RS-16-248, Byron/Braidwood Nuclear Stations, Revision 16 to Updated Final Safety Analysis Report, Technical Requirements Manual

Byron/Braidwood Nuclear Stations, Revision 16 to Updated Final Safety Analysis Report, Technical Requirements Manual

ML16357A545
Person / Time
Site:	Byron, Braidwood
Issue date:	12/15/2016
From:	Exelon Generation Co
To:	Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
Shared Package
ML16357A264	List: ML16357A507 ML16357A508 ML16357A509 ML16357A521 ML16357A523 ML16357A524 ML16357A526 ML16357A529 ML16357A530 ML16357A532 ML16357A533 ML16357A534 ML16357A535 ML16357A536 ML16357A537 ML16357A540 ML16357A545 ML16357A549 ML16357A579 ML16357A580 ML16357A583 ML16357A584 ML16357A591 ML16357A592 ML16357A593 ML17086A588 ML17086A589 ML17086A590 ML17086A591 ML17086A592 ML17086A593 ML17086A595 ML17086A596 ML17086A597 ML17086A598 ML17086A599 ML17086A600 ML17086A601 ML17094A865 ML17095A852 ML17095A866 RS-16-248, Braidwood Nuclear Station, Amendment 27 to Fire Protection Report, Chapter 3.0, Guidelines of BTP Cmeb 9.5-1 RS-16-248, Braidwood Nuclear Stations, Revision 16 to Updated Final Safety Analysis Report, Technical Requirements Manual RS-16-248, Braidwood Nuclear Stations, Revision 16 to Updated Final Safety Analysis Report, Technical Specifications Bases RS-16-248, Byron Nuclear Station, Amendment 27 to Fire Protection Report, Chapter 3.0, Guidelines of BTP Cmeb 9.5-1 RS-16-248, Byron Nuclear Stations, Revision 16 to Updated Final Safety Analysis Report, Technical Specifications Bases RS-16-248, Byron/Braidwood Nuclear Stations, Amendment 27 to Fire Protection Report, Chapter 4.0, Fire Protection Program RS-16-248, Byron/Braidwood Nuclear Stations, Revision 16 to Updated Final Safety Analysis Report, Appendix B - Construct Material Standards and Quality Control Procedures RS-16-248, Byron/Braidwood Nuclear Stations, Revision 16 to Updated Final Safety Analysis Report, Appendix C - Deleted RS-16-248, Byron/Braidwood Nuclear Stations, Revision 16 to Updated Final Safety Analysis Report, Appendix D - Computer Programs ... further results
References
RS-16-248
Download: ML16357A545 (613)

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Figure 3.9.a-1 0 20 40 60 80 100 120 140 160 180 200405060708090100110120130140Total Number of Assys OffloadedTime After Shutdown (Hrs)41 MBTU/hr SFP Background Heat Load Margin ICDT Unacceptable Region Acceptable Region I Acceptable Region II

Figure 3.9.a-2 0 20 40 60 80 100 120 140 160 180 200405060708090100110120130140Total Number of Assys OffloadedTime After Shutdown (Hrs) 42 MBTU/hr SFP Background Heat Load Margin ICDT Acceptable Region II Unacceptable Region Acceptable Region I

Figure 3.9.a-3 0 20 40 60 80 100 120 140 160 180 200405060708090100110120130Total Number of Assys OffloadedTime After Shutdown (Hrs) 43 MBTU/hr SFP Background Heat Load Margin ICDT Acceptable Region II Unacceptable Region Acceptable Region I

Figure 3.9.a-4 0 20 40 60 80 100 120 140 160 180 200405060708090100110120Total Number of Assys OffloadedTime After Shutdown (Hrs)44 MBTU/hr SFP Background Heat Load Margin ICDT Acceptable Region II Unacceptable Region Acceptable Region I

Figure 3.9.a-5 0 20 40 60 80 100 120 140 160 180 200405060708090100110120Total Number of Assys OffloadedTime After Shutdown (Hrs)45 MBTU/hr SFP Background Heat Load Margin ICDT Acceptable Region II Unacceptable Region Acceptable Region I

Figure 3.9.a-6 0 20 40 60 80 100 120 140 160 180 200405060708090100110Total Number of Assys OffloadedTime After Shutdown (Hrs)46 MBTU/hr SFP Background Heat Load Margin ICDT Acceptable Region II Unacceptable Region Acceptable Region I

Figure 3.9.a-7 0 20 40 60 80 100 120 140 160 180 200405060708090100110Total Number of Assys OffloadedTime After Shutdown (Hrs)47 MBTU/hr SFP Background Heat Load Margin ICDT Acceptable Region II Unacceptable Region Acceptable Region I

Figure 3.9.a-8 0 20 40 60 80 100 120 140 160 180 200405060708090100Total Number of Assys OffloadedTime After Shutdown (Hrs)48 MBTU/hr SFP Background Heat Load Margin ICDT Acceptable Region II Unacceptable Region Acceptable Region I

Figure 3.9.a-9 0 20 40 60 80 100 120 140 160 180 200405060708090100Total Number of Assys OffloadedTime After Shutdown (Hrs)49 MBTU/hr SFP Background Heat Load Margin ICDT Acceptable Region II Unacceptable Region Acceptable Region I

(continued)

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(CM-1)(CM-1)

Outdoor Bulk Diesel Fuel Oil Tanks:

0DO03T (125k gal.)

0DO12T (50k gal.)

EDG Storage Tanks:

1DO01TA/B/C/D (25k

gal.)

2DO01TA/B (50k gal.)

EDG Day Tanks:

1DO02TA/B (500 gal.)

2DO02TA/B (500 gal.)

_B AF Pump Diesel Day Tanks:

1DO10T (500 gal.)

2DO10T (500 gal.)

Diesel-Driven Fire Pump Diesel Fuel

Oil Day Tank:

0DO05T (650 gal.)

Security Diesel Generator Day

Tank:

0DO06T (500 gal.)

0A/0B SX Makeup Pump Diesel Day

Tanks:

0DO08TA (2000 gal.)

0DO08TB (2000 gal.)

Frequency -Monthly31 Days92 Days31 Days92 Days92 Days92 Days92 Days92 Days Parameter:

Flash Point XX (2)Cloud Point XX (3)Water and Sediment XX (1)(2)(5)X (5)X (5)X (5)X (5)XX (5)Ramsbottom

Carbon Residue XX (3)Ash XX (3)Kinetic Viscosity XX (2)XXXXXX Copper Strip Corrosion XX (3)Cetane Index XX (3)Sulfur XX (3)API Gravity XX (2)XXXXXX Distillation

Temperature XX (3)Bacteria X (5)X (5)X (5)X (5)X (5)X (5)Clear and Bright XX (1)(2)XXXXXX Color XX (1)(2)XXXXXX Heat Value X Total Particulate

Contamination XX (4)(5)XX (5)X (5)X (5)X (5)X (5)Removal of

Accumulated

WaterXXX (5)XX (5)X (5)X (5)X (5)Lubricity X XX = Technical Specification required testing performed X = Testing performed NOTES:(1)Water and Sediment orClear and Bright with Proper Color is required.(2)Technical Specifications require verifying within limits within 30 days prior to adding new fuel oil to storage tanks.

(3)Technical Specifications require verifying within 30 days following sampling and addition to storage tanks.(4)Required since the Outdoor Bulk Diesel Fuel Oil Tanks are considered the source of stored fuel for the _B AF Pump Diesel Day Tanks and the 0A/0B SX Makeup Pump Diesel Day Tanks (5) Testing required per the License Renewal Fuel Oil Chemistry Aging Management Program (CM-1)

Yes No No No Yes Yes

Yes Yes Yes No No No

FIGURE 2 SUPPORT/SUPPORTED SYSTEM DIAGRAM EXAMPLE 1 EXAMPLESTRAIN A TRAIN B System 8 System 8 System 4 System 4 System 9 System 9 System 2 System 2 System 10 System 10 System 5 System 5 System 11 System 11 System 1 System 1 System 12 System 12System 6 System 6 System 13 System 13 System 3 System 3 System 14 System 14 System 7 System 7 System 15 System 15

FIGURE 2 SUPPORT/SUPPORTED SYSTEM DIAGRAM EXAMPLE 2 EXAMPLESTRAIN A TRAIN B System 8 System 8 System 4 System 4 System 9 System 9 System 2 System 2 System 10 System 10 System 5 System 5 System 11 System 11 System 1 System 1 System 12 System 12 System 6 System 6 System 13 System 13 System 3 System 3 System 14 System 14 System 7 System 7 System 15 System 15

FIGURE 2 SUPPORT/SUPPORTED SYSTEM DIAGRAM EXAMPLE 3 EXAMPLESTRAIN A TRAIN B System 8 System 8 System 4 System 4 System 9 System 9 System 2 System 2 System 10 System 10 System 5 System 5 System 11 System 11 System 1 System 1 System 12 System 12 System 6 System 6 System 13 System 13 System 3 System 3 System 14 System 14 System 7 System 7 System 15 System 15 Support System TS Number Support System Supported System TS Number Supported System

Support System TS Number Support System Supported System TS Number Supported System

Support System TS Number Support System Supported System TS Number Supported System

Support System TS Number Support System Supported System TS Number Supported System

Support System TS Number Support System Supported System TS Number Supported System

Support System TS Number Support System Supported System TS Number Supported System

SUPPORT SYSTEM:LCO 3.7.7Component Cooling Water

CASE A:

Case B:

CASE C:

Case C1

Case C2

Appendix Q was deleted with Revision 46 - see plant review 06-008.

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CORE OPERATING LIMITS REPORT (COLR)

FOR BYRON UNIT 1 CYCLE 21

EXELON TRACKING ID:

COLR BYRON 1 REVISION 11

COLR BYRON 1 Revision 11 Page 1 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 1 CYCLE 21 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report (COLR) for Byron Station Unit 1 Cycle 21 has been prepared in accordance with the requirements of Technical Specification 5.6.5 (ITS).

The Technical Specification Safety Limits and Limiting Conditions for Operation (LCOs) affected by this report are listed below:

SL 2.1.1 Reactor Core Safety Limits (SLs)

LCO 3.1.1 SHUTDOWN MARGIN (SDM)

LCO 3.1.3 Moderator Temperature Coefficient (MTC)

LCO 3.1.4 Rod Group Alignment Limits LCO 3.1.5 Shutdown Bank Insertion Limits LCO 3.1.6 Control Bank Insertion Limits LCO 3.1.8 PHYSICS TESTS Exceptions - MODE 2 LCO 3.2.1 Heat Flux Hot Channel Factor (F Q(Z)) LCO 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F NH) LCO 3.2.3 AXIAL FLUX DIFFERENCE (AFD)

LCO 3.2.5 Departure from Nucleate Boiling Ratio (DNBR)

LCO 3.3.1 Reactor Trip System (RTS) Instrumentation LCO 3.3.9 Boron Dilution Protection System (BDPS)

LCO 3.4.1 Reactor Coolant System (RCS) Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.9.1 Boron Concentration

The portions of the Technical Requirements Manual (TRM) affected by this report are listed below:

TRM TLCO 3.1.b Boration Flow Paths - Operating TRM TLCO 3.1.d Charging Pumps - Operating

TRM TLCO 3.1.f Borated Water Sources - Operating

TRM TLCO 3.1.g Position Indication System - Shutdown TRM TLCO 3.1.h Shutdown Margin (SDM) - MODE 1 and MODE 2 with keff 1.0 TRM TLCO 3.1.i Shutdown Margin (SDM) - MODE 5 TRM TLCO 3.1.j Shutdown and Control Rods TRM TLCO 3.1.k Position Indication System - Shutdown (Special Test Exception)

COLRBYRON 1 Revision 11 Page 2of15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 1CYCLE21 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections.

These limits are applicable for the entire cycle unless otherwise identified.

These limits have been developed using the NRG-approved methodologies specified in Technical Specification 5.6.5. 2.1 Reactor Core Safety Limits (Sls) (SL 2.1.1) ,...--..., LL_ en Q) 0 ...__,...

Q) \....._ ::::; -+--' 0 \....._ Q) o_ E Q) I--Q) en 0 \....._ Q) > <( 2.1.1 In MODES 1 and 2, the combination of Thermal Power, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in Figure 2.1.1. 680 2471 p s ia ----660 2250 p s ia .... _ 640 2000 p s i a ............._

1 860 p s i a . ..........._

620 600

____________________

-+-________________

0 0.2 0.4 0.6 F ro ct i on of N o o.s in o l P owe r Figure 2.1.1: Reactor Core Limits 1.2 COLR BYRON 1 Revision 11 Page 3 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 1 CYCLE 21 2.2 SHUTDOWN MARGIN (SDM)

The SDM limit for MODES 1, 2, 3, and 4 is:

2.2.1 The SDM shall be greater than or equal to 1.3% k/k (LCOs 3.1.1, 3.1.4, 3.1.5, 3.1.6, 3.1.8, 3.3.9; TRM TLCOs 3.1.b, 3.1.d, 3.1.f, 3.1.h, and 3.1.j).

The SDM limit for MODE 5 is:

2.2.2 SDM shall be greater than or equal to 1.3% k/k (LCO 3.1.1, LCO 3.3.9; TRM TLCOs 3.1.i and 3.1.j).

2.3 Moderator Temperature Coefficient (MTC) (LCO 3.1.3)

The Moderator Temperature Coefficient (MTC) limits are:

2.3.1 The BOL/ARO/HZP-MTC upper limit shall be +2.035 x 10

-5 k/k/°F. 2.3.2 The EOL/ARO/HFP-MTC lower limit shall be -4.6 x 10

-4 k/k/°F. 2.3.3 The EOL/ARO/HFP-MTC Surveillance limit at 300 ppm shall be -3.7 x 10

-4 k/k/°F. 2.3.4 The EOL/ARO/HFP-MTC Surveillance limit at 60 ppm shall be -4.3 x 10

-4 k/k/°F. where: BOL stands for Beginning of Cycle Life ARO stands for All Rods Out HZP stands for Hot Zero Thermal Power EOL stands for End of Cycle Life HFP stands for Hot Full Thermal Power 2.4 Shutdown Bank Insertion Limits (LCO 3.1.5) 2.4.1 All shutdown banks shall be fully withdrawn to at least 224 steps.

2.5 Control Bank Insertion Limits (LCO 3.1.6) 2.5.1 The control banks, with Bank A greater than or equal to 224 steps, shall be limited in physical insertion as shown in Figure 2.5.1.

2.5.2 Each control bank shall be considered fully withdrawn from the core at greater than or equal to 224 steps.

2.5.3 The control banks shall be operated in sequence by withdrawal of Bank A, Bank B, Bank C and Bank D. The control banks shall be sequenced in reverse order upon insertion.

2.5.4 Each control bank not fully withdrawn from the core shall be operated with the following overlap limits as a function of park position:

Park Position (step) Overlap Limit (step) 231 115 COLR BYRON 1 Revision 11 Page 4 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 1 CYCLE 21 0 20 40 60 80100 120 140 160 180 200 2200102030405060708090100Rod Bank Position (Steps Withdrawn) R e l a t i v e P o w e r ( P e r c e n t ) Figure 2.5.1: Control Bank Insertion Limits Versus Percent Rated Thermal Power (26.52%, 224) (76.96%, 224) (100%, 161)

BANK B BANK C BANK D (0%, 163) (0%, 47) (30%, 0) 224 COLR BYRON 1 Revision 11 Page 5 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 1 CYCLE 21 2.6 Heat Flux Hot Channel Factor (F Q(Z)) (LCO 3.2.1) 2.6.1 Total Peaking Factor:

where: P = the ratio of THERMAL POWER to RATED THERMAL POWER

F Q RTP = 2.60 K(Z) is provided in Figure 2.6.1.

0.5 P for )(0.5 F (Z)F RTP Q Q Z xK 0.5 P for )(P F (Z)F RTP Q Q Z xK(0.0, 1.0) (6.0, 1.0) (12.0,0.924) 00.10.20.30.4 0.50.60.70.8 0.9 11.10123456789101112K(Z) Normalized F Q (Z) BOTTOM Core Height (ft) TOP Figure 2.6.1 K(Z) - Normalized F Q(Z) as a Function of Core Height LOCA Limiting Envelope(0.0, 1.0) (6.0, 1.0) (12.0,0.924) 00.1 0.2 0.3 0.4 0.50.60.7 0.8 0.9 11.10123456789101112K(Z) Normalized F Q (Z) BOTTOM Core Height (ft) TOP Figure 2.6.1 K(Z) - Normalized F Q(Z) as a Function of Core Height LOCA Limiting Envelope COLR BYRON 1 Revision 11 Page 6 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 1 CYCLE 21 2.6.2 W(Z) Values:

a) When the Power Distribution Monitoring System (PDMS) is OPERABLE, W(Z) = 1.00000 for all axial points.

b) When PDMS is inoperable, W(Z) is provided as:

1) Table 2.6.2.a are the normal operation W(Z) values for the full cycle and correspond to the AXIAL FLUX DIFFERENCE (AFD) Acceptable Operation Limits provided in Figure 2.8.1.a. The normal operation W(Z) values have been determined at burnups of 150, 6000, 14000, and 20000 MWD/MTU.

2) The EOL-only normal operation W(Z) values provided in Table 2.6.2.b may be used for cycle burnups 18000 MWD/MTU. The EOL-only W(Z) values correspond to the REDUCED AXIAL FLUX DIFFERENCE (AFD) Acceptable Operation Limits provided in Figure 2.8.1.b. The EOL-only normal operation W(Z) values have been determined at burnups of 18000 and 20000 MWD/MTU and the last column of W(Z) values is a duplicate of the 20000 MWD/MTU values. If invoked, the EOL-only W(Z) values are to be used for the remainder of the cycle unless superseded by a subsequent analysis.

Table 2.6.2.c shows the F C Q(z) penalty factors that are greater than 2% per 31 Effective Full Power Days (EFPD). These values shall be used to increase the F W Q(z) as per Surveillance Requirement 3.2.1.2. A 2% penalty factor shall be used at all cycle burnups that are outside the range of Table 2.6.2.c.

2.6.3 Uncertainty:

The uncertainty, U FQ, to be applied to the Heat Flux Hot Channel Factor F Q(Z) shall be calculated by the following formula UUUFQque where:

U qu = Base F Q measurement uncertainty = 1.05 when PDMS is inoperable (U qu is defined by PDMS when OPERABLE.)

U e = Engineering uncertainty factor = 1.03 2.6.4 PDMS Alarms:

F Q(Z) Warning Setpoint = 2% F Q(Z) Margin F Q(Z) Alarm Setpoint = 0% F Q(Z) Margin

COLR BYRON 1 Revision 11 Page 7 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 1 CYCLE 21 Table 2.6.2.a Full Cycle W(Z) versus Core Height for AFD Acceptable Operation Limits in Figure 2.8.1.a (Top and Bottom 8% Excluded per WCAP-10216)

Height 150 6000 14000 20000 (feet) MWD/MTU MWD/MTU MWD/MTU MWD/MTU 0.00 (core bottom) 1.3318 1.4685 1.4170 1.3511 0.20 1.3188 1.4244 1.3820 1.3254 0.40 1.3143 1.4056 1.3650 1.3132 0.60 1.3088 1.3888 1.3480 1.3027 0.80 1.2997 1.3656 1.3270 1.2984 1.00 1.2884 1.3526 1.3140 1.2936 1.20 1.2746 1.3369 1.2980 1.2868 1.40 1.2651 1.3201 1.2870 1.2801 1.60 1.2569 1.2987 1.2740 1.2694 1.80 1.2514 1.2807 1.2620 1.2607 2.00 1.2422 1.2592 1.2500 1.2510 2.20 1.2262 1.2369 1.2360 1.2364 2.40 1.2110 1.2154 1.2230 1.2247 2.60 1.1941 1.1920 1.2079 1.2091 2.80 1.1794 1.1737 1.1970 1.1945 3.00 1.1681 1.1692 1.1902 1.1833 3.20 1.1590 1.1645 1.1821 1.1764 3.40 1.1528 1.1603 1.1746 1.1820 3.60 1.1438 1.1538 1.1647 1.1852 3.80 1.1395 1.1480 1.1609 1.1886 4.00 1.1344 1.1418 1.1567 1.1908 4.20 1.1296 1.1328 1.1507 1.1900 4.40 1.1313 1.1251 1.1446 1.1927 4.60 1.1319 1.1151 1.1376 1.2011 4.80 1.1323 1.1060 1.1381 1.2075 5.00 1.1311 1.0989 1.1386 1.2110 5.20 1.1285 1.0955 1.1397 1.2143 5.40 1.1261 1.0912 1.1411 1.2170 5.60 1.1215 1.0868 1.1497 1.2354 5.80 1.1178 1.0897 1.1641 1.2507 6.00 1.1237 1.0955 1.1776 1.2640 6.20 1.1304 1.1003 1.1884 1.2724 6.40 1.1351 1.1061 1.1971 1.2779 6.60 1.1389 1.1147 1.2029 1.2796 6.80 1.1417 1.1214 1.2076 1.2794 7.00 1.1435 1.1271 1.2085 1.2753 7.20 1.1446 1.1338 1.2065 1.2662 7.40 1.1479 1.1447 1.2035 1.2562 7.60 1.1582 1.1543 1.1965 1.2404 7.80 1.1680 1.1635 1.1886 1.2255 8.00 1.1767 1.1728 1.1788 1.2077 8.20 1.1849 1.1830 1.1661 1.1873 8.40 1.1919 1.1934 1.1546 1.1774 8.60 1.1976 1.2023 1.1555 1.1646 8.80 1.2005 1.2107 1.1577 1.1519 9.00 1.2005 1.2186 1.1598 1.1415 9.20 1.2129 1.2365 1.1630 1.1764 9.40 1.2203 1.2536 1.1714 1.2100 9.60 1.2274 1.2687 1.1960 1.2560 9.80 1.2313 1.2837 1.2300 1.2960 10.00 1.2348 1.2979 1.2610 1.3320 10.20 1.2429 1.3104 1.2860 1.3640 10.40 1.2509 1.3227 1.3030 1.3900 10.60 1.2447 1.3314 1.3139 1.4156 10.80 1.2514 1.3363 1.3287 1.4364 11.00 1.2490 1.3290 1.3385 1.4492 11.20 1.2341 1.3207 1.3380 1.4300 11.40 1.2386 1.3260 1.3350 1.4170 11.60 1.2433 1.3329 1.3062 1.3900 11.80 1.2486 1.3394 1.2951 1.3849 12.00 (core top) 1.2572 1.3498 1.2878 1.3886 Note: W(Z) values at 20000 MWD/MTU may be applied to cycle burnups greater than 20000 MWD/MTU to prevent W(Z) function extrapolation

COLR BYRON 1 Revision 11 Page 8 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 1 CYCLE 21 Table 2.6.2.b EOL-only W(Z) versus Core Height for AFD Acceptable Operation Limits in Figure 2.8.1.b (Top and Bottom 8% Excluded per WCAP-10216) Height 18000 20000 24114 (feet) MWD/MTU MWD/MTU MWD/MTU 0.00 (core bottom) 1.2467 1.2264 1.2264 0.20 1.2208 1.2019 1.2019 0.40 1.2067 1.1875 1.1875 0.60 1.1960 1.1794 1.1794 0.80 1.1886 1.1788 1.1788 1.00 1.1826 1.1766 1.1766 1.20 1.1728 1.1722 1.1722 1.40 1.1668 1.1676 1.1676 1.60 1.1586 1.1599 1.1599 1.80 1.1523 1.1535 1.1535 2.00 1.1502 1.1471 1.1471 2.20 1.1448 1.1376 1.1376 2.40 1.1401 1.1294 1.1294 2.60 1.1334 1.1192 1.1192 2.80 1.1276 1.1102 1.1102 3.00 1.1240 1.1078 1.1078 3.20 1.1244 1.1136 1.1136 3.40 1.1282 1.1257 1.1257 3.60 1.1345 1.1409 1.1409 3.80 1.1447 1.1551 1.1551 4.00 1.1543 1.1673 1.1673 4.20 1.1630 1.1792 1.1792 4.40 1.1697 1.1891 1.1891 4.60 1.1749 1.1976 1.1976 4.80 1.1798 1.2056 1.2056 5.00 1.1833 1.2110 1.2110 5.20 1.1859 1.2143 1.2143 5.40 1.1884 1.2170 1.2170 5.60 1.2035 1.2354 1.2354 5.80 1.2193 1.2507 1.2507 6.00 1.2333 1.2640 1.2640 6.20 1.2433 1.2724 1.2724 6.40 1.2504 1.2779 1.2779 6.60 1.2536 1.2796 1.2796 6.80 1.2553 1.2794 1.2794 7.00 1.2530 1.2753 1.2753 7.20 1.2462 1.2662 1.2662 7.40 1.2381 1.2562 1.2562 7.60 1.2248 1.2404 1.2404 7.80 1.2115 1.2255 1.2255 8.00 1.1957 1.2077 1.2077 8.20 1.1767 1.1873 1.1873 8.40 1.1643 1.1774 1.1774 8.60 1.1568 1.1646 1.1646 8.80 1.1500 1.1519 1.1519 9.00 1.1445 1.1415 1.1415 9.20 1.1644 1.1764 1.1764 9.40 1.1863 1.2100 1.2100 9.60 1.2237 1.2560 1.2560 9.80 1.2626 1.2960 1.2960 10.00 1.2977 1.3320 1.3320 10.20 1.3276 1.3640 1.3640 10.40 1.3501 1.3900 1.3900 10.60 1.3694 1.4156 1.4156 10.80 1.3884 1.4364 1.4364 11.00 1.4013 1.4492 1.4492 11.20 1.3908 1.4300 1.4300 11.40 1.3815 1.4170 1.4170 11.60 1.3508 1.3900 1.3900 11.80 1.3416 1.3849 1.3849 12.00 (core top) 1.3391 1.3886 1.3886 Note: W(Z) values at 20000 MWD/MTU may be applied to cycle burnups greater than 20000 MWD/MTU to prevent W(Z) function extrapolation

COLR BYRON 1 Revision 11 Page 9 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 1 CYCLE 21 Table 2.6.2.c Penalty Factors in Excess of 2% per 31 EFPD Cycle Burnup Penalty Factor (MWD/MTU) F C Q(z) 0 1.0200 502 1.0200 1205 1.0427 1381 1.0450 1700 1.0450 1909 1.0422 2085 1.0392 2261 1.0356 2964 1.0214 3140 1.0200 14396 1.0200 14572 1.0204 15099 1.0208 15803 1.0213 16600 1.0218 17034 1.0218 17913 1.0215 18100 1.0213 18660 1.0200 24114 1.0200 Notes:

Linear interpolation is adequate for intermediate cycle burnups.

All cycle burnups outside the range of Table 2.6.2.c shall use a 2% penalty factor for compliance with the 3.2.1.2 Surveillance Requirements.

COLR BYRON 1 Revision 11 Page 10 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 1 CYCLE 21 2.7 Nuclear Enthalpy Rise Hot Channel Factor (F NH) (LCO 3.2.2) 2.7.1 F NH FH RTP[1.0 + PFH(1.0 - P)]

where: P = the ratio of THERMAL POWER to RATED THERMAL POWER (RTP)

FH RTP = 1.70 PFH = 0.3 2.7.2 Uncertainty:

The uncertainty, U FH, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor F NH shall be calculated by the following formula:

U FH = U FHm where: U FHm = Base F NH measurement uncertainty = 1.04 when PDMS is inoperable (U FHm is defined by PDMS when OPERABLE.)

2.7.3 PDMS Alarms:

F NH Warning Setpoint = 2% F NH Margin F NH Alarm Setpoint = 0% F NH Margin 2.8 AXIAL FLUX DIFFERENCE (AFD) (LCO 3.2.3) 2.8.1 When PDMS is inoperable, the AXIAL FLUX DIFFERENCE (AFD) Acceptable Operation Limits are provided in the Figures described below or the latest valid PDMS Surveillance Report, whichever is more conservative.

a) Figure 2.8.1.a is the full cycle AFD Acceptable Operation Limits associated with the full cycle W(Z) values in Table 2.6.2.a.

b) Figure 2.8.1.b is the Reduced AFD Acceptable Operation Limits which may be applied after 18000 MWD/MTU. The Reduced AFD Acceptable Operation Limits are associated with the EOL-only W(Z) values in Table 2.6.2.b. Prior to changing to Figure 2.8.1.b, confirm that the plant is within the specified AFD envelope.

2.8.2 When PDMS is OPERABLE, no AFD Acceptable Operation Limits are applicable.

2.9 Departure from Nucleate Boiling Ratio (DNBR) (LCO 3.2.5) 2.9.1 DNBRAPSL 1.563 The Axial Power Shape Limiting DNBR (DNBR APSL) is applicable with THERMAL POWER 50% RTP when PDMS is OPERABLE.

2.9.2 PDMS Alarms:

DNBR Warning Setpoint = 2% DNBR Margin DNBR Alarm Setpoint = 0% DNBR Margin

COLR BYRON 1 Revision 11 Page 11 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 1 CYCLE 21 Figure 2.8.1.a: Axial Flux Difference Limits as a Function of Rated Thermal Power (Full Cycle)

(-35, 50) (-15, 100) Acceptable Operation (+10, 100)

(+25, 50) Unacceptable Operation Unacceptable Operation 0 20 40 60 80 100 120-50-40-30-20-1001020304050% of RATED THERMAL POWER AXIAL FLUX DIFFERENCE (%) Axial Flux Difference Limits with PDMS Inoperable COLR BYRON 1 Revision 11 Page 12 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 1 CYCLE 21 Figure 2.8.1.b: Reduced Axial Flux Difference Limits as a Function of Rated Thermal Power (Cycle burnup 18000 MWD/MTU)

Axial Flux Difference LimitswithPDMS Inoperable 0 20 40 60 80 100 120-50-40-30-20-1001020304050AXIAL FLUX DIFFERENCE (%)% of RATED THERMAL POWE RUnacceptableOperationUnacceptableOperationAcceptableOperation(-20, 50)(-10, 100)(+10, 100)(+25, 50)

COLR BYRON 1 Revision 11 Page 13 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 1 CYCLE 21 2.10 Reactor Trip System (RTS) Instrumentation (LCO 3.3.1) - Overtemperature T Setpoint Parameter Values 2.10.1 The Overtemperature T reactor trip setpoint K 1 shall be equal to 1.325.

2.10.2 The Overtemperature T reactor trip setpoint T avg coefficient K 2 shall be equal to 0.0297 / °F.

2.10.3 The Overtemperature T reactor trip setpoint pressure coefficient K 3 shall be equal to 0.00135 / psi.

2.10.4 The nominal T avg at RTP (indicated) T shall be less than or equal to 588.0 °F.

2.10.5 The nominal RCS operating pressure (indicated) P shall be equal to 2235 psig.

2.10.6 The measured reactor vessel T lead/lag time constant 1 shall be equal to 8 sec.

2.10.7 The measured reactor vessel T lead/lag time constant 2 shall be equal to 3 sec.

2.10.8 The measured reactor vessel T lag time constant 3 shall be less than or equal to 2 sec. 2.10.9 The measured reactor vessel average temperature lead/lag time constant 4 shall be equal to 33 sec.

2.10.10 The measured reactor vessel average temperature lead/lag time constant 5 shall be equal to 4 sec.

2.10.11 The measured reactor vessel average temperature lag time constant 6 shall be less than or equal to 2 sec.

2.10.12 The f 1 (I) "positive" breakpoint shall be +10% I. 2.10.13 The f 1 (I) "negative" breakpoint shall be -18% I. 2.10.14 The f 1 (I) "positive" slope shall be +3.47% / % I. 2.10.15 The f 1 (I) "negative" slope shall be -2.61% / % I.

COLR BYRON 1 Revision 11 Page 14 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 1 CYCLE 21 2.11 Reactor Trip System (RTS) Instrumentation (LCO 3.3.1) - Overpower T Setpoint Parameter Values 2.11.1 The Overpower T reactor trip setpoint K 4 shall be equal to 1.072.

2.11.2 The Overpower T reactor trip setpoint T avg rate/lag coefficient K 5 shall be equal to 0.02 / °F for increasing Tavg. 2.11.3 The Overpower T reactor trip setpoint T avg rate/lag coefficient K 5 shall be equal to 0 / °F for decreasing T avg. 2.11.4 The Overpower T reactor trip setpoint T avg heatup coefficient K 6 shall be equal to 0.00245 / °F when T T. 2.11.5 The Overpower T reactor trip setpoint T avg heatup coefficient K 6 shall be equal to 0 / °F when T T. 2.11.6 The nominal T avg at RTP (indicated) T shall be less than or equal to 588.0 °F.

2.11.7 The measured reactor vessel T lead/lag time constant 1 shall be equal to 8 sec.

2.11.8 The measured reactor vessel T lead/lag time constant 2 shall be equal to 3 sec.

2.11.9 The measured reactor vessel T lag time constant 3 shall be less than or equal to 2 sec. 2.11.10 The measured reactor vessel average temperature lag time constant 6 shall be less than or equal to 2 sec.

2.11.11 The measured reactor vessel average temperature rate/lag time constant 7 shall be equal to 10 sec.

2.11.12 The f 2 (I) "positive" breakpoint shall be 0 for all I. 2.11.13 The f 2 (I) "negative" breakpoint shall be 0 for all I. 2.11.14 The f 2 (I) "positive" slope shall be 0 for all I. 2.11.15 The f 2 (I) "negative" slope shall be 0 for all I.

COLR BYRON 1 Revision 11 Page 15 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 1 CYCLE 21 2.12 Reactor Coolant System (RCS) Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits (LCO 3.4.1) 2.12.1 The pressurizer pressure shall be greater than or equal to 2209 psig.

2.12.2 The RCS average temperature (T avg) shall be less than or equal to 593.1 °F.

2.12.3 The RCS total flow rate shall be greater than or equal to 386,000 gpm.

2.13 Boron Concentration 2.13.1 The refueling boron concentration shall be greater than or equal to the applicable value given in the Table below (LCO 3.9.1). The reported "prior to initial criticality" value also bounds the end-of-cycle requirements for the previous cycle.

2.13.2 To maintain keff 0.987 with all shutdown and control rods fully withdrawn in MODES 3, 4, or 5 (TRM TLCO 3.1.g Required Action B.2 and TRM TLCO 3.1.k.2), the Reactor Coolant System boron concentration shall be greater than or equal to the applicable values given in the Table below.

COLR Section Conditions Boron Concentration (ppm) 2.13.1 a) prior to initial criticality 1653 b) for cycle burnups 0 MWD/MTU and < 16000 MWD/MTU 1764 c) for cycle burnups 1410 2.13.2 a) prior to initial criticality 1754 b) all other times in life 1980 CORE OPERATING LIMITS REPORT (COLR)

FORBYRON UNIT 2 CYCLE 20 EXELON TRACKING ID: COLR BYRON 2 REVISION 8 COLR BYRON 2 Revision 8 Page 1 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 2 CYCLE 20 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report (COLR) for Byron Station Unit 2 Cycle 20 has been prepared in accordance with the requirements of Technical Specification Safety Limits and Limiting Conditions for Operation (LCOs) 5.6.5 (ITS). The Technical Specification Safety Limits and Limiting Conditions for Operation (LCOs) affected by this report are listed below: SL 2.1.1 Reactor Core Safety Limits (SLs) LCO 3.1.1 SHUTDOWN MARGIN (SDM) LCO 3.1.3 Moderator Temperature Coefficient (MTC) LCO 3.1.4 Rod Group Alignment Limits LCO 3.1.5 Shutdown Bank Insertion Limits LCO 3.1.6 Control Bank Insertion Limits LCO 3.1.8 PHYSICS TESTS Exceptions - MODE 2 LCO 3.2.1 Heat Flux Hot Channel Factor (F Q(Z)) LCO 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F NH)LCO 3.2.3 AXIAL FLUX DIFFERENCE (AFD) LCO 3.2.5 Departure from Nucleate Boiling Ratio (DNBR) LCO 3.3.1 Reactor Trip System (RTS) Instrumentation LCO 3.3.9 Boron Dilution Protection System (BDPS) LCO 3.4.1 Reactor Coolant System (RCS) Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.9.1 Boron Concentration The portions of the Technical Requirements Manual (TRM) affected by this report are listed below:

TRM TLCO 3.1.b Boration Flow Paths - Operating TRM TLCO 3.1.d Charging Pumps - Operating TRM TLCO 3.1.f Borated Water Sources - Operating TRM TLCO 3.1.g Position Indication System - Shutdown TRM TLCO 3.1.h Shutdown Margin (SDM) - MODE 1 and MODE 2 with keff 1.0 TRM TLCO 3.1.i Shutdown Margin (SDM) - MODE 5 TRM TLCO 3.1.j Shutdown and Control Rods TRM TLCO 3.1.k Position Indication System - Shutdown (Special Test Exception)

COLRBYRON2 Revision 8 Page 2of15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 2 CYCLE 20 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections.

These limits are applicable for the entire cycle unless otherwise identified.

These limits have been developed using the NRG-approved methodologies specified in Technical Specification 5.6.5. 2.1 Reactor Core Safety Limits (Sls) (SL 2.1.1) .----... Li_ Q") Q) c:::::i .____.. Q) +--' 0 Q) o_ E Q) f-----Q) Q") 0 Q) > <( 2.1.1 In MODES 1 and 2, the combination of Thermal Power, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in Figure 2.1.1. 680 2471 p s ia 660 2250 p s ia 6 4 0 2000 p s ia ......_ 1 86 0 p s ia . ......_ 620 600 o 0.2 o.4 o.6 o.s 1.2 F ro ct i o n o f N om i n o I P owe r Figure 2.1.1: Reactor Core Limits COLR BYRON 2 Revision 8 Page 3 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 2 CYCLE 20 2.2 SHUTDOWN MARGIN (SDM) The SDM limit for MODES 1, 2, 3, and 4 is: 2.2.1 The SDM shall be greater than or equal to 1.3% k/k (LCOs 3.1.1, 3.1.4, 3.1.5, 3.1.6, 3.1.8, 3.3.9; TRM TLCOs 3.1.b, 3.1.d, 3.1.f, 3.1.h, and 3.1.j). The SDM limit for MODE 5 is: 2.2.2 SDM shall be greater than or equal to 1.3% k/k (LCO 3.1.1, LCO 3.3.9; TRM TLCOs 3.1.i and 3.1.j). 2.3 Moderator Temperature Coefficient (MTC) (LCO 3.1.3) The Moderator Temperature Coefficient (MTC) limits are: 2.3.1 The BOL/ARO/HZP-MTC upper limit shall be +2.721 x 10

-5k/k/°F. 2.3.2 The EOL/ARO/HFP-MTC lower limit shall be -4.6 x 10

-4k/k/°F. 2.3.3 The EOL/ARO/HFP-MTC Surveillance limit at 300 ppm shall be -3.7 x 10

-4k/k/°F. 2.3.4 The EOL/ARO/HFP-MTC Surveillance limit at 60 ppm shall be -4.3 x 10

-4k/k/°F. where: BOL stands for Beginning of Cycle Life ARO stands for All Rods Out HZP stands for Hot Zero Thermal Power EOL stands for End of Cycle Life HFP stands for Hot Full Thermal Power 2.4 Shutdown Bank Insertion Limits (LCO 3.1.5) 2.4.1 All shutdown banks shall be fully withdrawn to at least 224 steps. 2.5 Control Bank Insertion Limits (LCO 3.1.6) 2.5.1 The control banks, with Bank A greater than or equal to 224 steps, shall be limited in physical insertion as shown in Figure 2.5.1. 2.5.2 Each control bank shall be considered fully withdrawn from the core at greater than or equal to 224 steps. 2.5.3 The control banks shall be operated in sequence by withdrawal of Bank A, Bank B, Bank C and Bank D. The control banks shall be sequenced in reverse order upon insertion. 2.5.4 Each control bank not fully withdrawn from the core shall be operated with the following overlap limits as a function of park position: Park Position (step) Overlap Limit (step) 231 115 COLRBYRON2 Revision 8 Page 4of15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 2 CYCLE 20 Figure 2.5.1: Contro l Bank Insertion Limits Versus Percent Rated Thermal Power 224 (26.52%, 224) (76.96%, 224) 220 200 180 c: 160 nl .... "CJ ..c: 140 :s: Ill a. 120 c: 0 100 Ill 0 a.. 80 c: nl m "CJ 0 60 a: 40 20 I' I' / v / / /1 sANKB I v / / 1 (0%, 163) I L / / I BANKC I / / If' , / / v / I BANKD I / / / v / / v 'Ho%, 47) I / (30%, 0) 1/ (1 00%, 1 6 1) 0 v 0 10 20 30 40 50 60 70 80 90 100 Relative Power (Percent)

COLR BYRON 2 Revision 8 Page 5 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 2 CYCLE 20 2.6 Heat Flux Hot Channel Factor (F Q(Z)) (LCO 3.2.1) 2.6.1 Total Peaking Factor: where: P = the ratio of THERMAL POWER to RATED THERMAL POWER F Q RTP = 2.60 K(Z) is provided in Figure 2.6.1.

0.5 P for )(0.5 F (Z)F RTP Q Q Z xK 0.5 P for )(P F (Z)F RTP Q Q Z xK(0.0, 1.0) (6.0, 1.0) (12.0,0.924) 00.10.2 0.3 0.4 0.5 0.6 0.7 0.80.9 11.10123456789101112K(Z) Normalized F Q (Z)BOTTOM Core Height (ft) TOP Figure 2.6.1 K(Z) - Normalized F Q(Z) as a Function of Core Height LOCA Limiting Envelope COLR BYRON 2 Revision 8 Page 6 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 2 CYCLE 20 2.6.2 W(Z) Values: a) When the Power Distribution Monitoring System (PDMS) is OPERABLE, W(Z) = 1.00000 for all axial points. b) When PDMS is inoperable, W(Z) is provided as: 1) Table 2.6.2.a are the normal operation W(Z) values for the full cycle and correspond to the AXIAL FLUX DIFFERENCE (AFD) Acceptable Operation Limits provided in Figure 2.8.1.a. The normal operation W(Z) values have been determined at burnups of 150, 5000, 14000, and 20000 MWD/MTU. 2) The EOL-only normal operation W(Z) values provided in Table 2.6.2.b may be used for cycle burnups 18000 MWD/MTU. The EOL-only W(Z) values correspond to the REDUCED AXIAL FLUX DIFFERENCE (AFD) Acceptable Operation Limits provided in Figure 2.8.1.b. The EOL-only normal operation W(Z) values have been determined at burnups of 18000 and 20000 MWD/MTU and the last column of W(Z) values is a duplicate of the 20000 MWD/MTU values. If invoked, the EOL-only W(Z) values are to be used for the remainder of the cycle unless superseded by a subsequent analysis. Table 2.6.2.c shows the F C Q(z) penalty factors that are greater than 2% per 31 Effective Full Power Days (EFPD). These values shall be used to increase the F W Q(z) as per Surveillance Requirement 3.2.1.2. A 2% penalty factor shall be used at all cycle burnups that are outside the range of Table 2.6.2.c. 2.6.3 Uncertainty:

The uncertainty, U FQ, to be applied to the Heat Flux Hot Channel Factor F Q(Z) shall be calculated by the following formula UUUFQque where: U qu = Base F Q measurement uncertainty = 1.05 when PDMS is inoperable (U qu is defined by PDMS when OPERABLE.) U e = Engineering uncertainty factor = 1.03 2.6.4 PDMS Alarms:

F Q(Z) Warning Setpoint = 2% F Q(Z) Margin F Q(Z) Alarm Setpoint = 0% F Q(Z) Margin COLR BYRON 2 Revision 8 Page 7 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 2 CYCLE 20 Table 2.6.2.a Full Cycle W(Z) versus Core Height for AFD Acceptable Operation Limits in Figure 2.8.1.a (Top and Bottom 8% Excluded per WCAP-10216) Height 150 5000 14000 20000 (feet) MWD/MTU MWD/MTU MWD/MTU MWD/MTU 0.00 (core bottom) 1.3939 1.5083 1.3877 1.3456 0.20 1.3748 1.4558 1.3691 1.3355 0.40 1.3650 1.4434 1.3592 1.3329 0.60 1.3533 1.4285 1.3438 1.3302 0.80 1.3342 1.4182 1.3007 1.3254 1.00 1.3207 1.4008 1.2896 1.3188 1.20 1.3034 1.3591 1.2861 1.3097 1.40 1.2885 1.3440 1.2774 1.3007 1.60 1.2712 1.3283 1.2621 1.2873 1.80 1.2579 1.3124 1.2477 1.2753 2.00 1.2525 1.2962 1.2336 1.2613 2.20 1.2451 1.2813 1.2239 1.2428 2.40 1.2363 1.2621 1.2152 1.2263 2.60 1.2238 1.2411 1.2037 1.2066 2.80 1.2079 1.2199 1.1932 1.1888 3.00 1.1935 1.2077 1.1849 1.1755 3.20 1.1824 1.1982 1.1762 1.1715 3.40 1.1765 1.1929 1.1679 1.1734 3.60 1.1704 1.1852 1.1577 1.1786 3.80 1.1642 1.1786 1.1512 1.1822 4.00 1.1582 1.1716 1.1505 1.1852 4.20 1.1510 1.1623 1.1475 1.1878 4.40 1.1449 1.1529 1.1445 1.1998 4.60 1.1444 1.1426 1.1405 1.2099 4.80 1.1433 1.1320 1.1428 1.2177 5.00 1.1397 1.1218 1.1481 1.2240 5.20 1.1350 1.1151 1.1524 1.2274 5.40 1.1289 1.1138 1.1568 1.2303 5.60 1.1238 1.1179 1.1646 1.2473 5.80 1.1311 1.1240 1.1830 1.2651 6.00 1.1378 1.1281 1.2000 1.2787 6.20 1.1425 1.1308 1.2140 1.2874 6.40 1.1472 1.1360 1.2252 1.2954 6.60 1.1490 1.1411 1.2325 1.2965 6.80 1.1498 1.1434 1.2389 1.2967 7.00 1.1486 1.1461 1.2413 1.2929 7.20 1.1473 1.1511 1.2400 1.2843 7.40 1.1476 1.1550 1.2363 1.2734 7.60 1.1473 1.1605 1.2296 1.2584 7.80 1.1465 1.1664 1.2199 1.2416 8.00 1.1452 1.1660 1.2126 1.2231 8.20 1.1410 1.1658 1.2031 1.2026 8.40 1.1378 1.1653 1.1934 1.1879 8.60 1.1348 1.1646 1.1837 1.1743 8.80 1.1385 1.1660 1.1796 1.1673 9.00 1.1426 1.1661 1.1837 1.1666 9.20 1.1451 1.1790 1.1933 1.1858 9.40 1.1487 1.1996 1.2034 1.2362 9.60 1.1530 1.2183 1.2495 1.2856 9.80 1.1615 1.2370 1.2826 1.3280 10.00 1.1690 1.2536 1.3130 1.3638 10.20 1.1793 1.2690 1.3393 1.4005 10.40 1.1924 1.2849 1.3613 1.4279 10.60 1.2197 1.2948 1.3681 1.4505 10.80 1.2329 1.2962 1.3799 1.4671 11.00 1.2329 1.2890 1.3881 1.4805 11.20 1.2433 1.2698 1.3412 1.4699 11.40 1.2487 1.3068 1.4021 1.4788 11.60 1.2271 1.2748 1.3699 1.4428 11.80 1.2234 1.2788 1.3535 1.4267 12.00 (core top) 1.2244 1.2892 1.3399 1.4205 Note: W(Z) values at 20000 MWD/MTU may be applied to cycle burnups greater than 20000 MWD/MTU to prevent W(Z) function extrapolation COLR BYRON 2 Revision 8 Page 8 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 2 CYCLE 20 Table 2.6.2.b EOL-only W(Z) versus Core Height for AFD Acceptable Operation Limits in Figure 2.8.1.b (Top and Bottom 8% Excluded per WCAP-10216) Height 18000 20000 23337 (feet) MWD/MTU MWD/MTU MWD/MTU 0.00 (core bottom) 1.2299 1.2430 1.2430 0.20 1.2130 1.2208 1.2208 0.40 1.2051 1.2127 1.2127 0.60 1.1962 1.2060 1.2060 0.80 1.1846 1.2040 1.2040 1.00 1.1749 1.1879 1.1879 1.20 1.1720 1.1819 1.1819 1.40 1.1677 1.1757 1.1757 1.60 1.1595 1.1660 1.1660 1.80 1.1529 1.1568 1.1568 2.00 1.1459 1.1458 1.1458 2.20 1.1342 1.1317 1.1317 2.40 1.1255 1.1187 1.1187 2.60 1.1179 1.1104 1.1104 2.80 1.1142 1.1065 1.1065 3.00 1.1137 1.1084 1.1084 3.20 1.1141 1.1131 1.1131 3.40 1.1205 1.1267 1.1267 3.60 1.1281 1.1441 1.1441 3.80 1.1356 1.1596 1.1596 4.00 1.1471 1.1745 1.1745 4.20 1.1592 1.1875 1.1875 4.40 1.1705 1.1995 1.1995 4.60 1.1795 1.2095 1.2095 4.80 1.1871 1.2174 1.2174 5.00 1.1936 1.2240 1.2240 5.20 1.1980 1.2274 1.2274 5.40 1.2019 1.2303 1.2303 5.60 1.2151 1.2473 1.2473 5.80 1.2340 1.2651 1.2651 6.00 1.2500 1.2787 1.2787 6.20 1.2617 1.2874 1.2874 6.40 1.2715 1.2954 1.2954 6.60 1.2754 1.2965 1.2965 6.80 1.2785 1.2967 1.2967 7.00 1.2773 1.2929 1.2929 7.20 1.2714 1.2843 1.2843 7.40 1.2630 1.2734 1.2734 7.60 1.2507 1.2584 1.2584 7.80 1.2358 1.2416 1.2416 8.00 1.2216 1.2231 1.2231 8.20 1.2051 1.2026 1.2026 8.40 1.1919 1.1879 1.1879 8.60 1.1794 1.1743 1.1743 8.80 1.1733 1.1673 1.1673 9.00 1.1749 1.1666 1.1666 9.20 1.1897 1.1858 1.1858 9.40 1.2219 1.2362 1.2362 9.60 1.2717 1.2856 1.2856 9.80 1.3110 1.3280 1.3280 10.00 1.3453 1.3638 1.3638 10.20 1.3783 1.4005 1.4005 10.40 1.4037 1.4279 1.4279 10.60 1.4195 1.4505 1.4505 10.80 1.4349 1.4671 1.4671 11.00 1.4472 1.4805 1.4805 11.20 1.4194 1.4699 1.4699 11.40 1.4519 1.4788 1.4788 11.60 1.4176 1.4428 1.4428 11.80 1.3999 1.4267 1.4267 12.00 (core top) 1.3888 1.4205 1.4205 Note: W(Z) values at 20000 MWD/MTU may be applied to cycle burnups greater than 20000 MWD/MTU to prevent W(Z) function extrapolation COLR BYRON 2 Revision 8 Page 9 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 2 CYCLE 20 Table 2.6.2.c Penalty Factors in Excess of 2% per 31 EFPD Cycle Burnup (MWD/MTU) Penalty Factor F C Q(z) 0 1.0200 853 1.0210 1029 1.0230 1205 1.0410 1381 1.0425 1556 1.0410 1908 1.0360 2436 1.0320 3139 1.0270 3666 1.0200 11929 1.0200 12105 1.0202 12281 1.0200 13336 1.0200 13512 1.0201 13688 1.0205 13863 1.0209 14039 1.0211 14215 1.0205 14391 1.0200 Notes: Linear interpolation is adequate for intermediate cycle burnups. All cycle burnups outside the range of Table 2.6.2.c shall use a 2% penalty factor for compliance with the 3.2.1.2 Surveillance Requirements.

COLR BYRON 2 Revision 8 Page 10 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 2 CYCLE 20 2.7 Nuclear Enthalpy Rise Hot Channel Factor (F NH) (LCO 3.2.2) 2.7.1 F NH FH RTP[1.0 + PFH(1.0 - P)] where: P = the ratio of THERMAL POWER to RATED THERMAL POWER (RTP)

FH RTP = 1.70 PFH = 0.3 2.7.2 Uncertainty: The uncertainty, U FH, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor F NH shall be calculated by the following formula:

U FH = U FHm where: U FHm = Base F NH measurement uncertainty = 1.04 when PDMS is inoperable (U FHm is defined by PDMS when OPERABLE.) 2.7.3 PDMS Alarms:

F NH Warning Setpoint = 2% F NH Margin F NH Alarm Setpoint = 0% F NH Margin 2.8 AXIAL FLUX DIFFERENCE (AFD) (LCO 3.2.3) 2.8.1 When PDMS is inoperable, the AXIAL FLUX DIFFERENCE (AFD) Acceptable Operation Limits are provided in the Figures described below or the latest valid PDMS Surveillance Report, whichever is more conservative. a) Figure 2.8.1.a is the full cycle AFD Acceptable Operation Limits associated with the full cycle W(Z) values in Table 2.6.2.a. b) Figure 2.8.1.b is the Reduced AFD Acceptable Operation Limits which may be applied after 18000 MWD/MTU. The Reduced AFD Acceptable Operation Limits are associated with the EOL-only W(Z) values in Table 2.6.2.b. Prior to changing to Figure 2.8.1.b, confirm that the plant is within the specified AFD envelope. 2.8.2 When PDMS is OPERABLE, no AFD Acceptable Operation Limits are applicable. 2.9 Departure from Nucleate Boiling Ratio (DNBR) (LCO 3.2.5) 2.9.1 DNBRAPSL 1.563 The Axial Power Shape Limiting DNBR (DNBR APSL) is applicable with THERMAL POWER 50% RTP when PDMS is OPERABLE. 2.9.2 PDMS Alarms: DNBR Warning Setpoint = 2% DNBR Margin DNBR Alarm Setpoint = 0% DNBR Margin COLR BYRON 2 Revision 8 Page 11 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 2 CYCLE 20 Figure 2.8.1.a: Axial Flux Difference Limits as a Function of Rated Thermal Power (Full Cycle)(-35, 50) (-15, 100) Acceptable Operation (+10, 100) (+25, 50) Unacceptable Operation Unacceptable Operation 0 20 40 60 80 100 120-50-40-30-20-1001020304050% of RATED THERMAL POWER AXIAL FLUX DIFFERENCE (%) Axial Flux Difference Limits with PDMS Inoperable COLR BYRON 2 Revision 8 Page 12 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 2 CYCLE 20 Figure 2.8.1.b: Reduced Axial Flux Difference Limits as a Function of Rated Thermal Power (Cycle burnup 18000 MWD/MTU) Axial Flux Difference LimitswithPDMS Inoperable 0 20 40 60 80 100 120-50-40-30-20-1001020304050AXIAL FLUX DIFFERENCE (%)% of RATED THERMAL POWE RUnacceptableOperationUnacceptableOperationAcceptableOperation(-20, 50)(-10, 100)(+10, 100)(+25, 50)

COLR BYRON 2 Revision 8 Page 13 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 2 CYCLE 20 2.10 Reactor Trip System (RTS) Instrumentation (LCO 3.3.1) - Overtemperature T Setpoint Parameter Values 2.10.1 The Overtemperature T reactor trip setpoint K 1 shall be equal to 1.325. 2.10.2 The Overtemperature T reactor trip setpoint T avgcoefficient K 2 shall be equal to 0.0297 / °F. 2.10.3 The Overtemperature T reactor trip setpoint pressure coefficient K 3 shall be equal to 0.00135 / psi. 2.10.4 The nominal T avgat RTP (indicated) T shall be less than or equal to 588.0 °F. 2.10.5 The nominal RCS operating pressure (indicated) P shall be equal to 2235 psig. 2.10.6 The measured reactor vessel T lead/lag time constant 1shall be equal to 8 sec. 2.10.7 The measured reactor vessel T lead/lag time constant 2shall be equal to 3 sec. 2.10.8 The measured reactor vessel T lag time constant 3shall be less than or equal to 2 sec. 2.10.9 The measured reactor vessel average temperature lead/lag time constant 4shall be equal to 33 sec. 2.10.10 The measured reactor vessel average temperature lead/lag time constant 5shall be equal to 4 sec. 2.10.11 The measured reactor vessel average temperature lag time constant 6shall be less than or equal to 2 sec. 2.10.12 The f 1 (I) "positive" breakpoint shall be +10% I. 2.10.13 The f 1 (I) "negative" breakpoint shall be -18% I. 2.10.14 The f 1 (I) "positive" slope shall be +3.47% / % I. 2.10.15 The f 1 (I) "negative" slope shall be -2.61% / % I.

COLR BYRON 2 Revision 8 Page 14 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 2 CYCLE 20 2.11 Reactor Trip System (RTS) Instrumentation (LCO 3.3.1) - Overpower T Setpoint Parameter Values 2.11.1 The Overpower T reactor trip setpoint K 4 shall be equal to 1.072. 2.11.2 The Overpower T reactor trip setpoint T avgrate/lag coefficient K 5 shall be equal to 0.02 / °F for increasing Tavg. 2.11.3 The Overpower T reactor trip setpoint T avgrate/lag coefficient K 5 shall be equal to 0 / °F for decreasing T avg. 2.11.4 The Overpower T reactor trip setpoint T avgheatup coefficient K 6 shall be equal to 0.00245 / °F when T T. 2.11.5 The Overpower T reactor trip setpoint T avgheatup coefficient K 6 shall be equal to 0 / °F when T T. 2.11.6 The nominal T avgat RTP (indicated) T shall be less than or equal to 588.0 °F 2.11.7 The measured reactor vessel T lead/lag time constant 1shall be equal to 8 sec. 2.11.8 The measured reactor vessel T lead/lag time constant 2shall be equal to 3 sec. 2.11.9 The measured reactor vessel T lag time constant 3shall be less than or equal to 2 sec. 2.11.10 The measured reactor vessel average temperature lag time constant 6shall be less than or equal to 2 sec. 2.11.11 The measured reactor vessel average temperature rate/lag time constant 7shall be equal to 10 sec. 2.11.12 The f 2 (I) "positive" breakpoint shall be 0 for all I. 2.11.13 The f 2 (I) "negative" breakpoint shall be 0 for all I. 2.11.14 The f 2 (I) "positive" slope shall be 0 for all I. 2.11.15 The f 2 (I) "negative" slope shall be 0 for all I.

COLR BYRON 2 Revision 8 Page 15 of 15 CORE OPERATING LIMITS REPORT (COLR) for BYRON UNIT 2 CYCLE 20 2.12 Reactor Coolant System (RCS) Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits (LCO 3.4.1) 2.12.1 The pressurizer pressure shall be greater than or equal to 2209 psig. 2.12.2 The RCS average temperature (T avg) shall be less than or equal to 593.1 °F. 2.12.3 The RCS total flow rate shall be greater than or equal to 386,000 gpm. 2.13 Boron Concentration 2.13.1 The refueling boron concentration shall be greater than or equal to the applicable value given in the Table below (LCO 3.9.1). The reported "prior to initial criticality" value also bounds the end-of-cycle requirements for the previous cycle. 2.13.2 To maintain keff 0.987 with all shutdown and control rods fully withdrawn in MODES 3, 4, or 5 (TRM TLCO 3.1.g Required Action B.2 and TRM TLCO 3.1.k.2), the Reactor Coolant System boron concentration shall be greater than or equal to the applicable value given in the Table below.

COLRSection ConditionsBoron Concentration (ppm) 2.13.1 a) prior to initial criticality 1733 b) for cycle burnups 0 MWD/MTU and < 16000 MWD/MTU 1794 c) for cycle burnups 16000 MWD/MTU 1387 2.13.2 a) prior to initial criticality 1759 b) for cycle burnups 0 MWD/MTU and < 16000 MWD/MTU 1943 c) for cycle burnups 16000 MWD/MTU 1480 BYRON UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)(November 2015)

BYRON -UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT i Table of Contents Section Page1.0 Introduction1 2.0 RCS Pressure and Temperature Limits 1 2.1 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3) 13.0 Low Temperature Over Pressure Protection and Boltup7 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 ii List of Figures Figure Page2.1Byron Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates of 100

!F/hr) Applicable for 32 EFPY (WithoutMargins for Instrumentation Errors) 32.2Byron Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50, and 100

!F/hr) Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 43.1Byron 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 iii List of TablesTablePage2.1aByron Unit 1 Heatup Data Points at 32 EFPY (Without Margins for Instrumentation Errors) 52.1bByron Unit 1 Cooldown Data Points at 32 EFPY (Without Margins for Instrumentation Errors) 63.1Data Points for Byron Unit 1 Nominal PORV Setpoints for the LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) 94.1Byron Unit 1 Surveillance Capsule Withdrawal Summary115.1Byron Unit 1 Calculation of Chemistry Factors Using Surveillance Capsule Data 135.2Byron Unit 1 Reactor Vessel Material Properties145.3Summary of Byron Unit 1 Adjusted Reference Temperature(ART) Values at 1/4T and 3/4T Locations for 32 EFPY 155.4RT PTSCalculation for Byron Unit 1 Beltline Region Materials at EOL (32 EFPY) 16 BYRON -UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 11.0Reactor 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 (P/T) Limits; andTS-LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System.2.0RCS Pressure and Temperature LimitsThis 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", andd)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 11, documents the technical basis for the elimination of the flange requirements. 2.1RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3) 2.1.1The 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, andc)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 TEMPERATURE LIMITS REPORT 22.1.2The RCS P/T limits for heatup, inservice hydrostatic and leak testing,and criticality are specified by Figure 2.1 and Table 2.1a. The RCS P/T limits for cooldown are

shown in Figure 2.2 and Table 2.1b. 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.

BYRON -UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGINGLIMITING ART VALUES AT 32 EFPY:1/4T, 106

!F 3/4T, 97!F-250 0 250 500 750 1000 1250 1500 1750 2000 2250 2500050100150200250300350400450500550Calculated Pressure (PSIG)Moderator Temperature (Deg. F)

Acceptable Operation Unacceptable Operation Boltup Temp.60°F Heatup Rate 100 Deg. F/Hr Leak Test Limit Critical Limit

100 Deg. F/HrCriticality Limit based on inservice hydrostatic test temperature (166

°F) for the service period up to 32 EFPYThe lower limit for RCS pressure is -14.7 psig Figure 2.1 Byron Unit 1 Reactor Coolant System Heatup Limitations (Heatup rates of 100°F/hr) Applicable for 32EFPY (Without Margins for Instrumentation Errors)

BYRON -UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 4 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGINGLIMITING ART VALUES AT 32 EFPY:1/4T, 106

!F 3/4T, 97

!F -250 0 250 500 750 1000 1250 1500 1750 2000 2250 2500050100150200250300350400450500550Calculated Pressure (PSIG)Moderator Temperature (Deg. F)Acceptable Operation Unacceptable Operation Boltup Temp.

60°F CooldownRates, °F/Hrsteady-state,-25,

-50, and-100The lower limit for RCS

pressure is -14.7 psig Figure 2.2 Byron Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown rates of 0, 25, 50 and 100!!F/hr) Applicable for 32 EFPY (Without Margins for Instrumentation Errors)

BYRON -UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 5 Table 2.1a Byron Unit 1 Heatup Data Points at 32 EFPY (Without Margins for Instrumentation Errors)

Heatup Curve100 F HeatupCriticality Limit Leak Test Limit T (°F)P (p si g)T (°F)P (p si g)T (°F)P (p si g)60-14.7166-14.71492000607201667201662485657201667207072016672075720166720807201667208572016672090720166720957201667231007231667291057291667371107371667491157491667641207641667811257811708021308021758261358261808541408541858861458861909211509211959621559622001007160100720510571651057210111317011132151175175117522012441801244225132118513212301406190140623514991951499240160320016032451718205171825018442101844255198421519842602138220213826523082252308 BYRON -UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 6 Table 2.1b Byron Unit 1 Cooldown Data Points at 32 EFPY (Without Margins for Instrumentation Errors) Note: For each cooldown rate, the steady-state pressure values shall govern the temperature where no allowable pressure values are provided.

Cooldown Curves Steady State 25 !F Cooldown50

!F Cooldown100

!F Cooldown T (°F)P (p si g)T (°F)P (p si g) T (°F)P (p si g)T (°F)P (p si g)60-14.760-14.760-14.760-14.760753607096066560581657696572665685656067078770746707067063375806757677573075663808278079180757806978585185817857868573590877908469081990777959069587995855958231009371009141008951008741059731059541059401059311101011110997110989115105411510451151043120110212010991251154130121213512761401347145142515015121551607160171316518291701958175210118022581852433 BYRON -UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 73.0Low Temperature Overpressure Protection and BoltupThis 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.3.1LTOP 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 P/T 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.3.2LTOP 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.3.3Reactor Vessel Boltup Temperature (Non-Technical Specification)

The minimum boltup temperature for the Reactor Vessel Flange shall be

1. 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).

BYRON -UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 8 595 psig 2335 psig 541 psig 0 250 500 750 1000 1250 1500 1750 2000 2250 2500050100150200250300350400450Auctioneered Low RCS Temperature (DEG. F)Nominal PORV Pressure (PSIG)

PCV-456 PCV-455A Unacceptable Operation 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 9 Table 3.1 Data Points for Byron Unit 1 Nominal PORV Setpoints for the LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)PCV-455APCV-456(1TY-0413M)(1TY-0413P)AUCTIONEERED LOWRCS PRESSUREAUCTIONEERED LOWRCS PRESSURERCS TEMP. (DEG. F)(PSIG)RCS TEMP. (DEG. F)(PSIG)605416059530054130059540023354002335 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.)

BYRON -UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 104.0Reactor Vessel Material Surveillance ProgramThe 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, RT NDT, which is determined in accordance with ASME Boiler and Pressure Vessel Code,Section III, NB-2331. The empirical relationship between RT NDTand 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(CapsuleW)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 removedand placed in the spent fuel poolto avoid excessive fluence accumulation should they be needed to support life extension. The removal summaryis provided in Table 4.1.

BYRON -UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 11 Table 4.1 Byron Unit 1 Surveillance Capsule Withdrawal Summary (a)Capsule Capsule LocationLeadFactorWithdrawal EFPY (b)Fluence (n/cm 2,E > 1.0 MeV)

U 58.5!4.051.180.409 x 10 19 X 238.5!4.095.671.49 x 10 19 W 121.5!4.089.272.26 x 10 19 Z (c)301.5!4.1114.59 (EOC 12)3.34x10 19 V (c)61.0!3.8914.59 (EOC 12)3.16x10 19 Y (c)241.0!3.8518.81 (EOC 15)3.97x10 19 Notes:(a)Source document is CN-AMLRS-10-8 (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 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.

BYRON -UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 125.0Supplemental Data TablesThe following tables provide supplemental information on reactor vessel material properties and are provided to be consistent with Generic Letter 96-03. Some of the material property

values shown were used as inputs to the P/T limits.Table 5.1 shows the calculation of the surveillance material chemistry factors using surveillance capsule data.

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 3/4T locations for 32 EFPY.

Table 5.4 provides the RT PTS values forByron Unit 1 for 32 EFPY obtained from Reference 4.

BYRON -UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 13 Table 5.1 Byron Unit 1 Calculation of Chemistry Factors Using Surveillance Capsule Data (a)MaterialCapsuleCapsulef (b)(n/cm 2, E > 1.0MeV)

FF (c)RT NDT ( b)(!F)FF*RT NDT (!F)FF 2Intermediate Shell Forging (Tangential)U4.09 x10 190.75228.5521.470.57X1.49 x10 191.1109.8210.901.23W2.26 x10 191.22149.2060.061.49 Intermediate Shell Forging (Axial)U0.409 x10 190.75218.5213.930.57X1.49 x10 191.11053.0358.891.23W2.26 x10 191.22129.3435.821.49Sum:201.066.58 CFISForging= %(FF *RT NDT) &  %(FF 2) = (201.06)

& (6.58) =

30.6!FByron Unit 1 Surveillance Weld Material (Heat #442002)U4.09 x 10 19 0.752 11.22 (5.61)8.440.57X1.49 x 10 19 1.110 80.22 (40.11)89.081.23W2.26 x 10 19 1.221 102.68 (51.34)125.34 1.49Byron Unit 2 Surveillance Weld Material (Heat #442002)U0.406 x 10 19 0.750 16.88 (8.44)12.66 0.56W1.20 x 10 19 1.051 57.76 (28.88)60.701.10 X 2.18 x 10 19 1.211 108.02 (54.01)130.861.47SUM:427.086.42 CF Weld Metal

= %(FF

• RT NDT) &  %(FF 2) = (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; RT NDT values are the measured 30 ft-lb shift values taken from Reference13. RT NDT values for the surveillance weld dataare adjusted by a ratio of 2.0 (pre-adjusted values are listed in parentheses).c)FF = fluence factor = f(0.28 -0.10*log f).

BYRON -UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 14 Table 5.2 Byron Unit 1 Reactor Vessel Material Properties (a)Material DescriptionCu (%)Ni (%)Initial RT NDT (ºF)(b)Closure Head Flange 124K358VA1--0.7460Vessel Flange 123J219VA1--0.7310Nozzle Shell Forging 123J2180.050.7230Intermediate ShellForging 5P-59330.040.7440Lower Shell Forging 5P-59510.040.6410Intermediate to Lower Shell Forging Circ. Weld Seam WF-336 (Heat # 442002)0.040.63-30Nozzle Shell to Intermediate Shell Forging Circ. Weld Seam WF-501 (Heat # 442011)0.030.6710Byron Unit 1 Surveillance Program Weld Metal (Heat # 442002)0.020.69--Byron Unit 2 Surveillance Program Weld Metal (Heat # 442002)0.020.71--

Braidwood Units 1 & 2 Surveillance Program Weld Metals (Heat # 442011) 0.03 0.67, 0.71--a)Reference 7.b)The initial RT NDT values for the plates and welds are based on measured data.

BYRON -UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 15 Table 5.3 Summary of Byron Unit 1 Adjusted Reference Temperature (ART) Values at 1/4T and 3/4T Locations for 32 EFPY (a)Reactor Vessel Material Surface Fluence (n/cm 2 , E > 1.0 MeV) 32 EFPY1/4T ART (ºF)3/4T ART (ºF)

Nozzle Shell Forging0.598x 10 197459 Intermediate Shell Forging1.77x 10 199378Using non-credible surveillance data1.77x 10 1910285 Lower Shell Forging1.77x 10 196348 Nozzle to Intermediate Shell ForgingCirc. Weld Seam(Heat # 442011)0.598x 10 196949Using credible Braidwood Units1and 2 surveillance data0.598x 10 194735 Intermediate to Lower Shell ForgingCirc. Weld Seam(Heat # 442002)1.72x 10 197949Using credible surveillance data1.72x 10 196546 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.

BYRON -UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 16 Table 5.4 RT PTS Calculation for Byron Unit 1 Beltline Region Materials at EOL (32 EFPY) (a,b)Reactor Vessel Material R.G. 1.99, Rev. 2 Position CF (°F)Fluence (n/cm 2 , E > 1.0 MeV)FFIRT NTD (C)(°F)RT NTD (°F) u (c)(°F)(d)(°F)MarginRT PTS (°F)Nozzle Shell Forging1.1310.598x 10 190.856030 26.5 013.326.583 Intermediate Shell Forging1.1261.77x 10 191.156940 30.1 015.030.1100Using non-credible surveillance data2.130.61.77x 10 191.156940 35.4 01734 109 Lower Shell Forging1.1261.77x 10 191.156910 30.1 015.030.170 N ozzle to Intermediate Shell Forging Circ. Weld Seam (Heat # 442011)1.1410.598x 10 190.85601035.1017.535.180Using credible Braidwood Units 1 and 2 surveillance data2.126.10.598x10 190.856010 22.3011.222.355 Intermediate to Lower shell Forging Circ Weld Seam (Heat # 442002)1.1541.72x 10 191.1492-30 62.10285688Using credible surveillance data2.166.51.72x 10 191.1492-30 76.4 0142874 Notes:(a)The 10 CFR 50.61 methodology was utilized in the calculation of the RT PTS values.(b)The source document containing detailed calculations is CN-AMLRS-10-8 (Reference 4), Table 5.5-1.(c)Initial RT NDT values are based on measured data. Hence u = 0°F.(d)Per the guidance of 10 CFR 50.61, the base metal = 17°F for Position 1.1and for Position 2.1 with non-credible surveillance data; the weld metal = 28°F for Position 1.1 (without surveillancedata) and with credible surveillance data= 14°F for Position 2.1. However, need not to exceed 0.5*RT NTD BYRON -UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 176.0References1.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 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 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, to O. 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.

BYRON -UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 1810.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-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-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 Commonwealth Edison Company Byron Unit 1 Reactor Vessel Radiation Surveillance Program," T.J. Laubham, et al, January

1999.

BYRON UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)(November 2015)

BYRON -UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT i Table of Contents Section Page1.0 Introduction1 2.0 RCS Pressure and Temperature Limits 1 2.1 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3) 13.0 Low Temperature Over Pressure Protection and Boltup7 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 105.0 Supplemental Data Tables 12 6.0 References 17 BYRON -UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT ii List of Figures Figure Page2.1Byron Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 100

!F/hr) Applicable for 30.5EFPY (Without Margins for Instrumentation Errors) 32.2Byron Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50, and 100

!F/hr) Applicable for 30.5EFPY (Without Margins for Instrumentation Errors) 43.1Byron Unit 2 Nominal PORV Setpoints for the Low Temperature Overpressure Protection(LTOP) System Applicable for 30.5EFPY (Includes Instrumentation Uncertainty) 8 BYRON -UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT iii List of TablesTablePage2.1aByronUnit 2 Heatup Data Points at 30.5EFPY (Without Margins for Instrumentation Errors) 52.1bByron Unit 2 Cooldown Data Points at 30.5EFPY (Without Margins for Instrumentation Errors) 63.1Data Points for Byron Unit 2 Nominal PORV Setpoints for the LTOP System Applicable for 30.5EFPY (Includes Instrumentation Uncertainty) 94.1Byron Unit 2 Surveillance Capsule Withdrawal Summary115.1Byron Unit 2 Calculation of Chemistry Factors Using Surveillance Capsule Data 135.2Byron Unit 2 Reactor Vessel Material Properties145.3Summary of Byron Unit 2 Adjusted Reference Temperatures (ART)Valuesat 1/4T and 3/4T Locations for 32 EFPY 155.4RT PTSCalculation for Byron Unit 2 Beltline Region Materials atEOL(32 EFPY) 16 BYRON -UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 11.0Reactor 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 (P/T) Limits; and TS-LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System.2.0RCS Pressure and Temperature LimitsThis section provides the Byron Unit 2Heatup 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",

andd)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 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. 2.1RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3)2.1.1The 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, andc.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 22.1.2The RCS P/T limits for heatup, inservice hydrostatic and leak testing, and criticality are specified by Figure 2.1 and Table 2.1a. The RCS P/T limits for

cooldown are shown in Figure 2.2 and Table 2.1b. These limits are defined in

WCAP-15392, Revision 2 (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.

BYRON -UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 3 MATERIAL PROPERTY BASISLIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-447 & NOZZLE SHELL FORGINGLIMITING ART VALUES AT 30.5EFPY:1/4T, 107°F (N-588) & 52

!F ('96 App. G) 3/4T, 89°F (N-588) & 37

!F ('96 App. G)

Figure 2.1 Byron Unit 2 Reactor Coolant System Heatup Limitations (Heatup rates of 100

!!F/hr) Applicable for 30.5EFPY (Without Margins for Instrumentation Errors)

-250 0 250 500 750 1000 1250 1500 1750 2000 2250 2500050100150200250300350400450500550Calculated Pressure (PSIG)Moderator Temperature (Deg. F)

Acceptable Operation Unacceptable Operation Boltup Temp. 60°F Heatup Rate 100 Deg. F/Hr Leak Test Limit Critical Limit 100 Deg. F/HrCriticality Limit based on inservice hydrostatic test temperature (112

°F) for the service period up to 30.5 EFPYThe lower limit for RCS pressure is -14.7 psig BYRON -UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 4 MATERIAL PROPERTY BASISLIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-447 & NOZZLE SHELL FORGINGLIMITING ART VALUES AT 30.5EFPY:1/4T, 107°F (N-588) & 52

!F ('96 App. G) 3/4T, 89°F (N-588) & 37

!F ('96 App. G)

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

-250 0 250 500 750 1000 1250 1500 1750 2000 2250 2500050100150200250300350400450500550Calculated Pressure (PSIG)Moderator Temperature (Deg. F)

Acceptable Operation Unacceptable Operation Boltup Temp. 60°F CooldownRates, °F/Hrsteady-state,-25,

-50, and-100The lower limit for RCS pressure is -14.7 psig BYRON -UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 5 Table 2.1aByronUnit 2 Heatup Data Points at 30.5EFPY (Without Margins for Instrumentation Errors)

Heatup Curve100 F HeatupCriticality Limit Leak Test Limit T (°F)P (p si g)T (°F)P (p si g)T (°F)P (p si g)60-14.7112-14.795200060103011210781122485651078112112870112811511487511481201152801152125116285116213011809011801351205951205140123710012371451276105127615013231101323155137711513771601440120144016515111251511170159213015921751682135168218017841401784185189714518971902023150202319521201552120200222716022272052347165234721024801702480 BYRON -UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 6 Table 2.1bByron Unit 2 Cooldown Data Points at 30.5EFPY (Without Margins for Instrumentation Errors) Note: For each cooldownrate, the steady-state pressure values shall govern the temperature where no allowable pressure values are provided.

Cooldown Curves Steady State 25 !F Cooldown50

!F Cooldown100

!F Cooldown T (°F)P (p si g)T (°F)P (p si g) T (°F)P (p si g)T (°F)P (p si g)60-14.760-14.760-14.760-14.76010456010366010336510926510887011437512008012638513329014099514941001587105169111018051151932120207112522261302396 BYRON -UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 73.0Low Temperature Overpressure Protection and BoltupThis 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.3.1LTOP 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.3.2LTOP 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.3.3Reactor Vessel BoltupTemperature (Non-Technical Specification)

The minimum boltup temperature for the Reactor Vessel Flange shall be

1. 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).

BYRON -UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 8 Figure 3.1 Byron Unit 2 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for the 30.5EFPY (Includes Instrumentation Uncertainty) 639 psig 2335 psig 599 psig 0 250 500 750 1000 1250 1500 1750 2000 2250 2500050100150200250300350400450Auctioneered Low RCS Temperature (DEG. F)Nominal PORV Pressure (PSIG)

PCV 456 PCV 455A Unacceptable Operation BYRON -UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 9Table3.1 Data Points for Byron Unit 2 Nominal PORV Setpoints for the LTOP System Applicable for 30.5EFPY (Includes Instrumentation Uncertainty)PCV-455APCV-456(2TY-0413M)(2TY-0413P)AUCTIONEERED LOWRCS PRESSUREAUCTIONEERED LOWRCS PRESSURERCS TEMP. (DEG. F)(PSIG)RCS TEMP. (DEG. F)(PSIG)505995063930059930063940023354002335Note: 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.)

BYRON -UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 104.0Reactor Vessel Material Surveillance Program The pressure vessel material surveillance program (Reference 4) is in compliance with AppendixH to 10 CFR 50, "Reactor Vessel Radiation Surveillance Program." The material test requirements and the acceptance standard utilize the reference nil-ductility

temperature, RT NDT , which is determined in accordance with ASME Boiler and Pressure Vessel Code Section III, NB-2331. The empirical relationship between RT NDT 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. Theremaining threecapsules, V, Y and Z, wereremovedand placed in the spent fuel poolto avoid excessive fluence accumulation should they be needed to support life extension. The removal summary is provided in Table 4.1.

BYRON -UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 11 Table 4.1Byron Unit 2Surveillance Capsule Withdrawal Summary (a)Capsule Capsule LocationLeadFactorWithdrawal EFPY (b)Fluence (n/cm 2,E > 1.0 MeV)

U 58.5!4.021.190.406 x 10 19 W 121.5!4.074.671.20 x 10 19 X 238.5!4.148.632.18x 10 19 Z (c)301.5!4.1114.28 (EOC 11)3.25 x10 19 V (c)61.0!3.8814.28 (EOC 11)3.07x10 19 Y (c)241.0!3.8820.05 (EOC 15)4.19x10 19 Notes:(a)Source document is CN-AMLRS-10-8 (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.

BYRON -UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 125.0Supplemental Data Tables The following tables provide supplemental information on reactor vessel material properties and are provided to be consistent with Generic Letter 96-03. Some of the material property values shown were used as inputs to the P/T limits.Table 5.1 shows the calculation of the surveillance material chemistry factors using surveillance capsule data.

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)valuesat the 1/4T and 3/4T locations for 32 EFPY.Table 5.4provides the RT PTS values for Byron Unit 2 for 32 EFPY obtained from Reference 5.

BYRON -UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 13 Table 5.1 Byron Unit 2 Calculation of Chemistry Factors Using Surveillance Capsule Data (a)MaterialCapsuleCapsulef (b)(n/cm 2, E > 1.0MeV)

FF (c)RT NDT( b)(!F)FF*RT NDT (!F)FF 2Lower Shell Forging (Tangential)U4.06x 10 190.7500.0 (d)0.000.56W1.20 x 10 191.0513.653.841.10X2.18 x 10 191.21115.7519.081.47Lower Shell Forging (Axial)U04.06 x 10 190.75019.7614.820.56W1.20 x 10 191.05131.8833.501.10X2.18 x 10 191.21138.9147.141.47SUM:118.386.27 CFLSForging= %(FF *RT NDT) &  %( FF 2) = (118.38)

& (6.27) = 18.9!FByron Unit 1 Surveillance Weld Material (Heat #442002)U0.409 x10 19 0.752 11.22(5.61)8.440.57X1.49 x10 19 1.110 80.22 (40.11)89.081.23W2.26 x10 19 1.221102.68 (51.34)125.34 1.49Byron Unit 2 Surveillance Weld Material (Heat #442002)U0.406 x 10 19 0.750 16.88 (8.44)12.66 0.56W1.20 x 10 19 1.051 57.76 (28.88)60.701.10X2.18 x 10 19 1.211 108.02 (54.01)130.861.47SUM:427.086.42 CF Weld Metal

= %(FF

• RT NDT) &  %( FF 2) = (427.08)

& (6.42) = 66.5!F Notes:a)Source document is CN-AMLRS-10-8 (Reference 5), Table 5.2-2.b)f = fluence; RT NDT values are the measured 30 ft-lb shift values taken from Reference9. RT NDT values for the surveillance weld dataare adjusted by a ratio of 2.0 (pre-adjusted values are listed in parentheses).c)FF = fluence factor = f(0.28 -0.10*log f)

.d)Measured RT NDT value was determined to be negative, but physically a reduction should not occur; therefore a conservative value of zero is used.

BYRON -UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 14 Table 5.2 Byron Unit 2 Reactor Vessel Material Properties (a)Material DescriptionCu (%)Ni (%)

Initial RT NDT (ºF)(b)Closure Head Flange 5P7382 / 3P6407--0.710Vessel Flange 124L556VA1--0.7030Nozzle Shell Forging 4P-61070.050.7410Inter. Shell Forging [49D329/49C297]-1-10.010.70-20Lower Shell Forging [49D330/49C298]-1-10.060.73-20Circumferential Weld WF-447 (HT# 442002)0.040.6310Upper Circumferential Weld WF-562 (HT# 442011)0.030.6740Byron Unit 1 Surveillance Program Weld Metal (Heat # 442002)0.020.69--Byron Unit 2 Surveillance Program Weld Metal (Heat # 442002)0.020.71--

Braidwood Units 1 & 2 Surveillance Program Weld Metal (Heat # 442011) 0.03 0.67, 0.71--a)Reference 7.b)Initial RT NDT values are based on measured data.

BYRON -UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 15 Table 5.3 Summary of Byron Unit 2 Adjusted Reference Temperatures (ART) Values at1/4T and 3/4T Locations for32 EFPY (a)Reactor Vessel Material Surface Fluence (n/cm 2 , E > 1.0 MeV)32EFPY1/4T ART (ºF)3/4T ART (ºF)

Nozzle Shell Forging 0.549 x 10 195338 Intermediate Shell Forging 1.76 x 10 19219Lower Shell Forging1.76 x 10 195234 Using credible surveillance data 1.76 x 10 19168 Nozzle to Intermediate Shell Forging Circ. Weld Seam (Heat # 442011) 0.549 x 10 199777Using credible Braidwood Units 1 and 2 surveillance data 0.549 x 10 197664 Intermediate to Lower Shell Forging Circ. Weld Seam (Heat # 442002) 1.70 x 10 1911988 Using credible surveillance data 1.70 x 10 1910586 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 P/T limit curves.

BYRON -UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 16 Table 5.4 RT PTSCalculation for Byron Unit 2 Beltline Region Materials atEOL (32 EFPY) (a,b)Reactor Vessel Material R.G. 1.99, Rev. 2 Position CF (°F)Fluence (n/cm 2 , E > 1.0 MeV)FFIRT NTD (C)(°F)RT NTD (°F) u (c)(°F)(d)(°F)MarginRT PTS (°F)Nozzle Shell Forging1.1310.549 x 10 190.832310 25.8 012.925.862 Intermediate Shell Forging1.120 1.76 x 10 191.1554-20 23.1 011.623.126Lower Shell Forging1.137 1.76 x 10 191.1554-20 42.7 0173457 Using credible surveillance data2.118.9 1.76 x 10 191.1554-20 21.8 08.51719 Nozzle to Intermediate Shell Forging Circ. Weld Seam (Heat # 442011)1.1410.549 x 10 190.83234034.1017.134.1108Using credible Braidwood Units 1 and 2 surveillance data2.126.10.549 x10 190.832340 21.7 010.921.783 Intermediate to Lower shell Forging Circ Weld Seam (Heat # 442002)1.1541.70 x 10 191.146110 61.902856128 Using credible surveillance data2.166.5 1.70 x 10 191.146110 76.2 01428114 Notes:(a)The 10 CFR 50.61 methodology was utilized in the calculation of the RT PTSvalues.(b)The source document containing detailed calculations is CN-AMLRS-10-8 (Reference 5), Table 5.5-2.(c)Initial RT NDT values are based on measured data. Hence u = 0°F.(d)Per the guidance of 10 CFR 50.61, the base metal = 17°Ffor Position 1.1 (without surveillance data) and with credible surveillance data= 8.5°Ffor position 2.1; the weld metal = 28°Ffor position 1.1 (without surveillance data) and with credible surveillance data= 14°Ffor Position 2.1. However, need not to exceed 0.5*RT NTD.

BYRON -UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 176.0References1.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 CooldownLimit 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 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.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 O. 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-15176, 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.

BYRON -UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 1812.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 Byron/Braidwood Units 1 and 2," W. Bamford, et al., October 2014.