ML070680370

Pressure and Temperature Limits Reports (Ptlrs), Revision 4

ML070680370
Person / Time
Site:	Braidwood
Issue date:	03/08/2007
From:	Coutu T Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, NRC/NRR/ADRO
References
BW070019, TAC MC8693, TAC MC8694, TAC MC8695, TAC MC8696
Download: ML070680370 (52)

Text

____ __ _____ Exd_n E~hnGeneration Conpany, LC N uci ear

~raiawood St~t on c

March 8, 2007 SWO7001 9 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and 50-457

Subject:

Pressure and Temperature Limits Reports (PTLRs), Revision 4, Braidwood Station, Units 1 and 2

References:

(1) Letter from K. FL Jury (Exelon Generation Company, LLC) to U. S. NRC, License Amendment Request Regarding Reactor Coolant System Pressure and Temperature Limits Report and Request for Exemption from 10 CFR 50.60, Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation, dated October 3, 2005 (2) Letter from R. F. Kuntz (U. S. NRC) to C. M. Crane, Byron Station, Unit Nos.

1 and 2, and Braidwood Station, Unit Nos. 1 and 2 issuance of Amendments Re: Reactor Coolant System Pressure and Temperature Limits Report (TAC Nos. MC8693, MC8694, MC8695, MC8696), dated November 27, 2006 The purpose of this letter is to transmit the Pressure and Temperature Limits Reports (PTLRs) for Braidwood Station, Units 1 and 2 in accordance with Technical Specification (TS) 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR). The Braidwood Unit 1 and Unit 2 PTLRs were recently revised to extend the applicability of the heatup and cooldown curves to 32 effective full power years (EFPY). The methodology for developing the revised PTLRs is consistent with a recently approved method added to Braidwood IS 5.6.6 (Reference 2).

U. S. Nuclear Regulatory Commission Page 2 March 8, 2007 Please direct any questions you may have regarding this matter to Mr. Dale Ambler, Regulatory Assurance Manager, at (815) 417-2800.

Thomas Coutu Site Vice President Braidwood Station Attachments: 1. Braidwood Unit 1 Pressure and Temperature Limits Report (PTLR), Revision 4

2. Braidwood Unit 2 Pressure and Temperature Limits Report (PTLR), Revision 4

ATTACHMENT 1 Braidwood Unit 1 Pressure and Temperature Limits Report (PTLR), Revision 4

BRAIDWOOD UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

Revision 4

BRAIDWOOD UNIT 1 -

PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents Section Page 1.0 Introduction 1 2.0 RCS Pressure and Temperature Limits 1 2.1 RCS Pressure and Temperature (PiT) Limits (LCO 3.4.3) 1 3.0 Low Temperature Over Pressure Protection and Boliup 7 3.1 LTOP System Setpoints (LCO 3.4.12) 7 3.2 LTOP Enable Temperature 7 3.3 Reactor Vessel Boliup Temperature (Non-Technical Specification) 7 4.0 Reactor Vessel Material Surveillance Program 10 5.0 Supplemental Data Tables 12 6.0 References 19

BRAIDWOOl) UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT List of Figures Figure Page 2.1 Braidwood Unit I Reactor Coolant System Heatup Limitations (Heatup 3 Rate of l00°F/hr)Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 2.2 Braidwood Unit 1 Reactor Coolant System Cooldown Limitations 4 (Cooldown Rates of 0, 25, 50, and 100°F/hr)Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 3.1 Braidwood Unit 1 Nominal PORV Setpoints for the Low Temperature 8 Overpressure Protection (LTOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) ft

BRAIDWOOD UNIT 1 -

PRESSURE AND TEMPERATURE LIMITS REPORT List of Tables Table Page 2.la Braidwood Unit I Heatup Data Points at 32 EFPY (Without 5 Margins for Instrumentation Errors) 2.lb Braidwood Unit 1 Cooldown Data Points at 32 EFPY (Without 6 Margins for Instrumentation Effors) 3.1 Data Points for Braidwood Unit 1 Nominal PORV Setpoints for 9 the LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) 4.1 Braidwood Unit 1 Capsule Withdrawal Schedule 11 5.1 Braidwood Unit 1 Calculation ofChemistry Factors Using 13 Surveillance Capsule Data 5,2 Braidwood Unit 1 Reactor Vessel Material Properties 14 5.3 Summary ofBraidwood Unit I Adjusted Reference Temperatures 15 (ARTs) at 1/4T and 3/4T Locations for 32 EFPY 5.4 Braidwood Unit 1 Calculation of Adjusted Reference 16 Temperatures (ARTs) at 32 EFPY at the Limiting Reactor Vessel Material, Nozzle Shell Forging 5P-7016 5.5 RTprs Calculation for Braidwood Unit 1 Beltline Region 17 Materials at EOL (32 EFPY) 5.6 RTp~5Calculation for Braidwood Unit 1 Beltline Region 18 Materials at Life Extension (48 EFPY)

BRAIDWOOD UNIT I -

PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 Introduction This Pressure and Temperature Limits Report (PTLR) for Braidwood Unit I has been prepared in accordance with the requirements ofBraidwood Technical Specification (TS) 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR).

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

The Technical Specifications (TS) addressed in this report are listed below:

LCO 3.4.3 RCS Pressure and Temperature (PiT) Limits; and LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System.

2.0 RCS Pressure and Temperature Limits The PTLR limits for Braidwood Unit 1 were developed using a methodology specified in the Technical Specifications. The methodology listed in WCAP-14040-NP-A, Revision 2 (Reference 1) was used with the following exceptions:

a) Optional use ofASME Code Section XI, Appendix G, Article G-2000, 1996 Addenda, b) Use of ASME Code Case N-640, Alternative Reference Fracture Toughness for Development ofP-T Limit Curves,Section XI, Division 1, c) Use of ASME Code Case N-588, Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessel,Section XI, Division 1, and d) Elimination ofthe flange requirements documented in WCAP-16143-~P.

These exceptions to the methodology in WCAP 14040-NP-A, Revision 2 have been reviewed and accepted by the NRC in References 2, 8, 9 and 10.

WCAP 15364, Revision 2 (Reference 11), provides the basis for the Braidwood Unit 1 P/T curves, along with the best estimate chemical compositions, fluence projections and adjusted reference temperatures used to determine these limits. WCAP- I 6143-P.

Reference 12. documents the technical basis for the elimination of the flange requirements.

2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3) 2.1.1 The RCS temperature rate-of-change limits defined in WCAP- 15364, Revision 2 (Reference Il) are:

a. A maximum heatup of 100°Fin any i-hour period,

b. A maximum cooldown of 100°Fin any 1-hourperiod, and

BRAIDWOOD UNIT 1 -

PRESSURE AND TEMPERATURE LIMITS REPORT

c. A maximum temperature change ofless than or equal to 10°Fin any i-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

2.1.2 The RCS PIT limits forheatup, inservice hydrostatic and leak testing, and criticality are specified by Figure 2.1 and Table 2. ia. The RCS P/T limits for cooldown are shown in Figure 2.2 and Table 2.lb. These limits are defined in WCAP- 15364, Revision 2 (Reference II). Consistent with the methodology described in Reference 1 and exceptions noted in Section 2.0, the RCS P/T limits for heatup and cooldown shown in Figures 2.1 and 2.2 are provided without margins for instrument error. These limits were developed using ASME Boiler and Pressure Vessel Code Section XL Appendix G, Article G2000, 1996 Addenda.

The criticality limit curve specifies pressure-temperature limits forcore operation to provide additional margin during actual power production as specified in 10 CFR 50, Appendix G.

The P/T limits for core operation (except for low power physics testing) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40°Fhigher than the minimum permissible temperature in the corresponding P/T curve for heatup and cooldown.

____ __ _____ Exd_n E~hnGeneration Conpany, LC N uci ear

~raiawood St~t on c

March 8, 2007 SWO7001 9 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and 50-457

Subject:

Pressure and Temperature Limits Reports (PTLRs), Revision 4, Braidwood Station, Units 1 and 2

References:

(1) Letter from K. FL Jury (Exelon Generation Company, LLC) to U. S. NRC, License Amendment Request Regarding Reactor Coolant System Pressure and Temperature Limits Report and Request for Exemption from 10 CFR 50.60, Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation, dated October 3, 2005 (2) Letter from R. F. Kuntz (U. S. NRC) to C. M. Crane, Byron Station, Unit Nos.

1 and 2, and Braidwood Station, Unit Nos. 1 and 2 issuance of Amendments Re: Reactor Coolant System Pressure and Temperature Limits Report (TAC Nos. MC8693, MC8694, MC8695, MC8696), dated November 27, 2006 The purpose of this letter is to transmit the Pressure and Temperature Limits Reports (PTLRs) for Braidwood Station, Units 1 and 2 in accordance with Technical Specification (TS) 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR). The Braidwood Unit 1 and Unit 2 PTLRs were recently revised to extend the applicability of the heatup and cooldown curves to 32 effective full power years (EFPY). The methodology for developing the revised PTLRs is consistent with a recently approved method added to Braidwood IS 5.6.6 (Reference 2).

U. S. Nuclear Regulatory Commission Page 2 March 8, 2007 Please direct any questions you may have regarding this matter to Mr. Dale Ambler, Regulatory Assurance Manager, at (815) 417-2800.

Thomas Coutu Site Vice President Braidwood Station Attachments: 1. Braidwood Unit 1 Pressure and Temperature Limits Report (PTLR), Revision 4

2. Braidwood Unit 2 Pressure and Temperature Limits Report (PTLR), Revision 4

ATTACHMENT 1 Braidwood Unit 1 Pressure and Temperature Limits Report (PTLR), Revision 4

BRAIDWOOD UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

Revision 4

BRAIDWOOD UNIT 1 -

PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents Section Page 1.0 Introduction 1 2.0 RCS Pressure and Temperature Limits 1 2.1 RCS Pressure and Temperature (PiT) Limits (LCO 3.4.3) 1 3.0 Low Temperature Over Pressure Protection and Boliup 7 3.1 LTOP System Setpoints (LCO 3.4.12) 7 3.2 LTOP Enable Temperature 7 3.3 Reactor Vessel Boliup Temperature (Non-Technical Specification) 7 4.0 Reactor Vessel Material Surveillance Program 10 5.0 Supplemental Data Tables 12 6.0 References 19

BRAIDWOOl) UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT List of Figures Figure Page 2.1 Braidwood Unit I Reactor Coolant System Heatup Limitations (Heatup 3 Rate of l00°F/hr)Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 2.2 Braidwood Unit 1 Reactor Coolant System Cooldown Limitations 4 (Cooldown Rates of 0, 25, 50, and 100°F/hr)Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 3.1 Braidwood Unit 1 Nominal PORV Setpoints for the Low Temperature 8 Overpressure Protection (LTOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) ft

BRAIDWOOD UNIT 1 -

PRESSURE AND TEMPERATURE LIMITS REPORT List of Tables Table Page 2.la Braidwood Unit I Heatup Data Points at 32 EFPY (Without 5 Margins for Instrumentation Errors) 2.lb Braidwood Unit 1 Cooldown Data Points at 32 EFPY (Without 6 Margins for Instrumentation Effors) 3.1 Data Points for Braidwood Unit 1 Nominal PORV Setpoints for 9 the LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) 4.1 Braidwood Unit 1 Capsule Withdrawal Schedule 11 5.1 Braidwood Unit 1 Calculation ofChemistry Factors Using 13 Surveillance Capsule Data 5,2 Braidwood Unit 1 Reactor Vessel Material Properties 14 5.3 Summary ofBraidwood Unit I Adjusted Reference Temperatures 15 (ARTs) at 1/4T and 3/4T Locations for 32 EFPY 5.4 Braidwood Unit 1 Calculation of Adjusted Reference 16 Temperatures (ARTs) at 32 EFPY at the Limiting Reactor Vessel Material, Nozzle Shell Forging 5P-7016 5.5 RTprs Calculation for Braidwood Unit 1 Beltline Region 17 Materials at EOL (32 EFPY) 5.6 RTp~5Calculation for Braidwood Unit 1 Beltline Region 18 Materials at Life Extension (48 EFPY)

BRAIDWOOD UNIT I -

PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 Introduction This Pressure and Temperature Limits Report (PTLR) for Braidwood Unit I has been prepared in accordance with the requirements ofBraidwood Technical Specification (TS) 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR).

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

The Technical Specifications (TS) addressed in this report are listed below:

LCO 3.4.3 RCS Pressure and Temperature (PiT) Limits; and LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System.

2.0 RCS Pressure and Temperature Limits The PTLR limits for Braidwood Unit 1 were developed using a methodology specified in the Technical Specifications. The methodology listed in WCAP-14040-NP-A, Revision 2 (Reference 1) was used with the following exceptions:

a) Optional use ofASME Code Section XI, Appendix G, Article G-2000, 1996 Addenda, b) Use of ASME Code Case N-640, Alternative Reference Fracture Toughness for Development ofP-T Limit Curves,Section XI, Division 1, c) Use of ASME Code Case N-588, Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessel,Section XI, Division 1, and d) Elimination ofthe flange requirements documented in WCAP-16143-~P.

These exceptions to the methodology in WCAP 14040-NP-A, Revision 2 have been reviewed and accepted by the NRC in References 2, 8, 9 and 10.

WCAP 15364, Revision 2 (Reference 11), provides the basis for the Braidwood Unit 1 P/T curves, along with the best estimate chemical compositions, fluence projections and adjusted reference temperatures used to determine these limits. WCAP- I 6143-P.

Reference 12. documents the technical basis for the elimination of the flange requirements.

2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3) 2.1.1 The RCS temperature rate-of-change limits defined in WCAP- 15364, Revision 2 (Reference Il) are:

a. A maximum heatup of 100°Fin any i-hour period,

b. A maximum cooldown of 100°Fin any 1-hourperiod, and

BRAIDWOOD UNIT 1 -

PRESSURE AND TEMPERATURE LIMITS REPORT

c. A maximum temperature change ofless than or equal to 10°Fin any i-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

2.1.2 The RCS PIT limits forheatup, inservice hydrostatic and leak testing, and criticality are specified by Figure 2.1 and Table 2. ia. The RCS P/T limits for cooldown are shown in Figure 2.2 and Table 2.lb. These limits are defined in WCAP- 15364, Revision 2 (Reference II). Consistent with the methodology described in Reference 1 and exceptions noted in Section 2.0, the RCS P/T limits for heatup and cooldown shown in Figures 2.1 and 2.2 are provided without margins for instrument error. These limits were developed using ASME Boiler and Pressure Vessel Code Section XL Appendix G, Article G2000, 1996 Addenda.

The criticality limit curve specifies pressure-temperature limits forcore operation to provide additional margin during actual power production as specified in 10 CFR 50, Appendix G.

The P/T limits for core operation (except for low power physics testing) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40°Fhigher than the minimum permissible temperature in the corresponding P/T curve for heatup and cooldown.

BRAIDWOOD UNIT 1 -

PRESSURE AND TEMPERATURE LIMITS REPORT

~RASI I LLMJTING MATERIAL~NOZZLE SHELL FORGING 5P~7O~6 UMflING ART VALUES AT 32 EFPY: 1/41, 4g~F 3/4T, 35~F 2500 Vers~on~5,1 Run~29844J 2250 2000 _________

I I Unacceptabiij I Acceptable

[9peratlon L~°~I 1750 1500 - Heatup Rate.

l000eg.

1250 I

I 1000 750 Criticality Limit based on Inservice hydrostatic test temperature (108 F~)for the 500 sarvtceperiodupto3zEFPy 250 Boftup Temp 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 2.1 Braidwood Unit 1 Reactor Coolant System fleatup Limitations (Heatup Rate of 100°F/hr)

Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 3

BRAIDWOOD UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: NOZZLE FORGING SP-7016 LIMITING ART VALUES AT 32 EFPY: l/4T, 45~F 3i4T.35~F 2500 V&~onS f r~~ffv29a44j 2250 ___________

Unacceptable ~

Operation f

/

I I Acceptable 2000 Operation 1750

~.1500 11250 I 1000 Cooldown (F/Hr)

- ~tbady-~*ate.

~25.

-50, and 750 100 500 250 [~~J 0

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

Figure 2.2 Braidwood Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates of0, 25,50 and 100°F/hr)Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 4

BRAID WOOD UNIT 1 -

PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.la Braidwood Unit 1 Heatup Data Points at 32 EFPY (Without Margins for Instrumentation Errors)

Heatup Curve 100 F Heatup Criticality Leak Test Limit Limit I (°F~ P (nsi~ I (Ti P (nsi~ I (~F~ P insii~i 60 0 108 0 91 2000 60 1064 108 1114 108 2485 65 1114 110 1166 70 - 1166 115 1172 75 1172 120 1176 80 1176 125 1188 85 1188 130 1207 90 1207 135 1234 95 1234 140 1267 100 1267 145 1308 105 1308 150 1357 110 1357 155 1414 115 1414 160 1479 120 1479 65 1554 125 1554 1 70 1638 130 1638 175 1732 135 1732 180 1838 140 1838 185 1956 145 1956 190 2088 150 - 2088 195 2235 155 2235 200 2397 160 2397

BRAIDWOOD UNIT 1 -

PRESSURE AN!) TEMPERATURE LIMITS REPORT Table 2.lb Braidwood U-nit I Cooldown Data Points at 32 EFPY (Without Margins for Instrumentation Errors)

Cooldown Cur%es Steady State 25 T Cooldown 50 T Cooldown 100 ~F CooIdown T (°F) P (psig) I CF) P (psig) I ~ P (psig T CF) P (psig) 60 0 60 (3 60 0 60 0 60 1082 60 1078 - 60 1078* 60 1078*

65

• 1133 65 1133 65 1133 65 1133 70 1188 70 1188 70 1188 70 1188 75 1250 75 1250 75 1250 75 1250 80 1318 80 80 1318 80 1318 85 1393 85 1393 85 1393 85 1393 90 1476 90 1476 90 1476 90 1476 95 1568 95 1568 95 1568 95 1568 1(X) 1669 100 1669 I (3(3 1669 100 1669 105 178! 105 1781 105 1781 1(35 1781 11(3 1905 110 1905 110 1905 110 1905 115 2042 115 2042 115 2042 115 2(342 120 2194 120 2194 12(3 2194 120 2194 125 2361 125 2361 125 2361 - - 125 2361

• Refer to Reference 13 6

BRAiDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 3.0 Low Temperature Overpressure Protection and Boltup This section provides the Braidwood Unit 1 power operated relief valve lift settings, low temperature overpressure protection (LTOP) system arming temperature, and minimum reactor vessel boltup temperature.

3.1 LTOP System Setpoints (LCO 3,4.12)

The power operated relief valves (PORVs) shall each have maximum lift settings in accordance with Figure 3.1 and Table 3.1. These limits are based on References 3 and 4.

The LTOP setpoints are based on P/i limits which were established in accordance with 10 CFR 50, Appendix G without allowance for instrumentation error and in accordance with the methodology described in Reference 1. The LTOP PORV nominal lift settings shown in Figure 3.1 and Table 3.1 account for appropriate instrument error.

3.2 LTOP Enable Temperature Braidwood Unit 1 procedures governing the heatup and cooldown of the RCS require the arming of the LTOP System for RCS temperature of350°Fand below and disarming of LTOP for RCS temperature above 350°F.

Note that the last LTOP PORV segment in Table 3.1 extends to 400°Fwhere the pressure setpoint is 2335 psig. This is intended to prohibit PORV lift for an inadvertent LTOP system arming at power.

3.3 Reactor Vessel Boltup Temperature (Non-Technical Specification) j The minimum boltup temperature for the Reactor Vessel Flange shall be 60°F.Boltup is a condition in which the Reactor Vessel head is installed with tension applied to any stud, and with the RCS vented to atmosphere.

7

BRAID WOOD UNIT 1 -

PRESSURE AND TEMPERATURE LIMITS REPORT 2500------- ------- ------------

2335 ps~

2000~- --- ____- -- ____ -- _____

C, 1750 .- -------- - - - --- ______ ___________

~is00~

Unacceptab~sOperation 1250 h- - ______ _______

0 11000+-~ - - ------- -____ ~- - ____

I PCV-456

~  : ~ ~

541 ps~g 250 - _______

0 0 50 130 150 200 250 300 350 400 450 Auctioneered Low RCS Temperature (DEG. F)

Figure 3.1 Braidwood Unit 1 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) 8

BRAIDWOOD UNIT 1 -

PRESSURE AND TEMPERATURE LIMITS REPORT Table 3.1 Data Points for Braidwood Unit 1 Nominal P01W Setpoints for the LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) I PCV-455A PCV-456 (ITY-0413M) (1TY0413P)

AUCTIONEERED LOW RCS PRESSURE AUCTIONEERED LOW RCS PRESSURE -

RCS TEMP. (DEG, F) (PSIG) RCS TEMP. (DEG, F) (PSIG) 60 541 60 595 300 541 300 595 400 2335 400 2335 Note: To determine nominal lift setpoints for RCS Pressure and RCS Temperatures greater than 300°F,linearly interpolate between the 300°Fand 400°Fdata points shown above. (Setpoints extend to 400°Fto prevent PORV liftoff from an inadvertent LTOP system arming while at power.)

9

BRAIDWOOD UNIT 1 -

PRESSURE AND TEMPERATURE LIMITS REPORT 4.0 Reactor Vessel Material Surveillance Program The pressure vessel material surveillance program (Reference 5) is in compliance with Appendix H to 10 CFR 50, Reactor Vessel Radiation Surveillance Program, The material test requirements and the acceptance standard utilize the reference nil-ductility temperature. RTNDT, which is determined in accordance with ASME Boiler and Pressure Vessel Code Section III, NB-233 1. The empirical relationship between RTNur and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, Protection Against Non-Ductile Failure, to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E185-82.

The third and final reactor vessel material irradiation surveillance specimens (Capsule W) have been removed and analyzed to determine changes in material properties. The surveillance capsule testing has been completed for the original operating period.

10

BRAIDWOOD UNIT 1

• PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.1 I

Braidwood Unit 1 Capsule Withdrawal Schedule Capsule Vessel Location Capsule Lead Removal Time~ Estimated Capsule (Degrees) Factor~ (EFPY) Fluence (nlcm2) (a)

U 58.5° 4.37 1.10 3.87 x X 238.5° 4.23 4.234 1.24 x 10~

W 121.5° 4.20 7.61 2.09x iø~~

Z 301.5° 4.20 12.01 (d)

V 61° 3.92 Standby --

Y 241° 3.92 12.01 ~>

(a) Updated in Capsule W dosimetry analysis, (Reference 6).

(b) Effective Full Power Years (EFPY) from plant startup.

(C) Plant specific evaluation.

(d) Capsule removed and is stored in the spent fuel pool. Capsule has not been analyzed and therefore capsule fluence has not been estimated.

II

BRAIDWOOD UNIT I-PRESSURE AND TEMPERATURE LIMITS REPORT 5.0 Supplemental Data Tables The following tables provide supplemental information on reactor vessel material properties and are provided to be consistent with Generic Letter 96-03. Some of the material property values shown were used as inputs to the P/T limits.

Table 5.1 shows the calculation of the surveillance material chemistry factors using surveillance capsule data. The values ofthe CF listed in Table 5.1 are those obtained from the most recent Unit 1 Capsule data, Capsule W, (Reference 6).

Table 5.2 provides the reactor vessel material properties table.

Table 5.3 provides a summary of the Braidwood Unit I adjusted reference temperature (ARTs) at the 1/41 and 3/4T locations for 32 EFPY.

Table 5.4 shows the calculation ofARTs at 32 EFPY for the limiting Braidwood Unit 1 reactor vessel material, i.e. weld WF-562 (HT#442011, Based on Surveillance Capsules U and X Data).

Table 5.5 provides RTprs calculation for Braidwood Unit 1 BeItline Region Materials at EOL (32 EFPY), (Reference 7).

Table 5.6 provides RTp~scalculation for Braidwood Unit 1 Beltline Region Materials at Life Extension (48 EFPY), (Reference 7).

12

BRAIDWOOD UNIT 1 PRESSURE AN]) TEMPERATURE LIMITS REPORT Table 5.1 Braidwood Unit 1 Calculation of Chemistry Factors Using Surveillance Capsule Data 2

Material Capsule Capsule f~ FFd~ L~RT~DTIC} FF*~SRT,~~PT FF Lower Shell Forging U 0387 0.737 5,78 4.26 0.543 49D867/49C813-l X 1.24 1.060 38.23 40.52 1.124 (Tangential) W 2,09 1.201 24.14 28.99 1.442 Lower Shell U 0.387 0.737 0.0 0.0 0.543 Forging 49D867. 1 X 1.24 1.060 28.75 30.48 1.124 49C813-l (Axial) W 2.09 1.201 37.11 44.57 1.442 SUM: 148.82 6.218 CFForgüig = ~(W *L~RTNDT) + E( FF2) (148.82) + (6,218) 23.9°F Braidwood Unit I U 0.387 0.737 l7.06~ 12,57 0.543 Surv. Weld 1.24 1.060 30,l5~ 31.96 1.124 Material X W 2.09 1.201 49.6g~ 59.67 1.442 (Heat #442011)

Braidwood Unit 2 U 0.40 0.746 0.0 0.0 0.557 Surv. Weld X 1.23 1.058 26.3~0 27.83 1,119 Material 2.25 1.220 23.9~ 29.16 1.488 (Heat #442011) W

- SUM: 161.19 6273 CF = ~(FF * ~RT~) ÷ ~( FF2)=(161.19)÷(6.273) 25.7°F Notes:

(a) 1= Calculated tluence, ( x l0~n/cm2. E> 1.0 MeV)

(b) FF= fluence factor= ~02~0i!ogfl (c) ART.~~1values

. are the measured 30 ft-lb shift values, (d) The surveillance weld metal ART~~ values have not been adjusted.

13

BRAIDWOOD UNIT 1

• PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.2 I

Braidwood Unit 1 Reactor Vessel Material Properties Chemistry Initial Material Description Cu (%) Ni (%) Factor RT NOT (°F)

Closure Head Flange 0.11 0.67 -- -20 Heat # 5P7381/3P6406 Vessel Flange 0.77 -- -10 Heat # 122N357V Nozzle Shell Forging

• 0.04 0.73 2600F(b) 10 Heat_# 5P-7016 Intermediate Shell Forging

• 30 Heat #49D383-1/49C344-1 0.05 0.73 3l.O°F~

(also referred to as the Upper Shell forging)

Lower Shell Forging

• 3l,0o~b) 0.05 0.74 Heat # 49D867/49C813-I 23.9°F~ 20 Circumferential Weld

• o~(b)

(Intermediate Shell to Lower Shell) 0.03 0.67 410 25.7°F~ 40 WF-562 (HT#442011)

Upper Circumferential Weld *

(Nozzle Shell to Intermediate Shell) 0.04 0.46 54.0°F~ -25 WF-645_(HT# H4498)

• Beltline Region Materials a) The Initial RT~DTvalues for the plates and welds are based on measured data.

b) Chemistry Factor calculated for Cu and Ni values per Regulatory Guide 1,99, Rev. 2, Position I. I.

c) Chemistry Factor calculated for Cu and Ni values per Regulatory Guide 1,99, Rev. 2, Position 2.1, 14

BRAIDWOOD UNIT I -

PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.3 Summary ofBraidwood Unit 1 Adjusted Reference Temperatures (ARTs) at JJ4T and 3/4T Locations for 32 EFPY 32EFPY Material Description 1/4T ART(°F) 3/4T ART(°F)

Intermediate Shell Forging Heat # 49D383-l/49C344-1 36 18 (RG Position 1)

Lower Shell Forging Heat # 49D867/49C8 13-1 46 28 (RG Position 1)

Using Surveillance Data~ 31 17 (RG_Position_2(a))

Nozzle Shell Forging 48(b) 35 (b)

Heat # 5P-7016 Circumferential Weld (Intermediate Shell to Lower Shell) 126 103 WF-562 (HT# 442011)

(RG Position 1)

Using credible surveillance 94 79 Data (RG Position 2~)

(a) Calculated using a chemistry factor based on Regulatory Guide (RG) 1.99, Position 2.

(b) These ART values were used to generate the Braidwood Unit I Heatup and Cooldown curves, since they produced the most conservative curves (Reference I I).

IS

BRAID WOOD UNIT I

• PRESSURE AND TEMPERATURE LIMITS REPORT (a) The Braidwood Unit I reactor vessel wall thickness is 8.5 inches at the beltline region.

Table 5.4 Braidwood Unit 1 Calculation of Adjusted Reference Temperatures (ARTs) at 32 EFPY at the Limiting Reactor Vessel Material, Nozzle Shell Forging SP-7016 Parameter Values Operating Time 32 EFPY Location~ l/4T ART(°F) 3/4T ART(°F)

Chemistry Factor, CF (°F) 26.0 26.0 Fluence(O, n/cm2 3.65 ~ 10s 1.32 x108 (E> 1.0_Mev)~

Fluence Factor, FF 0.772 0.475

/.~.RTNDT=CFXFF(°F) 18.8 12.4 Initial RTNDT~,I(°F) 10 10 Margin. M (°F) 18.8 12.4 ART= I+/-(CF*FF)+M,°F 48 35 per RG 1.99, Revision 2 (b) Fluence f, is based upon ~ (E> 1.0 Mev) = 6.08 x l0~at 32 EFPY (Reference 11).

16

BRAIDWOOD UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.5 RTp~5Calculation for Braidwood Unit 1 Beltline Region Materials at EOL (32 EFPY)

Material Fluence FF CF ~RTp,~*~ Margin RTp.~

(1&n/cm2, (°F) (°F) (°F) (°F) (°F)

E>l.O MeY)

Intermediate Shell Forging 2.05 1.20 31,0 37,2 34 -30 41 Heat # 49D383- I /49C344- I Lower Shell Forging 2.05 1.20 31.0 37.2 34 -20 51 Heat # 49D867/49C8 13-1 Lower Shell Forging 2.05 1,20 23.9 28.7 17 -20 26 (Using S/C Data)

Nozzle Shell Forging 0,608 0.86 26.0 22.4 22.4 10 Heat # 5P.70 16 Inter, to Lower Shell Circ. Weld 19~ 1.19 41.0 48,8 48.8 40 138 WF-562 (HT# 44201!)

Inter, to Lower Shell Cire. Weld 1.99 1.19 25.7 30.6 28 40 99 Using S/C Data Nozzle Shell to Inter. Shell Cire. 0.608 0.86 54.0 46.5 46.5 -25 68 Weld WF-645 (HT# H4498) ~

(a) i.RTp15CF*FF (b) Initial RTNDT values are measured values.

(c) RT~= RTN~u)+ ~RT~ + Margin (P) 17

BRAIDWOOD UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.6 RT~Calculation for Braidwood Unit 1 Beltilne Region Materials at Life Extension (48 EFPY)

Material Fluence FE CF ~RT~1~ Margin RTNDT~I,)m RTpr~

(1&n/em~, (°F) (CF) (0F) (°F) (*F)

E>1.O MeV)

Intermediate Shell Forging 306 1,30 31.0 4(J,3 34 -30 44 Heat # 49D383- 1/49C344-1 Lower Shell Forging 306 54 1.30 31.0 40.3 34 -20 Heat # 49D867/49C8 13-I Lower Shell Forging 3.06 1.30 23.9 31.1 31.1 -20 42 Using S/C Data Nozzle Shell Forging 10 0.909 0.97 26.0 25.2 25.2 60 Heat #SP.70l6 Inter, to Lower Shell Circ. Weld 2.98 1.29 41.0 52,9 52.9 30 146 WF-562 (HT# 442011)

Inter, to Lower Shell Circ. Weld 2,98 1.29 25.7 33.2 28 40 101 Using S/C Data Nozzle Shell to Inter. Shell Cue. 0.909 0.97 54.0 524 52.4 .25 80 Weld WF-645 (HT# H4498)

(a) ~ CF *

(b) Initial RT~.values are measured values.

(c) RT~= RT~oT(u)+ ~RTprs + Margin (CF) 18

BRAIDWOOD UNIT 1-PRESSURE AND TEMPERATURE LIMITS REPORT 6.0 References

1. WCAP-14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, J.D.Aridrachek.

Ct al., January 1996.

2. NRC Letter from R. A. Capra to O.D. Kingsley. Commonwealth Edison Company, Byron Station Units 1 and 2 and Braidwood Station Units 1 and 2, Acceptance for referring of pressure temperature limits report, (M98799. M98800. M98801, and M98802), January 21 1998.

3. Westinghouse Letter to Exelon Nuclear, CAE-06-90/CCE-06-86, Transmittal of Byron and Braidwood Units I and 2 Revision 1 LTOPS Setpoints Analysis Reports for 22 and 32 EFPY (LTR-SCS-03-87, Revision 1 Attachment A) (LTR-SCS-03-87, Revision 1 Attachment B),

August 28, 2006.

4. Byron & Braidwood Design Information Transmittal DIT-BRW-2006-OOSl, Transmittal of Braidwood Unit 1 and Unit 2 Temperature and Pressure Uncertainties for Low Temperature Overpressure System (LTOPS) Power Operated Relief Valves (PORVS), Nathan (Joe)

Wolff Jr., July 18, 2006.

5. WCAP-9807, Commonwealth Edison Company, Braidwood Station Unit I Reactor Vessel Radiation Surveillance Program, S.E. Yanichko. et al., February 1981.

6. WCAP- 15316, Revision 1, Analysis of Capsule W from Commonwealth Edison Company Braidwood Unit I Reactor Vessel Radiation Surveillance Program, E. Terek, et al.,

December 1999.

7. WCAP-15365, Revision 1, Evaluation of Pressurized Thermal Shock for Braidwood Unit 1, J.H. Ledger, January 2002.

8. NRC Letter from 0. F. Dick, Jr., NRR, to C. Crane, Exelon Generation Company, LLC, Issuance of Amendments: Revised Pressure-Temperature Limits Methodology; Byron Station. Units I and 2, and Braidwood Station, Units 1 and 2, dated October 4, 2004.

9. NRC Letter from M. Chawla to O.D. Kingsley, Exelon Generation Company, LLC, Issuance ofexemption from the Requirements of 10 CFR 50 Part 60 and Appendix G for Byron Station, Units 1 and 2, and Braidwood Stations, Units 1 and 2, dated August 8, 2001.

10. NRC Letter from R. F. Kuntz, NRR, to C. M. Crane, Exelon Generation Company, LLC, Byron Station, Unit Nos. I and 2, and Braidwood Station, Unit Nos. I and 2 Issuance of Amendments Re: Reactor Coolant System Pressure and Temperature Limits Report (TAC Nos. MC8693, MC8694, MC8695. and MC8696), November 27, 2006.

19

BRAIDWOOD UNIT 1 -

PRESSURE AND TEMPERATURE LIMITS REPORT

11. WCAP-15364, Revision 2, Braidwood Unit I Heatup and Cooldown Limit Curves for Normal Operation, T.J. Laubham. November 2003.

12. WCAP-16143-P, Revision 0. Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for ByronlBraidwood Units I and 2, W. Bamford. et al., November 2003.

13. Westinghouse Letter to Exelon Nuclear, CCE-07-24, Braidwood Unit I and 2 RCS HU/CD Limit Curve Table Values, dated February 15, 2007.

20

ATTACHMENT 2 Braidwood Unit 2 Pressure and Temperature Limits Report (PTLR), Revision 4

BRAIDWOOD UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

Revision 4

BRAIDWOOD UNIT 2-PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents Section Page 1.0 Introduction 1 2.0 RCS Pressure Temperature Limits 1 2.1 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3) 1 3.0 Low Temperature Over Pressure Protection and Boltup 7 3.1 LTOP System Setpoints (LCO 3.4.12) 7 3.2 LTOP Enable Temperature 7 3.3 Reactor Vessel Boltup Temperature (Non-Technical Specification) 7 4.0 Reactor Vessel Material Surveillance Program 10 5.0 Supplemental Data Tables 12 6.0 References 19

BRAIDWOOD UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT List of Figures Figure Page 2.1 Braidwood Unit 2 Reactor Coolant System Heatup Limitations 3 (Heatup Rate of 100°F/hr)Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 2.2 Braidwood Unit 2 Reactor Coolant System Cooldown Limitations 4 (Cooldown Rates of 0, 25, 50 and lOO°F/hr)Applicable to 32 EFPY (Without Margins for Instrumentation Errors) 3.1 Braidwood Unit 2 Nominal PORV Setpoints for the Low Temperature 8 Overpressure Protection (LTOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)

II

BRAIDWOOD UNIT 2 -

PRESSURE AND TEMPERATURE LIMITS REPORT List of Tables Table Page

2. la Braidwood Unit 2 Heatup Data at 32 EFPY (Without 5 Margins for Instrumentation Errors)

2. lb Braidwood Unit 2 Cooldown Data Points 32 EFPY (Without 6 Margins for Instrumentation Errors) 3.1 Data Points for Braidwood Unit 2 Nominal PORV 9 Setpoints for the LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) 4.1 Braidwood Unit 2 Capsule Withdrawal Schedule 11 5.1 Braidwood Unit 2 Calculation ofChemistry Factors Using 13 Surveillance Capsule Data 5.2 Braidwood Unit 2 Reactor Vessel Material Properties 14 5.3 Summary of Braidwood Unit 2 Adjusted Reference 15 Temperatures (ARTs) at 1/4T and 3/4T Locations for 32 EFPY 5.4 Braidwood Unit 2 Calculation of Adjusted Reference 16 Temperatures (ARTs) at 32 EFPY at the Limiting Reactor Vessel Material, Nozzle Shell Forging 5P-7056 5.5 RTvrs Calculation for Braidwood Unit 2 Beltline Region 17 Materials at EOL (32 EFPY) 5.6 RT~Calculation for Braidwood Unit 2 Beltline Region 18 Materials at Life Extension (48 EFPY)

III

BRAIDWOOD UNIT 2-PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 Introduction This Pressure and Temperature Limits Report (PTLR) for Braidwood Unit 2 has been prepared in accordance with the requirements of Braidwood Technical Specification (TS) 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR). Revisions to the PTLR shall be provided to the NRC after issuance.

The Technical Specifications addressed in this report are listed below:

LCO 3.4.3 RCS Pressure and Temperature (PIT) Limits; and LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System.

2.0 RCS Pressure Temperature Limits The PTLR limits for Braidwood Unit 2 were developed using a methodology specified in the Technical Specifications. The methodology listed in WCAP-14040-NP-A, Revision 2 (Reference 1) was used with the following exception:

a) Optional use of ASME Code Section XI, Appendix G, Article G-2000, 1996 Addenda, b) Use ofASME Code Case N-640, Alternative Reference Fracture Toughness for Development of P-I Limit Curves,Section XI, Division 1, and c) Use of ASME Code Case N-588, Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessel,Section XI, Division 1, and d) Elimination of the flange requirements documented in WCAP-16 143-P.

This exception to the methodology in WCAP 14040-NP-A, Revision 2 has been reviewed and accepted by the NRC in References 2, 7, 9, and 10.

WCAP 15373, Revision 2 (Reference 11), provides the basis for the Braidwood Unit 2 P/T curves, along with the best estimate chemical compositions, fluence projections and adjusted reference temperatures used to determine these limits.

WCAP-l6 143-P. Reference 12, documents the technical basis for the elimination of the flange requirements.

2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3) 2.1.1 The RCS temperature rate-of-change limits defined in Reference 11 are:

a. A maximum heatup of 100°Fin any 1-hour period.

b. A maximum cooldown of 100°Fin any 1-hour period, and

BRAID WOOD UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT

c. A maximum temperature change of less than or equal to 10°Fin any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

2.1.2 The RCS PiT limits for heatup, inservice hydrostatic and leak testing, and criticality are specified by Figure 2.1 and Table 2.1a. The RCS P/T limits for cooldown are shown in Figure 2.2 and Table 2.lb. These limits are defined in WCAP-15373, Revision 2 (Reference 11). Consistent with the methodology described in Reference 1, with the exception noted in Section 2.0, the RCS P/T limits for heatup and cooldown shown in Figures 2.1 and 2.2 are provided without margins for instrument error. These limits were developed using ASME Boiler and Pressure Vessel Code Section XI, Appendix G, Article 02000, 1996 Addenda. The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in 10 CFR 50, Appendix G.

The P/T limits for core operation (except for low power physics testing) are that the reactor vessel must be at a temperature equal to orhigher than the minimum temperature required for the inservice hydrostatic test, and at least 40°Fhigher than the minimum permissible temperature in the corresponding PIT curve for heatup and cooldown.

2.

BRAIDWOOD UNIT 2 -

PRESSURE AND TEMPERATURE LIMITS REPORT Material Property Basis Limiting Material: Circumferential Weld WF-562 & Nozzle Shell Forging Limiting ART Values at 32 EFPY 1/41 93°F(N-588) & 67°F(96 App. G) 3141 79°F(N-588) & 54°F(96 App. G) 2500 __________________

/~,_.fi~akTestLmft ~0perflm Version 5 1 Run 190171 2250-Acceptable

~

____________ Operation Unacceptable Operation 1750 0 Heatup Rate 1500 l000eg.F/Hr CritIcal 100 Deg.Limit F/Hr 1250-31 1000 750 Criticality Limit based on Inservlce hydrostatic test 500 temperature (127 F) for the service period up to 32 EFPY 250 ______ rii~T -

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

Figure 2.1 Braidwood Unit 2 Reactor Coolant System Ileatup Limitations (Heatup Rate of 100°F/br)

Applicable to 32 EFPY (Without Margins for Instrumentation Errors) 3-

BRAIDWOOD UNIT 2 -

PRESSURE AND TEMPERATURE LIMITS REPORT Material Property Basis Limiting Material: Circumferential WeldWF-562 & NozzleShell Forging Limiting ART Values at 32 EFPY 1/41 93°F(N-588) & 67°F(96 App. G) 3/41 79°F(N-588) & 54°F(96 App. G) 2500 - _________________

[oper1~mVers~on:5.1Run:19o17J 2250 I Unacceptable I Acceptable 2000 Loperation J~ Operation 1750

~3.

~1500 -

J 1250 I 1000 - RatesF/Hr steady-~ste - - -

-25 - -

-50 750 -100 500 250 J~Tjtup - - - - - - -

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

Figure 2.2 Braidwood Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25,50 and 100°F/hr)Applicable to 32 EFPY (Without Margins of Instrumentation Errors) 4

BRAIDWOOD UNIT 2 -

PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.la Braidwood Unit 2 Heatup Data Points at 32 EFPY (Without Margins for Instrumentation Errors)

Heatup Curve 100 F Heatup Criticality Limit Leak Test Limit T(°F) P(psig) T(°F) P(psig) T(°F) P(psig) 60 0 127 0 110 2000 60 924 127 965 127 2485 65 965 127 977*

70 977 127 977 75 977 127 981 80 977 130 990 85 981 135 1005 90 990 140 1025 95 1005 145 1051 100 1025 150 1081 105 1051 155 1118 110 1081 160 1161 115 1118 165 1210 120 1161 170 1266 125 1210 175 1329 130 1266 180 1400 135 1329 185 1480 140 1400 190 1569 145 1480 195 1668 150 1569 200 1778 155 1668 205 1901 160 1778 210 2036 165 1901 215 2186 170 2036 220 2353 175 2186 180 2353

• Refer to Reference 13 5

BRAIDWOOD UNIT 2-PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.lb Braidwood Unit 2 Cooldown Data at 32 EFPY (Without Margins for Instrumentation Errors)

Cooldown Curves Steady State 25 °FCooldown 50 °FCooldown 100 °FCooldown T(°F) P(psig) T(°F) P(psig) T(°F) P(psig) T(°F) P(psig) 60 0 60 0 60 0 60 0 60 931 60 908 60 889 60 866 65 965 65 946 65 932 65 921 70 1003 70 989 70 980 70 980 75 1045 75 1036 75 1033 75 1033 80 1092 80 1088 80 1088 80 1088 85 1143 85 1143 85 1143 85 1143 90 1200 90 1200 90 1200 90 1200 95 1263 95 1263 95 1263 95 1263 100 1332 100 1332 100 1332 100 1332 105 1409 105 1409 105 1409 105 1409 110 1494 110 1494 110 1494 110 1494 115 1587 115 1587 115 1587 115 1587 120 1691 120 169! 120 1691 120 1691 125 1805 125 1805 125 1805 125 1805 130 1932 130 1932 130 1932 130 1932 135 207! 135 2071 135 2071 135 2071 140 2226 140 2226 140 2226 140 2226 145 2396 145 2396 145 2396 145 2396 6

BRAIDWOOD UNIT 2 -

PRESSURE AND TEMPERATURE LIMITS REPORT 3.0 Low Temperature Overpressure Protection and Boltup This section provides the Braidwood Unit 2 power operated relief valve lift settings, low temperature overpressure protection (LTOP) system arming temperature, and minimum reactor vessel boltup temperature.

3.1 LTOP System Setpoints (LCO 3.4.12).

The power operated relief valves (PORVs) shall each have nominal lift settings in accordance with Figure 3.1 and Table 3.1. These limits are based on References 3 and 8.

The LTOP setpoints are based on PIT limits that were established in accordance with 10 CFR 50, Appendix G without allowance for instrumentation error. The LTOP setpoints were developed using the methodology described in Reference 1.

The LTOP PORV nominal lift settings shown in Figure 3.1 and Table 3.1 account for appropriate instrument error.

3.2 LTOP Enable Temperature Braidwood Unit 2 procedures governing the heatup and cooldown ofthe RCS require the arming of the LTOP System for RCS temperature of 350°Fand below and disarming of LTOP for RCS temperature above 350°F.

Note that the last LTOP PORV segment in Table 3.1 extends to 400°Fwhere the pressure setpoint is 2335 psig. This is intended to prohibit PORV lift for an inadvertent LTOP system arming at power.

3.3 Reactor Vessel Boltup Temperature (Non-Technical Specification)

The minimum boltup temperature forthe Reactor Vessel Flange shall be 60°F.

Boltup is a condition in which the Reactor Vessel head is installed with tension applied to any stud, and with the RCS vented to atmosphere.

7

BRAIDWOOD UNIT 2 -

PRESSURE AND TEMPERATURE LIMITS REPORT 2500

/

2250 2000 1750 1500 I

~1250 Unacceptable Opetaftcn 2

F°°° PCV 456 z

639 psig 500 -

PCV 455A 250 0

0 50 100 150 200 250 300 350 400 450 Auctioneered Low RCS Temperature(DEG. F)

Figure 3.1 Braidwood Unit 2 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) 8

BRAIDWOOD UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 3.1 Data Points for Braidwood Unit 2 Nominal PORV Setpoints for the LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)

PCV-455A PCV-456 RCS TEMP. RCS Pressure RCS TEMP. RCS Pressure (DEG. F) (PSIG) (DEG. F) (PSIG) 60 599 60 639 300 599 300 639 400 2335 400 2335 Note: To determine nominal lift setpoints for RCS Pressure and RCS Temperatures greater than 300°F, linearly interpolate between the 300°Fand 400°Fdata points shown above. (Setpoints extend to 400°Fto prevent PORV liftoff from an inadvertent LTOP system arming while at power).

9

BRAIDWOOD UNIT 2 -

PRESSURE AND TEMPERATURE LIMITS REPORT 4.0 Reactor Vessel Material Surveillance Program The pressure vessel material surveillance program (Reference 4) is in compliance with Appendix H to 10 CFR 50, Reactor Vessel Radiation Surveillance Program. The material test requirements and the acceptance standards utilize the reference nil-ductility temperature, RTNDT, which is determined in accordance with ASME Boiler and Pressure Vessel Code,Section III, NB-233 1.

The empirical relationship between RTNDT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, Protection Against Non-Ductile Failure, to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E185-82.

The third and final reactor vessel material irradiation surveillance specimens (Capsule W) have been removed and analyzed to determine changes in material properties. The surveillance capsule testing has been completed for the original operating period.

10

BRAID WOOD UNIT 2 -

PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.1 1 Braidwood Unit 2 Capsule Withdrawal Schedule Capsule Location Capsule Lead Removal Time~ Estimated Capsule (Degrees) Factor(a) (EFPY) Fluence (n/cm2) ~

U 58.5° 4.41 1.15 4.OOx 1018~

X 238.5° 3.85 4.2 15 1.23 x lO~~

W 121.50 4.17 8.53 2.25 x i019~

Z 301.5° 4.17 12.78 (d)

~.

V 61.00 3.92 Standby Y 241.00 3.92 12.78 (~

Notes:

(a) Updated in Capsule W dosimetry analysis (Reference 5).

(b) Effective Full Power Years (EPPY) from plant startup.

(c) Plant specific evaluation.

(d) Capsule has been removed and stored in the spent fuel pool. Capsule has not been analyzed and therefore capsule fluence has not been estimated.

11 -

BRAIDWOOD UNIT 2 -

PRESSURE AND TEMPERATURE LIMITS REPORT 5.0 Supplemental Data Table The following tables provide supplemental information on reactor vessel material properties and are provided to be consistent with Generic Letter 96-03. Some of the material property values shown were used as inputs to the PIT limits.

Table 5.1 shows the calculation of the surveillance material chemistry factors using surveillance capsule data (Reference 5).

Table 5.2 provides the reactor vessel material properties table.

Table 5.3 provides a summary of the Braidwood Unit 2 adjusted reference temperatures (ARTs) at the 1/4T and 3/4T locations for 32 EFPY.

Table 5.4 shows the calculation of ARTs at 32 EFPY forthe limiting Braidwood Unit 2 reactor vessel material.

Table 5.5 provides RTprs Calculation for Braidwood Unit 2 Beltline Region Materials at EOL (32 EFPY), (Reference 6).

Table 5.6 provides RTprs Calculation for Braidwood Unit 2 Beltline Region Materials at Life Extension (48 EFPY), (Reference 6).

12 -

BRAIDWOOD UNIT 2 -

PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.1 Braidwood Unit 2 Calculation of Chemistry Factors Using Surveillance Capsule Data Material Capsule Capsule f(s) FF~ fl*~J~T~0i FF2 Lower Shell U 0.400 0.746 0.0 0.0 0.557 Forging (50D102- -_________

X 1.23 1.058 0.0 0.0 1.119 l/50C97- I)

(Tangential) W 2.25 1.220 4.53 5.53 1.488 Lower Shell U 0.400 0.746 0.0 0.0 0.557 Forging (SOD 102-X 1.23 1.058 33.94 35.91 1.119 l/50C97-l)

(Axial) W 2.25 1.220 33.2 40.50 1.488 SUM: 81.94 6.328 CFForging = ~(FF *L~RTNIYF) ÷~( ~ (81.94 ) ÷(6.328) 12.9°F Braidwood Unit I U 0.387 0.737 l7.O6~ 12.57 0.543 Surv. Weld X 1.24 1.060 30.15~ 31.96 1.124 Material (Heat #442011) W 2.09 1.201 49.68k> 59.67 1.442 Braidwood Unit 2 U 0.40 0.746 0.0 0.0 0.557 Surv. Weld X 1.23 1.058 26.3~ 27.83 1.1 19 Material (Heat #442011) W 2.25 1.220 23.9~ 29.16 1.488 SUM: 161.19 6.273 I CF= E(FF * ~.RTN~) + ~( FF2) =(16l.19) + (6.273) = 25.7°F Notes:

(a) f = Calculated tluence, ( x I 0~n/cm2, E> 1.0 MeV)

(b) FE fluence factor = 0llog I)

(c) ~ values are the measured 30 ft-lb shift values.

(d) The surveillance weld metal ~tRTNm~ values have not been adjusted.

13 -

BRAIDWOOD UNIT 2 -

PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.2 Braidwood Unit 2 Reactor Vessel Material Properties Chemistry Initial Material Description Cu (%) Ni ~

Factor RI NOT (°F)~

Closure Head Flange Heat # 3P6566/5P7547/4P6986 -- 0.75 -- 20 Serial# 2031-V-i -________

Vessel Flange 0.07 0.70 -- 20 Heat #_124P455 Nozzle Shell Forging

• 0.04 0.90 2600~

b) 30 Heat # 5P7056 Intermediate Shell Forging

• 20.O°F(b)

Heat # 49D963/49C9O4-1-1) 0.03 0.71 -30 (also_referred_to as_the_Upper Shell_forging)

Lower Shell Forging

• 37.0°F(b) 0.06 0.76 -30 Heat # SOD 102/50C97-1-1 12.9°F(c)

Circumferential Weld *

(Intermediate Shell to Lower Shell) - 41.0 F(b)

Weld Seam WF-562 0.03 0.67 40

- 25.7F(c)

Heat#442011 ,

Circumferential Weld *

(Nozzle Shell to Intermediate Shell)

Weld Seam WF-645 0.04 0.46 54.O°F(b) -25 Heat # H4498 I

• Beltline Region Materials a) The Initial RTN~values for the plates and welds are based on measured data.

b) Chemistry Factor calculated for Cu and Ni values per Regulatory Guide 1.99, Rev. 2, Position 1.1.

C) Chemistry Factor calculated for Cu and Ni values per Regulatory Guide 1.99, Rev. 2, Position 2.1 14 -

BRAIDWOOD UNIT 2

• PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.3 Summary of Braidwood Unit 2 Adjusted Reference Temperatures (ARTs) at 1/4T and 3/4T Locations for 32 EFPY 32 EFPY Material Description I/4T ART(°F) 3/41 ART(°F)

Intermediate Shell Forging Heat # 49D963/49C904-1-l) 12 0 Lower Shell Forging 43 26 Heat # 50D102/5OC97-1-1

-Using Surveillance Data 18 14 Circumferential Weld (Intermediate Shell to Lower 125 102 Shell)Weld Seam WF-562 Heat 442011

-Using Surveillance Data 93 79 Circumferential Weld (Nozzle Shell to Intermediate 51 25 Shell)Weld Seam WF-645 Heat # H4498 Nozzle Shell Forging 67(a) 54(a)

Heat # 5P7056 (a) These ART values were used to calculate the Heatup and Cooldown curves in Figures 2.1 and 2.2 using the 1996 Appendix G Methodology since they produced the most conservative curves (Reference II).

15

BRAIDWOOD UNIT 2 -

PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.4 Braidwood Unit 2 Calculation of Adjusted Reference Temperatures (ARTs) at 32 EFPY at the Limiting Reactor Vessel Material, Nozzle Shell Forging 5P-7056 Parameter Values Operating Time 32 EFPY Location~ 1/41 ART (°F) 3/41 ART(°F)

Chemistry Factor, CF (°F) 26.0 26.0 F~e~t~

3.4OxlO8 1.23x108 fluence Factor, FF 0.703 0.460 z~.RTNDT=CFxFF(°F~ 18.3 12.0 Initial RT NOT,, I(°F) 30 30 Margin, M(°F~ 18.3 12.0 ART= I~l~(CF*FF)+M, °F 67 54 per RG 1.99, Revision 2 a) The Braidwood Unit 2 reactor vessel wall thickness is 8.5 inches at the beltline region.

b) Fluence, f, is the calculated peak clad/base metal interface fluence (E> 1.0 Mev) =5.67x 8 n/cm2 at 32 EFPY (Reference 11).

16

BRAIDWOOD UNIT 2 -

PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.5 RTprs Calculation for Braidwood Unit 2 Beltline Region Materials at EOL (32 EFPY)

Material Fluence FE CF iSRT,~2> Margin RTNDT(L)(b) RT~.~°

- (1O1~n/cm2, (°F) (°F) (°F) (°F) (°F)

E>1.O MeV)

Intermediate Shell Forging 1.96 1.18 20 23.6 23.6 -30 17 Heat # 49D963/49C904- 1-1 -__________

Lower Shell Forging 1.96 34 -30 1.18 37 43.7 48 Heat # SOD l02/50C97-l-l Lower Shell Forging 1.% 1.18 12.9 15.2 34 -30 19 (Using S/C Data) (d)

Nozzle Shell Forging 0.567 0.841 26 21.9 21.9 30 74 Heat # 5P-7056 Circumferential Weld (Intermediate Shell to Lower Shell)

Weld WF-562 1.89 1.17 41.0 48.0 48.0 40 136 Heat #442011 Circumferential Weld (Intermediate Shell to Lower Shell) 1.89 1.17 25.7 30.1 28 40 98 (Using S/C Data)

Circumferential Weld (Nozzle Shell to Intermediate Shell)

Weld WF-645 0.567 0.84 1 54 45.4 45.4 -25 66 Heat#H4498 (a) Initial RT5~-values are measured values.

(b) RT~= RT~(u)+ ~RT~ + Margin (°F).

(c) z~RT~= CF

• FE (d) Surveillance data is considered not credible. In addition, the Table chemistry factor is conservative and would normally be used for calculating RT ~. However, because the chemistry factor predicted by the Regulatory Guide 1.99 Position 2.1 for the forging surveillance data was greater that the Position 1.1 chemistry factor, then the Position 2.1 chemistry factor will be used to determine the RT~with a full a~margin term.

17

BRAIDWOOD UNIT 2 -

PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.6 RTjqs Calculation for Braidwood Unit 2 Beltline Region Materials at Lire Extension (48 EFPY)

Material Fluence FE CF i~RT~~ Margin RTNDT(tJ)(b) RT~~

(1&n/cm2, (°F) (°F) (°F) (°F) (°F)

E>1.O MeV)

Intermediate Shell Forging 2.94 1.29 20 25.8 25.8 -30 22 Heat # 49D963/49C904 Lower Shell Forging 2.94 1.29 37 47.7 34 -30 52 Heat # 50D102/50C97-1-l Lower Shell Forging 2.94 1.29 12.9 16.6 (Using S/C Data) (d) 34 -30 21 Nozzle Shell Forging 0.849 0.954 26 Heat # SP-7056 24.8 24.8 30 80 Circumferential Weld (Intermediate Shell to Lower Shell) 2.83 1.28 41.0 52.9 52.9 40 145 Weld Seam WF-562 Heat #442011 Circumferential Weld (Intermediate Shell to Lower Shell) 2.83 1.28 25.7 32.9 28 40 101 (Using S/C Data)

Circumferential Weld (Nozzle Shell to Intermediate Shell) 0.849 0.954 54 51.3 51.5 -25 78 Weld Seam WF-645 Heat # 114498 (a) Initial RTNDT values are measured values.

(b) RT~= RTN~y~(u) + t~RT~ + Margin (°F)

(c) ~RTvrs CF*FF (d) Surveillancedata is considered not credible. In addition the Table chemistry factor is conservative and would normally be used for calculating RT~.However, because the chemistry factor predicted by the Reg. Guide 1.99 Position 2.1 for the forging surveillance data was greater than the Position 1.1 chemistry factor then the Position 2.1 chemistry factor will be used to determine the RT~with a full a~margin term.

18 -

BRAIDWOOD UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 6.0 References

1. WCAP- 14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, J.D.

Andrachek, et al., January 1996.

2. NRC Letter from R. A. Capra to O.D. Kingsley, Commonwealth Edison Company, Byron Station Units 1 and 2 and Braidwood Station Units 1 and 2, Acceptance for referring of pressure temperature limits report, (M98799, M98800, M98801, and M98802), January 21, 1998.

3. Westinghouse Letter to Exelon Nuclear, CAE-06-90/CCE-06-86, Transmittal of Byron and Braidwood Units 1 and 2 Revision 1 LTOPS Setpoints Analysis Reports for 22 and 32 EFPY (LTR-SCS-03-87, Revision 1 Attachment A) (LTR-SCS-03-87, Revision 1 Attachment B), August 28, 2006.

4. WCAP- 11188, Commonwealth Edison Company, Braidwood Station Unit 2 Reactor Vessel Surveillance Program, December 1986.

5. WCAP-15369, Analysis of Capsule W from the Commonwealth Edison Company Braidwood Unit 2 Reactor Vessel Radiation Surveillance Program, March 2000.

6. WCAP-15381, Evaluation of Pressurized Thermal Shock for Braidwood Unit 2, T.J.

Laubham, September 2000.

7. NRC Letter from G. F. Dick, Jr., NRR, to C. Crane, Exelon Generation Company, LLC, Issuance ofAmendments: Revised Pressure-Temperature Limits Methodology; Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, dated October 4, 2004.

8. Byron & Braidwood Design Information Transmittal DIT-BRW-2006-005 1, Transmittal of Braidwood Unit 1 and Unit 2 Temperature and Pressure Uncertainties for Low Temperature Overpressure System (LTOPS) Power Operated Relief Valves (PORVS), Nathan (Joe) Wolff Jr., July 18, 2006.

9. NRC Letter from M. Chawla to O.D. Kingsley, Exelon Generation Company, LLC, Issuance of exemption from the Requirements of 10 CFR 50 Part 60 and Appendix G for Byron Station, Units 1 and 2, and Braidwood Stations, Units 1 and 2, dated August 8, 2001.

10. NRC Letter from R. F. Kuntz, NRR, to C. M. Crane, Exelon Generation Company, LLC, ByTon Station, Unit Nos. 1 and 2, and Braidwood Station, Unit Nos. 1 and 2 Issuance of Amendments Re: Reactor Coolant System Pressure and Temperature Limits Report (TAC Nos. MC8693, MC8694, MC8695, and MC8696), November 27, 2006.

11. WCAP-15373, Revision 2, Braidwood Unit 2 Heatup and Cooldown Limits for Normal Operation, T.J. Laubham et al., November 2003.

19

BRAIDWOOD UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT

12. WCAP-16143-P, Revision 0, Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for ByronlBraidwood Units 1 and 2, W. Bamford, et al.,

November 2003.

13. Westinghouse Letter to Exelon Nuclear, CCE-07-24, Braidwood Unit 1 and 2 RCS HU/CD Limit Curve Table Values, dated February 15, 2007.

20