Authority Sequoyah Unit 2 Reactor Vessel Radiation Surveillance Program", T.J. Laubharn, et. al., Dated December 1999.
- 8. 1989 Section m, Division I of the ASME Boiler and Pressure Vessel Code, Paragraph NB-2331, "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 Q( "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.
Chicots, dated 11/12/92. File #WOG-108/4-18 (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. TVAletterNumberN9664, "TASKN99-017 -Reactor Coolant System Pressure and Temperature Limit Report Development - N2N-048," W. M. Justice, August 17,1999.
- 17. TVAletterNumberN9667, "TASKN99-017 - Reactor Coolant System Pressure and Temperature Limit Report Development - N2N-048," W. M. Justice, August 20, 1999.
- 18. Westinghouse Letter TVA-93-105.
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 1 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 350'F), 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°F 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.
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 transient 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 pressureltemperature 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, 10CFR50 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 A-1.
The nominal upper limit (based on the minimum of the steady-state OCFR50 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.
A-3 The previous Sequoyah analyses of multiple mass input cases were used to determine the relationship between setpressures and resulting overshootstundershoots.
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 Final Setpoint Selection Appendix G limits described in Section 1.2.1, were conservatively adjusted accounting for the pressure difference (AP) 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 10CFR50, 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 50F 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 0 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 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 WCAP-15321
A-4 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 undershoots discussed in Sections 1.2.2 and 1.2.3 to ensure that they protect against the RCP # 1 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 OCFRS0 (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-1 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'F.
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 2000 F or at coolant temperatures less than a temperature corresponding to a reactor vessel metal temperature less than RTNDT + 50'F, whichever is greater. RT1.= 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 2250 F at 32 EFPY of operation.
A-5 Table A-1 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 WCAP-15321
A-6 Sequoyah Unit 2 LTOPS Setpoint Selection E
2
.5 0
0 t
0 0 50 100 150 200 250 300 350 400 450 500 Reactor Coolant System Temperature (F)
-Appendix G Appendix G Unitw/4 RCPs Runnin
- --- RO Seal Lnit - Maxiun Allow able SP POR0#2-IFnal - PORV#1 Figure A-1 Sequoyah Unit 2 LTOPS Setpoint Selection WCAP-15321
A-7 Sequoyah Unit 2 LTOPS Selected Setpolnts 25-_
200-0 SO 100 150 200 250 300 350 400 450 500 Rseactor Coolant System Temperature (F) l ---- POR#2-nal a PORI#1 Figure A Sequoyah Unit 2 LTOPS Selected Setpoints WCAP-15321
B-O APPENDIX B PRESSURIZED THERMAL SHOCK (PTS) RESULTS WCAP-15321
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 RTm , accepted by the NRC, for each reactor vessel behtline material for the EOL fluence of the material. This assessment must specify the basis for the projected value of RTpTs 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 RTs, or upon request for a change in the expiration date for operation of the facility. (Changes to RTm 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 RTmTr, 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-1 and B-2 contain the results of the calculations for each of the beitline 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
B-2 TABLE B-I RTps Calculations for Sequoyah Unit 2 Beitline Region Materials at 32 EFPY (e)
Material Fluence FF CF ARTpr.° Margin RTNDT° RTp'rsb
. ~~~~~~~~~B/cnM2e
,E1.0 (OF) (OF) (OF) (OF) (OF)
_ _ __ __ _ _ _ _ _ _ _ _ M eV )
IntermediateShellForgingO5 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 S/C Data)
Lower Shell Forging 04 1.82 1.164 104 121.1 34(4) -22 133 Cinmferential WeldMetal 1.82 1.164 63 73.3 56 -4 125 Circumferential Weld Metal 1.82 1.164 77.8 90.6 56(4 -4 143 (Using S/C Data)
Notes:
(a) Initial RTmyr values are measured values (b) RTprs = RTNDTM + ARTpm + Margin (OF)
(c) ARTm = CF
- FF (d) Data deemed not-credible (See Reference 7), thus the fill crA will be used to determine margin.
(e) Neutron Fluence value used for all material is the highest value from Table 4 for 32 EFPY.
TABLE B-2 RTpm Calculations for Sequoyah Unit 2 BeItline Region Materials at 48 EFPY (e)
Material Fluence FF CF ARTprs ) Margin RTNDTcu) RTrSb (nkm/, E>1.0 (OF) (OF) (F) (OF) (OF)
_ __ __ _ _ _ _ _ _ _ _ _ _ M eV)
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)
Lower Shell Forging 04 2.71 1.266 104 131.7 34(d -22 144 Circumferential Weld Metal 2.71 1.266f 63 79.8 56 -4 132 Circumferential Weld Metal 2.71 1266 77.8 98.5 561d -4 151 (Using S/C Data) II Not (a) Initial RTmT values are measured values (b) RTprs = RTNDM + ARTpTs + Margin (IF)
(c) ARTpr = CF
- FF (d) Data deemed not-credible (See Reference 7), thus the full cA will be used to determine margin.
(e) Neutron Fluence value used for all material is the highest value from Table 4 for 48 EFPY.
All of the beltline materials in the Sequoyah Unit 2 reactor vessel are below the screening criteria values of 270°F and 300°F at 32 and 48 EFPY.
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-15320M 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 1.
Combining this measurement data base with the plant-specific calculations, the best estimate vessel exposure is obtained from the following relationship:
OB.,r,, = K Dacd,.
where:
OBestIt. = The best estiInate 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.
'cac = 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 O(E > 1.0 MeV),
O(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.
C-2 Table C-1 Azimuthal Variations Of The Neutron Exposure Projections On The Reactor Vessel Clad/Base Metal Interface At Core Midplane Calculated 00 150 300 450[]
10.54 EFPY E>1.0 MeV 2.11E+18 3.36E+18 4.26E+18 6.37E+18 E>0.I MeV 5.37E+18 8.50E+18 1.11E+19 1.70E+19 dpa 3.43E-03 5.39E-03 6.88E-03 1.03E-02 20 EFPY E>1.0 MeV 3.80E+18 6.00E+18 7.73E+18 1.16E+19 E>0.1 MeV 9.65E+18 1.52E+19 2.01E+19 3.10E+19 dpa 6.16E-03 9.61E-03 1.25E-02 1.88E-02 32 EFPY E>1.0 MeV 5.93E+18 9.34E+18 1.21E+19 1.82E+19 E>0.I MeV 1.51E+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-3 Table C-2 Comparison Of Calculated And Best Estimate Integrated Neutron Exposure Of Sequoyah Unit 2 Surveillance Capsules T, U, X, and Y CAPSULE T Calculated Best Estimate BE/C
<4(E> 1.0 MeV) [n/cm 2 ] 2.61E+18 2.57E+18 0.98 O(E> 0.1 MeV) [n/cm 2 ] 8.74E+18 8.98E+18 1.03 dpa 4.34E-03 4.36E-03 1.01 CAPSULE U Calculated Best Estimate BE/C (Z)(E> 1.0 MeV) [n/cm2 ] 6.92E+18 6.03E+18 0.87 4P(E>0. MeV) [n/cm2 ] 2.31E+19 2.11E+19 0.91 dpa 1.15E-02 1.03E-02 0.90 CAPSULE X Calculated Best Estimate BE/C 4I(E> 1.0 MeV) [n/cm2 ] 1.22E+19 1.04E+19 0.85
<D (E> 0.1 MeV) [nlcm2 ] 4.09E+19 3.63E+19 0.89 dpa 2.03E-02 1.77E-02 0.87 CAPSULE Y Calculated Best Estimate BE/C 1.0 MeV) [n/cm 2] 2.14E+19 _(E> 2.18E+19 1.02 O(E>0.IMeV) [n/cm 2 ] 7.14E+19 7.72E+19 1.08 dpa 3.54E-02 3.70E-02 1.05 AVERAGE BEIC RATIOS BE/C 4P(E > 1.0 MeV) [n/cm2 ] 0.93
<(E> 0.1 MeV) [n/cm2 ] 0.98 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-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 (x 10 n/cm2) (F) (a) (0F) () (%) (a) (%)(c)
Intemiediate Shell T 0.261 60.33 63.65 17 12 Forging 05 U 0.692 85.22 79.31 21 16 (Tangential) X 1.22 100.23 85.7 23 8 (Heat # Y 2.14 114.67 134.12 26 22 288757/981057)
Intermediate Shell T 0.261 60.33 48.73 17 7 Forging 05 U 0.692 8522 66.06 21 9 (Axial) X 1.22 100.23 110.04 23 2 (Heat# Y 2.14 114.67 89.21 26 22 288757/981057) .
Weld Metal T 0.261 43.12 74.56 20 2 U 0.692 60.91 130.38 25 6 (Heat # 4278)m' 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 usmg 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 El 85-82.
(d) Surveillance Weld was fabricated from weld wire type SMIT 89, Heat # 4278, Flux Type SMIT 89, Lot #
1211.
E-0 APPENDIX E REACTOR VESSEL BELTLINE MATERIAL PROJECTED END OF LICENSE UPPER SHELF ENERGY VALUES WCAP-15321
E-1 TABLE EI 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(a) Decrease EOL USE (1019 n/cmn2 )
_________________ (fi-I) (%) (ft-Ib) hitemediate Shell Forging 05 0.13 1.10 93 18.5 76 Using S/C Data _ _
Lower Shell Forging 04 0.14 1.10 100 23 77 Intermediae to Lower Shell 0.12 1.10 102 35 66 Circumferential Weld Seam Using S/C Data Notes:
(a) These values were obtained fom Reference 12.
F-O APPENDIX F UPDATED SURVEILLANCE CAPSULE REMOVAL SCHEDULE WCAP-15321
F-1 The following surveillance capsule removal schedule meets the requirements ofASTM E185-82 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-1 Sequoyah Unit 2 Reactor Vessel Surveillance Capsule Withdrawa Schedule Removal Time Fluence Capsule Location Lead Factor() (EFPY)(4 (n/cm2 ,E>1.0 MeV))
T 400 3.33 1.04 2.61 x 10" (c)
U 1400 3.40 2.93 6.92 x 10"' (c)
X 2200 3.39 5.36 1.22 x 10" (c)
Y 3200 3.35 10.54 2.14 x 10'9 (c,d)
S 40 1.09 Standby (de)
V 1760 1.09 Standby (de)
W 1840 1.09 Standby (de)
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 1019 (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.
0-0 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 2000 F or at coolant temperatures less than a temperature corresponding to a reactor vessel metal temperature less than RTNDT + 500 F, whichever is greater. RTNDT 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 l/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, operim.film File) the maximum AT.,W is:
Cooldown Rate (Steady-State Cooldown):
max (AT,.w) at 1/4T = 0F Heatup Rate of 1000 F/Hr:
max (ATmw) at 1/4T = 28.9240 F Enable Temperature (ENBT) = RTNDT + 50 + max (ATbO, IF
= (142 + 50 + 28.924) IF
= 220.924IF The minimum required enable temperature for the Sequoyah Unit 2 Reactor Vessel is 2250 F at 32 EFPY of operation.