ML070240261
| ML070240261 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 01/23/2007 |
| From: | Hoots D Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Byron 2007-0013 | |
| Download: ML070240261 (26) | |
Text
n I
I~ar January 23, 2007 LTR:
BYRON 2007-0013 File:
1.10.0101 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 Byron Station, Unit I Facility Operating License No, NPF-37 NRC Docket No, STN 50-454
Subject:
Byron Station Unit 1 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)
In accordance with Technical Specification 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), we are submitting the December 2006 revision to the Unit 1 PTLR. The PTLR was revised to extend the applicability of the heatup and cooldown curves out to 32 Effective Full Power Years.
Should you have any questions concerning this report, please contact William Grundmann, Regulatory Assurance Manager, at (815) 406-2800.
Respectfully, D vid M F~1oots Site Vice President Byron Nuc~earGenerating Station
Attachment:
Byron Station Unit 1 PTLR DMH/JELIrah
ATTACHMENT Byron Station Unit 1 Reactor Coolant System (RCS)
Pressure and Temperature Limits Report (PTLR)
(December 2006)
BYRON UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
(December 2006)
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents Section Page 1.0 Introduction 1
2.0 RCS Pressure and Temperature Limits 1
2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3) 1 3.0 Low Temperature Over Pressure Protection and Boltup 7
3.1 LTOP System Setpoints (LCO 3.4.12) 7 3.2 LTOP Enable Temperature 7
3.3 Reactor Vessel Boltup Temperature (Non-Technical Specification) 7 4.0 Reactor Vessel Material Surveillance Program 10 5.0 Supplemental Data Tables 12 6.0 References 19
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°Fihr)Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 2.2 Byron Unit I Reactor Coolant System Cooldown Limitations 4
(Cooldown Rates of 0, 25, 50, and 100°F/hr)Applicable for 32 EFPY (Without Margins for instrumentation Errors) 3.1 ByTon Unit 1 Nominal PORV Setpoints for the Low Temperature 8
Overpressure Protection (LTOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)
II
BYRON - UNIT 1 PRESSURE AN!) TEMPERATURE LIMITS REPORT List of Tables Table Page 2.1 a Byron Unit 1 Heatup Data Points at 32 EFPY (Without Margins 5
for Instrumentation Errors) 2.lb Byron Unit 1 Cooldown Data Points at 32 EFPY (Without 6
Margins for instrumentation Errors) 3.1 Data Points for Byron Unit 1 Nominal PORV Setpoints for the 9
LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) 4.1 Byron Unit 1 Capsule Withdrawal Schedule 11 5.1 Byron Unit 1 Calculation of Chemistry Factors Using 13 Snrveillance Capsule Data 5.2 Byron Unit 1 Reactor Vessel Material Properties 14 5.3 Summary ofByron Unit 1 Adjusted Reference Temperatures 15 (ARTs) at 114T and 3/4T Locations for 32 EFPY 5.4 Byron Unit 1 Calculation of Adjusted Reference Temperatures 16 (ARTs) at 32 EFPY at the Limiting Reactor Vessel Material, Intermediate Shell Forging SP-5933 (Conservatively Based on Surveillance Capsule Data) 5.5 RTpr5 Calculation for Byron Unit I Beitline Region Materials at 17 EOL (32 EFPY) 5.6 RTPTS Calculation for Byron Unit I Beitline Region Materials at 18 Life Extension (48 EFPY)
In
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)
This Pressure and Temperature Limits Report (PTLR) for Byron Unit I has been prepared in accordance with the requirements of Byron TS 5.6.6 (RCS Pressure and Temperature Limits Report). Revisions to the PTLR shall be provided to the NRC after issuance.
The Technical Specifications addressed in this report are listed below:
TS-LCO 3.4.3 RCS Pressure and Temperature (PIT) Limits; and TS-LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System.
2.0 RCS Pressure and Temperature Limits This section provides the Byron Unit I Heatup and Cooldown Limitations.
The PTLR limits for Byron Unit I 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 ofASME Code Case N-588, Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessel,Section XI, Division 1, and d) Elimination ofthe 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 I PIT curves, along with the best estimate chemical compositions, fluence projections, and adjusted reference temperatures used to determine these limits. The weld metal data integration for ByTon and Braidwood Units 1 and 2 is documented in Reference 2. WCAP-16143-P, Reference 13, documents the technical basis for the elimination of the flange requirements.
2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3) 2.1.1 The RCS temperature rate of-change limits defined in Reference 7 are:
a)
A maximum heatup of 100°Fin any 1-hour period.
b) A maximum cooldown of 100°Fin any 1 -hour period, and c)
A maximum temperature change of less than or equal to 10°Fin 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 2.1.2 The RCS P/T limits for heatup, inservice hydrostatic and leak testing, and criticality are specified by Figure 2.1 and Table 2. Ia. The RCS P/T limits for cooldown are shown in Figure 2.2 and Table 2.lb. These limits aie defined in WCAP 15391. Rev.
I (Reference 7). Consistent with the methodology described in Reference 1, the RCS PIT limits for heatup and cooldown shown in Figures 2.1 and 2.2 are provided without margins for instrument error. These limits were developed using ASME Boiler and Pressure Vessel Code Section XI, Appendix G, Article G-2000, 1996 Addenda. The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in 10 CFR 50, Appendix G.
The PIT limits for core operation (except for low power physics testing) are that the reactor vessel must be at a temperature equal to orhigher than the minimum temperature required for the inservice hydrostatic test, and at least 40°Fhigher than the minimum permissible temperature in the corresponding PIT curve for heatup and cooldown.
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING LIMITING ART VALUES AT 32 EFPY:
1/41. 1O6~F 2500 2250 2000 1750 1250 750 500 250
-~
3/41. 97SF 0
50 100 150 200 250 300 350 400 Moderator Temperature (Deg. F)
Figure 2.1 Byron Unit 1 Reactor Coolant System Heatup Limitations (Heatup rates of 100°F/br)
Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 302616034
~iJnacceptable Operation Acceptable Operation 1500 I
1000 0
Criticality Limit based on inservice hydrostatic test temperature (166 F) for the service Deriod u~to 32 EFPY 450 500 550 3
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT MATERiAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING LIMITING ART VALUES AT 32 EFPY:
1141. 106°F 3/4T. 97°F 2500 2250 2000 1750 1500 1250 I
1000 (5
750 500 250 0
Figure 2.2 Byron Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown rates of0,25, 50 and 100°F/br)Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F) 4
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.la Byron Unit 1 HeatLip Data Points at 32 EFPY (Without Margins for Instrumentation Errors)
Heatup Curve 100 FHeatup Criticality Limit Leak Test Limit T(°F~ Ptn~ie~ T°F)
P(nsi~ T°F~ P(nsigl 60 0
166 0
149 2000 60 720 166 720 166 2385 65 720 166 720 70 720 166 720 75 720 166 720 80 720 166 720 85 720 166 720 90 720 166 720 95 720 166 723 100 723 166 729 105 729 166 737 110 737 166 749
15 749 166 764 20 764 166 781 125 781 170 802 130 802 175 826 135 826 180 854 140 854 185 886 145 886 190 921 150 921 195 962 155 962 200 1007 160 1007 205 1057 165 1057 210 1113 170 1113 215 1175 175 1175 220 1244 80 1244 225 1321 85 1321 230 406 90 1406 235
,499 195 1499 240 1603 200 1603 245 1718 205 718 250 1844 210 844 255 1984 215 1984 260 2138 220 2138 265 2308 225 2308
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.lb Byron Unit 1 Cooldown Data Points at 32 EFPY (Without Margins for Instrumentation Errors)
Cooklown Curves Steady State 25 F Cooldown 50 °FCooldown 100 F Cooldown T(~fl P(nsip~ T(°F~P~nsi~T( F~ P(nsi~ T(F~
P(nsi~
60 0
60 0
60 0
60 0
60 753 60 709 60 665 60 581 65 769 65 726 65 685 65 606 70 787 70 746 70 706 70 633 75 806 75 767 75 730 75 663 80 827 80 791 80 757 80 697 85 85!
85 817 85 786 85 735 90 877 90 846 90 819 90 777 95 906 95 879 95 855 95 823 100 937 100 914 100 895 100 874 105 973 105 954 105 940 105 931 110 1011 110 997 110 989 115 1054 115 1045 115 1043 120 1102 120 1099 125 1154 130 1212 135 1276 140 1347 145 1425 ISO 1512 155 1607 160 1713 165 1829 170 1958 175 2101 180 2258 185 2433 Note: For each cooldown rate, the stead)-state pressure values shall govern the temperature where no allowable pressure values are provided.
6
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 3.0 Low Temperature Overpressure Protection and Boltup This section provides the Byron Unit 1 power operated relief valve lift settings, low temperature overpressure protection (LTOP) system arming temperature, and minimum reactor vessel boltup temperature.
3.1 LTOP System Setpoints (LCO 3.4.12)
The power operated relief valves (PORVs) shall each have maximum lift settings in accordance with Figure 3.1 and Table 3.!. 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.
3.2 LTOP Enable Temperature The required enable temperature for the PORVs shall be 350°FRCS temperature. (Byron Unit I procedures governing the heatup and cooldown of the RCS require the arming of the LTOP System for RCS temperature of 350°Fand below and disarming of LTOP for RCS temperature above 3 50°F).
Note that the last LTOP PORV segment in Table 3.1 extends to 400°Fwhere the pressure setpoint is 2335 psig. This is intended to prohibit PORV lift for an inadvertent LTOP system arming at power.
3.3 Reactor Vessel Boltup Temperature (Non-Technical Specification)
The minimum boltup temperature for the Reactor Vessel Flange shall be 60°F.
Boltup is a condition in which the Reactor Vessel head is installed with tension applied to any stud, and with the RCS vented to atmosphere (Reference 7).
7
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 2500 2250 2000 -~-
1750 (0
e 1500 0
0
~ 1250 00.
~
1000 I E0z 750 595 psig 500 -L 541 psig 250 ~
350 400 450 Figure 3.1 Byron Unit 1 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertaint~)
2335 psig Unacceptable Operation PCV~456 PCV-455A 100 150 200 250 300 Auctioneered Low ACS Temperature (DEG, F) 8
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 3.1 Data Points for Byron Unit 1 Nominal PORV Setpoints for the LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)
PCV-455A (1 TY-04 1 3M)
AUCTIONEERED LOW RCS PRESSURE RCS TEMP. (DEG. F)
(PSIG) 60 541 300 400 541 2335 PCV-456 (ITY-041W)
AUCTIONEERED LOW RCS TEMP. (DEG. F) 60 300 400 RCS PRESSURE (PSIG) 595 595 2335 Note: To determine nominal lift setpoints for RCS Pressure and RCS Temperatures greater than 300°F,linearly interpolate between the 300°Fand 400°Fdata points shown above.
(Setpoints extend to 400°Fto prevent PORV liftoff from an inadvertent LTOP system arming while at power.)
9
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 4.0 Reactor Vessel Material Surveillance Program The pressure vessel material surveillance program (Reference 14) is in compliance with Appendix H to 10 CFR 50, Reactor Vessel Radiation Surveillance Program. The material test requirements and the acceptance standards utilize the reference nil-ductility temperature.
RT~DT,which is determined in accordance with ASME Boiler and Pressure Vessel Code,Section III, NB-2331. The empirical relationship between RTNDT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, Protection Against Non-Ductile Failure, to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E185-82.
The third and final reactor vessel material irradiation surveillance specimens 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. Other capsules will be removed to avoid excessive fluence accumulation should they be needed to support life extension. The removal schedule is provided in Table 4.1. The time of specimen withdrawal may be modified to coincide with those refueling outages or reactor shutdowns most closely approaching the withdrawal schedule.
10
BYRON - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.1 Byron Unit 1 Capsule Withdrawal Schedule Capsule Vessel Location (Degrees)
Capsule Lead Factor Renioval Time~
(EFPY)
Estimated Capsule Fluence (n/cm2)
U 58.5° 4.22 1.15 (Removed) 4.04x 10s X
238.5° 4.27 5.64 (Removed) 1.57 x 1019 W
121.5° 4.20 9.24 (Removed) 2.43 x Z
301.5° 4.20 B1R12~
(d)
V 61.0° 3.97 B1R12 (C)
(d)
Y 241.0° 3.97 Standby (c)
.~
a)
Effective Full Power Years (EPPY) from plant startup.
b)
Maximum end of license (32 EFPY) inner vessel wall fluence is estimated to be 2.02 x 1 0~nlcm2.
c)
Standby capsule to be used for future license renewal (Derived from WCAP 15123, Rev. 1) (Reference 15).
d)
Capsule removed and is stored in the spent fuel pool. Capsule has not been analyzed and therefore capsule fluence has not been estimated.
11
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 5.0 Supplemental Data Tables The following tables provide supplemental information on reactor vessel material properties and are provided to be consistent with Generic Letter 96-03. Some of the material property values shown were used as inputs to the P/T limits.
Table 5.! 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 (ARTs) at the 1/4T and 3/4T locations for 32 EFPY.
Table 5.4 shows the calculation of ARTs at 32 EFPY for the limiting Byron Unit 1 reactor vessel material (Intermediate Shell Forging 5P-5933).
Table 5.5 provides RTp1s values for Byron Unit 1 for 32 EFPY obtained from Reference 4.
Table 5.6 provides RTprs values for Byron Unit 1 for 48 EFPY obtained from Reference 4.
12
n I
I~ar January 23, 2007 LTR:
BYRON 2007-0013 File:
1.10.0101 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 Byron Station, Unit I Facility Operating License No, NPF-37 NRC Docket No, STN 50-454
Subject:
Byron Station Unit 1 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)
In accordance with Technical Specification 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), we are submitting the December 2006 revision to the Unit 1 PTLR. The PTLR was revised to extend the applicability of the heatup and cooldown curves out to 32 Effective Full Power Years.
Should you have any questions concerning this report, please contact William Grundmann, Regulatory Assurance Manager, at (815) 406-2800.
Respectfully, D vid M F~1oots Site Vice President Byron Nuc~earGenerating Station
Attachment:
Byron Station Unit 1 PTLR DMH/JELIrah
ATTACHMENT Byron Station Unit 1 Reactor Coolant System (RCS)
Pressure and Temperature Limits Report (PTLR)
(December 2006)
BYRON UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
(December 2006)
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents Section Page 1.0 Introduction 1
2.0 RCS Pressure and Temperature Limits 1
2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3) 1 3.0 Low Temperature Over Pressure Protection and Boltup 7
3.1 LTOP System Setpoints (LCO 3.4.12) 7 3.2 LTOP Enable Temperature 7
3.3 Reactor Vessel Boltup Temperature (Non-Technical Specification) 7 4.0 Reactor Vessel Material Surveillance Program 10 5.0 Supplemental Data Tables 12 6.0 References 19
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°Fihr)Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 2.2 Byron Unit I Reactor Coolant System Cooldown Limitations 4
(Cooldown Rates of 0, 25, 50, and 100°F/hr)Applicable for 32 EFPY (Without Margins for instrumentation Errors) 3.1 ByTon Unit 1 Nominal PORV Setpoints for the Low Temperature 8
Overpressure Protection (LTOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)
II
BYRON - UNIT 1 PRESSURE AN!) TEMPERATURE LIMITS REPORT List of Tables Table Page 2.1 a Byron Unit 1 Heatup Data Points at 32 EFPY (Without Margins 5
for Instrumentation Errors) 2.lb Byron Unit 1 Cooldown Data Points at 32 EFPY (Without 6
Margins for instrumentation Errors) 3.1 Data Points for Byron Unit 1 Nominal PORV Setpoints for the 9
LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) 4.1 Byron Unit 1 Capsule Withdrawal Schedule 11 5.1 Byron Unit 1 Calculation of Chemistry Factors Using 13 Snrveillance Capsule Data 5.2 Byron Unit 1 Reactor Vessel Material Properties 14 5.3 Summary ofByron Unit 1 Adjusted Reference Temperatures 15 (ARTs) at 114T and 3/4T Locations for 32 EFPY 5.4 Byron Unit 1 Calculation of Adjusted Reference Temperatures 16 (ARTs) at 32 EFPY at the Limiting Reactor Vessel Material, Intermediate Shell Forging SP-5933 (Conservatively Based on Surveillance Capsule Data) 5.5 RTpr5 Calculation for Byron Unit I Beitline Region Materials at 17 EOL (32 EFPY) 5.6 RTPTS Calculation for Byron Unit I Beitline Region Materials at 18 Life Extension (48 EFPY)
In
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)
This Pressure and Temperature Limits Report (PTLR) for Byron Unit I has been prepared in accordance with the requirements of Byron TS 5.6.6 (RCS Pressure and Temperature Limits Report). Revisions to the PTLR shall be provided to the NRC after issuance.
The Technical Specifications addressed in this report are listed below:
TS-LCO 3.4.3 RCS Pressure and Temperature (PIT) Limits; and TS-LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System.
2.0 RCS Pressure and Temperature Limits This section provides the Byron Unit I Heatup and Cooldown Limitations.
The PTLR limits for Byron Unit I 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 ofASME Code Case N-588, Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessel,Section XI, Division 1, and d) Elimination ofthe 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 I PIT curves, along with the best estimate chemical compositions, fluence projections, and adjusted reference temperatures used to determine these limits. The weld metal data integration for ByTon and Braidwood Units 1 and 2 is documented in Reference 2. WCAP-16143-P, Reference 13, documents the technical basis for the elimination of the flange requirements.
2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3) 2.1.1 The RCS temperature rate of-change limits defined in Reference 7 are:
a)
A maximum heatup of 100°Fin any 1-hour period.
b) A maximum cooldown of 100°Fin any 1 -hour period, and c)
A maximum temperature change of less than or equal to 10°Fin 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 2.1.2 The RCS P/T limits for heatup, inservice hydrostatic and leak testing, and criticality are specified by Figure 2.1 and Table 2. Ia. The RCS P/T limits for cooldown are shown in Figure 2.2 and Table 2.lb. These limits aie defined in WCAP 15391. Rev.
I (Reference 7). Consistent with the methodology described in Reference 1, the RCS PIT limits for heatup and cooldown shown in Figures 2.1 and 2.2 are provided without margins for instrument error. These limits were developed using ASME Boiler and Pressure Vessel Code Section XI, Appendix G, Article G-2000, 1996 Addenda. The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in 10 CFR 50, Appendix G.
The PIT limits for core operation (except for low power physics testing) are that the reactor vessel must be at a temperature equal to orhigher than the minimum temperature required for the inservice hydrostatic test, and at least 40°Fhigher than the minimum permissible temperature in the corresponding PIT curve for heatup and cooldown.
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING LIMITING ART VALUES AT 32 EFPY:
1/41. 1O6~F 2500 2250 2000 1750 1250 750 500 250
-~
3/41. 97SF 0
50 100 150 200 250 300 350 400 Moderator Temperature (Deg. F)
Figure 2.1 Byron Unit 1 Reactor Coolant System Heatup Limitations (Heatup rates of 100°F/br)
Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 302616034
~iJnacceptable Operation Acceptable Operation 1500 I
1000 0
Criticality Limit based on inservice hydrostatic test temperature (166 F) for the service Deriod u~to 32 EFPY 450 500 550 3
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT MATERiAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING LIMITING ART VALUES AT 32 EFPY:
1141. 106°F 3/4T. 97°F 2500 2250 2000 1750 1500 1250 I
1000 (5
750 500 250 0
Figure 2.2 Byron Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown rates of0,25, 50 and 100°F/br)Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F) 4
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.la Byron Unit 1 HeatLip Data Points at 32 EFPY (Without Margins for Instrumentation Errors)
Heatup Curve 100 FHeatup Criticality Limit Leak Test Limit T(°F~ Ptn~ie~ T°F)
P(nsi~ T°F~ P(nsigl 60 0
166 0
149 2000 60 720 166 720 166 2385 65 720 166 720 70 720 166 720 75 720 166 720 80 720 166 720 85 720 166 720 90 720 166 720 95 720 166 723 100 723 166 729 105 729 166 737 110 737 166 749
15 749 166 764 20 764 166 781 125 781 170 802 130 802 175 826 135 826 180 854 140 854 185 886 145 886 190 921 150 921 195 962 155 962 200 1007 160 1007 205 1057 165 1057 210 1113 170 1113 215 1175 175 1175 220 1244 80 1244 225 1321 85 1321 230 406 90 1406 235
,499 195 1499 240 1603 200 1603 245 1718 205 718 250 1844 210 844 255 1984 215 1984 260 2138 220 2138 265 2308 225 2308
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.lb Byron Unit 1 Cooldown Data Points at 32 EFPY (Without Margins for Instrumentation Errors)
Cooklown Curves Steady State 25 F Cooldown 50 °FCooldown 100 F Cooldown T(~fl P(nsip~ T(°F~P~nsi~T( F~ P(nsi~ T(F~
P(nsi~
60 0
60 0
60 0
60 0
60 753 60 709 60 665 60 581 65 769 65 726 65 685 65 606 70 787 70 746 70 706 70 633 75 806 75 767 75 730 75 663 80 827 80 791 80 757 80 697 85 85!
85 817 85 786 85 735 90 877 90 846 90 819 90 777 95 906 95 879 95 855 95 823 100 937 100 914 100 895 100 874 105 973 105 954 105 940 105 931 110 1011 110 997 110 989 115 1054 115 1045 115 1043 120 1102 120 1099 125 1154 130 1212 135 1276 140 1347 145 1425 ISO 1512 155 1607 160 1713 165 1829 170 1958 175 2101 180 2258 185 2433 Note: For each cooldown rate, the stead)-state pressure values shall govern the temperature where no allowable pressure values are provided.
6
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 3.0 Low Temperature Overpressure Protection and Boltup This section provides the Byron Unit 1 power operated relief valve lift settings, low temperature overpressure protection (LTOP) system arming temperature, and minimum reactor vessel boltup temperature.
3.1 LTOP System Setpoints (LCO 3.4.12)
The power operated relief valves (PORVs) shall each have maximum lift settings in accordance with Figure 3.1 and Table 3.!. 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.
3.2 LTOP Enable Temperature The required enable temperature for the PORVs shall be 350°FRCS temperature. (Byron Unit I procedures governing the heatup and cooldown of the RCS require the arming of the LTOP System for RCS temperature of 350°Fand below and disarming of LTOP for RCS temperature above 3 50°F).
Note that the last LTOP PORV segment in Table 3.1 extends to 400°Fwhere the pressure setpoint is 2335 psig. This is intended to prohibit PORV lift for an inadvertent LTOP system arming at power.
3.3 Reactor Vessel Boltup Temperature (Non-Technical Specification)
The minimum boltup temperature for the Reactor Vessel Flange shall be 60°F.
Boltup is a condition in which the Reactor Vessel head is installed with tension applied to any stud, and with the RCS vented to atmosphere (Reference 7).
7
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 2500 2250 2000 -~-
1750 (0
e 1500 0
0
~ 1250 00.
~
1000 I E0z 750 595 psig 500 -L 541 psig 250 ~
350 400 450 Figure 3.1 Byron Unit 1 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertaint~)
2335 psig Unacceptable Operation PCV~456 PCV-455A 100 150 200 250 300 Auctioneered Low ACS Temperature (DEG, F) 8
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 3.1 Data Points for Byron Unit 1 Nominal PORV Setpoints for the LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)
PCV-455A (1 TY-04 1 3M)
AUCTIONEERED LOW RCS PRESSURE RCS TEMP. (DEG. F)
(PSIG) 60 541 300 400 541 2335 PCV-456 (ITY-041W)
AUCTIONEERED LOW RCS TEMP. (DEG. F) 60 300 400 RCS PRESSURE (PSIG) 595 595 2335 Note: To determine nominal lift setpoints for RCS Pressure and RCS Temperatures greater than 300°F,linearly interpolate between the 300°Fand 400°Fdata points shown above.
(Setpoints extend to 400°Fto prevent PORV liftoff from an inadvertent LTOP system arming while at power.)
9
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 4.0 Reactor Vessel Material Surveillance Program The pressure vessel material surveillance program (Reference 14) is in compliance with Appendix H to 10 CFR 50, Reactor Vessel Radiation Surveillance Program. The material test requirements and the acceptance standards utilize the reference nil-ductility temperature.
RT~DT,which is determined in accordance with ASME Boiler and Pressure Vessel Code,Section III, NB-2331. The empirical relationship between RTNDT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, Protection Against Non-Ductile Failure, to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E185-82.
The third and final reactor vessel material irradiation surveillance specimens 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. Other capsules will be removed to avoid excessive fluence accumulation should they be needed to support life extension. The removal schedule is provided in Table 4.1. The time of specimen withdrawal may be modified to coincide with those refueling outages or reactor shutdowns most closely approaching the withdrawal schedule.
10
BYRON - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.1 Byron Unit 1 Capsule Withdrawal Schedule Capsule Vessel Location (Degrees)
Capsule Lead Factor Renioval Time~
(EFPY)
Estimated Capsule Fluence (n/cm2)
U 58.5° 4.22 1.15 (Removed) 4.04x 10s X
238.5° 4.27 5.64 (Removed) 1.57 x 1019 W
121.5° 4.20 9.24 (Removed) 2.43 x Z
301.5° 4.20 B1R12~
(d)
V 61.0° 3.97 B1R12 (C)
(d)
Y 241.0° 3.97 Standby (c)
.~
a)
Effective Full Power Years (EPPY) from plant startup.
b)
Maximum end of license (32 EFPY) inner vessel wall fluence is estimated to be 2.02 x 1 0~nlcm2.
c)
Standby capsule to be used for future license renewal (Derived from WCAP 15123, Rev. 1) (Reference 15).
d)
Capsule removed and is stored in the spent fuel pool. Capsule has not been analyzed and therefore capsule fluence has not been estimated.
11
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 5.0 Supplemental Data Tables The following tables provide supplemental information on reactor vessel material properties and are provided to be consistent with Generic Letter 96-03. Some of the material property values shown were used as inputs to the P/T limits.
Table 5.! 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 (ARTs) at the 1/4T and 3/4T locations for 32 EFPY.
Table 5.4 shows the calculation of ARTs at 32 EFPY for the limiting Byron Unit 1 reactor vessel material (Intermediate Shell Forging 5P-5933).
Table 5.5 provides RTp1s values for Byron Unit 1 for 32 EFPY obtained from Reference 4.
Table 5.6 provides RTprs values for Byron Unit 1 for 48 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 La)
Material Capsule Capsule f~
FF(C)
L~RT~DT~
d)
FF*ART~DT FF2 Inter. Shell Forging5P-5933 (Tangential)
U 4.O4~l0~
0.748 28.55 21.36 0.560 X
l.57xl09 1.124 9.82 11.04 1.263 W
2.43xl09 1.239 49.20 60.96 1.535 Inter. Shell ForgingSP-5933 (Axial)
U 4.04xl0~
0.748 18.52 13.85 0.560 X
l.57x109 1.124 53.03 59.61 1.263 W
2.43x109 1.239 29.34 36.35 1.535 Sum:
203.17 6.716 CFFo~gjng= ~
~ARTNDT) ÷ ~( ~2)
(203.17) ÷ (6.7 16) 30.3°F Byron I Weld Metal WF-336 (Heat#442002)
U 4.04x108 0.749 11.22 (5.61) ~
8.40 0.561 X
1.S7xlO9 1.125 80.22(40.1l)~
90.25 1.266 W
2.43x109 1.239 102.68 (51.34) ~
127.22 1.535 Byron 2 Weld Metal WF-447 (Heat#442002)
U 4.OSx 1018 0.749 16.88 (8.44) Le) 12.64 0.56 1 W
1.27 xlO9 1.067 57.76 (28.88) ~
61.63 l.138 X
2.30x i~~
1.225 108.02 (54.0l)~
132.32 1.500 SUM:
- ART\\Dr) +/- ~( FF2)
(432.46) ÷ (6.561) = 65.9°F a)
Reference 7, Table 4-9 b) f = Calculated fluence, ( x I ~
n/cm~,E> 1.0 MeV) c)
FE = fluence factor = ~
0.1 log ~
d)
ART~01values are the measured 30 ft-lb shift values e~ Adjusted ART~,)lper Ratio Procedure of Regulatory Guide 1.99. Rev. 2. Ratio 2.0. See Table 4-9 of WCAP 15391, Rev.
I (Reference 7).
13
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.2 Byron Unit 1 Reactor Vessel 1\\Iaterial Properties (a)
Material Description Cu (%)
Ni (%)
Initial RT NDT oFb Closure Head Flange 124K358VA1 0.74 60 Vessel Flange 123J219VA1 0.73 10 Nozzle Shell Forging 123J218 0.05 0.72 30 Intermediate Shell Forging 5P-5933 0.04 0.74 40 Lower Shell Forging 5P-5951 0.04 0.64 10 Intermediate to Lower Shell Forging Circ.
Weld Seam WF-336 (Heat # 442002) 0.04 0.63
-30 Nozzle Shell to Intermediate Shell Forging Circ. Weld Seam WF-501 (Heat #442011) 0.03 0.67 10 Byron Unit 1 Surveillance Program Weld Metal (Heat # 442002) 0.02 0.69 Byron Unit 2 Surveillance Program Weld_Metal_(Heat_# 442002) 0.02 0.71 Braidwood Units 1 & 2 Surveillance Program Weld Metals (Heat #442011) 0.03 0.67, 0.71 a)
Reference 7.
b)
The initial RTNDT values for the plates and welds are based on measured data.
14
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5,3 Summary of Byron Unit 1 Adjusted Reference Temperatures (ARTs) at 1/4T and 3/4T Locations for 32 EFPY~
Material Description 32 EFPY I/4T ART(°F) 3/4T ART(°F)
Intermediate Shell Forging 5P-5933
- Using Surveillance Data(b) 95 80 106~
97(c)
Lower Shell Forging 5P-5951 65 50 Circumferential Weld WF-336 82 52
- Using Credible Surveillance Data~
67 48 Circumferential Weld WF-501
- Using Credible Surveillance Data from Braidwood 1 and 2 69 49 47 34 Nozzle Shell Forging 123J21 8 75 59 (a)
Fluence, f, is based upon ~U[f (E> 1.0MeV) = 2.02x 1010 at 32 EFPY (Reference 7>.
(b)
Calculated using a chemistry factor based on Regulatory Guide (RG) 1.99, Revision 2. Position 2 along with a full margin since it was determined that this data was not credible and the Table chemistry factor was non conservative (Reference 8).
(c)
These ART values were used to generate the Byron Unit 1 32 EFPY heatup and cooldown curves (Reference 7).
(d)
Calculated using the chemistry factor from the Byron Unit I and 2 integrated surveillance data as reported in WCAP-l539l (Reference 7).
15
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.4 Byron Unit 1 Calculation of Adjusted Reference Temperatures (ARTs) at 32 EFPY at the Limiting Reactor Vessel Material, Intermediate Shell Forging SP-5933 (Conservatively Based on Surveillance Capsule Data)~
Parameter Values Operating Time 32 EFPY Location~
Chemistry Factor, CF (°F) 30.3 30.3 Fluence(f), n/cm l.21x109 4.37x108 Fluence Factor, FF 1.053 0.770 LXRTNDT= CFxFF (°F) 31.9 23,3 Initial RT NDT,, I (°F) 40 40 Margin. M (°F) 34 34 ART= I+(CF*FF)+M,°F per_RG_1.99,_Revision 2 106 97 a)
WCAP 15123 (Reference 15) b)
The Byron Unit 1 reactor vessel wall thickness is 8.5 inches at the beltline region.
c)
Fluence, f. is based upon f~LLf,(E>l.0Mev) 2.O2x 10~at 32 EFPY (Reference 7).
16
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.5 RTPTS Calculation for Byron Unit 1 Beitline Region Materials at EOL (32 EFPY) (a)
Material Fluence (10~n/cm2, E>1.O MeV)
(°F)
L~RTPTs~
(°F)
Margin (°F)
RT\\DT)c)1°
(°F)
RTPTS(d)
(°F)
Intermediate Shell ForgingSP-5933 Intermediate Shell Forging 5P-5933 using S/C Data~° 1.95 ~ IO~
1.95 ~ 10~
1.18 1.18 26.0 30.3 30.7 35.8 30.7 34 40 30 101 110 Lower shell Forging 5P-5951 1.95 ~ i~~
1.18 26.0 30.7 30.7 10 71 Inter. To Lower Shell Circ. Weld Metal WF-336 (442002)
Inter. To Lower Shell Circ. Weld Metal (442002) using S/C Data~° 1.95 ~ 10 1.95 ~ 1010 1.18 1.18 54.0 65.9 63.7 67.6 56 28
-30
-30 90 66 Nozzle Shell Forging 173J2l8 5.83 ~ lO~
0.849 31.0 26.3 26.3 30 83 Nozzle Shell to Inter. Shell Circ.
Weld Metal WF-50l (442011) 5.83 ~ 1O~
0.849 41.0 34.8 34.8 10 80 Nozzle Shell to inter. Shell Circ.
Weld Metal (442011) using S/C Data 5.83 x 10s 0.849 16.7 14,2 14.2 10 38 (a)
Limiting RTN-s is significantly less than the PTS Screening Criteria of 270 °F.
(b) ARTprs
= CF*FF (c)
Initial RT~Dlvalues are measured values.
(d) RTvrs = RT\\,)T)tJ) + ART,~5+ Margin ( F)
Ic)
Surveillance data is considered not credible, however, since the chemistry factor tCF) from the Reg. Guide Tables (Pos. 1.1) is lower (i.e. CF via Pos. 2.1 > CF via Pos. 1.1), then the Pos. 2.1 CF is used to determine PTS with a full o~margin term. i.e. 17 T.
(f)
Based on Byron Unit 1 and 2 integrated surveillance data chemistry factor from WCAP-15 178 (Reference 9).
17
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.6 RTpTS Calculation for Byron Unit 1 Beltline Region Materials at Life Extension (48 EFPY) ~
Material Fluence (IO9n/cm2, E>1.0 MeV)
(°F)
~RTp1s~
(°F)
Margin
(°F)
R t~DT)~)
(°F)
RT~15~
(°F)
Intermediate Shell Forging 5P-5933 Intermediate Shell Forging SP-5933 using S/C Data~
2.91 ~ l0~
2.91 ~ 10 1.28 1.28 26.0 30.3 33,3 38.8 33,3 34 40 40 107 113 Lower shell Forging 5P-595l 2.91
.~ 10 1.28 26.0 33,3 33,3 10 77 Inter. To Lower Shell Circ. Weld Metal WF-336 (442002)
Inter. To Lower Shell Circ. Weld Metal (442002) using S/C Data~
2.91 ~ ~
2.91 ~ 10 1.28 1.28 54.0 65.9 69.1 84.4 56 28
-30
-30 95 82 Nozzle Shell Forging l23J2l8 8.70x l0~
0.961 31.0 29.8 29.8 30 90 Nozzle Shell to inter. Shell Circ.
Weld Metal WF-501 (442011)
Nozzle Shell to Inter. Shell Circ.
Weld Metal (44201l)using S/C Data 8.70 ~ lO~
8.70x 1018 0.961 0.961 4l.0 16.7 39,4 16.0 39,4 16.0 10 10 89 42 (a) The fluence for 48 EFPY (Reference 4) did not incorporate the 5% increase. However, this fluence value is greater than the end-of-life fluence (32 EFPY).
(b) Limiting RT,rrs is significantly less than the PTS Screening Criteria of 270 °F.
(c) ~\\RT~5 = CF*FF (d) Initial RT~,values are measured values.
(C) RT~15 RT~DT(t~)
+ ~RT~15+ Margin (°F)
(f) Sur~ei1lancedata is considered not credible, however, since the chemistry factor (CF) from the Reg. Guide Tables (Pos.
I. 1) is lower (i.e. CF via Pos. 2.1 > CF via Pos. 1.1), then the Pos. 2.1 CF is used to determine PTS with a full o~margin term. i.e. 17 F.
(g) Based on Byron Unit I and 2 integrated surveillance data chemistry factor from WCAP-15178 (Reference 9).
18
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 6.0 References 1.
WCAP-14040-NP-A, Revision 2, Methodology Used to Develop Cold Ovcrprcssure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, J.D. Andrachek, et al., January 1996.
2.
WCAP-14824, Revision 2, Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld Metal Integration for Byron & Braidwood, November 1997 with Westinghouse errata letters CAE-97-220, dated November 26, 1997 and CAE 23 1/CCE-97-3 14 and CAE-97-233/CCE-97 316, dated January 6, 1998.
3.
Westinghouse Letter to Commonwealth Edison Company, CAE-06-90/CCE-06-86, Transmittal of Byron and Braidwood Units 1 and 2 Revision I LTOPS Setpoints Analysis Reports for 22 and 32 EFPY (LTR-SCS-03-87, Revision 1 Attachment A) (LTR-SCS-03-87, Revision 1 Attachment B), August 28, 2006.
4.
WCAP-15125, Evaluation of Pressurized Thermal Shock for Byron Unit 1, Revision 0, T.
J. Laubham et a!., November 1998.
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 0. D. Kingsley, Commonwealth Edison Co., Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, Acceptance for Referencing of Pressure Temperature Limits Report (TAC Numbers M98799, M98800, M98801, and M98802), January 21, 1998.
7.
WCAP-15391, Revision I, Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation, T. J. Laubham, et al., November 2003.
8.
WCAP-15183, Revision 0, Commonwealth Edison Company Byron Unit 1 Surveillance Program Credibility Evaluation, T. J. Laubham, et al., June 1999.
9.
WCAP-15178, Revision 0, Byron Unit 2 Heatup and Cooldown Limit Curves for Normal Operation, T. J. Laubham, et al., June 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 I and 2, and Braidwood Stations, Units 1 and 2, dated August 8, 2001.
19
BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 12.
NRC Letter from R. F. Kuntz, NRR, to C. M. Crane, Exelon Generation Company, LLC, Byron Station, Unit Nos. 1 and 2, and Braidwood Station, Unit Nos. 1 and 2 - Issuance of Amendments Re: Reactor Coolant System Pressure and Temperature Limits Report (TAC Nos. MC8693, MC8694. MC8695, and MC8696j, November 27. 2006.
- 13. WCAP-16143-P, Revision 0, Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units I and 2, W. Bamford, et al., November 2003.
- 14. WCAP-9517, Commonwealth Edison Company, Byron Station Unit I Reactor Vessel Surveillance Program, J.A. Davidson, July 1979.
- 15. WCAP-15123, Revision 1, Analysis of Capsule W from Common Wealth Edison Company Byron Unit I Reactor Vessel Radiation Surveillance Program, T.J. Laubham, et al, January 1999.
20