ML070680370
| ML070680370 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 03/08/2007 |
| From: | Coutu T Exelon Generation Co, Exelon Nuclear |
| To: | Document Control Desk, NRC/NRR/ADRO |
| References | |
| BW070019, TAC MC8693, TAC MC8694, TAC MC8695, TAC MC8696 | |
| Download: ML070680370 (52) | |
Text
___________Exd_n E~hn Generation Conpany, LC N uci ear~raiawood St~t on c March 8, 2007 SWO7001 9 U.S.Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos.NPF-72 and NPF-77 NRC Docket Nos.STN 50-456 and 50-457
Subject:
Pressure and Temperature Limits Reports (PTLRs), Revision 4, Braidwood Station, Units 1 and 2
References:
(1)Letter from K.FL Jury (Exelon Generation Company, LLC)to U.S.NRC,"License Amendment Request Regarding Reactor Coolant System Pressure and Temperature Limits Report and Request for Exemption from 10 CFR 50.60,"Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation,"
dated October 3, 2005 (2)Letter from R.F.Kuntz (U.S.NRC)to C.M.Crane,"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, MC8696),"dated November 27, 2006 The purpose of this letter is to transmit the Pressure and Temperature Limits Reports (PTLRs)for Braidwood Station, Units 1 and 2 in accordance with Technical Specification (TS)5.6.6,"Reactor Coolant System (RCS)Pressure and Temperature Limits Report (PTLR)."The Braidwood Unit 1 and Unit 2 PTLRs were recently revised to extend the applicability of the heatup and cooldown curves to 32 effective full power years (EFPY).The methodology for developing the revised PTLRs is consistent with a recently approved method added to Braidwood IS 5.6.6 (Reference 2).
U.S.Nuclear Regulatory Commission Page 2 March 8, 2007 Please direct any questions you may have regarding this matter to Mr.Dale Ambler, Regulatory Assurance Manager, at (815)417-2800.Thomas Coutu Site Vice President Braidwood Station Attachments:
1.Braidwood Unit 1 Pressure and Temperature Limits Report (PTLR), Revision 4 2.Braidwood Unit 2 Pressure and Temperature Limits Report (PTLR), Revision 4 ATTACHMENT 1 Braidwood Unit 1 Pressure and Temperature Limits Report (PTLR), Revision 4 BRAID WOOD UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)Revision 4 BRAID WOOD-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 Boliup 7 3.1 LTOP System Setpoints (LCO 3.4.12)7 3.2 LTOP Enable Temperature 7 3.3 Reactor Vessel Boliup Temperature (Non-Technical Specification) 7 4.0 Reactor Vessel Material Surveillance Program 10 5.0 Supplemental Data Tables 12 6.0 References 19 BRAID WOOl)-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT List of Figures Figure Page 2.1 Braidwood Unit I Reactor Coolant System Heatup Limitations (Heatup 3 Rate of l00°F/hr)Applicable for 32 EFPY (Without Margins for Instrumentation Errors)2.2 Braidwood Unit 1 Reactor Coolant System Cooldown Limitations 4 (Cooldown Rates of 0, 25, 50, and 100°F/hr)Applicable for 32 EFPY (Without Margins for Instrumentation Errors)3.1 Braidwood Unit 1 Nominal PORV Setpoints for the Low Temperature 8 Overpressure Protection (LTOP)System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) ft BRAID WOOD-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT List of Tables Table Page 2.la Braidwood Unit I Heatup Data Points at 32 EFPY (Without 5 Margins for Instrumentation Errors)2.lb Braidwood Unit 1 Cooldown Data Points at 32 EFPY (Without 6 Margins for Instrumentation Effors)3.1 Data Points for Braidwood Unit 1 Nominal PORV Setpoints for 9 the LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)
4.1 Braidwood
Unit 1 Capsule Withdrawal Schedule 11 5.1 Braidwood Unit 1 Calculation of Chemistry Factors Using 13 Surveillance Capsule Data 5,2 Braidwood Unit 1 Reactor Vessel Material Properties 14 5.3 Summary of Braidwood Unit I Adjusted Reference Temperatures 15 (ARTs)at 1/4T and 3/4T Locations for 32 EFPY 5.4 Braidwood Unit 1 Calculation of Adjusted Reference 16 Temperatures (ARTs)at 32 EFPY at the Limiting Reactor Vessel Material, Nozzle Shell Forging 5P-7016 5.5 RTprs Calculation for Braidwood Unit 1 Beltline Region 17 Materials at EOL (32 EFPY)5.6 RTp~5 Calculation for Braidwood Unit 1 Beltline Region 18 Materials at Life Extension (48 EFPY)
BRAID WOOD-UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 Introduction This Pressure and Temperature Limits Report (PTLR)for Braidwood Unit I has been prepared in accordance with the requirements of Braidwood Technical Specification (TS)5.6.6,"Reactor Coolant System (RCS)Pressure and Temperature Limits Report (PTLR)".Revisions to the PTLR shall be provided to the NRC after issuance.The Technical Specifications (TS)addressed in this report are listed below: LCO 3.4.3 RCS Pressure and Temperature (PiT)Limits;and LCO 3.4.12 Low Temperature Overpressure Protection (LTOP)System.2.0 RCS Pressure and Temperature Limits The PTLR limits for Braidwood Unit 1 were developed using a methodology specified in the Technical Specifications.
The methodology listed in WCAP-14040-NP-A, Revision 2 (Reference 1)was used with the following exceptions:
a)Optional use of ASME Code Section XI, Appendix G, Article G-2000, 1996 Addenda, b)Use of ASME Code Case N-640,"Alternative Reference Fracture Toughness for Development of P-T Limit Curves,Section XI, Division 1", c)Use of ASME Code Case N-588,"Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessel,Section XI, Division 1", and d)Elimination of the flange requirements documented in WCAP-16143-~P.
These exceptions to the methodology in WCAP 14040-NP-A, Revision 2 have been reviewed and accepted by the NRC in References 2, 8, 9 and 10.WCAP 15364, Revision 2 (Reference 11), provides the basis for the Braidwood Unit 1 P/T curves, along with the best estimate chemical compositions, fluence projections and adjusted reference temperatures used to determine these limits.WCAP-I 6143-P.Reference 12.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 WCAP-15364, Revision 2 (Reference Il)are: a.A maximum heatup of 100°F in any i-hour period, b.A maximum cooldown of 100°F in any 1-hour period, and BRAID WOOD-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 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.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.ia.The RCS P/T limits for cooldown are shown in Figure 2.2 and Table 2.lb.These limits are defined in WCAP-15364, Revision 2 (Reference II).Consistent with the methodology described in Reference 1 and exceptions noted in Section 2.0, 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 XL Appendix G, Article G2000, 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.
___________Exd_n E~hn Generation Conpany, LC N uci ear~raiawood St~t on c March 8, 2007 SWO7001 9 U.S.Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos.NPF-72 and NPF-77 NRC Docket Nos.STN 50-456 and 50-457
Subject:
Pressure and Temperature Limits Reports (PTLRs), Revision 4, Braidwood Station, Units 1 and 2
References:
(1)Letter from K.FL Jury (Exelon Generation Company, LLC)to U.S.NRC,"License Amendment Request Regarding Reactor Coolant System Pressure and Temperature Limits Report and Request for Exemption from 10 CFR 50.60,"Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation,"
dated October 3, 2005 (2)Letter from R.F.Kuntz (U.S.NRC)to C.M.Crane,"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, MC8696),"dated November 27, 2006 The purpose of this letter is to transmit the Pressure and Temperature Limits Reports (PTLRs)for Braidwood Station, Units 1 and 2 in accordance with Technical Specification (TS)5.6.6,"Reactor Coolant System (RCS)Pressure and Temperature Limits Report (PTLR)."The Braidwood Unit 1 and Unit 2 PTLRs were recently revised to extend the applicability of the heatup and cooldown curves to 32 effective full power years (EFPY).The methodology for developing the revised PTLRs is consistent with a recently approved method added to Braidwood IS 5.6.6 (Reference 2).
U.S.Nuclear Regulatory Commission Page 2 March 8, 2007 Please direct any questions you may have regarding this matter to Mr.Dale Ambler, Regulatory Assurance Manager, at (815)417-2800.Thomas Coutu Site Vice President Braidwood Station Attachments:
1.Braidwood Unit 1 Pressure and Temperature Limits Report (PTLR), Revision 4 2.Braidwood Unit 2 Pressure and Temperature Limits Report (PTLR), Revision 4 ATTACHMENT 1 Braidwood Unit 1 Pressure and Temperature Limits Report (PTLR), Revision 4 BRAID WOOD UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)Revision 4 BRAID WOOD-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 Boliup 7 3.1 LTOP System Setpoints (LCO 3.4.12)7 3.2 LTOP Enable Temperature 7 3.3 Reactor Vessel Boliup Temperature (Non-Technical Specification) 7 4.0 Reactor Vessel Material Surveillance Program 10 5.0 Supplemental Data Tables 12 6.0 References 19 BRAID WOOl)-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT List of Figures Figure Page 2.1 Braidwood Unit I Reactor Coolant System Heatup Limitations (Heatup 3 Rate of l00°F/hr)Applicable for 32 EFPY (Without Margins for Instrumentation Errors)2.2 Braidwood Unit 1 Reactor Coolant System Cooldown Limitations 4 (Cooldown Rates of 0, 25, 50, and 100°F/hr)Applicable for 32 EFPY (Without Margins for Instrumentation Errors)3.1 Braidwood Unit 1 Nominal PORV Setpoints for the Low Temperature 8 Overpressure Protection (LTOP)System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) ft BRAID WOOD-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT List of Tables Table Page 2.la Braidwood Unit I Heatup Data Points at 32 EFPY (Without 5 Margins for Instrumentation Errors)2.lb Braidwood Unit 1 Cooldown Data Points at 32 EFPY (Without 6 Margins for Instrumentation Effors)3.1 Data Points for Braidwood Unit 1 Nominal PORV Setpoints for 9 the LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)
4.1 Braidwood
Unit 1 Capsule Withdrawal Schedule 11 5.1 Braidwood Unit 1 Calculation of Chemistry Factors Using 13 Surveillance Capsule Data 5,2 Braidwood Unit 1 Reactor Vessel Material Properties 14 5.3 Summary of Braidwood Unit I Adjusted Reference Temperatures 15 (ARTs)at 1/4T and 3/4T Locations for 32 EFPY 5.4 Braidwood Unit 1 Calculation of Adjusted Reference 16 Temperatures (ARTs)at 32 EFPY at the Limiting Reactor Vessel Material, Nozzle Shell Forging 5P-7016 5.5 RTprs Calculation for Braidwood Unit 1 Beltline Region 17 Materials at EOL (32 EFPY)5.6 RTp~5 Calculation for Braidwood Unit 1 Beltline Region 18 Materials at Life Extension (48 EFPY)
BRAID WOOD-UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 Introduction This Pressure and Temperature Limits Report (PTLR)for Braidwood Unit I has been prepared in accordance with the requirements of Braidwood Technical Specification (TS)5.6.6,"Reactor Coolant System (RCS)Pressure and Temperature Limits Report (PTLR)".Revisions to the PTLR shall be provided to the NRC after issuance.The Technical Specifications (TS)addressed in this report are listed below: LCO 3.4.3 RCS Pressure and Temperature (PiT)Limits;and LCO 3.4.12 Low Temperature Overpressure Protection (LTOP)System.2.0 RCS Pressure and Temperature Limits The PTLR limits for Braidwood Unit 1 were developed using a methodology specified in the Technical Specifications.
The methodology listed in WCAP-14040-NP-A, Revision 2 (Reference 1)was used with the following exceptions:
a)Optional use of ASME Code Section XI, Appendix G, Article G-2000, 1996 Addenda, b)Use of ASME Code Case N-640,"Alternative Reference Fracture Toughness for Development of P-T Limit Curves,Section XI, Division 1", c)Use of ASME Code Case N-588,"Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessel,Section XI, Division 1", and d)Elimination of the flange requirements documented in WCAP-16143-~P.
These exceptions to the methodology in WCAP 14040-NP-A, Revision 2 have been reviewed and accepted by the NRC in References 2, 8, 9 and 10.WCAP 15364, Revision 2 (Reference 11), provides the basis for the Braidwood Unit 1 P/T curves, along with the best estimate chemical compositions, fluence projections and adjusted reference temperatures used to determine these limits.WCAP-I 6143-P.Reference 12.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 WCAP-15364, Revision 2 (Reference Il)are: a.A maximum heatup of 100°F in any i-hour period, b.A maximum cooldown of 100°F in any 1-hour period, and BRAID WOOD-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 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.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.ia.The RCS P/T limits for cooldown are shown in Figure 2.2 and Table 2.lb.These limits are defined in WCAP-15364, Revision 2 (Reference II).Consistent with the methodology described in Reference 1 and exceptions noted in Section 2.0, 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 XL Appendix G, Article G2000, 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.