ML18150A387

License Amendment Request - Revised Technical Specifications to Allow Two Safety Relief Valves / Safety Values to Be Out-of-Service with Increased Reactor Pressure Safety Limit

ML18150A387
Person / Time
Site:	Peach Bottom
Issue date:	05/30/2018
From:	David Helker Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML18150A385	List: JAFP-18-0052, Proposed Alternative to Utilize Code Case N-879 ML18150A387
References
Download: ML18150A387 (55)

Text

200 Exelon Way Exelon Generation ~; Kennett Square, PA 19348 www.exeloncorp.com PROPRIETARY INFORMATION - WITHOLD UNDER 10 CFR 2.390 10 CFR 50.90 May 30, 2018 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Peach Bottom Atomic Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278

Subject:

License Amendment Request- Revise Technical Specifications to Allow Two Safety Relief Valves/Safety Valves to be Out-of-Service with Increased Reactor Pressure Safety Limit

References:

1. NRC letter to Exelon, "Peach Bottom Atomic Power Station ,

Units 2 and 3 - Issuance of Amendments Re: Extended Power Uprate (TAC NOS. ME9631 and ME9632), dated August 25, 2014 (ADAMS Accession No. ML14133A046)

2. NRC letter to Exelon, "Peach Bottom Atomic Power Station, Units 2 and 3 - Issuance of Amendments Re: Measurement Uncertainty Recapture Power Uprate (CAC Nos. MF9289 and MF9290; EPID L-2017-LLS-0001)," dated November 15, 2017 (ADAMS Accession No. ML17286A013)

In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC), proposes a change to the Technical Specifications (TS), Appendix A of Renewed Facility Operating License Nos. DPR-44 and DPR-56 for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3. Specifically, the proposed changes revise the Renewed Operating License and Technical Specifications (TS) to allow continued operation with two Safety Relief Valves/Safety Valves (SRVs/SVs) out-of-service and to increase the Reactor Coolant System Pressure Safety Limit.

The proposed changes are based on taking credit for the additional Safety Valve (SV) that was installed on each unit per the Extended Power Uprate amendments for PBAPS Unit 2 and Unit 3 (Reference 1) and a re-evaluation of the transient pressure analysis at the current licensed thermal power authorized by the Measurement Uncertainty Recapture (MUR) uprate amendments (Reference 2). This analysis is provided in Attachment 4 and concludes that positive margin remains between the reactor vessel overpressure analysis and the overpressure limits.

Attachment 4 transmitted herewith contains Proprietary Information.

When separated from Attachment 4, this document is decontrolled.

U.S. Nuclear Regulatory Commission License Amendment Request Revise TS for Two SRVs/SVs Out-of-Service Docket Nos. 50-277 and 50-278 May 30, 2018 Page 2 The proposed change has been reviewed by the PBAPS Plant Operations Review Committee in accordance with the requirements of the EGC Quality Assurance Program.

EGC requests approval of the proposed amendments by May 30, 2019. Once approved, the amendments shall be implemented within 60 days.

EGC has concluded that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92, "Issuance of amendment."

This amendment request contains no regulatory commitments. provides the evaluation of the proposed changes. Attachment 2 provides a copy of the marked-up TS pages that reflects the proposed changes. Attachment 3 provides a copy of the marked-up TS Bases pages that reflects the proposed changes (for information only). Attachment 4 provides a proprietary version of the GE Hitachi Nuclear Energy (GEH) report documented in 004N6240-P, "Peach Bottom Units 2 & 3 Two Safety Relief Valves Out-of-Service Evaluation," Revision 1. Attachment 5 provides an affidavit executed by GEH for withholding certain information contained in Attachment 4. Attachment 6 provides a non-proprietary version of Attachment 4. In accordance with 10 CFR 2.390, "Public inspections, exemptions, requests for withholding," EGC requests withholding Attachment 4 from public disclosure.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), EGC is notifying the Commonwealth of Pennsylvania and the State of Maryland of this request by transmitting a copy of this letter along with the Attachments to the designated State Officials.

Should you have any questions concerning this submittal, please contact David Neff at 267-533-1132.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 3Q1h day of May 2018.

Respectfully, David P. Helker Manager, Licensing & Regulatory Affairs Exelon Generation Company, LLC

U.S. Nuclear Regulatory Commission License Amendment Request Revise TS for Two SRVs/SVs Out-of-Service Docket Nos. 50-277 and 50-278 May 30, 2018 Page 3 Attachments: 1. Evaluation of Proposed Changes

2. Markup of Proposed Technical Specifications Pages

3. Markup of Proposed Technical Specifications Bases Pages (For Information Only)

4. GE Hitachi Nuclear Energy 004N6240-P, "Peach Bottom Units 2 & 3 Two Safety Relief Valves Out-of-Service Evaluation," Revision 1, (Proprietary Version)

5. GE Hitachi Nuclear Energy Affidavit Supporting Withholding Attachment 4 from Public Disclosure

6. GE Hitachi Nuclear Energy 004N6240-NP, "Peach Bottom Units 2 & 3 Two Safety Relief Valves Out-of-Service Evaluation," Revision 1, (Non-Proprietary Version) cc: Regional Administrator - NRC Region I w/ attachments II NRC Senior Resident Inspector - PBAPS II NRC Project Manager, NRR - PBAPS R.R. Janati, Pennsylvania Bureau of Radiation Protection w/NP attachments II S.T. Gray, State of Maryland

ATT AC HM ENT 1 License Amendment Request Peach Bottom Atomic Power Station, Units 2 and 3 Docket Nos. 50-277 and 50-278 EVALUATION OF PROPOSED CHANGES

Subject:

License Amendment Request to Revise Technical Specifications to Allow Two Safety Relief Valves/Safety Valves to be Out-of-Service with Increased Reactor Pressure Safety Limit 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedence 4.3 No Significant Hazards Consideration 4.4 Conclusions

6.0 REFERENCES

License Amendment Request Attachment 1 Revise TS For Two SRVs/SVs Out-of-Service Page 1 of 13 Docket Nos. 50-277 and 50-278 Evaluation of Proposed Changes 1.0

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC), proposes a change to the Technical Specifications (TS), Appendix A of Renewed Facility Operating License (RFOL) Nos. DPR-44 and DPR-56 for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 respectively.

Specifically, the proposed changes revise the RFOL and TS to allow continued operation with two Safety Relief Valves/Safety Valves (SRVs/SVs) out-of-service and to increase the Reactor Coolant System (RCS) Pressure Safety Limit (SL).

The proposed amendments would revise TS Section 3.4.3 to lower the required number of operable Safety Relief Valves (SRVs) and Safety Valves (SVs) from 13 to 12. The proposed amendments will also revise TS Section 2.1.2 to raise the RCS Pressure SL Limit from 1325 to 1340 psig. This license amendment request (LAR) is based on a GE Hitachi Nuclear Energy (GEH) report documented in 004N6240-P, "Peach Bottom Units 2 & 3 Two Safety Relief Valves Out-of-Service Evaluation," (Attachment 4 to this LAR) that concludes positive margin remains to the reactor vessel overpressure protection limits. This analysis is based on the current licensed thermal power (CLTP) authorized by the Measurement Uncertainty Recapture (MUR) uprate amendments (Reference 1) and considers the Maximum Extended Load Line Limit Analysis-Plus (MELLLA+) amendments (Reference 2),

2.0 DETAILED DESCRIPTION Overpressure protection of the reactor pressure vessel (RPV) at PBAPS is provided by 11 SRVs and three SVs located on the main steam lines between the reactor vessel and the first isolation valve within the drywell. The SRVs can actuate by either an overpressure condition (above the SRV setpoint) or by remote operation. Along with overpressure protection, five SRVs are part of the Automatic Depressurization System (ADS) designed within the Emergency Core Cooling System (ECCS) which is governed by TS Section 3.5.1, Emergency Core Cooling System -

Operating. In the ADS function, the five SRVs are used to depressurize the nuclear boiler system to allow the low pressure coolant injection systems to function. The SRVs discharge to the suppression pool. The spring-loaded SVs actuate during an overpressure condition and discharge to the drywell. TS Section 3.4.3 currently requires 13 of the 14 SRVs/SVs, in any combination, to be operable during operation in Modes 1, 2, and 3. Amendment 315 has been implemented on PBAPS Unit 2 that allows for operation with 12 of the 14 SRVs/SVs operable at reduced power during cycle 22 (Reference 10), and is scheduled to expire in the fall of 2018.

The proposed amendment would revise TS Section 3.4.3 to lower the required number of operable Safety Relief Valves (SRVs) and Safety Valves (SVs) from 13 to 12. The Technical Evaluation section below provides an evaluation showing that 12 SRVs and SVs provide sufficient safety function capability, with margin, for the RCS overpressure protection during operation in Modes 1, 2 and 3 and with the current licensed thermal power (CLTP) level of 4,016 MWt. To accommodate the lower number of required SRVs/SVs and the impact on the RCS overpressure analysis, the proposed amendment also revises TS Section 2. 1.2 to raise the RCS SL value from 1325 to 1340 psig.

License Amendment Request Attachment 1 Revise TS For Two SRVs/SVs Out-of-Service Page 2of13 Docket Nos. 50-277 and 50-278 Evaluation of Proposed Changes The proposed change includes the following TS revisions:

TS Section 3.4.3, "Safety Relief Valves (SRVs) and Safety Valves (SVs)," Limiting Condition for Operation (LCO) 3.4.3 - This LCO is being revised to lower the minimum number of SRVs/SVs required to be operable from 13 to 12.

TS Section 2.1.2, "Reactor Coolant System Pressure SL," -This section is being revised to raise the reactor steam dome pressure limit from 1325 to 1340 psig. provides the existing TS pages marked-up to show the proposed changes.

Marked-up TS Bases are provided in Attachment 3 for information only. The TS Bases change will be processed in accordance with the PBAPS, TS Bases Control Program (TS 5.5.10). provides supporting analysis performed by GEH specifically for this LAR.

3.0 TECHNICAL EVALUATION

The proposed amendment would revise TS Section 3.4.3 to lower the required number of operable SRVs and SVs from 13 to 12 when operating in Modes 1, 2 and 3 with the CLTP level of 4,016 MWt and considering the MELLLA+ operating domain. The proposed amendments will also revise TS Section 2.1.2 to raise the RCS Pressure SL Limit from 1325 to 1340 psig. This technical evaluation is based on a GEH report provided in Attachment 4 to this LAR that concludes positive margin remains to the RCS overpressure protection limits.

Reactor Vessel Overprotection Equipment Description The Nuclear Boiler System (NBS) transports the steam generated in the 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 safety valves (SVs) and 11 dual function safety relief valves (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 Reactor Coolant Pressure Boundary (RCPB) is 1250 psig. The acceptance limit for pressurization events is the American Society of Mechanical Engineers (ASME) code allowable peak pressure of 1,375 psig (110% of design value). The SVs and SRVs are designed to meet the requirements for reactor vessel overpressure protection to conform to ASME Boiler and Pressure Vessel Code (B&PVC), Section Ill, Article 9. The nuclear system pressure relief system design includes 11 SRVs with opening setpoints of 1,135 psig (4), 1,145 psig (4), and 1,155 psig (3), and 3 SVs with opening setpoints of 1,260 psig, assuming a +/-3% tolerance setting on each valve .

The SRVs are Target Rock three-stage pilot operated safety/relief valves. The SVs are Dresser spring loaded safety valves. The SRVs can actuate by either of two modes: the safety (overpressure) 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

License Amendment Request Attachment 1 Revise TS For Two SRVs/SVs Out-of-Service Page 3of13 Docket Nos. 50-277 and 50-278 Evaluation of Proposed Changes when steam pressure at the inlet overcomes the spring force holding the valve disc closed. This satisfies the ASME Code requirement.

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 ADS.

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 reactor coolant pressure boundary. 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 ADS.

The pressure relief system prevents over pressurization of the nuclear system during Anticipated Operational Occurrence transient events (AOOs}, which includes the plant ASME code upset overpressure protection event, and postulated Anticipated Transient Without Scram (ATWS) events. The SRVs and SVs, along with other functions, provide this protection.

Overpressure Protection Function During Power Operation (Limiting AOO Event)

For PBAPS, the limiting overpressure AOO event is the main steam isolation valve closure with scram on high flux (MSIVF). The case of MSIVF is analyzed during every cycle-specific reload and was re-evaluated at MUR conditions (Reference 1) to assure that the ASME code allowable value for peak vessel pressure is not violated. The MSIVF case analysis conservatively assumed that the MSIV position scram fails and the event terminates on a high neutron flux scram signal.

The closure of all MSIVs causes a rapid pressure increase in the reactor vessel. The pressure increase is mitigated by the actuation of the SRVs and SVs. The MELLLA+ amendment (Reference 2) and the Reload Analysis for the current operating cycles for PBAPS Units 2 and 3 (References 3 and 4) confirmed that the MSIVF event at the CLTP of 4,016 MWt remains the limiting overpressure event with 13 of 14 operable SRVs/SVs. The overpressure analyses assume that the out-of-service relief valve is an SRV with the lowest pressure setpoint. This assumption is conservative since this minimizes the initial pressure relief capacity and results in the highest peak pressure value for the overpressure analyses. Hence, the analyses are bounding for any SRV or SV out-of-service. This analysis forms the current overpressure analysis.

For this LAR, the overpressure analysis for the MSIVF event was re-analyzed using the most recent reload licensing analysis inputs for PBAPS Units 2 and 3 at CLTP conditions with GNF2 fuel, with the SRVs/SVs configuration at a +/-3% tolerance setting for each valve and with two SRVs at the lowest pressure setpoint out-of-service (two SRVOOS). This selection of SRVs minimizes the initial pressure relief capacity and bounds any combination of SRVs and/or SVs out-of-service for overpressure analysis. This analysis was performed using the TRACG AOO methodology (References 5 and 6) at the currently licensed power level of 4,016 MWt, with consideration for the MELLLA+ operating domain. The MSIVF event was analyzed at both minimum (85% of rated) and maximum (110% of rated) core flow, which are bounding conditions for reactor vessel overpressure calculations. Both PBAPS units have reactor cores with only

License Amendment Request Attachment 1 Revise TS For Two SRVs/SVs Out-of-Service Page 4of13 Docket Nos. 50-277 and 50-278 Evaluation of Proposed Changes GNF2 fuel assemblies. The TRACG statistical pressure adder was calculated for GNF2 fuel for use in the ASME code overpressure analysis per the TRACG AOO methodology.

The PBAPS Unit 2 results for one SRVOOS from Reference 3 and the results from the new analysis performed with two SRVOOS are provided in Table 1. The PBAPS Unit 3 results for one SRVOOS from Reference 4 and the results from the new analysis performed with two SRVOOS are provided in Table 2. The current one SRVOOS peak pressure results includes a pressure adder from Reference 7. The new two SRVOOS peak pressure results include an updated pressure adder from Reference 8.

Table 1 PBAPS Unit 2 Cycle 22 MUR Overpressure Analysis Results Peak MSIVF Peak Limiting Number Peak Dome SRV/SV Vessel Vessel Domain of Pressure Configuration Pressure Pressure SRVOOS (psig)

(psial Adder (csid)

Increased Core Flow (ICF, 110% Core 1 1,324 1,353 1 SRVOOS Flow)

See Attachment (Hard Bottom Burn (HBB)) 4 Table 1 MELLLA+

(Core Flow 1 1,325 1,349 1SRVOOS 85%)

(HBB}

ICF (HBB) 2 1,325 1,354 2 SRVOOS See Attachment MELLLA+ 4 Table 1 2 1,326 1,351 2 SRVOOS (HBB}

Table 2 PBAPS Unit 3 Cycle 22 MUR Overpressure Analysis Results Peak MSIVF Peak Limiting Number Peak Dome SRV/SV Vessel Vessel Domain of Pressure Configuration Pressure Pressure SRVOOS (psig)

(psial Adder (psid)

ICF 1 1,322 1,352 1 SRVOOS (HBB)) See Attachment MELLLA+ 4 Table 2 1 1,324 1,349 1 SRVOOS (HBB)

ICF (HBB) 2 1,324 1,353 2 SRVOOS See Attachment MELLLA+ 4 Table 2 2 1,327 1,351 2 SRVOOS (HBB)

License Amendment Request Attachment 1 Revise TS For Two SRVs/SVs Out-of-Service Page 5of13 Docket Nos. 50-277 and 50-278 Evaluation of Proposed Changes The reload licensing analysis results for one SRVOOS show that the margin to the current dome pressure safety limit of 1,325 psig is less than 5 psi; however, for the peak vessel pressure, there is over 20 psi margin to the ASME code overpressure limit of 1,375 psig. This current analysis of record includes a TRACG statistical pressure adder in accordance with Reference 7. For the two SRVOOS analysis, a revised TRACG statistical pressure adder was used as discussed in and Reference 8. With two SRVOOS, the peak steam dome pressure is higher than the current dome pressure safety limit. However, for the peak vessel pressure, there is still over 20 psi margin to the ASME code overpressure limit. This margin includes the penalty due to the TRACG statistical pressure adder required to be included in the analysis results.

Given the maximum calculated pressure difference between the steam dome pressure and the peak vessel pressure, raising the dome pressure safety limit by 15 psi from 1,325 to 1,340 psig still provides sufficient margin (approximately 5 psi) for the peak pressure vessel pressure and thus continues to support the ASME code overpressure limit requirements. This is further illustrated in Table 3. See Attachment 4 section 2.1.2 and 2.1.3 and Reference 6 for additional technical information regarding the TRACG analysis.

Table 3 Peak Pressure Vessel Margin Summary Assuming SL Pressure Existing Proposed Maximum Maximum ASME Code Available Steam SL Analytical Pressure Reactor Vessel Overpressure Margin Dome Increase Difference - Steam Pressure (psig) Limit (psig) (psi)

Pressure (psi) Dome and Reactor SL (psiQ) Vessel (psi) 1325 + 15 + 29 = 1369 1375 6 Compliance with the ASME upset code requirements for vessel overpressure protection would still be ensured with this change to the dome pressure safety limit (1,340 psig) to support operation with two SRVOOS.

Overpressure Protection Function During Anticipated Transients without Scram (Limiting Special Event)

The evaluation of the PBAPS response to postulated ATWS events is not a design basis requirement. However, evaluation is necessary to demonstrate that permitting full-power operation with two SRVOOS does not have an unacceptable effect on the mitigation capability of the plant during postulated ATWS events.

The peak vessel pressure analysis is performed because the ATWS acceptance criteria most challenged by 2 SRVOOS is the peak vessel pressure. The Peak Cladding Temperature (PCT),

peak containment pressure, and suppression pool temperature acceptance criteria are not affected by the additional SRVOOS. These long-term transient response parameters are driven by the core power generation and the relief capacity of the operable SRVs and SVs. Based on the ATWS analysis performed in support of the PBAPS MUR Amendment (Reference 1) and the discussion in Attachment 4 (Section 2.2.1, 2nd paragraph) the steam flow following the post initial pressure peak is well within the capacity of the remaining nine in-service SRVs.

License Amendment Request Attachment 1 Revise TS For Two SRVs/SVs Out-of-Service Page 6of13 Docket Nos. 50-277 and 50-278 Evaluation of Proposed Changes Calculations were performed to determine the peak vessel pressure during an ATWS event to demonstrate compliance with the ASME Code Service Level C limit of 1,500 psig (Emergency Condition). The analysis was performed using the TRACG transient analysis code (References 5 and 9) with the reactor operating at the MUR bounding high power level (4,018 MWt) and limiting MELLLA+ core flow conditions. Two events were considered: Main Steam Isolation Valve Closure (MSIVC) and Pressure Regulator Failure Open (PRFO). Each of these events has the potential to yield the maximum vessel overpressure result. In the MSIVC event, the main steam isolation valves (MSIVs) on all four steam lines close simultaneously, while the normal direct scram on full MSIVC fails. In addition, scram on high neutron flux and high vessel pressure are assumed to fail. In the PRFO event, the pressure regulator failure produces the maximum steam flow demand. Reactor pressure drops and the MSIVs close on a low steam line pressure signal. Similarly, scram on full MSIV closure, high flux and high pressure all fail. The ultimate shutdown of the plant is accomplished through the actuation of the Standby Liquid Control System (SLCS). The initial reduction in power occurs as a result of the ATWS high dome pressure Recirculation Pump Trip (RPT). After the ATWS RPT and following the opening of the SRVs/SVs, the event is terminated for overpressure considerations.

The A TWS system response is simulated with two SRVOOS and the SRV/SV opening setpoints at +3% of the nominal setpoint. The resulting peak vessel pressure values and comparison between the one and two SRVOOS conditions for the MSIVC and PRFO events are provided in (Section 2.2.2, 1st paragraph). The results for the two SRVOOS condition are still below the 1,500 psig limit.

Compliance with the ASME Code Service Level C limit of 1,500 psig (Emergency Condition) for vessel overpressure protection in an ATWS event would still be ensured with this change to support operation with two SRVOOS.

Safety Systems Evaluation The other safety systems that may be affected by a higher peak vessel pressure have been analyzed to determine the effect that two SRVOOS will have on the ability of each system to meet design basis requirements. These systems are the High Pressure Coolant Injection (HPCI}, Reactor Core Isolation Cooling (RCIC), Standby Liquid Control (SLC) and the Control Rod Drive (CRD) systems.

HPCI and RCIC Systems Evaluation The HPCI and RCIC systems are required to provide injection into the reactor pressure vessel at the lowest group of SRV setpoints including the +3% tolerance drift provided there are at least two functional SRVs in the lowest group. For the two SVROOS condition, an evaluation is provided in Attachment 4, Section 3.1. Because there are four SRVs in the lowest group, taking two SRVOOS in this group still leaves two SRVs. Therefore, the HPCI and RCIC systems' injection capability is not affected by an additional SRVOOS and there is sufficient injection capability to support two SRVOOS.

SLC System Evaluation The SLC system is designed to inject over a wide range of reactor operating pressures and a peak injection pressure is calculated for the ATWS event. Based on the ATWS analysis performed in support of the PBAPS MUR Amendment (Reference 1) and the discussion in

License Amendment Request Attachment 1 Revise TS For Two SRVs/SVs Out-of-Service Page 7 of 13 Docket Nos. 50-277 and 50-278 Evaluation of Proposed Changes (Section 2.2.1, 2nd paragraph and Section 3.2), the steam flow following the post initial pressure peak is well within the capacity of the remaining nine in-service SRVs. For the two SRVOOS condition, an evaluation provided in Attachment 4, Section 3.2 concludes that the peak vessel pressure calculated for SLC system injection for both MUR and MELLLA+ conditions will not be exceeded with an additional SRVOOS. Therefore, the SLCS is adequate to support two SRVOOS.

CRD System Evaluation The CRD system is designed to shutdown the reactor by inserting control rods. The Control Rod Drive Mechanism (CROM) is also part of the reactor pressure boundary. For the two SRVOOS condition, an evaluation is provided in Attachment 4, Section 3.3. This evaluation concludes that the effect of the additional SRVOOS on control rod injection times is bounded by the current analysis. The evaluation also concludes that, for CRD integrity, the CROM has been analyzed (Attachment 4 Section 3.3) to a value that bounds the peak vessel pressures shown in Tables 1 and 2 provided above. Therefore, the CRD system is adequate to support two SRVOOS.

ECCS Loss of Coolant Accident (LOCA) Analvsis For large break LOCA cases, depressurization of the vessel occurs due to mass and energy release out the break. Therefore , further relief through the SRVs is not required to arrive at post-accident conditions whereby ECCS assets can be delivered to re-cover the core and acceptably arrest the temperature excursion on the cladding.

For small break LOCA cases, the SRVs are relied upon for depressurization through the Automatic Depressurization System (ADS). The SRVs can be actuated by alternate means, mechanical or pneumatic, so a valve declared out-of-service for one actuation may still be available for the alternate actuation. ADS availability is stipulated by TS Section 3.5.1, "ECCS-Operating," with assurance that the five SRVs supporting ADS would remain available despite changes to allow an additional SRVOOS (mechanically, for overpressure protection) under the provisions of TS Section 3.4.3, "SRVs and SVs." With this confirmation, and noting that the five SRVs available for ADS function conform to the basis of the most recent ECCS-LOCA analysis (Reference 2, as cited as the basis for Reference 1) for limiting small break cases, it is concluded that there would be no effect on the ECCS-LOCA analysis results, and continued compliance to the acceptance criteria of 10 Code of Federal Regulations (CFR) 50.46 would be ensured.

The SRVs are modeled in the ECCS- LOCA analysis, but a change to allow two SRVOOS will have no effect on the LOCA analysis.

Conclusion Analysis confirms that raising the number of out-of-service SRVs/SVs from one to two does not have an adverse impact on 1) the overpressure protection for the reactor pressure vessel, 2) the ability of HPCI, RCIC, SLC and CRD safety systems to perform their design basis requirements, and 3) the ECCS LOCA analysis. The analysis also confirms that for the peak vessel pressure, there is still over 20 psi margin to the ASME code overpressure limit. This margin also includes the penalty due to the TRACG statistical pressure adder required to be included in the analysis results, thereby providing additional analytical margin. Raising the dome pressure safety limit by 15 psi from 1,325 to 1,340 psig still provides sufficient margin (approximately 5 psi) for the peak

License Amendment Request Attachment 1 Revise TS For Two SRVs/SVs Out-of-Service Page 8of13 Docket Nos. 50-277 and 50-278 Evaluation of Proposed Changes pressure vessel pressure and thus continues to support the ASME code overpressure limit requirements. Compliance with the ASME upset code requirements for vessel overpressure protection is still ensured with this change to the dome pressure safety limit (1,340 psig) to support operation with two SRVOOS.

TS Bases Sections The TS Bases for the RCS Pressure SL (TS Section 2.1.2) will be revised to describe the increase in the reactor steam dome pressure limit value and the analysis used to determine the peak pressure for the reactor vessel pressure.

The TS Bases for the SRVs and SVs (TS Section 3.4.3) will be revised to describe the lowering of the required number of operable SRVs/SVs and the analysis used to determine the adequacy of the overpressure protection.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria As stated in Appendix H of the Peach Bottom Atomic Power Station (PBAPS) Updated Final Safety Analysis Report (UFSAR), the plant design was evaluated against the draft General Design Criteria proposed by the Atomic Energy Commission (AEC) in July 1967.

It was concluded that the design of Units 2 and 3 conforms with the intent of the proposed criteria.

Relief and safety valves and the Reactor Protection System (RPS) provide overpressure protection for the RCPB during power operation. The regulatory 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) draft GDC-33 insofar as it requires the reactor coolant pressure boundary to be capable of accommodating static and dynamic loads resulting from an inadvertent and sudden release of energy to the coolant.

The overpressure protection system must accommodate the most severe pressurization transient. Each PBAPS Unit has 11 installed Safety Relief Valves (SRVs) and three Safety Valves (SVs) of which a total of 13 SRVs/SVs are currently required to be operable. Evaluations have determined that the most severe transient is the closure of all main steam isolation valves (MSIVs), followed by reactor scram on high neutron flux (i.e.,

failure of the direct scram associated with MSIV positions, or abbreviated as MSIVF). For the purpose of the analyses, a total of 12 SRVs and SVs are assumed to operate in the safety mode (i.e., any combination of 2 SRVs or SVs out-of-service). The MSIVF event is conservatively analyzed assuming a failure of the MSIV position scram. The analysis results demonstrate that the design SRV and SV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure (110% x

=

1,250 psig 1,375 psig). The Technical Specifications (TS) Limiting Condition for Operation (LCO) for the SRVs and SVs ensures that the acceptance limit of 1,375 psig is met during the Design Basis Event. From an overpressure standpoint, the design basis events are bounded by the MSIV closure with flux scram event described above.

License Amendment Request Attachment 1 Revise TS For Two SRVs/SVs Out-of-Service Page 9of13 Docket Nos. 50-277 and 50-278 Evaluation of Proposed Changes Although not a design basis event, the Anticipated Transient Without Scram (ATWS) analysis demonstrates that peak vessel bottom pressure is less than the ASME Code Service Level C limit of 1,500 psig, which is the ATWS acceptance criterion for overpressure protection.

The Safety Limit (SL) on reactor steam dome pressure protects the Reactor Coolant System (RCS) against overprotection. For purposes of the analysis, a total of 12 SRVs and SVs are assumed to operate in the safety mode. The analysis results demonstrate that the design SRV and SV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure with margin and supports raising the RCS SL from 1,325 to 1,340 psig to support operation with two SRVs/SVs out-of-service.

4.2 Precedence There is no specific precedent for lowering the number of SRVs/SVs required to be operable. However, the NRC has issued the following amendments to Facility Operating Licenses allowing operation with increased RCS SL of at least 1,340 psig.

1. Pilgrim Nuclear Power Station, Amendment 235, dated March 28, 2011 (ADAMS Accession No. ML110650009). The RCS SL was increased to 1340 psig.

2. Dresden Nuclear Power Station, Unit 2, Amendment 75, dated April 7, 1983 (ADAMS Accession No. ML021150150). The RCS SL was increased to 1345 psig.

3. Quad Cities Nuclear Power Station, Unit 1, Amendment 83, dated December 15, 1982 (ML020940289). The RCS SL was increased to 1345 psig.

4.3 No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC), proposes a change to the Technical Specifications (TS), Appendix A of the Renewed Facility Operating License Nos. DPR-44 and DPR-56 for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3.

The proposed amendment would revise TS Section 3.4.3 to lower the required number of operable Safety Relief Valves (SRVs) and Safety Valves (SVs) from a total of 13 to 12 and raise the Reactor Coolant System (RCS) Safety Limit (SL) from 1,325 to 1,340 psig.

EGC has evaluated the proposed changes, using the criteria in 10 CFR 50.92, "Issuance of amendment," and has determined that the proposed changes do not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards consideration.

License Amendment Request Attachment 1 Revise TS For Two SRVs/SVs Out-of-Service Page 10of13 Docket Nos. 50-277 and 50-278 Evaluation of Proposed Changes

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change would revise TS Section 3.4.3 to lower the required number of operable Safety Relief Valves (SRVs) and Safety Valves (SVs) from a total of 13 to 12, and raise the Reactor Coolant System (RCS) Safety Limit (SL) from 1,325 to 1,340 psig. Analysis confirms that raising the number of out-of-service SRVs/SVs from one to two does not have an adverse impact on 1) the overpressure protection for the reactor pressure vessel , 2) the ability of High Pressure Coolant Injection (HPCI), Reactor Core Isolation Cooling (RCIC),

Standby Liquid Control (SLC) and Control Rod Drive (CRD) safety systems to perform their design basis requirements, and 3) the Emergency Core Cooling System (ECCS) Loss of Coolant Accident (LOCA) analysis. The analysis also confirms that for the peak vessel pressure in the overpressure event, there is still over 20 psi margin to the American Society of Mechanical Engineer (ASME) code overpressure limit. This margin also includes the penalty due to the TRACG statistical pressure adder required to be included in the analysis results, thereby providing additional analytical margin. Raising the dome pressure safety limit by 15 psi from 1,325 to 1,340 psig still provides sufficient margin (approximately 5 psi) for the peak pressure vessel pressure and thus continues to support the ASME code overpressure limit requirements. Compliance with the ASME upset code requirements for vessel overpressure protection is still ensured with this change to the dome pressure safety limit (1,340 psig) to support operation with two SRVOOS.

This analysis covers the plant response to the design basis accidents, Anticipated Operational Occurrence (AOO) events and Special Events. The proposed change does not require any new or unusual operator actions. The proposed change does not introduce any new failure modes that could result in a new or different accident. The SRVs and SVs are not being modified or operated differently and will continue to operate to meet the design basis requirements for RPV overpressure protection. The proposed change does not alter the manner in which the RPV overpressure protection system is operated and functions and thus, there is no significant impact on reactor operation. There is no change being made to safety limits or limiting safety system settings that would adversely affect plant safety as a result of the proposed change.

For PBAPS, the limiting overpressure AOO event is the main steam isolation valve closure with scram on high flux (MSIVF). The PBAPS A TWS Special Event analysis considered the limiting cases for RPV overpressure and is analyzed under two cases: (1) Main Steam Isolation Valve Closure (MSIVC) and (2)

Pressure Regulator Failure Open (PRFO). These events were analyzed under the proposed conditions and it was confirmed that the existing analyses remain bounding for the condition of adding a second SRV/SV Out-of-Service.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

License Amendment Request Attachment 1 Revise TS For Two SRVs/SVs Out-of-Service Page 11 of 13 Docket Nos. 50-277 and 50-278 Evaluation of Proposed Changes

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed change would revise TS Section 3.4.3 to lower the required number of operable SRVs and SVs from a total of 13 to 12, and raise the RCS SL from 1,325 to 1,340 psig. The RPV overpressure protection capability of the 12 operable SRVs and SVs is adequate to ensure the ASME code allowable peak pressure limits are not exceeded. The SRVs and SVs are not being modified or operated differently and will continue to operate to meet the design basis requirements for RPV overpressure protection. The proposed change does not introduce any new failure modes that could result in a new or different accident.

The proposed change does not alter the manner in which the RPV overpressure protection system is operated and functions and thus, there is no new failure mechanisms for the overpressure protection system . The plant response to the design basis accidents, AOO events and Special Events remains bounded by existing analyses. These events were analyzed under the proposed conditions and it was confirmed that the existing analyses remain bounding for the condition of adding a second SRV/SV Out-of-Service at the current licensed thermal power.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response : No The margin of safety is established through the design of the plant structures, systems and components, the parameters within which the plant is operated, and the establishment of setpoints for the actuation of equipment relied upon to respond to an event. The proposed change does not change the setpoints at which the protective actions are initiated. The proposed change would revise TS Section 3.4.3 to lower the required number of operable SRVs and SVs and raise the RCS SL from 1,325 to 1,340 psig. The RPV overpressure protection capability of the 12 operable SRVs and SVs is adequate to ensure the ASME code allowable peak pressure limits are not exceeded. The plant response to the design basis accidents, AOO events and Special Events remains bounded by existing analyses. These events were analyzed under the proposed conditions and it was confirmed that for the peak vessel pressure in the overpressure event, there is still over 20 psi margin to the American Society of Mechanical Engineer (ASME) code overpressure limit. This margin also includes the penalty due to the TRACG statistical pressure adder required to be included in the analysis results, thereby providing additional analytical margin. Raising the dome pressure safety limit by 15 psi from 1,325 to 1,340 psig still provides sufficient margin for the peak pressure vessel pressure and thus continues to support the ASME code overpressure limit requirements. Compliance with the ASME upset code

License Amendment Request Attachment 1 Revise TS For Two SRVs/SVs Out-of-Service Page 12of13 Docket Nos. 50-277 and 50-278 Evaluation of Proposed Changes requirements for vessel overpressure protection is still ensured with this change to the dome pressure safety limit (1,340 psig) to support operation with two SRVOOS.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above evaluation, EGC concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92, paragraph (c}, and accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

EGC has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, "Standards for Protection Against Radiation ." However, the proposed amendment does not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review,"

paragraph (c)(9). Therefore, pursuant to 10 CFR 51.22, paragraph (b}, no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

License Amendment Request Attachment 1 Revise TS For Two SRVs/SVs Out-of-Service Page 13of13 Docket Nos. 50-277 and 50-278 Evaluation of Proposed Changes

6.0 REFERENCES

1. NRC letter to Exelon, "Peach Bottom Atomic Power Station, Units 2 and 3 -

Issuance of Amendments Re: Measurement Uncertainty Recapture Power Uprate (CAC Nos. MF9289 and MF9290; EPID L-2017-LLS-0001)," dated November 15, 2017 (ADAMS Accession No. ML17286A013).

2. NRC letter to Exelon, "Peach Bottom Atomic Power Station, Units 2 and 3 -

Issuance of Amendments Re: Maximum Extended Load Line Limit Analysis Plus (CAC NOS. MF4760 AND MF4761)," dated March 21, 2016 (ADAMS Accession No. ML16034A372).

3. Global Nuclear Fuel, "Supplemental Reload Licensing Report for Peach Bottom Unit 2 Reload 21 Cycle 22 Mid-Cycle Thermal Power Optimization (TPO),"

004N2488, Revision 0, October 2017.

4. Global Nuclear Fuel, "Supplemental Reload Licensing Report for Peach Bottom Unit 3 Reload 21 Cycle 22," 003N1452, Revision 0, September 2017.

5. GE Hitachi Nuclear Energy, "Migration to TRACG04 I PANAC11 from TRACG02 I PANAC10 for TRACG AOO and ATWS Overpressure Transients," NEDE-32906P Supplement 3-A, Revision 1, April 2010.

6. GE Nuclear Energy, "TRACG Application for Anticipated Operational Occurrences (AOO) Transient Analyses," NEDE-32906P-A, Revision 3, September 2006.

7. Letter, J. F. Harrison (GEH) to NRC Document Control Desk, "Event-Specific LiCPR/ICPR Biases and Uncertainties and Peak Pressure Adders for AOO Licensing Applications," MFN 16-030, May 12, 2016(ML16133A082 and ML16133A083).

8. Letter, J. F. Harrison (GEH) to NRC Document Control Desk, "Event-Specific LiCPR/ICPR Biases and Uncertainties and Peak Pressure Adders for AOO Licensing Applications," M180061, March 23, 2018 (ML18082A047 and ML18082A048).

9. GE Nuclear Energy, "TRACG Application for Anticipated Transient Without Scram Overpressure Transient Analyses," NEDE-32906P Supplement 1-A, November 2003.

10. NRC letter to Exelon, "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 (ADAMS Accession No. ML17249A151).

ATTACHMENT 2 License Amendment Request Peach Bottom Atomic Power Station, Units 2 and 3 Docket Nos. 50-277 and 50-278 License Amendment Request - Revise Technical Specifications to Allow Two Safety Relief Valves/Safety Valves to be Out-of-Service with Increased Reactor Pressure Safety Limit Markup of Proposed Technical Specifications Page Unit 2 and Unit 3 TS Pages 2.0-1 3.4-8

SLs 2.0 2.0 SAFETY LIMITS CSLs)

2. 1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure< 700 psia or core flow< 10% rated core flow:

THERMAL POWER shall bes 22.6% RTP.

2.1.1.2 With the reactor steam dome pressure~ 700 psia and core flow~ 10% rated core flow:

MCPR shall be~ 1.15 for two recirculation loop operation or ~ 1.15 for single recirculation loop operation.

2 .1.1. 3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

~

2 .1. 2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be s ~ psig.

2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

continued PBAPS UN IT 2 2.0-1 Amendment No. 316

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 I3- 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 OPERABLE~ 3358 MWt during operating cycle 22.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.l 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 UNIT 2 3.4-8 Amendment No. 315

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.l Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure< 700 psia or core flow< 10% rated core flow:

THERMAL POWER shall be~ 22.6% RTP.

2.1.1.2 With the reactor steam dome pressure~ 700 psia and core flow~ 10% rated core flow:

MCPR shall be~ 1.15 for two recirculation loop operation or~ 1.15 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant System Pressure SL ~

Reactor steam dome pressure shall be~ ~ psig.

2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.l Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

continue PBAPS UNIT 3 2.0-1 Amendment No. 319

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 +/- valves (any combination of SRVs and SVs) shall be OPERABLE.

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 UNIT 3 3.4-8 Amendment No. 296

ATTACHMENT 3 License Amendment Request Peach Bottom Atomic Power Station, Units 2 and 3 Docket Nos. 50-277 and 50-278 License Amendment Request - Revise Technical Specifications to Allow Two Safety Relief Valves/Safety Valves to be Out-of-Service with Increased Reactor Pressure Safety Limit Markup of Proposed Technical Specifications Bases Pages Unit 2 and Unit 3 TS Bases Page B 2.0-8 B 2.0-10 B 3.4-16 B 3.4-18

RCS Pressure SL B 2.1.2 The SL has been determined to be adequate to ensure the RCS pressure does not exceed the 1375 psig RCS pressure limit (Refs. 7 and 8) .

BASES APPLICABLE The RCS pressure SL has been selected such that it is a SAFETY ANALYSES pressure below which it can be shown that the integrity (continued) the system is not endangered. The reactor pressure vess is designed to Section III, 1965 Edition of the ASME, Bo and Pressure Vessel Code, including Addenda through the 1340 psig is winter of 1965 (Ref. 5), which permits a maximum pressure tr

• n of l10%, 1375 psi g, of design pressure 1250 psi The SL o 132& psig, as measured in the reactor steam domeT is eqijivalent tG 137& psig at the lgwest elevatiGn gf the

~ The RCS is designed to the ASME Section III, 1980 Edition, including Addenda through winter of 1981 (Ref. 6),

for the reactor recirculation piping, which permits a maximum pressure transient of 110% of design pressures of 1250 psig for suction piping and 1500 psig for discharge piping. The RCS pressure SL is selected to be the lowest transient overpressure allowed by the applicable codes.

SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure vessel under the ASME Code,Section III, is 110% of design pressure. The maximum transient pressure allowable in the RCS piping, valves, and fittings is 110% of design pressures of 1250 psig for suction piping and 1500 psig for discharge piping. The most limiting of these allowances is the 110 ~

of design pressures of 1250 psig; therefore, the SL o ~ ~

maximum allowable RCS pressure is established at~ psig, as measured at the reactor steam dome.

APPLICABILITY SL 2.1.2 applies in all MODES.

SAFETY LIMIT VIOLATIONS continu d P8APS UN IT 2 B 2.0 -8 Revision No. 57

RCS Pressure SL B 2.1.2 BASES REFERENCES 3. ASME, Boiler and Pressure Vessel Code,Section XI, (continued) Article IW-5000.

4. 10 CFR 50.67.

5. ASME, Boiler and Pressure Vessel Code,Section III, 1965 Edition, including Addenda to winter of 1965.

6. ASME, Boiler and Pressure Vessel Code,Section III, 1980 Edition, Addenda to winter of 1981.

7. G-080-VC-413, "Reactor Vessel Overpressure Protection ," GE Hitachi Nuclear Energy, 26AB321 , Revision 1.

8. G-080-VC-468, "Peach Bottom Units 2 & 3 Two Safety Relief Valves Out-of-Service Evaluation ," GE Hitachi Nuclear Energy, 004N6240-P, Revision 1.

PBAPS UN IT 2 B 2.0-10 Revision No. 75

SRVs and SVs B 3.4.3 BASES (continued) Refs. 1, 4 and 5 APPLICABLE The overpressure protection s stem must ace modate the most SAFETY ANALYSES severe pressurization transi nt. Evaluatio s have determined that the most sev re transient i the closure of all main steam isolation va ves (MSIVs), fo lowed by reactor scram on high neutron flux (i.e., failure the direct scram associated with MSIV osition) ( Ref, 1). For the purpose of the analyses, ~ SRVs and SVs are assumed to operate in the safety mode. The analysis results demonstrate that the design SRV and SV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure (110% x 1250 psig =

1375 psig). This LCO helps to ensure that the acceptance limit of 1375 psig is met during the Design Basis Event.

From an overpressure standpoint, the design basis events are bounded by the MSIV closure with flux scram event described above. Reference 2 discusses additional events that are expected to actuate the SRVs and SVs. Although not a design basis event, the ATWS analysis demonstrates that peak vessel bottom pressure is less than the ASME Service Level C limit of 1,500 psig.

SRVs and SVs satisf Criterion 3 of the NRC Policy Statement. 1, 2, 4 and 5 LCO The safety function of a combination of ~ SRVs and SVs are required to be OPERAB C to satisfy the assumptions of the safety analysis (Refs. ). Regarding the SRVs, the requirements of this LCO are applicable only to their capability to mechanically open to relieve excess pressure when the lift setpoint is exceeded (safety mode).

The SRV and SV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied. The ASME Code specifications require the lowest safety valve setpoint to be at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressurization conditions. The transient evaluations in the UFSAR are based on these setpoints, but also include the additional uncertainties of

+ 3% of the nominal setpoint to provide an added degree of conservatism.

Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being exceeded.

(continued)

PBAPS UN IT 2 B 3.4-16 Revision No. 142

SRVs and SVs B 3.4.3 BASES SURVEILLANCE SR 3.4.3.2 REQUIREMENTS (continued) The pneumatic actuator of each SRV valve is stroked to verify that the second stage pilot disc rod is mechanically displaced when the actuator strokes. Second stage pilot rod movement is determined by the measurement of actuator rod travel. The total amount of movement of the second stage pilot rod from the valve closed position to the open position shall meet criteria established by the SRV supplier. If the valve fails to actuate due only to the failure of the solenoid, but is capable of opening on overpressure, the safety function of the SRV is considered OPERABLE.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. NEDC-33566P, "Safety Analysis Report for Exel on Peach Bottom Atomic Power Station, Units 2 and 3, Constant Pressure Power Uprate," Revision 0.

2. UFSAR, Chapter 14.

3. NEDC-32988-A, Revision 2, Technical Justification to Support Risk-Informed Modification to Se l ected Required

~ End States for BWR Plants, December 2002.

4. G-080-VC-413, "Reactor Vessel Overpressure Protection," GE Hitachi Nuclear Energy, 26A8321, Revision 1.

5. G-080-VC-468, "Peach Bottom Units 2 & 3 Two Safety Relief Valves Out-of-Service Evaluation," GE Hitachi Nuclear Energy, 004N6240-P, Revision 1.

PBAPS UN IT 2 B 3.4-18 Revision No. 114

RCS Pressure SL

__T_h_e_S_L_h_a_s_b_e_e_n_d_e-te_r_m_i_n-ed-to_b_e_a_d_e_q_u-at_e_t_o_e_n_s_u_re-th_e_R_C_S_p_r_e_ss_u_r_e_d_o_e_s_., B 2

• 1. 2 not exceed the 1375 psig RCS pressure limit (Refs. 7 and 8).

BASES 1340 psig is APPLICABLE The RCS pressure SL has been sel cted such that i is at a SAFETY ANALYSES pressure below which it can be hown that the in egrity of (continued) the system is not endangered. The reactor pres ure vessel is designated to Section III, 965 Edition of he ASME, Boiler and Pressure Vessel C e, including Ad enda through the summer of 1966 (Ref. 5), which permits a pressure transient of 110% 1375 psig, of d 1250 psig. The SL of 1325 psi§ , as measur in the reactor steam dome, is eq~ivalent to 1375 psi§ at tAe lowest elevation of tAe RCS . The RCS is designed to ASME Section III, including Addenda through the winter of 1981 (Ref. 6),

for the reactor recirculation piping, which permits a maximum pressure transient of 110% of design pressures of 1250 psig for suction piping and 1500 psig for discharge piping. The RCS pressure SL is selected to be the lowest transient overpressure allowed by the applicable codes.

SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure vessel under the ASME Code,Section III, is 110% of design pressure. The maximum transient pressure allowable in the RCS piping, valves, and fittings is 110% of design pressures of 1250 psig for suction piping and 1500 psig for discharge piping. The most limiting of these allowances is the 110~

of design pressures of 1250 psig; therefore, the SL o ~ '----------'

maximum allowable RCS pressure is established at ~ psig as measured at the reactor steam dome.

APPLICABILITY SL 2.1.2 applies in all MODES.

SAFETY LIMIT VIOLATIONS PBAPS UNIT 3 B 2.0-8 Revision No. 58

RCS Pressure SL B 2.1.2 BASES REFERENCES 3. ASME, Boil er and Pressure Vessel Code,Section XI, (continued) Article IW-5000.

4. 10 CFR 50.67.

5. ASME, Boiler and Pressure Vessel Code,Section III, 1965 Edition, including Addenda to summer of 1966.

6. ASME, Boiler and Pressure Vessel Code,Section III, 1980 Edition, Addenda to winter of 1981.

7. G-080-VC-413, "Reactor Vessel Overpressure Protection," GE Hitachi Nuclear Energy, 26A8321, Revision 1.

8. G-080-VC-468, "Peach Bottom Units 2 & 3 Two Safety Relief Valves Out-of-Service Evaluation," GE Hitachi Nuclear Energy, 004N6240-P, Revision 1.

PBAPS UN IT 3 B 2.0-10 Revision No. 76

SRVs and SVs B 3.4.3 Refs . 1, 4 and 5 BASES (continued)

APPLICABLE The overpressure protection stem must ac ommodate the most SAFETY ANALYSES severe pressurization transi nt. Evaluat*ons have determined that the most sev re transient is the closure of all main steam isolation va ves (MSIVs), fo ll owed by reactor scram on high neutron flux i.e., fail u of the direct scram associated with MSIV osition) ( Ref . 1 ) . For the purpose of the analyses, SRVs and SVs are assumed to operate in the safety mode . The analysis results demonstrate that the design SRV and SV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure (110% x 1250 psig =

1375 psig). This LCO helps to ensure that the acceptance limit of 1375 psig is met during the Design Basis Event.

From an overpressure standpoint, the design basis events are bounded by the MSIV closure with flux scram event described above. Reference 2 discusses additional events that are expected to actuate the SRVs and SVs. Although not a design basis event, the ATWS analysis demonstrates that peak vessel bottom pressure is less than the ASME Service Level C limit of 1,500 psig .

SRVs and SVs satisfy Criterion 3 of the NRC Policy Statement. d 1, 2 . 4 an 5 LCO The safety function any combination of -?r SRVs and SVs are required to be OPE LE to satisfy the assumptions of the safety analysis (Refs. 1 aAd 2). Regarding the SRVs, the requirements of this LCO are applicable only to their capability to mechanically open to relieve excess pressure when the lift setpoint is exceeded (safety mode) .

The SRV and SV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied. The ASME Code specifications require the lowest safety valve setpoint to be at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressurization conditions . The transient evaluations in the UFSAR are based on these setpoints, but also include the additional uncertainties of

+ 3% of the nominal setpoint to provide an added degree of conservatism.

Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being exceeded.

(continued)

PBAPS UNIT 3 B 3.4-16 Revision No. 119

SRVs and SVs B 3.4.3 BASES SURVEILLANCE SR 3.4.3.2 REQUIREMENTS (continued) The pneumatic actuator of each SRV valve is stroked to verify that the second stage pilot disc rod is mechanically displaced when the actuator strokes. Second stage pilot rod movement is determined by the measurement of actuator rod travel. The total amount of movement of the second stage pilot rod from the valve closed position to the open position shall meet criteria established by the SRV supplier. If the valve fails to actuate due only to the failure of the solenoid, but is capable of opening on overpressure, the safety function of the SRV is considered OPERABLE.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. NEDC-33566P, "Safety Analysis Report for Exelon Peach Bottom Atomic Power Station, Units 2 and 3, Constant Pressure Power Uprate," Revision 0.

2. UFSAR, Chapter 14.

3. NEDC-32988-A, Revision 2, Technical Justification to Support Risk-Informed Modification to Selected Required

~ nd States for BWR Plants, December 2002.

4. G-080-VC-413, "Reactor Vessel Overpressure Protection," GE Hitachi Nuclear Energy, 26A8321, Revision 1.

5. G-080-VC-468, "Peach Bottom Units 2 & 3 Two Safety Relief Valves Out-of-Service Evaluation," GE Hitachi Nuclear Energy, 004N6240-P, Revision 1.

PBAPS UNIT 3 B 3.4 -18 Revision No. 119

ATTACHMENT 5 License Amendment Request Peach Bottom Atomic Power Station, Units 2 and 3 Docket Nos. 50-277 and 50-278 License Amendment Request - Revise Technical Specifications to Allow Two Safety Relief Valves/Safety Valves to be Out-of-Service with Increased Reactor Pressure Safety Limit GE Hitachi Nuclear Energy Affidavit Supporting Withholding Attachment 4 from Public Disclosure

9 HITACHI GE Hitachi Nuclear Energy 004N6240-P Revision 1 March 2018 GEH Proprietary Information - Class II (Internal)

Peach Bottom Units 2 & 3 Two Safety Relief Valves Out-of-Service Evaluation Copyright 2018, GE-Hitachi Nuclear Energy Americas LLC All Rights Reserved

004N6240-P Revision 1 GEH Proprietary Information - Class II (Internal)

PROPRIETARY INFORMATION NOTICE This document contains proprietary information of GE-Hitachi Nuclear Energy Americas, LLC (GEH) and is furnished in confidence solely for the purpose(s) stated below in the notice regarding the contents of this report. No other use, direct or indirect, of the document or the information it contains is authorized. Furnishing this document does not convey any license, express or implied, to use any patented invention or, except as specified above, any GEH proprietary information disclosed herein or any right to publish the document without prior written permission from GEH.

The header of each page in this document carries the notation "GEH Proprietary Information -

Class II (Internal)." GEH proprietary information within the text and tables is identified by a dotted underline inside double square brackets. ((Ihi~--~~n!~m:_~j~--~D--~-~-~mpJ.~,_f_~~)) In all cases, the superscript notation 13 1 refers to Paragraph (3) of the enclosed affidavit that provides the basis for the proprietary determination.

IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The design, engineering, and other information contained in this document are furnished in accordance with the contract between Exelon and GEH, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than Exelon, or for any purpose other than that for which it is furnished by GEH is not authorized; and with respect to any unauthorized use, GEH makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.

11

GE-Hitachi Nuclear Energy Americas LLC AFFIDAVIT I, Lisa K. Schichlein, state as follows:

(1) I am a Senior Project Manager, NPP/Services Licensing, Regulatory Affairs, GE-Hitachi Nuclear Energy Americas LLC (GEH), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in GEH proprietary report 004N6240-P, "Peach Bottom Units 2 & 3 Two Safety Relief Valves Out-of-Service Evaluation,"

Revision 1, dated March 2018. GEH proprietary information in 004N6240-P Revision 1 is identified by a dotted underline inside double square brackets. [lT.hi~---~~nt~n9~.J~_Jm

~~w.m~l~,~~-lJ]. GEH proprietary information in figures and large objects is identified by double square brackets before and after the object. In each case, the superscript notation t3 l refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.

(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 U.S.C. §552(b)(4), and the Trade Secrets Act, 18 U.S.C.

§1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for trade secrets (Exemption 4). The material for which exemption from disclosure is here sought also qualifies under the narrower definition of trade secret, within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975 F.2d 871 (D.C. Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F .2d 1280 (D.C. Cir. 1983).

(4) The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a and (4)b. Some examples of categories of information that fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEH's competitors without a license from GEH constitutes a competitive economic advantage over other companies;

b. Information that, if used by a competitor, would reduce its expenditure of resources or improve its competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information that reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEH; 004N6240-P Revision I Affidavit Page I of 3

GE-Hitachi Nuclear Energy Americas LLC

d. Information that discloses trade secret or potentially patentable subject matter for which it may be desirable to obtain patent protection.

(5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GEH, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GEH, not been disclosed publicly, and not been made available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions for proprietary or confidentiality agreements or both that provide for maintaining the information in confidence. The initial designation of this information as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in the following paragraphs (6) and (7).

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, who is the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or who is the person most likely to be subject to the terms under which it was licensed to GEH.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GEH are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary and/or confidentiality agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because it contains detailed results and conclusions regarding supporting evaluations for a GEH Boiling Water Reactor (BWR). The analysis utilized GEH data and testing, and analytical models and methods, including computer codes, which GEH has developed, obtained NRC approval of, and applied to perform evaluations for a GEH BWR. The accumulation of this data and the development of the evaluation process along with the interpretation and application of the analytical results were derived from the extensive experience database that constitutes a major GEH asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GEH's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.

The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply 004N6240-P Revision I Affidavit Page 2 of 3

GE-Hitachi Nuclear Energy Americas LLC the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GEH. The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial. GEH's competitive advantage will be lost if its competitors are able to use the results of the GEH experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GEH would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on this 20th day of March 2018.

Lisa K. Schichlein Senior Project Manager, NPP/Services Licensing Regulatory Affairs GE-Hitachi Nuclear Energy Americas LLC 3901 Castle Hayne Road Wilmington, NC 28401 Lisa.Schichlein@ge.com 004N6240-P Revision 1 Affidavit Page 3 of 3

ATTACHMENT 6 License Amendment Request Peach Bottom Atomic Power Station, Units 2 and 3 Docket Nos. 50-277 and 50-278 License Amendment Request - Revise Technical Specifications to Allow Two Safety Relief Valves/Safety Valves to be Out-of-Service with Increased Reactor Pressure Safety Limit GE Hitachi Nuclear Energy 004N6240-NP, "Peach Bottom Units 2 & 3 Two Safety Relief Valves Out-of-Service Evaluation," Revision 1, (Non-Proprietary Version)

- HITACHI GE Hitachi Nuclear Energy 004N6240-NP Revision 1 March 2018 Non-Proprietary Information - Class I (Public))

Peach Bottom Units 2 & 3 Two Safety Relief Valves Out-of-Service Evaluation Copyright 2018, GE-Hitachi Nuclear Energy Americas LLC All Rights Reserved

004N6240-NP Revision 1 Non-Proprietary Information - Class I (Public)

INFORMATION NOTICE This is a non-proprietary version of the document 004N6240-P Revision 1, which has the proprietary information removed. Portions of the document that have been removed are indicated by an open and closed bracket as shown here (( )).

IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The design, engineering, and other information contained in this document are furnished in accordance with the contract between Exelon and GEH, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than Exelon, or for any purpose other than that for which it is furnished by GEH is not authorized; and with respect to any unauthorized use, GEH makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.

II

004N6240-NP Revision 1 Non-Proprietary Information - Class I (Public)

REVISION

SUMMARY

Revision Revision Summary 0 Initial release 1 Revised to mark specific instances of GEH proprietary information Ill

004N6240-NP Revision 1 Non-Proprietary Information - Class I (Public)

TABLE OF CONTENTS

1.0 INTRODUCTION

............................................................................................................... 1 1.1 Purpose ........................................................................................................................ 1 1.2 Scope ............................................................................................................................ 1 2.0 TRANSIENT PRESSURE ANALYSIS ............................................................................ 1 2.1 ASME Overpressure Analysis ..................................................................................... 1 2.1.1 Analysis Overview ............................................................................................ 1 2.1.2 Analysis Results ................................................................................................ 2 2.1.3 Conclusion ........................................................................................................ 3 2.2 A TWS Overpressure Analysis ..................................................................................... 4 2.2.1 Analysis Overview ............................................................................................ 4 2.2.2 Analysis Results ................................................................................................ 4 2.2.3 Conclusion ........................................................................................................ 5 3.0 SYSTEMS EVALUATION ................................................................................................ 5 3.1 HPCI and RCIC Evaluation ......................................................................................... 5 3.2 SLC Evaluation ............................................................................................................ 5 3.3 CRD Evaluation ........................................................................................................... 6 4.0 ECCS LOCA ANALYSIS .................................................................................................. 6 4.1 Analysis Overview and Results ................................................................................... 6 4.2 Conclusion ................................................................................................................... 7 5.0 ACRONYMS ....................................................................................................................... 8

6.0 REFERENCES

.................................................................................................................. 10 LIST OF TABLES Table 1 PBAPS Unit 2 Cycle 22 TPO Overpressure Analysis Results .................................... 2 Table 2 PBAPS Unit 3 Cycle 22 TPO Overpressure Analysis Results .................................... 3 Table 3 Beginning of Cycle (BOC) MSIVC Event Sequence ................................................ 11 Table 4 BOC PRFO Event Sequence ..................................................................................... 11 LIST OF FIGURES Figure 1 BOC MSIVC (Short-term) ........................................................................................ 12 Figure 2 BOC PRFO (Short-term) ........................................................................................... 13 iv

004N6240-NP Revision I Non-Proprietary Information - Class I (Public)

1.0 INTRODUCTION

1.1 Purpose This report provides the GE Hitachi Nuclear Energy (GEH) evaluation of a proposed change to allow two Safety Relief Valves Out-of-Service (SRVOOS) at Peach Bottom Atomic Power Station (PBAPS) Units 2 and 3.

1.2 Scope The scope of this evaluation includes the following:

1. A transient safety analysis to determine the effect of two SRVOOS on the American Society of Mechanical Engineers (ASME) overpressure and Anticipated Transient Without Scram (ATWS) overpressure results for PBAPS. The TRACG statistical pressure adder was recalculated for GNF2 fuel for use in the ASME overpressure analysis for PBAPS.

GEH addresses the effect of having two SRVOOS on Abnormal Operational Occurrence (AOO) transients and other ATWS criteria. This analysis is based on the Thermal Power Optimization (TPO) power level and considers the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) operating domain.

2. An evaluation to demonstrate that the existing safety analysis (based on the 1,325 psig dome pressure limit) is overly conservative for fulfilling the ASME upset requirement vessel bottom pressure limit of 1,375 psig. The results of this evaluation will inform Exelon's decision to raise the dome pressure safety limit to some value higher than 1,325 psig.

3. A qualitative evaluation of the effect two SRVOOS will have on the ability of the following systems to meet design basis requirements: High Pressure Coolant Injection (HPCI),

Reactor Core Isolation Cooling (RCIC), Standby Liquid Control (SLC), and Control Rod Drive (CRD).

4. A qualitative analysis of Loss-of-Coolant Analysis (LOCA).

2.0 TRANSIENT PRESSURE ANALYSIS 2.1 ASME Overpressure Analysis This section describes the effect of the additional SRVOOS on the PBAPS ASME overpressure analysis.

2.1.1 Analysis Overview An evaluation was performed for PBAPS to determine the effect of having two SRVOOS on the ASME required analysis for reactor vessel overpressure protection. The TRACG statistical pressure adder was recalculated for GNF2 fuel for use in the ASME overpressure analysis for PBAPS. This analysis is based on the TPO power level and considers the MELLLA+ operating domain.

This analysis was performed using the TRACG AOO methodology (References I and 2) at the currently licensed TPO power level of 4,016 MW th. The limiting overpressure event is the Main

004N6240-NP Revision 1 Non-Proprietary Information - Class I (Public)

Steam Isolation Valve Closure with Flux Scram (MSIVF) event. This event was analyzed with two SRVOOS in accordance with the methodologies described in References 2 and 3.

The results from this evaluation will help inform Exelon's decision to increase the Technical Specification (TS) dome pressure safety limit for both units if it is demonstrated that the existing 1,325 psig dome pressure limit is overly conservative for fulfilling the ASME upset condition requirements.

2.1.2 Analysis Results The overpressure analyses were performed with two SRVOOS using the most recent reload licensing analysis inputs for PBAPS Units 2 and 3. The most recent MSIVF transients for PBAPS (References 4 and 5) were analyzed with one SRVOOS for the current Safety Relief Valve (SRV)

I Safety Valve (SY) configuration.

The PBAPS Unit 2 results for one SRVOOS from Reference 4 and the results from the new analysis performed with two SRVOOS are provided in Table l. The PBAPS Unit 3 results for one SRVOOS from Reference 5 and the results from the new analysis performed with two SRVOOS are provided in Table 2. The current one SRVOOS peak pressure results include a pressure adder from Reference 6. The new two SRVOOS peak pressure results include an updated pressure from Reference 7.

Table 1 PBAPS Unit 2 Cycle 22 TPO Overpressure Analysis Results Number Peak Dome Peak Vessel MSIVFPeak Limiting SRV/SV of Pressure Pressure Vessel Pressure Domain Configuration SRVOOS (psi2) (pshd Adder (osid)

Increased Core Flow (ICF) l 1,324 1,353 Current (Hard Bottom

((

Bum (HBB))

MELLLA+

l 1,325 1,349 Current (HBB)

ICF (HBB) 2 1,325 1,354 New MELLLA+ ))

2 1,326 1,351 New (HBB) 2

004N6240-NP Revision 1 Non-Proprietary Information - Class I (Public)

Table 2 PBAPS Unit 3 Cycle 22 TPO Overpressure Analysis Results Peak MSIVFPeak Number Peak Vessel Limiting Dome SRV/SV Vessel of Pressure Domain Pressure Configuration Pressure SRVOOS (psig)

(psie:) Adder (psid)

ICF (HBB) 1 1,322 1,352 Current MELLLA+ ((

l 1,324 l,349 Current (HBB)

ICF (HBB) 2 1,324 1,353 New MELLLA+ ))

2 1,327 1,351 New (HBB)

The reload licensing analysis results for one SRVOOS show that the margin to the current dome pressure safety limit of 1,325 psig is less than 5 psi; however, there is over 20 psi margin to the ASME overpressure limit of 1,375 psig. With two SRVOOS, the current dome pressure safety limit is violated for both PBAPS units; however, there is still over 20 psi margin to the ASME overpressure limit. ((

))

The safety limit is placed on the dome pressure to have a plant measurable parameter to demonstrate compliance with the vessel pressure limit of 1,3 75 psig. The results shown in Tables 1 and 2 demonstrate that there is, at most, a 30 psi difference between the calculated peak vessel pressure and peak dome pressure. This indicates that the current dome pressure safety limit, which is set to 1,325 psig, has -20 psi excess conservatism. Establishing a new dome pressure safety limit at 1,340 psig provides margin to the ASME overpressure limit of 1,3 75 psig based on an assessment of the calculated peak pressures in Tables l and 2. A dome pressure safety limit of 1,340 psig would provide a 35 psi margin to the ASME overpressure limit, which is 5 psi higher than the observed maximum pressure difference for the limiting ASME overpressure event.

2.1.3 Conclusion The MSIVF event was analyzed for TPO conditions using the most recent reload licensing analysis inputs for PBAPS Units 2 and 3 with two SRVOOS. ((

)) The peak vessel pressure and dome pressure results from the overpressure analysis were compared to the related ASME and TS limits, and were shown to support an increase to the dome pressure safety limit in the TS from 1,325 psig to 1,340 psig. Compliance with the ASME upset code requirements for vessel overpressure protection would still be ensured with this change to the dome pressure safety limit (1,340 psig) to support operation with two SRVOOS.

The results from this evaluation are not applicable for design use to support current operation until the TS dome pressure safety limit change has been approved by the Nuclear Regulatory Commission (NRC).

3

004N6240-NP Revision 1 Non-Proprietary Information - Class I (Public) 2.2 ATWS Overpressure Analysis This section describes the effect of the additional SRVOOS on the PBAPS ATWS overpressure analysis.

2.2.1 Analysis Overview The evaluation of the PBAPS response to postulated ATWS events is not a design basis requirement. However, evaluation is necessary to demonstrate that permitting full-power operation with two SRVOOS does not have an unacceptable effect on the mitigation capability of the plant during postulated A TWS events.

The peak vessel pressure analysis is performed because that is the ATWS acceptance criteria significantly affected by the SRV relief capacity decrease. The Peak Cladding Temperature (PCT),

peak containment pressure, and suppression pool temperature acceptance criteria are not affected by the additional SRVOOS. These long-term transient response parameters are driven by the core power generation and the relief capacity of the SRVs and SVs. Based on the ATWS analysis performed in support of the PBAPS TPO project (Reference 8), after the ((

)). This is well within the capacity of the remaining nine in-service SRVs.

Calculations were performed to determine the peak vessel pressure during an ATWS event to demonstrate compliance with the ASME Code Service Level C limit of 1,500 psig (Emergency Condition). The analysis was performed using the TRACG transient analysis code (References 1 and 9) with the reactor operating at the TPO bounding high power level (4,018 MWt) and limiting MELLLA+ core flow conditions. Two events were considered: Main Steam Isolation Valve Closure (MSIVC) and Pressure Regulator Failure Open (PRFO). Each of these events has the potential to yield the maximum vessel overpressure result. In the MSIVC event, the isolation valves on all four steam lines close simultaneously, while the normal direct scram on full MSIVC fails. In addition, scram on high neutron flux and high vessel pressure are assumed to fail. In the PRFO event, the pressure regulator failure produces the maximum steam flow demand. Reactor pressure drops and the Main Steam Isolation Valves (MSIVs) close on a low steam line pressure signal. Similarly, scram on full MSIV closure, high flux and high pressure all fail. The ultimate shutdown of the plant is accomplished through the actuation of the Standby Liquid Control System (SLCS). The initial reduction in power occurs as a result of the ATWS high dome pressure Recirculation Pump Trip (RPT). After the ATWS RPT and following the opening of the SRVs/SVs, the event is terminated for overpressure considerations.

2.2.2 Analysis Results The ATWS system response is simulated with two SRVOOS and the SRV/SV opening setpoints at +3% of the nominal setpoint. The resulting peak vessel pressure values for the MSIVC and PRFO events are (( )), respectively. These results are ((

)) as compared to the PBAPS TPO results (Reference 8), but are still below the 1,500 psig limit. The short-term sequence of events are given in Tables 3 and 4 and the transient response plots are shown in Figures 1 and 2.

4

004N6240-NP Revision 1 Non-Proprietary Information - Class I (Public) 2.2.3 Conclusion During the most limiting ATWS event, the peak vessel pressure does not exceed the licensing basis criterion of 1,500 psig. Therefore, with respect to ATWS overpressure, technical justification that both PBAPS Units 2 and 3 may operate with two SRVOOS and a +3% SRV/SV setpoint tolerance is demonstrated.

3.0 SYSTEMSEVALUATION This section describes the qualitative evaluation of the effect two SRVOOS will have on the ability of the HPCI, RCIC, SLC, and CRD systems to meet design basis requirements.

3.1 HPCI and RCIC Evaluation The HPCI system is designed to pump water into the reactor vessel over a wide range of operating pressures. The primary purpose of the HPCI system is to maintain reactor vessel coolant inventory in the event of a small break LOCA that does not immediately depressurize the reactor vessel. In this event, the HPCI system maintains reactor water level and helps depressurize the reactor vessel.

The adequacy of the HPCI system is demonstrated in the Emergency Core Cooling System (ECCS) performance discussion in Section 4.0. HPCI also services as a backup to the RCIC system. The HPCI backup function, for a RCIC system failure, is unaffected because the injection rate of the HPCI system is significantly greater than the RCIC system. The HPCI system is required to provide injection into the reactor pressure vessel at the lowest group of SRV setpoints including the +3% tolerance drift providing there are at least two functional SRVs in the lowest group.

Because there are four SRVs in the lowest group, taking two SRVOOS in this group still leaves two SRVs. Therefore, the HPCI system injection capability is not affected by an additional SRVOOS.

The RCIC system is required to maintain sufficient water inventory in the reactor to permit adequate core cooling following a reactor vessel isolation event accompanied by loss of flow from the feedwater system. The system design injection rate must be sufficient to comply with the system limiting criteria to maintain the reactor water level above Top of Active Fuel (T AF) at TPO conditions. The RCIC system is designed to pump water into the reactor vessel over a wide range of operating pressures. The RCIC system is required to provide injection into the reactor pressure vessel at the lowest group of SRV setpoints including the+3% tolerance drift providing there are at least two functional SRVs in the lowest group. Because there are four SRVs in the lowest group, taking two SRVOOS in this group still leaves two SRVs. With two SRVOOS ((

)). Therefore, the PBAPS RCIC system injection capability is adequate to support two SRVOOS.

3.2 SLC Evaluation SLCS is designed to shutdown the reactor from rated power conditions to cold shutdown in the postulated situation that some or all of the control rods cannot be inserted. This manually operated system pumps a highly enriched sodium pentaborate solution into the reactor vessel to provide 5

004N6240-NP Revision 1 Non-Proprietary Information - Class I (Public) neutron absorption and achieve a subcritical reactor condition. SLCS is designed to inject over a wide range of reactor operating pressures. The original SLCS design criteria for the maximum system operating pressure was based on the SRVs operating in the relief mode at the upper analytical setpoint limit. SLCS is not dependent upon any other SRV operating modes. This criterion has been generally replaced by the use of plant-specific ATWS transient pressure data occurring during the time the SLCS is analyzed to be in operation in consideration of NRC Information Notice 2001-13 (Reference 10). ATWS specific pressure data generally exceeds the original injection pressure. This injection pressure was calculated for both EPU and MELLLA+

and is within the capability of the SLCS. For the case of two SRVOOS, the ATWS results in Section 2.2.1 of this report indicate that the SRVs are cycling with a maximum relief flow of less than (( )) of rated steam flow. This relief flow is well within the capacity of the remaining nine in-service SRVs, and it can be concluded that the peak reactor vessel pressure calculated for SLCS injection for both TPO and MELLLA+ will not be exceeded with an additional SRVOOS.

Therefore, the PBAPS SLCS is adequate to support two SRVOOS.

3.3 CRD Evaluation The CRD system is designed to shutdown the reactor by inserting the control rods (scram). The Control Rod Drive Mechanism (CRDM) is also part of the reactor coolant pressure boundary.

Therefore, the potential effect on scram performance and CRD pressure boundary integrity are evaluated.

As discussed in Section 2.1.2, ((

)), the additional SRVOOS results in an (( )) in the peak reactor vessel pressure (bottom head and steam dome). The net (( )) in peak reactor pressure is (( ))

with the additional SRVOOS. For PBAPS, which is a BWR/4 plant, scram times decrease with an increase in transient reactor pressure (Reference 11 ); therefore, the effect of the additional SRVOOS is bounded by current analysis. For CRD integrity, the CRDM has been analyzed ((

)) bounds the peak bottom head pressures shown in Tables 1 and 2. Therefore, the PBAPS CRD system is adequate to support two SRVOOS.

4.0 ECCS LOCA ANALYSIS This section provides the qualitative analysis which addresses the effect of the additional SRVOOS on the ECCS-LOCA analysis.

4.1 Analysis Overview and Results For large break LOCA cases, depressurization of the vessel occurs due to mass and energy release out the break. Therefore, further relief through the SRVs is not required to arrive at post-accident conditions whereby ECCS assets can be delivered to recover the core and acceptably arrest the temperature excursion on the cladding.

6

004N6240-NP Revision 1 Non-Proprietary Information - Class I (Public)

For small break LOCA cases, the SRVs are relied upon for depressurization through the Automatic Depressurization System (ADS). The SRYs can be actuated by alternate means, mechanical or pneumatic, so a valve declared out-of-service for one actuation may still be available for the alternate actuation. ADS availability is stipulated by TS 3.5.1 with assurance that the five SRVs supporting ADS would remain available despite changes to allow an additional SRVOOS (mechanically, for overpressure protection) under the provisions of TS 3.4.3. With this confirmation, and noting that the five SRVs available for ADS function conform to the basis of the most recent ECCS-LOCA analysis (Reference 12, as cited as the basis for Reference 8) for limiting small break cases, it is concluded that there would be no effect on the ECCS-LOCA analysis results, and continued compliance to the acceptance criteria of 10 Code of Federal Regulations (CFR) 50.46 would be ensured.

4.2 Conclusion The SRVs are modelled in the ECCS-LOCA analysis, but a change to allow two SRVOOS will have no effect.

7

004N6240-NP Revision 1 Non-Proprietary Information - Class I (Public) 5.0 ACRONYMS Acronym Explanation ADS Automatic Depressurization System AOO Abnormal Operational Occurrence ASME American Society of Mechanical Engineers ATWS Anticipated Transient Without Scram BOC Beginning of Cycle CFR Code of Federal Regulations CRD Control Rod Drive CRDM Control Rod Drive Mechanism ECCS Emergency Core Cooling System EPU Extended Power Uprate GEH GE Hitachi Nuclear Energy HBB Hard Bottom Burn HPCI High Pressure Coolant Injection ICF Increased Core Flow LOCA Loss-of-Coolant Accident MELLLA+ Maximum Extended Load Line Limit Analysis MSIV Main Steam Isolation Valve MSIVC Main Steam Isolation Valve Closure MS IVF Main Steam Isolation Valve Closure with Flux Scram NRC Nuclear Regulatory Commission PBAPS Peach Bottom Atomic Power Station PCT Peak Cladding Temperature PRFO Pressure Regulator Failure Open RCIC Reactor Core Isolation Cooling RPT Recirculation Pump Trip SLC Standby Liquid Control SLCS Standby Liquid Control System SRV Safety Relief Valve SRVOOS Safety Relief Valve Out-of-Service 8

004N6240-NP Revision I Non-Proprietary Information - Class I (Public)

Acronym Explanation sv Safety Valve TAF Top of Active Fuel TCV Turbine Control Valve TPO Thermal Power Optimization TS Technical Specification 9

004N6240-NP Revision 1 Non-Proprietary Information - Class I (Public)

6.0 REFERENCES

1. GE Hitachi Nuclear Energy, "Migration to TRACG04 I PANACl 1 from TRACG02 I PANAClO for TRACG AOO and ATWS Overpressure Transients," NEDE-32906P Supplement 3-A, Revision 1, April 2010.

2. GE Nuclear Energy, "TRACG Application for Anticipated Operational Occurrences (AOO) Transient Analyses," NEDE-32906P-A, Revision 3, September 2006.

3. Global Nuclear Fuel, "General Electric Standard Application for Reactor Fuel (GEST AR II)," NEDE-24011-P-A-25, and Supplement for United States, NEDE-24011-P-A-25-US, August 2017.

4. Global Nuclear Fuel, "Supplemental Reload Licensing Report for Peach Bottom Unit 2 Reload 21 Cycle 22 Mid-Cycle Thermal Power Optimization (TPO)," 004N2488, Revision 0, October 2017.

5. Global Nuclear Fuel, "Supplemental Reload Licensing Report for Peach Bottom Unit 3 Reload 21Cycle22," 003N1452, Revision 0, September 2017.

6. Letter, J. F. Harrison (GEH) to NRC Document Control Desk, "Event-Specific 8CPR/ICPR Biases and Uncertainties and Peak Pressure Adders for AOO Licensing Applications," MFN 16-030, May 12, 2016.

7. Letter, J. F. Harrison (GEH) to NRC Document Control Desk, "Event-Specific 8CPR/ICPR Biases and Uncertainties and Peak Pressure Adders for AOO Licensing Applications," M 180061, March 23, 2018.

8. GE Hitachi Nuclear Energy, Safety Analysis Report for Peach Bottom Atomic Power Station Units 2 and 3 Thermal Power Optimization," NEDC-33873P, Revision 0, February 2017.

9. GE Nuclear Energy, "TRACG Application for Anticipated Transient Without Scram Overpressure Transient Analyses," NEDE-32906P Supplement 1-A, November 2003.

10. NRC Information Notice 2001-13, "Inadequate Standby Liquid Control System Relief Valve Margin," August 10, 2001.

11. GE Hitachi Nuclear Energy, "Safety Analysis Report for Exelon Peach Bottom Atomic Power Station Units 2 and 3 Constant Pressure Power Uprate," NEDC-33566P, Revision 0, September 2012.

12. GE Hitachi Nuclear Energy, "Safety Analysis Report for Peach Bottom Atomic Power Station Units 2 & 3 Maximum Extended Load Line Limit Analysis Plus," NEDC-33720P, Revision 0, September 2014.

10

004N6240-NP Revision 1 Non-Proprietary Information - Class I (Public)

Table3 Beginning of Cycle (BOC) MSIVC Event Sequence BOC Event Item Event Time (sec) 1 MSIV Isolation Initiated rr 2 High Pressure A TWS Setpoint 3 MSIVs Fully Closed 4 Peak Neutron Flux 5 Recirculation Pumps Trip 6 Ooening of the First Relief Valve 7 Peak Heat Flux 8 Peak Vessel Pressure 11 Table 4 BOC PRFO Event Sequence BOC Event Item Event Time (sec) 1 Turbine Control Valve (TCV) and Bypass Valves Start Open rr 2 MSIV Closure Initiated by Low Steam Line Pressure 3 MSIV s Fully Closed 4 Peak Neutron Flux 5 High Pressure A TWS Setpoint 6 Recirculation Pumps Trip 7 Opening of the First Relief Valve 8 Peak Heat Flux 9 Peak Vessel Pressure 11 11

004N6240-NP Revision 1 Non-Proprietary Information - Class I (Public)

((

))

Figure 1 BOC MSIVC (Short-term) 12

004N6240-NP Revision I Non-Proprietary Information - Class I (Public)

[(

))

Figure 2 BOC PRFO (Short-term) 13