ML15344A159

Submittal of Pressure and Temperature Limits Reports (Ptlrs), Revision 8 and Braidwood, Unit 2 - Pressure and Temperature Limits Reports (Ptlrs), Revision 7

ML15344A159
Person / Time
Site:	Braidwood
Issue date:	12/10/2015
From:	Marchionda M Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BW150112
Download: ML15344A159 (49)

Text

Exelon Generation

< December 10, 2015 BW150112 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555-0001 Bra i dwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket No s. STN 50-456 and 50-4 5 7 Exelon Gonorotion Company, LLC Brmdwood Station 35100 South Route 53, Suito 64 Bracovitlo, IL 60401*9619

Subject:

Braidwood Unit 1 -Pressure and Temperature Limits Reports (PTLRs), Revision 8 Braidwood Unit 2 -Pressure and Tempera t ure Limits Reports (PTLRs), Revision 7

Reference:

Letter from U.S. NRC to B.C. Hanson (Exelon Generation Company, LLC), "Issuance of Amendments to Utilize WCAP-16143

-P, Revision 1, "Reactor Vessel Closure HeadNessel Flange Requirements Evaluation for Byron/Braidwood Units 1 and 2," dated October 16 , 2014," da t ed October 28, 2015 The purpose of this letter is to transmit the Pressure and Temper a ture Limits Report s (PTLRs) for Braidwood Station, Units 1 and 2 in accordance with Technic a l Spec ifi cation (TS) 5.6.6, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)." The Braidwood Unit 1 and Unit 2 PTLRs were revised to update the reference to WCAP-16143-P, "Reactor Vessel Closure HeadNessel Flange Requirements E valuation for Byron/Braidwood Unit s 1 and 2," from Revision Oto Revision 1 for consistency with the referenced Technical Specification Amendment.

Please direct any questions you may have regarding this matter to Mr. Phillip Raush, Regulatory Assurance Manager, at (815) 417-2800.

Marri Ma hionda-Palmer Site Vice resident Braidwood Station Attachments:

1. Braidwood Unit 1 Pressure Tempera t ure Li mit s Report (PTLR), Revision 8 2 Braidwood Unit 2 Pressure Temperature Limits Report (PTLR), Revision 7 ATTACHMENT 1 Braidwood Unit 1 Pressure and Temperature Limits Report (PTLR), Revision 8 BRAIDWOOD UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) Revision 8 BRAIDWOOD

-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents Section Page l .O Introduction 2.0 RCS Pressure and Temperature Limits 2.1 RCS Pressure and Temperature (Pff) Limits (LCO 3.4.3) 3.0 Low Temperature Over Pressure Protection and Boltup L 1 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 18 Figure 2.1 2.2 3.1 BRAIDWOOD

-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT List of Figures Braidwood Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 100°F/hr)

Applicable for 32 EFPY (Without Margins for Instrumentation Errors) Braidwood Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50, and l00°F/hr)

Applicable for 32 EFPY (Without Margins for Instrumentation Errors) Braidwood Unit 1 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for 32 EFPY (Includes Instrumentat i on Uncertainty) i i Page 3 4 8 BRAIDWOOD

• UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT List of Tables Table Page 2.la Braidwood Unit l 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 Errors) 3.1 Data Points for Braidwood Unit l Nominal PORV Setpoints for 9 the L TOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) 4.1 Braidwood Unit l Surveillance Capsule Withdrawal Summary 11 5.1 Braidwood Unit l Calculation of Chemistry Factors Using 13 Surveillance Capsule Data 5.2 Braidwood Unit 1 Reactor Vessel Material Properties 14 5.3 Summary of Braidwood Unit 1 Adjusted Reference Temperature 15 (ART) Values 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 SP-7016 5.5 RTPTs Calculation for Braidwood Unit 1 Beltline Region 1 7 Materials at EOL (32 EFPY) i ii BRAIDWOOD*

UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 Introduction This Pressure and Temperature Limits Report (PTLR) for Braidwood Unit 1 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 (TS) addressed in this report are listed below: LCO 3.4.3 RCS Pressure and Temperature (Pff) 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 l 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 of ASME Code Section XI, Appendix G, Article G-2000, 1996 Addenda, b) Use of ASME Code Case N-640, .. Alternative Reference Fracture Toughness for Development of P-T Limit Curves,Section XI, Di vision I", 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 l ", and d) Elimination of the 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 I 0. WCAP 15364, Revision 2 (Reference 11), provides the basis for the Braidwood Unit l Pff curves, along with the best estimate chemical compositions, fluence projections and adjusted reference temperatures used to determine these limits. WCAP-16143-P, Reference 12, documents the technical basis for the elimination of the flange requirements.

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

11) are: a A maximum heatup of 100°F in any 1-hour period, b. A maximum cooldown of 100°F in any I-hour period, and BRAIDWOOD

-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT c. A maximum temperature change of less than or equal to l0°F in any I-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves. 2.1.2 The RCS Pff limits for heatup , in s ervice hydrostatic and leak testing, and criticality are specified by Figure 2.1 and Table 2. la. The RCS Pff limits for cooldown are shown in Figure 2.2 and Table 2. lb. These limits are defined in WCAP-15364, Revision 2 (Reference 11). Consistent with the methodology described in Reference 1 and exceptions noted in Section 2.0, the RCS Pff lim i ts for heatup and cooldown shown in Figures 2.1 and 2.2 are provided without margins for instrument error. The s e limits were developed using ASME Boiler and Pressure Vessel Code Section XI, Appendix G, Article G2000, 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 Pff 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°F higher than the minimum pennissible temperature in the corresponding Pff curve for heatup and cooldown.

2 BRAIDWOOD

-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS ' LIMITING MATERIAL:

NO:ZZLE SHELL FORGING SP-7016 LIMITING ART VALUES AT 32 EFPY: l/4T, 48°F 3/4T, 35°F 2500 *. 2250 . 2000 . j Unacceptable Operation 1750 -C!J us c. -Heatup Rate *I-j I 1500 Cl) ... :::J fl) fl) 1250 100 Deg. F/Hr ) ! c. 'ti ! 1000 ..!!! . ,_-! ,./ . :::J u . "i 0 750 . -500 -250 . . .. 0 *. ' r ;Leak Test Limit I OpeM l m V erslon:5.1 Run:29844 Acceptable Operation j Critical Limit L--"' 100 Deg. F/Hr I Crltlcallty Limit based on lnaervlce hydrostatic test temperature (1 OS'F) for the service period up to 32 EFPY Boltup Temp IThe lower llmlt for RCS 1 pressure lso psla n . .. ' 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 Heatup Limitations (Heatup Rate of 100°F/hr)

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

• UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: N07ZLE FORGING SP-7016 LIMITING ART VALUES AT32 El-"PY: l/4T, 48°F 3/4T, 35°F 2500 ' 2250 . Unacceptable Operation 2000 . 1750 . -CJ ii5 a. 1500 -. J Q) ... :J fl) fl) .. 1250 Q) ... a. 1 /1 Cooldown Rates . 1000 ca '3 u 'ii 750 0 (°F/Hr) . steady.state, -25, -50, and -100 500 . . I Bolt up 250 -I Temp. I 0 IJ!he lower llmlt for RCS pressure la O psla I ' .. I Oper11m V ersl0'1: 5.1 Run: 2984 4 Acceptable Operation I I ' ' ' I 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 of 0, 25, 50 and 100°F/hr)

Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 4 BRAIDWOOD

-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 Te s t Limit Limit T (°F) P <osie:) T(°F) P (osi2) T<°F) P (o si iz) 60 Note I 108 Note I 91 2 000 60 1064 108 1114 108 24 85 65 1114 I to 1166 70 1166 115 1172 75 1172 120 1176 80 1176 125 1188 85 1188 130 1207 90 1207 135 1 2 34 95 1234 140 1267 100 1267 145 1308 105 1308 150 1357 110 1357 155 1414 115 1414 160 1479 120 1479 165 1554 125 1554 170 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 160 239 7 Note l :The Minimum acceptable pressure is 0 psi a BRAIDWOOD

-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.lb Braidwood Unit 1 Cooldown Data Points at 32 EFPY (Without Margins for Instrumentation Errors) Cooldown Curves Sleady Stale 25 °F Cooldown 50 °F Cooldown 100 °F Cooldown T (°F) P (psig) T(°F) P (psig) T (°F) P (psi g) T (°F) P (psi g) 60 Note 1 60 Note I 60 Note I 60 Note I 60 1082 60 1078 60 l078* 60 1078* 65 1133 65 1133 65 1133 65 1133 70 1188 70 t 188 70 1188 70 1188 75 1250 75 1250 75 1250 75 1250 80 1318 80 1318 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 100 1669 JOO 1669 100 1669 JOO 1669 105 1781 105 1781 105 1781 105 1781 110 1905 110 1905 110 1905 110 1905 115 2042 115 2042 115 2042 115 2042 120 2194 120 2194 120 2194 120 2194 125 2361 125 2361 125 2361 125 2361

• Refer to Reference 13 Note 1 :The Minimum acceptable pressure is 0 psia 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 opemted 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. l. These limits are based on References 3 and 4. The LTOP setpoints are based on Pff 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

l. The LTOP PORV nominal lift settings shown in Figure 3.1 and Table 3.1 account for appropriate instrument error. 3.2 L TOP Enable Temperature Braidwood Unit 1 procedures governing the heatup and cooldown of the RCS require the arming of the LTOP System for RCS temperature of 350°F and below and disarming of L TOP for RCS temperature above 350°F. Note that the last LTOP PORV segment in Table 3.1 extends to 400°F where 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 for the Reactor Vessel Flange shall 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 2250 2000 _ 1750 S! ! 1500 I a. 1250 > IC 2 .. 1000 c *e 750 500 0 0 BRAIDWOOD*

UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT I 2335p s1g I I I Unacceplable Operation I I PCV-456 I I 595 pslg i II . 541 pslg "'PCV-455A mo 150 2 00 2 50 300 350 Auctlaneenid Low RCS Temperaluni (DEG. F) Figure 3.1 Braidwood Unit 1 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (L TOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) 8 BRAIDWOOD-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT PCV-455A (ITY-0413M)

AUCTIONEERED LOW RCS TEMP. (DEG. F) 60 300 400 Table 3.1 Data Points for Braidwood Unit 1 Nominal PORV Setpoints for the LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)

PCV-456 (ITY-0413P)

RCS PRESSURE AUCTIONEERED LOW RCS PRESSURE (PSIG) RCS TEMP. (DEG. F) (PSIG) 541 60 595 541 300 595 2335 400 2335 Note: To determine nominal lift setpoints for RCS Pressure and RCS Temperatures greater than 300°F, linearly interpolate between the 300°F and 400°F data points shown above. (Setpoints extend to 400°F to 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, RT NDT, which is determined in accordance with ASME Boiler and Pressure Vessel Code Section III, NB-2331. The empirical relationship between RTNoT 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 El85-82. 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. The remaining three capsules, V, Y, and Z, were removed and placed in the spent fuel pool to avoid excessive fluence accumulation should they be needed to support life extension.

The removal summary is provided in Table 4.1. 1 0 BRAIDWOOD

-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT .Table 4.1 Braidwood Unit 1 Capsule Withdrawal Summary(*)

Capsule Capsule Lead Factor Withdrawal EFPY Fluence Location (n/cm 2 , E>l.O MeV) u 58.5° 4.02 1.16 0.388 x 10 19 x 238.5° 4.06 4.30 1.17 x 10 19 w 121.5° 4.05 7.79 l.98 x 10 19 z<c) 301.5° 4.09 12.01 (EOC 10) 2.79 x 10 19 y(c) 61.0° 3.92 17.69 (EOC 14) 3.71 x 10 19 y<c) 241.0° 3.81 12.01 (EOC 10) 2.60 x 10 19 Notes: (a) Source document is CN-AMLRS-I0-7 (Reference 14), Table 5.7-3. (b) Effective Full Power Year s (EFPY) from plant startup. (c) Standby Capsules Z, V, and Y were removed and plac ed in the spent fu e l pool. No t esti ng or analysis has been performed on th ese capsules.

If license renewal is sou g ht , one of these s tandby capsules may nt!ed to be tested to determine the effect of neutron irradiation on the reactor vessel s urveillance materials during th e period of ex1ended operation.

ll BRAIDWOOD

-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 5.0 Supplemental Data Tables The foJlowing 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 Pff limits. Table 5.1 shows the calculation of the surveillance material chemistry factors using surveillance capsule data. Table 5.2 provides the reactor vessel material properties table. Table 5.3 provides a summary of the Braidwood Unit I adjusted reference temperature (ART) values at the l/4T and 3/4T locations for 32 EFPY. Table 5.4 shows the calculation of ARTs at 32 EFPY for the limiting Braidwood Unit I reactor vessel material, i.e. weld WF-562 ( HT # 442011, Based on Surveillance Capsules U and X Data). Table 5.5 provides the RTns calculation for Braidwood Unit 1 Beltline Region Materials at EOL (32 EFPY), (Reference 7). 12 BRAIDWOOD*

UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.1 Braidwood Unit 1 Calculation of Chemistry Factors Using Surveillance Capsule Data<11> Material Capsule Capsule f<bl FF<c1 .1.RT NllTlbl FP.1.RTNDT (n/cm 2 , E > 1.0 MeV) (oF) (oF) Lower Shell u 0.388 x 10 19 0.738 5.78 4.26 Forging x 1.17 x 10 19 1.044 38.23 39.91 (Tangential) w 1.98 x 10 19 1.186 24.14 28.64 Lower Shell u 0.388 x 10 19 0.738 o.0 1 d 1 0.00 Forging x 1.17 x 10 19 1.044 28.75 30.01 (Axial) w 1.98 x 10 19 1.186 37.11 44.03 SUM: 146.8 5 CF LSForJini

= L(FF *.1.RT NoT) + L(FF 2) = (146.85) + (6.08) = 24.1°F Braidwood Unit 1 u 0.388 x 10 19 0.738 17.06 12.59 Surveillance Weld x 1.17 x 10 19 1.044 30.15 31.47 Material w 1.98 x 10 19 1.186 49.68 58.94 Braidwood Unit 2 u 0.388 x 10 19 0.738 o.o<d> 0.00 Surveillance Weld x 1.15 x 10 1 9 1.039 26.3 27.33 Material w 2.07 x 10 19 1.198 23.9 28.63 SUM: 158.96 CF Wckll'.lcw

= L(FF "'ARTNoT)

+ L(FF 2) = (158.96) + (6.10) = 26.l°F (a) Source document is CN-AMLRS-10

-7 (Reference 14), Table 5.2-1. (b) (c) f = fluence; .1.RT NOT v a lues are the measured 30 ft-lb s hift values taken from Reference

6. FF = fluen ce factor = r 0*28

• o.io*1og I) FF 2 0.54 1.09 1.41 0.54 1.09 1.41 6.08 0.54 1.09 1.41 0.54 1.08 1.44 6.10 (d) Measured ART Nm values were determined to be negative , but physically a reduction should not occur; therefore, conservative values of zero are used. IJ BRAIDWOOD

-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.2 .Braidwood Unit 1 Reactor Vessel Material Properties Material Description Cu(%) Ni(%) Chemistry Initial Factor RT NOT {°F)(a) Closure Head Flange 0.1 l 0.67 ---20 Heat# 5P738 l/3P6406 Vessel Flange Heat # 122N357V --0.77 ---10 Nozzle Shell Forging

• 0.04 0.73 26.0°Pb> 10 Heat# 5P-7016 Intermediate Shell Forging

• 0.05 0.73 3 l.0°Pb> -30 Heat # [ 49D383/49C344

)-1-1 Lower Shell Forging

• 0.05 0.74 31.0°F 0' -20 Heat# [49D867/49C813]-l-l 24.l ope> Circumferential Weld* 4l.0°Pb> (Intermediate Shell to Lower Shell) 0.03 0.67 26.1 ope> 40 WF-562 (HT# 442011) Upper Circumferential Weld* 54.oopbl (Nozzle Shell to Intermediate Shell) 0.04 0.46 -25 WF-645 (HT# H4498)

• Beltline Region Materials a) The Initial RTNoTvalues 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 1 PRESSURE AND TEMPERATURE LIMITS REPORT .Table 5.3 Summary of Braidwood Unit 1 Adjusted Reference Temperature (ART) Values at l/4T and 3/4T Locations for 32 EFPv 1*> 32 EFPY Surface Flucncc Reactor V cssel Material (n/cm2, E>l.O McV) 1/4TART 3/4TART (OF) (oF) Nozzle Shell Forging 0.586 lt 10 19 47 34 Intermediate Shell Forging 1.76 x 10 19 33 15 Lower Shell Forging 1.76 x 10 19 43 25 -.Using credible surveillance data 1.76 x 10 19 21 15 Nozzle to Intermediate Shell Forging Circ. Weld Seam 0.586x 10 19 52 25 (Heat # H4498) Intermediate to Lower Shell Forging Circ. Weld Seam 1.70 x 10 19 122 99 (Heat# 442011) -Using credible surveillance data 1.70 x 10 1 9 93 78 (a) The source document containing detailed calculations is CN-AMLRS-10-7 (Reference 14), Tables 5.3.1-1 and 5.3.1-2. The ART values summarized in this t a ble utilize the most recent fluence projections and materials data, but were not used in development of the Pff limit curves. See Figures 2.1 and 2.2 of this PTLR for the ART values used in development of the Pff limit curve s. 1 5 BRAIDWOOD

-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT .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 Locationtai l/4T ART(°F) 3/4T ART(°F) Chemistry Factor, CF (°F) 26.0 26.0 Fluence(t), n/cml (E> 1.0 Mev) 3.65 x 10 18 l.32 xl0 18 Fluence Factor , FF 0.772 0.475 i1RT Nor= CFxFF(°F) 18.8 12.4 Initial RT Nor .* 1(°F) 10 10 Margin , M (°F) 18.8 12.4 ART= I+(CF*FF)+M,°F 48 35 per RG 1.99, Revision 2 (a) The Braidwood Unit I reactor v ess el wall thickness is 8.5 inches at the beltline re gi on. (b) Flucnce r, is based upon f , wf (E > 1.0 Mev) = 6.08 x 10 18 at 32 EFPY (Reference 11). l6 BRAIDWOOD*

UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.5 RTPTS Calculation for Braidwood Unit 1 Beltline Region Materials at EOL (32 EFPYi*,h>

(c R.G.1.99, IRTNDT Reactor Vessel Material Rev.2 CF Fluence FF I .ARTNDT Position (oF) (n/cm 2 , E>l.O MeV) (OF) Nozzle Shell Forging 1.1 26 0.586 x 10 19 0.8504 10 Intennediate Shell Forging l.l 31 1.76 x 10 19 1.1554 -30 Lower Shell Forging 1.1 31 1.76 x 10 19 1.1554 +Using credible surveillance data 2.1 24.l l.76x 10 19 l.l554 -20 Nozzle to Intermediate Shell Forging Circ. Weld Seam I.I 54 0.586x 10 19 0.8504 -25 (Heat # H4498) Intermediate to Lower Shell Forging Circ. Weld Seam I.I 41 1.70 x 10 1 9 1.1461 40 (Heat# 442011) -+Using credible surveillance data 2.1 26.l 1.70 x 10 19 1.1461 40 Notes: (a) The 10 CFR 50.61 methodology was uti.l ized in the calculation of the RT PTS values. (b) The source document containing detailed calculations is CN-AMLRS-10-7 (Reference 14), Table 5.5-l. (c) Initial RT NDT va l ues are based on measured data. Hence , Gu= 0°F. (oF) 22.l 35.8 35.8 2 7.8 45.9 47.0 29.9 Cfu<c> CfJ.(d) (Of) {of) 0 l l.l 0 17 0 17 0 8.5 0 23.0 0 23.5 0 14 Margin (oF) 22.l 34 34 17 45.9 47.0 28 (d) Per th e guidance of 10 CFR 50.61, the base metal a 11 = l 7°F for Position 1.1 (without s urveillance data) and with cred i ble s urveill ance data a 4 = 8.5°F for Position 2.1; the weld m e tal a 4 = 28°F for Position I.I (without surveillance data) and with credible surveill ance data cr 4 = 14 °F for Position 2.1. However, cr 4 need not exceed 0.5*.6.RT NDT* 17 RTFrS (Of) 54 40 50 25 67 134 98 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. 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-10-MUR-197, Revision 0, "Low Temperature Overpressure Protection (LTOP) System Evaluation Final Letter Report," M.P. Rudakewiz, September 8, 2010. 4. Byron & Braidwood Design Information Transmittal DIT-BRW-2006-0051, "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 1 Reactor Vessel Radiation Surveillance Program," S.E. Yanichko, et al., February 1981. 6. WCAP-15316, Revision l, "Analysis of Capsule W from Commonwealth Edison Company Braidwood Unit l Reactor Vessel Radiation Surveillance Program," E. Terek, et al., December 1999. 7. WCAP-15365, Revision l, "Evaluation of Pressurized Thermal Shock for Braidwood Unit l," J.H. Ledger, January 2002. 8. NRC Letter from G. F. Dick, Jr., NRR, to C. Crane, Exelon Generation Company, LLC, "Issuance of Amendments:

Revised Pressure-Temperature Limits Methodology; Byron Station, Units 1 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 of exemption from the Requirements of 10 CFR 50 Part 60 and Appendix G for Byron Station, Units l and 2, and Braidwood Stations, Units 1 and 2," dated August 8, 200 l. 10. NRC Letter from R. F. Kuntz, NRR, to C. M. Crane, Exelon Generation Company, LLC, "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, and MC8696)," November 27, 2006. 18 BRAIDWOOD

-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 11. WCAP-15364, Revision 2, "Braidwood Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," T.J. Laubham, November 2003. 12. WCAP-16143-P, Revision I, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units I and 2," W. Bamford, et al., October 2014. 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. 14. Westinghouse Calculation Note CN-AMLRS-10-7, Revision 0, "Braidwood Units 1 and:? Measurement Uncc11ainty Recapture (MUR) Uprate: Reactor Vessel Integrity Evaluations," A.E. Leicht. September 2010. 19 ATTACHMENT 2 Braidwood Unit 2 Pressure and Temperature Limits Report (PTLR), Revision 7 BRAIDWOOD UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) Revision 7 BRAIDWOOD

-UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents Section Page 1.0 Introduction 1 2.0 RCS Pressure Temperature Limits 2.1 RCS Pressure and Temperature (P/f) Limits (LCO 3.4.3) 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 18 Figure 2.1 2.2 3.1 BRAIDWOOD

-UNIT 2 PRESSURE AND TEMPERA TORE LIMITS REPORT List of Figures Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of l 00°F/hr) Applicable for 32 EFPY (Without Margins for Instrumentation Errors) Braidwood Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and 100°F/hr)

Applicable to 32 EFPY (Without Margins for Instrumentation Errors) Braidwood Unit 2 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)

Page 3 4 8 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.l b Braidwood Unit 2 Cooldown Data Points 32 EFPY (Without 6 Margins for lnstrumentation Errors) 3. l 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 Summary 11 5.1 Braidwood Unit 2 Calculation of Chemistry 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 Temperature (ART) Values a t l/4T and 3/4T Locations for 32 EFPY 5.4 Braidwood Unit 2 Calculation of Adjusted Reference 16 Temperature (ARTs) at 32 EFPY at the Limiting Reactor Vessel Material, Nozzle Shell Forging SP-7056 5.5 RT PTS Calculation for Braidwood Unit 2 Beltline Region 17 Materials at EOL (32 EFPY)

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 (Pff) 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 I) was used with the following exception:

a) Optional use of ASME Code Section XI, Appendix G, Article G-2000, 1996 Addenda, b) Use of ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves,Section XI, Division l", 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-16143-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 Pff curves, along with the best estimate chemical compositions, fluence projections and adjusted reference temperatures used to detennine these limits. WCAP-16143-P, Reference 12, documents the technical basis for the elimination of the flange requirements. 2.1 RCS Pressure and Temperature (Pff) Limits (LCO 3.4.3) 2.1. l The RCS temperature rate-of-change limits defined in Reference 11 are: a. A maximum heatup of 100°F in any I-hour period. b. A maximum cooldown of 100°F in any I-hour period, and 1 BRAIDWOOD

-UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT . c. A maximum temperature change of less than or equal to l0°F in any I-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves. 2.1.2 The RCS Pff limiL'i for heatup, inservice hydrostatic and leak testing , and criticality are specified by Figure 2.1 and Table 2.1 a. The RCS Pff limits for cooldown are shown in Figure 2.2 and Table 2.1 b. These limits are defined in WCAP-15373, Revision 2 (Reference 11). Consistent with the methodology described in Reference l, with the exception noted in Section 2.0, the RCS Pff 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 G2000, 1996 Addenda. The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in LO CFR 50, Appendix G. The Pff 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°F higher than the minimum permissible temperature in the corresponding Pff curve for heatup and cooldown.

2 BRAIDWOOD*

UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT M a teri a l Proocrty B a sis Limiting M a terial: Circumferential Weld WF-562 & Nozzle Shell Forging Limiting ART Values at 32 EFPY l/4T 93°F (N-588) & 67°F ('96 App. G) 3/4T 79°F (N-588) & 54°F ('96 App. G) 2500 *. Leak Test Limit ..... 2250 . , ' -2000 . j Unacceptable 1750 -CJ Ci) 1500 CL -CD ... U) 1250 U) CD ... CL "Cl 1000 CD -..!!! *-Operation J I ) --...........

IJ . 1 DO Deg. F/Hr t r-Crltlcal Limit 100 Deg. F/Hr . . _V I . ,..-... u ca 750 . Crltlcallty Limit b a sed on lnservlca hydrostatic test 0 -temperature (127"F) for the , service period up to 32 EFPY 500 . -Boltup , Temp 250 . 0 !The lower llmlt for RCS ,_ .....-1 presaura Is o psla . . .. . ' . . . I Oper1im V ers l on: 5.1 Aun: 19017 Acceptable Operation . 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F) Figure 2.1 Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100°F/hr)

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

-UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Material Prop e rty Basis Limiting M a leri a l: Circumferential W e ld WF-562 & No z zle Shell Forging Limiting ART V a lue s at 32 EFPY 1/4T 93°F (N-588) & 67°F ('96 App. G) 3/4T 79°F (N-588) & 54°F ('96 App. G) 2500 f 2250 -. Unacceptable Operation 2000 -1750 CJ Ci5 a. -1500 u ... en J -I I ,,, ! 1250 a. "C u -m 1000 u J -Ii Cooldown Rates -(°F/Hr) , steady-state, 'ii 0 750 *25, *50, and *100 500 . I Bolt up I I Temp. 250 LJ;rh* lower limit f or RCS I -pressure lso psla --I I 0 . . . ' . ' ' Oper11m Version:5.1 Run: 19017 I Acceptable Operation

' ' 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F) Figure 2.2 Braidwood Un i t 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of O, 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 Note I 127 Note I 110 2000 60 924 127 965 127 2485 65 965 127 977* 70 977 127 977 75 977 127 981 80 977 1 3 0 990 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 1 45 14 8 0 19 5 1 668 1 5 0 1 5 6 9 2 00 1 77 8 1 55 166 8 2 0 5 190 1 16 0 177 8 2 10 2 0 36 16 5 1901 2 1 5 2 1 8 6 170 2 0 36 22 0 2353 175 2 1 86 1 8 0 2353

• Refer to Reference l 3 Note 1: The Minimum acceptable pressure is 0 psia 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 °F Cooldown 50 °F Cooldown I 00 °F Cooldown T ("F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 60 Note I 60 Note I 60 Note I 60 Note 1 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 l088 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 too 1332 100 1332 !05 1409 !05 1409 l05 1409 105 1409 110 1494 110 1494 110 1494 110 1494 115 1587 115 1587 115 1587 115 1587 120 1691 120 1691 120 1691 120 1691 125 1805 125 1805 125 1805 125 1805 130 1932 130 1932 130 1932 130 1932 135 2071 135 2071 135 2071 135 2071 140 2226 140 2226 140 2226 140 2226 145 2396 145 2396 145 2396 14 5 23 96 Note 1: The Minimum acceptable pressure is 0 psia 6

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 (L TOP) system arming temperature, and minimum reactor vessel boltup tempemture. 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 L TOP setpoints are based on Pff limits that were established in accordance with IO CFR 50, Appendix G without allowance for instrumentation error.

The L TOP 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 of the RCS require the arming of the LTOP System for RCS temperature of 350°F and 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°F where the pressure setpoint is 2335 psig. This is intended to prohibit PORV lift for an inadvertent L TOP system arming at power. 3.3 Reactor Vessel Boltup Temperature (Non-Technical Specification)

The minimum boltup temperature for the Reactor Vessel Flange shall 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 2500 2250 2000 m;o 8 ill Q. i tiSOO i Q. 1 250 > a: 0 Q. l 1000 *e 0 BRAIDWOOD

-UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT I i33 Spslg I I I Unocceplable Opera tion I I PCV456 I ! 639 psig I 599pelg '\. " PCV4 5 5A 0 100 1 50 200 2 50 300 350 400 Auctlon .. red Low RCS Temp1mure IDEG. F) Figure 3.1 Braidwood Unit 2 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (L TOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) 8 BRAIDWOOD

-UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT PCV-455A RCS TEMP. (DEG. F) 60 300 400 Table 3.1 Data Points for Braidwood Unit 2 Nominal PORV Setpoints for the LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)

PCV-456 RCS Pressure RCS TEMP. RCS Pressure (PSIG) (DEG. F) (PSIG) 599 60 639 599 300 639 2335 400 2335 Note: To determine nominal lift setpoints for RCS Pressure and RCS Temperatures greater than 300°F, linearly interpolate between the 300°F and 400°F data points shown above. (Setpoints extend to 400°F to 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, RT NOT, which is determined in accordance with ASME Boiler and Pressure Vessel Code,Section III, NB-2331. The empirical relationship between RT NOT 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 E 185-82. 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. The remaining three capsules, V, Y , and Z, were removed and placed in the spent fuel pool to avoid excessive fluence accumulation should they be needed to support life extension.

The removal summary is provided in Table4.1.

IO BRAIDWOOD

-UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.1 Braidwood Unit 2 Capsule Withdrawal Summary'0> Capsule Capsule Lead Factor Withdrawal EFPY<"> Fluence Location (n/cm 2 , E>l.O MeV) u 58.5° 4.08 l.18 0.388 x 10 19 x 238.5° 4.03 4.24 1.15 x 10 19 w 121.5° 4.06 8.56 2.07 x 10 19 zCc) 301.5° 4.14 12.78 (EOC 10) 2.83 x 10 19 y(c) 61.0° 3.92 18.42 (EOC 14) 3.73 x 10 19 y<c> 241.0° 3.89 12.78 (EOC 10) 2.66 x 10 19 Notes: (a) Source document is CN-AMLRS-10-7 (Reference 14), Table 5.7-4. (b) Effective Full Power Years (EFPY) from plant startup. (c) Standby Capsules Z, V, a nd Y were removed and placed in the spent fuel pool. No testing or analysis has been performed on these capsules.

If license renewal is sought, one of these standby capsules may need to be tested to detennine the effec t of neulron irrad i ation on the reactor vessel surveillance materials during the period of extended operation. 11 BRAIDWOOD*

UNIT 2 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 Pff limits. Table 5.1 shows the calculation of the surveillance material chemistry factors using surveillance capsule data. Table 5.2 provides the reactor vessel material properties table. Table 5.3 provides a summary of the Braidwood Unit 2 adjusted reference temperature (ART) values at the l/4T and 3/4T locations for 32 EFPY. Table 5.4 shows the calculation of ARTs at 32 EFPY for the limiting Braidwood Unit 2 reactor vessel material.

Table 5.5 provides the RT PTS Calculation for Braidwood Unit 2 Beltline Region Materials at EOL (32 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<0> Material Capsule Capsule rb 1 FF< cl ARTNDT(b)

FF*ARTNDT (n/cm 1 , E > 1.0 MeV) (oF) (oF) Lower Shell u 0.388 x 10 19 0.738 o.o<d1 0.00 Forging x 1.15 x 10 19 1.039 o.o<d> 0.00 (Tangential) w 2.07 x 10 19 1.198 4.53 5.43 Lower Shell u 0.388 x 10 19 0.738 o.o<d1 0.00 Forging x 1.15 x rn 19 1.039 33.94 35.26 (Axial) w 2.07 x 10 19 1.198 33.2 39.78 SUM: 80.47 CF LS Forging= L(FF *ART NO'T) + :E(FF 2) = (80.47) + (6.12) = 13.2°F Braidwood Unit I u 0.388 x rn 19 0.738 17.06 12.59 Surveillance Weld x 1.17 x 10 19 1.044 30.15 31.47 Material w 1.98 x 10 19 1.186 49.68 58.94 Braidwood Unit 2 u 0.388 x 10 1 9 0.738 o.o<d> 0.00 Surveillance Weld x 1.15 x 10 19 1.039 26.3 27.33 Material w 2.07 x 10 19 1.198 23.9 28.63 SUM: 158.96 CF Weld Mew= L(FF *ART Nor)+ L(FF 2) = (158.96) + (6.IO) = 26.PF Notes: (a) Source document is CN-AMLRS 7 (Reference 14 ), Table 5.2-2. (b) (c) (d) f= fluence; ARTNDT values are the measured 30 ft-lb shift values taken from Reference

5. FF= fluence faclor = (0*28

• o.io*i oa 0 Measured ART NDT values were determined to be negative, but phy s ically a reduction should not occur; therefore, conservative values of zero are used. 1 3 FF 2 0.54 1.08 1.44 0.54 1.08 1.44 6.12 0.54 1.09 1.41 0.54 1.08 1.44 6.10 BRAIDWOOD

-UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.2 Braidwood Unit 2 Reactor Vessel Material Properties Material Description Cu(%) Ni(%) Chemistry Initial Factor RT NDT (°F)(o) Closure Head Flange Heat# 3P6566/5P7547/4P6986

--0.75 --20 Serial# 2031-V-1 Vessel Flange 0.07 Heat # 124P455 0.7 0 --20 Nozzle Shell Forging

• 0.04 0.90 26.0°Fbl 30 Heat # SP-7056 lntennediate Shell Forging

• 20.0°Fb) Heat # [ 49D963/49C904

]-1-1 0.03 0.71 -30 Lower Shell Forging

• 0.06 0.76 37.0°F 0' -30 Heat# [500102/50C97]-l-l 13.2°Fc) Circumferential Weld * (Intermediate Shell to L ower Shell) 0.03 0.67 41.opb> 40 Weld Seam WF-562 26.lFc> Heat# 442011 Circumferential Weld* (Nozzle Shell to Intermediat e Shell) 0.04 0.46 54.0°phl -25 Weld Seam WF-645 Heat#H4498

• Beltline Region M a terials a) The Initial RT NOT 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 Temperature (ART) Values at l/4T and 3/4T Locations for 32 EFPv**> Surface Fluence 32EFPY Reactor Vessel Material (n/crn 2 , E>l.O MeV) l/4T ART (°F) 3/4T ART (°F) Nozzle Shell Forging 0.559 x 10 19 66 54 Intermediate Shell Forging 1.73 x 10 19 JO -I Lower Shell Forging 1.73 x 10 19 41 24 -+Using non-credible surveillance data 1.73 x 10 19 1 l Nozzle lo Intermediate Shell Forging 0.559x 10 19 Circ. Weld Seam 51 24 (Heat # H4498) Intermediate to Lower Shell Forging Circ. Weld Seam 1.67 x 10 19 122 99 (Heat# 442011) -+Using credible surveillance data 1.67 x 10 19 92 78 Notes: (a) The source document containing detailed calculations is CN-AMLRS-10-7 (Reference 14), Tables 5.3.1-3 and 5.3.1-4. The ART values summarized in this table utilize the most recent lluence projections and materials data, but were not used in development of the Prr limit c urves. See Figures 2.1 and 2.2 of this PTLR for the ART values used in development of the Pff limit curves. 1 5 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 SP-7056 Parameter Values Operating Time 32 EFPY l/4T ART (°F) 3/4T ART(0 F) Chemistry Factor, CF (°F) 26.0 26.0 Fluence(f), n/cm..: 3.40xl0 18 l.23xl0 18 (E>l.0 Mevib> Fluence Factor, FF 0.703 0.460 ART Nor-CFxFF(°F) 18.3 12.0 Initial RT NOT .* 1(°F) 30 30 Marl?in, M(°F) 18.3 12.0 ART= I+(CF*FF)+M, °F 67 54 per RO l.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 fluencc (E>l.0 Mev) =5.67xl0 18 n/cm 2 at 32 EFPY (Reference 11). 16 BRAIDWOOD

-UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.5 RTPTS Calculation for Braidwood Unit 2 Beltline Region Materials at EOL (32 EFPYia.b)

R.G.1.99, CF Flue nee IRT <c) Reactor Vessel Material Rev.2 FF NDT Position (oF) (nlcm 1 , E>l.O MeV) (oF) Nozzle Shell Forging 1.1 26 0.559 x 10 19 0.8373 30 Intermediate Shell Forging I.I 20 1.73 x 10 19 1.1508 -30 Lower Shell Forging l.l 37 l.73x 10 19 1.1508 Using non-credible surveillance data 2.l 13.2 1.73 x 10 1 9 1.1508 -30 Nozzle to Intermediate Shell Forging Circ. Weld Seam 1.1 54 0.559x 10 19 0.8373 -25 (Heal # H4498) Intermediat e 10 Lower Shell Forging Circ.

Weld Seam 1.1 41 1.67 x 10 19 1.1413 40 (Heal # 442011) -Using credible surveillance data 2.1 26.l 1.67 x 10 19 l.1413 40 (a) The lO CFR 50.61 methodology was utilized in the calcul a tion of the RT !"TS values. (b) The source document containing detailed calculations is CN-AMLRS-10-7 (Reference 14), Table 5.5-2. (c) Initial RT NOT values are based on measured data. Henc e, Gu= 0°F. 4RTNDT Gu Cc) al<dl (of) (OF) (OF) 21.8 0 10.9 23.0 0 11.5 42.6 0 17 15.2 0 7.6 45.2 0 22.6 46.8 0 23.4 29.8 0 14 Margin (Of) 21.8 23.0 34 15.2 45.2 46.8 28 (d) Per the guidance of 10 CFR 50.61, the base metal a A = 17°F for Position 1.1 (without surveillance data) and for Position 2.1 with non-credible surveillance data; the weld metal GA= 28°F for Position I.I (without surveillance data) and with credible surveillance d ata GA= 14°F for Position 2.1. Howev er , a A n eed not exceed O.S*aRT NDT* 17 RTPTS (oF) 74 16 47 0 65 134 98 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-10-MUR-197, Revision 0, "Low Temperature Overpressure Protection (LTOP) System Evaluation Final Letter Report," M.P. Rudakewiz, September 8, 2010. 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 of Amendments:

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-2006-0051, "Transmittal of Braidwood Unit l 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, "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, 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. 1 8 BRAIDWOOD

-UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 12. WCAP-16143-P, Revision l, .. Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units 1 and 2," W. Bamford, et al., October 2014. 13. Westinghouse Letter to Exelon Nuclear , CCE-07-24 , .. Braidwood Unit 1and2 RCS HU/CD Limit Curve Table Values," dated February 15, 2007. 14. Westinghouse Calculation Note CN-AMLRS-10-7 , Revision O. "Braidwood Units I and 2 Measurement Uncertainty Recapture (MURJ Uprate: Reactor Vessel Integrity Evaluations." A.E. Leicht, September 2010, and Westinghouse evaluation MCOE-LTR-13-102 Rev. 0, .. Byron and Braidwood Closure Head/Vessel Flange Region: MUR Uprate Assessment," November 2013. l9 Exelon Generation

< December 10, 2015 BW150112 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555-0001 Bra i dwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket No s. STN 50-456 and 50-4 5 7 Exelon Gonorotion Company, LLC Brmdwood Station 35100 South Route 53, Suito 64 Bracovitlo, IL 60401*9619

Subject:

Braidwood Unit 1 -Pressure and Temperature Limits Reports (PTLRs), Revision 8 Braidwood Unit 2 -Pressure and Tempera t ure Limits Reports (PTLRs), Revision 7

Reference:

Letter from U.S. NRC to B.C. Hanson (Exelon Generation Company, LLC), "Issuance of Amendments to Utilize WCAP-16143

-P, Revision 1, "Reactor Vessel Closure HeadNessel Flange Requirements Evaluation for Byron/Braidwood Units 1 and 2," dated October 16 , 2014," da t ed October 28, 2015 The purpose of this letter is to transmit the Pressure and Temper a ture Limits Report s (PTLRs) for Braidwood Station, Units 1 and 2 in accordance with Technic a l Spec ifi cation (TS) 5.6.6, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)." The Braidwood Unit 1 and Unit 2 PTLRs were revised to update the reference to WCAP-16143-P, "Reactor Vessel Closure HeadNessel Flange Requirements E valuation for Byron/Braidwood Unit s 1 and 2," from Revision Oto Revision 1 for consistency with the referenced Technical Specification Amendment.

Please direct any questions you may have regarding this matter to Mr. Phillip Raush, Regulatory Assurance Manager, at (815) 417-2800.

Marri Ma hionda-Palmer Site Vice resident Braidwood Station Attachments:

1. Braidwood Unit 1 Pressure Tempera t ure Li mit s Report (PTLR), Revision 8 2 Braidwood Unit 2 Pressure Temperature Limits Report (PTLR), Revision 7 ATTACHMENT 1 Braidwood Unit 1 Pressure and Temperature Limits Report (PTLR), Revision 8 BRAIDWOOD UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) Revision 8 BRAIDWOOD

-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents Section Page l .O Introduction 2.0 RCS Pressure and Temperature Limits 2.1 RCS Pressure and Temperature (Pff) Limits (LCO 3.4.3) 3.0 Low Temperature Over Pressure Protection and Boltup L 1 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 18 Figure 2.1 2.2 3.1 BRAIDWOOD

-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT List of Figures Braidwood Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 100°F/hr)

Applicable for 32 EFPY (Without Margins for Instrumentation Errors) Braidwood Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50, and l00°F/hr)

Applicable for 32 EFPY (Without Margins for Instrumentation Errors) Braidwood Unit 1 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for 32 EFPY (Includes Instrumentat i on Uncertainty) i i Page 3 4 8 BRAIDWOOD

• UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT List of Tables Table Page 2.la Braidwood Unit l 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 Errors) 3.1 Data Points for Braidwood Unit l Nominal PORV Setpoints for 9 the L TOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) 4.1 Braidwood Unit l Surveillance Capsule Withdrawal Summary 11 5.1 Braidwood Unit l Calculation of Chemistry Factors Using 13 Surveillance Capsule Data 5.2 Braidwood Unit 1 Reactor Vessel Material Properties 14 5.3 Summary of Braidwood Unit 1 Adjusted Reference Temperature 15 (ART) Values 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 SP-7016 5.5 RTPTs Calculation for Braidwood Unit 1 Beltline Region 1 7 Materials at EOL (32 EFPY) i ii BRAIDWOOD*

UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 Introduction This Pressure and Temperature Limits Report (PTLR) for Braidwood Unit 1 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 (TS) addressed in this report are listed below: LCO 3.4.3 RCS Pressure and Temperature (Pff) 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 l 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 of ASME Code Section XI, Appendix G, Article G-2000, 1996 Addenda, b) Use of ASME Code Case N-640, .. Alternative Reference Fracture Toughness for Development of P-T Limit Curves,Section XI, Di vision I", 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 l ", and d) Elimination of the 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 I 0. WCAP 15364, Revision 2 (Reference 11), provides the basis for the Braidwood Unit l Pff curves, along with the best estimate chemical compositions, fluence projections and adjusted reference temperatures used to determine these limits. WCAP-16143-P, Reference 12, documents the technical basis for the elimination of the flange requirements.

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

11) are: a A maximum heatup of 100°F in any 1-hour period, b. A maximum cooldown of 100°F in any I-hour period, and BRAIDWOOD

-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT c. A maximum temperature change of less than or equal to l0°F in any I-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves. 2.1.2 The RCS Pff limits for heatup , in s ervice hydrostatic and leak testing, and criticality are specified by Figure 2.1 and Table 2. la. The RCS Pff limits for cooldown are shown in Figure 2.2 and Table 2. lb. These limits are defined in WCAP-15364, Revision 2 (Reference 11). Consistent with the methodology described in Reference 1 and exceptions noted in Section 2.0, the RCS Pff lim i ts for heatup and cooldown shown in Figures 2.1 and 2.2 are provided without margins for instrument error. The s e limits were developed using ASME Boiler and Pressure Vessel Code Section XI, Appendix G, Article G2000, 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 Pff 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°F higher than the minimum pennissible temperature in the corresponding Pff curve for heatup and cooldown.

2 BRAIDWOOD

-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS ' LIMITING MATERIAL:

NO:ZZLE SHELL FORGING SP-7016 LIMITING ART VALUES AT 32 EFPY: l/4T, 48°F 3/4T, 35°F 2500 *. 2250 . 2000 . j Unacceptable Operation 1750 -C!J us c. -Heatup Rate *I-j I 1500 Cl) ... :::J fl) fl) 1250 100 Deg. F/Hr ) ! c. 'ti ! 1000 ..!!! . ,_-! ,./ . :::J u . "i 0 750 . -500 -250 . . .. 0 *. ' r ;Leak Test Limit I OpeM l m V erslon:5.1 Run:29844 Acceptable Operation j Critical Limit L--"' 100 Deg. F/Hr I Crltlcallty Limit based on lnaervlce hydrostatic test temperature (1 OS'F) for the service period up to 32 EFPY Boltup Temp IThe lower llmlt for RCS 1 pressure lso psla n . .. ' 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 Heatup Limitations (Heatup Rate of 100°F/hr)

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

• UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: N07ZLE FORGING SP-7016 LIMITING ART VALUES AT32 El-"PY: l/4T, 48°F 3/4T, 35°F 2500 ' 2250 . Unacceptable Operation 2000 . 1750 . -CJ ii5 a. 1500 -. J Q) ... :J fl) fl) .. 1250 Q) ... a. 1 /1 Cooldown Rates . 1000 ca '3 u 'ii 750 0 (°F/Hr) . steady.state, -25, -50, and -100 500 . . I Bolt up 250 -I Temp. I 0 IJ!he lower llmlt for RCS pressure la O psla I ' .. I Oper11m V ersl0'1: 5.1 Run: 2984 4 Acceptable Operation I I ' ' ' I 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 of 0, 25, 50 and 100°F/hr)

Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 4 BRAIDWOOD

-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 Te s t Limit Limit T (°F) P <osie:) T(°F) P (osi2) T<°F) P (o si iz) 60 Note I 108 Note I 91 2 000 60 1064 108 1114 108 24 85 65 1114 I to 1166 70 1166 115 1172 75 1172 120 1176 80 1176 125 1188 85 1188 130 1207 90 1207 135 1 2 34 95 1234 140 1267 100 1267 145 1308 105 1308 150 1357 110 1357 155 1414 115 1414 160 1479 120 1479 165 1554 125 1554 170 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 160 239 7 Note l :The Minimum acceptable pressure is 0 psi a BRAIDWOOD

-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.lb Braidwood Unit 1 Cooldown Data Points at 32 EFPY (Without Margins for Instrumentation Errors) Cooldown Curves Sleady Stale 25 °F Cooldown 50 °F Cooldown 100 °F Cooldown T (°F) P (psig) T(°F) P (psig) T (°F) P (psi g) T (°F) P (psi g) 60 Note 1 60 Note I 60 Note I 60 Note I 60 1082 60 1078 60 l078* 60 1078* 65 1133 65 1133 65 1133 65 1133 70 1188 70 t 188 70 1188 70 1188 75 1250 75 1250 75 1250 75 1250 80 1318 80 1318 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 100 1669 JOO 1669 100 1669 JOO 1669 105 1781 105 1781 105 1781 105 1781 110 1905 110 1905 110 1905 110 1905 115 2042 115 2042 115 2042 115 2042 120 2194 120 2194 120 2194 120 2194 125 2361 125 2361 125 2361 125 2361

• Refer to Reference 13 Note 1 :The Minimum acceptable pressure is 0 psia 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 opemted 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. l. These limits are based on References 3 and 4. The LTOP setpoints are based on Pff 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

l. The LTOP PORV nominal lift settings shown in Figure 3.1 and Table 3.1 account for appropriate instrument error. 3.2 L TOP Enable Temperature Braidwood Unit 1 procedures governing the heatup and cooldown of the RCS require the arming of the LTOP System for RCS temperature of 350°F and below and disarming of L TOP for RCS temperature above 350°F. Note that the last LTOP PORV segment in Table 3.1 extends to 400°F where 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 for the Reactor Vessel Flange shall 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 2250 2000 _ 1750 S! ! 1500 I a. 1250 > IC 2 .. 1000 c *e 750 500 0 0 BRAIDWOOD*

UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT I 2335p s1g I I I Unacceplable Operation I I PCV-456 I I 595 pslg i II . 541 pslg "'PCV-455A mo 150 2 00 2 50 300 350 Auctlaneenid Low RCS Temperaluni (DEG. F) Figure 3.1 Braidwood Unit 1 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (L TOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) 8 BRAIDWOOD-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT PCV-455A (ITY-0413M)

AUCTIONEERED LOW RCS TEMP. (DEG. F) 60 300 400 Table 3.1 Data Points for Braidwood Unit 1 Nominal PORV Setpoints for the LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)

PCV-456 (ITY-0413P)

RCS PRESSURE AUCTIONEERED LOW RCS PRESSURE (PSIG) RCS TEMP. (DEG. F) (PSIG) 541 60 595 541 300 595 2335 400 2335 Note: To determine nominal lift setpoints for RCS Pressure and RCS Temperatures greater than 300°F, linearly interpolate between the 300°F and 400°F data points shown above. (Setpoints extend to 400°F to 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, RT NDT, which is determined in accordance with ASME Boiler and Pressure Vessel Code Section III, NB-2331. The empirical relationship between RTNoT 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 El85-82. 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. The remaining three capsules, V, Y, and Z, were removed and placed in the spent fuel pool to avoid excessive fluence accumulation should they be needed to support life extension.

The removal summary is provided in Table 4.1. 1 0 BRAIDWOOD

-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT .Table 4.1 Braidwood Unit 1 Capsule Withdrawal Summary(*)

Capsule Capsule Lead Factor Withdrawal EFPY Fluence Location (n/cm 2 , E>l.O MeV) u 58.5° 4.02 1.16 0.388 x 10 19 x 238.5° 4.06 4.30 1.17 x 10 19 w 121.5° 4.05 7.79 l.98 x 10 19 z<c) 301.5° 4.09 12.01 (EOC 10) 2.79 x 10 19 y(c) 61.0° 3.92 17.69 (EOC 14) 3.71 x 10 19 y<c) 241.0° 3.81 12.01 (EOC 10) 2.60 x 10 19 Notes: (a) Source document is CN-AMLRS-I0-7 (Reference 14), Table 5.7-3. (b) Effective Full Power Year s (EFPY) from plant startup. (c) Standby Capsules Z, V, and Y were removed and plac ed in the spent fu e l pool. No t esti ng or analysis has been performed on th ese capsules.

If license renewal is sou g ht , one of these s tandby capsules may nt!ed to be tested to determine the effect of neutron irradiation on the reactor vessel s urveillance materials during th e period of ex1ended operation.

ll BRAIDWOOD

-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 5.0 Supplemental Data Tables The foJlowing 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 Pff limits. Table 5.1 shows the calculation of the surveillance material chemistry factors using surveillance capsule data. Table 5.2 provides the reactor vessel material properties table. Table 5.3 provides a summary of the Braidwood Unit I adjusted reference temperature (ART) values at the l/4T and 3/4T locations for 32 EFPY. Table 5.4 shows the calculation of ARTs at 32 EFPY for the limiting Braidwood Unit I reactor vessel material, i.e. weld WF-562 ( HT # 442011, Based on Surveillance Capsules U and X Data). Table 5.5 provides the RTns calculation for Braidwood Unit 1 Beltline Region Materials at EOL (32 EFPY), (Reference 7). 12 BRAIDWOOD*

UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.1 Braidwood Unit 1 Calculation of Chemistry Factors Using Surveillance Capsule Data<11> Material Capsule Capsule f<bl FF<c1 .1.RT NllTlbl FP.1.RTNDT (n/cm 2 , E > 1.0 MeV) (oF) (oF) Lower Shell u 0.388 x 10 19 0.738 5.78 4.26 Forging x 1.17 x 10 19 1.044 38.23 39.91 (Tangential) w 1.98 x 10 19 1.186 24.14 28.64 Lower Shell u 0.388 x 10 19 0.738 o.0 1 d 1 0.00 Forging x 1.17 x 10 19 1.044 28.75 30.01 (Axial) w 1.98 x 10 19 1.186 37.11 44.03 SUM: 146.8 5 CF LSForJini

= L(FF *.1.RT NoT) + L(FF 2) = (146.85) + (6.08) = 24.1°F Braidwood Unit 1 u 0.388 x 10 19 0.738 17.06 12.59 Surveillance Weld x 1.17 x 10 19 1.044 30.15 31.47 Material w 1.98 x 10 19 1.186 49.68 58.94 Braidwood Unit 2 u 0.388 x 10 19 0.738 o.o<d> 0.00 Surveillance Weld x 1.15 x 10 1 9 1.039 26.3 27.33 Material w 2.07 x 10 19 1.198 23.9 28.63 SUM: 158.96 CF Wckll'.lcw

= L(FF "'ARTNoT)

+ L(FF 2) = (158.96) + (6.10) = 26.l°F (a) Source document is CN-AMLRS-10

-7 (Reference 14), Table 5.2-1. (b) (c) f = fluence; .1.RT NOT v a lues are the measured 30 ft-lb s hift values taken from Reference

6. FF = fluen ce factor = r 0*28

• o.io*1og I) FF 2 0.54 1.09 1.41 0.54 1.09 1.41 6.08 0.54 1.09 1.41 0.54 1.08 1.44 6.10 (d) Measured ART Nm values were determined to be negative , but physically a reduction should not occur; therefore, conservative values of zero are used. IJ BRAIDWOOD

-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.2 .Braidwood Unit 1 Reactor Vessel Material Properties Material Description Cu(%) Ni(%) Chemistry Initial Factor RT NOT {°F)(a) Closure Head Flange 0.1 l 0.67 ---20 Heat# 5P738 l/3P6406 Vessel Flange Heat # 122N357V --0.77 ---10 Nozzle Shell Forging

• 0.04 0.73 26.0°Pb> 10 Heat# 5P-7016 Intermediate Shell Forging

• 0.05 0.73 3 l.0°Pb> -30 Heat # [ 49D383/49C344

)-1-1 Lower Shell Forging

• 0.05 0.74 31.0°F 0' -20 Heat# [49D867/49C813]-l-l 24.l ope> Circumferential Weld* 4l.0°Pb> (Intermediate Shell to Lower Shell) 0.03 0.67 26.1 ope> 40 WF-562 (HT# 442011) Upper Circumferential Weld* 54.oopbl (Nozzle Shell to Intermediate Shell) 0.04 0.46 -25 WF-645 (HT# H4498)

• Beltline Region Materials a) The Initial RTNoTvalues 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 1 PRESSURE AND TEMPERATURE LIMITS REPORT .Table 5.3 Summary of Braidwood Unit 1 Adjusted Reference Temperature (ART) Values at l/4T and 3/4T Locations for 32 EFPv 1*> 32 EFPY Surface Flucncc Reactor V cssel Material (n/cm2, E>l.O McV) 1/4TART 3/4TART (OF) (oF) Nozzle Shell Forging 0.586 lt 10 19 47 34 Intermediate Shell Forging 1.76 x 10 19 33 15 Lower Shell Forging 1.76 x 10 19 43 25 -.Using credible surveillance data 1.76 x 10 19 21 15 Nozzle to Intermediate Shell Forging Circ. Weld Seam 0.586x 10 19 52 25 (Heat # H4498) Intermediate to Lower Shell Forging Circ. Weld Seam 1.70 x 10 19 122 99 (Heat# 442011) -Using credible surveillance data 1.70 x 10 1 9 93 78 (a) The source document containing detailed calculations is CN-AMLRS-10-7 (Reference 14), Tables 5.3.1-1 and 5.3.1-2. The ART values summarized in this t a ble utilize the most recent fluence projections and materials data, but were not used in development of the Pff limit curves. See Figures 2.1 and 2.2 of this PTLR for the ART values used in development of the Pff limit curve s. 1 5 BRAIDWOOD

-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT .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 Locationtai l/4T ART(°F) 3/4T ART(°F) Chemistry Factor, CF (°F) 26.0 26.0 Fluence(t), n/cml (E> 1.0 Mev) 3.65 x 10 18 l.32 xl0 18 Fluence Factor , FF 0.772 0.475 i1RT Nor= CFxFF(°F) 18.8 12.4 Initial RT Nor .* 1(°F) 10 10 Margin , M (°F) 18.8 12.4 ART= I+(CF*FF)+M,°F 48 35 per RG 1.99, Revision 2 (a) The Braidwood Unit I reactor v ess el wall thickness is 8.5 inches at the beltline re gi on. (b) Flucnce r, is based upon f , wf (E > 1.0 Mev) = 6.08 x 10 18 at 32 EFPY (Reference 11). l6 BRAIDWOOD*

UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.5 RTPTS Calculation for Braidwood Unit 1 Beltline Region Materials at EOL (32 EFPYi*,h>

(c R.G.1.99, IRTNDT Reactor Vessel Material Rev.2 CF Fluence FF I .ARTNDT Position (oF) (n/cm 2 , E>l.O MeV) (OF) Nozzle Shell Forging 1.1 26 0.586 x 10 19 0.8504 10 Intennediate Shell Forging l.l 31 1.76 x 10 19 1.1554 -30 Lower Shell Forging 1.1 31 1.76 x 10 19 1.1554 +Using credible surveillance data 2.1 24.l l.76x 10 19 l.l554 -20 Nozzle to Intermediate Shell Forging Circ. Weld Seam I.I 54 0.586x 10 19 0.8504 -25 (Heat # H4498) Intermediate to Lower Shell Forging Circ. Weld Seam I.I 41 1.70 x 10 1 9 1.1461 40 (Heat# 442011) -+Using credible surveillance data 2.1 26.l 1.70 x 10 19 1.1461 40 Notes: (a) The 10 CFR 50.61 methodology was uti.l ized in the calculation of the RT PTS values. (b) The source document containing detailed calculations is CN-AMLRS-10-7 (Reference 14), Table 5.5-l. (c) Initial RT NDT va l ues are based on measured data. Hence , Gu= 0°F. (oF) 22.l 35.8 35.8 2 7.8 45.9 47.0 29.9 Cfu<c> CfJ.(d) (Of) {of) 0 l l.l 0 17 0 17 0 8.5 0 23.0 0 23.5 0 14 Margin (oF) 22.l 34 34 17 45.9 47.0 28 (d) Per th e guidance of 10 CFR 50.61, the base metal a 11 = l 7°F for Position 1.1 (without s urveillance data) and with cred i ble s urveill ance data a 4 = 8.5°F for Position 2.1; the weld m e tal a 4 = 28°F for Position I.I (without surveillance data) and with credible surveill ance data cr 4 = 14 °F for Position 2.1. However, cr 4 need not exceed 0.5*.6.RT NDT* 17 RTFrS (Of) 54 40 50 25 67 134 98 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. 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-10-MUR-197, Revision 0, "Low Temperature Overpressure Protection (LTOP) System Evaluation Final Letter Report," M.P. Rudakewiz, September 8, 2010. 4. Byron & Braidwood Design Information Transmittal DIT-BRW-2006-0051, "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 1 Reactor Vessel Radiation Surveillance Program," S.E. Yanichko, et al., February 1981. 6. WCAP-15316, Revision l, "Analysis of Capsule W from Commonwealth Edison Company Braidwood Unit l Reactor Vessel Radiation Surveillance Program," E. Terek, et al., December 1999. 7. WCAP-15365, Revision l, "Evaluation of Pressurized Thermal Shock for Braidwood Unit l," J.H. Ledger, January 2002. 8. NRC Letter from G. F. Dick, Jr., NRR, to C. Crane, Exelon Generation Company, LLC, "Issuance of Amendments:

Revised Pressure-Temperature Limits Methodology; Byron Station, Units 1 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 of exemption from the Requirements of 10 CFR 50 Part 60 and Appendix G for Byron Station, Units l and 2, and Braidwood Stations, Units 1 and 2," dated August 8, 200 l. 10. NRC Letter from R. F. Kuntz, NRR, to C. M. Crane, Exelon Generation Company, LLC, "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, and MC8696)," November 27, 2006. 18 BRAIDWOOD

-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 11. WCAP-15364, Revision 2, "Braidwood Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," T.J. Laubham, November 2003. 12. WCAP-16143-P, Revision I, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units I and 2," W. Bamford, et al., October 2014. 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. 14. Westinghouse Calculation Note CN-AMLRS-10-7, Revision 0, "Braidwood Units 1 and:? Measurement Uncc11ainty Recapture (MUR) Uprate: Reactor Vessel Integrity Evaluations," A.E. Leicht. September 2010. 19 ATTACHMENT 2 Braidwood Unit 2 Pressure and Temperature Limits Report (PTLR), Revision 7 BRAIDWOOD UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) Revision 7 BRAIDWOOD

-UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents Section Page 1.0 Introduction 1 2.0 RCS Pressure Temperature Limits 2.1 RCS Pressure and Temperature (P/f) Limits (LCO 3.4.3) 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 18 Figure 2.1 2.2 3.1 BRAIDWOOD

-UNIT 2 PRESSURE AND TEMPERA TORE LIMITS REPORT List of Figures Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of l 00°F/hr) Applicable for 32 EFPY (Without Margins for Instrumentation Errors) Braidwood Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and 100°F/hr)

Applicable to 32 EFPY (Without Margins for Instrumentation Errors) Braidwood Unit 2 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)

Page 3 4 8 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.l b Braidwood Unit 2 Cooldown Data Points 32 EFPY (Without 6 Margins for lnstrumentation Errors) 3. l 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 Summary 11 5.1 Braidwood Unit 2 Calculation of Chemistry 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 Temperature (ART) Values a t l/4T and 3/4T Locations for 32 EFPY 5.4 Braidwood Unit 2 Calculation of Adjusted Reference 16 Temperature (ARTs) at 32 EFPY at the Limiting Reactor Vessel Material, Nozzle Shell Forging SP-7056 5.5 RT PTS Calculation for Braidwood Unit 2 Beltline Region 17 Materials at EOL (32 EFPY)

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 (Pff) 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 I) was used with the following exception:

a) Optional use of ASME Code Section XI, Appendix G, Article G-2000, 1996 Addenda, b) Use of ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves,Section XI, Division l", 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-16143-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 Pff curves, along with the best estimate chemical compositions, fluence projections and adjusted reference temperatures used to detennine these limits. WCAP-16143-P, Reference 12, documents the technical basis for the elimination of the flange requirements. 2.1 RCS Pressure and Temperature (Pff) Limits (LCO 3.4.3) 2.1. l The RCS temperature rate-of-change limits defined in Reference 11 are: a. A maximum heatup of 100°F in any I-hour period. b. A maximum cooldown of 100°F in any I-hour period, and 1 BRAIDWOOD

-UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT . c. A maximum temperature change of less than or equal to l0°F in any I-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves. 2.1.2 The RCS Pff limiL'i for heatup, inservice hydrostatic and leak testing , and criticality are specified by Figure 2.1 and Table 2.1 a. The RCS Pff limits for cooldown are shown in Figure 2.2 and Table 2.1 b. These limits are defined in WCAP-15373, Revision 2 (Reference 11). Consistent with the methodology described in Reference l, with the exception noted in Section 2.0, the RCS Pff 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 G2000, 1996 Addenda. The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in LO CFR 50, Appendix G. The Pff 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°F higher than the minimum permissible temperature in the corresponding Pff curve for heatup and cooldown.

2 BRAIDWOOD*

UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT M a teri a l Proocrty B a sis Limiting M a terial: Circumferential Weld WF-562 & Nozzle Shell Forging Limiting ART Values at 32 EFPY l/4T 93°F (N-588) & 67°F ('96 App. G) 3/4T 79°F (N-588) & 54°F ('96 App. G) 2500 *. Leak Test Limit ..... 2250 . , ' -2000 . j Unacceptable 1750 -CJ Ci) 1500 CL -CD ... U) 1250 U) CD ... CL "Cl 1000 CD -..!!! *-Operation J I ) --...........

IJ . 1 DO Deg. F/Hr t r-Crltlcal Limit 100 Deg. F/Hr . . _V I . ,..-... u ca 750 . Crltlcallty Limit b a sed on lnservlca hydrostatic test 0 -temperature (127"F) for the , service period up to 32 EFPY 500 . -Boltup , Temp 250 . 0 !The lower llmlt for RCS ,_ .....-1 presaura Is o psla . . .. . ' . . . I Oper1im V ers l on: 5.1 Aun: 19017 Acceptable Operation . 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F) Figure 2.1 Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100°F/hr)

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

-UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Material Prop e rty Basis Limiting M a leri a l: Circumferential W e ld WF-562 & No z zle Shell Forging Limiting ART V a lue s at 32 EFPY 1/4T 93°F (N-588) & 67°F ('96 App. G) 3/4T 79°F (N-588) & 54°F ('96 App. G) 2500 f 2250 -. Unacceptable Operation 2000 -1750 CJ Ci5 a. -1500 u ... en J -I I ,,, ! 1250 a. "C u -m 1000 u J -Ii Cooldown Rates -(°F/Hr) , steady-state, 'ii 0 750 *25, *50, and *100 500 . I Bolt up I I Temp. 250 LJ;rh* lower limit f or RCS I -pressure lso psla --I I 0 . . . ' . ' ' Oper11m Version:5.1 Run: 19017 I Acceptable Operation

' ' 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F) Figure 2.2 Braidwood Un i t 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of O, 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 Note I 127 Note I 110 2000 60 924 127 965 127 2485 65 965 127 977* 70 977 127 977 75 977 127 981 80 977 1 3 0 990 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 1 45 14 8 0 19 5 1 668 1 5 0 1 5 6 9 2 00 1 77 8 1 55 166 8 2 0 5 190 1 16 0 177 8 2 10 2 0 36 16 5 1901 2 1 5 2 1 8 6 170 2 0 36 22 0 2353 175 2 1 86 1 8 0 2353

• Refer to Reference l 3 Note 1: The Minimum acceptable pressure is 0 psia 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 °F Cooldown 50 °F Cooldown I 00 °F Cooldown T ("F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 60 Note I 60 Note I 60 Note I 60 Note 1 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 l088 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 too 1332 100 1332 !05 1409 !05 1409 l05 1409 105 1409 110 1494 110 1494 110 1494 110 1494 115 1587 115 1587 115 1587 115 1587 120 1691 120 1691 120 1691 120 1691 125 1805 125 1805 125 1805 125 1805 130 1932 130 1932 130 1932 130 1932 135 2071 135 2071 135 2071 135 2071 140 2226 140 2226 140 2226 140 2226 145 2396 145 2396 145 2396 14 5 23 96 Note 1: The Minimum acceptable pressure is 0 psia 6

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 (L TOP) system arming temperature, and minimum reactor vessel boltup tempemture. 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 L TOP setpoints are based on Pff limits that were established in accordance with IO CFR 50, Appendix G without allowance for instrumentation error.

The L TOP 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 of the RCS require the arming of the LTOP System for RCS temperature of 350°F and 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°F where the pressure setpoint is 2335 psig. This is intended to prohibit PORV lift for an inadvertent L TOP system arming at power. 3.3 Reactor Vessel Boltup Temperature (Non-Technical Specification)

The minimum boltup temperature for the Reactor Vessel Flange shall 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 2500 2250 2000 m;o 8 ill Q. i tiSOO i Q. 1 250 > a: 0 Q. l 1000 *e 0 BRAIDWOOD

-UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT I i33 Spslg I I I Unocceplable Opera tion I I PCV456 I ! 639 psig I 599pelg '\. " PCV4 5 5A 0 100 1 50 200 2 50 300 350 400 Auctlon .. red Low RCS Temp1mure IDEG. F) Figure 3.1 Braidwood Unit 2 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (L TOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) 8 BRAIDWOOD

-UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT PCV-455A RCS TEMP. (DEG. F) 60 300 400 Table 3.1 Data Points for Braidwood Unit 2 Nominal PORV Setpoints for the LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)

PCV-456 RCS Pressure RCS TEMP. RCS Pressure (PSIG) (DEG. F) (PSIG) 599 60 639 599 300 639 2335 400 2335 Note: To determine nominal lift setpoints for RCS Pressure and RCS Temperatures greater than 300°F, linearly interpolate between the 300°F and 400°F data points shown above. (Setpoints extend to 400°F to 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, RT NOT, which is determined in accordance with ASME Boiler and Pressure Vessel Code,Section III, NB-2331. The empirical relationship between RT NOT 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 E 185-82. 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. The remaining three capsules, V, Y , and Z, were removed and placed in the spent fuel pool to avoid excessive fluence accumulation should they be needed to support life extension.

The removal summary is provided in Table4.1.

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-UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.1 Braidwood Unit 2 Capsule Withdrawal Summary'0> Capsule Capsule Lead Factor Withdrawal EFPY<"> Fluence Location (n/cm 2 , E>l.O MeV) u 58.5° 4.08 l.18 0.388 x 10 19 x 238.5° 4.03 4.24 1.15 x 10 19 w 121.5° 4.06 8.56 2.07 x 10 19 zCc) 301.5° 4.14 12.78 (EOC 10) 2.83 x 10 19 y(c) 61.0° 3.92 18.42 (EOC 14) 3.73 x 10 19 y<c> 241.0° 3.89 12.78 (EOC 10) 2.66 x 10 19 Notes: (a) Source document is CN-AMLRS-10-7 (Reference 14), Table 5.7-4. (b) Effective Full Power Years (EFPY) from plant startup. (c) Standby Capsules Z, V, a nd Y were removed and placed in the spent fuel pool. No testing or analysis has been performed on these capsules.

If license renewal is sought, one of these standby capsules may need to be tested to detennine the effec t of neulron irrad i ation on the reactor vessel surveillance materials during the period of extended operation. 11 BRAIDWOOD*

UNIT 2 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 Pff limits. Table 5.1 shows the calculation of the surveillance material chemistry factors using surveillance capsule data. Table 5.2 provides the reactor vessel material properties table. Table 5.3 provides a summary of the Braidwood Unit 2 adjusted reference temperature (ART) values at the l/4T and 3/4T locations for 32 EFPY. Table 5.4 shows the calculation of ARTs at 32 EFPY for the limiting Braidwood Unit 2 reactor vessel material.

Table 5.5 provides the RT PTS Calculation for Braidwood Unit 2 Beltline Region Materials at EOL (32 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<0> Material Capsule Capsule rb 1 FF< cl ARTNDT(b)

FF*ARTNDT (n/cm 1 , E > 1.0 MeV) (oF) (oF) Lower Shell u 0.388 x 10 19 0.738 o.o<d1 0.00 Forging x 1.15 x 10 19 1.039 o.o<d> 0.00 (Tangential) w 2.07 x 10 19 1.198 4.53 5.43 Lower Shell u 0.388 x 10 19 0.738 o.o<d1 0.00 Forging x 1.15 x rn 19 1.039 33.94 35.26 (Axial) w 2.07 x 10 19 1.198 33.2 39.78 SUM: 80.47 CF LS Forging= L(FF *ART NO'T) + :E(FF 2) = (80.47) + (6.12) = 13.2°F Braidwood Unit I u 0.388 x rn 19 0.738 17.06 12.59 Surveillance Weld x 1.17 x 10 19 1.044 30.15 31.47 Material w 1.98 x 10 19 1.186 49.68 58.94 Braidwood Unit 2 u 0.388 x 10 1 9 0.738 o.o<d> 0.00 Surveillance Weld x 1.15 x 10 19 1.039 26.3 27.33 Material w 2.07 x 10 19 1.198 23.9 28.63 SUM: 158.96 CF Weld Mew= L(FF *ART Nor)+ L(FF 2) = (158.96) + (6.IO) = 26.PF Notes: (a) Source document is CN-AMLRS 7 (Reference 14 ), Table 5.2-2. (b) (c) (d) f= fluence; ARTNDT values are the measured 30 ft-lb shift values taken from Reference

5. FF= fluence faclor = (0*28

• o.io*i oa 0 Measured ART NDT values were determined to be negative, but phy s ically a reduction should not occur; therefore, conservative values of zero are used. 1 3 FF 2 0.54 1.08 1.44 0.54 1.08 1.44 6.12 0.54 1.09 1.41 0.54 1.08 1.44 6.10 BRAIDWOOD

-UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.2 Braidwood Unit 2 Reactor Vessel Material Properties Material Description Cu(%) Ni(%) Chemistry Initial Factor RT NDT (°F)(o) Closure Head Flange Heat# 3P6566/5P7547/4P6986

--0.75 --20 Serial# 2031-V-1 Vessel Flange 0.07 Heat # 124P455 0.7 0 --20 Nozzle Shell Forging

• 0.04 0.90 26.0°Fbl 30 Heat # SP-7056 lntennediate Shell Forging

• 20.0°Fb) Heat # [ 49D963/49C904

]-1-1 0.03 0.71 -30 Lower Shell Forging

• 0.06 0.76 37.0°F 0' -30 Heat# [500102/50C97]-l-l 13.2°Fc) Circumferential Weld * (Intermediate Shell to L ower Shell) 0.03 0.67 41.opb> 40 Weld Seam WF-562 26.lFc> Heat# 442011 Circumferential Weld* (Nozzle Shell to Intermediat e Shell) 0.04 0.46 54.0°phl -25 Weld Seam WF-645 Heat#H4498

• Beltline Region M a terials a) The Initial RT NOT 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 Temperature (ART) Values at l/4T and 3/4T Locations for 32 EFPv**> Surface Fluence 32EFPY Reactor Vessel Material (n/crn 2 , E>l.O MeV) l/4T ART (°F) 3/4T ART (°F) Nozzle Shell Forging 0.559 x 10 19 66 54 Intermediate Shell Forging 1.73 x 10 19 JO -I Lower Shell Forging 1.73 x 10 19 41 24 -+Using non-credible surveillance data 1.73 x 10 19 1 l Nozzle lo Intermediate Shell Forging 0.559x 10 19 Circ. Weld Seam 51 24 (Heat # H4498) Intermediate to Lower Shell Forging Circ. Weld Seam 1.67 x 10 19 122 99 (Heat# 442011) -+Using credible surveillance data 1.67 x 10 19 92 78 Notes: (a) The source document containing detailed calculations is CN-AMLRS-10-7 (Reference 14), Tables 5.3.1-3 and 5.3.1-4. The ART values summarized in this table utilize the most recent lluence projections and materials data, but were not used in development of the Prr limit c urves. See Figures 2.1 and 2.2 of this PTLR for the ART values used in development of the Pff limit curves. 1 5 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 SP-7056 Parameter Values Operating Time 32 EFPY l/4T ART (°F) 3/4T ART(0 F) Chemistry Factor, CF (°F) 26.0 26.0 Fluence(f), n/cm..: 3.40xl0 18 l.23xl0 18 (E>l.0 Mevib> Fluence Factor, FF 0.703 0.460 ART Nor-CFxFF(°F) 18.3 12.0 Initial RT NOT .* 1(°F) 30 30 Marl?in, M(°F) 18.3 12.0 ART= I+(CF*FF)+M, °F 67 54 per RO l.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 fluencc (E>l.0 Mev) =5.67xl0 18 n/cm 2 at 32 EFPY (Reference 11). 16 BRAIDWOOD

-UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.5 RTPTS Calculation for Braidwood Unit 2 Beltline Region Materials at EOL (32 EFPYia.b)

R.G.1.99, CF Flue nee IRT <c) Reactor Vessel Material Rev.2 FF NDT Position (oF) (nlcm 1 , E>l.O MeV) (oF) Nozzle Shell Forging 1.1 26 0.559 x 10 19 0.8373 30 Intermediate Shell Forging I.I 20 1.73 x 10 19 1.1508 -30 Lower Shell Forging l.l 37 l.73x 10 19 1.1508 Using non-credible surveillance data 2.l 13.2 1.73 x 10 1 9 1.1508 -30 Nozzle to Intermediate Shell Forging Circ. Weld Seam 1.1 54 0.559x 10 19 0.8373 -25 (Heal # H4498) Intermediat e 10 Lower Shell Forging Circ.

Weld Seam 1.1 41 1.67 x 10 19 1.1413 40 (Heal # 442011) -Using credible surveillance data 2.1 26.l 1.67 x 10 19 l.1413 40 (a) The lO CFR 50.61 methodology was utilized in the calcul a tion of the RT !"TS values. (b) The source document containing detailed calculations is CN-AMLRS-10-7 (Reference 14), Table 5.5-2. (c) Initial RT NOT values are based on measured data. Henc e, Gu= 0°F. 4RTNDT Gu Cc) al<dl (of) (OF) (OF) 21.8 0 10.9 23.0 0 11.5 42.6 0 17 15.2 0 7.6 45.2 0 22.6 46.8 0 23.4 29.8 0 14 Margin (Of) 21.8 23.0 34 15.2 45.2 46.8 28 (d) Per the guidance of 10 CFR 50.61, the base metal a A = 17°F for Position 1.1 (without surveillance data) and for Position 2.1 with non-credible surveillance data; the weld metal GA= 28°F for Position I.I (without surveillance data) and with credible surveillance d ata GA= 14°F for Position 2.1. Howev er , a A n eed not exceed O.S*aRT NDT* 17 RTPTS (oF) 74 16 47 0 65 134 98 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-10-MUR-197, Revision 0, "Low Temperature Overpressure Protection (LTOP) System Evaluation Final Letter Report," M.P. Rudakewiz, September 8, 2010. 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 of Amendments:

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-2006-0051, "Transmittal of Braidwood Unit l 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, "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, 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. 1 8 BRAIDWOOD

-UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 12. WCAP-16143-P, Revision l, .. Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units 1 and 2," W. Bamford, et al., October 2014. 13. Westinghouse Letter to Exelon Nuclear , CCE-07-24 , .. Braidwood Unit 1and2 RCS HU/CD Limit Curve Table Values," dated February 15, 2007. 14. Westinghouse Calculation Note CN-AMLRS-10-7 , Revision O. "Braidwood Units I and 2 Measurement Uncertainty Recapture (MURJ Uprate: Reactor Vessel Integrity Evaluations." A.E. Leicht, September 2010, and Westinghouse evaluation MCOE-LTR-13-102 Rev. 0, .. Byron and Braidwood Closure Head/Vessel Flange Region: MUR Uprate Assessment," November 2013. l9