Text

Pressure and Temperature Limits Report (PTLR)
	ML15323A005
Person / Time
Site:	Byron
Issue date:	11/19/2015
From:	Kanavos M; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	BYRON 201 5-0128
	Download: ML15323A005 (45)
	v • d • e

-

Byron Generating Station Exeton Generation 4450 No th German Church Rd Byron, www.exeloncorp.com November 19, 2015 LTR: BYRON 201 5-0128 File: 2.01.0300 (5A.101) 1.10.0101 (1D.101)

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 Byron Station, Units 1 and 2

References:

Letter from David Gullott (Exelon Generation Company, LLC) to U. S. NRC, License Amendment Request to Utilize WCAP-16143-P, Revision 1, as an Analytical Method to Determine the Reactor Coolant System Pressure and Temperature Limits, dated October 16, 2014 In accordance with Technical Specification 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), we are submitting the November 2015 revisions to the Byron Station Units 1 and 2 PTLR documents. The purpose is to update the PTLRs to reference revision 1 of WCAP-16143-P as applied by the License Amendment Request for Byron Units 1 and 2 submitted under letter RS-14-284, dated October 16, 2014.

Should you have any questions concerning this report, please contact Mr. Douglas Spitzer, Regulatory Assurance Manager, at (815) 406-2800.

Respectfully, Mark E. Kanavos Site Vice President Byron Generating Station MEKIGC/sg Attachments: 1. Byron Unit 1 Pressure and Temperature Limits Report, November 2015

2. Byron Unit 2 Pressure and Temperature Limits Report, November 2015 cc: Regional Administrator NRC Region Ill NRC Senior Resident Inspector Byron Station

BYRON UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

(November 2015)

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

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/hr) Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 3.1 Byron Unit 1 Nominal PORV Setpoints for the Low Temperature $

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 l/4T and 3/4T Locations for 32 EFPY 5.4 RTPTS Calculation for Byron Unit 1 Beltline Region Materials at 16 EOL (32 EFPY)

In

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 0, Article 0-2000, 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-58$, Alternative to Reference Flaw Orientation of Appendix 0 for Circumferential Welds in Reactor Vessel,Section XI, Division 1, and d) Elimination of the flange requirements documented in WCAP-16 143-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 11, 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.

BYRON UNIT 1 PRESSURE AND TEMPERAT URE 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 PIT 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 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 1 PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING LIMITING ART VALUES AT 32 EFPY: 1/4T, 106°F 3/4T, 97°F 2500 2250 2000 1750 1500 a

C) 1250

o 1000 C)

C) 3 U

Cu 750 C

500 250 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°Ffhr)

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 2250 2000 H

L:J 1750 Acceptable L Operation 1500 0

a) 7250 Cooldown Rates, CF/Hr v 1000 a)

Cu I

750 500 6O0F

] Boltup Temp.

250 The lower limit for RCS 0 : 1pressure is -14.7 psig 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

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

Heatup Curve 100 F Heatup Criticality Leak Test Limit Limit T (OF) P (psig) T (°F) P (psig) T (°F) P (psig) 60 -14.7 166 -14.7 149 2000 60 720 166 720 166 2485 65 720 166 720 70 720 166 720 75 720 166 720 80 720 166 720 85 720 166 720 90 720 166 720 95 720 166 723 100 723 166 729 105 729 166 737 110 737 166 749 115 749 166 764 120 764 166 781 125 781 170 802 130 802 175 826 135 826 180 854 140 854 185 886 145 886 190 921 150 921 195 962 155 962 200 1007 160 1007 205 1057 165 1057 210 1113 170 1113 215 1175 175 1175 220 1244 180 1244 225 1321 185 1321 230 1406 190 1406 235 1499 195 1499 240 1603 200 1603 245 1718 205 1718 250 1844 210 1844 255 1984 215 1984 260 2138 220 2138 265 2308 225 2308 5

BYRON UNIT 1-PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.lb Byron Unit 1 Cooldown Data Points at 32 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 (OF) P (psig) T (°F) P (psig) 60 -14.7 60 -14.7 60 -14.7 60 -14.7 60 753 60 709 60 665 60 581 65 769 65 726 65 685 65 606 70 787 70 746 70 706 70 633 75 806 75 767 75 730 75 663 80 827 80 791 80 757 80 697 85 851 $5 817 85 786 85 735 90 877 90 846 90 819 90 777 95 906 95 879 95 855 95 823 100 937 100 914 100 895 100 874 105 973 105 954 105 940 105 931 110 1011 110 997 110 989 115 1054 115 1045 115 1043 120 1102 120 1099 125 1154 130 1212 135 1276 140 1347 145 1425 150 1512 155 1607 160 1713 165 1829 170 1958 175 2101 180 2258 185 2433 Note: For each cooldown rate, the steady-state pressure values shall govern the temperature where no allowable pressure values are provided.

6

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 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 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 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 1 -

PRESSURE AND TEMPERATURE LIMITS REPORT 2335 psig 2000 1750 C,,

1500 a Unacceptable Operation a

° 1250 0

0

1000 C

E PCV-456 0

z 750 500 541 psig 250 0

50 100 150 200 250 300 350 400 450 0

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)

BYRON UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 3.1 Data Points for Byron Unit 1 Nominat 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 12)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 E1$5-$2.

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 (b)

Location (n/cm2 , E> 1.0 MeV)

U 58.50 4.05 1.18 0.409 x i0 X 238.5° 4.09 5.67 1.49 x i09 W 121.5° 4.08 9.27 2.26 x Z 301.5° 4.11 14.59 (EOC 12) 3.34x iO9 V 61.0° 3.89 14.59(EOC12) 3.16x y(c) 241.00 3.85 18.81 (EOC 15) 3.97 x io Notes:

(a) Source document is CN-AMLRS-1O-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 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 1 adjusted reference temperature (ART) values at the 1/4T and 3/41 locations for 32 EFPY.

Table 5.4 provides the RTpT5 values for Byron Unit 1 for 32 EFPY obtained from Reference 4.

12

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.1 Ct Byron Unit 1 Calculation of Chemistry Factors Using Surveillance Capsule Data b)

Capsule f FF(C) ZRTXDT1 FF*ARTNDT Material Capsule FF2 (n/cm2, E> 1.0 MeV) (°f) (°F)

U 4.09 x 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 2.26x 1019 1.221 49.20 60.06 1.49 U 0.409 x IO9 0.752 18.52 13.93 0.57 Intermediate

. X l.49x 1019 1.110 53.03 58.89 1.23 Shell Forging (Axial) W 2.26 x 1019 1.221 29.34 35.82 1.49 Sum: 201.06 6.58 CF IS forging = Z(FF

ARTNDT) ÷ Z(Ff) (201.06)÷ (6.58)= 30.6°F 11.22

. U 4.09 x b9 0.752 (5.61) 8.44 0.57 Byron Unit 1 Surveillance 80.22 Weld Material X J.49x 1019 1.110 (40.11) 89.08 1.23 (Heat #442002) 102.68 W 2.26x i 1.221 (51.34) 125.34 1.49 16.88

. U 0.406 x i0 0.750 (8.44) 12.66 0.56 Byron Unit 2 Surveillance 57.76 Weld Material W 1.20 x iO9 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 CF weld Metal = Z(FF

RTDT) ÷ Z(FF2) (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; RTNDT values are the measured 30 ft-lb shift values taken from Reference 13.

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

FF = fluence factor = O.2i O,iO°Iog f) c)

13

BYRON UNIT 1-PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.2 (a)

Byron Unit 1 Reactor Vessel Material Properties

. . . . Initial Material Description Cu (%) Ni (%) ro1\(b) 11 NDT 1)

Closure Head Flange 124K35$VA1 -- 0.74 60 Vessel Flange 123J219VA1 -- 0.73 10 Nozzle Shell Forging 123J21$ 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, --

0 03 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 (a) 1/4T and 314T Locations for 32 EFPY Surface Fluence 32 EFPY Reactor Vessel Material (n/cm2, E> 1.0 MeV) 1/4T ART (°F)_]_3/4T ART (°F) 0.59$ x 1019 74 59 Nozzle Shell Forging Intermediate Shell forin 1.77x109 93 78

Using non-credible 85 1.77x 1019 102 surveillance data Lower Shell forging 1.77 1019 63 4$

Nozzle to Intermediate Shell Forging 0.598 l0 69 49 Circ. Weld Seam (Heat # 44201 1)

Using credible Braidwood Units 0.598 x iO9 47 35 I and 2 surveillance data Intermediate to Lower Shell Forging l.72x 1019 t 79 49 Circ. Weld Seam (Heat # 442002)

Using credible surveillance data 1.72 x 1019 65 46

]

Note:

ence 4),

(a) The source document containing detailed calculations is CN-AMLRS-1O-8 (Refer recent Tables 5.3.1-1 and 5.3.1-2. The ART values summarized in this table utilize the most fluence projections and materials data, but were not used in development of the PIT 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 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.4 e Region Materials at EOL (32 EFPY)

RTPTS Calculation for Byron Unit 1 Beltiln (C) (c) (d)

Ff IRTNTD ARTNTD Margin RTPTS R.G. 1.99, CF Fluence Rev. 2 (°F) (n/cm2, E> 1.0 MeV) (°F) (°F) (°F) (°F) (°F)

Reactor Vessel Material Position 0.598 x 10 0.8560 30 26.5 0 13.3 26.5 83 Nozzle Shell Forging 1.1 31 1.77 X iOW 1.1569 40 30.1 0 15.0 30.1 100 Intermediate Shell Forging 1.1 26

+Using non-credible 2.1 30.6 1.77 x 1019 1.1569 40 35.4 0 17 34 109 surveillance data 1.77 X i0 1.1569 10 30.1 0 15.0 30.1 70 Lower Shell Forging 1.1 26 Nozzle to Intermediate Shell 35.1 80 1.1 41 0.598 x 1019 0.8560 10 35.1 0 17.5 Forging Circ. Weld Seam (Heat #442011)

+Using credible Braidwood Units 2.1 26.1 0.598 x1019 0.8560 10 22.3 0 11.2 22.3 55 I and 2 surveillance data Intermediate to Lower shell 56 88 1.1 54 1.72x i 1.1492 -30 62.1 0 28 Forging Circ Weld Seam (Heat_#_442002) 1.72x i 1.1492 -30 76.4 0 14 28 74

Using credible surveillance data 2.1 66.5 Notes:

(a) The 10 CFR 50.61 methodology was utilized in the calculation of the RTPTS values.

5.5-I.

(b) The source document containing detailed calculations is CN-AMLRS-l0-8 (Reference 4), Table (c) Initial RTNDT values are based on measured data. Hence a0°F.

2.1 with non-credible surveillance data; the weld metal y28°F for (d) Per the guidance of 10 CFR 50.61, the base metal c = 17°F for Position 1.1 and for Position

, need not to exceed O.5*RTNTD Position 1.1 (without surveillance data) and with credible surveillance data 14°f for Position 2.1. However, 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 elTata letters CAE-97-220, dated November 26, 1997 and CAE 231/CCE-97-314 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. 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 Dff-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, 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, M98$01, and M98802), January 21, 1998.

7. WCAP- 15391, Revision 1, Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation, T. I. 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 MC$696), November 27, 2006.

11. WCAP-16143-P, Revision 1, Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units 1 and 2, W. Bamford, et al., October 2014.

12. WCAP-95 17, Commonwealth Edison Company, Byron Station Unit 1 Reactor Vessel Surveillance Program, l.A. Davidson, July 1979.

13. WCAP-15 123, Revision 1, Analysis of Capsule W from Commonwealth Edison Company Byron Unit 1 Reactor Vessel Radiation Surveillance Program, T.J. Laubham, et al, January 1999.

1$

BYRON UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

(November 2015)

BYRON UNiT 2 PRESSURE AND TEMPERAT URE 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 TEMPERAT URE 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/hr) 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 RTPTS Calculation for Byron Unit 2 Beltline Region Materials at 16 EOL (32 EFPY) 111

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- 16143-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 PIT 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.

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 0, Article 0-2000, 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.

BYRON UNIT 2-PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-447 & NOZZLE SHILL FORGING LIMITING ART VALUES AT 30.5 EFPY: I/4T, 107°F (N-588) & 52°F (96 App. G) 3/4T, 89°f (N582) & 37°F (96 App. G) 2500 2250 2000 1750 U) 1500 0

1250 v 1000

) 750 0

500 250 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 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: 114T, 107°F (N-588) & 52°F (96 App. G) 3/4T, 89°F (N-588) & 37°F (96 App. G) 2500 2250

j Acceptable 2000 .

Operation 1750 c

5 1500 0

0 1250 Cooldown 4 Rates, °F/Hr

1000 steady-state, 0 -25,

-50, and

-100 C.) 750 0

500 250 The lower limit for RCS 0 ._pressure is -14.7 psig mm 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°FIhr) Applicable for 30.5 EFPY (Without Margins for Instrumentation Errors) 4

BYRON UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.la Byron Unit 2 Heatup Data Points at 30.5 EFPY (Without Margins for Instrumentation Errors)

Heatup Curve 100 F Heatup Criticality Leak Test Limit Limit T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 60 -14.7 112 -14.7 95 2000 60 1030 112 1078 112 2485 65 1078 112 1128 70 1128 115 1148 75 1148 120 1152 80 1152 125 1162 85 1162 130 1180 90 1180 135 1205 95 1205 140 1237 100 1237 145 1276 105 1276 150 1323 110 1323 155 1377 115 1377 160 1440 120 1440 165 1511 125 1511 170 1592 130 1592 175 1682 135 1682 180 1784 140 1784 185 1897 145 1897 190 2023 150 2023 195 2120 155 2120 200 2227 160 2227 205 2347 165 2347 210 2480 170 2480 5

BYRON UNIT 2-PRESSURE AND TEMPERAT URE 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 -14.7 60 -14.7 60 -14.7 60 -14.7 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.

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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 PIT 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 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).

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BYRON UNIT 2 -

PRESSURE AND TEMPERATURE LIMITS REPORT 2500 2335 psig 2250 2000 1750 0

0 0

1500 -----

Unacceptable Operation 1250 0

C

.1000 -- ----- --

PCV 456 750 599 psig 500-PCV 455A 250 0

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

Figure 3.1 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 AND TEMPERATURE LIMITS REPORT Table 3.1 Data Points for Byron Unit 2 Nominal PORV Setpoints for the LTOP System Applicable for 30.5 EFPY (Inc]udes Instrumentation Uncertainty)

PCV-455A PCV-456 (2TY-04 1 3M) (2TY-04 1 3P)

AUCTIONEERED LOW RCS PRESSURE AUCTIONEERED LOW RCS PRESSURE RCS TEMP. (DIG. 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.)

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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 El$5-$2.

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.

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BYRON UNIT 2-PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.1 Byron Unit 2 Surveillance Capsule Withdrawal Summary Capsule Capsule Lead Factor Withdrawal EFPY fluence Location (n/cm2, E> 1.0 MeV)

U 58.50 4.02 1.19 0.406x i W 121.50 4.07 4.67 1.20x 1019 X 238.5° 4.14 8.63 2.18 x iü z 301.5° 4.11 14.28 (EOC 11) 3.25 x 10 v 61.0° 3.88 14.28 (EOC 11) 3.07 x i0 y(c) 241.0° 3.88 20.05 (EOC 15) 4.19 x io Notes:

(a) Source document is CN-AMLRS-1O-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.

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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 1/4T and 3/4T locations for 32 EFPY.

Table 5.4 provides the RT15 values for Byron Unit 2 for 32 EFPY obtained from Reference 5.

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BYRON UNIT 2-PRESSURE AND TEMP ERATURE LIMITS REPORT Table 5.1 llance Capsule Data Byron Unit 2 Calculation of Chemistry Factors Using Survei b) fF*RTNDT ARTDT1 FF2 Material Capsule FF (n/2E>1O*iiIV)

U 4.06x 1019 0.750 0.0 0.00 0.56 Lower Shell forging W 1.20 io 1.05 1 3.65 3.84 1.10 (Tangential) 1.211 15.75 19.08 1.47 X 2.l$x l0 U 04.06 x iO9 0.750 19.76 14.82 0.56 Lower Shell Forging W 1.20x 1019 1.051 31.88 33.50 1.10 (Axial) 1.211 38.91 47.14 1.47 X 2.18xl019 SUM: 118.38 6.27 CFLs Forging = 11 *ARTNDT) ÷ ( FF2) = (118.38) ÷ (6.27) 18.9°F 11.22 U 0.409 x 1019 0.752 (5 6l 8.44 0.57 Byron Unit I .

Surveillance Weld x 1.49x i 1.110 89.08 1.23 (40 U)

Material 102.68 (Heat#442002) W 2.26x 1019 1.221 (51.34) 125.34 1.49 16.88 U 0.406 x i0 0.750 (8 44) 12.66 0.56 Byron Unit 2 Surveillance Weld W 1.20x 1019 1.051 (88) 60.70 1.10 Material 108.02 (Heat#442002) x 2.18x i0 1.211 (5401) 130.86 1.47 SUM: 427.08 6.42 CfWeld Metal = (f1 *

\RTNDT) ÷ ( fF2) = (427.08) (6.42) = 66.5°f Notes:

a) Source document is CN-AMLRS-l0-$ (Reference 5), Table 5.2-2.

b) f = fluence; RTNDT values are the measured 30 ft-lb shift values taken from Reference 9.

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

= f(O28 OlOlog fl FF = fluence factor c)

U) Measured LRTNDT 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 AND TEMPERATURE LIMITS REPORT Table 5.2 (a)

Byron Unit 2 Reactor Vessel Material Properties

. JiLitial Matenal Description Cu (%) Ni (%) (b)

Ri NDT(T)

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]-1-1 0.01 0.70 -20 Lower Shell Forging [49D330/49C298j-1-l 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 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 Program 0.67, --

0 03 Weld Metal (Heat #442011) 0.71 a) Reference 7.

b) Initial RTNDT values are based on measured data.

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BYRON UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.3 Summary of Byron Unit 2 Adjusted Reference Temperature s (ART) Values at (a) 1/4T and 314T Locations for 32 EFPY Reactor Vessel Material

[ Surface Fluence 32 EFPY (n/cm , E> 1.0 MeV) 1/4T ART (°F) 3/4T ART (°F)

Nozzle Shell Forging 0.549 1019 53 38 Intermediate Shell forging 1.76 1019 21 9 Lower Shell Forging 1.76 X 10 52 34 Using credible surveillance data 1.76 10 16 8 Nozzle to Intermediate Shell Forging Circ. Weld Seam 0.549 x 1019 97 77 (Heat #442011)

Using credible Braidwood Units 1 0.549 10 76 64 and 2 surveillance data Intermediate to Lower Shell Forging Circ. Weld Seam 1.70 x 1019 119 88 (Heat # 442002) 1.70 X 1019 105 86

Using credible surveillance data Note:

(a) The source document containing detailed calculations is CN-AMLRS-1O-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 PIT 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 AND TEMPERATURE LIMITS REPORT Table 5.4 RTPTS Calculation for Byron Unit 2 Beltline Region Materials at EOL (32 EFPY)

(C) (c)

R.G. 1.99, CF Fluence FF IRTNTD IRTNTD a (d) Margin RTPTS Rev. 2 (°F) (°F) (°F) (°F) (°F)

Reactor Vessel Material (nlcm2, (°F)

Position E 1.0 MeV)

Nozzle Shell forging 1.1 31 0.549x l0 0.8323 10 25.8 0 12.9 25.8 62 Intermediate Shell forging 1.1 20 1.76 X i0 1.1554 -20 23.1 0 11.6 23.1 26 Lower Shell Forging 1.1 37 1.76x 10 1.1554 -20 42.7 0 17 34 57 Using credible surveillance data 2.1 18.9 1.76x 1019 1.1554 -20 21.8 0 8.5 17 19 Nozzle to Intermediate Shell forgingCirc.WeldSeam 1.1 41 0.549x 1019 0.8323 40 34.1 0 17.1 34.1 108 (HeaL_#442011)

UsingcredibleBraidwoodUnits 2.1 26.1 O.549x109 0.8323 40 21.7 0 10.9 21.7 83 I and 2 surveillance data Intermediate to Lower shell forging Circ Weld Seam 1.1 54 1.70x i 1.1461 10 61.9 0 28 56 128 (Heat # 442002)

Usingcrediblesurveiflancedata 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 RTPTS 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

= 8.5°F for position 2.1; the weld metal a = 28°F for position 1.1 (without surveillance data) and with credible surveillance data

= 14°f for Position 2.1. However, a need not to exceed O.SRTNTD 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-$, 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

BYRON UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT

12. 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.

13. WCAP-16143-P, Revision 1, Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for ByronfBraidwood Units 1 and 2, W. Bamford, et al., October 2014.

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