BYRON 2021-0063, Reactor Coolant System Pressure and Temperature Limits Report

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Reactor Coolant System Pressure and Temperature Limits Report
ML21271A242
Person / Time
Site: Byron Constellation icon.png
Issue date: 09/28/2021
From: Kowalski J
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
BYRON 2021-0063
Download: ML21271A242 (26)
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Text

Byron Generating Station Exelon Generation 4450 North German Church Rd Byron. IL 61010-9794 www exeloncorp.com September 28, 2021 LTR: BYRON 2021-0063 File: 2.01 .0300 (5A.101) 1.10.0101 (1D.101)

United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Byron Station Unit 1 Renewed Facility Operating License No. NPF-37 NRC Docket No. STN 50-454

Subject:

Byron Station Unit 1 Reactor Coolant System Pressure and Temperature Limits Report In accordance with Technical Specification 5.6.6, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)", we are submitting the Unit 1 PTLR.

Should you have any questions concerning this report, please contact Ms. Lisa Zurawski, Regulatory Assurance Manager, at (815) 406-2800.

Respectfully, John J. Kowalski Site Vice President Byron Generating Station

Attachment:

Byron Station Unit 1 RCS PTLR JJK/LZ/mf cc: Regional Administrator- NRC Region Ill NRC Senior Resident Inspector - Byron Station

BYRONUNITl PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

(October 2020)

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 Overpressure 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 IO 5.0 Supplemental Data Tables 12 6.0 References 20

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/hr) Applicable for 57 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 57 EFPY (Without Margins for Instrumentation Errors) 3.1 Byron Unit 1 Nominal PORV Setpoints for the Low Temperature 8 Overpressure Protection (LTOP) System Applicable for 57 EFPY (Includes Instrumentation Uncertainty) ii

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT List of Tables Table Page 2.la Byron Unit 1 Heatup Data Points at 57 EFPY (Without Margins 5 for Instrumentation Errors) 2.lb Byron Unit 1 Cooldown Data Points at 57 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 57 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 I Reactor Vessel Material Properties 14 5.3 Summary of Byron Unit 1 Adjusted Reference Temperature 16 (ART) Values at 1/4T and 3/4T Locations for 57 EFPY 5.4 RTrrs Calculation for Byron Unit 1 Beltline Region Materials at 18 EOLE (57 EFPY) iii

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 Introduction 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; 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-A, Revision 4 (Reference 1) was used with the following exceptions:

a) Elimination of the flange requirements documented in WCAP-16143-P.

b) The initial reference temperature of the outlet nozzle forging to shell weld (WF-419) is determined using BAW-2308 in lieu of the ASME NB-2300 requirements.

WCAP-18371-NP, Revision 0, Reference 4, 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 "Master Curve" fracture toughness properties from BA W-2308 Revision 1-A Safety Evaluation (SE) and Revision 2-A SE (Reference 2) are used for one outlet nozzle to upper shell forgings weld. WCAP-16143-P ( Reference 5) documents the technical basis for the elimination of the flange requirements. These exceptions to the methodology in WCAP-14040-A, Revision 4 have been reviewed and accepted by the NRC in References 6, 7, and 10.

2.1 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3) 2.1. l The RCS temperature rate-of-change limits defined in Reference 4 are:

a) A maximum heatup of 100°F in any I-hour period.

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

BYRON - UNIT 1 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.la. The RCS PIT limits for cooldown are shown in Figure 2.2 and Table 2.1 b. These limits were developed in WCAP-18371-NP, Rev. 0 (Reference 4) using the limiting material between Byron Units 1 and 2. This approach is conservative. 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, 1998 Edition through 2000 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: BYRON UNIT I INTERMEDIATE SHELL FORGING SP-5933 LIMITING ART VALUES AT 57 EFFECTIVE FULL POWER YEARS (EFPY): 1/4T, I 02°F 3/4T, 87°F 2500 -r-- ------;------============.i OperlimAnalysis Version:5.4 Run:13859 Operlim.xlsm Version: 5.4.1 ILeak Test Limitl 2250 '-

2000 Unacceptable 0 eratlon 1750

( !)

in 1500

-e a.

,n 1250 Critical Limit 100 De . F/Hr

,n e

a.

"C 1000 s Acceptable

.!! 0 eration CJ "ii 750 0

500 Criticality Limit based on inservice hydrostatic test temperature (162°F) for the service period up to 57 EFPY 250 0

Lower limit for RCS pressure is 0 psia

-250 ----~...........~...........~...........~...........~............~.......,_~.......,_~.......,_~.......,_~..........................

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

Figure 2.1 Byron Unit 1 Reactor Coolant System Heatup Limitations (Heatup rates of 100°F/hr)

Applicable for 57 EFPY (Without Margins for Instrumentation Errors) 3

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: BYRON UNIT I INTERMEDIATE SHELL FORGING 5P-5933 LIMITING ART VALUES AT 57 EFPY: 1/4T, I 02°F 3/4T, 87°F 2500 ,---- - - - - -- -- - -============ OperlimAnalysis Version :5.4 Run :13859 Operlim.xlsm Version : 5.4.1 2250 2000 Unacceptable 0 eration 1750

<!> 1500 U)

-e Q.

~ 1250 in in Acceptable e

Q. 0 eration i:, 1000

.s ftS ooldown

i Rates u teady-state ci 750 0 25°F/hr 50°F/hr 100°F/hr 500 250 0

~ Lower limit for RCS pressure is 0 psia I

-250 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°F/hr) Applicable for 57 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 57 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 Note I 162 Note 1 145 2000 60 764 162 839 162 2485 65 764 165 851 70 764 170 877 75 764 175 908 80 764 180 943 85 764 185 982 90 764 190 1027 95 766 195 1076 100 772 200 1132 105 781 205 I 194 110 793 210 1263 115 809 215 1339 120 828 220 1424 125 851 225 1518 130 877 230 1622 135 908 235 1737 140 943 240 1864 145 982 245 2004 150 1027 250 2159 155 1076 255 2330 160 1132 165 1194 170 1263 175 1339 180 1424 185 1518 190 1622 195 1737 200 1864 205 2004 210 2159 215 2330 Note:

1. The minimum acceptable pressure is Opsia.

5

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.lb Byron Unit 1 Cooldown Data Points at 57 EFPY (Without Margins for Instrumentation Errors)

Cooldown Curves Steady State 25 °F Cooldown 50 °F Cooldown I00 °F Cooldown T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T(°F) P (psig) 60 Note I 60 Note 1 60 Note 1 60 Note I 60 766 60 723 60 681 60 601 65 783 65 742 65 702 65 628 70 802 70 763 70 726 70 658 75 823 75 786 75 752 75 691 80 846 80 812 80 780 80 727 85 871 85 840 85 812 85 768 90 900 90 872 90 848 90 814 95 931 95 907 95 887 95 864 100 965 100 945 100 930 100 920 105 1003 105 988 105 979 105 979 110 1045 110 1035 110 1032 110 1032 115 1092 115 1088 115 1088 115 1088 120 1143 120 1143 120 1143 120 1143 125 1200 125 1200 125 1200 125 1200 130 1263 130 1263 130 1263 130 1263 135 1332 135 1332 135 1332 135 1332 140 1409 140 1409 140 1409 140 1409 145 1494 145 1494 145 1494 145 1494 150 1587 150 1587 150 1587 150 1587 155 1691 155 1691 155 1691 155 1691 160 1805 160 1805 160 1805 160 1805 165 1932 165 1932 165 1932 165 1932 170 2071 170 2071 170 2071 170 2071 175 2226 175 2226 175 2226 175 2226 180 2396 180 2396 180 2396 180 2396 Note:

I. The minimum acceptable pressure is O psia.

6

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

3.1 LTOP System Setpoints (LCO 3.4.12)

Two PORVs shall have maximum lift settings in accordance with Figure 3.1 and Table 3.1.

These settings are based on the LTOP calculation in Reference 3.

The LTOP setpoints are based on PIT limits that were established in accordance with IO CFR 50, Appendix G without allowance for instrumentation error. The LTOP setpoints were developed using the methodology described in Reference I . The LTOP PORV lift settings shown in Figure 3.1 and Table 3.1 account for appropriate instrument error.

3.2 LTOP Enable Temperature Byron Unit I procedures governing the heatup and cooldown of the RCS require the arming of the LTOP system for RCS temperature less than 350°F and disarming of the L TOP system for RCS temperature of 350°F and above.

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

3.3 Reactor Vessel Boltup Temperature (Non-Technical Specification)

The minimum boltup temperature for the Reactor Vessel Flange shall be ~ 60°F. Boltup is a condition in which the Reactor Vessel head is installed with tension applied to any stud, and with the RCS vented to atmosphere.

7

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT

- PCV 455A - - PCV 456 2300.00 1800.00 Cl)

Unacceptable Operation ct 1300.00 VJ

~

Acceptable Operation 800.00 -

300.00 0 so 100 150 200 250 300 350 400 450 Indicated RCS Temperature (°F)

Figure 3.1 Byron Unit 1 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for 57 EFPY (Includes Instrumentation Uncertainty) 8

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 3.1 Data Points for Byron Unit 1 Nominal PORV Setpoints for the LTOP System Applicable for 57 EFPY (Includes Instrumentation Uncertainty)

PCV-455A PCV-456 (I TY-0413M) (ITY-0413P)

AUCTIONEERED LOW RCS PRESSURE AUCTIONEERED LOW RCS PRESSURE RCS TEMP. (DEG. F) (PSIG) RCS TEMP. (DEG. F) (PSIG) 60 541 60 618 300 541 300 618 310 720 310 789 320 899 320 961 330 1079 330 1133 340 1258 340 1304 350 1438 350 1476 360 1617 360 1648 400 2335 400 2335 Note: 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 8) 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, RTNoT, which is determined in accordance with ASME Boiler and Pressure Vessel Code,Section III, NB-2331. The empirical relationship between RTNoT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, "Protection Against Non-Ductile Failure," to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM El 85-82.

The fourth reactor vessel material irradiation surveillance specimens (Capsule Y) have been removed and analyzed to determine changes in the reactor vessel material properties. The surveillance capsule testing has been completed for the licensed operating period. The remaining two capsules, V 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<a>

Capsule Withdrawal EFPY(b) Fluence Capsule Lead Factor Location (n/cm 2, E > 1.0 MeV) u 58.5° 4.03 1.18 (EOC l)<d> 0.409 X 10 19 19 X 238.5° 4.08 5.67 (EOC 5) 1.49 X 10 w 121.5° 4.08 9.27 (EOC 8) 2.26 X 10 19 y 241.0° 3.87 18.81 (EOC 15) 3.97 X 10 19 z(c) 301.5° 4.11 14.59 (EOC 12) 3.34 X 10 19 y(c) 61.00 3.89 14.59 (EOC 12) 3.16xl0 19 Notes:

(a) Source document is WCAP-18054-NP (Reference 9), Table 7-1 .

(b) EFPY from plant startup.

(c) Standby Capsules Zand V were removed and placed in the spent fuel pool. No testing or analysis has been performed on these capsules.

(d) EOC = end-of-cycle.

II

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 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 57 EFPY.

Table 5.4 provides the Reference Temperature for Pressurized Thermal Shock (RTrrs) values for Byron Unit 1 for 57 EFPY obtained from Reference 4.

12

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.1 Byron Unit 1 Calculation of Chemistry Factors Using Surveillance Capsule Data <*>

Capsule r<h) LiRTNDT(b) FF*LiRTNoT Material Capsule FF<c> FF2 (n/cm2, E > 1.0 MeV) (OF) (OF) u 0.409 X 10 19 0.752 28.7 21.58 0.57 Intennediate X J.49 X 10 19 1.110 18.3 20.32 1.23 Shell Forging (Tangential) w 2.26 X 10 19 1.221 49.5 60.42 1.49 y 19 3.97x 10 1.355 27.8 37.66 1.83 u 0.409 X 10 19 0.752 18.6 13.99 0.57 Intennediate X J.49 X 10 19 1.110 54.6 60.63 1.23 Shell Forging (Axial) w 2.26 X 10 19 1.221 29.5 36.01 1.49 y 3.97x 10 19 1.355 I 1.7 15.85 1.83 SUM: 266.46 10.25 CF JS Forging= L(FF

  • l\RTNoT) + L(FF 2) = (266.46) + (10.25) = 26.0°F u 0.409 X 10 19 0.752 10.4 (5.2) 7.82 0.57 Byron Unit I X J.49 X 10 19 1.110 80.2(40.1) 89.06 1.23 Surveillance w 2.26x 10 19 1.221 101.2 (50.6) 123.54 1.49 Weld Material (Heat #442002) y 19 1.355 153.4 (76.7) 207.79 1.83 3.97 X 10 u 0.406x 10 19 0.750 17.4(8.7) 13.05 0.56 Byron Unit 2 w J.21 X 10 19 1.053 57.6 (28.8) 60.66 I.II Surveillance X 19 1.211 I08.4 (54.2) 131.32 1.47 Weld Material 2.18x 10 (Heat #442002) y 4.19x 10 19 1.366 117.4(58.7) 160.39 1.87 SUM : 793.63 10.13 2

CF Weld Mc1al = L(FF

  • l\R TNDT) + L(FF ) = (793 .63) + (I 0.13} = 78.3°F Notes:

a) Source document is WCAP-18371-NP (Reference 4), Table 5-1 and Table 5-3 .

b) f= fluence; LiRTNorvalues are the measured 30 ft-lb shift values taken from References 9md 11.

.6.RTNDT values for the surveillance weld data are adjusted by a ratio of2.0 to account for chemistry differences between the surveillance weld and the vessel weld. (Pre-adjusted values are listed in parentheses.)

c) FF= fluence factor= t< 0*28

  • 0*10 ' 108 f)_

13

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.2 Byron Unit 1 Reactor Vessel Material Properties (a)

Chemistry Initial Material Description Cu(%) Ni(%) Factor RT NDT (°Fib)

Closure Head Flange, Heat# 124K358VA1 -- 0.74 -- 60 Vessel Flange, Heat # 123J2 l 9V A 1 -- 0.73 -- 10 Inlet Nozzle 03-001, Heat# 1V4684-3V1320 0.12 0.82 86ld) -10 Inlet Nozzle 03-002, Heat# 1V4684-3V1320 0.12 0.82 86(d) -20 Inlet Nozzle 04-001. Heat# 1V4695 0.13 0.79 95.8(d) -20 Inlet Nozzle 04-002. Heat # 1V4695 0.12 0.78 85.7(d) -20 Outlet Nozzle O1-001. Heat # 1V4656 0.11 0.84 77(d) 0 Outlet Nozzle 01-002. Heat# 1V4656 0.11 0.84 77(d) -20 Outlet Nozzle 02-001. Heat# 2V2557 0.11 0.85 77(d) -20 Outlet Nozzle 02-002. Heat # 2V2557 0.11 0.84 77(d) -10 Nozzle Shell Forging, Heat# 1231218 0.05 0.72 31 (d) 30 Intermediate Shell Forging, Heat# SP-5933 0.04 0.74 26(d) 40 Lower Shell Forging, Heat# 5P-595 l 0.04 0.64 26(d) 10 Intermediate to Lower Shell Forging Circ. 54(d)

Weld Seam WF-336 (Heat# 442002) 0.04 0.63 78.3(e) ' -30 Nozzle Shell to Intermediate Shell Forging Circ. 4lld)

Weld Seam WF-501 (Heat# 442011) 0.03 0.67 31.2(c)(I)' 10 Byron Unit 1 Surveillance Program 27(d)

Weld Metal (Heat# 442002) 0.02 0.69 --

Byron Unit 2 Surveillance Program 27(d)

Weld Metal (Heat# 442002) 0.02 0.71 --

Braidwood Units 1 & 2 Surveillance 0.67, 41 (d)

Program Weld Metals (Heat# 442011) 0.03 0.71 Inlet Nozzle to Nozzle Shell Forging Circumferential Weld Seams WF-337 0.15 0.56 139.2(d)(g) -10 (Heat# 442002)

Outlet Nozzle to Nozzle Shell Forging Circumferential Weld Seams WF-419 0.178 0.69 168.3 -48.6(c)

(Heat# 1P5412)

Outlet Nozzle to Nozzle Shell Forging Circumferential Weld Seams WF-406 0.054 0.80 73.6(d) 10 (Heat# 504)

Notes contained on the following page:

14

BYRON- UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Notes:

a) Data taken from Reference 4.

b) The initial RT NDT values for the forgings and welds are based on measured data.

c) Generic value taken from BA W-2308 (Reference 2).

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

e) Chemistry Factor calculated per Regulatory Guide 1.99, Rev. 2, Position 2.1.

f) The Position 2.1 CF uses credible surveillance data from Braidwood in WCAP-18370-NP (Reference 12).

g) The surveillance weld material is not representative of this weld material even though the two welds share the same material heat number, flux type, and lot number. The reactor vessel beltline and surveillance welds have low weight-percent copper values due to restrictions on copper content in the beltline region; whereas, the nozzle circumferential weld seams do not have the low weight percent copper restrictions. The embrittlement behavior for these two welds would not be the same.

Therefore, surveillance weld data will not be applied to this weld.

15

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.3 Summary of Byron Unit 1 Adjusted Reference Temperature (ART) Values at l/4T and 3/4T Locations for 57 EFPY (a)

Surface Fluence 57 EFPY Reactor Vessel Material (n/cm2, E > 1.0 MeV) l/4T ART (°F) I 3/4T ART (°F) 17 Inlet Nozzle 03-001 J.33 X 10 J2.8(b)

Inlet Nozzle 03-002 J.33 X 10 17 2.8(b)

Inlet Nozzle 04-00 I 1.33 X 10 17 5.4(b)

Inlet Nozzle 04-002 J.33 X 10 17 2.7(b)

Outlet Nozzle 0 1-00 I I.OJ X 10 17 ) 7.Q(h)

Outlet Nozzle 0 1-002 I.QI X 10 17 -3.Q(h)

Outlet Nozzle 02-00 I I.OJ X 10 17 -3.Q(h)

Outlet Nozzle 02-002 I.OJ X 10 17 7.Q(h)

Nozzle Shell Forging 85.6 68.6 l.15x 10 19 Intermediate Shell Forging 3.)9x10 19 101.2 86.6

-.. Using non-credible surveillance data 3.19x 10 19 101.2 86.6 Lower Shell Forging 3.)8x)0 19 71.2 56.6 Nozzle to Intermediate Shell Forging Circ. Weld Seam (Heat# 442011) 1.)5 X 10 19 83.5 61.1

_..Using credible Braidwood Units 1.15 X )0 19 66.0 48.9 I and 2 surveillance data Intermediate to Lower Shell Forging Circ. Weld Seam (Heat# 442002) 3.07 X 10 19 89.1 65.6

-.. Using credible surveillance data 3.07 X 10 19 89.4 67.3 Inlet Nozzle to Nozzle Shell Forging Circumferential Weld Seams WF-337 1.33 X 10 17 26.9(b)

<Heat # 442002)

Outlet Nozzle to Nozzle Shell Forging Circumferential Weld Seams I.OJ X 10 17 36.6(b)

WF-419 (Heat# I P5412)

Outlet Nozzle to Nozzle Shell Forging Circumferential Weld Seams I.OJ X 10 17 26.2(b)

WF-406 (Heat# 504)

Notes contained on the following page:

16

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Notes:

(a) The source document containing detailed calculations is WCAP-18371-NP (Reference 4), Table 7-1, Table 7-3, Table 7-4 and Table 7-5.

(b) The ART values for the extended beltline materials are conservatively calculated at the surface, i.e. without attenuation of the fluence .

17

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.4 RTns Calculation for Byron Unit 1 Beltline Region Materials at End-of-Life Extension (EOLE) (57 EFPY) (a,b)

R.G. 1.99, CF Fluence FF IRTNDT(o) .&RTNoT CJ (o) u CJ~ (d) Margin RTPTs Rev.2 (OF) (n/cm2, E > 1.0 MeV) (OF)

Reactor Vessel Material {°F) (OF) (OF) (OF)

Position Inlet Nozzle 03-00 l 1.1 86 l.33x 10 17 0.133 -10 11.4 0 5.7 11.4 12.8 Inlet Nozzle 03-002 1.1 86 1.33 X 10 17 0.133 -20 11.4 0 5.7 11.4 2.8 17 Inlet Nozzle 04-001 I.I 95.8 l.33x 10 0.133 -20 12.7 0 6.4 12.7 5.4 17 Inlet Nozzle 04-002 I.I 85.7 1.33 X )0 0.133 -20 11.4 0 5.7 11.4 2.7 17 Outlet Nozzle O1-00 I I.I 77 1.01 X 10 0.110 0 8.5 0 4.3 8.5 17.0 17 Outlet Nozzle 01-002 1.1 77 I.OJ X 10 0.110 -20 8.5 0 4.3 8.5 -3.0 17 Outlet Nozzle 02-00 I 1.1 77 I.OJ X 10 0.110 -20 8.5 0 4.3 8.5 -3.0 Outlet Nozzle 02-002 I.I 77 J.0) X 10 17 0.110 -10 8.5 0 4.3 8.5 7.0 19 Nozzle Shell Forging I.I 31 J.15 X 10 1.039 30 32.2 0 16.1 32.2 94.4 19 Intennediate Shell Forging 1.1 26 3.19xl0 1.305 40 33.9 0 17.0 33.9 107.9

_,. Using non-credible 2.1 26.0 3.19xl0 19 1.305 40 33.9 0 17.0 33.9 107.9 surveillance data Lower Shell Forging I.I 26 3.18x10 19 1.304 10 33.9 0 17.0 33.9 77.8 Nozzle to Intennediate Shell 19 Forging Circ. Weld Seam 1.1 41 1.15 X 10 1.039 10 42.6 0 21.3 42.6 95 .2 (Heat # 442011)

_,. Using credible Braidwood Units 1.15 x)0 19 32.4 2.1 31.2 1.039 10 0 14.0 28.0 70.4 I and 2 surveillance data Intennediate to Lower shell Forging Circ Weld Seam 1.1 54 3.07 X 10 19 1.296 -30 70.0 0 28.0 56.0 96.0 (Heat# 442002)

_,. Using credible surveillance data 2.1 78.3 3.07x 10 19 1.296 -30 101.5 0 14.0 28.0 99.5 Inlet Nozzle to Nozzle Shell Forging Circumferential Weld Seams 1.1 139.2 l.33x 10 17 0.133 -10 18.5 0 9.2 18.5 26.9 WF-337 (Heat# 442002)

Outlet Nozzle to Nozzle Shell Forging Circumferential Weld Seams 1.1 168.3 I.OJ X 10 17 0.110 -48.6<*> 18.6 18.0<*> 28.o<*> 66.6 36.6 WF-419 (Heat# IP5412)

Outlet Nozzle to Nozzle Shell Forging Circumferential Weld Seams 1.1 73.6 I.OJ X 10 17 0.110 10 8.1 0 4.1 8.1 26.2 WF-406 (Heat # 504) 12

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 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 WCAP-18371-NP (Reference 4), Table E-1.

(c) Initial RTNDT values are based on measured data, unless noted otherwise. Hence Gu= 0°F.

(d) Per the guidance of JO CFR 50.6 l, the base metal er 6 a l 7°F for Position l.l and for Position 2.1 with non-credible surveillance data; the weld metal er 6 a 28°F for Position l.l (without surveillance data) and with credible surveillance data er A

  • 14°F for Position 2.1. However, er A need not to exceed 0.5*~RTNDT (e) The IRTNoT values are based on BAW-2308 (Reference 2). Use ofBAW-2308 as an exemption to the 10 CFR 50.61 methodology was approved in Reference
10. BA W-2308 requires the use of a,= 18°F and cr6 = 28°F.

IQ

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 6.0 References

1. WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", J.D. Andrachek, et al., May 2004.
2. AREVA Document, BAW-2308, Revision 1-A and 2-A, "Initial RTNoT of Linde 80 Weld Materials," August 2005 and March 2008.
3. LTR-SCS-19-13, Revision 0, "Byron Units 1 and 2 Low Temperature Overpressure Protection System (LTOPS) Analysis for 57 EFPY," December 10, 2019.
4. WCAP-18371-NP, Revision 0, "Byron Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," September 2019.
5. WCAP-16143-P, Revision 1, "Reactor Vessel Closure HeadNessel Flange Requirements Evaluation for Byron/Braidwood Units 1 and 2," W. Bamford, et al., October 2014.
6. 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 -

Exemption from the Requirements of 10 CFR Part 50, Appendix G (TAC Nos. MC8697, MC8698, MC8699, and MC8700)," November 22, 2006. [ADAMS Accession Number ML061890003}

7. NRC Letter from J. S. Wiebe, NRR, to B.C. Hanson, Exelon Generation Company, LLC, "Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2 - Issuance of Amendments to Utilize WCAP-16143-P, Revision 1 "Reactor Vessel Closure HeadNessel Flange Requirements Evaluation for Byron/Braidwood Units 1 and 2,"

Dated October 16, 2014 (CAC Nos. MF5033, MF5034, MF5035 and MF5036),"

October 28, 2015. [ADAMS Accession Number MLJ 5232A441}

8. WCAP-9517, "Commonwealth Edison Company, Byron Station Unit 1 Reactor Vessel Surveillance Program", J.A. Davidson, July 1979.
9. WCAP-18054-NP, Revision 1, "Analysis of Capsule Y from the Exelon Generation Byron Unit 1 Reactor Vessel Radiation Surveillance Program," Septem}?er 2018.
10. NRC Letter from J. S. Wiebe, NRR, to B.C. Hanson, Exelon Generation Company, LLC, "Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2- Issuance of Amendment Nos. 217,217,221, and 221 Regarding Reactor Coolant System Pressure and Temperature Limits Report Technical Specifications (EPID L-2019-LLA-0215)," September 18, 2020. [ADAMS Accession Number ML20163A046].
11. WCAP-18056-NP, Revision 1, "Analysis of Capsule Y from the Exelon Generation Byron Unit 2 Reactor Vessel Radiation Surveillance Program," September 2018.

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BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT

12. WCAP-18370-NP, Revision 0, "Braidwood Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," June 2019.

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