ML17249A151

Peach Bottom Atomic Power Station, Unit 2 - Issuance of Amendment Safety Relief Valve and Safety Valve Operability for Cycle 22 (CAC No. MF9705; EPID L-2017-LLA-0229)

ML17249A151
Person / Time
Site:	Peach Bottom
Issue date:	10/25/2017
From:	Richard Ennis Plant Licensing Branch 1
To:	Bryan Hanson Exelon Nuclear
Ennis R, NRR/DORL/LPL1, 415-1420
References
CAC MF9705, EPID L-2017-LLA-0229
Download: ML17249A151 (19)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Bryan C. Hanson October 25, 2017 President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

PEACH BOTTOM ATOMIC POWER STATION, UNIT 2 -ISSUANCE OF AMENDMENT RE: SAFETY RELIEF VALVE AND SAFETY VALVE OPERABILITY FOR CYCLE 22 (CAC NO. MF9705, EPID L-2017-LLA-0229)

Dear Mr. Hanson:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 315 to Renewed Facility Operating License No. DPR-44 for Peach Bottom Atomic Power Station, Unit 2. This amendment consists of changes to the technical specifications in response to your application dated May 19, 2017, as supplemented by letter dated August 29, 2017. The amendment revises the technical specifications to decrease the number of safety relief valves and safety valves required to be operable when operating at a power level less than or equal to 3,358 megawatts thermal.

This change is applicable only to the current Peach Bottom Atomic Power Station, Unit 2, Cycle 22, which is scheduled to end in October 2018. A copy of our Safety Evaluation is also enclosed.

Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Docket No. 50-277

Enclosures:

1. Amendment No. 315 to Renewed DPR-44 2. Safety Evaluation cc w/

Enclosures:

Distribution via Listserv Sincerely, Richard B. Ennis, Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation 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. 315 Renewed License No. DPR-44 1. The U.S. 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 May 19, 2017, as supplemented by letter dated August 29, 2017, 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. 315, 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 5 days.

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance:

October 2 5, 2O1 7 FOR THE NUCLEAR REGULATORY COMMISSION James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation ATTACHMENT TO LICENSE AMENDMENT NO. 315 PEACH BOTTOM ATOMIC POWER STATION, UNIT 2 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 Insert 3 Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change. Remove 3.4-8 Insert 3.4-8 Paqe 3 (5) Exelon Generation

Company, pursuant to the Act and 1 O 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 I: 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 3951 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 315, 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 plans1, 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 1 O 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 1 O CFR 50.90 and 10 CFR 50.54(p).

The Exelon Generation Company CSP was approved by License Amendment No. 281 and modified by Amendment No. 301. (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. 315 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety Relief Valves (SRVs) and Safety Valves (SVs) SRVs and SVs 3.4.3 LCO 3.4.3 The safety function of 13 valves (any combination of SRVs and SVs) shall be OPERABLE.

---

NOTE---------------------------

The safety function of 12 valves (any combination of SRVs and SVs) are required to be 3358 MWt during operating cycle 22. APPLICABILITY:

MODES 1, 2, and 3. ACTIONS CONDITION A. One or more required SRVs or SVs inoperable.

PBAPS UNIT 2 A.l A.2 REQUIRED ACTION COMPLETION TIME Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> 3.4-8 Amendment No. 315 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 315 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-44 EXELON GENERATION

COMPANY, LLC PSEG NUCLEAR LLC PEACH BOTTOM ATOMIC POWER STATION, UNIT 2 DOCKET NO. 50-277

1.0 INTRODUCTION

By application dated May 19, 2017 (Reference 1 ), as supplemented by letter dated August 29, 2017 (Reference 2), Exelon Generation

Company, LLC (Exelon, the licensee) submitted a license amendment request (LAR) for Peach Bottom Atomic Power Station (PBAPS),

Unit 2. The amendment would revise the Technical Specifications (TSs) to decrease the number of safety relief valves (SRVs) and safety valves (SVs) required to be operable when operating at a power level less than or equal to 3,358 megawatts thermal (MWt). This change would be in effect for the current PBAPS, Unit 2, Cycle 22, which is scheduled to end in October 2018. The supplemental letter dated August 29, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally

noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC or the Commission) staff's original proposed no significant hazards consideration determination as published in the Federal Register on July 5, 2017 (82 FR 31094). 2.0 REGULATORY EVALUATION 2.1 General Design Criteria 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 General Design Criteria (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"). Appendix Hof. 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. Enclosure 2 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 O CFR) Part 50, "General Design Criteria for Nuclear Power Plants" (hereinafter referred to as the '1inal GDC"). Differences between the draft GDC and final GDC included a consolidation from 70 to 64 criteria.

As discussed in the NRG Staff Requirements Memorandum for SECY-92-223, dated September 18, 1992 (Reference 3), 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 1 O 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, Unit 2, 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.

The NRG staff identified the following GDC as being applicable to this LAR:

• Draft GDC 6, "Reactor Core Design (Category A)," which requires that the reactor core be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits, which has been stipulated and justified.

• Draft GDC 9, "Reactor Coolant Pressure Boundary (Category A)," which requires that the reactor coolant pressure boundary (RCPB) be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage throughout its design lifetime.

• Draft GDC 14, "Core Protection Systems (Category B)," which requires that core protection

systems, together with associated equipment, be designed to act automatically to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits.

• Draft GDC 15, "Engineered Safety Features Protection Systems (Category B)," which requires that protection systems be provided for sensing accident situations and initiating the operation of necessary engineered safety features (ESFs).

• Draft GDC-29,"Reactivity Shutdown Capability (Category A)," which requires that at least one of the reactivity control systems be provided capable of making the core subcritical under any condition (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage limits. Shutdown margins greater than the minimum worth of the most effective control rod when fully withdrawn shall be provided.

• Draft GDC-33, "Reactor Coolant Pressure Boundary Capability (Category A)," which requires that the RCPB be capable of accommodating static and dynamic loads resulting from an inadvertent and sudden release of energy to the coolant.

• Draft GDC 37, "Engineered Safety Features Basis for Design (Category A)," which requires, in part, that ESFs be provided to back up the safety provided by the core design, the RCPB, and their protective systems.

• Draft GDC 40, "Missile Protection (Category A)," which requires that protection for engineered safety features shall be provided against dynamic effects and missiles that might result from plant equipment failures.

• Draft GDC 41, "Engineered Safety Features Performance Capability (Category A)," which requires, in part, that ESFs such as emergency core cooling and containment heat removal systems provide the required safety function, assuming a failure of a single active component.

• Draft GDC 42, "Engineered Safety Features Components Capability (Category A)," which requires that ESFs be designed so that the capability of each component and system to perform its required function is not impaired by the effects of a loss-of-coolant accident (LOCA).

• Draft GDC 44, "Emergency Core Cooling Systems Capability (Category A)," which requires that each emergency core cooling system (ECCS) and the core shall be designed to prevent fuel and clad damage that would interfere with the emergency core cooling function.

• Final GDC 15, "Reactor coolant system design,"

which requires the reactor coolant system and associated auxiliary,

control, and protection systems be designed with sufficient margin to assure that the design conditions of the RCPB are not exceeded during any condition of normal operation, including anticipated operational occurrences (AOOs).

• Final GDC 31, "Fracture prevention of reactor coolant pressure boundary,"

which requires the RCPB be designed with sufficient margin to assure that when stressed under operating, maintenance,

testing, and postulated accident conditions, (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized.

2.2 Technical Specification Requirements In 1 O CFR 50.36, "Technical specifications,"

the NRC established its regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following categories:

(1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements; (4) design features; and (5) administrative controls.

The regulation does not specify the particular requirements to be included in a plant's TSs. As discussed in 1 O CFR 50.36(c)(2),

LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

When LCOs are not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the LCOs can be met. 2.3 Other Regulatory Requirements The NRC staff identified the following regulatory requirements as being applicable to this LAA: 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors,"

which, in part, establishes standards for the calculation of ECCS accident performance and acceptance criteria for that calculated performance. 1 O CFR 50.62, "Requirements for reduction of risk from anticipated transients without scram (A TWS) events for light-water-cooled nuclear power plants,"

which requires, in part that: (1) Each boiling water reactor (BWR) must have an alternate rod injection 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. (2) Each BWR must have a standby liquid control (SLC) system with the capability of injecting into the reactor pressure vessel a borated water solution with reactivity control at least equivalent to the control obtained by injecting 86 gallons per minute of a 13 weight-percent sodium pentaborate decahydrate solution at the natural boron-1 O isotope abundance into a 251-inch inside diameter reactor pressure vessel for a given core design. (3) Each BWR must have equipment to trip the reactor coolant recirculating pumps automatically under conditions indicative of an ATWS. ATWS is defined as an AOO followed by the failure of the reactor trip portion of the protection system. 10 CFR 50.63, "Loss of all alternating current power," paragraph (a)(2), which requires, in part that: The reactor core and associated

coolant, control, and protection systems, including station batteries and any other necessary support systems, must provide sufficient capacity and capability to ensure that the core is cooled and appropriate containment integrity is maintained in the event of a station blackout for the specified duration.

3.0 TECHNICAL EVALUATION 3.1 Proposed TS Change The LCO for TS 3.4.3, "Safety Relief Valves (SRVs) and Safety Valves (SVs)," currently requires that the safety function of 13 valves (any combination of SRVs and SVs) be operable in Modes 1, 2, and 3. The proposed amendment would modify the LCO to add the following note: The safety function of 12 valves (any combination of SRVs and SVs) are required to be OPERABLE

5 3358 MWt during operating cycle 22. The current licensed thermal power level is 3,951 MWt. As such, the proposed note would be applicable when the plant is operating at less than or equal to 85 percent of the currently licensed power level (i.e., 3,358/3,951

). 3.2 System Description PBAPS, Unit 2, is a BWR of General Electric BWR/4 design with a Mark I containment.

The nuclear boiler system transports the steam generated in the reactor pressure vessel (RPV) through the primary containment by means of a piping system (consisting of four 26-inch main steam lines with two pneumatically

operated, globe type isolation valves in each steam line) from the RPV nozzles to the outboard main steam isolation valves (MSIVs).

Between the RPV and the MSIVs, three SVs and 11 dual function SRVs are mounted on the steam lines which, in conjunction with reactor scram, assist in limiting peak pressure in the primary system during plant transient conditions.

The design pressure of the reactor vessel and RCPB is 1,250 pounds per square inch gauge (psig). The acceptance limit for pressurization events is the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) allowable peak pressure of 1,375 psig (11 O percent of design value). The SVs and SRVs are designed to meet the requirements for reactor vessel overpressure protection to conform to the ASME Code, Section Ill, Article 9. The nuclear system pressure relief system is designed with 11 SRVs with opening setpoints of 1, 135 psig, 1, 145 psig, and 1, 155 psig, and 3 SVs with opening setpoints of 1,260 psig. 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. Each of the 11 SRVs discharge steam through a discharge line to a point below the minimum water level in the suppression pool. The three SVs discharge steam directly to the drywell.

In the depressurization mode, each SRV is opened by a pneumatic

actuator, which opens the second stage disc. The main valve then opens as described above for the safety mode. The depressurization mode is initiated either manually by the operator or automatically by the automatic depressurization system. Unlike the safety mode, the depressurization mode does not rely on the pilot stage and is independent of the bellows.

The depressurization mode provides a method for depressurization of the RCPB. All 11 of the SRVs function in the safety mode and have the capability to operate in the depressurization mode via manual actuation.

Five of the SRVs are allocated to the automatic depressurization system. The safety objective of the pressure relief system is to prevent overpressurization of the nuclear system; this protects the RCPB from failure, which could result in the uncontrolled release of fission products.

In addition, the automatic depressurization feature of the pressure relief system acts in conjunction with ECCS for reflooding the core. This protects the reactor fuel cladding from failure due to overheating.

3.3 NRC Staff Evaluation As described in the LAR, the basis for the licensee to propose the TS change is one of the 11 SRVs (the 2K SRV) becoming inoperable in the current cycle of operation.

A second inoperable valve would require a plant shutdown in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in accordance with TS LCO 3.4.3, Required Action A.1. If the proposed TS change is approved, the licensee could avoid an unnecessary reactor shutdown and avoid the reactivity control challenges that can occur during startup from a Hot Shutdown condition following a short shutdown.

The licensee identified the safety analyses that are potentially affected by the proposed TS change. In summary, the affected analysis areas as identified in the LAR include ASME overpressure protection, A TWS, AOO, ECCS/LOCA performance, and high pressure system performance.

The NRC staff reviewed the licensee's evaluation on these areas below against the regulatory requirements as identified in Section 2.0 of this safety evaluation. 3.3.1 Development of Allowable Condition with One Additional SRV or SV Out of Service (i.e., Two SRV/SV Out of Service (2 SRV/SVOOS))

Overpressure protection for the RCPB during power operation is provided by SRVs and SVs and the reactor protection system. The N RC'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; (2) final GDC-15, insofar as it requires that the reactor coolant system and associated auxiliary,

control, and protection systems be designed with sufficient margin to assure that the design conditions of the RCPB are not exceeded during any condition of normal operation, including AOOs; and (3) final GDC-31, insofar as it requires that the RCPB be designed with sufficient margin to assure that it behaves in a non-brittle manner and that the probability of rapidly propagating fracture is minimized.

As discussed in Attachment 1 to the licensee's application dated May 19, 2017, for the PBAPS, Unit 2, current Cycle 22 operation, the results of the reload licensing analysis (from the Supplemental Reload Licensing Report) indicate that the calculated peak reactor dome pressure due to ASME overpressure compliance analysis is 1,321 psig. That pressure has a 4 pounds per square inch (psi) margin to the TS safety limit of 1,325 psig. The limiting transient is the MSIV's closure with scram on high neutron flux (MSIVF) initiated at 100 percent of rated core power (3,951 MWt) and 83 percent of rated core flow with one SRV out of service.

The SRV and SV safety mode lift setpoint tolerance is assumed to be +3 percent.

For a MSIVF initiated at 100 percent power/83 percent flow, the RPV starts overpressurization from 1,035 psig due to the stoppage of steam flow by the closure of MSIVs and the continuous generation of steam with decay heat inside the RPV. The RPV overpressurization continues and reaches a peak pressure at 1,321 psig (4 psi to the TS safety limit) when a total of 13 SRVs/SVs open to balance the steam generation.

On average, one SRV/SV opening has an effect of holding overpressurization down by about 22 psi. The 4 psi margin cannot afford the overpressurization induced by MSIVF at 100 percent power/83 percent flow with one more SRV/SV out of service.

Reactor overpressure results from the stoppage of steam flow in the main steam lines due to the closure of the MSIVs and continuous steam generation from fission heat (prior to scram), reactor sensible heat, and decay heat (post scram), etc. With the steam flow and steam generation rate being proportional to the reactor power, it is reasonable for the licensee to propose a reduction of reactor operating power in order to allow one additional SRV or SV to be out of service.

The licensee has conservatively estimated that one SRV/SV accounts for about 7 percent of rated reactor power. Therefore, the licensee proposed an allowable maximum reactor power of 85 percent of rated power (i.e., 3,358 MWt) for 2 SRV/SVOOS.

The licensee provided an evaluation of an MSIVF event with 2 SRV/SVOOS to justify the above-proposed reactor operating power to accommodate 2 SRV/SVOOS.

One analysis was performed during Reload 19 (Cycle 20) with initial core power of 100 percent of the pre-extended power uprate rated power level (3,514 MWt) and 2 SRV/SVOOS.

The calculated peak steam dome pressure of 1,298 psig is 27 psi below the TS safety limit. Since the analysis power level of 3,514 MWt is 4.6 percent higher than the proposed allowable maximum power of 3,358 MWt, the licensee considers the proposed TS change justifiable.

However, the ASME overpressure analysis for Cycle 20 is based on+ 1 percent instead of +3 percent SRV/SV lift setpoint tolerance.

In addition, the high neutron flux scram signal used in that analysis is based on a rated power of 3,514 MWt instead of 3,951 MWt. (Note: The NRC staff approved license amendments for PBAPS, Units 2 and 3, after Cycle 20 to increase the allowable as-found SRV and SV lift setpoint tolerance from +/-1 percent to +/-3 percent (Reference 4). The licensee developed the allowable operating power to accommodate 2 SRV/SVOOS without explicitly considering these effects that will potentially increase the peak vessel pressure during an MSIVF event. To resolve the above concerns, the NRC staff determined that an independent confirmatory analysis was prudent, and a request for additional information (RAI), SRXB-RAl-1, was issued to the licensee to collect the information required to perform the analysis.

In response to the request (Reference 2), the licensee provided the information that enabled the staff to perform the confirmatory analysis with a general purpose thermal hydraulic computer code. The calculated results showed that the amount of reactor core coolant mass and enthalpy just before the SRV/SV lifting for the 85 percent power case is, as expected to be, less than that for the 100 percent power case. In terms of average RPV pressure, the difference between these two cases is about 40 psi. In other words, if the 100 percent power case had its peak dome pressure below the TS limit with one SRV/SVOOS, then the pressure margin for the 85 percent power case versus the 100 percent power case (i.e., 40 psi), should be able to accommodate an additional SRV/SVOOS.

As expected, the calculated peak reactor dome pressure from an MSIVF event initiated from 85 percent of rated power (i.e., 3,358 MWt) with 2 SRV/SVOOS is below the TS safety limit by an acceptable margin. In the same RAI response, the licensee provided the last two PBAPS, Unit 2, SRVs/SVs test results.

It indicated that all except one of the tested SRVs/SVs would open not only within allowed tolerance but also mostly lower than the opening setpoint.

The one failed test was due to an out-of-tolerance condition, but in the lower setpoint direction.

If the same test result trend is maintained, then the lost relief capacity due to the additional SRV/SVOOS can be compensated.

Based on the considerations discussed above, the NRC staff concludes that there is reasonable assurance that the PBAPS, Unit 2, RCPB will be protected by the SRVs and SVs, while allowing 2 SRV/SVOOS when the reactor power is limited to 85 percent of rated thermal power (i.e., 3,358 MWt) during any condition of normal operation, including the AOOs, consistent with draft GDC-9, final GDC-15, and final GDC-31. 3.3.2 A TWS Overpressure Analysis An A TWS is defined as an AOO followed by the failure of the reactor portion of the protection system specified in draft GDCs 14 and 15. Requirements related to A TWS events are specified in 1 O CFR 50.62. The NRC staff review includes confirming that the peak reactor vessel bottom pressure will be less than the ASME Service Level C limit of 1,500 psig during an ATWS overpressure event as protected by the SRVs/SVs and systems required in accordance with 10 CFR 50.62 (e.g., alternate rod injection system, standby liquid control system).

As discussed in Section 9.3.1.1 of Attachment 5 to the licensee's letter dated September 4, 2014 (ADAMS Accession No. ML 14247A503),

the PBAPS plant-specific ATWS overpressure analysis is performed using the NRC-approved overpressure methodology documented in General Electric Topical Report NEDE-32906P-A, Supplement 3-A (Reference 5). The limiting A TWS event for vessel overpressure is the MSIV closure (MSIVC) event at beginning of cycle initiated from 100 percent of current licensed thermal power (3,951 MWt) and 83 percent of rated core flow. With one SRV/SVOOS, the licensee determined that the peak vessel bottom pressure is 1,419 psig, which is below the ASME Service Level C limit of 1,500 psig. The peak vessel bottom pressure response is dependent on several inputs, including the SRV upper tolerance

(+3 percent drift tolerance) assumed in the ATWS an*alysis.

The NRG staff issued SRXB-RAl-2 requesting clarification about the current ATWS licensing analysis information and its applicability to the new proposed operating condition (85 percent power, 2 SRV/SVOOS).

In the response to the SRXB-RAl-2 (Reference 2), the licensee justified that the MSIVC is the limiting ATWS overpressure event and is also applicable to the new operating condition.

The staff finds the pertinent data for ATWS analysis (e.g., MSIV flow area as a function of closure time) is similar to but less severe than that for ASME overpressure

analysis, although the A TWS overpressure is more severe than ASME overpressure due to the loss of direct scram. Based on the above considerations, the NRG staff concludes that the results of the licensee's A TWS overpressure analysis for the limiting event, under the proposed new operating condition, provides reasonable assurance that the peak reactor vessel bottom pressure will be less than the ASME Service Level C limit of 1,500 psig. 3.3.3 Other Assessments The licensee stated that assessments of the impact of the proposed TS change on the current cycle's thermal limits, ECCS-LOCA performance and high pressure systems performance, were performed, and it was confirmed that the existing analyses remained bounding.

However, no justification was provided in the LAR package.

The NRG staff's RAls, SRXB-RAl-3 and SRXB-RAl-4, were issued to request the justification.

The licensee provided its response in Reference

2. The staff's evaluation of the licensee's assessment and associated response is provided below. Thermal Limits Assessment Draft GDCs 6, 14, 15, and 29 provide requirements related to core design and protection systems in order to assure that acceptable fuel damage limits are not exceeded.

Operating limits are established to ensure that regulatory limits, including safety limits, are not exceeded for a range of postulated events (transients and accidents).

The safety limit minimum critical power ratio (SLMCPR) ensures that 99.9 percent of the fuel rods are protected from boiling transition during normal operation or AOO. The operating limit minimum critical power ratio (OLMCPR) ensures that the SLMCPR will not be exceeded as a result of an AOO. The maximum average planar linear heat generation rate (MAPLHGR) and linear heat generation rate (LHGR) limits ensure that the plant does not exceed regulatory limits established in 1 O CFR 50.46 or the fuel design limit. The MAPLHGR limit is determined by analyzing the limiting LOCA for the given plant. The LHGR limit is determined by the fuel rod thermal-mechanical design. In response to SRXB-RAl-3 (Reference 2), the licensee stated that the OLMCPR multiplier, Kp, and power dependent LHGR multiplier (LHGRFAC(P))

for 85 percent of rated power are determined by evaluating the limiting AOO transients (e.g., load rejection no bypass (LRNBP) and feedwater controller failure (FWCF)) for each equipment out-of-service combination.

No one particular AOO transient establishes the multipliers Kp or LHGRFAC(P) values for all combinations or for all power levels for any one out-of-service combination.

For instance, in the Base combination of PBAPS, Unit 2, Cycle 22 core operating limits report (COLA) (Reference

6) Table 4-2, the LRNBP is limiting at some power levels, but the FWCF is limiting at other power levels. In the turbine bypass system out-of-service combination, the FWCF is typically limiting for most, but not at all power levels. For COLA Table 5-3, the FWCF is typically, but not always, limiting for LHGRFAC(P).

The Kp and LHGRFAC(P) multipliers are developed in a manner that ensures all AOO transients are bounded, since no one AOO transient event is limiting at all off-rated conditions.

The licensee determined that the most limiting MCPR and maximum LHGR values during pressurization events are calculated to occur prior to the opening of any SRVs or SVs. The limiting MCPR occurs approximately 0.7 to 0.8 seconds prior to SRV opening, and the maximum LHGR occurs approximately 1.1 seconds prior to SRV opening.

Therefore, the most limiting MCPR and maximum LHGR values are not impacted by SRVs or SVs that may be out of service.

The NRC staff reviewed the licensee's response and found it acceptable because:

(1) the licensee applied NRG-approved licensing methodologies and analytical methods and codes for determination of the thermal limits for Cycle 22, and (2) the new operating condition (85 percent power, 2 SRV/SVOOS) has no impact on the existing analysis results because the thermal limits occur prior to the lifting of any SRVs/SVs.

ECCS-LOCA Performance Assessment The NRC's acceptance criteria related to ECCS and LOCA performance is based, in part, on: (1) 1 O CFR 50.46, insofar as it establishes standards for the calculation of ECCS performance and acceptance criteria for that calculated performance; (2) draft GDCs 40 and 42, insofar as they require that protection be provided for engineered safety features against the dynamic effects that might result from plant equipment

failures, as well as the effects of a LOCA; and (4) draft GDCs 37, 41, and 44, insofar as they require that a system to provide abundant emergency core cooling be provided so that fuel and clad damage that would interfere with the emergency core cooling function will be prevented.

The PBAPS, Unit 2, ECCS is designed to provide protection against postulated LOCAs caused by ruptures in the primary system piping. The ECCS performance under all LOCA conditions and the analysis models must satisfy the requirements of 10 CFR 50.46 and Appendix K to 1 O CFR Part 50. For the analytical methods and codes, the ECCS-LOCA analysis is performed in accordance with the NRG-approved methodology specified in Global Nuclear Fuel Topical Report NEDE-24011-P-A (Reference 7). For PBAPS, Unit 2, a base ECCS-LOCA

analysis, with a full-scope break spectrum, forms the initial SAFER/GESTR-LOCA analysis-of-record for the rated power level. During reload analyses, the licensee evaluated the cycle-specific MAPLHGR limits to confirm that the MAPLHGR limit based on the ECCS-LOCA analysis-of-record remains bounding.

In a response to RAI SRXB-RAl-4 (Reference 2), the licensee stated that the impact of SRV/SVOOS on the ECCS/LOCA performance had been addressed in the technical report NEDC-33533P (Reference

8) submitted for the SRV/SV lift setpoint tolerance LAA that was approved (Reference 4). Reference 8 included evaluations for containment pressure and temperature for design-basis accident (DBA) LOCA, small steam line breaks, intermediate and small line break accidents, suppression pool temperature, and DBA-LOCA hydrodynamic loads. Those evaluations concluded that the ECCS/LOCA results were not affected by SRV/SV setpoint tolerance relaxation and/or an SRV/SVOOS.

Reference 8 addressed the impact on operation with one SRV/SVOOS on SRV dynamic loads. The report states that having an SRV/SVOOS does not have an effect on SRV dynamic loads. SRV loads are driven by SRV opening pressure and the SRV discharge line water level at the time of the second SRV actuation.

The SRV discharge line water level is, in turn, a function of the time between the closure of the valves at the end of the first actuation and the time of the second actuation.

Having an SRV/SVOOS will not affect the operable valves' setpoints or the time between the initial and second valve actuations.

As such, having an SRV/SVOOS has no effect on SRV dynamic loads. SVs have no impact on suppression pool dynamic loads. These conclusions do not change for an additional SRV/SVOOS.

High Pressure System Performance Assessment The high pressure systems include the high pressure coolant injection (HPCI), reactor core isolation cooling (RCIC), and SLC systems.

The NRC's acceptance criteria for HPCI are based on: (1) draft GDCs 40 and 42, insofar as they require that protection be provided for ESFs against the dynamic effects that might result from plant equipment

failures, as well as the effects of a LOCA; and (2) draft GDCs 37, 41, and 44, insofar as they require that a system to provide abundant emergency core cooling be provided so that fuel and clad damage that would interfere with the emergency core cooling function will be prevented.

The NRC's acceptance criteria for RCIC are based, in part, on: (1) draft GDC-40, insofar as it requires that protection be provided for ESFs against dynamic effects;

{2) 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; and (3) 10 CFR 50.63, insofar as it requires that the plant withstand and recover from an SBO of a specified duration.

For the SLC system, the NRC's acceptance criteria are based, in part, on: 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 RPV at a boron concentration, boron enrichment, and flow rate that provides a set level of reactivity control.

In response to RAI SRXB-RAl-4 (Reference 2), the licensee stated that according to Reference 8, the most significant potential effect of an SRV/SVOOS on the HPCI and RCIC systems' operations is the maximum reactor pressure at which they are required to deliver water to the reactor.

Both systems are required to provide injection into the RPV at the pressure corresponding to the lowest group of SRV lift setpoints (including drift). PBAPS, Unit 2, includes four SRVs in the lowest setpoint group. The condition of an SRV/SVOOS does not change the SRV lift setpoints.

This statement applies equally to either one or two SRV/SVOOS.

Therefore, it is concluded that the maximum reactor pressure at which the HPCI and RCIC systems are required to deliver water to the reactor would not be adversely impacted by operation with an additional (up to two) SRV/SVOOS.

The limiting A TWS case for long-term

results, including peak reactor lower plenum pressure during SLC system injection, has been evaluated for the PBAPS current license thermal power conditions.

At the time of SLC system injection for the limiting MSIVC case, the reactor power is in the range of approximately 25 percent power. This power level is well within the available steam relief system capacity, even with an additional (up to two) SRV/SVOOS.

Therefore, the licensee concluded that peak reactor lower plenum pressure during SLC system injection would not increase significantly, and the SLC system performance would not be adversely impacted by operation with an additional (up to two) SRV/SVOOS.

The NRG staff reviewed the above RAI response, and the evaluation bases and conclusions in Reference 8 and determined that the licensee's assessment is acceptable because the impact of the new operating condition has been addressed in Reference

8. The staff concludes that the proposed amendment would not impact the capability of HPCI, RCIC, and SLC systems from performing their intended design functions, consistent with the applicable regulatory requirements stated above. 3.4 Technical Evaluation Conclusion Based on the considerations discussed in Section 3.3 of this safety evaluation, the NRC staff concludes that the proposed changes to the LCO for TS 3.4.3 will continue to meet the requirements of 1 O CFR 50.36( c)(2) in that the LCO describes the lowest functional capability or performance level of the SRVs and SVs required for safe operation of the facility.

Therefore, the staff further concludes that the proposed amendment is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Pennsylvania State official was notified of the proposed issuance of the amendment on September 19, 2017. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 1 O CFR Part 20. The NRC staff has determined that the amendment involves 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 amendment involves no significant hazards consideration, and there has been no public comment on such finding (July 5, 2017; 82 FR 31094). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 1 O CFR 51.22(c)(9).

Pursuant to 1 O CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

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 amendment 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 Exelon to NRC, dated May 19, 2017, "Peach Bottom Atomic Station, Unit 2, License Amendment Request-Revise Technical Specifications Section 3.4.3 (SRVs/SVs) for the Remainder of the Current Operating Cycle for Unit 2" (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 17139D357).

2. Letter from Exelon to NRC, dated August 29, 2017, "Peach Bottom Atomic Power Station, Unit 2, License Amendment Request-Revise Technical Specifications Section 3.4.3 (SRVs/SVs) for the Remainder of the Current Operating Cycle for Unit 2 -Supplement 1 Response to Request for Additional Information,"

dated May 19, 2017 (ADAMS Accession No. ML17241A194). 3. NRC Staff Requirements Memorandum for SECY-92-223, "Resolution of Deviations Identified During the Systematic Evaluation Program,"

dated September 18, 1992 (ADAMS Accession No. ML003763736).

4. Letter from NRC to Exelon, dated May 5, 2014, "Peach Bottom Atomic Station, Units 2 and 3 -Issuance of Amendments Re: Safety Relief Valve and Safety Valve Lift Setpoint Tolerance" (ADAMS Accession No. ML 14079A102).

5. GE Hitachi Nuclear Energy, Licensing Topical Report, NEDE-32906P, Supplement 3-A, Revision 1, "Migration to TRACG04/PANAC11 from TRACG02/PANAC10 for TRACG AOO and ATWS Overpressure Transients,"

April 2010 (ADAMS Package Accession No. ML 110970401

). 6. Letter from Exelon to NRC, dated November 18, 2016, "Peach Bottom Atomic Station, Unit 2, Issuance of the Core Operating Limits Report for Reload 21, Cycle 22, Revision 11" (ADAMS Accession No. ML 16327 A068). 7. Global Nuclear Fuel, Technical Reports NEDE-24011-P-A-20 and NEDE-24011-P-A-20-US, "General Electric Standard Application for Reactor Fuel (GESTAR-11)"

(latest approved version).

8. GE Hitachi Nuclear Energy Report NEDC-33533P, Revision 1, "Peach Bottom Atomic Power Station, Units 2 and 3, Safety Valve Setpoint Tolerance Increase Safety Analysis Report,"

May 2013. (Note: The subject report is Attachment 3 to the cover letter included in ADAMS Package Accession No. ML 131750144 and is non-public.

Attachment 4 (ADAMS Accession No. ML 13175A 110) to the letter contains a public version of the report.)

Principal Contributor:

S. Peng Date: October 25, 2017

SUBJECT:

PEACH BOTTOM ATOMIC POWER STATION, UNIT 2 -ISSUANCE OF AMENDMENT RE: SAFETY RELIEF VALVE AND SAFETY VALVE OPERABILITY FOR CYCLE 22 (CAC NO. MF9705, EPID L-2017-LLA-0229)

DATED OCTOBER 25, 2017 DISTRIBUTION:

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