Text

Proposed License Amendment Update PT Curve and LTOP for 55 EFPY
	ML080290135
Person / Time
Site:	Saint Lucie
Issue date:	01/23/2008
From:	Johnston G; Florida Power & Light Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-2007-198
	Download: ML080290135 (86)
	v • d • e

Florida Power & Light Company, 6501 S. Ocean Drive, Jensen Beach, FL 34957 0 January 23, 2008 F=PL L-2007-198 10 CFR 50.90 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 RE: St. Lucie Unit 2 Docket No. 50-3 89 Proposed License Amendment Update PT Curve and LTOP for 55 EFPY Pursuant to 10 CFR 50.90, Florida Power & Light Company (FPL) requests to amend Facility Operating License NPF-16 for St. Lucie Unit 2. FPL proposes to replace the current St. Lucie Unit 2 Technical Specification pressure/temperature (P/T) limit curves with new P/T limit curves applicable to 55 effective full power years (EFPY). The low temperature overpressure protection (LTOP) requirements, which are based on the P/T limits, will also be applicable to 55 EFPY.

Attachment 1 is a description of the proposed changes and the supporting justification including the Determination of No Significant Hazards and Environmental Considerations. Attachment 2 provides marked up copies of the proposed Technical Specification changes. Attachment 3 provides copies of the word processed TS pages. Attachment 4 provides information only copies of the marked up TS Bases pages. Enclosure 1 is WCAP-16817-NP, Rev 2, "St. Lucie Unit 2 RCS Pressure and Temperature Limits and Low Temperature Overpressure Protection Report For 55 Effective Full Power Years."

The proposed amendment has been reviewed in accordance with the FPL QATR requirements.

In accordance with 10 CFR 50.91 (b)(1), a copy of the proposed amendment is being forwarded to the State Designee for the State of Florida.

FPL requests that the proposed license amendment be issued by January 31, 2009, effective immediately on issuance with implementation within 60 days.

Please contact Ken Frehafer at (772) 467-7748 if there are any questions about this submittal.

an FPL Group company

L-2007-198 Page 2 I declare under penalty of perjury that the foregoing is true and correct.

Executed on the ,_ _, day of .WVJ W 2008.

Sincerely, Site Vice Presii St. Lucie Plant GLJ[KWF Attachments cc: Mr. William A. Passetti, Florida Department of Health

St. Lucie Unit 2 L-2007-198 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 1 of 6 Update PT Curve and LTOP for 55 EFPY

1. PURPOSE/SCOPE The current St. Lucie Unit 2 P/T limit curves, also called the heatup and cooldown curves, are applicable for 21.7 effective full power years (EFPY) of cumulative operation.

[Reference 1]. Accordingly, the Technical Specifications require revision prior to Unit 2 reaching 21.7 EFPY, which is projected to occur approximately 14 months after the start of cycle 17, or in February 2009.

This license amendment request (LAR) will replace the current St. Lucie Unit 2 Technical Specification pressure/temperature (P/T) limit curves with new P/T limit curves applicable to 55 EFPY. The low temperature overpressure protection (LTOP) requirements, which are based on the P/T limits, will also be applicable to 55 EFPY. This license amendment request should provide P/T limit curves through the remainder of the current 60 year license.

2. BACKGROUND Pressure/temperature limits are developed to satisfy 10 CFR Part 50 Appendix A, Design Criteria 14 and 31. These design criteria require that the reactor coolant pressure boundary be designed, fabricated, erected and tested in order to have an extremely low probability of abnormal leakage, and of rapid or gross failure. The criteria also require that the reactor coolant pressure boundary be designed with sufficient margin to assure that when stressed the boundary behaves in a non-brittle manner and the probability of rapidly propagating fracture is minimized.

10 CFR 50 Appendix G "Fracture Toughness Requirements" describes the requirements for developing P/T limits and the basis for the limitations. The margins of safety against fracture provided by the P/T limits using the requirements of 10 CFR 50 Appendix G are equivalent to those recommended in ASME Section XI, Appendix G. The NRC indicated in the NRC Regulatory Issue Summary (RIS) 2004-04, dated April 5, 2004, that the ASME Section XI, Appendix G, 1998 Edition with the 2000 Addenda may be used without the need for exemption. This Code version utilizes the K1, curve in place of the Kia curve which maintains margin against brittle fracture while improving the window of operation for heatup and cooldown when operating in the lower temperature ranges. Using the approved KI.

methodology, the new 55 EFPY P/T limit curves are less restrictive (shifted higher and to the left) than the current 21.7 EFPY P/T limit curves that were prepared using the Kia methodology, even with more embrittled material properties associated with the vessel. The new P/T limit curves also account for pressure and temperature instrument uncertainty.

The method to predict the reactor vessel material irradiation damage is provided in Regulatory Guide 1.99 Rev. 2, "Radiation Embrittlement of Reactor Vessel Materials." The period of applicability for the new P/T limit curves is based on projections of fluence using the methods in Regulatory Guide 1.190 [Reference 3] and irradiation embrittlement for the reactor vessel beltline limiting materials.

St. Lucie Unit 2 L-2007-198 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 2 of 6 Update PT Curve and LTOP for 55 EFPY Overpressure protection is provided to keep the reactor coolant system (RCS) pressure below the P/T limits after the initiation of assumed energy-addition and mass-addition transients, while operating at low temperatures, in accordance with the NRC Standard Review Plan 5.2.2.

3. DESCRIPTION OF PROPOSED CHANGE The new proposed P/T limits will be applicable to 55 EFPY and will ensure that all RCS components will be able to withstand the effects of transient loads due to system temperature and pressure changes without their functions or performance being impaired. These loads are introduced by normal load transients, reactor trips, and startup and shutdown Operations.

The LTOP system, provided by the power operated relief valves (PORVs) and also by the shut down cooling relief valves (SDCRVs), ensures RCS over pressurization below certain temperatures would be prevented, thus maintaining reactor coolant pressure boundary integrity. The LTOP analysis yields Limiting Conditions for Operation (LCO) that constitutes LTOP System alignments for the period of applicability.

The new P/T limits and LTOP requirements were prepared in accordance with the ASME Section XI, Appendix G, 1998 Edition with the 2000 Addenda using KIc methodology and existing Limiting Conditions for Operation (LCO) that constitutes LTOP System alignments for the period of applicability.

The proposed changes are as follows:

LCO 3.4.9.1 currently provides the pressure and temperature limits in terms of Figures 3.4-2, 3.4-3, and 3.4-4 for the RCS (except the pressurizer) during heatup, cooldown, criticality, and inservice leak and hydrostatic testing for 21.7 EFPY. Figure 3.4-4 identifies maximum cooldown rates once inside the LTOP enable range for 21.7 EFPY. The new 55 EFPY heatup and cooldown curves, Figures 3.4'2 and 3.4-3, are less restrictive (curves shifted up and to the left) than the previous curves due to the use of the ASME Section XI, Appendix G, 1998 Edition with the 2000 Addenda. The new heatup curve, Figure 3.4-2, includes instrument uncertainty and the inservice hydrotest line. The new cooldown curve, Figure 3.4-3, only has 100°F/hr and 75°F/hr cooldown rates since the lower rates allow higher pressures than are allowed by the lowest service temperature and flange limitation restriction.

As a result, the complexity of 4 different cooldown rates has been reduced to just two and the separate figure, 3.4-4, to identify the maximum allowable cooldown rates is not needed. The maximum cooldown rates that were identified on Figure 3.4-4 for 21.7 EFPY, are now in a box on Figure 3.4-3. Figure 3.4-4 has been eliminated.

St. Lucie Unit 2 L-2007-198 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 3 of 6 Update PT Curve and LTOP for 55 EFPY LCO 3.4.9.3 identifies the requirements for the low temperature overpressure protection (LTOP) system. The PORV lift setpoint has been raised from < 470 psia to < 490 psia. This change will allow increased operating margin.

LCO 3.4.9.3 also identifies the enable temperature ranges for the LTOP system in Table 3.4-3 and the minimum temperature for PORV use for LTOP in Table 3.4-4. The enable temperatures in Table 3.4-3 have been lowered slightly for the 55 EFPY period to correspond to the less restrictive P/T limit curves. The use of the PORVs has been expanded in Table 3.4-4 to allow the PORVs to be used for LTOP during the entire heatup and to a lower temperature during cooldown which will allow greater flexibility with the SDC system.

TS Bases section 3/4.4.9 has been revised to reflect the new 55 EFPY period of applicability and the combining of the P/T limit curves and minimum cooldown figures from 3 figures (Figures 3.4-2; 3.4-3, & 3.4-4) into 2 figures (Figures 3.4-2 & 3.4-3).

The marked up Technical Specification and Bases changes are provided in Attachments 2 and 4.

.4. BASIS AND JUSTIFICATION OF PROPOSED CHANGE The current P/T Limit curves and LTOP analysis will expire at 21.7 EFPY. Westinghouse was contracted to update the P/T Limit curves and LTOP analysis to 55 EFPY based on the current ASME methods and Regulatory requirements. The results are P/T limit curves and LTOP setpoints and enable temperatures that are less restrictive than the current TS curves and requirements. The new curves also incorporate instrument uncertainty margin.

The basis for the proposed change is provided in WCAP-16817-NP [Reference 2], which is a compilation of the calculation notes for the P/T limit curves, LTOP analysis, and supporting calculations and evaluations. The fluence projections were prepared using the guidance of Regulatory Guide 1.190. [Reference 3] The reactor pressure vessel beltline pressure-temperature limits are based upon the irradiation damage prediction methods of Regulatory Guide 1.99 Revision 2. [Reference 4]

5. NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION The Nuclear Regulatory Commission has provided requirements in 10 CFR 50.92 for determining whether a request for amendment involves a no significant hazards consideration. The regulation states that no significant hazards considerations are involved if the operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety.

St. Lucie Unit 2 L-2007-198 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 4 of 6 Update PT Curve and LTOP for 55 EFPY Florida Power & Light (FPL) has reviewed this proposed license amendment request and determined that its adoption would satisfy the requirements of 10 CFR 50.92(c) and, accordingly, a no significant hazards consideration finding is justified. The conclusions of this determination are summarized below:

(1) Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes have been determined in accordance with the methodologies set forth in the regulations to provide an adequate margin of safety to ensure that the reactor vessel will withstand the effects of normal startup and shutdown cyclic loads due to system temperature and pressure changes as well as the loads associated with reactor trips. The regulations of 10 CFR Part 50 Appendix A, Design Criterion 14 and Design Criterion 31 remains satisfied. The pressure-temperature (P/T) limit curves in the Technical Specifications are conservatively generated in accordance with the fracture toughness requirements of the ASME Code Section XI, Appendix G. The margins of safety against fracture provided by the P/T limits using the requirements of 10 CFR 50 Appendix G are equivalent to those recommended in ASME Section XI, Appendix G. The Adjusted Reference Temperature (ART) values are based on the guidance of RG 1.99 [Reference 4].

The proposed changes will not result in physical changes to structures, systems or components SSCs or to event initiators or precursors. Changing the heatup and cooldown curves and the pressure relief setpoints to reflect 55 EFPY does not affect the ability to control the RCS at low temperatures such that the integrity of the reactor coolant pressure boundary would not be compromised by violating the P/T limits.

The proposed changes will not impact assumptions and conditions previously used in the radiological consequence evaluations nor affect mitigation of these consequences due to an accident described in the UFSAR. Also, the proposed changes will not impact a plant system such that previously analyzed SSCs might be more likely to fail.

The initiating conditions and assumptions for accidents described in the UFSAR remain as analyzed.

Thus, based on the above, reasonable assurance is provided that the proposed amendment does not significantly increase the probability or consequences of accidents previously evaluated.

(2) Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

St. Lucie Unit 2 L-2007-198 Docket No. 50-389 Attachment 1 Proposed License Amendment . Page 5 of 6 Update PT Curve and LTOP for 55 EFPY The requirements for P/T limit curves and LTOP have been in place since the beginning of plant operation. The revised curves are based on a later edition of Section XI of the ASME Code that incorporates current industry standards for P/T curves. The revised curves also are based on reactor vessel irradiation damage predictions using RG 1.99 methodology. No new failure modes are identified nor are any SSCs required to be operated outside of their design bases. Consequently, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

(3) Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

The proposed P/T curves continue to maintain the safety margins of 10 CFR 50 Appendix G by defining the limits of operation which prevent nonductile failure of the reactor pressure vessel. Analyses have demonstrated that the fracture toughness requirements are satisfied and that conservative operating restrictions are maintained for the purpose of low temperature overpressure protection. The P/T limit curves provide assurance that the RCS pressure boundary will behave in a ductile manner and that the probability of a rapidly propagating fracture is minimized. Therefore, operation in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

Summary: Based on the above discussion and the analysis performed, FPL has determined that the amendment request does not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the probability of a new and different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety; and therefore does not involve a significant hazards consideration.

6. ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION 10 CFR 51.22(c)(9) provides criteria for identification of licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not:

(i) involve a significant hazards consideration, (ii) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, and

St. Lucie Unit 2 L-2007-198 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 6 of 6 Update PT Curve and LTOP for 55 EFPY (iii) result in a significant increase in individual or cumulative occupational radiation exposure.

FPL has reviewed the proposed amendment and concluded that it involves no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite, and no significant increase in individual or cumulative occupational radiation exposure. The proposed amendment also involves no significant hazards consideration and meets the criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), an environmental impact statement or environmental assessment need not be prepared in connection with issuance of the amendments.

7. REFERENCES

1. St. Lucie Unit 2, Technical Specification through Amendment 148, dated 11-2-07.

2.. WCAP-16817-NP, "St. Lucie Unit 2 RCS Pressure and Temperature Limits and Low Temperature Overpressure Protection Report For 55 Effective Full Power Years,"

Revision 2, Westinghouse Electric Corp, LLC, October 2007.

3. US NRC Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence (ML010890301)," March 2001.

4. US NRC Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials,"

Revision 2, May 1988.

St. Lucie Unit 2 L-2007-198 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 1 of 12 Update PT Curve and LTOP for 55 EFPY Technical Specification Change Mark Ups Page Index XXI Page Index XXII Page 3/4 4-29 Page 3/4 4-30 Page 3/4 4-31a Page 3/4 4-3 1b Page 3/4 4-32 Page 3/4 4-35 Page 3/4 4-37a

St. Lucie Unit 2 L-2007-198 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 2 of 12 Update PT Curve and LTOP for 55 EFPY INDEX LIST OF FIGURES FIGURE PAGE 2.1-1 REACTOR CORE THERMAL MARGIN SAFETY LIMIT LINES FOUR REACTOR COOLANT PUMPS OPERATING ......................................................... 2-3 2.2-1 LOCAL POWER DENSITY- HIGH TRIP SETPOINT PART 1 (FRACTION OF RATED THERMAL POWER VERSUS QR 2) ................................. 2-7 2.2-2 LOCAL POWER DENSITY- HIGH TRIP SETPOINT PART 2 (Q R2 V E R S US Yi) .................................................................................................. 2-8 2.2-3 THERMAL MARGIN/LOW PRESSURE TRIP SETPOINT PART 1 (Y 1, V E RS US A 1).................................................................................................... 2-9 2.2-4 THERMAL MARGIN/LOW PRESSURE TRIP SETPOINT PART 2 (FRACTION OF RATED THERMAL POWER VERSUS QR1) ............................... 2-10 3.1-1 MINIMUM BORIC ACID STORAGE TANK VOLUME AS A FUNCTION OF STORED BORIC ACID CONCENTRATION .............................................. 3/4 1-15

.3.1-la iDELETED - ... .....

3.1-2 DELETED ........... .............................. .. . .......... .......... .... .......

3.2-1 DELETED ..... .................. ........ ..............

3.2-2 DELETED...... ....... --

32-3 DELETED... ...

.. ~ T U UIT'JI RECOR COOLANT 4.2-1 DELETED .... .......... HEAj P CORCRITCAL 7 -ICZ NINE TEST 3,2-4 DELETED ~ ~

34-1 DOSE EQUIVALENT 1-131 PRIMARY COOLAINT SPECIFIC ACTIVITY LLIMITS VERSU S PERC ENT OF RAT ED TH ERMAL POWE RWITHITH ./

PRIMARY COOLANT SPECIFIC ACTIVITY > 1 IIC/G RAM DOSE EQ UIVALENT -131. ............... ............................. ............ ........... 3/4 4 28 314-2, . ECO OLANT GCST-EM PREGCURE TEMPERATURE LIMiT-ATjEONS FrR 21.7 EFPY, .1ICATUP ANe e~e C,,RITICAt-..........3/4 4-31 a ST LO:t'6c~IEi*b i ; i ili: ! i ; !i117!

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8.53,73,92,112,[!`*i~i`;;ii~ii;*

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St. Lucie Unit 2 L-2007-198 Docket No. 50-3 89 Attachment 2 Proposed License Amendment Page 3 of 12 Update PT Curve and LTOP for 55 EFPY 5.6-id REQUIRED FUEL ASSEMBLY BURNUP vs INITIAL ENRICHMENT and DECAY TIME, REGION 1, 1.82 wlo .................................................................... 5-4E 5.6-1e REQUIRED FUEL ASSEMBLY BURNUP vs INITIAL ENRICHMENT, R E G IO N 1,2.82 w/o ............................................................................................ 5-4 F 5.6-1f REQUIRED FUEL ASSEMBLY BURNUP vs INITIAL ENRICHMENT, REGION II CASK PIT STORAGE RACK ............................................................ 5-4G 6 .2-1 DE LET E D............................................................................................................. 6-3 6 .2-2 . DE LE T E D .................................................................................................. ........ 6-4 ST. LUCIE - UNIT 2 XXII Amendment No. 8. 29, 53, 4,4,142, 4-T-7,135

St. Lucie Unit 2 L-2007-198 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 4 of 12 Update PT Curve and LTOP for 55 EFPY

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St. Lucie Unit 2 L-2007-198 Docket No. 50-3 89 Attachment 2 Proposed License Amendment Page 5 of 12 Update PT Curve and LTOP for 55 EFPY

-REACTOR COOLANT SYSTEM SURVE51IANCE REO-UIRMET uCngiied) i 44 9 12 The reactor vessel matenial irradiation surveillance specimens shall be removed anc examined, to determine changes in material propeities as required by 10 CFR Apendix H, The resuts of these examralons shall be .O used tudei ur_

'Tl LUCI -i 2! 3;4

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St. Lucie Unit 2 L-2007-198 Docket No. 50-3 89 Attachment 2 Proposed License Amendment Page 6 of 12 Uvdate PT Curve and LTOP for 55 EFPY

%4 e,-31, A~riencme7,i No. ;, 4~, 112

St. Lucie Unit 2 L-2007-198 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 7 of 12 Update PT Curve and LTOP for 55 EFPY REPLACE FIGURE 34-2 WITH THE FOLLOWING:

FIGURE 3.4-2 ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR 55 EFPY, HEATUP, CORE CRITICAL, AND INSERVICE TEST

!;ii:i*::*iiiiiiiiii 2500 1I Y DR2~I ff'J H STAvT, I 4

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-Includes Instrumnent Uncertainty

St. Lucie Unit 2 L-2007-198 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 8 of 12 TT ý,,X P*T r.,,p rnA T r'UD C-C ;; IQV1*V

St. Lucie Unit 2 L-2007-198 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 9 of 12 S1; IT7 V ILL UP TO 7 5 F/IR F*- --

St. Lucie Unit 2 L-2007-198 Docket No. 50-3 89 Attachment 2 Proposed License Amendment Page 10 of 12 T T-~A~

-DT Wr - A TrlrD C. r 55 EFPY OOLC 100

OR COOLN AXIMUM COOLDOWN

St. Lucie Unit 2 L-2007-198 Docket No. 50-3 89 Attachment 2 Proposed License Amendment Page 11 of 12 Update PT Curve and LTOP fc . I I1U DIV qR0 luulT.

2nteclion device toi OR

'ERPRESý

St. Lucie Unit 2 L-2007-198 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 12 of 12 I T-*Aý- DT 0'* ,, -A T TCD f,- <1<1 DV A"-,- N-11 e, 1-2

St. Lucie Unit 2 L-2007-198 Docket No. 50-3 89 Attachment 3 Proposed License Amendment Page 1 of 10 Update PT Curve and LTOP for 55 EFPY Word-Processed Technical Specification Changes Page Index XXI Page Index XXII Page 3/4 4-29 Page 3/4 4-30 Page 3/4 4-31a Page 3/4 4-3 1b Page 3/4 4-32 Page 3/4 4-35 Page 3/4 4-37a

St. Lucie Unit 2 L-2007-198 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 2 of 10 Update PT Curve and LTOP for 55 EFPY INDEX

ýýT OF FIGURES FIGURE PG 21-1 REACTOR CORE THERMAL MARGIN SAFETY LIMIT LINES FOUR REACTOR COOLANT PUMPS OPERATING .......... .... . ...... 2-3 2.2-1 LOCAL POWER DENSITY - HIGH TRIP SETPOINT*PART. 1 (FRACTION OF, RATED THERMAL POWER VERSUS R). ...... 2-7 2.2-2 LOCAL POWER DENSITY - HIGH TRIP SETPOINT PART 2 (QR 2 VERSUS Yj).... .............-.......... ............................ ................. 2-8 2.2-3 THERMAL MARGINILOW PRESSURE TRIP SETPOINT PART 1 (Y 1,VERSUSA 1 )...................... ................. ........ ... 2-9 2,2-4 THERMAL MARGINILOW PRESSURE TRIP SETPOINT PART 2 (FRACTION OF RATED THERMAL POWER VERSUS OR,) .......... .......... 2-...

3.1-1 MINIMUM BORIC ACID STORAGE TANK VOLUME AS A FUNCTION OF STORED BORIC ACID CONCENTRATION. ..... .................. _ ..... 3141-15 3.1,-1la D E LET E D ............... .................. .............. ......................... ..... ... .... ......... ..........

3.1-2 DELETED . _ ____ . ..... ............... : ._

32-1 DELETED ................ ... ... . ........... ....... .

3 .2-2 D ELE T E D .... .... --- -------------- ------- _ - ---- --_ ---

3.2 -3 D ELET E D .........

4.2-1 DELETED ......... ... .... ........ .....

3,2-4 DELETED __ . 1/2 ----- _-.-

3.4-1 DOSE EQUIVALENT 1-131 PRIMARY COOLANT SPECIFIC ACTIVITY' LIMITS VERSUS PERCENT OF RATED THERMAL POWER WTH THE PRIMARY COOLANT SPECIFIC ACTIVITY >. 1 tCi/GRAM DOSE EQUIVALENT 1-13 1. ý.......... ......... ....... ............. 314 4-28 3.4-2 ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR 55 EFPY, HEATUP, CORE CRITICAL, AND INSERVIC E T .... ..... ...... ....... _ ....314 4-31a ST LUCIE-UNIT2 xiAmnendment Wo 8.ý:Z 92,442

St. Lucie Unit 2 L-2007-198 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 3 of 10 Update PT Curve and LTOP for 55 EFPY FUNCTIONAL TEST RNUP vs INITIAL ENf RNUP vs INITIAL ENf

-Y BURI*

Iwfo..

-GIC UNIT 2

St. Lucie Unit 2 L-2007-198 Docket No. 50-3 89 Attachment 3 Proposed License Amendment Page 4 of 10 DDV ana t ther itions or be e RCS Tý,g ST LUCIE UNIT 2 Arnmn

Amendment I-i, 4C.,

St. Lucie Unit 2 L-2007-198 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 5 of 10 Update PT Curve and LTOP for 55 EFPY REACTOR COOLANT SYSTEM

ýSURVEILLANCE REQUiREMENTS (Contud 4 4.9,1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10 CFR 50 Appendix H.-The results of these examinations shall be used to update Figures 3.4-2 and 3,4-3. I1 ST, LUCIE - UNIT 2 314 4-30 ST..L cIE - UNIT2:3/4 4 ment No.,

If. 31, 0.4,

St. Lucie Unit 2 L-2007-198 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 6 of 10 Undate PT Curve and LTOP for 55 EFPY FIGURE 3.4-2 ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR 55 EFPY, HEATUP, CORE CRITICAL, AND INSERVICE TEST 2500 A~~~fl ~ .34 7 7J J4 34 Ai J.L. L I33i3 3Ii13 l37,;

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St. Lucie Unit 2 L-2007-198 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 7 of 10 Update PT Curve and LTOP for 55 EFPY FIGURE 3.4-3 ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE-TEPERATURE LIMITS FOR 66 EPPY, COOLDOWN AND INSERVICE TEST 2500 SAU(liL UUIS N~S&N ICE HYttIOSTA TCZTEST cL 2000

'U 100 75ooA- ~~~UT00I.Pl IL 0 00 1W00, 20 300 400WIL 500DMRAE Tc NDIATD RACOR OOANTSYTEM Fij TEMPERATURE z Icue Lntumn 7nce<a10t ST LUCIE - UNIT 2 314 4-31b STLCIE-NIT23/4

-3bA~mendmefnt No 37-4,04.42,

St. Lucie Unit 2 L-2007-198 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 8 of 10 Update PT Curve and LTOP for 55 EFPY DEL 3M 4-32, 3/4 4-3 Am-endment

St. Lucie Unit 2 L-2007-198 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 9 of 10 TT-At- DT 0-- .- A I TCWD C-F 1ý1I7DV DIO; StoC vent area

- .*I-:,*.

thin the LOW ible 3,4-3,

St. Lucie Unit 2 L-2007-198 Docket No. 50-3 89 Attachment 3 Proposed License Amendment Page 10 of 10 Update PT Curve and LTOP for 55 EFPY TABLE 3.4-3 LOW TEMPERATURE, RC$ OVERPREýSSURE PROTECTION RANGE Operating Cold Leg Temperature, F0 Period, During During

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MINIMUM COLD LEG TEMPERATURE FOR PORV USE FOR LTOP Oerating T:DId, 1 F' Period During During EFPY Cooldown

<55 80 132 ST. LUCIE - UV4ýT2 ý1ii34 4_ý37ai ~Am~endmenrt 3~44-7 No. 3.1,46, 442,

St. Lucie Unit 2 L-2007-198 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 1 of3 Update PT Curve and LTOP for 55 EFPY TS Bases Changes (Information Only)

St. Lucie Unit 2 L-2007-198 Docket No. 50-3 89 Attachment 4 Proposed License Amendment Page 2 of 3 Update PT Curve and LTOP for 55 EFPY SECTIO NO.: TITLE: TECHNICAL SPECIFICATINPAGE T 4,.4 BASES ATTACHMENT'6 OF ADM2.42 f3 REV(SDN NO.: REACTOR COOLANT BYSE 5 ST. LUCIE UNIT 2 33.4 REACTOR COOLANT SYSTEM (continued)-

BASES (continued) 33.4.9 PRESSURE/TEMPERATURE LIMITS All components inthe Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 5.2 of the FSAR During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldovyn rates are consistentwith the design assumptions and satisfy the stress limits for cyclic operation.

During heatup, the thermal gradients through the reactorvessel wall produce thermal stresses whiich are compressive atthe reactor vessel inside surface and are tensile at the reactorvessel outside surface. Since reactor vessel internal pressure always produces tensile stresses at both the inside and outside surface locations, the total applied stress is greatest at the outside surface location. However, since neutron irradiation damage is larger at the inside surface location when compared to the outside surface, the inside surface flaw may be more limiting. Consequently, for the heatup anallysis both the inside and outside surface flaw locations must be analyzed for the spec ific pressuire, and thefrmal oad ings to determi re wh ich is more lim titi ng.

During cooldawri, the thermal gradients through the reactor vessel wall produce thermal stresses which are tensile at the reactor vessel inside-surface and which are compressive at the reactor vessel outside surface.

Since reactor vessel internal pressure always produces tensile stresses at both the inside and outside surface locations, the total applied stress is greatest at the inside surface location. Since the neutron indication damage is also greatest at the inside surfacelocation the inside surface flaw is the limiting location. Consequently, only the inside surface flaw must be evaluated for the cooldown analysis.

The heatup and cooldown Iimit curves Figures 3.4- 2, 3.4-3 and 3.4-41are composite curves which were prepared by ldetermining the most conservative case, with eitherthe inside or outside wall controlling, for any heatup rate of

<up to 50 degrees F per hour or cooldown rate orfup to 100 degrees F per hour. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at21EF and they include 'adjustments for pressure differences between the reacto vessel beltline and pressurizer instrumnent taps,

,Figures3.4-2 and 2.-I-

St. Lucie Unit 2 .L-2007-198 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 3 of 3 Update PT Curve and LTOP for 55 EFPY ed) als n are st and I

ire

L-2007-198 ENCLOSURE 1 WCAP-16817-NP REV2

Protection Repor fo r 55 Effective, Full Power Ye Q Westinghouse

LEGAL NOTICE This report was prepared as an account of work performed by Westinghouse Electric Company LLC for Florida Power and Light Company. Neither Westinghouse Electric Company LLC, nor any person acting on its behalf:

A. Makes any warranty or representation, express or implied including the warranties of fitness for a particular purpose or merchantability, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this report.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-16817-NP, Rev 2 St. Lucie Unit 2 RCS Pressure and Temperature Limits and Low Temperature Overpressure Protection Report for 55 Effective Full Power Years October 2007 Author: F. P. Ferraraccio Reviewer: S. Maeby Approved: M. Gancarz Electronically approved records are authenticated in the Electronic Document Management System.

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WCAP-16817-NP, Rev 2 Page ii October 2007

RECORD OF REVISIONS Rev Date Revision Description Author Reviewed!

Approved S. Maeby 0 9/2007 Original Issue M. Thibodeau B.R. Ganta M. Gancarz Document Edited for WCAP format corrections S. Maeby 1 9/2007 only. F. Ferraraccio Revision 0 was not released to the customer. M. Gancarz F. Ferraraccio S. Maeby 2 See Sec. 2.2-1, pg 8; Sec. 3.3-1, pg 17; Table 8, pg EDMS 30. M. Gancarz WCAP-16817-NP, Rev 2 Page iii October 2007

TABLE OF CONTENTS SECTION and TITLE PAGE LIST O F TA B LE S .................................................................................................................................. V LIST O F F IG U RE S ................................................................................................................................ v 1 IN TR O D U C TION ...................................................................................................................... 1 2 PRESSURE - TEMPERATURE LIMITS .............................................................................. 3 2.1 ADJUSTED REFERENCE TEMPERATURE PROJECTIONS ........................................... 3 2.2 GENERAL APPROACH FOR CALCULATING PRESSURE - TEMPERATURE LIMITS ... 5 2.2-1 Application of Pressure Correction Factors ........................................................................ 7 2.3 THERMAL ANALYSIS METHODOLOGY .......................................................................... 8 2.4 COOLDOWN LIMIT ANALYSIS ......................................................................................... 9 2.5 HEATUP LIM IT AN ALY SIS .................................................................................................. 9 2.6 HYDROSTATIC TEST AND CORE CRITICAL LIMIT ANALYSIS ................................ 10 2.7 LOWEST SERVICE TEMPERATURE, MINIMUM BOLTUP TEMPERATURE, FLANGE LIMIT TEMPERATURE, AND MINIMUM PRESSURE LIMITS ................... 12 2 .8 D A TA ..................................................................................................................................... 13 3 LOW TEMPERATURE OVERPRESSURE PROTECTION .............................................. 15 3.1 G E NE R A L ................................................................................................................................ 15 3.2 METHOD AND ASSUMPTIONS ....................................................................................... 15 3.3 PRESSURE TRANSIENT ANALYSES .............................................................................. 16 3.3-1 Energy A ddition Transients ................................................................................................ 16 3.3-2 M ass A ddition Transients. ..................................................................................................... 17 3.3-3 C ontrolling Pressures ............................................................................................................. 19 3.4 LIMITING CONDITIONS FOR OPERATION .................................................................... 19 3.5

SUMMARY

OF PROPOSED CHANGES .......................................................................... 20 4 RE FE REN C E S ......................................................................................................................... 21 APPENDIX A - TECHNICAL SPECIFICATION FIGURES ...................................................... 41 WCAP-16817-NP, Rev 2 Page iv October 2007

LIST OF TABLES NO. TITLE PAGE 1 St. Lucie Unit 2 Reactor Vessel Beltline M aterials ............................................................... 23 2 St. Lucie Unit 2 Controlling Materials and Their Adjusted Reference Temperatures ............. 24 3 St. Lucie Unit 2 Cooldown and Heatup Allowable Pressure 55 EFPY, Adjusted to Actual Pressurizer Pressure, APCF ................................................................................................. 25 4 St. Lucie Unit 2 Cooldown and Heatup Allowable Pressure 55 EFPY, Adjusted to Indicated Pressurizer Pressure; IPCF.... ................................................ 26 5 St. Lucie Unit 2 Hydrostatic Test Pressure-Temperature Limit Data .................. 27 6 M axim um Transient Pressures .............................................................................................. 28 7 Summ ary of Controlling Pressures ............................... ........................................................ 29 8 LTOP Requirements, 55 EFPY ................ .............................. 30 9 Heatup Allowable Pressures, Uncorrected ........................................................................... 31 10 Cooldown Allowable Pressures, Uncorrected ..................................................................... 32 LIST OF FIGURES NO. TITLE PAGE 1 St. Lucie Unit 2 Cooldown P-T Limits 55 EFPY, APCF ...................................................... 33 2 St. Lucie Unit 2 Cooldown P-T Limits 55 EFPY, IPCF ........................... 34 3 St. Lucie Unit 2 Heatup P-T Limits 55 EFPY, APCF .......................................................... 35 4 St. Lucie Unit 2 Heatup P-T Limits 55 EFPY, IPCF ............................................................ 36 49 5 St. Lucie Unit 2 Energy Addition Transient w/ PORV, PSET= 0 PSIA ......................... 37 6 St. Lucie Unit 2 Energy Addition Transient w/ SDCRV, PSET= 3 50 PSIA ............................ 38 7 St. Lucie Unit 2 LTOP Mass Addition Transient, PORV Flowrate Vs Pressure ................... 39 8 St. Lucie Unit 2 LTOP Mass Addition, SDCRV Flowrate Vs Pressure ................................ 40 WCAP-1681 7-NP, Rev 2 Page v October 2007

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1 INTRODUCTION The following sections describe the basis for development of reactor vessel beltline pressure-temperature limitations and low temperature overpressure protection requirements for the St. Lucie, Unit 2, Nuclear Generating Station. These limits are calculated to meet the regulations of 10 CFR Part 50 Appendix A (Reference 1), Design Criterion 14 and Design Criterion 31. These design criteria require that the reactor coolant pressure boundary be designed, fabricated, erected and tested in order to have an extremely low probability of abnormal leakage, of rapid propagating failure, and of gross rupture. The criteria also require that the reactor coolant pressure boundary be designed with sufficient margin to assure that when stressed under operating, maintenance and testing the boundary behaves in a non-brittle manner and the probability of rapidly propagating fracture is minimized.

The pressure-temperature (P-T) limits are developed using the requirements of 10 CFR .50 Appendix G (Reference 2). This appendix describes the requirements for developing the pressure-temperature limits and provides the general basis for these limitations. The margins of safety against fracture provided by the pressure-temperature limits using the requirements of 10 CFR Part 50 Appendix G are equivalent to those recommended in the ASME Boiler and Pressure Vessel Code Section III, Appendix G "Fracture Toughness Criteria for Protection against Failure" (Reference 3). The general guidance provided in those procedures has been utilized to develop the St. Lucie Unit 2 pressure-temperature limits with the requisite margins of safety for the heatup and cooldown conditions.

The reactor pressure vessel beltline pressure-temperature limits are based upon the irradiation damage prediction methods of Regulatory Guide 1.99 Revision 2 (Reference 4). This methodology has been used to calculate the limiting material Adjusted Reference Temperatures for St. Lucie Unit 2 and has utilized fluence values for 55 Effective Full Power Years (EFPY).

This report provides reactor vessel beltline pressure-temperature limits in accordance with 10 CFR 50 Appendix G for 55 EFPY. The events analyzed are the isothermal, 10 through 100°F/hr cooldown conditions and the 50°F/hr heatup conditions. These conditions were analyzed to provide a data base of reactor vessel P-T limits for use in establishing low temperature overpressure protection requirements.

Low temperature overpressure protection (LTOP) requirements are established based upon the guidance provided in USNRC Standard Review Plan (SRP) 5.2.2 (Reference 5). Using this guidance the limiting transient pressures have been determined for mass and energy addition transients to establish the appropriate LTOP setpoints, heatup and cooldown rates, and administrative requirements.

Based upon the P-T limit analyses and LTOP requirements provided within this report, no limiting vessel operability issues are anticipated to exist.

WCAP-16817-NP, Rev 2 Page I October 2007

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2 PRESSURE - TEMPERATURE LIMITS 2.1 ADJUSTED REFERENCE TEMPERATURE PROJECTIONS In order to develop pressure-temperature limits over the design life of the reactor vessel, Adjusted Reference Temperatures (ART) for the controlling beltline material need to be determined. The adjusted reference temperatures of reactor vessel beltline materials for St. Lucie Unit 2 have been calculated at the 1/4t and 3/4t locations after 55 EFPY operation. By comparing ART data for each material, the controlling material for St.

Lucie Unit 2 has been determined.

The Adjusted Reference Temperature (ART) values have been calculated using the procedures in Regulatory Position 1.1 of Regulatory Guide 1.99 Revision 2 (Reference 4). The calculative procedure for the ART values for each material in the beltline is given by the following expression:

ART = Initial RTNDT + LRTNDT + Margin The fluence is attenuated through the vessel wall using the nominal reactor vessel thickness of 8.625 in.

(Reference 7), conservatively neglecting cladding thickness. The St. Lucie Unit 2 reactor vessel is plate limited, with the limiting 1/4t and 3/4t RTNDT values of 160'F and 137°F, respectively. The credible surveillance plate M-605-1 is shown in Table 1 for comparison. No margin reduction is taken for credible surveillance data (plate) since the reduction may be eliminated by proposed (ASTM E900 Draft) revisions to Regulatory Guide 1.99. The 55 EFPY projections for all beltline materials are provided in Table 1.

The following information provides the basis for the calculated ART values for St. Lucie Unit 2:

I. Material data were obtained from Reference 7, including copper content, nickel content and initial reference temperature (Initial RTNDT). These data are summarized in Table 1 for St. Lucie Unit 2.

2. Peak neutron fluence for the Unit 2 beltline region was determined to be 4.56 x 1019 n/cm2 (E>I MeV) at 55 EFPY (Reference 7).

3. Reactor vessel beltline thickness is 8.625 in. (Reference 7).

4. Calculations were based on the procedures in NRC Regulatory Guide 1.99, Rev. 2 (Reference 4).

Adjusted reference temperatures for all beltline materials at the 1/4t and 3/4t locations after 55 EFPY were calculated using Regulatory Guide 1.99 Revision 2 and the results of the calculation are listed in Table 1 for.

St. Lucie Unit 2. The controlling materials are shown in Table 2; the term "controlling" means having the highest ART for a given time and position within the vessel wall. The highest, or limiting, ARTs are then used to develop the pressure-temperature limits for the corresponding time period.

In the case of St. Lucie Unit 2, the intermediate shell plate M-605-2 is controlling at the 1/4t location and M-605-1 at the 3/4t location after 55 EFPY based on the predicted ART values of 160'F and 137°F respectively (Reference 7).

WCAP-16817-NP, Rev 2 Page 3 October 2007

According to Position 1.1 of Regulatory Guide 1.99, Revision 2 (Reference 4), the uncertainty in the value of initial RTNDT is to be estimated from the precision of test method when a "measured" value of initial RTNDT is available. RTNTT is derived in accordance with NB2300 of the ASMIE Boiler and Pressure Vessel Code,Section III. It involves both a series of drop weight (ASTM E208) and Charpy impact (ASTM E23) tests on

,the material. The RTNDT resulting from this two test method evaluation is conservatively biased. The elements of this conservatism include:

1) Choice for RTNDT is the higher of NDT or Tcv -60'F. The drop-weight test is performed to obtain NDT and a full Charpy impact curve is developed to obtain Tcv for a given material. The combination of the two test methods gives protection against the possibility of errors in conducting either test and, with the full Charpy. curve, demonstrates that the material is typical, of reactor pressure vessel steel. Choice of the more conservative of the two (i .e., the higher of NDT or Tcv -

60 0 F) assures that tests at temperatures above the reference temperature will yield increasing values of toughness, and verifies the temperature dependence of the fracture toughness implicit in the KIR curve (ASME Code,Section III, Appendix G).

2) Selection of the most adverse Charpy results for Tcv. In accordance with NB2300, a temperature, Tcv is established at which three Charpy specimens exhibit at least 35 mils lateral expansion and not less than 50 ft-lb absorbed energy. The three specimens will typically exhibit a range of lateral expansion and absorbed energy consistent with the variables inherent in the test: Specimen temperature, testing equipment, operator, and test specimen (e.g., dimensional tolerance and material homogeneity). All of these variables are controlled using process and procedural controls, calibration and operator training, and they are conservatively bounded by using the lowest measurement of the three specimens. Furthermore, two related criteria are used; lateral expansion and absorbed energy, where consistency between the two measurements provides further assurance that they are realistic and the material will exhibit the intended strength, ductility and toughness implicit in the KIR curve.

3) Inherent conservatism in the protocol used in performing the drop-weight test. The drop-weight test procedure was carefully designed to assure attainment of explicit values of deflection and stress concentration, eliminating a specific need to account for below nominal test conditions and thereby guaranteeing a conservative direction of these uncertainty components. In addition, the test protocol calls for decreasing temperature until the first failure is encountered, followed by increasing the test temperature 10°F above the point where the last failure is encountered. This in fact assures that one has biased the resulting estimate toward a low failure probability region of the temperature versus failure rate function. The effect of this protocol is to conservatively accommodate the integrated uncertainty components.

Given the three elements of conservatism described above, values of initial RTNDT obtained in accordance with NB2300 will result in a conservative measure of the reference temperature.

The conservative bias of the NB2300 methodology and the drop-weight test protocol essentially eliminate the uncertainty which might result from the precision of an individual drop-weight or Charpy impact test.

Therefore, when measured values of RTNDT are available, the estimate of uncertainty in initial RTNDT is taken as zero.

WCAP-16817-NP, Rev 2 Page 4 October 2007

2.2 GENERAL APPROACH FOR CALCULATING PRESSURE-TEMPERATURE LIMITS The analytical procedure for developing reactor vessel pressure-temperature limits utilizes the methods of Linear Elastic Fracture Mechanics (LEFM) and guidance found in the ASME Boiler and Pressure Vessel Code Section XI, Appendix G (Reference 3) in accordance with the requirements of 10 CFR Part 50 Appendix G. For these analyses, the Mode I (opening mode, according to fracture mechanics terminology) stress intensity factors are used for the solution basis.

The St. Lucie, Unit 2, 55 EFPY P-T Limits analysis utilizes a Westinghouse developed and quality assured computer code to generate P-T Limits for the reactor beltline region. That computer code uses superposition technique and influence coefficients to calculate these curves.

The reactor coolant system (RCS) P-T limit curves are based on the beltline P-T limits for a set of heatup and cooldown rates. These curves are then adjusted to represent pressurizer pressure conditions (the adjustment addresses both the RCS hydraulic pressure drop due to flow and pressurizer-to-beltline region elevation) and, where appropriate, adjusted for temperature and pressure instrumentation uncertainties.

The final P-T limits include the minimum bolt-up temperature, lowest service temperature, and the flange limit. Minimum bolt-up and lowest service temperatures are both specified in Reference 7.

LTOP enable temperatures are also determined using heat transfer results from the P-T limits analysis and applying ASME Boiler and Pressure Vessel Code Section XI, Appendix G methodology.

The temperature distribution through the reactor vessel wall is characterized by a partial differential equation, defined for the applicable boundary conditions and geometry, and solved numerically. Numerical solution uses a finite element model to determine wall temperature as a function of radius, time, and thermal rate.

This method utilizes three-noded, isoparametric finite elements suitable for one-dimensional, axisymmetric radial conduction-convection heat transfer. The wall is divided into 11 elements. The first element represents cladding, and the remaining 10 elements represent base metal. The analysis code utilizes convective boundary conditions on the inside wall and an insulation boundary on the outside wall of the reactor vessel. Variation of material properties through the wall is permitted. This allows for the change in material thermal properties between the cladding and the base metal.

The reactor vessel beltline region is analyzed assuming a semi-elliptical surface flaw oriented in the axial direction, with a depth of one-quarter of the reactor vessel beltline thickness. The assumed flaw has an aspect ratio of one to six. The postulated flaw is analyzed at both the inside diameter location (referred to as the 1/4 t location) and the outside diameter location (referred to as the 3/4 t location) to ensure the most limiting condition is achieved.

At each of the postulated flaw locations, the Mode I stress intensity factor, K1, produced by each of the specified loadings, is calculated. The summation of the K, values is compared to reference stress intensity, KIc, which is the critical value of K, for the involved material and temperature. The result of this method is a WCAP-16817-NP, Rev 2 Page 5 October 2007

relationship of pressure versus temperature for reactor vessel operating limits, which conservatively precludes brittle fracture.

KIR is obtained from a reference fracture toughness curve for reactor vessel low alloy steels, and is defined in Appendices A and G of Section XI of the ASME Code (References 13 and 3). In this calculation, KIR is defined as Klc, and it is the lower bound of static initiation critical K1 values measured as a function of temperature. This governing curve is defined by Equation 3 below.

For operational events, P-T limits are calculated using the equation:

2KLW +KIT KIc Equation 1 where:

KIM = Membrane (pressure) stress intensity factor, ksii~n KIT = Thermal stress intensity factor, ksiJn K1c = Reference stress intensity factor, ksijn Rearranging the terms in the above equation:

K~v = Kic -- KIT 2 Equation 2 Allowable pressure is then computed using the allowable membrane stress intensity factor from Equation 2 and the pressure influence coefficients. The fracture toughness is:

KlC = 33.2 + 2 0. 7 3 4 e°°2(T-RTArT) Equation 3 For the hydrostatic test limits, the structural factor 2 in Equation 1 is replaced by 1.5:

1.5KM < KIc Equation 4 1.5 Equation 5 For any instant during the postulated heatup or cooldown, K1c is calculated at the metal temperature and at the adjusted RTNDT at the tip of the flaw. The temperature distribution and the temperature at the flaw tip are calculated using a one-dimensional, three-noded isoparametric finite element suitable for one-dimensional radial conduction-convection heat transfer analysis.

The fracture mechanics algorithms use a superposition technique using influence coefficients to calculate the Mode I stress intensity factors.

WCAP-16817-NP, Rev 2 Page 6 October 2007

At the conditions of 55 EFPY, isothermal and transient conditions were analyzed. The cooldown transients analyzed at rates of I 0F/hr, 30'F/hr, 50'F/hr, 75°F/hr, and 1000 F/hr begin at a bulk coolant temperature of 550'F and terminate at 70'F. The heatup transient analyzed has a rate of 50°F/hr and begins at a bulk temperature of 70'F and terminates at 550'F. The hydrostatic limits were obtained for the isothermal condition only.

2.2-1 Application of Pressure Correction Factors The P-T limits, as directly calculated by ASME methodology, typically represent the limiting material conditions at the reactor vessel beltline. However, these beltline P-T limits cannot be used directly by the plant operations staff, since pressure measurement in the RCS is limited to the pressurizer and, as such, the beltline values require adjustment to representative values relative to the pressurizer location.

For adjustment of the beltline P-T limits to the pressurizer, pressure correction factors (PCF) are used. These PCF were updated for the current plant operations for this analysis and consist of:

(1) The pressure difference due to the static head of fluid between the pressurizer pressure instrument nozzle elevation and reactor vessel beltline lowest point; (2) The flow-induced pressure drop between the applicable point in the reactor vessel and hot leg surge line nozzle, due to flow resulting from operating reactor coolant pumps (RCP); and (3) The uncertainty associated with the pressure instrumentation, as applicable.

These correction factors are applied to the beltline P-T limits in two manners. An actual pressure correction factor (APCF) is applied to the beltline P-T limits to provide representative actual (or analysis) values relative to the pressurizer location. An APCF is developed from plant data associated with items (1) and (2) in the prior paragraph's explanation. While it is possible to develop multiple APCF to represent multiple plant operating conditions (e.g. combinations of operating RCP) a bounding static head condition (item (1))

and the single bounding consideration of three operating RCP (item (2)) were selected. This choice to have a single PCF makes it applicable throughout the LTOP range where any combination of three or less RCP are operating. This updated APCF was also developed to be bounding of both the plant conditions of original and replacement steam generators. The APCF value is 74.5 psid.

The potential for up to three operating RCP condition fully bounds the plant operating conditions within the LTOP applicable range. Current plant procedures limit the operation of four RCP to greater than 500'F, with the possible future plant change to permit this limitation to be lowered to approximately 400'F. Inspection of the P-T limits, Figures 1 and 3, the most limiting pressures are greater than 2500 psia at any temperature value above 300'F. Thus the plant will not be in a operating condition where a 4-RCP APCF is necessary.

However, an informational value was prepared, and noted in the Data, Section 2.8 Due to uncertainties in the pressurizer pressure instrument loop components, indicated pressurizer pressure observed by control room operators can differ from actual pressurizer pressure. If unaccounted for, actual pressurizer pressure can be greater than indicated pressurizer pressure, which could potentially lead to violation of the actual P-T limits. To prevent this violation, an indicated pressure correction factor (IPCF) of 91.7 psid is applied to the beltline P-T limits. This accounts for this instrumentation uncertainty, in addition to the previously described adjustments for actual limits, to represent indicated P-T limits.

WCAP-16817-NP, Rev 2 Page 7 October 2007

In conditions where the indicated P-T limits are developed (IPCF are applied), corresponding conservative adjustment of the temperature values are accommodated by a temperature correction factor, which acknowledges the possible uncertainty of the temperature indication loop. A value of 10'F is applied to adjust the P-T limits as presented in the Technical Specification figures, as this represents the control room instrument error. A value if 14'F is applied to the LTOP Enable Temperature values (Table 8) since these values are associated with the PORV actuation channels.

2.3 THERMAL ANALYSIS METHODOLOGY The thermal stress intensity factors are found using the temperature differences through the wall as a function of transient time. Then, they are subtracted from the available KIR value to calculate the allowable pressure stress intensity factor and, consequently, the allowable pressure.

Equation 1 provides the expression used to derive Pressure-Temperature limits. The superposition technique used is temperature profile-based rather than the commonly used stress profile-base. A third-order polynomial fit to the temperature distributions in the wall was used (Reference 20):

T(x) = C +C 1 (-h)+ XXC 2 (1-E 2 +C3 (1- X3 Equation 6 where:

T(x) = temperature at radial location x from inside wall surface CO-C 3 = coefficients in polynomial fit x = distance through beltline wall, inches h = beltline wall thickness, inches The unit K1 values are calculated for each term of the polynomial using a two-dimensional finite element code. These unit values are used to determine the total K1 value for the applied loads under any general temperature profile in the wall that occurs during the thermal transient.

The thermal stress intensity factor is represented by Equation 7:

3 KIT (a) =ZCIKi i=0 Equation 7 where KIT = thermal stress intensity factor Ci = coefficients in polynomial fit Ki = polynomial influence coefficients Temperature-based influence coefficients for determination of the thermal stress intensity factor, KIT, are used. Using Reference 18 methods, these coefficients were computed using a two-dimensional reactor vessel model with a crack adjusted to account for three-dimensional effects.

WCAP-16817-NP, Rev 2 Page 8 October 2007

2.4 COOLDOWN LIMIT ANALYSIS During cooldown, membrane and thermal bending stresses act together in tension at the reactor vessel inside wall. This results in the pressure stress intensity factor, KIM, and the thermal stress intensity factor, KIT, acting in unison to create high stress intensity. At the reactor vessel outside wall, the tensile pressure stress and the compressive thermal stress act in opposition, resulting in a lower total stress than at the inside wall location. Also, neutron embrittlement, the shift in RTNDT, and the reduction in fracture toughness are less severe at the outside wall when compared to the inside wall location. Consequently, the inside flaw location is more limiting for the cooldown event.

Utilizing the material metal temperature and adjusted RTNDT at the 1/4t and 3/4t locations, the reference stress intensities are determined. From the finite element method used for the heat transfer analysis, the through-wall temperature gradient is calculated for the assumed'cooldown rate to determine the thermal stress intensity factor. The thermal stress intensity factors are found using the temperature difference through the wall as a function of transient time. Then, they are subtracted from the available K1c value to calculate the allowable pressure stress intensity factor and, consequently, the allowable pressure.

The cooldown pressure-temperature curves are thus generated by calculating the allowable pressure on the reference flaw at the 1/4t and 3/4t locations. This is based upon Equation 2 of Section 2.2.

To develop a minimum pressure-temperature limit for the cooldown event, the isothermal pressure-temperature limit must be calculated. Then, the isothermal pressure-temperature limit is compared to the pressure-temperature limit associated with a cooling rate. The more restrictive allowable pressure-temperature limit is chosen, resulting in a minimum limit curve for the reactor vessel beltline.

Table 10 presents the pressure-temperature limits results for conditions at the beltline (without applied correction factors) for cases for isothermal and I0F/hr, 30°F/hr, 50°F/hr, 75°F/hr, and 100°F/hr cooldown.

Table 3 and Table 4 provide the results that include the APCF and IPCF respectively. APCF data is used for comparison to the design basis LTOP transient results, which are also referenced to the pressurizer pressure location. IPCF data is used for the recommended Technical Specification P-T Limit figure changes.

Uncorrected values are provided for completeness.

2.5 HEATUP LIMIT ANALYSIS During heatup, the thermal bending stress is compressive at the reactor vessel inside wall and is tensile at the reactor vessel outside wall. Internal pressure creates a tensile stress at the inside wall and outside wall locations. Consequently, the outside wall location, when compared to the inside wall, has the larger total stress. However, neutron embrittlement, shift in material RTNDT, and reduction in fracture toughness are greater at the inside location. Therefore, results from both the inside and outside flaw locations must be compared to ensure recognition of the most limiting condition.

As described in the cooldown case, the reference stress intensity is calculated at the metal temperature and the adjusted RTNDT is calculated at the tip of the flaw. Using a finite element method' for the heat transfer analysis, the temperature profile through the wall and the metal temperatures at the tip of the flaw are calculated for the transient history. This information, in conjunction with the calculated wall gradient and thermal influence coefficients, is used to calculate the thermal stress intensity factor at the 1/4t and 3/4t WCAP-16817-NP, Rev 2 Page 9 October 2007

locations. Then, the allowable pressure stress intensity is determined by superposition of the thermal stress intensity factor, with the available reference stress intensity, at the flaw tip. Allowable pressure is derived from the calculated allowable pressure stress intensity factor.

A sign change occurs in the thermal stress through the reactor vessel beltline wall. Assuming a reference flaw at the 1/4t location, the thermal stress tends to alleviate the pressure stress. This alleviation indicates that the isothermal steady state condition would represent the limiting P-T limit. However, the isothermal condition may not always provide the limiting pressure-temperature limit for the 1/4t location during a heatup transient. This is due to the difference between the. base metal temperature and the RCS fluid temperature at the inside wall. For a given heatup rate (non-isothermal), the differential temperature through the clad and film increases as a function of thermal rate, resulting in a crack tip temperature which is lower than the RCS fluid temperature. Therefore, to ensure accurate representation of the 1/4t pressure-temperature limit during heatup, both the isothermal and heatup rate dependent pressure-temperature limits are calculated. This also ensures that the limiting condition is recognized. These limits, in conjunction with the cooling limits, account for clad and film differential temperatures, as well as the gradual buildup of wall differential temperatures with time.

To develop minimum pressure-temperature limits for the heatup transient, the isothermal conditions at 1/4t and 3/4t, 1/4t heatup, and 3/4t heatup pressure-temperature limits are compared for a given thermal rate.

Then, the most restrictive pressure-temperature limits are combined, resulting in a minimum limit curve for the reactor vessel beltline for the heatup event.

Table 9 presents the P-T results for conditions at the beltline, without applied correction factors, for isothermal and 50°F/hr heatup pressure-temperature limits. Table 3 provides the results with APCF. Table 4 provides results with IPCF, which include temperature and pressure correction factors. Tables 3 and 4 supply the allowable pressurizer pressure values versus reactor coolant temperature. APCF data is used for comparison to the design basis LTOP transient results, which are also referenced to the pressurizer pressure location. IPCF data is used for the recommended Technical Specification P-T Limit figure changes.

Uncorrected values are provided for completeness.

2.6 HYDROSTATIC TEST AND CORE CRITICAL LIMIT ANALYSIS Hydrostatic test limits have been calculated for 55 EFPY using the methodology of the ASME Boiler and Pressure Vessel Code, Section X1, Appendix G. The governing equation for determination of hydrostatic test limits is Equation 4 from Section 2.2.

The procedure is similar to calculating normal operations' heatup and cooldown limits. The one exception is the factor of safety that is applied to the allowable pressure stress intensity (KIM). To account for this exception, the analysis method utilized for this calculation modified the applied factor of safety from 2.0 (for normal operation) to 1.5 for hydrostatic limits.

The purpose of the hydrostatic test limit is to establish the minimum temperature required at the corresponding hydrostatic test pressure. Westinghouse recommends that the in-service hydrostatic test for Combustion Engineering (CE) Nuclear Steam Supply Systems (NSSS) designs be performed at a test pressure corresponding to 1.1 times the operating pressure, with the reactor core not critical. Under these conditions, 10CFR50, Appendix G requires that the minimum temperature for the reactor vessel be at least as WCAP-16817-NP, Rev 2 Page 10 October 2007

high as the RTNDT for the limiting material in the closure flange region plus 90'F. However, the beltline hydrostatic test, at the recommended test pressure, has greater limitations. Therefore, it is only necessary to control plant operations to the beltline in-service hydrostatic test limits in the vicinity of this pressure.

To define minimum temperature criteria for core critical operation, Appendix G of 10 CFR 50 specifies the following pressure-temperature limits:

If the RCS pressure is less than or equal to 20% of the Pre-service Hydrostatic Test Pressure (PHTP),

the minimum temperature requirement for the reactor vessel must be at least as high as the RTNDT for the limiting material in the closure flange region stressed by bolt preload plus 40'F, or the minimum permissible temperature for the in-service hydrostatic pressure test, whichever is larger.

If the RCS pressure is greater than 20% of the PHTP, the minimum temperature requirement for the reactor vessel must be at least as high as the RTNDT for the limiting material in the closure flange region stresses by bolt preload plus 160'F, or the minimum permissible temperature for the in-service hydrostatic pressure test, whichever is larger.

According to Appendix G to IOCFR50, the following calculation specifies pressure-temperature limits for core critical operation to provide additional margin during actual power operation:

In-service hydrostatic pressure =

= (1.1 x operating pressure) + instrumentation uncertainty

= (1.1 x (2,250-14.7) + 14.7 psia) + 0 psi = 2,473.5 psia Pressure instrumentation uncertainty is not included. Furthermore, the factor 1.1 is used for the gauge units (psig) of operating pressure instead of the absolute units (psia).

The minimum temperature for the core critical operation and the hydrostatic test is the temperature corresponding to the in-service hydrostatic pressure. The minimum temperature for the hydrostatic and leak test cases is 217.5°F. This temperature value was obtained from Table 5 (unadjusted, beltline data) by interpolating the temperature values to the pressure given above.

Hydrostatic test limits are tabulated in Table 5, and are adjusted using the correction factors for both the APCF and IPCF cases.

For both the APCF and IPCF cases, the specified beltline heatup P-T limit is more restrictive at temperatures above 217.5°F. Consequently, the core critical limits have been established as a combination of this temperature and the specified heatup P-T limit from ASME Appendix G, plus 40'F.

The core critical limits established are based solely on fracture mechanics considerations, and do not consider core physics safety analyses. Core physics safety analyses can control the temperature at which the core can be brought critical.

WCAP-16817-NP, Rev 2 Page 11 October 2007

2.7 LOWEST SERVICE TEMPERATURE, MINIMUM BOLTUP TEMPERATURE, FLANGE LIMIT TEMPERATURE, AND MINIMUM PRESSURE LIMITS In addition to the computation of the reactor vessel beltline P-T limits, additional limits have been provided for reference. These additional limits are the Lowest Service Temperature, Minimum Boltup Temperature, Flange Limit Temperature, and Minimum Pressure Limit. These limits are described below.

The Lowest Service Temperature (LST) is defined in ASME Section III, NB-3211 as the minimum temperature for piping, pumps and valves (the remainder of the RCS) in the RCS in order to exceed the 20%

of the pre-service hydrostatic test pressure. The LST shall be established as a temperature not less than RTNDT of the remainder of the RCS + 100'F. From UFSAR Tables 5-2-8, 5.2-9, and 5.2-12 the highest RTNDT for the RCS is +60'F, therefore the LST is 160'F. No uncertainty need be applied to this value since it is entirely behind the LTOP window.

When the pressure exceeds 20% of pre-service hydrostatic test pressure, the temperature of the closure flange regions must exceed the initial RTNDT of the material by at least 120'F for normal operation and by 90'F for hydrostatic and leak testing.

The Minimum Pressure Limit is applicable between the Minimum Boltup Temperature, Lowest Service Temperature, and the Flange Limit Temperature. Defined by the ASME Boiler and Pressure Vessel Code as 20% of the pre-operational hydrostatic test pressure, the minimum pressure is as follows:

20% of Pre-Service Hydrostatic Test = (1.25 x Design Pressure) x 0.20

= 1.25 x (2,500-14.7) x 0.20 + 14.7 = 636 psia With the correction factors (APAPCF = 74.5 psid, APIPCF = 91.7 psid), this pressure is adjusted to 561.5 psia for APCF and 544.3 psia for the IPCF cases.

The scale factor used on the design pressure in the previous calculation is the gauge value (psig) instead of the absolute pressure (psia).

The minimum boltup temperature is defined as 80'F, which provides margin to protect the vessel head, vessel flange and upper shell from being stressed at a temperature below the RTNDT of those materials. The PT curves include a I0F margin shift for indicated instrument uncertainty so that the operator does not need to account for the instrument error at bolt up. For steady state, a 30'F margin on minimum bolt up temperature is already in place since the lowest RTNDT of the flange region is 50F.

WCAP-16817-NP, Rev 2 Page 12 October 2007

2.8 DATA Reaqctor Vessel-. Data R eference.

Reactor Vessel Data Reference Design Pressure = 2500 psia 7 Design Temperature = 650OF 7 Operating Pressure = 2250 psia 7 Beltline Thickness = 8.625 in 7 Inside Radius = 86.964 in 7 Outside Radius = 95.81 in 7 Cladding Thickness = 0.2187 in 7 Material-SA 533 Grade B Reference Thermal Conductivity = 23.8 BTU/hr-ft-°F 15 Youngs Modulus = 28 x 106 psi 15 6

Coefficient of Thermal = 7.8 x 10- in/in-°F 15 Expansion Specific Heat = 0.122 BTU/lb- 0F Density = 49Olb / ft3 Stainless Steel Cladding Reference Thermal Conductivity = 10.1 BTU/hr-ft- °F 15 Adiusted Reference Temperature Values 55 EFPY Reference 1/4t 160OF 7 3/4t 137 0 F 7 Film coefficient on inside surface = 1000 BTU/hr-ft2 -°F WCAP-16817-NP, Rev 2 Page 13 October 2007

Pressure Correction Factors for Elevation and Flow Applicable to all plant condition with three or less RCP in operation and with either OSG or RSG:

Actual Pressure Correction Factor APCF = 74.5 psid Indicated Pressure Correction Factor:

Narrow Range Pressure Instruments: IPCF = 91.7 psid Wide Range Pressure Instruments IPCF = 133.4 psid Corresponding Information values for four operating RCP are:

Actual Pressure Correction Factor APCF = 75.1 psid Indicated Pressure Correction Factor:

Narrow Range Pressure Instruments: IPCF = 92.3 psid Wide Range Pressure Instruments IPCF = 134.0 psid WCAP-16817-NP, Rev 2 Page 14 October 2007

3 LOW TEMPERATURE OVERPRESSURE PROTECTION 3.1 GENERAL The primary objective of LTOP systems is to preclude violation of applicable Technical Specification P-T limits during startup and shutdown conditions. These P-T limits are usually applicable to a finite time period of operation and are based upon the irradiation damage prediction by the end of the period. Accordingly, each time new P-T limits are to become effective, the LTOP system needs to be re-analyzed and modified, if necessary, to continue its function.

The objective of the LTOP system is to prevent violation of the RCS brittle fracture P-T limits in case of an overpressure event within the LTOP temperature range. A reactor coolant pump (RCP) start overpressure event is one of two design basis events for the LTOP system. The RCP start is referred to as the Energy Addition event. The other design basis event is the mass addition transient, typically based on an inadvertent Safety Injection Actuation Signal (SIAS) in the LTOP temperature range.

A typical LTOP system includes pressure relieving devices and a number of administrative and operational controls. At St. Lucie Unit 2, the current LTOP system makes use of two shutdown cooling relief valves (SDCRV) at the low RCS temperatures, starting at the boltup temperature, and two power operated relief valves (PORVs), in the remaining part of the LTOP temperature range up to the LTOP enable temperature.

The SDCRVs (Tag Nos. V3666 and V3667) are spring operated relief valves with a lift setting of 350 psia each. The PORVs (Tag Nos. V1474 and V1475) are power operated relief valves with a nominal opening setpoint of 470 psia.

These relief valves, in combination with certain other limiting conditions for operation contained in Technical Specifications, comprise the St. Lucie Unit 2 LTOP System.

Since the new P-T limits described in this report cover the operating period ending at 55 EFPY, the existing LTOP system was re-analyzed to determine if modifications are required or improvements can be implemented in order for the system to provide adequate LTOP through 55 EFPY. The LTOP system was analyzed for both OSG (Original Steam Generator) and RSG (Replacement Steam Generator) conditions.

The following sections document the method, assumptions and results of the new analyses.

3.2 METHOD AND ASSUMPTIONS The approach taken was to analyze the existing SDCRV and PORV :setpoints. Accordingly, the existing SDCRV lift setting of 350 psia and the nominal PORV setpoint of 470 psia were used. Additional PORV setpoints were analyzed as described below, as possible alternatives to 470 psia setpoint.

The following existing general assumptions are used in the LTOP analyses:

1. Only one SDCRV or PORV is available.

2. RCS is in a water solid condition.

3. Letdown flow paths are isolated.

4. Pressurizer heater input and decay heat input are considered as additional energy sources.

WCAP-16817-NP, Rev 2 Page 15 October 2007

5. There is no heat absorption or metal expansion at the primary pressure boundaries.

The PORV opening characteristics are adjusted for control circuit uncertainty and valve response time. This is addressed in the following manner:

1. RCS pressure just prior to PORV opening was conservatively assumed to be greater than the nominal PORV setpoint, due to relative pressure instrument uncertainty between the pressure indication and the PORV actuation channels. This 21 psi uncertainty was provided in Reference 7.

2. PORV opening time was assumed to equal 1.30 seconds which envelopes the opening times observed in applicable tests with a Garrett PORV. Based on an evaluation of test data it was assumed that this total opening time consisted of solenoid delay time of 0.15 seconds and stroke time of 1.15 seconds. The computer code used to model the energy addition transient cannot model a ramped PORV opening. To account for a ramp opening during stroke time, a delay in PORV opening equal to a sum of the solenoid delay time (0.15 seconds) and one half of the stroke time (0.575 seconds) was assumed in the energy addition transient analysis. This delay was assumed to be followed by instantaneous opening. The PORV opening setpoint used in the energy addition transient analysis code is adjusted based on this delay time, assuming a bounding pressure ramp rate prior to valve opening. In the Mass Addition transient analysis, the PORV is opened in time steps following the solenoid delay. PORV capacity x (time passed/stroke time) is credited as stroke time passes, until the PORV is full open.

3. The impact of PORV opening time was taken into account in the Energy Addition transient analysis by adding transient-specific pressure accumulation during 0.725 seconds (0.15 seconds plus 0.575 seconds) to the opening pressure to arrive at the maximum opening pressure. Pressure accumulation was assumed to be a function of an applicable pressurization ramp rate moments prior to reaching the valve setpoint.

Based on the existing analyses, modified assumptions and inputs, new maximum transient pressures for the same design basis transients were determined as appropriate. Out of these, the most limiting pressures in given temperature ranges were selected as "controlling" the limiting temperatures for LTOP.

Finally, by comparing these controlling pressures to the P-T limit curves for 55 EFPY, limiting conditions for operation were identified.

3.3 PRESSURE TRANSIENT ANALYSES 3.3-1 Energy Addition Transients The energy addition analysis determines the peak pressure which would occur as a result of the RCS pressure transient caused by a RCP start with an initial steam generator-to-reactor vessel temperature differential of 40'F (Reference 7, Section 5.1.3) during RCS water-solid, low temperature conditions.

LTOP system overpressure protection at St. Lucie Unit 2 is provided by PORVs or SDCRVs in accordance with Reference 10, LCO 3.4.9.3. This calculation analyzes cases with either a single PORV or single SDCRV providing LTOP overpressure protection.

WCAP- 16817-NP, Rev 2 Page 16 October 2007

An analysis methodology, consistent with the transient analysis of record, was used with inputs provided in accordance with Reference 7. Plant specific volumes, masses, decay heat, RCP heat, pressurizer heater contributions, selected initial temperatures, and heat transfer coefficients are incorporated into the analysis input, which produces results in the form of RCS system pressure values versus time.

The SDCRV mitigated energy addition transient for OSGs and RSGs was analyzed for the existing SDCRV setpoint of 350 psia (Reference 7, Section 5.3.1). The PORV mitigated energy addition transient analysis for OSGs and RSGs was performed for the existing PORV setpoint of 470 psia (Reference 7, Section 5.3.2).

Based on the results of this analysis, the limiting RSG case was analyzed for additional PORV setpoints of 450 psia and 490 psia.

The analysis assumed that the pressure transient was taking place in the pressurizer. The effect of the PORV inlet piping on the analysis results was taken into account by determining PORV flow rates at PORV inlet pressure, which was corrected from pressurizer pressure for elevation difference and flow losses. This correction reduces PORV discharge, thus maximizing the transient pressures.

The following major assumptions were used in the analysis of the RCP start transient, in addition to the assumptions mentioned above and in Section 3.2:

1) PORV opening occurs at an opening pressure that is greater than the nominal setpoint by a sum of the pressure instrument uncertainty and pressure accumulation due to finite opening time. This assumption maximizes RCS pressure at the PORV opening.

2) The initial RCS fluid temperature is 290'F. This temperature is consistent with that assumed in existing analyses, and bounds the updated LTOP enable temperatures of 246°F for heat up and 224°F for cooldown (Table 8).

3) The initial RCS pressure is 300 psia, consistent with existing analyses.

4) The historical St. Lucie Unit 2 LTOP Energy Addition analyses only consider the water and metal masses inside and including the Steam Generator (SG) downcomer to contribute as heat sources.

This analysis maintains the assumption for the Original Steam Generators (OSGs). However, when evaluating cases with Replacement Steam Generators (RSGs), the entire secondary side SG water mass was conservatively used due to the unavailability of a detailed RSG water mass breakdown.

5) RCP heat input is considered as an additional energy source.

The PORV mitigated (490 psia setpoint) pressure transient is graphically illustrated in Figure 5 and the resulting maximum transient pressure of 541.6 psia is provided in Table 6. The SDCRV mitigated pressure transient is graphically illustrated in Figure 6 and the resulting maximum transient pressure adjusted to a pressurizer pressure of 368.2 psia is provided in Table 6.

3.3-2 Mass Addition Transients The RCS pressure transient due to an inadvertent safety injection actuation is the design basis mass addition transient. The most severe mass addition transient results from an actuation of two HPSI pumps with a simultaneous operation of all three charging pumps, with letdown isolated. This transient, however, is only analyzed at RCS temperature above 200'F, consistent with existing LTOP controls on HPSI pump availability limitations in the Technical Specifications, LCO 3.5.3; Reference 10. As a result, at RCS WCAP-16817-NP, Rev 2 Page 17 October 2007

temperature below 200'F, the most limiting mass addition transient is due to one HPSI and three charging pumps input.

The following major assumptions were used in the analysis of the mass addition transients, in addition to the assumptions mentioned above and in Section 3.2:

1) It is assumed that the SDCS will be aligned below 200'F. In this configuration, one HPSI and 3 Charging pumps may be aligned.

2) The configuration with the SDCS isolated may allow two HPSI pumps and 3 Charging pumps to be aligned. The PORV is the primary LTOP protection device.

3) In all transient cases, only a single pressure protection relief valve is assumed.

4) The RCS volume with OSGs (and 200 tubes plugged) is bounding for both OSG and RSG conditions and is used in this analysis. Using a smaller RCS volume maximizes the effect of the mass addition and is conservative. Increasing S/G Tube Plugging to as high as 40% tubes plugged would lower the RCS volume, adding even more conservatism. However, this would not change the outcome of the analysis, as explained below.

5) As many as 3 RCP's are operational at startup and during fill and vent and could be operating during the LTOP Mass addition transient. However, the RCP heat input for the mass addition transient (consistent with current methodology) need not be considered since the transient initiates with the plant in a steady state condition (operator controlled heatup or cooldown) and instantaneous RCP start is not a credible transient input.

6) PZR initial conditions are assumed to be 300 psia and 417.3°F (saturation), consistent with existing analyses.

The analysis updates the existing design inputs and assumptions to more accurately represent the current operating configuration. RCS volume expansion due to contributions from decay heat and full pressurizer heater heat are taken into account. PORV discharge flowrate vs. pressurizer pressure is plotted in Figure 7.

The SDCRV release vs. pressurizer pressure curve is plotted separately in Figure 8. The mass addition events (including the RCS volume expansions) are plotted together with the PORV (Figure 7) and SDCRV (Figure 8) cases and equilibrium pressures are determined. An equilibrium pressure is the pressure at which the mass inputs match the relief valve discharge. PORV transient analyses are performed to determine maximum transient pressures for three different PORV set pressures: 360, 440, and 470 psia (at the request of FPL). The transient analysis calculates RCS pressure over time steps until an equilibrium is reached between HPSI and charging pumps inflow and PORV outflow.

The equilibrium pressures are as follows:

Transient Equilibrium Pressure (PORV Mitigation) 2 HPSI + 3 Charging Pumps 549 psia 1 HPSI + 3 Charging Pumps 377 psia The equilibrium pressure of 549 psia is the maximum (peak) pressure for this transient. For all three PORV setpoints analyzed, 360, 440 and 470 psia, the peak transient pressure is calculated to be between 548 and 549 psia, rounded up to 549 psia. This is expected; as the equilibrium pressure is much greater than any of the three PORV setpoints analyzed. In all cases, the system pressure continues to rise past the PORV WCAP-16817-NP, Rev 2 Page 18 October 2007

setpoint until the RCS back-pressure is high enough to limit the HPSI injection total mass addition to the capacity of the PORV. This is independent of PORV setpoint and only a function of PORV capacity.

For the I HPSI case, the equilibrium pressure is 377 psia. The peak pressure for the 1 HPSI case is calculated to be between 509 and 510 psia, rounded up to 510 psia. A sensitivity case is also run, to justify Assumption 4. The sensitivity case is run on the most limiting transient; 2 HPSI / 3 Charging pumps and a PORV setpoint of 470 psia. A 40% S/G tube plugging RCS volume of 10,068.8 ft3 is used, and new pressure iterations performed for each time step. The final peak/equilibrium pressure remains at 548.23 psia, proving that maximizing tube plugging would not affect the outcome of this analysis.

Unlike the PORV, there is no delay time or setpoint overshoot associated with' the SDCRV. The maximum transient pressure is therefore equal to the equilibrium pressure. The following equilibrium pressures for SDCRV mitigated transient are determined from Figure 8:

1. 366 psia (in PZR) for the 1 HPSI and 3 Charging pump lineup;

2. 384 psia (in PZR) for the 2 HPSI and 3 Charging pump lineup (Analysis applicable >200'F).

The final results of the mass addition transient analysis are provided in Table 6.

3.3-3 Controlling Pressures The pressure transient analysis results contained in Table 6 were evaluated to identify the controlling pressures and applicable temperature ranges. The controlling pressures are the maximum transient pressures of all applicable transients in a particular temperature region. Since a PORV mitigated 1 HPSI & 3 charging pump mass addition transient with a PORV setpoint of 490 psia was not analyzed it is assumed based on the results of the 470 psia PORV setpoint transients that the energy addition peak pressure remains the limiting pressure for PORV mitigated transients when indicated RCS temperature is less than 2000F.

Table 7 contains a summary of controlling pressures that were utilized in the determination of limiting conditions for operation that result from LTOP requirements. These limiting conditions for operation are provided in Section 3.4.

3.4 LIMITING CONDITIONS FOR OPERATION The temperature requirements for aligning the SDCRVs and PORVs for LTOP and the limitations on heatup and cooldown rates are provided in Table 8. These requirements are based on a PORV setpoint of 490 psia, but are also valid (bounding) for a PORV setpoint of 470 psia.

It should be noted that during heatup, the LTOP function can be transferred from the SDCRVs to the PORVs at any temperature above the minimum cold leg temperature for PORV use for LTOP (e.g., 80'F in Table 8).

During cooldown, however, the SDCRVs must take over the LTOP function upon reaching the indicated temperature of 132°F in Table 8.

The existing Technical Specification LTOP requirements related to the limitations on RCP starts, operating RCP and HPSI pump alignment to the RCS remain unchanged except for the temperature range of applicability for the RCP start limitations during heatup or cooldown.

WCAP-16817-NP, Rev 2 Page 19 October 2007

3.5

SUMMARY

OF PROPOSED CHANGES The proposed LTOP system is designed in accordance with the requirements set forth in the NRC Branch Technical Position RSB 5-2, Reference 5.

The proposed system is adequate to prevent violation of Appendix G P-T limits during the operating period ending at 55 EFPY. In order to implement the proposed LTOP system the following is required:

Modification of appropriate Technical Specifications and

Modification of appropriate plant operating procedures.

The implementation of the proposed LTOP system will not result in a reduction in the margin of safety presently afforded by Technical Specifications.

WCAP-16817-NP, Rev 2 Page 20 October 2007

4 REFERENCES

1. Code of Federal Regulations, 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants", January 2006.

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

December 1995.

3. ASME Boiler and Pressure Vessel Code Section XI, Appendix Q "Fracture Toughness Criteria for Protection against Failure", 1998 Edition with the 2000 Addenda.

4. Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials", U.S. Nuclear Regulatory Commission, Revision 2, May 1988.

5. U. S. Nuclear Regulatory Commission Standard Review Plan (SRP) 5.2.2, Overpressure Protection, Revision 3, March 2007.

6. Not used.

7. Florida Power and Light Report, PSL-ENG-SESJ-05-03 1, Rev. 2, "Engineering Evaluation, Input Parameters for 55 EFPY Pressure Temperature (P-T) Limits and LTOP Requirements, St. Lucie Nuclear Plant Unit 2", December 20, 2006.

8. Not used.

9. Not used.

10. St. Lucie Plant Unit No. 2 Technical Specifications, Amendment 146, June 19, 2006.

11. Not used.

12. Not used.

13. ASME Boiler and Pressure Vessel Code Section XI, Appendix A, "Analysis of Flaws", 1998 Edition with the 2000 Addenda.

14. Not used.

15. ASME Boiler and Pressure Vessel Code Section III, Appendix I, "Design Stress Intensity Values, Allowable Stresses, Material Properties, and Fatigue Design Curves", 1989 Edition.

16. Not used.

17. Not used.

WCAP-16817-NP, Rev 2 Page 21 October 2007

18. J. Heliot, R.C. Labbens, and Pellisser-Tanon, "Semi-Elliptical Cracks in a Cylinder Subjected to Stress Gradients", ASTM Special Technical Publication 677, August 1979.

19. Not used.

20. Westinghouse Report, CE-NPSD-683-A Task- 1174, Rev. 06, "Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Requirements from the Technical Specifications", April 2001.

WCAP- 16817-NP, Rev 2 Page 22 October 2007

TABLE 1 ST. LUCIE UNIT 2 REACTOR VESSEL BELTLINE MATERIALS 55 EFPY P-T/LTOP Embrittlement Material input data (Limiting Values Enlarged and in Bold) 55 EFP-Y PEAK 55 YEAR 55 YEAR CALC, TABLE I=lA LEUINCE. El 19 tT 3/47 ART g, ART@

LOCATION ID u % CFC F :CF RTndi MARGIN i Mar* n MVelhio& njcrA32 FLUENCE FLUjENCE 1T:4 3/4 Lowerse s pl +3ate Position 1.1 f,14 116- G- "I 2 ý, 0 7 37.0 2C, I 241 i Sur. data) 4.56E+19 2.7E-12 E ,I9 P 11,C8 101 'F 91 'F Lower s-clilat 14,41IlB-2'A?-. 0.8 -. C 201, 2F 4 'F7 Position 1.1 4.5E+1 2Ž19 9 .8E18 11-OF Lowecr s,el pat 1,141 tO-3ý A.6 44.0ý, 20' F 34'1 Position 1 4.SBEil 22E19 lbO 1 1t IF 'F, S Itseoae'1~ 'A-84 K1,2 :.1 8 7-P 01ý

'F 34 'F Position 1.1 4.58Ee19 2.Ec1 98&rl 1&

f'F 137 OF its l~oat&~05Position 1-"2 IlSuive !lance data A-8490ý-2 C1,01 87690 30 IF 17 IF iSiv dxata 4.56E419 272 19 98'1S 1"9:'F 134 'F hi ciIse

-41." -~5~

01'08 Jf 24& s 1 160 'F 1135 IF I6-l2lteP-t5 4150-21 Oil 081 74 01, 24' Position 1.1ý 4.SBCE4 1O 2.72-19: QOE iE 12M 07 Lowe( rs~lIaal welds 10 14AHS 3.7 0.0 007 34 1 50 IF 15 0~ :PosiritýlTP 11 4,56E+19 2.727E-12 O8ý 1 4P 'F 40F I.%hell aeiai' welds (101- 1'4A6.Ci 82842 005 D f4 956'F P85,51 Position 1.1 4.SSE+ 19 2 3w.2 F §I qv I3z 61F' In, i sh11 taxial weld s.

(101-;24C~ R PAIR 638421; - Position 1.1 ONY8.37 0'I 50'ýF 801 InSur.,data 4.56E 09 7222:

41,9 aed, 1a 4' 0' hit ýo Loner girthvwe'ds 3.871 -Position 1.1 fl0l17i7 3P731'7 0.071 50' F no SUNvdata

(~F 4.58E+ 19 . 2.7 2 7 E- 412 GE88I*1! T _'F ______

WCAP-16817-NP, Rev 2 Page 23 October 2007

TABLE 2 ST. LUCIE UNIT 2 CONTROLLING MATERIALS AND THEIR ADJUSTED REFERENCE TEMPERATURES REACTOR MATERIAL INITIAL ART AT 55 EFPY, TF VESSEL PLATE LOCATION ID NO RTNDT OF 1/4T 3/4T Intermediate M-605-1 30 158 137 Intermediate M-605-2 10 160 135 WCAP-16817-NP, Rev 2 Page 24 October 2007

TABLE 3 St. Lucie Unit 2 Cooldown and Heatup Allowable Pressure 55 EFPY, Adjusted to Actual Pressurizer Pressure, APCF Cooldown Heatup RCS 100 F~hlr 75 F/hr. 50 F!hr 30 Ffhr 10. F/hr Isothermal RCS .50. F/Ir Temp P-All P-All P-All P-All P-All P-All Temp P-All P-.AllI

'- sia i ! K(ksia, (ks aYI (ksia) : (kiM iksia). (ksia) 0 0,308 0,381 * .461 0.529 0,600 0.637 :*'i:38 0 1i!!! 0,637.. 0-634 90 0.34,18 0o414 0.488 '0,552. 0,619 0.654 90 0.634 100 0,396 0.454 0.:20 0.579 0,642 0.675 1003 0.675 0.634.

110 0.458 .502.

0, 0.559 0.612 0-670. 0.701 110 0,70*1 0-634 120) 0.532 D.H52 0.607 0.653 0.704.*" 0.732 0.7i~iO!321:i~i 120... 0.6 41 130 0.621 0635 0 0,702 03746 0.771 130 0.771 0.657 140 0.731' 0.723 0.38 C0.7 63 0.797F 0.818 140 01.818 0,684 150 0&865 0.832 0.826 0.837 0:860 0.875 150 0.875. 0.721*...

160 0,945 0-945 0.933 0,927 0.936-K 0945 160 :0.945 0.770:;

170 11.031 1.031 1.031 1.031 1 029 1.031 170 1.031 0.833:

180 11,35 1-135 1*1 35 1,135. -135 1.135 180. 1*.135.. 0,912 190 1.:263 1-263 1.263 1.263 1.263 1263 190 1.263 .1.011 200- 1,419 1.41:_19 1.419 1,419 1.419 1.419 200 1.*419 1,132_

-210 1.609 1,609 11.609 1,609 1,609 1.609 210 1.609 1*281 220 1.841 1-841 1.*841 1.ý841 1,841 1.841 i;,i!1.841*,i*

220 1 1, -"11i:2

i!!* 1,464 230 2.125 2,125 2.125 2,125 2*125 2.125 230 2,125. 1.689 240* 2.472. 2,472 2.472 2,472 2.472 . '2*472 240 2*472 1,963 250 2.89695j 2,896P, 2.896'3 2.8968- 2.896 2.896 250.. 2.896 2.299 2*60' 3239 3-3343 13414 3,414 3,414 13414:: 260. 3.414 2.644

ý270,3 3.240 3,343 3.4450 3.537T .3.624 3.667 270 3.667 3.057 280 3,242 35343 3.450 3.537 3.624

- 31667 280 3.6 6-7. 3.562.

29o0 3,243 3.344, 3.450 13537 . 3.624 3,667 290 3.667 3.667 300 3.244 3,344 . 3.450 3.537 3.624 3.667_ 300 3,667 .3.667 WCAP-16817-NP, Rev 2 Page 25 October 2007

TABLE 4 St. Lucie Unit 2 Cooldown and Heatup Allowable Pressure 55 EFPY, Adjusted to Indicated Pressurizer Pressure, IPCF Cooldown Heatup RCS 100 Fflhr 75 F/hr "50 Fhr 30 F/hr 10 F/hr IIsotherma -.

RCS Iso~thermall 50 F/hr Tem p P-AN P-AUl P-All~ P-All P-All P-All Temp P,All P-All 80 0.258 0.03377 0.423 0.494 0.568 0,606 .80 : :,:

0.606: i!*:

90 0,291 0,364 0,444 0.512 0-583 ý0620 90 0.620.

100 0,331 0.397 0.471 0.535 0,602 0,637,. 100 110 0.381 0A37 0,503 0-562 0.625 0.658 110 0.658 0,617 0.617 120 0,441 0.485 0,542 0,595 0,653 0.684 120- 0.684 130 0,5It55 0E545 0,59D 0.636 0,687 0-715 130 0.715. 0.624 140 0.604 0.618 0.649 0.685 0.729 0.754 '140. 0.754 0.640 150 0,714 0.706 0.721 0.746 0,780 0.801' 150 0.801 0.667 160 0.848 0M815 0,809 *0820 0,843 0.858 160 .0.858 0.704 170C 0928 0.928 0.916 0.910 0.919 . 0,928 ,170 0.928 0.753 180 11014 1,J014 1.014 1.014 1.012 1.014 180 1*014 0*.816::

1901.118 1.118 1.t18 .1.118 1.118 1.118' 190* .1118 200 1,246 1.246 1.246 1.246 1.246 1.246. 200- 1,246 0.994 2 10) 1.402 1.402 1.402 1.402 1.402 1.402 210 1,402 1.11 220 1.592 1.592 1.592 1,592 1-592 1.592 -220 1.592 1-264 230 1.82414 1.824 1.824 1.824 1.824 1.824 230 .1,824..

240, 2.108 2.108 2,108 2.108 2.10 2.108 240 2.108 1,672 250 ",2,4552.455 2*.455 2,455 2.455ý:* 2.455 250: 2.455. ..1.946 260" Z879 2.879 2.879 2.879 2.879:1 ::2.879 260 2,879. 2.282:.

270- 3"222 I3326 3397 3.397 3,397 3.397 270. 3-397 2.627:

280 3.223 13326 3.433 3.520 3.607i :3650C '1 280. 3.650 3.040 290 3.225 -3.326:.: 3.433 3.520 3.607 3.650 290 3*650 3,545 300 3.226 3.327 3.433 3.520 13607 3.650 300 :3.650 :

31650 WCAP-16817-NP, Rev 2 Page 26 October 2007

TABLE 5 St. Lucie Unit 2 Hydrostatic Test Pressure-Temperature Limit Data Actual Pressurizer, Indicated Pressurizer, Beltline APCF IPCF Conditions Conditions Conditions RCS -Isothermal RCS Isothermal CS Isothermnal Temp. P-All Temp P-all Temp P-all

{ksia) (ksia)

.0.869 80 0.?,3 3 0,9 4 90 &0892 . 0.852 1)00 0,375 0.9-54 110

0. 9037 120 0.996 120- 110 0.93 579 130 I1.043 130 120 12;3. .

140. 1.031 140 150 1.26E2.

S60 1.280 1860 1.170 .6n. 1.,3 55 170 1.394 170' 1,263 170 1._4._-9 1.-377. 1.'6C08 10 _1.6

_200n-2475 210 '2148. 2107 2_30 2140- 3-316 240 2),P37 227 250C -3.881 2501 9343 260C 4.,571 .. '3.64 2-70 4.909. 4,554 3. 9N6 28 20 4.89ý2 4.646 _

4_8ý92 300 4.909 3300 4.892 WCAP- 16817-NP, Rev 2 Page 27 October 2007

TABLE 6 MAXIMUM TRANSIENT PRESSURES Maximum Pressure PORV Mitigation SDCRV Mitigation Transient Setpoint Setpoint Setpoint 470 psia 490 psia 350 psia RCP Start 521.6 psia 541.6 psia 368.2 psia 2 HPSI + 3 Charging Pumps 549 psia 549 psia 384 psia 1 HPSI + 3 Charging Pumps 510 psia Not Analyzed 366 psia WCAP-16817-NP, Rev 2 Page 28 October 2007

TABLE 7

SUMMARY

OF CONTROLLING PRESSURES Controlling Pressure PORV Mitigation SDCRV Mitigation RCS Temperature Setpoint Setpoint Setpoint 470 psia 490 psia 350 psia

< 200°F 521.6 psia 541.6 psia 368.2 psia

> 200°F 549 psia 549 psia 384 psia WCAP-16817-NP, Rev 2 Page 29 October 2007

TABLE 8 LTOP REQUIREMENTS, 55 EFPY LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE Operating Cold Leg Temperature, OF Period, EFPY During Heatup During Cooldown

<55 <246 < 224 MINIMUM COLD LEG TEMPERATURE FOR PORV USE FOR LTOP Operating Cold Leg Temperature, 'F Period, EFPY During Heatup During Cooldown

<55 80 132 MAXIMUM ALLOWABLE HEATUP RATES 50-F /hr, at all temperatures MAXIMUM ALLOWABLE COOLDOWN RATES 75-F /hr, at T,< 105 0F 100-F /hr, at T,,> 105 0F A RCP shall not be started with two idle loops,.unless the secondary water temperature of each steam generator is less than 40'F above each of the RCS cold leg temperatures. (This is an existing limitation).

Applicability: During heatup, at T, < 246°F During cooldown, at T, < 224°F

One HPSI pump shall be rendered inoperable prior to entering Mode 5 during cooldown. (This is an existing limitation).

WCAP-16817-NP, Rev 2 Page 30 October 2007

TABLE 9 Heatup Allowabl~e Pressures, Uncorrected RCS Isothermal 50 tF!hr Temp P-All P-All (ksia) (Ksia)l 80D 0312 0ý709 90 0.729 0.709 100 0.750 0.709 1i10 0,776! 0.709 1,20 0.807 0.716 1303 0,546 0,732 140 ".893 0.759 1503 0.950 0,796 160 1.020 1 0,845 170 .. 0 - 0.908 1I .1 0.987 200 1,494 1,207

______ 1,6E4 1.356 220 1.916 1-S39 230 2.200 1.ý764 240 .2.547 2038 1

250 2,971 - 2.374 2760 3.4&9 2.719.

2:70 3.742 3.132 2810 3.742 3.637 2903 3.742 3.742 300 3.742 13.742 WCAP-16817-NP, Rev 2 Page 31 October 2007

TABLE 10 Coldown Allowable Preýsures, U ncorrected RCS~ 100 Ffh r 75 F~hr 50 F/,hr 30 F/,hr 10 F/hr Isothermal, Temp P-All P-All P-All P-All P-All P-All

'Fksia) f(ksia) f{kc i ) (ksia) (Ksi a !ksia) 80 0.383 C,456 a.536- 0.604 0-675 0,712' 90 0.423 DB049 0,563 0.627 0-694 0.729 100 . 473 0.529 C" 0,595 0.654 0.717 0.750 110 ' 0.533 0.834 0.577 0687 0.745 0.7769 120 . 0-1,60.807 7-7 0.8*0.

68- 728 0.779 0,807 130 E096 01 0 0.741 0777 10.82 0,8,46 140 01806*i 0.796 (3 E 3 0.838 0.872 0.893 I150 0.9404 D 0.907 .*90t1 0.912 0. 935I 0.950 16 1.02 1.2 .008 1,002 1.011 1,020 170 1.106 11,06 I. :1.106 106 1.i04 11.06 180 1.210 1-210 1.210 1C210 1.210' 1.210 1 190 1.338. 1,338 1. 33 1 338 1.-335 1.333 200 1.4.94 1.4.94 1,49L 1.494 1.494 1.94 2-10 ~1.F64 1.6884 1.688. 1.68,4 1.ý684 '1,884 220 1.91i6 1.916 1.91E 1.916 1.916 1.916 1230 2.200 2.20' 22.2002.200 2.2200 .,. 2,2003 240 .5*.7 2.547 2.54.7 2.2547 2,54.7 2.547 250 2.971 2.971 2,971 2.9 711 2.971 2.971 260 3.3t14 3 4186 3.48. 3,489 3.489 3.489 70 3.31i 5 3.418 3.525 3,612 3.699 33742 33117

,280 3.418 3, 5 2 3.12:

.525 6 3.699- 3.742 290 3.3183 3.419 3.525 3.612 31699 3.742 300 331,9- 13,19 3.525. 3.612:

2 ,3.6991 3.742 WCAP- 16817-NP, Rev 2 Page 32 October 2007

FIGURE 1 ST. LUCIE UNIT 2 COOLDOWN P-T LIMITS 55 EFPY, APCF Adjusted to Pressurizer Pressure Lowes Ser:ice t IIi 2 D~

's 10i~iilTemip ,160-

, i i iii, , i,

H i t I*

.&=.;. ..

00 RV Flange M1r4 Temperature Ii Requirements 170 F I:5'

:T*: I * :;i*

I ]- Core Criticýality Based on Inservice

I Hydrostatic Test Temperature (2 175;F) for Specified RTNOT I*,. * :* :, ;.L @ :* , '. = :' :. . :: .

0 50*i~!ii! 100 1S0 : 200 250 300 35ýJ 400 RCS Temperature ('F)

WCAP-16817-NP, Rev 2 Page 33 October 2007

FIGURE 2 ST. LUCIE UNIT 2 COOLDOWN P-T LIMITS 55 EFPY, IPCF Adjusted to Indicated Pressurizer Pressure

!3))2:

2:-5 Lowest Service I~

2.0 JTemp,l160'F]

2iih~;,,

i

-Its

_eprtr-l'qIl ill i

,;i!!. *. i . ':: ::::.: : 7!!*, : : : : :: ::* >::]!,. !, !:*i**u!*Ui!!!}**!,*,' :*W, *, : :G .* *iu *

L : : * ' :J *'i : *I' 150 :: !:ii 1ii i::ii: :! [ : :? i :i:i: I i

i~ : *: i i iiii :F i ii:i:i:iiiii* :1I : iii: ~

iii: i:i::: : :T *i : : ":: i : i: i*

TempeureicTestuemr--entur 10-Tempos.ti Tes Temperaturerm 5443 psi~a 0 150D 200 25-0 300 3H0 4X00 RC.S Temperature ('

WCAP-16817-NP, Rev 2 Page 34 October 2007

FIGURE 3 ST. LUCIE UNIT 2 HEATUP P-T LIMITS 55 EFPY, APCF Adjusted to Pressurizer Pressure WCAP-16817-NP, Rev 2 Page 35 October 2007

FIGURE 4 ST. LUCIE UNIT 2 HEATUP P-T LIMITS 55 EFPY, IPCF Adjusted to Indicated Pressurizer Pressure RLecommendcd H-ýtfostatit' es*

Lowest Servi *4 15 Temp., leD RV Fkr,,ge Mmý Te'nper1r1~~

Requ;irvmcne~s -7'CFI "I

0 0.... ... ... ["!!!

.10:

ý2 cc152 2IC 250u 350D 400 R'STciem rzire~ir-)

WCAP-16817-NP, Rev 2 Page 36 October 2007

FIGURE 5 ST. LUCIE 2, ENERGY ADDITION TRANSIENT w/ PORV, PSET= 4 9 0 PSIA RCS Pressure vs. Time RSG, PORV setpoint 490 psia 600 4:~.500 Ld i'450 S400 350 TIME, SEC.

WCAP-16817-NP, Rev 2 Page 37 October 2007

FIGURE 6 ST. LUCIE 2, ENERGY ADDITION TRANSIENT w/ SDCRV, PSET= 3 5 0 PSIA SDC Relief Inlet Pressure vs. Time RSG, SDC Relief Valve setpoint = 350 psia 370 1 360 U) 350

a. ... 1:,!**

i!'lii~*I :*lii'*,i H!:':

: !:l ':iiiiliil,,li:** il l*:l:

340 330 ILL 7

LU 320 310 300 0 2 4 6 8 10 12 TIME, SEC.

WCAP-16817-NP, Rev 2 Page 38 October 2007

FIGURE 7 ST. LUCIE UNIT 2 LTOP MASS ADDITION TRANSIENT PORV Flowrate Vs Pressure Case 1 Case2 500 300 800 1000O 1200 1400 1 ýr Flowrate (gpm)

Case 1 - Single HPSI + 3 Charging Pumps Case 2 - Both HPSI + 3 Charging Pumps WCAP- I 6817-NP, Rev 2 Page 39 October 2007

FIGURE 8 ST. LUCIE UNIT 2 LTOP MASS ADDITION 1 00020 1600F, lowrate (gpo Case 1 - Single HPSI + 3 Charging Pumps Case 2 - Both HPSI + 3 Charging Pumps WCAP-16817-NP, Rev 2 Page 40 October 2007

APPENDIX A TECHNICAL SPECIFICATION FIGURES APPENDIX A Technical Specification Figures WCAP-16817-NP, Rev 2 Page 41 October 2007

APPENDIX A TECHNICAL SPECIFICATION FIGURES FIGURE 3.4-2 ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR 55 EFPY, HEATUP, CORE CRITICAL, AND INSERVICE TEST 2500

- * --4 -I

-- III-I-

- *--* F--I FI - -F I -I-- I -F - -- -k - I I-11 I - I--FI J -- I -- F- '-- I-- I-- i-+- I-I --

11i 1 1 1 7 7 -/ r I F III - FI 1 1-I II - 11I 111 I 11 1T 11 11 T11 11 w-i*--rI--I-I T*--

-I--I-F- t-l---I-I-- -r'l I- F 1-l-l "r7 -I F T * -I-I-- " --I -FT -I--

.I I I I 1I l It' I . .I - 1- 1 I I1 I, I 1I1I1I

'I 1111III

'1S 7 1' -4 7 1' 1_SO ITT** 4 HJ7ML-L *,i_ -7 T,

-,I*

-' **-*- 1 .. .. .. . ...... _

_L 1_'L_

1....

.. .... L.. J.L..

... L..U J. LL....

Lu 2000 FI mI rFTI 1 Im I T II TI~I -II TI

-I I _ _- - - - -I - - I-I7NS '1BTNI'EFr- 7-Y1STFcTT-'1i F F T T- IF,"-

z LJ_ _LL 1 J__LL r

_ILL J . J

- HH*

_ L _L 4 -H IJ

-I-I-AJLLLAIL HI-H---+-*-I-I-- -f-I-I-HI-+

ALI _L I I

-I-I--

L I 1 1 I I I II I I I I I I I I.

w 1.Ev I-CEHYDROSTATIC TEST- - r -- ---l-I II II I II I I 17I I ' E 1I 1 L L J I L I J I L 1 - IL -I - - L I L W 1500 it i ii ii i i

i I T I I I I I I I I I I I I I I I I i I I N -1 - - *1--*1 IAL- L J" I--LL'-

IL I I -L - A-4 - OREGRT&L*'IIHI--1

+ _ 1 I L - Il L

-TI I -FI-,- - I- I -I- I II " - I L I I I I II II I I *E I I- I r - -1 LOWEST r-, - ,- r i" rr rrF F "T *-I-CrT l -I-I-C 4.

" 1 -I-I- FT I--I--F H

-I -I-HI-H+@-I-I-Hk 4.4-I-I-HI--I4-I-I-I-C, _£ J SBW ICE J__ LL I.A. _ILL J I L w T- TEMPERATURE I- -r

-t I-r -i IL- " I- r- T 1- -I r 7 - r--

- T - I- - F A J1l-160OF 21 4-l L L -4 A - +

0~ IF_1 1-1 - 1 I I I I I I I I I I I I I I

" I I I I _ _II w

w C_Fl 1T CI - A1 -i F T 1-1 r " --- T l- T r L I- L L_ -I- L - A -2 I- L I 1 - -2I- L - A _II L LA-t_ UI L I H HH - I H 4 II_ _I ST, I

170 mlv~ATR 2FFFm miJI III LL 1

544II PSI'A* II 7L12 I I I AI IICF I i ALLOWABLE HEATUP RATES

- F -I - I- L F -I- - -i i 170 0.F7 -_

z I IL L4 I I L II I~

pA TE, °F/H R TEMP. LIMIT 'F S500 50 AT ALL TEMPERATURES IULLAJL iJ 4 7~F~ MIN. BOLTUP TEMP.

I80 0F" IIF L I II-r-I TT-I--1 I I I li I

At I- I

-I-:- I-

: I ff -I-II I I- L

+- I A-I -I

!- I Lj IJ-- L II I I-H -I- - I- I IT I F T AT"

- -T -

F1HI--- . .- -.t H. ". I.

0 0 100 200 300 400 500 0

Tc - INDICATED REACTOR COOLANT SYSTEM TEMPERATURE, F

- Includes Instrument Uncertainty WCAP-16817-NP, Rev 2 Page 42 October 2007

APPENDIX A TECHNICAL SPECIFICATION FIGURES FIGURE 3.4-3 ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR 55 EFPY, COOLDOWN AND INSERVICE TEST 2500 I' . "T4 4- F .- .4--- .L -t-i-Ll---W-i-i- T44-1-i.L-TT-t-.i-I II I&' I I III L ': I I I I I 111 1

_111 "7 1I -I T *- I- T *-I -i T T ji 3-I -IF I til L T -I- Ilii 1-, 1- 1 ij71- --i- 1 1- T 1 1- T I II-r-

I -I I I 7 I I I I I FT I I T I I FTI if I I -I-I rrIr I TI FTI JI I 2000 D .L J -J-l_ L 1 -1__I - 1 I L ý -_ I J L--LLI J _-:-L -I J _ LLL J _ I_ J- __I t L -L J -1 L 1I1I I I1 lilt'Iii IiI I IiIiIIII iI IiIi Ito Ij I J 1I L 1 -1 I I I 1I J l LI I I I I I I I I-I I J 1--- L- J--I- L-J-J-i--

r -I-t- 16 111 ° i- F J--l--l--

77q 1- - --4-i-t--

TI'l3 r -I* -I- -* T 1

-- l-I--

T 7 --Ii I l I--

TF T

l----I

-I-i I-"-I- i F T-T T7-I-rF

--I- L i ii --- i i i/

a 4_i__-

Ji Titiii7-1i-i-----I-Liil L F

7- 1-

---l i/ T III--

T i 7-i- I rII T T--II-r"i-F II I---

i r lO EISIi I Coo I Iii I 1

T I-iii I r T 1

I 1

-I- Iii 1 1 1 1 6 I1 11F 1 1 1 7 01 -1 U 1500 LOWEST N "T T -- IT'L

I*--+*--- U TO 1OOL OWN T-/IHR--IT -IIr WW L-ti- -i - .LL. -- i. W- -- W-- ,

UJI r- t- rr--+ i-r i--- r-I- i-/ i-- - --i- F-i-i- F -1 -i-i- Ft7 -i-I-F -i -i-i r T-I -I- rF w I - LAI 1. L I I .I.L1I II II L li i i t'1 1-1 1

- i I 1 (L t - i-i- +-- - - 1---- --

I--- i-i-i o 1000 COLOWE N I I2'TI_I I I tI_ OLLOW A AB RATES w

TEMF/ATURE 1 - UTO--0oFi-Rk-,-i-

- UP.FT170'FI-I I II I-L RATE, 0 F/H R II TEMP. LIMIT 0OF I1 --- i-Fr - 1-rii-- ti-i-i-i- -1 t- i-[-r 100 >!105 z 0 .OW L D .N -I -LI1 . L L J 75 < 105 UP TO75F/HR* IAI I 4IpA*II I I t i I F 500

(- T II LI t I ...... ...... .. .* -- '- l-i-- t I- I--

-l-T MINIMUM-BOLTUPTE M-----Im-r T-- ---------------- UR--80---- *Ti-i-II *--I

_iJJ--i- -I:-- -- t t1 I-I-i L -I--L F Ij I ', I _ _ II tLL -I 1- - I L [ "Il "L "i I .

rT -l 1 rriT i -- T T TT' II IT -l iT7FFT TT -II I' I F I I I I I: I it -i-I-- ti-:---- -'t if-l-i-i- tif-i-i- F'- II-ii i-- -- Ft if-i--l-tim-i -I ti-i-i-- rt *I-lilt , , t ill..-- ... -- . -4.. 4 .-1 0 0 100 200 300 400 500 0

Tc - INDICATED REACTOR COOLANT SYSTEM TEMPERATURE, F

- Includes Instrument Uncertainty WCAP-16817-NP, Rev 2 Page 43 October 2007

WCAP-16817-NP, Rev 2 Non-Proprietary, Class 3 SWestinghouse Westinghouse Electric Company, LLC 20 International Drive Windsor, Connecticut 06095

L-2007-198, Proposed License Amendment Update PT Curve and LTOP for 55 EFPY

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