ML042250516
| ML042250516 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 08/26/2004 |
| From: | Dick G NRC/NRR/DLPM/LPD3 |
| To: | Crane C Exelon Generation Co |
| Dick G F, NRR/DLPM, 415-3019 | |
| Shared Package | |
| ML042250531 | List: |
| References | |
| TAC MB9760, TAC MB9761, TAC MB9762, TAC MB9763 | |
| Download: ML042250516 (22) | |
Text
August 26, 2004 Mr. Christopher M. Crane, President and Chief Nuclear Officer Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
BYRON STATION, UNITS 1 AND 2, AND BRAIDWOOD STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS, RE: PRESSURIZER SAFETY VALVE SETPOINTS, (TAC NOS. MB9762, MB9763, MB9760, AND MB9761)
Dear Mr. Crane:
The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 138 to Facility Operating License No. NPF-37 and Amendment No. 138 to Facility Operating License No. NPF-66 for the Byron Station, Unit Nos. 1 and 2, respectively, and Amendment No. 131 to Facility Operating License No. NPF-72 and Amendment No. 131 to Facility Operating License No. NPF-77 for the Braidwood Station, Unit Nos. 1 and 2, respectively. The amendments are in response to the Exelon Generation Company, LLC, application dated June 27, 2003, (ML031810344) as supplemented by letters dated January 29, 2004, (ML040300843), March 3, 2004, (ML040650522), June 4, 2004, (ML041600213), and August 11, 2004 (ML042260330).
The amendments revise the band of the pressurizer safety valve tolerance specified in the technical specifications. Specifically, the changes revise TS 3.4.10, Pressurizer Safety Valves, by changing the existing pressurizer safety valve lift settings from 2460 psig and
2510 psig, to 2411 psig and 2509 psig. The change represents a change in lift tolerance from +/-1 percent around a lift setting of 2485 psig to +/-2 percent around a lift setting of 2460 psig.
C. Crane A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commissions biweekly Federal Register notice.
Sincerely,
/RA/
George F. Dick, Jr., Project Manager, Section 2 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. STN 50-454, STN 50-455, STN 50-456 and STN 50-457
Enclosures:
- 1. Amendment No. 138 to NPF-37
- 2. Amendment No. 138 to NPF-66
- 3. Amendment No. 131 to NPF-72
- 4. Amendment No. 131 to NPF-77
- 5. Safety Evaluation cc w/encls: See next page
C. Crane A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commissions biweekly Federal Register notice.
Sincerely,
/RA/
George F. Dick, Jr., Project Manager, Section 2 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. STN 50-454, STN 50-455, STN 50-456 and STN 50-457
Enclosures:
- 1. Amendment No. 138 to NPF-37
- 2. Amendment No. 138 to NPF-66
- 3. Amendment No. 131 to NPF-72
- 4. Amendment No. 131 to NPF-77
- 5. Safety Evaluation cc w/encls: See next page DISTRIBUTION:
PUBLIC PDIII-2 R/F WRuland AMendiola GDick PCoates OGC ACRS GHill (8)
TBoyce AStone, RIII DLPM DPR SSun WPoertner ADAMS Accession Number: ML042250531 (Package)
ADAMS Accession Number: ML042250516 (Amendment)
ADAMS Accession Number: ML04 (Technical Specifications)
- via memo OFFICE PM:LPD3-2 LA:LPD3-2 SC:EMEB SC:SRXB SC:IROB OGC w/nlo SC:LPD3-2 NAME GDick PCoates SSun*
WPoertner*
TBoyce RHoefling AMendiola DATE 08/17/04 08/17/04 7/15/04 7/15/04 8/17/04 08/24/04 08/25/04 OFFICIAL RECORD COPY
Byron/Braidwood Stations cc:
Ms. C. Sue Hauser, Project Manager Westinghouse Electric Corporation Energy Systems Business Unit Post Office Box 355 Pittsburgh, PA 15230 Joseph Gallo Gallo & Ross 1025 Connecticut Ave., NW, Suite 1014 Washington, DC 20036 Howard A. Learner Environmental Law and Policy Center of the Midwest 35 East Wacker Dr., Suite 1300 Chicago, IL 60601-2110 U.S. Nuclear Regulatory Commission Byron Resident Inspectors Office 4448 N. German Church Road Byron, IL 61010-9750 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, IL 60532-4351 Ms. Lorraine Creek RR 1, Box 182 Manteno, IL 60950 Chairman, Ogle County Board Post Office Box 357 Oregon, IL 61061 Mrs. Phillip B. Johnson 1907 Stratford Lane Rockford, IL 61107 George L. Edgar Morgan, Lewis and Bockius 1800 M Street, NW Washington, DC 20036-5869 Attorney General 500 S. Second Street Springfield, IL 62701 Illinois Emergency Management Agency Division of Disaster Assistance &
Preparedness 110 East Adams Street Springfield, IL 62701-1109 Byron Station Plant Manager Exelon Generation Company, LLC 4450 N. German Church Road Byron, IL 61010-9794 Site Vice President - Byron Exelon Generation Company, LLC 4450 N. German Church Road Byron, IL 61010-9794 U.S. Nuclear Regulatory Commission Braidwood Resident Inspectors Office 35100 S. Rt. 53, Suite 79 Braceville, IL 60407 Chairman Will County Board of Supervisors Will County Board Courthouse Joliet, IL 60434 Braidwood Station Plant Manager Exelon Generation Company, LLC 35100 S. Rt. 53, Suite 84 Braceville, IL 60407-9619 Ms. Bridget Little Rorem Appleseed Coordinator 117 N. Linden Street Essex, IL 60935
Byron/Braidwood Stations Document Control Desk - Licensing Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 Site Vice President - Braidwood Exelon Generation Company, LLC 35100 S. Rt. 53, Suite 84 Braceville, IL 60407-9619 Senior Vice President - Nuclear Services Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 Vice President of Operations - Mid-West Pressurized Water Reactors Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 Director - Licensing and Regulatory Affairs Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 Regulatory Assurance Manager - Braidwood Exelon Generation Company, LLC 35100 S. Rt. 53, Suite 84 Braceville, IL 60407-9619 Regulatory Assurance Manager - Byron Exelon Generation Company, LLC 4450 N. German Church Road Byron, IL 61010-9794 Associate General Counsel Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 Vice President - Licensing and Regulatory Affairs Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 Manager Licensing - Braidwood, Byron and LaSalle Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555
EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-454 BYRON STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 138 License No. NPF-37 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Exelon Generation Company, LLC (the licensee) dated June 27, 2003, as supplemented by letters dated January 29, 2004, March 3, 2004, June 4, 2004, and August 11, 2004, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-37 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 138 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license.
The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance and shall be implemented within 30 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Anthony J. Mendiola, Chief, Section 2 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: August 26, 2004
EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-455 BYRON STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 138 License No. NPF-66 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Exelon Generation Company, LLC (the licensee) dated June 27, 2003, as supplemented by letters dated January 29, 2004, March 3, 2004, June 4, 2004, and August 11, 2004, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-66 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A (NUREG-1113), as revised through Amendment No. 138 and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF-37, dated February 14, 1985, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance and shall be implemented within 30 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Anthony J. Mendiola, Chief, Section 2 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: August 26, 2004
ATTACHMENT TO LICENSE AMENDMENT NOS. 138 AND 138 FACILITY OPERATING LICENSE NOS. NPF-37 AND NPF-66 DOCKET NOS. STN 50-454 AND STN 50-455 Replace the following page of the Appendix A Technical Specifications with the attached page. The revised page is identified by amendment number and contain marginal lines indicating the area of change.
Remove Page Insert Page 3.4.10-1 3.4.10-1
EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-456 BRAIDWOOD STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 131 License No. NPF-72 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Exelon Generation Company, LLC (the licensee) dated June 27, 2003, as supplemented by letters dated January 29, 2004, March 3, 2004, June 4, 2004, and August 11, 2004, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-72 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 131 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance and shall be implemented within 30 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Anthony J. Mendiola, Chief, Section 2 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: August 26, 2004
EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-457 BRAIDWOOD STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 131 License No. NPF-77 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Exelon Generation Company, LLC (the licensee) dated June 27, 2003, as supplemented by letters dated January 29, 2004, March 3, 2004, June 4, 2004, and August 11, 2004, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-77 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 131 and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF-72, dated July 2, 1987, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date if its issuance and shall be implemented within 30 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Anthony J. Mendiola, Chief, Section 2 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: August 26, 2004
ATTACHMENT TO LICENSE AMENDMENT NOS. 131 AND 131 FACILITY OPERATING LICENSE NOS. NPF-72 AND NPF-77 DOCKET NOS. STN 50-456 AND STN 50-457 Replace the following page of the Appendix "A" Technical Specifications with the attached page. The revised page is identified by amendment number and contain marginal lines indicating the area of change.
Remove Page Insert Page 3.4.10-1 3.4.10-1
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 138 TO FACILITY OPERATING LICENSE NO. NPF-37, AMENDMENT NO. 138 TO FACILITY OPERATING LICENSE NO. NPF-66, AMENDMENT NO. 131 TO FACILITY OPERATING LICENSE NO. NPF-72, AND AMENDMENT NO. 131 TO FACILITY OPERATING LICENSE NO. NPF-77 EXELON GENERATION COMPANY, LLC BYRON STATION, UNIT NOS. 1 AND 2 BRAIDWOOD STATION, UNIT NOS. 1 AND 2 DOCKET NOS. STN 50-454, STN 50-455, STN 50-456 AND STN 50-457
1.0 INTRODUCTION
By application dated June 27, 2003, (ML031810344) as supplemented by letters dated January 29, 2004, (ML040300843), March 3, 2004, (ML040650522), June 4, 2004, (ML041600213), and August 11, 2004, (ML042260330), Exelon Generation Company, LLC (the licensee) requested changes to the Technical Specifications (TS) for Byron Station, Units 1 and 2, (Byron) and Braidwood Station, Units 1, and 2 (Braidwood). The August 11, 2004, letter was a request from the licensee regarding normalizing the amendment numbers for Byron. The request was administrative in nature and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register on September 30, 2003 (68 FR 56343). The other supplements provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination.
The proposed changes would revise the band of the pressurizer safety valve tolerance specified in the TS. Specifically the proposed changes would revise TS 3.4.10, Pressurizer Safety Valves, by changing the existing pressurizer safety valve (PSV) lift settings from 2460 psig and 2510 psig, to 2411 psig and 2509 psig. The change represents a change in lift tolerance from +/-1 percent around a lift setting of 2485 psig to +/-2 percent around a lift setting of 2460 psig.
2.0 REGULATORY EVALUATION
General Design Criteria (GDC) 10, Reactor Design, found in Appendix A of Title 10 of the Code of Federal Regulations (10 CFR), Part 50 requires that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. GDC 15, Reactor Coolant System Design, requires that the that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including the effects of anticipated op-operational occurrences.
Additionally, 10 CFR 50.36(c)(1) requires that plant TS will include safety limits, limiting safety system settings, and limiting control settings. 10 CFR 50.36(c)(2)(ii)(C) specifies that a limiting condition for operation be established for a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure or presents a challenge to the integrity of a fission product barrier. Pressurizer safety valves (PSVs) are part of the primary success path to mitigate consequences of design-basis events (DBEs), and are credited in the Braidwood and Byron Final Safety Analysis Reports (UFSARs) analyses.
3.0 TECHNICAL EVALUATION
3.1 Background
The current Braidwood and Byron TS 3.4.10 requires that three PSVs shall be operable with as-found lift settings 2460 psig and 2510 psig. The current TS values represent a +/-1 percent setpoint tolerance around a nominal lift setting of 2485 psig for the PSVs. The licensee indicated that there had been many instances where one or more of the Braidwood and Byron PSVs were found to have setpoints outside the +/-1 percent setpoint tolerance, which resulted in PSVs being declared inoperable. It also indicated that most of the as-found lift settings had not exceeded +/-2 percent of the nominal pressure setting. Therefore, the licensee proposed TS changes to reduce the nominal setpoint and increase the setpoint tolerance for the PSVs to minimize TS violations caused by setpoint drift.
The licensee proposed to change the existing as found PSV lift settings to a range of 2411 psig and 2509 psig. The proposed TS reflects changes in the allowed PSV tolerance from +/-1 percent to +/-2 percent and a reduction in the nominal lift setting from 2485 psig to 2460 psig.
The TS change allows a decrease in the valve actuation pressure and, therefore, provides the potential for earlier pressurizer relief at a reduced reactor coolant system (RCS) pressure. The licensee would not change the +/-1 percent as-left setpoint tolerance for the PSVs specified in the current TS Surveillance Requirement (SR) 3.4.10.1. The proposed change would revise the associated Bases for SR 3.4.10.1 to reflect the proposed PSV settings.
In support of the proposed TS changes, the licensee provided the results of its technical evaluation (Ref. 1) and responses (Refs. 2 through 4) to the staff requests for additional information for the staff to review.
3.2 Evaluation The staff has reviewed the licensees proposed TS changes and the associated supporting analysis (Refs. 1 and 4) for the Braidwood and Byron plants, and prepared the following evaluation.
3.2.1 PSV Setpoint with Positive Tolerance Analysis Each unit of the Braidwood and Byron plants has three spring-loaded PSVs with a relief capacity of 420,000 lb/hr for each valve. As described in Section 5.4.13 of the Braidwood and Byron UFSARs, the PSVs provide over pressure protection for the RCS. Together with the reactor protection system, the PSVs ensure that the RCS pressure meets the GDC 15 requirement in terms of the RCS design pressure safety limit. Compliance with the GDC 15 requirement is demonstrated in the analysis of the DBEs. In assessing the effects of the TS changes on the DBE analysis, the licensee evaluated the analysis of record (AOR) and identified that six DBEs relied on PSV actuation to limit the pressure increase to below the pressure safety limit of 110 percent of the design pressure. The events are: (1) uncontrolled rod withdrawal from full power; (2) loss of reactor coolant flow; (3) loss of external electrical load; (4) loss of normal feedwater; (5) loss of all AC power to station auxiliaries; and (6) reactor coolant pump locked rotor.
The licensee indicated that the AOR for the events listed above assumed that the PSVs would fully open when the calculated pressurizer pressure reaches 2534.7 psig, which corresponds to the PSV setpoint of 2485 psig with the associated tolerance of +1 percent and an additional
+1 percent allowance to account for a pressure shift due to operation with water-filled pressurizer loop seals. Modeling the proposed PSV setpoint changes would result in a PSV full open pressure of 2533.8 psig, which is based on the proposed PSV lift setting of 2460 psig with the associated +2 percent tolerance and an additional +1 percent to account for pressure shift.
The setpoint of 2534.7 psig assumed in the AOR generates a higher effective PSV opening pressure and results in a higher peak pressurizer pressure during an event. Therefore, the staff determined that the AOR remains bounding for over pressure protection and is valid for supporting the proposed PSV lift setting of 2460 psig with an upper tolerance of +2 percent.
3.2.2 PSV Setpoint with Negative Tolerance Analysis Use of PSV setpoints with negative tolerances lowers effective PSV opening pressures, which would cause an earlier opening of the PSVs and a slower increase in RCS pressure during over pressurization events. PSV actuation at a lower RCS pressure could result in a lower departure from nucleate boiling ratio (DNBR) that reduces the margin to the safety DNBR limit, and it could also result in a higher pressurizer water level that increases the potential to overfill the pressurizer with water. The licensee evaluated the AOR and identified that the following events assumed PSV opening with negative tolerance modeled:
(1) loss of load (LOL) / turbine trip (TT),
(2) rod withdrawal at power (RWAP),
(3) loss of alternating current (LOAC) with reactor coolant pump (RCP) seal injection, (4) loss of normal feedwater (LONF), and (5)
LOAC.
The licensee reanalyzed the five events for both Units 1 (with Babcock and Wilcox International steam generators) and Units 2 (with Westinghouse Model D5 steam generators). During the course of the review, the staff requested the licensee to discuss the methods used for the reanalysis and identify input parameters that were different from those assumed in the AOR. In response (Ref. 2), the licensee indicated that the computer code (LOFTRAN) and methods used in the reanalysis are the same as those used in the AOR. For events 1 through 3, changes in input parameters are associated with the proposed change in PSV setpoint and tolerance. For Events 4 and 5, changes in input parameters are associated with a decrease in initial feedwater temperature and the proposed change in PSV setpoint and tolerance. The reduction in initial feedwater temperature causes a slower increase in RCS pressure and results in a higher pressurizer water level. Therefore, the assumption of a lower initial feedwater temperature is conservative with respect to the calculated pressurizer water level.
3.2.3 DNBR Reanalysis The licensee performed the DNBR reanalysis for the LOL/TT and RWAP events that were identified in the AOR as the limiting cases that resulted in the lowest DNBR values. The results of the reanalysis (Ref. 1) indicated that for the LOL/TT event, although the calculated minimum DNBR values were slightly lower than those calculated in the AOR, they remained well above the DNBR safety limit. For the RWAP event, the analysis indicated that the results for Units 2 remained limiting and that the minimum DNBR remained above that calculated in the AOR.
Since the licensees reanalysis of the limiting events showed that the calculated DNBRs are either bounded by the AOR or meet the DNBR safety limits, the staff determined that the reanalysis satisfies the GDC 10 requirement related to fuel rod integrity criteria, and concluded that the reanalysis is acceptable to support the proposed TS.
3.2.4 Pressurizer Water Inventory Reanalysis The licensee performed a reanalysis of pressurizer water level for the LOL/TT, LONF, LOAC and LOAC with RCP seal injection events that were identified in the AOR as the limiting cases that resulted in the highest pressurizer water levels. The results of the reanalysis indicated that for the LOL/TT, LONF, or LOAC events, although the calculated peak pressurizer water level was slightly higher than that calculated in the AOR, it did not reach the top of the pressurizer, and thus, the results demonstrated that no water discharge from the pressurizer occurred.
As for the LOAC with RCP seal injection event, the licensees analysis indicated that continued injection of water into the RCS through the RCP seals would result in a water-solid pressurizer and water discharge through the PSVs. The proposed PSV setpoint tolerance assuming negative tolerance would result in a lower PSV lift setpoint. With the lower setpoint, the PSV would open earlier and a larger number of PSV water cycles and a lower water discharge temperature could result during the transient. The licensee performed an analysis of the LOAC with RCP seal injection event and determined (Ref. 1) the revised PSV setpoint would result in an increase of about one PSV water cycle and a reduction in the liquid discharge temperature of about 0.5 oF.
The licensee also performed a qualitative evaluation (Ref. 1) of the effect of the proposed TS changes on the spurious safety injection (SI) at power event, the limiting AOR event with respect to water inventory addition to the pressurizer. Based on its qualitative evaluation, the licensee claimed that the event would show results similar to those of the LOAC with RCP seal injection event in terms of the number of PSV water cycles and the PSV discharge water temperature. However, the licensee did not provide the results of a quantitative analysis of the spurious SI event to support its position. The staff requested the licensee to quantify the effect of the lower PSV setpoint on the AOR limiting event. In response, the licensee performed a reanalysis (Ref. 2) and showed that the revised PSV setpoint would result in an increase of two PSV water cycles and a reduction in the liquid discharge temperature of no more than 3.0 oF.
A comparison of the reanalysis showed that the spurious SI event remained the limiting event since it resulted in a greater increase in the number of PSV water cycles (two cycles vs. one cycle) and a greater decrease in the PSV discharge water temperature (3.0 oF vs. 0.5 oF) than that calculated for the LOAC with RCP seal injection event. As indicated in reference 6, the water discharge temperature in the AOR for the spurious SI event was 590 oF. The lowest discharge water temperature for the spurious SI event with the revised PSV setpoint is 587 oF (i.e., 590 oF - 3.0 oF ). The staff found that the calculated water discharge temperature (587 oF) is significantly higher than the discharge water temperature of 530 oF that was used to support operability of the PSVs as discussed in the AOR (Ref. 6). Therefore, the staff concludes that the reanalysis is acceptable to assure that the PSVs will remain operable following a spurious SI event.
Since (1) the reanalyses used the same methods as those used in the AOR, (2) the values of input parameters, except for the PSV opening pressure and the conservative assumption of a lower initial feedwater temperature, used in the reanalyses were the same as that assumed in the AOR, and (3) the results of the reanalyses showed that the calculated minimum DNBR did not exceed the DNBR safety limits, and that the calculated PSV operating conditions did not exceed the AOR PSV operability range previously approved by the staff, the staff concludes that the reanalyses are acceptable.
3.3 Margin Between High Pressure Reactor Trip and Opening of PSVs The licensee indicated that the PSV setpoints were established to be above the setpoint of the high pressure reactor trip to minimize challenges to the PSVs. During the review, the staff requested that the licensee specify the pressure measurement uncertainties associated with the high pressure reactor trip and the PSVs, and confirm that they were appropriately considered in the error analysis such that a reactor trip would occur prior to a PSV actuation. In the licensee response (Ref. 2), the error analysis of the pressure measurement uncertainties showed that the lowest lifting pressure for a PSV (with the proposed setpoint of 2460 psig) is 2411 psig and the highest pressure for a reactor trip (with a nominal setpoint of 2385 psig) is 2427 psig. The licensee concluded that the probability of having a PSV lift before achieving a high pressure trip is less than 1% for any given pressure based on instrument uncertainties and that a reactor trip was expected to occur prior to a PSV actuation. Based on this information, the staff requested that the licensee evaluate the impact on the appropriate accident analysis acceptance criteria.
In response, the licensee evaluated the AOR and identified (Ref. 3) that only the peak pressure cases for the LOL/TT and RWAP analyses resulted in a reactor trip on high pressurizer pressure. The licensee reanalyzed (Ref. 4) these two events by assuming that the PSVs would lift at the low end of the tolerance band with and without including a one second delay in PSV lifting and subsequent steam relief through the PSVs to account for PSV loop seal clearance.
The PSVs operate with a water loop seal during normal operation and the loop seal results in a nominal 1 second delay (as documented in WCAP-12910) in PSV lifting while the water is purged from the loop seal. When the PSV loop seal purge delay was included in the analysis, the high pressurizer pressure trip was reached prior to steam relief through the PSVs. When the loop seal purge delay was not included in the analysis, the PSVs relieved prior to receiving the high pressurizer pressure reactor trip signal. The licensee analysis determined that a reactor trip would occur on over temperature delta-T (OTDT) in the LOL/TT case, and on high neutron flux in the RWAP case if credit was not assumed for the loop seal purge delay. The results of the reanalysis showed that the reactor trip occurred only a short time after it would have occurred on high pressurizer pressure (i.e., rod motion would start approximately 7.6 seconds later for the LOL/TT event and approximately 0.3 seconds later for the RWAP event).
The reanalysis further confirmed that all acceptance criteria remained satisfied, in particular, the peak RCS pressure and steam generator secondary side pressure remained less than the allowable limits and the pressurizer did not reach a water-solid condition. Based on the above analysis, the staff determined that a reactor trip would occur during the analyzed events and that all acceptance criteria remained satisfied.
3.5 NUREG-1431 Consistency The staff found that the proposed tolerance of +/-2 percent for the PSV setpoint is within the allowable range specified in NUREG-1431 (Ref. 4), Standard Technical Specifications for Westinghouse Plants. Specifically, the Bases for SR 3.4.10.1 allow a tolerance range of +/-3 percent for the PSV setpoint.
3.6 Summary Since (1) the proposed changes to the PSV setpoint tolerances are adequately reflected in the acceptable analyses that satisfy the requirements of GDC 10 for fuel integrity and GDC 15 for integrity of the RCS pressure boundary, and (2) the proposed PSV setpoint tolerances are within the range allowed for Westinghouse-designed plants, such as the Braidwood and Byron plants, as specified in the Standard Technical Specifications, NUREG-1431 (Revision 2), the staff concludes that the proposed PSV setpoint tolerances are acceptable.
As stated in the Bases for SR 3.4.10.1, the licensee would not change the +/-1 percent as-left setpoint tolerance for the PSVs. Setting the PSV to this tolerance helps reduce the overall setpoint drift over time, which is acceptable to the staff.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Illinois State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (68 FR 56343). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
7.0 REFERENCES
1.
Letter from K. A. Ainger (Exelon Generation Company, LLC) to NRC, "Request for a License Amendment to Revise the Pressurizer Safety Valves Lift Settings, dated June 27, 2003.
2.
Letter from K. A. Ainger (Exelon Generation Company, LLC) to NRC, "Request for Additional Information Regarding a License Amendment Request to Revise the Pressurizer Safety Valves Lift Settings, dated January 29, 2004.
3.
Letter from K. A. Ainger (Exelon Generation Company, LLC) to NRC, "Request for Additional Information Regarding a License Amendment Request to Revise the Pressurizer Safety Valves Lift Settings, dated March 3, 2004.
4.
Letter from K. A. Ainger (Exelon Generation Company, LLC) to NRC, "Request for Additional Information Regarding a License Amendment Request to Revise the Pressurizer Safety Valves Lift Settings, dated June 4, 2004.
5.
NUREG-1431, Revision 2, Standard Technical Specifications - Westinghouse Plants, April 30, 2001.
6.
Letter from G. F. Dick (NRC) to O. D. Kingsley (EGC), Issuance of Amendments; Increase in Reactor Power, Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, dated May 4, 2001.
Principle Contributors: S. Sun W. Poertner Date: August 26, 2004