BYRON 2014-0003, Pressure and Temperature Limits Report (PTLR) for Measurement Uncertainty Recapture (Mur) Power Uprate

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Pressure and Temperature Limits Report (PTLR) for Measurement Uncertainty Recapture (Mur) Power Uprate
ML14044A019
Person / Time
Site: Byron  Constellation icon.png
Issue date: 02/13/2014
From: Kearney F
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
1.10.0101, 2.01.0300, BYRON 2014-0003
Download: ML14044A019 (45)
v • d • e


Text

{{#Wiki_filter:Byron Generating Station __f__j 4450 North German Church Road Exe[on Generat i on February 13, 2014 LTR: BYRON 201 4-0003 File: 2.01 .0300 1.10.0101 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 Byron Station, Units 1 and 2 Facility Operating License Nos, NPF-37 and NPF-66 NRC Docket No, STN 50-454 and 50-455

Subject:

Pressure and Temperature Limits Report (PTLR) for Measurement Uncertainty Recapture (MUR) Power Uprate Byron Station, Units 1 and 2

References:

Letter from Craig Lambert (Exelon Generation Company, LLC) to U. S. NRC, Request for License Amendment Regarding Measurement Uncertainty F ecapture Power Uprate, dated June 23, 2011 In accordance with Technical Specification 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), we are submitting the January 2014 revisions to the Byron Station Units 1 and 2 PTLR documents. The purpose is to address Pressure and Temperature impacts due to the Measurement Uncertainty Recapture (MUR) Power Uprate License Amendment Request for Byron Units 1 and 2 submitted under letter RS-1 1-099, dated June 23, 2011. Should you have any questions concerning this report, please contact Steven Gackstetter, Regulatory Assurance Manager, at (815) 406-2800. Respectfully, Site Vice Pfsident Byron NuclearGenerating Station FAK!GC/sg Attachments: 1 Byron Unit 1 Pressure and Temperature Limits Report, January 2014 2 Byron Unit 2 Pressure and Temperature Limits Report, January 2014 cc: Regional Administrator NRC Region Ill NRC Senior Resident Inspector Byron Station

BYRON UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (January 2014)

BYRON UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents Section Page 1.0 Introduction 1 2.0 RCS Pressure and Temperature Limits 1 2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3) 1 3.0 Low Temperature Over Pressure Protection and 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 17 I

BYRON UNIT 1-PRESSURE AND TEMPERATURE LIMITS REPORT List of Figures Figure Page 2.1 Byron Unit 1 Reactor Coolant System Heatup Limitations (Heatup 3 Rates of 100°F/br) Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 2.2 Byron Unit 1 Reactor Coolant System Cooldown Limitations 4 (Cooldown Rates of 0, 25, 50, and 100°F/br) Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 3.1 Byron Unit 1 Nominal PORV Setpoints for the Low Temperature 8 Overpressure Protection (LTOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) 11

BYRON UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT List of Tables Table Page 2.la Byron Unit 1 Heatup Data Points at 32 EFPY (Without Margins 5 for Instrumentation Errors) 2.lb Byron Unit 1 Cooldown Data Points at 32 EFPY (Without 6 Margins for Instrumentation Errors) 3.1 Data Points for Byron Unit 1 Nominal PORV Setpoints for the 9 LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) 4.1 Byron Unit 1 Surveillance Capsule Withdrawal Summary 11 5.1 Byron Unit 1 Calculation of Chemistry Factors Using 13 Surveillance Capsule Data 5.2 Byron Unit 1 Reactor Vessel Material Properties 14 5.3 Summary of Byron Unit 1 Adjusted Reference Temperature 15 (ART) Values at 1/4T and 3/4T Locations for 32 EFPY 5.4 RTPTS Calculation for Byron Unit 1 Beitline Region Materials at 16 EOL (32 EFPY) 111

BYRON UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) This Pressure and Temperature Limits Report (PTLR) for Byron Unit 1 has been prepared in accordance with the requirements of Byron TS 5.6.6 (RCS Pressure and Temperature Limits Report). Revisions to the PTLR shall be provided to the NRC after issuance. The Technical Specifications addressed in this report are listed below: TS-LCO 3.4.3 RCS Pressure and Temperature (PIT) Limits; and TS-LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System. 2.0 RCS Pressure and Temperature Limits This section provides the Byron Unit 1 Heatup and Cooldown Limitations. The PTLR limits for Byron Unit 1 were developed using a methodology specified in the Technical Specifications. The methodology listed in WCAP-14040-NP-A, Revision 2 (Reference 1) was used with the following exceptions: a) Optional use 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 1, c) Use of ASME Code Case N-588, Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessel, Section XI, Division 1, and d) Elimination 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 6, 10, 11 and 12. WCAP-15391, Revision 1, Reference 7, provides the basis for the Byron Unit 1 P/T curves, along with the best estimate chemical compositions, fluence projections, and adjusted reference temperatures used to determine these limits. The weld metal data integration for Byron and Braidwood Units 1 and 2 is documented in Reference 2. WCAP-16143-P, Reference 13, documents the technical basis for the elimination of the flange requirements. 2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3) 2.1.1 The RCS temperature rate-of-change limits defined in Reference 7 are: a) A maximum heatup of 100°F in any 1-hour period. b) A maximum cooldown of 100°F in any 1-hour period, and c) A maximum temperature change of less than or equal to 10°F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves. 1

BYRON UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 2.1.2 The RCS P/T limits for heatup, inservice hydrostatic and leak testing, and criticality are specified by Figure 2.1 and Table 2.la. The RCS P/T limits for cooldown are shown in Figure 2.2 and Table 2.lb. These limits are defined in WCAP-15391, Rev. 1 (Reference 7). Consistent with the methodology described in Reference 1, the RCS P/T limits for heatup and cooldown shown in Figures 2.1 and 2.2 are provided without margins for instrument error. These limits were developed using ASME Boiler and Pressure Vessel Code Section XI, Appendix G, Article G-2000, 1996 Addenda. The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in 10 CFR 50, Appendix G. The P/T limits for core operation (except for low power physics testing) are that the reactor vessel must be at a temperature equal to 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 P/T curve for heatup and cooldown. 2

BYRON UNIT 1-PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING LIMITING ART VALUES AT 32 EFPY: 114T, 106°F 314T, 97°F 2500 L 302616034 2250 est Limit 2000 1750 UnaccepDable Acceptable Operation 1 I Operation 1500 Heatup Rate 1250 1OODeg.)Hr - CriticalLimit 1000 750 500 4 Fticaiity Limit based on BOItUP inservice hydrostatic test 250 Temp. temperature (166 F) for the service period up to 32 EFPY 0 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F) Figure 2.1 Byron Unit 1 Reactor Coolant System Heatup Limitations (Heatup rates of 100°F/hr) Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 3

BYRON UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING LIMITING ART VALUES AT 32 EFPY: 114T, 106°F 3/4T, 97°F 2500 302616034 2250 2000 Unacceptab] Operation 1750 Acceptable Operation 1500 F-------- 1250 1000 Cooldown Rates (FIHr) steady-state, 

                             -25, 750                     -50, and
                             -100 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F) Figure 2.2 Byron Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown rates of 0,25,50 and 100°FIhr) Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 4

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BYRON UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 3.0 Low Temperature Overpressure Protection and Boltup This section provides the Byron Unit 1 power operated relief valve lift settings, low temperature overpres sure 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 (PORV5) shall each have maximum lift settings in accordance with Figure 3.1 and Table 3.1. These limits are based on References 3 and 5. The LTOP setpoints are based on PIT limits that were established in accordance with 10 CFR 50, Appendix G without allowance for instrumentation error. The LTOP setpoints were developed using the methodology described in Reference 1. The LTOP PORV nominal lift settings shown in Figure 3.1 and Table 3.1 account for appropriate instrument error. 3.2 LTOP Enable Temperature The required enable temperature for the PORVs shall be 350°F RCS temperature. (Byron Unit 1 procedures governing the heatup and cooldown of the RCS require the arming of the LTOP System for RCS temperature of 3 50°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 LTOP system arming at power. 3.3 Reactor Vessel Boltup Temperature (Non-Technical Specification) The minimum boltup temperature for the Reactor Vessel Flange shall be 60°F. Boltup is a condition in which the Reactor Vessel head is installed with tension applied to any stud, and with the RCS vented to atmosphere (Reference 7). 7

BYRON UNIT 1 - PRESSURE AND TEMPERATURE LIMITS REPORT 2500 2250 2000 1750 C, 0 0. 1500 1250 0 0.

  • 1000 2

0 z 750 500 250 0 0 50 100 150 200 250 300 350 400 450 Auctioneered Low RCS Temperature (DEG. F) Figure 3.1 Byron Unit 1 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) 8

BYRON UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 3.1 Data Points for Byron Unit 1 Nominal PORV Setpoints for the LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) PCV-455A PCV-456 (1TY-0413M) (1TY-0413P) AUCTIONEERED LOW RCS PRESSURE AUCTIONEERED LOW RCS PRESSURE RCS TEMP. (DEG. F) (PSIG) RCS TEMP. (DEG. F) (PSIG) 60 541 60 595 300 541 300 595 400 2335 400 2335 Note: To determine nominal lift setpoints for RCS Pressure and RCS Temperatures greater than 300°F, linearly interpolate between the 300°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

BYRON UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 4.0 Reactor Vessel Material Surveillance Program The pressure vessel material surveillance program (Reference 14) is in compliance with Appendix H to 10 CFR 50, Reactor Vessel Radiation Surveillance Program. The material test requirements and the acceptance standards utilize the reference nil-ductility temperature, RTNDT, which is determined in accordance with ASME Boiler and Pressure Vessel Code, Section ifi, NB-2331. The empirical relationship between RTNDT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, Protection Against Non-Ductile Failure, to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E185-82. The third and final reactor vessel material irradiation surveillance specimens (Capsule W) have been removed and analyzed to determine changes in the reactor vessel 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. 10

BYRON UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.1 Byron Unit 1 Surveillance Capsule Withdrawal Summary Capsule Fluence Capsule Lead Factor Withdrawal EFPY° Location , E> 1.0 MeV) 2 (n/cm U 58.5° 4.05 1.18 0.409 x i0 X 23 8.5° 4.09 5.67 1.49 x 1019 W 121.5° 4.08 9.27 2.26 x 1019 z 301.5° 4.11 14.59 (EOC 12) 3.34 x v(c) 61.0° 3.89 14.59 (EOC 12) 3.16 x 1019 y(c) 3.85 18.81 (EOC 15) 3.97 x i0 241.0° Notes: (a) Source document is CN-AMLRS-lO-8 (Reference 4), Table 5.7-3. (b) Effective Full Power Years (EFPY) from plant startup. (c) Standby Capsules Z, V, and Y were removed and placed in the spent fuel pooi. 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 determine the effect of neutron irradiation on the reactor vessel surveillance materials during the period of extended operation. 11

BYRON UNIT 1 PRESSURE AN]) 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 PIT 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 Byron Unit 1 adjusted reference temperature (ART) values at the 114T and 314T locations for 32 EFPY. Table 5.4 provides the RTp 5 values for Byron Unit 1 for 32 EFPY obtained from Reference 4. 12

BYRON UNIT 1 - PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.1 Byron Unit 1 Calculation of Chemistry Factors Using Surveillance Capsule Data (a) f(b) Capsule ZiRTN1yr FF*LRTN1JT Material Capsule FF 2 FF 

                                  ,E 2

(n/cm > 1.0 MeV) (°F) (°F) U 4.09x 1019 0.752 28.55 21.47 0.57 Intermediate X 1.49x 1019 1.110 9.82 10.90 1.23 Shell Forging (Tangential) W 2.26 x 1.221 49.20 60.06 1.49 U 0.409 x l0 0.752 18.52 13.93 0.57 Intermediate X 1.49 x i0 1.110 53.03 58.89 1.23 Shell Forging (Axial) W 2.26 x l0 1.22 1 29.34 35.82 1.49 Sum: 201.06 6.58 CF IS Forging = * (201.06) ÷ (6.58) 30.6°F tRTT) ÷ (FF ) 2 = = 11.22 U 4.09 x 1019 0.752 (5.61) 8.44 0.57 Byron Unit 1 Surveillance 80.22 Weld Material X 1.49 x i0 1.110 (40.11) 89.08 1.23 (Heat #442002) 102.68 W 2.26 x 1019 1.221 (51.34) 125.34 1.49 16.88 U 0.406 x 0.750 (8.44) 12.66 0.56 Byron Unit 2 Surveillance 57.76 Weld Material W 1.20x 1019 1.051 (28.88) 60.70 1.10 (Heat #442002) 108.02 X 2.18 x 1019 1.211 (54.01) 130.86 1.47 SUM: 427.08 6.42 

                                                                   ÷ (2)

CF Weld Metal =

  • z\RTNi) = (427.08) ÷ (6.42) = 66.5°F Notes:

a) Source document is CN-AMLRS-10-8 (Reference 4), Table 5.2-1. b) f = fluence; L\RTNDT values are the measured 30 ft-lb shift values taken from Reference 15. ARTNDT values for the surveillance weld data are adjusted by a ratio of 2.0 (pre-adjusted values are listed in parentheses). 

                               =         O.1Olog c)         = fluence factor 13

BYRON UNIT 1-PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.2 Byron Unit 1 Reactor Vessel Material Properties (a) Initial Material Description Cu (%) Ni (%) oFb RT NDT Closure Head Flange 124K358VA1 -- 0.74 60 Vessel Flange 123J219VA1 -- 0.73 10 Nozzle Shell Forging 123J218 0.05 0.72 30 Intermediate Shell Forging 5P-5933 0.04 0.74 40 Lower Shell Forging 5P-5951 0.04 0.64 10 Intermediate to Lower Shell Forging Circ. 0.04 0.63 -30 Weld_Seam WF-336_(Heat # 442002) Nozzle Shell to Intermediate Shell Forging 0.03 0.67 10 Circ. Weld Seam WF-501 (Heat # 442011) Byron Unit 1 Surveillance Program 0.02 0.69 Weld Metal_(Heat # 442002) Byron Unit 2 Surveillance Program 0.02 0.71 Weld_Metal_(Heat # 442002) Braidwood Units 1 & 2 Surveillance 0.67, 003 Program Weld Metals (Heat #442011) 0.71 -- a) Reference 7. b) The initial RT NDT values for the plates and welds are based on measured data. 14

BYRON UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.3 Summary of Byron Unit 1 Adjusted Reference Temperature (ART) Values at 1/4T and 314T Locations for 32 EFPY (a) Surface Fluence 32 EFPY Reactor Vessel Material 

                                           , E> 1.0 MeV) 2 (n/cm                1/4T ART (°F)  3/4T ART (°F)

Nozzle Shell Forging 0.598 x 1019 74 59 Intermediate Shell Forging 1.77 i 93 78

  • Using non-credible surveillance data 1.77x l0 102 85 Lower Shell Forging 1.77 x 1019 63 48 Nozzle to Intermediate Shell Forging Circ. Weld Seam (Heat #442011) 0.598 x iO 9 69 49
  • Using credible Braidwood Units 0.598 x i 47 35 1 and 2 surveillance data Intermediate to Lower Shell Forging Circ. Weld Seam (Heat # 442002) 1.72x iO 9 79 49 Using credible surveillance data 1.72 x 65 46 Note:

(a) The source document containing detailed calculations is CN-AMLRS- 10-8 (Reference 4), Tables 5.3.1-1 and 5.3.1-2. The ART values summarized in this table utilize the most recent fluence projections and materials data, but were not used in development of the P/T limit curves. See Figures 2.1 and 2.2 of this PTLR for the ART values used in development of the P/T limit curves. 15

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.4 (a,b) RTPTS Calculation for Byron Unit 1 Beitline Region Materials at EOL (32 EFPY) (C) (C) (d) R.G. 1.99, CF Fluence FF IRTNTD G, Margin RTPTS Rev. 2 (°F) 2 (n/cm 

                                                                    , E> 1.0 MeV)

Reactor Vessel Material (°F) (°F) (°F) (°F) (°F) Position Nozzle Shell Forging 1.1 31 0.598 x 1019 0.8560 30 26.5 0 13.3 26.5 83 Intermediate Shell Forging 1.1 26 1.77 x 1.1569 40 30.1 0 15.0 30.1 100 

       .Using non-credible 2.1        30.6           1.77 x 1019       1.1569         40        35.4         0           17      34        109 surveillance data Lower Shell Forging                 1.1         26            1.77 x i0 19         1.1569         10        30.1         0          15.0    30.1       70 Nozzle to Intermediate Shell Forging Circ. Weld Seam                1.1         41           0.598 x 1019       0.8560         10        35.1         0          17.5    35.1       80 (Heat #442011)
  • Using credible Braidwood Units 2.1 26.1 9 0.598 xlO 0.8560 10 22.3 0 11.2 22.3 55 1 and_2_surveillance data Intermediate to Lower shell Forging Circ Weld Seam 1.1 54 1.72 x 1.1492 -30 62.1 0 28 56 88 (Heat # 442002)
 +Using credible surveillance data              2.1        66.5           1.72 x            1.1492        -30        76.4         0           14      28        74 Notes:

(a) The 10 CFR 50.61 methodology was utilized in the calculation of the RT 5 values. (b) The source document containing detailed calculations is CN-AMLRS- 10-8 (Reference 4), Table 5.5-1. (c) Initial RTNDT values are based on measured data. Hence a 0°F.= (d) Per the guidance of 10 CFR 50.61, the base metal a =17°F for Position 1.1 and for Position 2.1 with non-credible surveillance data; the weld metal GA = 28°F for Position 1.1 (without surveillance data) and with credible surveillance data a =14°F for Position 2.1. However, a A need not to exceed 0.5*ARTNm 16

BYRON 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. WCAP- 14824, Revision 2, Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld Metal Integration for Byron & Braidwood, November 1997 with Westinghouse errata letters CAE-97-220, dated November 26, 1997 and CAE 23 1/CCE-97-3 14 and CAE-97-233/CCE-97-3 16, dated January 6, 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. Westinghouse Calculation Note CN-AMLRS-10-8, Revision 0, Byron Units 1 and 2 Measurement Uncertainty Recapture (MUR) Uprate: Reactor Vessel Integrity Evaluations, A. E. Leicht, September 2010.
5. Byron Station Design Information Transmittal DIT-BYR-06-046, Transmittal of Byron Unit 1 and Unit 2 Temperature and Pressure Uncertainties for Low Temperature Overpressure System (LTOPS) Power Operated Relief Valves (PORVS), David Neidich, August 15, 2006.
6. NRC Letter from R. A. Capra, NRR, toO. D. Kingsley, Commonwealth Edison Co., Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, Acceptance for Referencing of Pressure Temperature Limits Report (TAC Numbers M98799, M98800, M98801, and M98802), January 21, 1998.
7. WCAP- 15391, Revision 1, Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation, T. J. Laubham, et al., November 2003.
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 1 and 2, and Braidwood Stations, Units 1 and 2, dated August 8, 2001.

17

BYRON UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT

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- 16 143-P, Revision 0, Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units 1 and 2, W. Bamford, et al., November 2003.
12. WCAP-9517, Commonwealth Edison Company, Byron Station Unit 1 Reactor Vessel Surveillance Program, J.A. Davidson, July 1979.
13. WCAP-15123, Revision 1, Analysis of Capsule W from Common Wealth Edison Company Byron Unit 1 Reactor Vessel Radiation Surveillance Program, T.J. Laubham, et al, January 1999.

18

BYRON UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (January 2014)

BYRON UNIT 2 PRESSURE AN]) TEMPERATURE LIMITS REPORT Table of Contents Section Page 1.0 Introduction 1 2.0 RCS Pressure and Temperature Limits 1 2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3) 1 3.0 Low Temperature Over Pressure Protection and 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 17

BYRON UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT List of Figures Figure Page 2.1 Byron Unit 2 Reactor Coolant System Heatup Limitations (Heatup 3 Rates of 100°F/br) Applicable for 30.5 EFPY (Without Margins for Instrumentation Errors) 2.2 Byron Unit 2 Reactor Coolant System Cooldown Limitations 4 (Cooldown Rates of 0, 25, 50, and 100°F/br) Applicable for 30.5 EFPY (Without Margins for Instrumentation Errors) 3.1 Byron Unit 2 Nominal PORV Setpoints for the Low Temperature 8 Overpressure Protection (LTOP) System Applicable for 30.5 EFPY (Includes Instrumentation Uncertainty) 11

BYRON UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT List of Tables Table Page 2.la Byron Unit 2 Heatup Data Points at 30.5 EFPY (Without Margins 5 for Instrumentation Errors) 2.lb Byron Unit 2 Cooldown Data Points at 30.5 EFPY (Without 6 Margins for Instrumentation Errors) 3.1 Data Points for Byron Unit 2 Nominal PORV Setpoints for the 9 LTOP System Applicable for 30.5 EFPY (Includes Instrumentation Uncertainty) 4.1 Byron Unit 2 Surveillance Capsule Withdrawal Summary 11 5.1 Byron Unit 2 Calculation of Chemistry Factors Using 13 Surveillance Capsule Data 5.2 Byron Unit 2 Reactor Vessel Material Properties 14 5.3 Summary of Byron Unit 2 Adjusted Reference Temperatures 15 (ART) Values at 1/4T and 3/4T Locations for 32 EFPY 5.4 RTprs Calculation for Byron Unit 2 Beltline Region Materials at 16 EOL (32 EFPY) in

BYRON UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) This Pressure and Temperature Limits Report (PTLR) for Byron Unit 2 has been prepared in accordance with the requirements of Byron TS-5.6.6 (RCS Pressure and Temperature Limits Report). Revisions to the PTLR shall be provided to the NRC after issuance. The Technical Specifications addressed in this report are listed below: TS-LCO 3.4.3 RCS Pressure and Temperature (PIT) Limits; and TS-LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System. 2.0 RCS Pressure and Temperature Limits This section provides the Byron Unit 2 Heatup and Cooldown Limitations. The PTLR limits for Byron Unit 2 were developed using a methodology specified in the Technical Specifications. The methodology listed in WCAP- 14040-NP-A, Revision 2 (Reference 1) was used with the following exception: a) 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 1, c) Use of ASME Code Case N-588, Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels, Section XI, Division 1, and d) Elimination of the flange requirements documented in WCAP-16 143-P. This exception to the methodology in WCAP-14040-NP-A, Revision 2 has been reviewed and accepted by the NRC in References 8, 10, 11 and 12. WCAP-15392, Revision 2 (Reference 7), provides the basis for the Byron Unit 2 P/T curves, along with the best estimate chemical compositions, fluence projections, and adjusted reference temperatures used to determine these limits. The weld metal data integration for Byron and Braidwood Units 1 and 2 is documented in Reference 2. WCAP- 16 143-P, Reference 13, documents the technical basis for the elimination of the flange requirements. 2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3) 2.1.1 The RCS temperature rate-of-change limits defined in Reference 7 are:

a. A maximum heatup of 100°F in any 1-hour period,
b. A maximum cooldown of 100°F in any 1-hour period, and
c. A maximum temperature change of less than or equal to 10°F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

1

BYRON UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 2.1.2 The RCS PIT limits for heatup, inservice hydrostatic and leak testing, and criticality are specified by Figure 2.1 and Table 2.1 a. The RCS P/T limits for cooldown are shown in Figure 2.2 and Table 2. lb. These limits are defined in WCAP-15392, Revision 2 (Reference 7). Consistent with the methodology described in Reference 1, the RCS PIT 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 G-2000, 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 PIT 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 PIT curve for heatup and cooldown. 2

BYRON UNIT 2 - PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-447 & NOZZLE SHELL FORGING LIMITING ART VALUES AT 30.5 EFPY: 1/4T, 107°F (N-588) & 52°F (96 App. G) 314T, 89°F (N-588) & 37°F (96 App. G) 2500 [2EerIim_Version:5.1_Run:12458 LeakTest Limit 2250 2000 f j Acceptable 

                     !                                                Operation Unacceptable 1750 Operation      I 0

(I) Heatup Rate 0. 1500 1250 ZEiZ!ot!!zEZZ 

  .1-i Cs
  . 1000 Cs C) 750 f1,/_

[iticality L mit based on t j temperature (112 F) for the ervice period up to 30.5 EFPY 1 [ 500 ff 250 0 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F) Figure 2.1 Byron Unit 2 Reactor Coolant System Heatup Limitations (Heatup rates of 100°FIhr) Applicable for 30.5 EFPY (Without Margins for Instrumentation Errors) 3

BYRON UMT 2 PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITiNG MATERIAL: CIRCUMFERENTIAL WELD WF-447 & NOZZLE SHELL FORGING LiMITiNG ART VALUES AT 30.5 EFPY: 114T, 107°F (N-588) & 52°F (96 App. G) 314T, 89°F (N-588) & 37°F (96 App. G) 2500

  • Operlim Version:5.1 Run:1245J

[ 2250 Unacceptable Operation L 1 ,. I 2000 :_H 1750 1500 2! j 1250 I I 1000 Rates FIHr steady-state 

                               -25
                               -50
                               .100 750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 

                           -.             Moderator Temperature (Deg. F)

Figure 2.2 Byron Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0,25,50 and 100°F/hr) Applicable for 30.5 EFPY (Without Margins for Instrumentation Errors) 4

0 C C Ui C UiCUiC(JC C Ui C C C rj t%) b tJ t) I- .I. C..) t.) . C 00 - C* Ui Ui .- U) U) t.) tJ k.) . - C C j 00 -1 k.) t3 J C. 00 00 C.C -. LJ . U) C 00 C Ui . - U) c, C - C U) . 4 J J C - U) 0. -. Ui C J 000000 C a - C C .0 .0 00 00 - 0 0. Ui Ui 4 . U) U) J k) CUiUiCUiUiUij i- - t-. O IT. l) I.) t) t-J t-a IJ - - .- . 4). U) J C 00 . 0 Ui Ui 4). U) U) .) -..) .) C 00 4). L.) t) LJ .0 00 00 .0 i- .1 -. i U) C 00 Oi Ui 4 .) -. ci C .1 C U) -) . i-) -) C U) 0 - Ui C k) t-J 000000 I C INJ 00 C ci UiC r

BYRON UNIT 2-PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.lb Byron Unit 2 Cooldown Data Points at 30.5 EFPY (Without Margins for Instrumentation Errors) Cooldown Curves Steady State 25 °F Cooldown 50 °F Cooldown 100 °F Cooldown T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 60 0 60 0 60 0 60 0 60 1045 60 1036 60 1033 65 1092 65 1088 70 1143 75 1200 80 1263 85 1332 90 1409 95 1494 100 1587 105 1691 110 1805 115 1932 120 2071 125 2226 130 2396 Note: For each cooldown rate, the steady-state pressure values shall govern the temperature where no allowable pressure values are provided. 6

BYRON UNIT 2-PRESSURE AND TEMPERATURE LIMITS REPORT 3.0 Low Temperature Overpressure Protection and Boltup This section provides the Byron Unit 2 power operated relief valve lift settings, low temperature overpressure protection (LTOP) system arming temperature, and minimum reactor vessel boltup temperature. 3.1 LTOP System Setpoints (LCO 3.4.12) The power operated relief valves (PORVs) shall each have maximum lift settings in accordance with Figure 3.1 and Table 3.1. These limits are based on References 3 and 6. The LTOP setpoints are based on P/T limits which were established in accordance with 10 CFR 50, Appendix G without allowance for instrumentation error and in accordance with the methodology described in Reference 1. The LTOP PORV nominal lift settings shown in Figure 3.1 and Table 3.1 account for appropriate instrument error. 3.2 LTOP Enable Temperature The required enable temperature for the PORVs shall be S 350°F RCS temperature. (Byron 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 LTOP system arming at power. 3.3 Reactor Vessel Boltup Temperature (Non-Technical Specification) The minimum boltup temperature for the Reactor Vessel Flange shall be 60°F. Boltup is a condition in which the Reactor Vessel head is installed with tension applied to any stud, and with the RCS vented to atmosphere (Reference 7). 7

BYRON UNIT 2 - PRESSURE AND TEMPERATURE LIMITS REPORT 2335 psig 2250 2000 1750 ci) a. 1500 Unacceptable Operation 1250 0 0.

  • 1000 PCV 456 0

z 750 639 psig 599 psig PCV 455A 250 0 0 50 100 150 200 250 300 350 400 450 Auctioneered Low RCS Temperature (DEG. F) Figure 3.1 Byron Unit 2 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for the 30.5 EFPY (Includes Instrumentation Uncertainty) 8

BYRON UNIT 2 PRESSURE AN]) TEMPERATURE LIMITS REPORT Table 3.1 Data Points for Byron Unit 2 Nominal PORV Setpoints for the LTOP System Applicable for 30.5 EFPY (Includes Instrumentation Uncertainty) PCV-455A PCV-456 (2TY-041 3M) (2TY-041 3P) AUCTIONEERED LOW RCS PRESSURE AUCTIONEERED LOW RCS PRESSURE RCS TEMP. (DEG. F) (PSIG) RCS TEMP. (DEG. F) (PSIG) 50 599 50 639 300 599 300 639 400 2335 400 2335 Note: To determine nominal lift setpoints for RCS Pressure and RCS Temperatures greater than 300°F, linearly interpolate between the 300°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

BYRON 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 standard utilize the reference nil-ductility temperature, RTNDT, which is determined in accordance with ASME Boiler and Pressure Vessel Code Section III, NB-233 1. The empirical relationship between RTNDT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, Protection Against Non-Ductile Failure, to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E185-82. The third and final reactor vessel material irradiation surveillance specimens (Capsule W) have been removed and analyzed to determine changes in the reactor vessel 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 sunmiary is provided in Table 4.1. 10

BYRON UNIT 2-PRESSURE AN]) TEMPERATURE LIMITS REPORT Table 4.1 Byron Unit 2 Surveillance Capsule Withdrawal Summary Capsule Capsule Lead Factor Withdrawal EFPY Fluence Location , E> 1.0 MeV) 2 (n/cm U 58.5° 4.02 1.19 0.406 x 1019 W 12 1.5° 4.07 4.67 1.20 x X 238.5° 4.14 8.63 2.18 x 1019 z(c) 301.5° 4.11 14.28 (EOC 11) 3.25 x v 61.0° 3.88 14.28 (EOC 11) 3.07 x 1019 y(c) 241.0° 3.88 20.05 (EOC 15) 4.19 x 10 Notes: (a) Source document is CN-AMLRS-lO-8 (Reference 5), Table 5.7-4. (b) Effective Full Power Years (EFPY) from plant startup. (c) Standby Capsules Z, V, and 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 determine the effect of neutron irradiation on the reactor vessel surveillance materials during the period of extended operation. 11

BYRON 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 PIT 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 Byron Unit 2 adjusted reference temperature (ART) values at the 114T and 314T locations for 32 EFPY. Table 5.4 provides the RTpTS values for Byron Unit 2 for 32 EFPY obtained from Reference 5. 12

BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.1 Byron Unit 2 Calculation of Chemistry Factors Using Surveillance Capsule Data Capsule f(b) pj(c) ARTN1JT FF*RTT Material Capsule 2 pj 

                                    , E> 1.0 MeV) 2 (n/cm                                        (°F)           (°F)

U 4.06 x 1019 0.750 0.0 0.00 0.56 Lower Shell Forging W 1.20 1.051 3.65 3.84 1.10 (Tangential) X 2.18x i 1.211 15.75 19.08 1.47 Lower Shell U 04.06 x iO9 0.750 19.76 14.82 0.56 Forging W l.20x l0 1.051 31.88 33.50 1.10 (Axial) X 9 2.18x10 1.211 38.91 47.14 1.47 SUM: 118.38 6.27 1 Forging CF = (FF */RTNDT) ÷ ( FF) = (118.38) ÷ (6.27) = 18.9°F 11.22 U 0.409 x 1019 0.752 (5.61) 8.44 0.57 Byron Unit 1 Surveillance Weld 80.22 X 1.49x 1019 1.110 (40.11) 89.08 1.23 Material 102.68 (Heat #442002) W 2.26 x iO 9 1.22 1 125.34 1.49 (51.34) 16.88 U 0.406 x 1019 0.750 (8.44) 12.66 0.56 Byron Unit 2 Surveillance Weld 57.76 W 1.20x 1019 1.051 (28.88) 60.70 1.10 Material 108.02 (Heat #442002) X 2.18 x 1019 1.211 130.86 1.47 (54.01) SUM: 427.08 6.42 (FF

  • CFweld Metal = zRTNDT) ÷ ( FF
                                                                        ) = (427.08) ÷ (6.42) = 66.5°F 2

Notes: a) Source document is CN-AMLRS-10-8 (Reference 5), Table 5.2-2. b) f = fluence; L.RTNDT values are the measured 30 ft-lb shift values taken from Reference 9. 

            .RTNDT values for the surveillance weld data are adjusted by a ratio of 2.0 (pre-adjusted values are listed in parentheses).

O.1O*log c) FF = fluence factor = d) Measured ARTNDT value was determined to be negative, but physically a reduction should not occur; therefore a conservative value of zero is used. 13

BYRON - UNIT 2 PRESSURE AN]) TEMPERATURE LIMITS REPORT Table 5.2 (a) Byron Unit 2 Reactor Vessel Material Properties 

                      .        .  .                               .        Initial Material Description                      Cu (%) Ni (%)  om      o(b) 1
                                                                        £1 NDT. 1)

Closure Head Flange 5P7382 / 3P6407 -- 0.71 0 Vessel Flange 124L556VA1 -- 0.70 30 Nozzle Shell Forging 4P-6 107 0.05 0.74 10 Inter. Shell Forging [49D329/49C297]-l-l 0.01 0.70 -20 Lower Shell Forging [49D330149C298]-1-1 0.06 0.73 -20 Circumferential Weld WF-447 (HT# 442002) 0.04 0.63 10 Upper Circumferential Weld WF-562 (HT# 442011) 0.03 0.67 40 Byron Unit 1 Surveillance Program 002 0 69 Weld_Metal_(Heat # 442002) Byron Unit 2 Surveillance Program 0 02 071 Weld Metal_(Heat #442002) Braidwood Units 1 & 2 Surveillance Program 0.67, 003 Weld Metal (Heat #442011) 0.71 a) Reference 7. b) Initial RTNDT values are based on measured data. 14

BYRON UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.3 Summary of Byron Unit 2 Adjusted Reference Temperatures (ART) Values at 1/4T and 314T Locations for 32 EFPY (a) Surface Fluence 32 EFPY Reactor Vessel Material (n/cm2 , E> 1.0 MeV) 1/4T ART (°F) 3/4T ART (°F) Nozzle Shell Forging 0.549 o 53 38 Intermediate Shell Forging 1.76 i 21 9 Lower Shell Forging 1.76 x iO 9 52 34 Using credible surveillance data 1.76 16 8 Nozzle to Intermediate Shell Forging Circ. Weld Seam 0.549 x iO 9 97 77 (Heat #442011) 

  *Using credible Braidwood Units 1 0.549    o           76            64 and 2 surveillance data Intermediate to Lower Shell Forging Circ.WeldSeam                      1.70x i             119           88 (Heat # 442002)
  +Using credible surveillance data             1.70 X 9 iO          105            86 Note:

(a) The source document containing detailed calculations is CN-AMLRS- 10-8 (Reference 5), Tables 5.3.1-3 and 5.3.1-4. The ART values summarized in this table utilize the most recent fluence projections and materials data, but were not used in development of the P/T limit curves. See Figures 2.1 and 2.2 of this PTLR for the ART values used in development of the PIT limit curves. 15

BYRON UNIT 2-PRESSURE A1) TEMPERATURE LIMITS REPORT Table 5.4 RT 5 Calculation for Byron Unit 2 Beltline Region Materials at EOL (32 EFPY) (C) (d) R.G. 1.99, CF Fluence FF IRTNTD ART NTD GA Margin RTPTS Rev. 2 (°F) Reactor Vessel Material , 2 (n/cm (°F) (°F) (°F) (°F) (°F) Position E> 1.0 MeV) Nozzle Shell Forging 1.1 31 0.549 x 1019 0.8323 10 25.8 0 12.9 25.8 62 Intennediate Shell Forging 1.1 20 1.76 x 1019 1.1554 -20 23.1 0 11.6 23.1 26 LowerSheilForging 1.1 37 1.76x iO 9 1.1554 -20 42.7 0 17 34 57 -* Using credible surveillance data 2.1 18.9 1.76 x l0 1.1554 -20 21.8 0 8.5 17 19 Nozzle to Intermediate Shell Forging Circ. Weld Seam 1.1 41 0.549 x i 0.8323 40 34.1 0 17.1 34.1 108 (Heat # 442011) 

 +UsingcredibleBraidwoodUnits 2.1        26.1     9 O.549x10       0.8323          40         21.7          0      10.9      21.7   83 1 and 2 surveillance data Intermediate to Lower shell Forging Circ Weld Seam                1.1         54       1.70x i0       1.1461          10        61.9          0        28       56    128 (Heat # 442002)

+Using credible surveillance data 2.1 66.5 1.70x 1019 1.1461 10 76.2 0 14 28 114 Notes: (a) The 10 CFR 50.61 methodology was utilized in the calculation of the RT 5 values. (b) The source document containing detailed calculations is CN-AMLRS-10-8 (Reference 5), Table 5.5-2. (c) Initial RTNDT values are based on measured data. Hence a, = 0°F. (d) Per the guidance of 10 CFR 50.61, the base metal a =17°F for Position 1.1 (without surveillance data) and with credible surveillance data 

      =                                           =

GA 8.5°F for position 2.1; the weld metal a A 28°F for position 1.1 (without surveillance data) and with credible surveillance data GA =14°F for Position 2.1. However, a A need not to exceed O.S*ARTNTD 16

BYRON 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, Andrachek, J.D.,

et al., January 1996.

2. WCAP- 14824, Revision 2, Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld Metal Integration for Byron & Braidwood, November 1997 with Westinghouse errata letters CAE-97-220, dated November 26, 1997 and CAE 23 1/CCE-97-3 14 and CAE-97-233/CCE-97-316, dated January 6, 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-10398, Commonwealth Edison Company, Byron Station Unit 2 Reactor Vessel Radiation Surveillance Program, Singer, L.R., December 1983.
5. Westinghouse Calculation Note CN-AMLRS-10-8, Revision 0, Byron Units 1 and 2 Measurement Uncertainty Recapture (MUR) Uprate: Reactor Vessel Integrity Evaluations, A. E. Leicht, September 2010.
6. Byron Station Design Information Transmittal DIT-BYR-06-046, Transmittal of Byron Unit 1 and Unit 2 Temperature and Pressure Uncertainties for Low Temperature Overpressure System (LTOPS) Power Operated Relief Valves (PORVS), David Neidich, August 15, 2006.
7. WCAP- 15392, Revision 2, Byron Unit 2 Heatup and Cooldown Limit Curves for Normal Operation, T. J. Laubham, et al., November 2003.
8. NRC Letter from R. A. Capra, NRR, to 0. D. Kingsley, Commonwealth Edison Co., Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, Acceptance for Referencing of Pressure Temperature Limits Report (TAC Numbers M98799, M98800, M98801, and M98802), January 21, 1998.
9. WCAP-15 176, Revision 0, Analysis of Capsule X from Commonwealth Edison Company Byron Unit 2 Reactor Vessel Radiation Surveillance Program, T. J. Laubham, et al., March 1999.
10. 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.
11. 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.

17

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