ML14079A102
| ML14079A102 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 05/05/2014 |
| From: | Richard Ennis Plant Licensing Branch 1 |
| To: | Pacilio M Exelon Nuclear |
| Ennis R, NRR/DORL/LPL1-2, 415-1420 | |
| References | |
| TAC MF1970, TAC MF1971 | |
| Download: ML14079A102 (35) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 5, 2014
SUBJECT:
PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 AND 3 -ISSUANCE OF AMENDMENTS RE: SAFETY RELIEF VALVE AND SAFETY VALVE LIFT SETPOINT TOLERANCE (TAC NOS. MF1970 AND MF1971)
Dear Mr. Pacilio:
The Commission has issued the enclosed Amendment Nos. 290 and 293 to Renewed Facility Operating License Nos. DPR-44 and DPR-56 for Peach Bottom Atomic Power Station, Units 2 and 3. These amendments consist of changes to the Technical Specifications (TSs) and Facility Operating Licenses in response to your application dated June 10, 2013, as supplemented by letter dated November 6, 2013.
The amendments revise the TSs to: (1) increase the allowable as-found safety relief valve (SRV) and safety valve (SV) lift setpoint tolerance from +/- 1% to +/- 3%; (2) increase the required number of operable SRVs and SVs from 11 to 12; and (3) increase the Standby Liquid Control System pump discharge pressure from 1255 pounds per square inch gauge (psig) to 1275 psig.
A copy of the safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's Biweekly Federal Register notice.
Sincerely, Richard B. Ennis, Senior Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-277 and 50-278
Enclosures:
- 1. Amendment No. 290 to Renewed DPR-44
- 2. Amendment No. 293 to Renewed DPR-56
- 3. Safety Evaluation cc w/encls: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC PSEG NUCLEAR LLC DOCKET NO. 50-277 PEACH BOTTOM ATOMIC POWER STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 290 Renewed License No. DPR-44
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Exelon Generation Company, LLC (Exelon Generation Company), and PSEG Nuclear LLC (the licensees), dated June 10, 2013, as supplemented by letter dated November 6, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 Commission's 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 Commission's regulations and all applicable requirements have been satisfied.
Enclosure 1
- 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 Renewed Facility Operating License No. DPR-44 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 290, are hereby incorporated in the license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION Meena K. Khanna, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications and Facility Operating License Date of Issuance: May 5 , 2 0 1 4
ATTACHMENT TO LICENSE AMENDMENT NO. 290 RENEWED FACILITY OPERATING LICENSE NO. DPR-44 DOCKET NO. 50-277 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove 3
Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert 3.1-23 3.1-23 3.4-8 3.4-8 3.4-9 3.4-9
(5) Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not to separate, such byproduct and special nuclear material as may be produced by operation of the facility, and such Class B and Class C low-level radioactive waste as may be produced by the operation of Limerick Generating Station, Units 1 and 2.
C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:
(1) Maximum Power Level Exelon Generation Company is authorized to operate the Peach Bottom Atomic Power Station, Unit 2, at steady state reactor core power levels not in excess of 3514 megawatts thermal.
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 290, are hereby incorporated in the license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications.
(3) Physical Protection Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans 1, submitted by letter dated May 17, 2006, is entitled: "Peach Bottom Atomic Power Station Security Plan, Training and Qualification Plan, Safeguards Contingency Plan, and Independent Spent Fuel Storage Installation Security Program, Revision 3." The set contains Safeguards Information protected under 10 CFR 73.21.
Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP),
including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Exelon Generation Company CSP was approved by License Amendment No. 283.
(4) Fire Protection The Exelon Generation Company shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report for the facility, and as approved in the NRC Safety Evaluation Report (SER) dated May 23, 1979, and Supplements dated August 14, September 15, October 10 and November 24, 1980, and in the NRC SERs dated September 16, 1993, and August 24, 1994, subject to the following provision:
1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.
Renewed License No. DPR-44 Revised by letter dated October 28, 2004 Revised by letter dated May 29, 2007 Amendment No. 290 Page 3
SLC System
- 3. 1 . 7 SURVEILLANCE REQUIREMENTS Ccontinuedl SURVEILLANCE FREQUENCY SR 3.1.7.7 Verify the quantity of B-10 stored in the In accordance SLC tank is? 162.7 lbm. with the Surveillance Frequency Control Program.
SR 3.1.7.8 Verify each pump develops a flow rate In accordance
? 43.0 gpm at a discharge pressure with the
- > 1275 psig. Inservice Testing Program SR 3.1.7.9 Verify flow through one SLC subsystem from In accordance pump into reactor pressure vessel. with the Surveillance Frequency Control Program.
SR 3.1.7.10 Ver*i fy sodium pentaborate atom percent B-10 Once within 8 enrichment is within the 1 i mi ts of hours after Table 3.1.7-1. addition to SL_C tank PBAPS UNIT 2 3.1-23 Amendment No. 290
SRVs and SVs 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety Relief Valves (SRVs) and Safety Valves (SVs)
LCO 3. 4. 3 The safety function of 12 valves (any combination of SRVs and SVs ) s hal l be 0 PERA BLE.
APPLICABILITY: MODES l, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A. 1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SRVs or SVs inoperable.
A.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> PBAPS UN IT 2 3.4-8 Amendment No. 290
SRVs and SVs 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoints In accordance of the required SRVs and SVs are as with the follows: lnservice Testing Program Number of Setpoint SRVs (psi g) 4 1135 +/- 34.1 4 1145 +/- 34.4 3 1155 +/- 34.7 Number of Setpoint SVs ( Rs i g) 2 1260 +/- 37.8 Following testing, lift settings shall be within+/- 1%.
SR 3.4.3.2 Verify each required SRV actuator strokes In accordance when manually actuated in the with the depressurization mode. Surveillance Frequency Control Program.
PBAPS UN IT 2 3.4-9 Amendment No. 290
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC PSEG NUCLEAR LLC DOCKET NO. 50-278 PEACH BOTTOM ATOMIC POWER STATION, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 293 Renewed License No. DPR-56
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Exelon Generation Company, LLC (Exelon Generation Company), and PSEG Nuclear LLC (the licensees), dated June 10, 2013, as supplemented by letter dated November 6, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 Commission's 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 Commission's regulations and all applicable requirements have been satisfied.
Enclosure 2
- 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 Renewed Facility Operating License No. DPR-56 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No. 293, are hereby incorporated in the license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION
~
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Meena K. Khanna, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications and Facility Operating License Date of Issuance: May 5 , 2 0 1 4
ATTACHMENT TO LICENSE AMENDMENT NO. 293 RENEWED FACILITY OPERATING LICENSE NO. DPR-56 DOCKET NO. 50-278 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove Insert 3 3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert 3.1-23 3.1-23 3.4-8 3.4-8 3.4-9 3.4-9
(5) Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not to separate, such byproduct and special nuclear material as may be produced by operation of the facility, and such Class B and Class C low-level radioactive waste as may be produced by the operation of Limerick Generating Station, Units 1 and 2.
C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:
(1) Maximum Power Level Exelon Generation Company is authorized to operate the Peach Bottom Atomic Power Station, Unit No. 3, at steady state reactor core power levels not in excess of 3514 megawatts thermal.
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 293, are hereby incorporated in the license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications. 1 (3) Physical Protection Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans 2 , submitted by letter dated May 17, 2006, is entitled: "Peach Bottom Atomic Power Station Security Plan, Training and Qualification Plan, Safeguards Contingency Plan, and Independent Spent Fuel Storage Installation Security Program, Revision 3." The set contains Safeguards Information protected under 10 CFR 73.21.
Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP),
including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Exelon Generation Company CSP was approved by License Amendment No. 283.
1 Licensed power level was revised by Amendment No. 250, dated November 22, 2002, and will be implemented following the 141h refueling outage currently scheduled for Fall 2003.
2 The training and Qualification Plan and Safeguards Contingency Plan and Appendices to the Security Plan.
Renewed License No. DPR-56 Revised by letter dated October 28, 2004 Revised by letter dated November 5, 2004 Revised by letter dated May 29, 2007 Amendment No. 293 Page 3
SLC System 3 1. 7 0
SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.1.7.7 Verify the quantity of B-10 stored in the In accordance SLC tank is~ 162.7 lbm. with the Surveillance Frequency Control Program.
SR 3.1.7.8 Verify each pump develops a flow rate In accordance 2 43.0 gpm at a discharge pressure with the 2: 12/5 psig. Inservice Testing Program SR 3.1.7.9 Ver' i fy flow U1rough one SLC subsystem from In accordance pump into reactor pressure vessel. with the Survei 11 ance Frequency Control Program.
SR 3.1.7.10 Verify sodium pentaborate atom percent B-10 Once within 8 enrichment is within the limits of hours after Table 3.1.7-1. addition to SLC tank PBAPS UN IT 3 3.1-23 Amendment No. 2 93
SRVs and SVs 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety Relief Valves (SRVsJ and Safety Valves (SVs)
The safety function of 12 valves Cany combination of SRVs and SVsl shall be OPERABLE.
APPLICABILITY: MODES 1 ' ')
L' and 3.
ACf IONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A. 1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SRVs or SVs inoperable. AND A.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> PBAPS UN IT 3 3.4-8 Amendment No. 2 93
SRVs and SVs 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoints In accordance of the required SRVs and SVs are as with the follows: Inservice Testing Program Number of Setpoint SRVs (psi q l 4 1135+/-34.1 4 1145 +/- 34.4 3 1155 +/- 34.7 Number of Setpoint SVs ( RS i q) 2 1260 +/- 37.8 Following testing, lift settings shall be within+/- 1%.
SR 3.4.3.2 Verify each required SRV actuator strokes In accordance when manually actuated in the with the depressurization mode. Surveillance Frequency Control Pr'ogram.
PBAPS UNIT 3 3.4-9 Amendment No. 293
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 290 AND 293 TO RENEWED FACILITY OPERATING LICENSE NOS. DPR-44 AND DPR-56 EXELON GENERATION COMPANY, LLC PSEG NUCLEAR LLC PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 AND 3 DOCKET NOS. 50-277 AND 50-278
1.0 INTRODUCTION
By application dated June 10, 2013, as supplemented by letter dated November 6, 2013, (Agencywide Documents Access and Management System (ADAMS) Accession Nos.
ML131750144, and ML13317A294, respectively), Exelon Generation Company, LLC (Exelon, the licensee), requested changes to the Technical Specifications {TSs) and Facility Operating Licenses (FOLs) for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3.
The proposed amendments would revise the TSs to: (1) increase the allowable as-found safety relief valve (SRV) and safety valve (SV) lift setpoint tolerance from +/- 1% to +/- 3%; (2) increase the required number of operable SRVs and SVs from 11 to 12; and (3) increase the Standby Liquid Control (SLC) system pump discharge pressure from 1255 pounds per square inch gauge (psig) to 1275 psig.
The supplement dated November 6, 2013, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC or the Commission) staff's original proposed no significant hazards consideration determination as published in the Federal Register on December 26, 2013 (78 FR 78406).
2.0 REGULATORY EVALUATION
2.1 Background to the licensee's application dated June 10, 2013, provides background information, which explains the purpose of the SRVs and SVs, and provides historical context explaining why this request has been submitted. The following is excerpted from the attachment:
Enclosure 3
The [American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)], Section Ill, requires the reactor pressure vessel to be protected from overpressure during upset conditions by self-actuated safety valves. As part of the nuclear pressure relief system, the size and number of SRVs and SVs are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB).
Eleven (11) valves (any combinations of SRVs and SVs) are currently required to be operable by TS 3.4.3. In support of this proposed change, the required number of operable SRVs/SVs will be increased from 11 to 12.
The SRVs are Target Rock three-stage pilot operated safety/relief valves. The SVs are Dresser spring loaded safety valves. The SRVs and SVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell. The SRVs can actuate by either of two modes: the safety mode or the depressurization mode. In the safety mode, the pilot disc opens when steam pressure at the valve inlet expands the bellows to the extent that the hydraulic seating force on the pilot disc is reduced to zero. Opening of the pilot stage allows a pressure differential to develop across the second stage disc which opens the second stage disc, thus venting the chamber over the main valve piston. This causes a pressure differential across the main valve piston which opens the main valve. The SVs are spring loaded valves that actuate when steam pressure at the inlet overcomes the spring force holding the valve disc closed. This satisfies the Code requirement. The proposed changes do not impact the depressurization mode function of the SRVs.
As discussed in Attachment 1 to the licensee's application, most boiling-water reactor (BWR)
TSs originally required that the safety mode pressure setpoints for SRVs and SVs remain within a +/-1% tolerance band. It was subsequently identified that nuclear power plant licensees were experiencing difficulty in meeting the +/-1% setpoint tolerance criterion. As a result, the BWR Owners' Group (BWROG) developed a licensing topical report (LTR) to support increasing the setpoint tolerance from +/-1% to +/-3%. The LTR, NEDC-31753P, "BWROG In-Service Pressure Relief Technical Specification Revision Licensing Topical Report," dated February 1990, was approved by the NRC staff in a safety evaluation (SE) dated March 8, 1993 (ADAMS Legacy Library Accession No. 9702070012, non-publicly available). The NRC staff's SE stated that each licensee choosing to implement the TS modifications, associated with increasing the setpoint tolerance from +/-1% to +/-3%, must provide certain plant-specific analyses and evaluations. Specifically, the staff's approval contained the following conditions/limitations:
- 1. Transient analysis of all abnormal operational occurrences as described in NEDC-31753P, should be performed utilizing a +/-3% setpoint tolerance for the safety mode of the SSVs and SRVs. In addition, the standard reload methodology or other method approved by the staff) should be used for this analysis.
- 2. Analysis of the design basis overpressurization event using the 3% tolerance limit for the SRV setpoint is required to confirm that the vessel pressure does not exceed the ASME pressure vessel code upset limit.
- 3. The plant specific analyses described in Items 1 and 2 should assure that the number of SSVs, SRVs, and relief valves (RVs) included in the analyses correspond to the number of valves required to be operable in the technical specification.
- 4. Re-evaluation of the performance of high pressure systems (pump capacity, discharge pressure, etc.), motor-operated valves, and vessel instrumentation and associated piping must be completed, considering the 3% tolerance limit.
- 5. Evaluation of the +/-3% tolerance on any plant specific alternate operating modes (e.g., increased core flow, extended operating domain, etc.) should be completed.
- 6. Evaluation of the effect of the 3% tolerance limit on the containment response during loss of coolant accidents and the hydrodynamic loads on the SRV discharge lines and containment should be completed.
In support of the proposed amendments for PBAPS, Units 2 and 3, General Electric Hitachi Nuclear Energy (GEH) performed plant-specific analyses and evaluations to address the six conditions/limitations listed above. GEH documented the analyses and evaluations in proprietary report, NEDC-33533P, Revision 1, dated May 2013, "Peach Bottom Atomic Power Station, Units 2 and 3, Safety Valve Setpoint Tolerance Increase Safety Analysis Report," which was provided in Attachment 3 to the licensee's application dated June 10, 2013. A non-proprietary version of the report, NED0-33533, was provided in Attachment 4 to the application.
2.2 Licensee's Proposed TS Changes The proposed amendments would revise the PBAPS, Units 2 and 3, TSs as shown in to the licensee's application dated June 10, 2013. The specific changes are discussed in SE Sections 2.2.1 and 2.2.2.
2.2.1 TS 3.4.3 -"Safety Relief Valves (SRVs) and Safety Valves (SVs)"
Limiting Condition for Operation (LCO) 3.4.3 currently requires that:
The safety function of 11 valves (any combination of SRVs and SVs) shall be OPERABLE.
The proposed amendments would revise LCO 3.4.3 to increase the required number of operable SRVs and SVs from 11 to 12.
In addition to the change to the LCO, the amendments would revise surveillance requirement (SR) 3.4.3.1. This SR requires that the safety function lift setpoints of the required SRVs and SVs be verified on a frequency in accordance with the lnservice Testing (1ST) Program. The
proposed amendments would not alter the setpoints, but would change the setpoint tolerance (in terms of psig) to reflect an increase from +/-1% to +/-3%. Specifically, the setpoints are currently follows:
Number of Set point SRVs (psi g) 4 1135 +/- 11.0 4 1145 +/- 11.0 3 1155 +/- 12.0 Number of Set point SVs (psi g) 2 1260 +/- 13.0 The amendment would change the setpoints to read as follows:
Number of Set point SRVs (psi g) 4 1135 +/- 34.1 4 1145 +/- 34.4 3 1155 +/- 34.7 Number of Setpoint SVs (psi g) 2 1260 +/- 37.8 The SR would also be revised to require that, following testing, the lift settings be within +/- 1%.
2.2.2 TS 3.1.7- "Standby Liquid Control (SLC) System" SR 3.1.7.8 currently requires that each SLC pump develop a flow rate of greater than or equal to 43 gallons per minute (gpm) at a discharge pressure of greater than or equal to 1255 psig.
The proposed amendments would change the discharge pressure from 1255 psig to 1275 psig.
2.3 Regulatory Requirements and Guidance 2.3.1 General Design Criteria (GDC)
The construction permit for PBAPS, Units 2 and 3, was issued by the Atomic Energy Commission (AEC) on January 31, 1968. As discussed in Appendix H to the PBAPS Updated Final Safety Analysis Report (UFSAR), during the construction/licensing process, both units were evaluated against the then-current AEC draft of the 27 GDC issued in November 1965.
On July 11, 1967, the AEC published for public comment, in the Federal Register (32 FR 10213), a revised and expanded set of 70 draft GDC (hereinafter referred to as the "draft GDC"). The Federal Register notice stated that "[t]hese General Design Criteria would not add any new requirements but are intended to describe more clearly present Commission requirements ... " Appendix H of the PBAPS UFSAR contains an evaluation of the design basis
of PBAPS, Units 2 and 3, against the draft GDC. The licensee concluded that PBAPS, Units 2 and 3, conforms to the intent of the draft GDC.
On February 20, 1971, the AEC published in the Federal Register (36 FR 3255), a final rule that added Appendix A to Title 10 of the Code of Federal Regulations (1 0 CFR) Part 50, "General Design Criteria for Nuclear Power Plants" (hereinafter referred to as the final GDC").
Differences between the draft GDC and final GDC included a consolidation from 70 to 64 criteria. As discussed in the NRC's Staff Requirements Memorandum for SECY-92-223, dated September 18, 1992 (ADAMS Accession No. ML003763736), the Commission decided not to apply the final GDC to plants with construction permits issued prior to May 21, 1971. At the time of promulgation of Appendix A to 10 CFR Part 50, the Commission stressed that the final GDC were not new requirements and were promulgated to more clearly articulate the licensing requirements and practice in effect at that time. Each plant licensed before the final GDC were formally adopted was evaluated on a plant-specific basis, determined to be safe, and licensed by the Commission.
The licensees for PBAPS, Units 2 and 3, have made changes to the facility over the life of the plant that may have invoked the final GDC. The extent to which the final GDC have been invoked can be found in specific sections of the UFSAR and in other plant-specific design and licensing basis documentation.
2.3.2 Plant-Specific Analysis The NRC staff reviewed the plant-specific analysis contained in NEDC-33533P to identify regulatory requirements and review criteria applicable to the proposed amendment. Further guidance is provided in specific sections of the NRC's Standard Review Plan (SRP)
(i.e., NUREG-0800). This section of the SE identifies the appropriate SRP section and summarizes the applicable regulatory requirements.
Vessel Overpressure Evaluation Chapter 2 of NEDC-33533P, "Vessel Overpressure Evaluation," describes the effect of the increased setpoint tolerance on the ASME-required reactor vessel overpressure protection. The NRC staff review guidance for overpressure analyses is provided in SRP Section 5.2.2, "Overpressure Protection" (ADAMS Accession No. ML070540076). The NRC's acceptance criteria are based on: (1) draft GDC-9, insofar as it requires that the RCPB be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage throughout its design lifetime; and (2) final GDC-31, insofar as it requires that the RCPB be designed with sufficient margin to assure that it behaves in a nonbrittle manner and that the probability of rapidly propagating fracture is minimized.
Thermal Limits Chapter 3 of NEDC-33533P, "Thermal Limits," describes the effect of the increased setpoint tolerance on the PBAPS transient fuel thermal limits analysis. Thermal limits are established to ensure that the fuel is designed and operated with sufficient margin to prevent fuel system damage as a result of anticipated operational occurrences (AOOs). The NRC staff review guidance for thermal limits assessment is provided in SRP Section 4.4, "Thermal and Hydraulic
Design" (ADAMS Accession No. ML070550060). NRC's acceptance criteria are based on final GDC-1 0, insofar as it requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits (SAFDLs) are not exceeded during any condition of normal operation, including the effects of AOOs.
Anticipated Transients Without Scram (A TWS) Mitigation Analysis Chapter 4 of NEDC-33533P, "ATWS Mitigation Analysis," describes the effect of the increased setpoint tolerance on the PBAPS ATWS analysis. Although the ATWS event is considered beyond design-basis, the plant must show that the increased SRV/SV setpoint tolerance would not cause the reactor pressure vessel to exceed the ASME Service Level C limit of 1500 psig under ATWS conditions. The NRC staff review guidance for ATWS mitigation is provided in SRP Section 15.8, "Anticipated Transients Without Scram" (ADAMS Accession No. ML070570008). ATWS is defined as an AOO followed by the failure of the reactor portion of the protection system specified in draft GDCs 14 and 15. The regulation in 10 CFR 50.62, "Requirements for reduction in risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plant," requires that:
- Each BWR have an alternate rod injection (ARI) system that is designed to perform its function in a reliable manner and be independent (from the existing reactor trip system) from sensor output to the final actuation device.
- Each BWR have a standby liquid control system with the capability of injecting into the reactor vessel a borated water solution with reactivity control at least equivalent to the control obtained by injecting 86 gpm of a 13 weight-percent sodium pentaborate decahydrate solution at the natural boron-1 0 isotope abundance into a 251-inch inside diameter reactor vessel. The system initiation must be automatic.
- Each BWR have equipment to trip the reactor coolant recirculation pumps automatically under conditions indicative of an ATWS.
High Pressure Systems Performance Chapter 6 of NEDC-33533P, "High Pressure Systems Performance," summarizes the evaluation of high pressure systems, including the high pressure coolant injection (HPCI) system, the reactor core isolation cooling (RCIC) system, and the SLC system The NRC staff review guidance review guidance for the HPCI system is provided in SRP Section 6.3, "Emergency Core Cooling System" (ADAMS Accession No. ML070550068). The NRC's acceptance criteria are based on: (1) draft GDCs 40 and 42, insofar as they require that protection be provided for engineered safety features (ESFs) against the dynamic effects that might result from plant equipment failures, as well as the effects of a loss-of-coolant accident (LOCA); and (2) draft GDC-37, 41, and 44, insofar as they require that a system provide abundant emergency core cooling so that fuel and clad damage that would interfere with the emergency core cooling function will be prevented.
The NRC staff review guidance for the RCIC system is provided in SRP Section 5.4.6, "Reactor Core Isolation Cooling System" (ADAMS Accession No. ML070540102). The NRC's
acceptance criteria are based on: (1) draft GDC-40 insofar as it requires that protection be provided for ESFs against dynamic effects; (2) draft GDC-4, insofar as reactor facilities shall not share systems or components unless it is shown safety is not impaired by the sharing; (3) draft GDC-37, insofar as it requires that ESFs be provided to back up the safety provided by the core design, the RCPB, and their protective systems; (4) draft GDC-51, and 57, insofar as they require that piping systems penetrating containment be designed with appropriate features as necessary to protect from an accidental rupture outside containment and the capability to periodically test the operability of the isolation valves to determine if valve leakage is within acceptable limits; and (5) 10 CFR 50.63, insofar as it requires that the plant withstand and recover from a station blackout (SBO) of a specified duration.
NRC staff review guidance for the SLC system is provided in SRP Section 9.3.5, "Standby Liquid Control System (BWR)" (ADAMS Accession No. ML070680186). The NRC's acceptance criteria are based on (1) draft GDC-27 and 28 insofar as they require that at least two independent reactivity control systems be provided, with both systems capable of making and holding the core subcritical from any hot standby or hot operating condition sufficiently fast to prevent exceeding acceptable fuel damage limits; (2) draft GDC-29 and 30, insofar as they require that at least one of the reactivity control systems be capable of making and holding the core subcritical under any condition sufficiently fast to prevent exceeding acceptable fuel damage limits; and (3) 10 CFR 50.62(c)(4), insofar as it requires that the SLC system be capable of reliably injecting a borated water solution into the reactor pressure vessel at a boron concentration, boron enrichment, and flow rate that provides a set level of reactivity control.
2.3.3 Pumps and Valves The NRC staff's review also included certain safety-related pumps and valves typically designated as Class 1, 2, or 3 under Section Ill of the ASME Code and within the scope of the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code). The NRC staff's review focused on the effects of the proposed TS changes on the required functional performance of these pumps and valves at the PBAPS. The review also covered any impacts that the proposed TS changes may have on the licensee's motor-operated valve (MOV) programs related to Generic Letter (GL) GL 89-10, "Safety-Related Motor-Operated Valve Testing and Surveillance", GL 96-05, "Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves". The NRC's acceptance criteria are based on: (1) draft GDC-1, insofar as it requires that structures, systems and components (SSCs) important to safety be designed and tested to quality standards commensurate with the importance of the safety functions to be performed; and (2) 10 CFR 50.55a(f), insofar as it requires that pumps and valves subject to that section must meet the inservice testing program requirements identified in that section. Specific review criteria are contained in SRP Section 3.9.3, "ASME Code Class 1, 2, and 3 Components, and Component Supports, and Core Support Structures" (ADAMS Accession No. ML12334A360) and Section 3.9.6, "Functional Design, Qualification, and lnservice Testing Programs for Pumps, Valves, and Dynamic Restraints" (ADAMS Accession No. ML070720041 ).
3.0 TECHNICAL EVALUATION
As discussed above, in support of the proposed amendment for PBAPS, Units 2 and 3, GEH performed plant-specific analyses and evaluations to address the six conditions/limitations
associated with the NRC's approval of LTR, NEDC-31753P. GEH documented the analyses and evaluations in report NEDC-33533P which was provided in the licensee's application dated June 10, 2013.
The NRC's review of the proposed amendment focused on the plant-specific analyses and evaluations provided in NEDC-31753P to determine if the six conditions/limitations associated with the NRC's approval of LTR, NEDC-31753P were met. The conditions/limitations are listed above in SE Section 2.1.
3.1 Overpressure Analysis The SRVs and SVs, along with the reactor protection system, provide overpressure protection for the RCPB during power operation. The NRC staff's review covered the effect that the requested setpoint tolerances and valve operability requirements would have on the results of the primary system overpressure analysis.
The licensee provided an overview of the vessel overpressure evaluation (NEDC-33533P, page 2-1 ). The limiting overpressure event, which the licensee stated was the postulated main steamline isolation valve closure with flux scram (MSIVF) event, was analyzed using the GEH transient analysis code ODYN. The licensee modeled all SRV/SV opening setpoints 3% above the nominal lift setpoint, at increased core flow (ICF) conditions. The licensee stated that the results were applicable to PBAPS, Unit 3, Cycle 18, and that the event is re-analyzed on a cycle-specific basis for both Units 2 and 3 in accordance with NRC-approved reload licensing methodology presented in NEDE-24011 P-A, Revision 16, "General Electric Standard Application for Reactor Fuel" (GESTAR II), October 2007 (non-publicly available, ADAMS Accession No. ML091340082).
The following summarizes the general sequence of events for the postulated MSIVF as stated on page 2-1 of NEDC-33533P:
The event is initiated with the Main Steam Isolation Valves (MSIVs) on all four steam lines rapidly closing. As a result of the reactor vessel isolation, the reactor pressure rises and collapses the core coolant voids, which in turn, increases the neutron flux, causing the reactor to scram on high neutron flux. System pressure subsequently reaches the ATWS Recirculation Pump Trip (RPT) setpoint. The reduction in coolant flow increases the core void fraction and accelerates the power reduction. The vessel pressure continues to rise until the SRV/SV opening pressures are reached, and the SRV/SV actuations terminate the pressurization transient.
Several acceptance criteria are applicable to the overpressure analysis. First, the analysis must demonstrate that the SVs have sufficient capacity to limit the pressure to less than 11 0% of the RCPB design pressure during the most severe AOO with reactor scram, with sufficient margin to account for uncertainties in the design and operation of the plant, provided that: (1) the reactor is operating at the power level that will produce the most severe over-pressurization transient; (2) all system and core parameters have values within the normal operating range, including uncertainties and TS limits that produce the highest anticipated pressure; and (3) the second safety-grade signal from the reactor protection system initiates the scram (reference
SRP 5.2.2, page 5.2.2-6). Second, the analysis must also demonstrate that the PBAPS reactor coolant system safety limit, established by TS 2.1.2, is not exceeded. The design pressure for the RPV is 1250 psi, and the peak acceptable pressure is 1.1 *1250, or 1375 psi. The safety limit requires the peak dome pressure to be less than or equal to 1325 psi.
The NRC staff evaluated the initial conditions addressed by items (1) and (2) above. These items ensure that the overpressure evaluation is sufficiently conservative. In large part, the use of an NRC-approved methodology addresses the initial conditions. The NRC staff reviewed GESTAR II and observed that the basis for the analytic assumptions employed in the ODYN analysis are addressed by the NRC staff SE approving TR NEDE-24154-P-A, "Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors," dated August 1986 (non-publicly available; ADAMS Accession No. ML102290145). The staff's review of ODYN concluded that the generic modeling approach was acceptable, and that the method produced an adequately conservative calculation of the peak reactor system pressure. Therefore, the NRC staff verified that the licensee's overpressure evaluation for the SRV/SV setpoint tolerance increase request was acceptable with respect to an adequately conservative selection of initial conditions, based on the fact that the analytic methods are generically approved.
The NRC staff did not identify, in its review of GESTAR II and NEDE-24154-P-A, a clear disposition for the increased core flow operating domain. Since the SE approving NEDC-31753P requires that the entire operating domain be considered (Condition/Limitation 5),
the NRC staff requested that the licensee provide information to address the selection of the increased core flow statepoint for the analysis of this event.
Prior to requesting additional information, the NRC staff performed an independent evaluation of the overpressure transient. The NRC staff determined that at the other extent of the power-flow operating domain, at the Maximum Extended Load Line Limit (MELLL) statepoint, characterized by significantly reduced recirculation flow, the steady-state initial void fraction could be higher at the fully licensed thermal power level, resulting in a greater void collapse and a higher pre-scram flux spike. The flux spike could result in the delivery of greater energy to the coolant, causing a faster pressurization and more severe result.
The NRC staff provided the above description of its evaluation in a request for additional information (RAI). The licensee responded to the RAI in its November 6, 2013, letter. The licensee acknowledged the accuracy of the staff evaluation, stating, "[w]hile the increased core flow (ICF) initial condition often results in the limiting overpressure transient, it is not uncommon for the low core flow ... initial condition to be comparable or even limiting due to the dynamic thermal-hydraulic and nuclear effects observed during the overpressure transient response."
The licensee's response described the event dynamics in greater detail than the staff evaluation, providing information about other mitigating phenomena that cause the ICF statepoint to be more limiting. The licensee also provided results of overpressure analyses for both ICF and MELLL statepoints, demonstrating that the entire full-power operating domain had, in fact, been considered in the analysis. Since the information addressed the staff concerns about the transient dynamics and demonstrated that the licensee had considered the entire operating domain, the NRC staff concluded that the licensee had addressed Condition/Limitation 5 of the SE approving NEDC-31753P satisfactorily.
The analysis assumes that the event is terminated by the flux scram. The flux scram is indirect; a prior signal would be generated by the MSIV position indication. Therefore, the NRC staff concluded that the licensee addressed Item (3) of the SRP acceptance criterion, which requires that the event be terminated by the second safety-grade signal from the reactor protection system.
The licensee provided the results of the analysis, and the NRC staff confirmed that the results met both the reactor dome pressure safety limit and 110% design pressure limit; hence, the NRC staff determined that the results were acceptable.
Condition/Limitation 3 of the SE approving NEDC-31753P requires that the analysis assure that it includes the number of SRVs/SVs required to be operable by the TSs. The licensee proposes to increase the number of operable SRVs/SVs required per TS LCO 3.4.3 from 11 to 12. On page 5 of Attachment 1 to the application dated June 10, 2013, the licensee stated that the analyses performed in NEDC-33533P assume that one SRV/SV is out-of-service. Since PBAPS has 13 SRV/SVs, the analysis assumes the operability of the 12 valves required per proposed TS LCO 3.4.3. Based on the fact that the analysis is based on the TS LCO operability requirements, the NRC staff determined that the licensee satisfied Condition/Limitation 3 with respect to the overpressure analysis.
In conclusion, the NRC staff determined that the licensee's overpressure analysis was acceptable for the following reasons: (1) the licensee has used NRC-approved computer codes and analytic methods to analyze the overpressure event; (2) the analysis has acceptably addressed the review guidance contained in SRP Section 5.2.2; (3) the results of the analysis meet the applicable acceptance criteria; and (4) the licensee's analysis appropriately addressed the licensed operating domain and reflects the proposed TS configuration. Based on these considerations, the NRC staff concluded that the overpressure analysis satisfies Limitations/Conditions 2, 3, and 5 of theSE approving NEDC-31753P, and that the proposed SRV/SV setpoint tolerance increase is hence acceptable with respect to overpressure protection, consistent with the requirements of draft GDC 9 and final GDC 31.
3.2 Thermal Limts Assessment Condition/Limitation 1 of the SE approving NEDC-31753P requires that the licensee provide transient analyses, using NRC-approved methods, of AOOs as described in NEDC-31753P utilizing a +/-3% setpoint tolerance for the safety mode of the SVs and SRVs. Rather than directly satisfy this condition/limitation, the licensee instead provided an assessment of the effect of the setpoint tolerance increase on the thermal limits. In particular, the licensee evaluated the effect of the setpoint tolerance increase on the minimum critical power ratio (MCPR) for the limiting transients analyzed in the reload licensing analysis, to confirm that the MCPR remained above the MCPR safety limit.
The NRC staff reviewed the thermal limits assessment to confirm that the assessment: (1) is accomplished with or based on acceptable analytic methods; and (2) demonstrates acceptable margins of safety from conditions which would lead to fuel damage during normal reactor operation and AOOs in light of the proposed SRV/SV setpoint tolerances and operability requirements.
The licensee's thermal limits assessment is provided in Chapter 3 of NEDC-33533P. The licensee considered the most limiting events for MCPR considerations at PBAPS, which included the load rejection without bypass (LRNBP), the turbine trip without bypass (TTNBP),
and the feedwater controller failure (FWCF) with increasing flow event. The licensee provided results from the PBAPS, Unit 3, Cycle 18 Supplemental Reload Licensing Report, which was simulated with no SRV out-of-service and the SRV/SV opening setpoints at +3% of the nominal setpoint. Since the SE approving NEDC-31753P requires licensees to assure that analyses supporting the requested setpoint tolerance increase reflect the TS operability requirements (Condition/Limitation 3), and the proposed PBAPS TSs would allow one SRV/SV to be out-of-service, the licensee stated the following:
Changing the SRV setpoint tolerance and/or the number of SRVs out-of-service could only affect the protection of the MCPR safety limit if it worsened the reactor pressure increase before the peak surface heat flux and the minimum MCPR occur.
The licensee also stated that opening the SRVs at 3% below the nominal setpoint would result in premature actuation of the SRVs, reducing the rate of vessel pressurization and decreasing the rate of void collapse, ultimately resulting in a smaller change in critical power ratio.
The licensee provided a table showing the timing of the minimum MCPR and the first SRV lift for the limiting transients that had been analyzed for PBAPS, Unit 3, Cycle 18 (i.e., no SRV out-of-service, +3% lift setpoint tolerance), and concluded that the analyses demonstrate that the PBAPS operation with one SRV out-of-service and an increased lift setpoint is acceptable.
The NRC staff performed an independent evaluation of the limiting AOOs considered by the licensee. In the case of the fast transients, the nuclear flux, void reactivity, reactor vessel water level, and steam flow are all expected to oscillate with periods of 1-2 seconds immediately following the initiating event, and the system pressure may continue increasing (potentially causing additional void collapse) for a short period of time following the actuation of the SRVs.
These trends would all suggest that the transient may not be returning to a stable condition at the time the analyses performed by the licensee indicate that maximum heat flux has been achieved.
The NRC staff summarized the issues described above in an RAI, requesting that the licensee provide additional information to justify its thermal limits assessment, or alternatively, satisfy Conditions 1, 3, and 5 of the approving SE for NEDC-31753P directly by providing analyses that reflect the proposed TS operability requirements and SRV/SV setpoint tolerance. In its response letter dated November 6, 2013, the licensee provided a more detailed description of the fast transient dynamics, which explained why the potential for flux increase, which the NRC identified, would be mitigated. This additional consideration identified by the licensee is the reactor scram, which occurs prior to SRV lift. The scram will ensure, as asserted in NEDC-33533P, that the peak surface heat flux occurs early in the transient, such that the MCPR safety limit is not challenged at the SRV opening time. Based on this additional information, the NRC determined that the licensee's evaluation was acceptable, even considering the different out-of-service conditions reflected in the proposed TS and the Unit 3, Cycle 18 analyses.
In addition to the issues that the NRC staff identified in its independent evaluation, the NRC staff also concluded that the information presented did not characterize the limited transients presented in the context of the entire suite of analyses performed, including the extent of the licensed operating domain and the equipment operability options permitted for PBAPS, including various scram speeds. Therefore, while the licensee had provided a limited assessment based on previously analyzed limiting events, the licensee had not demonstrated that these events would be reasonably unaffected by the requested setpoint tolerance increase and valve operability requirements, nor had the licensee demonstrated that other, non-limiting events would become limiting with the SRV/SV changes explicitly analyzed.
The licensee presented results for a current Unit 3 PBAPS operating cycle, and stated that the events are re-analyzed on a cycle-specific basis for both Units 2 and 3. Since the licensee's assessment is based on a present analysis that does not reflect the proposed SRV/SV operability requirements, and Condition/Limitation 3 of theSE approving NEDC-31753P requires assurance that the analysis reflects the proposed TS operability requirements for SRVs and SVs, the NRC staff requested in an RAI that the licensee explain how this assurance will be provided on a cycle-specific basis for both Units 2 and 3. In its response to the RAI, the licensee stated that the limiting AOOs would be re-analyzed on a cycle-specific basis reflecting the actual TS-required plant configuration and equipment flexibility options. This information provides additional assurance that the acceptability of the proposed SRV and SV lift setpoint tolerances will be confirmed in accordance with NRC-approved reload safety analysis methodology, and that the full complement of equipment flexibility options will be reflected in the safety analyses.
The NRC staff determined that the licensee's decision not to analyze a reduced SRV/SV lift setpoint was acceptable based on the discussion provided in NEDC-33533P. Because an earlier SRV/SV opening would reduce the rate of vessel pressurization and either reduce the rate of void collapse or allow the formation of new voids, either of which would reduce the amount of moderator present and tend to suppress the flux, the NRC staff finds that the net effect would lessen the overall severity of the transient.
Based on the evaluation provided by the licensee, the NRC staff determined that the thermal limits, specifically the MCPR limits, would not be adversely affected by the proposed SRV/SV lift setpoint tolerance increase, with 12 SRVs/SVs required to be operable. Therefore, the NRC staff finds that the licensee has acceptably addressed Condition/Limitation 1 of theSE approving NEDC-31753, and that the proposed SRV/SV setpoint tolerances and operability requirements are acceptable with respect to the effects on thermal limits, consistent with the requirements of final GDC-1 0.
3.3 ATWS Mitigation Analysis Condition/Limitation 2 of theSE approving NEDC-31753P requires analysis of the design basis overpressure event using the increased tolerance limit for the SRV/SV setpoints to confirm that the vessel pressure does not exceed ASME pressure vessel upset limits. Although an ATWS event is technically considered beyond the PBAPS design basis, ATWS mitigation must still ensure that the vessel pressure does not exceed ASME Service Level C limits under the conditions associated with the most severe ATWS event. For analytic purposes, Service Level C limits are commonly accepted as 120% of the vessel design pressure, or 1500 psi.
The NRC staff's review was conducted to ensure that the mitigating capabilities for protection from ATWS events are not unacceptably affected by the proposed SRV/SV setpoint tolerance limits and operability requirements.
The licensee analyzed the main steamline isolation valve closure (MSIVC) and pressure regulator failure open (PRFO) to maximum demand ATWS events. The events were analyzed using the ODYN transient analysis code with the reactor operating at the full-power MELLL boundary. In each event, the reactor protection system is assumed to fail, and the plant shutdown is accomplished through SLC system actuation. A more immediate power reduction occurs due to an automatic recirculation pump trip. The licensee stated that the PRFO was the limiting ATWS event, and that the results of the limiting event, assuming a single SRV out-of-service, were less than the 1500 psi acceptance criterion.
The NRC staff observed that, while the ASME overpressure AOO analysis considered initial conditions in the increased core flow extent of the operating domain, this analysis conversely considered conditions at the MELLL boundary. The NRC staff requested additional information to confirm that the licensee appropriately considered ATWS mitigation throughout the entire operating domain, as Condition/Limitation 5 of the SE approving NEDC-31753P would require.
In response to the RAI, in the letter dated November 6, 2013, the licensee provided a disposition justifying the selection of this statepoint. The disposition reflected consideration of the ICF extent of the operating domain, and explained why the MELLL statepoint was bounding during an ATWS analysis.
Since the licensee explicitly addressed the limiting ATWS events by analyzing them using the NRC-approved ODYN code, and since the results of the analysis were less than the 1500 psi acceptance criterion, the NRC staff determined that the licensee's disposition for ATWS mitigation was acceptable. Therefore, the licensee has addressed Condition/Limitation 2 of the SE approving NEDC-31753P with respect to an ATWS event. The NRC staff concludes that Condition/Limitation 3 of theSE approving NEDC-31753P is satisfied because the analysis reflects the assumption that one SRV is out-of-service, consistent with the proposed TS operability requirements. The licensee's RAI response addressed Condition/Limitation 5 of the SE approving NEDC-31753P, since it explained how the MELLL analysis reflects the limiting extent of the allowed operating domain. In conclusion, the licensee has addressed Conditions/Limitations 2, 3, and 5 with respect to ATWS mitigation analysis.
3.4 High Pressure Systems Performance Condition/Limitation 4 of theSE approving NEDC-31753P requires the re-evaluation of the performance of high pressure systems (e.g., pump capacity, discharge pressure, etc.), motor-operated valves, and vessel instrumentation and associated piping considering the 3% tolerance limit. In Chapter 6 of NEDC-33533P, "High Pressure Systems Performance," the licensee identified the HPCI, RCIC and SLC systems for re-evaluation.
The HPCI system, which is part of the emergency core cooling system (ECCS), is designed to provide sufficient core cooling to prevent excessive fuel cladding temperature in the event of a small-break LOCA that does not depressurize the reactor quickly enough to permit timely operation of the low pressure ECCS. The HPCI system accomplishes this function by injecting
coolant makeup water into the pressure vessel with a turbine driven pump. The HPCI system also serves as a backup to the RCIC system to maintain the nuclear boiler in the standby condition in the event the vessel becomes isolated from the main condenser and feedwater makeup flow.
The RCIC system provides makeup water to the reactor pressure vessel whenever the vessel is isolated from the main condenser and feedwater make up. The RCIC system can also provide makeup water during shutdown whenever the normal water supply is unavailable. RCIC uses a turbine driven pump to maintain adequate reactor vessel water level.
The SLC system consists of a boron solution storage tank, two positive displacement pumps, two explosive valves that are provided in parallel for redundancy, and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel. The SLC system is manually initiated from the main control room, as directed by the emergency operating procedures, if the operator believes the reactor cannot be shut down, or kept shut down, with the control rods.
The NRC's review of the HPCI and RCIC systems is provided in SE Section 3.4.1. The NRC review of the SLC system is provided in SE Section 3.4.2.
3.4.1 HPCI and RCIC Systems Section 6.1 of NEDC-33533P provides this assessment for the HPCI and RCIC systems. With regard to the common functional requirements for HPCI and RCIC, the report states as follows:
Both systems are required to provide injection into the reactor pressure vessel at the lowest group of SRV setpoints (including drift).
Consistent with the applicable GDCs set forth in theSE Section 2.3.2, the evaluation considers the pump flow and head, system boundary components, each system's pump and turbine, instrumentation, motor-operated valves and the transient effects of pump start.
The licensee evaluated the pump flow and head requirements for the HPCI and RCIC systems in light of the requested setpoint tolerance increase. This evaluation concluded that the flow requirements, required and available net positive suction head, and calculated pressure losses due to pipe friction would remain unchanged. The NRC staff finds this assessment acceptable, since the HPCI and RCIC flow requirements are based on the decay heat load, which is unaffected by the requested setpoint tolerance increase. The evaluation also considered the existing conservative, bounding assumptions for dynamic and static head losses to determine existing total dynamic head requirements. The licensee stated that analyses of the piping configuration for both systems was performed to produce more accurate results than the previous analyses, and that these results confirmed that existing total developed head requirement remains both bounding and conservative for both systems. The evaluation also concluded that the HPCI turbine speed remains adequate, but concluded that an RCIC speed turbine increase may be necessary. However, consistent with the discussion in Section 6.1.2 of NEDC-33533P, the licensee reviewed design documentation for the RCIC system, identified existing margin in the RCIC total developed head requirements, and concluded that the turbine speed change was unnecessary. Based on its review of the licensee's evaluation, the NRC
staff concluded that crediting existing design margins and performing more accurate analyses of the HPCI and RCIC system total developed head requirements has demonstrated that the pump flow and head requirements are acceptable in consideration of the requested SRV/SV lift setpoint tolerance increase.
As discussed in Section 6.1.5 of NEDC-33533P, instrument specifications have been reviewed for application to the HPCI and RCIC required pressure increase. Because the maximum reactor pressure and temperature design parameters remain unchanged, and exceed those conditions required for the SRV setpoint tolerance increase, there is no change to the design pressure requirements of the instrumentation. In addition, because there is no increase in flow for either make up or steam, no change is required for leak detection or flow instruments. The NRC staff concludes that the licensee has adequately evaluated the proposed change with respect to the impact on instrumentation.
As discussed in Attachment 1 to the application dated June 10, 2013, the licensee evaluated the impact on MOVs due to the potential for increased reactor vessel and system pressure as a result of the proposed increase in SRV/SV tolerance. The licensee stated that proposed change was evaluated in accordance with the GL 89-10/96-05 program, NRC Bulletin 85-03, and NEDC-31322, "BWR Owners' Group Report on the Operational Design Basis of Selected Safety Related Motor-Operated Valves." The licensee determined that the impact on MOV operation is acceptable. The MOV methodology uses the reactor pressure corresponding to the spring setpoint of the SRV/SV with the lowest nominal setpoint for differential pressure determinations. Therefore, the MOV methodology is not impacted by the SRV/SV setpoint tolerance revision. The NRC staff concludes that the licensee has adequately evaluated the proposed change with respect to the impact on MOVs.
3.4.2 SLC System Section 6.2 of NEDC-33533P provides this assessment for the SLCS. With regard to the functional requirements for the SLCS, the report states as follows:
The SLCS [SLC system] is designed to shut down the reactor from rated power conditions to cold shutdown in the postulated situation that some or all of the control rods cannot be inserted. The SLCS is also required to inject the sodium pentaborate solution to control reactor power in the event of an ATWS event to meet the injection rate requirements of 10 CFR 50.62(c)(4).
The proposed SRV/SV lift setpoint tolerance increase has the potential to affect SLC system performance in two noteworthy ways: {1) the peak pressure against which the system must inject may increase; and (2) the SLC system discharge piping and interconnected valves, including relief valves, must remain functional when considering line losses and pump pulsations. Additional information concerning Item (2) above, is discussed in NRC Information Notice (IN) 2001-13, "Inadequate Standby Liquid Control System Relief Valve Margin," dated August 10,2001 (ADAMS Accession No. ML012210146). The NRC staff reviewed the licensee's assessment to confirm that the proposed SLC system requirements provide for acceptable operation with the proposed SRV/SV lift setpoint tolerance increase.
In its evaluation of SLC system performance, the licensee considered the design margin available in the system, in terms of the required injection pressure. This evaluation considered the SLC system relief valve setpoint, the pressure fluctuations associated with the positive displacement SLC pumps, SLC system line losses associated with single pump operation, the calculated maximum lower plenum pressure at a time when SLC system injection is credited during A TWS conditions, and the maximum lower plenum pressure that would prevent the SLC system relief valve from recirculation. The licensee's evaluation concluded that the SRV tolerance relaxation would not adversely affect the ability of the SLC system to meet its design requirement.
The licensee did determine, however, that the SLC pump discharge pressure surveillance requirement (i.e., TS SR 3.1. 7.8) would need to be increased from 1255 psig to 1275 psig. The licensee determined that the 20 psig increase would correspond to the approximate 20 psig increase in the maximum allowable opening pressure associated with the lowest setpoint bank of SRVs at PBAPS.
The NRC staff reviewed the licensee's evaluation of SLC system design margins by considering it against the conditions identified in IN 2001-13. The NRC staff determined that the license's evaluation accounted for all of the issues described in the IN, namely, the discharge line losses associated with the required pump operating configuration, the SLC system relief valve setpoints, and the required SLC system injection pressure during the limiting A TWS. Since the licensee accounted for these issues, the NRC staff concluded that the licensee has acceptably addressed the potential impact that an increased SRV/SV setpoint tolerance could have on SLC system operation.
In its review of the licensee's proposed revision toTS SR 3.1. 7.8, which would increase the required discharge pressure developed by the SLC system pump from 1255 psig to 1275 psig, the NRC staff observed that the only parameter changing in the SLC system evaluation, due to the SRV/SV setpoint tolerance increase, is the maximum lift pressure of the SRVs/SVs. The 20 psig increase reflects the increased maximum lift setpoint tolerance of the SRVs/SVs.
Therefore, the NRC staff determined that the proposed change to the SR is acceptable since it accounts for the proposed SRV/SV setpoint tolerance increase.
3.4.3 Conclusion Based on the discussion in SE Sections 3.4.1 and 3.4.2, the NRC staff concludes that the licensee re-evaluation of high pressure systems acceptably addresses Condition/Limitation 4 of theSE approving NEDC-31753.
3.5 Containment Response during LOCAs and SRV Hydrodynamic Loads Condition/Limitation 6 of theSE approving NEDC-31753P requires the evaluation of the effect of the 3% tolerance limit on the containment response during LOCAs and the hydrodynamic loads on the SRV discharge lines and containment. Chapter 5, "ECCS/LOCA Performance Evaluation," provided the associated evaluation.
3.5.1 Containment Response during LOCAs The licensee stated that the LOCA analysis was performed for PBAPS using the NRC-approved SAFERJGESTR LOCA application methods. The licensee considered the effects of the proposed change on DBA LOCA containment pressure and temperature and on the DBA LOCA suppression pool and wetwell pressure. The licensee determined the proposed change had no effect on these parameters.
The licensee evaluated the impact of the proposed change on small steam line breaks to determine the drywell temperature with respect to environmental qualification of equipment.
The licensee determined that the proposed increase in the SRV/SV setpoint tolerance would have no effect on the environmental qualification of equipment. The licensee also determined that the proposed change in SRV/SV setpoint tolerance would have no impact on the existing analysis for intermediate and small line break accidents.
The licensee also concluded that the DBA LOCA hydrodynamic loads are not affected due to the proposed change in SRV/SV setpoint tolerance.
3.5.2 SRV Hydrodynamic Loads As discussed in Section 5.6 of NEDC-33533P, an increase in the SRV setpoint tolerance from 1% to 3% may increase the pressure at which the SRVs open. The resulting increase in SRV opening pressure could increase the SRV flow rate, which could result in an increase in the hydrodynamic loads. As such, further evaluations were required to determine the impact of the proposed setpoint tolerance increase on the SRV hydrodynamic loads. These evaluations were summarized in Attachment 1 to the application dated June 10, 2013. The licensee stated in that the SRV discharge piping has been re-analyzed for the effects of increasing the SRV setpoint tolerance in combination with the existing design basis loads. The licensee stated that, in order to accomplish this analysis, new SRV forcing functions based on the revised lift tolerance were applied to the current piping models. The licensee found that the resulting stresses remain less than the original ANSI/ASME Code allowable stress values and, therefore, was acceptable.
In addition, restraint loads and deflections were evaluated against the current restraint designs.
This evaluation determined that the existing restraints can accommodate the proposed increase in the SRV tolerance. The evaluation included the SRV T -Quencher supports, which are considered anchors in the piping analysis. The licensee's evaluation found that the existing pipe support designs, including the T-Quencher supports and their components, can accommodate the effects of the proposed increase in the SRV tolerance and remain structurally adequate to perform their intended design function. Therefore, the licensee determined that the support design functions were acceptable.
The NRC staff concludes that the licensee provided reasonable assurance that the affected components will remain structurally adequate to perform their intended functions with respect to any change in hydrodynamic loads following implementation of the proposed amendments.
3.5.3 Conclusion Based on the discussion in SE Sections 3.5.1 and 3.5.2, the NRC staff concludes that the licensee evaluation of the effect of the 3% tolerance limit on the containment response during LOCAs and the hydrodynamic loads on the SRV discharge lines and containment acceptably addresses Condition/Limitation 6 of theSE approving NEDC-31753.
3.6 Proposed TS Changes Based on the discussion in SE Sections 3.1 through 3.5, the NRC staff concludes that the licensee has acceptably addressed all 6 Conditions/Limitations of the NRC's SE approving NEDC-31753. As such, the staff concludes that the proposed change to SR 3.4.3.1 to reflect an increase in the setpoint tolerance for the SRVs and SVs from +/-1% to +/-3% is acceptable.
The licensee has also proposed to revise SR 3.4.3.1 to require that, following testing, the lift settings be within +/-1 %. This proposed TS requirement is consistent with the NRC staff's SE which approved NEDC-31753. As such, this change is acceptable.
As discussed in SE Section 3.1, the licensee's analysis in NEDC-33533P assumes that 1 SRV/SV is out-of-service. Since PBAPS has 13 SRV/SVs, the analysis assumes the operability of the 12 valves. The proposed change to LCO 3.4.3 to increase the required number of operable SRVs and SVs from 11 to 12 is consistent with the analysis in NEDC-33533P. As such, the NRC staff concludes the proposed change to LCO 3.4.3 is acceptable.
As discussed in SE Section 3.4.2, the NRC staff concludes that the proposed revision toTS SR 3.1.7.8, which would increase the required discharge pressure developed by the SLC system pump from 1255 psig to 1275 psig, is acceptable.
Based on the above, the NRC staff concludes that the proposed amendment is acceptable.
The licensee's application dated June 10, 2013, provided revised TS Bases pages to be implemented with the associated TS changes. These pages were provided for information only and will be revised by the licensee in accordance with the TS Bases Control Program.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Pennsylvania State official was notified of the proposed issuance of the amendments. 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 and change surveillance requirements. 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 (78 FR 78406). 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) there is reasonable assurance that such activities will be conducted in compliance with the Commission's 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.
Principal Contributors: B. Parks J. Billerbeck A. Tsirigotis R. Ennis Date: May 5, 2014
May 5, 2014 Mr. Michael J. Pacilio President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 AND 3 -ISSUANCE OF AMENDMENTS RE: SAFETY RELIEF VALVE AND SAFETY VALVE LIFT SETPOINT TOLERANCE (TAC NOS. MF1970 AND MF1971)
Dear Mr. Pacilio:
The Commission has issued the enclosed Amendment Nos. 290 and 293 to Renewed Facility Operating License Nos. DPR-44 and DPR-56 for Peach Bottom Atomic Power Station, Units 2 and 3. These amendments consist of changes to the Technical Specifications (TSs) and Facility Operating Licenses in response to your application dated June 10, 2013, as supplemented by letter dated November 6, 2013.
The amendments revise the TSs to: (1) increase the allowable as-found safety relief valve (SRV) and safety valve (SV) lift setpoint tolerance from +/- 1% to +/- 3%; (2) increase the required number of operable SRVs and SVs from 11 to 12; and (3) increase the Standby Liquid Control System pump discharge pressure from 1255 pounds per square inch gauge (psig) to 1275 psig.
A copy of the safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's Biweekly Federal Register notice.
Sincerely, IRA/
Richard B. Ennis, Senior Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-277 and 50-278
Enclosures:
- 1. Amendment No. 290 to Renewed DPR-44
- 2. Amendment No. 293 to Renewed DPR-56
- 3. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:
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