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License Amendment Request to Reduce High Pressure Service Water System Design Pressure and Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, and 3.7.1 for Temporary Extension of ...
	ML18275A023
Person / Time
Site:	Peach Bottom
Issue date:	09/28/2018
From:	David Helker; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML18275A023 (75)
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Exelon Generation 200 Exelon Way Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.90 September 28, 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 NRG Docket Nos. 50-277 and 50-278

Subject:

License Amendment Request to Reduce High Pressure Service Water System Design Pressure and Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, and 3.7.1 for Temporary Extension of Completion Times In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit,u Exelon Generation Company, LLC (Exelon) requests an amendment to Renewed Facility Operating License Nos. DPR-44 and DPR-56 for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3.

The proposed changes revise the PBAPS, Units 2 and 3, design and licensing basis described in the Updated Final Safety Analysis Report (UFSAR) to reduce the design pressure rating of the High Pressure Service Water (HPSW) system. This change will provide additional corrosion margin in the HPSW system pipe wall thickness, increasing the margin of safety for the existing piping.

The proposed changes will also temporarily revise the PBAPS, Units 2 and 3, Technical Specifications (TS) Sections 3.6.2.3, "Residual Heat Removal (RHR) Suppression Pool Cooling," TS 3.6.2.4, "Residual Heat Removal (AHR) Suppression Pool Spray," TS 3.6.2.5, "Residual Heat Removal (AHR) Drywall Spray," and TS 3.7.1, "High Pressure Service Water (HPSW) System," as follows:

TS Section 3.6.2.3 will be revised to extend, on a one-time basis, four (4) allowable Completion Times (CTs} of Required Action A.1 for one AHR suppression pool cooling subsystem inoperable, from 7 days to 1O days, for Unit 2; and three (3)

CTs of Required Action A.1 for one AHR Suppression Pool Cooling subsystem inoperable, from 7 days to 1O days, for Unit 3, with an additional one (1) CT of up to 14 days for the A-C subsystem only, to allow for repairs to the Unit 3 AHR Heat Exchanger 3CE024.

U.S. Nuclear Regulatory Commission License Amendment Request Reduce HPSW System Design Pressure and Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, and 3.7.1 for Temporary Extension of Completion Times DPR-44 and DPR-56 September 28, 2018 Page2 TS Section 3.6.2.4 will be revised to extend, on a one-time basis, four (4) allowable CTs of Required Action A.1 for one RHR Suppression Pool Spray subsystem inoperable, from 7 days to 1O days, for Unit 2; and three (3) CTs of Required Action A.1 for one RHR suppression pool spray subsystem inoperable, from 7 days to 1O days, for Unit 3, with an additional one (1) CT of up to 14 days for the A-C subsystem only, to allow for repairs to the Unit 3 RHR Heat Exchanger 3CE024.

TS Section 3.6.2.5 will be revised to extend, on a one-time basis, four (4) allowable CTs of Required Action A.1 for one RHR drywall spray subsystem inoperable, from 7 days to 1O days, for Unit 2; and three (3) CTs of Required Action A.1 for one RHR drywell spray subsystem inoperable, from 7 days to 10 days, for Unit 3, with an additional one (1) CT of up to 14 days for the A-C subsystem only, to allow for repairs to the Unit 3 RHR Heat Exchanger 3CE024.

TS Section 3.7.1 will be revised to extend, on a one-time basis, four (4) allowable CTs of Required Action A.1 for one HPSW subsystem inoperable, from 7 days to 1O days, for Unit 2; and three (3) CTs of Required Action A.1 for one HPSW subsystem inoperable, from 7 days to 1O days, for Unit 3, with an additional one (1) CT of up to 14 days for the A-C subsystem only, to allow for repairs to the Unit 3 RHR Heat Exchanger 3CE024. Also, TS 3.7.1 will be revised to extend, on a one-time basis, two (2) allowable CTs of Required Action B.1 for one HPSW subsystem inoperable, from 7 days to 10 days, for Unit 2; and one (1) allowable CT of Required Action 8.1 for one HPSW subsystem inoperable, from 7 days to 1O days, for Unit 3, with an additional one (1) CT of up to 14 days for the A-C subsystem only, to allow for repairs to the Unit 3 RHR Heat Exchanger 3CE024.

These temporary TS changes are needed to allow sufficient time to perform physical modifications of the PBAPS, Units 2 and 3, HPSW systems to support the proposed reduction of the HPSW design pressure and to allow for timely repairs to Unit 3 RHR Heat Exchanger 3CE024.

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

The proposed changes have been reviewed by the PBAPS Plant Operations Review Committee in accordance with the requirements of the Exelon Quality Assurance Program.

Attachment 1 provides the evaluation of the proposed changes related to the HPSW design pressure. Attachment 2 provides a flow diagram of the HPSW system. Attachment 3 provides a diagram showing the portion of the HPSW system affected by the proposed change to the design pressure. Attachment 4 provides a copy of the marked-up UFSAR pages reflecting the proposed change to the design pressure (for information only).

Attachment 5 provides the evaluation of the proposed changes supporting the temporary extension of completion times for TS Sections 3.6.2.3, 3.6.2.4, 3.6.2.5, and 3.7.1.

U.S. Nuclear Regulatory Commission License Amendment Request Reduce HPSW System Design Pressure and Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, and 3.7.1 for Temporary Extension of Completion Times DPR-44 and DPR-56 September 28, 2018 Page 3 provides a copy of the marked-up TS pages that reflect the proposed changes. Attachment 7 provides the marked-up TS Bases pages (for information only).

This amendment request contains one regulatory commitment to implement the compensatory measures discussed in Section 3.3 of Attachment 5 during the extended TS Limiting Condition for Operation (LCO) CTs. This commitment is listed in Attachment 8.

Exelon requests approval of the proposed changes by September 28, 2019. Once approved, the temporary changes to TS Sections 3.6.2.3, 3.6.2.4, 3.6.2.5, and 3.7.1 will be implemented as required and will expire on December 31, 2020, for Unit 2 and on December 31, 2020, for Unit 3. Once approved, the reduction of HPSW design pressure will be implemented following completion of all physical modifications to the HPSW system, on a per unit basis. These modifications will be completed by December 31, 2020, for Unit 2 and by December 31, 2020, for Unit 3. The plant does not require this amendment to allow continued safe full power operation, although approval is required to support plant modifications required to reduce HPSW system design pressure.

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

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

Should you have any questions concerning this submittal, please contact Richard Gropp at 61 0-765-5557.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 28th day of September 2018.

Respectfully, David P. Helker Manager, Licensing and Regulatory Affairs Exelon Generation Company, LLC Attachments: 1. Evaluation of Proposed Changes - License Amendment Request to Reduce High Pressure Service Water System Design Pressure

2. Peach Bottom HPSW Flow Diagram

3. HPSW System Diagram Showing Current and Modified HPSW Configuration to Reduce Design Pressure

4. Markup of Proposed UFSAR Pages (For Information Only)

U.S. Nuclear Regulatory Commission License Amendment Request Reduce HPSW System Design Pressure and Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, and 3.7.1 for Temporary Extension of Completion Times DPR-44 and DPR-56 September 28, 2018 Page4

5. Evaluation of Proposed Changes - License Amendment Request to Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, and 3.7.1 for Temporary Extension of Completion Times

6. Markup of Technical Specifications Pages

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

8. Summary of Regulatory Commitments cc: w/ Attachments Regional Administrator- NRC Region I U.S. NRC Senior Resident Inspector- Peach Bottom Atomic Power Station U.S. NRC Project Manager, NRR- Peach Bottom Atomic Power Station R. R. Janati, Pennsylvania Bureau of Radiation Protection D. Tancabel, State of Maryland

ATTACHMENT 1 License Amendment Request Peach Bottom Atomic Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 EVALUATION OF PROPOSED CHANGES

Subject:

License Amendment Request to Reduce High Pressure Service Water System Design Pressure 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Background 2.2 Proposed Change

3.0 TECHNICAL EVALUATION

3.1 HPSW System Description 3.2 Design Analysis for Proposed Changes 3.3 Implementation

4.0 REGULATORY EVALUATION

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

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

License Amendment Request Attachment 1 Evaluation of Proposed Changes Page 1 of 16 Reduce HPSW System Design Pressure DPR-44 and DPR-56 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (Exelon) requests an amendment to Renewed Facility Operating License Nos. DPR-44 and DPR-56 for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3.

The proposed amendment revises the PBAPS Units 2 and 3 design and licensing basis described in the Updated Final Safety Analysis Report (UFSAR) to reduce the design pressure rating of the High Pressure Service Water (HPSW) system. This change will provide additional corrosion margin in the HPSW system pipe wall thickness, increasing the margin of safety for the existing piping.

2.0 DETAILED DESCRIPTION 2.1 Background The PBAPS Units 2 and 3 HPSW systems have exhibited a history of degradation similar to raw fresh water systems throughout the nuclear industry. In March 2014, Exelon submitted a request to the NRG, in accordance with 10 CFR 50.55a (a)(3)(i)

(References 1 and 2). This request proposed an alternative from the ASME Code,Section XI, IWD-3120(b) requirement to perform repair/replacement activities for degraded HPSW piping which has a maximum operating pressure in excess of 275 psig. The proposed alternative was to apply the evaluation methods of ASME Code Case N-513-3, "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 1," to the Class 3 HPSW system piping having a maximum operating pressure greater than 275 psig but less than or equal to 375 psig. The objective was to avoid accruing additional personnel radiation exposure and increased plant risk associated with a plant shutdown to comply with ASME Code,Section XI requirements. In addition, the use of an acceptable alternative analysis method in lieu of immediate action for a degraded condition allowed Exelon to perform additional extent of condition examinations on the affected systems while allowing time for safe and orderly long-term repair actions if necessary.

The NRG approved the relief request on March 19, 2015 (Reference 3). Exelon is currently implementing the monitoring requirements of Code Case N-513-3 at PBAPS Units 2 and 3, with an allowable leak rate of 5 gpm for the HPSW system piping. The NRG authorized use of the relief request (RR 14R-55) until the end of the fourth ISi interval, the acceptance criteria of Code Case N-513-3 are exceeded, or the leak rate exceeds the allowable, whichever occurs first. The fourth ISi interval ends on December 31, 2018.

Exelon submitted an additional relief request on March 26, 2018 (Reference 4) to address the fifth ISi interval, which ends on December 31, 2028.

This license amendment request proposes a permanent change to the HPSW system that will permit the plant to maintain compliance with the ASME Code,Section XI, ISi requirements and not require use of the relief request.

License Amendment Request Attachment 1 Evaluation of Proposed Changes Page 2 of 16 Reduce HPSW System Design Pressure DPR-44 and DPR-56 2.2 Proposed Changes The proposed amendment revises the PBAPS Units 2 and 3 design and licensing basis described in the UFSAR to reduce the design pressure rating of the HPSW system. The design pressure rating of the HPSW system supply piping from the HPSW pumps 2(3)AP042, BP042, CP042, and DP042 up to and including the Residual Heat Removal (AHR) heat exchanger HPSW throttle MOVs M0-2(3)-10-089A/B/C/D will be reduced from the current rating of 450 psig to a design pressure rating of approximately 200 psig. The final value is expected to fall between 190 psig and 230 psig in the detailed design with final vendor inputs.

The existing RHR heat exchangers 2(3)AE024, BE024, CE024, and DE024 located in the affected piping have a shell and tube design pressure rating of 450 psig. This pressure rating will remain unchanged. As a result of the proposed change, the operating pressure of the HPSW system will be reduced to a value below the maximum AHR system operating pressure.

The design pressure reduction will be achieved through physical plant modifications to reduce the operating pressure of the system. These include modifications to HPSW pumps to deliver less head, replacement of control valves M0-2(3)-10-089A/B/C/D (M0-089 valves) and removal of flow restricting orifices downstream of the M0-089 valves. The physical modifications that will implement the proposed change on PBAPS Units 2 and 3 will be addressed under the 10 CFR 50.59 process.

There are no permanent changes required to the PBAPS Units 2 and 3 Technical Specifications (TS) to implement the proposed change.

The proposed reduction in design pressure will provide for additional corrosion margin in pipe wall thickness, increasing the margin of safety for the existing piping.

3.0 TECHNICAL EVALUATION

3.1 HPSW System Description The HPSW system is described in UFSAR Section 10.7, and in TS Section 3.7.1, and TS Bases B.3. 7 .1. The safety objective of the HPSW system is to provide a reliable supply of cooling water for AHR under post-accident conditions. The major flow paths of the HPSW system consist of two independent parallel flow loops serving each unit, designated Division I and Division II. Each flow loop (i.e., subsystem) contains two HPSW pumps which discharge to a common header serving two AHR heat exchangers, connected in parallel. Division I contains HPSW pumps A and C, and Division II contains HPSW pumps Band D. In the normal open loop alignment, the HPSW pumps take suction from the Conowingo Pond through the Service Water Bay and the HPSW loops discharge through a common pipe for each unit to the discharge pond. The discharge pipe contains a normally open motor-operated isolation valve to the pond and a pipe connection to the emergency cooling water (ECW) system to provide an alternate discharge in the unlikely event that the Conowingo Dam fails or the pond floods. When the alternate discharge is

License Amendment Request Attachment 1 Evaluation of Proposed Changes Page 3 of 16 Reduce HPSW System Design Pressure DPR-44 and DPR-56 used, the emergency cooling tower (ECT) serves as a supply to the HPSW pumps through the pump bay in a closed loop alignment. See Attachment 2 for a flow diagram of the Peach Bottom HPSW system.

Each of the four (4) HPSW pumps on each unit are designed to deliver 4,500 gpm at a Total Developed Head (TOH) of nominally 700 feet (-303 psid). The maximum pump shutoff head is 1,025 feet (-444 psid) at zero flow. The HPSW pumps are driven by 1,000 hp electric motors. The pumps are 6-stage vertical pumps.

The HPSW system contains four (4) RHR heat exchangers per unit. They are vertically-oriented shell and tube heat exchangers, with HPSW flow through the tube side and RHR flow through the shell side. The heat exchanger design pressure is 450 psig on both the tube and shell sides.

The RHR heat exchanger HPSW outlet valves M0-089A/8/C/D are normally closed when the HPSW system is in standby with idle pumps. When establishing HPSW flow to a RHR heat exchanger, the M0-089 valve is opened, and then the HPSW pump is started. The M0-089 valves are throttled as necessary to achieve the required HPSW flow rate and to maintain HPSW pressure at least 20 psid higher than RHR pressure. At maximum expected RHR system pressures, a pressure of 233 psig at the HPSW pump discharge is required to ensure HPSW side pressure exceeds RHR system pressure at the RHR heat exchanger.

HPSW flow restricting orifices are installed downstream of the M0-089 valves in order to increase the backpressure at the M0-089 valves and at the RHR heat exchangers. This serves to suppress cavitation at the valves by reducing the required pressure drop across the throttled valves and increases margin above vapor pressure.

A divisional cross-tie line connecting the two HPSW loops on each unit is provided with a normally closed motor-operated isolation valve (M0-2-32-2344 and M0-3-32-3344). The divisional cross-tie is provided with redundant power supplies and is required to be operable during Modes 1, 2 and 3 to allow HPSW pumps from one loop to supply RHR heat exchangers in the opposite loop, if required. A unit cross-tie line with two normally closed manual isolation valves HV-2-32-516A and HV-3-32-5168 is also provided between one Unit 2 HPSW loop (8/D) and one Unit 3 HPSW loop (8/D). The cross-tie lines provide the flexibility to establish alternate flow alignments, if needed, under emergency conditions. A supply connection from the HPSW system to the RHR system, through two normally closed motor-operated valves, is provided from one HPSW loop per unit to permit the HPSW system furnishing a backup water supply to RHR for post-accident containment flooding.

The RHR and HPSW systems are designed such that HPSW operates at a higher pressure than RHR at the RHR heat exchanger interface; however, during standby conditions the RHR system pressure is maintained greater than HPSW. The RHR and HPSW systems are standby systems that typically operate during testing or plant shutdown. With this design, if there is an internal leak within a RHR heat exchanger during standby conditions, RHR water, which is normally torus water, leaks into the HPSW

License Amendment Request Attachment 1 Evaluation of Proposed Changes Page 4 of 16 Reduce HPSW System Design Pressure DPR-44 and DPR-56 system until the HPSW system pressure exceeds that of the AHR system. The HPSW system, when in operation, is maintained at a higher pressure than the AHR system at the AHR HX interface which prevents AHR water leakage into HPSW.

Each AHR heat exchanger contains a tube-to-shell differential pressure alarm, which alerts the operator if there is insufficient differential pressure between HPSW and AHR when the associated HPSW throttle valve M0-089 is not closed. Additionally, there are radiation monitors (2(3)AS331 and 2(3)BS331) that sample the HPSW system both upstream and downstream of the AHR heat exchangers, to indicate if there is cross-system leakage. These alarms and established Operations and Chemistry procedures are utilized to identify and address any RHR/HPSW system leakage issues.

The HPSW system is designed to seismic Class I criteria and is designed to be operable during flood conditions and the loss of pond event (Special Event), and during a loss of offsite power. The system is designed with capacity redundancy to supply cooling water under post-accident conditions with a limiting single active failure of an emergency diesel generator (EOG).

3.2 Design Analysis for Proposed Changes The objective of the proposed change is to reduce the design pressure rating of the HPSW system at PBAPS Units 2 and 3. The reduction in design pressure will provide for additional corrosion margin in pipe wall thickness. This change to the HPSW system will permit the plant to maintain compliance with the ASME Code,Section XI, ISi requirements and not require use of the Code Case N-513-3 relief request.

Current Design The current design pressure rating of the HPSW supply piping is 450 psig up to the first restricting orifice downstream of the M0-089 heat exchanger outlet valves. The design pressure bounds the maximum expected pump discharge pressure at pump shutoff head.

For HPSW subsystems with multiple restricting orifices (Units 2/3A and 2/30), the design pressure is reduced to 300 psig between orifices. The design pressure is reduced to 150 psig downstream of the last restricting orifice to the M0-2-32-2486 and M0-3-32-3486 valves and to 50 psig downstream of these valves to the discharge pond. The ECW system piping, which supplies HPSW flow to the ECT, is connected to the HPSW header between the last restricting orifice and the M0-2(3)-32-2(3)486 valves and is rated to 150 psig.

The HPSW piping system is designed to an original code of record (United States of America Standard [USAS] 831.1 - 1967 edition). The HPSW piping from the pumps to the M0-2(3)-32-2(3)486 valves is pipe class GB for carbon steel pipe A-106 Gr. 8, and carbon steel ANSI 300 lb. rated flanges and valves.

License Amendment Request Attachment 1 Evaluation of Proposed Changes Page 5 of 16 Reduce HPSW System Design Pressure DPR-44 and DPR-56 New Design The new design pressure rating of the HPSW supply piping will be reduced to a value of approximately 200 psig. The final value is expected to fall between 190 psig and 230 psig in the detailed design with final vendor inputs. Calculations will be revised to establish the new piping minimum wall thickness requirements at the reduced design pressure. The design pressure reduction will be achieved through physical modifications to reduce the operating pressure of the system, which include the following:

Replace or modify the HPSW pumps to deliver lower head (all four pumps on both units). Piping design pressures are reduced with the reduction in pump shutoff head and maximum sustained operating pressure.

Replace M0-089 throttle valves to achieve higher flow capacity (less pressure drop) with cavitation-resistant trim.

Remove flow restricting orifices in the HPSW piping downstream of the AHR heat exchanger M0-089 throttle valves. These orifices currently generate a significant portion of the system hydraulic resistance. For HPSW subsystems with multiple restricting orifices (Units 2/3A and 2/3D), the design pressure of the modified piping in this part of the system will also be reduced from 300 psig to 150 psig, consistent with the design pressure of the existing downstream piping.

Additional modifications to HPSW system components such as the overpressure protection relief valves, and pressure monitoring instrumentation will be made, as required, to accommodate the proposed change.

Attachment 3 shows the current HPSW system piping design pressures and the proposed change to reduce design pressure.

There will be no change to the design basis requirement for HPSW flow delivery to the AHR heat exchangers for post-accident containment cooling or other design basis functions. The HPSW system on each unit contains two independent and redundant loops (subsystems), where each loop contains two HPSW pumps. Either loop is capable of providing the required post-accident containment cooling capacity with two (2) HPSW pumps and two (2) RHR heat exchangers. TS 3.7.1 requires that two (both) HPSW loops and the HPSW divisional cross-tie be operable during Modes 1, 2 and 3. Per TS 3.7.1 Bases, manual initiation of the operable HPSW loop is assumed to occur ten (10) minutes after a design basis accident (OBA), with one (1) HPSW pump supplying cooling water to one (1) AHR heat exchanger. At one (1) hour after a OBA, a second HPSW pump is credited to supply cooling water to a second AHR heat exchanger for long term containment cooling. Each of two (2) credited HPSW pumps must deliver a minimum of 4,500 gpm to each of two (2) AHR HXs for post-LOCA containment cooling. At one (1) hour after a OBA, one (1) HPSW pump is also credited to deliver 4,500 gpm to one (1) RHR HX on the non-accident unit for safe shutdown.

The replacement HPSW pumps will be sized to deliver the required flow with margin, at a reduced discharge pressure. Since there is no change to the design basis HPSW flow delivery to the AHR heat exchangers and no change to HPSW supply temperature, the

License Amendment Request Attachment 1 Evaluation of Proposed Changes Page 6 of 16 Reduce HPSW System Design Pressure DPR-44 and DPR-56 proposed change will have no impact on AHR heat exchanger thermal performance or containment heat removal for normal operation and design basis events.

No changes are required to the UFSAR Chapter 14 accident analyses. The calculated radiological consequences of analyzed accidents are unchanged because the design basis radiological analysis, in conformance with Regulatory Guide 1.183, does not consider any liquid effluent release pathway through a heat exchanger interface. No new release pathway is introduced with the proposed change with respect to the design basis radiological analysis.

Though not required per the station design basis, a technical evaluation was performed to consider the radiological impact of a post-LOCA plant release into the environment via a liquid release pathway from a AHR heat exchanger leaking into HPSW. There is no regulatory guidance for a post-LOCA liquid release pathway and it is considered to be outside of the design bases. The most limiting dose consequence from AHR heat exchanger leakage into HPSW is that associated with post-LOCA control room dose due to airborne iodine release fraction. Based on a conservative analysis and maximum allowable AHR heat exchanger leak rate, the post-LOCA control room 30-day dose remains below the regulatory limit (5 rem TEDE). In addition, the evaluation considered the dose consequence from AHR heat exchanger leakage into HPSW at the nearest potable water sources and concluded that the 30-day dose to the public remains below the 25 rem TEDE limit at the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ).

An evaluation was also performed to consider the radiological impact due to a potential increase in AHR heat exchanger leakage during normal operations resulting from periods of operation in which HPSW pressure may be less than AHR pressure. A parametric study was conducted, using conservative AHR heat exchanger leakage rates, and concluded that the Offsite Dose Calculation Manual (ODCM) limits would not be exceeded.

There will be no new time critical operator actions for design basis events required as a result of the change. HPSW system surveillance procedures will be updated to reflect the new design pressure rating.

There are no changes anticipated to the HPSW system radiation monitors, associated Control Room annunciation, or established procedures for responding to a RHR heat exchanger tube leak. If AHR heat exchanger leakage is detected, the leakage is monitored and included in the Annual Radiological Effluent Release Report and the heat exchanger is repaired.

The PBAPS licensing basis for High Energy Line Breaks (HELB), as described in UFSAR Appendix A.10, defines high energy piping systems as those containing fluids with a service temperature > 200°F and design pressure > 275 psig. PBAPS piping systems are only considered high energy when both criteria (high pressure and high temperature) are satisfied. Thus, HPSW is not considered to be a high energy system in the PBAPS current licensing basis. Therefore, reducing the design pressure would not change the current characterization of HPSW with regard to pipe rupture hazards.

License Amendment Request Attachment 1 Evaluation of Proposed Changes Page 7 of 16 Reduce HPSW System Design Pressure DPR-44 and DPR-56 In the event of a loss of the normal heat sink (Conowingo Pond), the HPSW system is aligned to the Emergency Heat Sink (EHS) which utilizes the ECT in a closed loop mode for shutdown cooling of both units. The EHS is credited for safe shutdown of both units following loss of pond with dual unit LOOP, but is not credited for mitigating a DBA-LOCA or any other design basis accident. The loss of normal heat sink is classified as a Special Event in the PBAPS UFSAR (Table 1.4.2). The current analysis for this event credits HPSW flow from each unit to the ECT. The proposed modifications to the HPSW system will ensure that this design capability is maintained.

3.3 Implementation The physical modifications required to reduce HPSW system design pressure are expected to be installed with the plant on-line (Mode 1). PBAPS TS 3.7.1 requires that two HPSW subsystems shall be operable during Modes 1, 2 and 3. It has been determined that the current HPSW TS 3.7.1, Required Action A.1 Completion Time (CT) of 7 days is not sufficient time to replace both HPSW pumps in a loop, or to replace M0-089 valves and remove flow orifices from a single subsystem. Also, the TS 3.7.1 Required Action B.1 CT, for the HPSW cross-tie line inoperable, is not sufficient time to replace M0-089 valves and remove flow orifices from a single subsystem. An extension of the CTs will be required for TS 3.7.1, and for TS 3.6.2.3, 3.6.2.4, and 3.6.2.5, due to their dependency on the HPSW system for operability. Based on evaluation of the installation tasks, it has been determined that 1O days will provide sufficient time, with margin, to perform each of the installation tasks for a HPSW subsystem. Therefore, a temporary CT extension from 7 days to 1O days is requested for each of these TS, with the exception of one CT extension from 7 days to 14 days for the Unit 3 A-C subsystem only, to allow for repairs to the Unit 3 AHR Heat Exchanger 3CE024, as described in Attachment 5.

Once approved, the reduction of HPSW design pressure will be implemented following completion of all physical modifications to the HPSW system, on a per unit basis. These modifications will be completed by 12/31 /2020 for Unit 2 and by 12/31 /2020 for Unit 3.

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.

AEC Criterion 6 - Reactor Core Design (Category A)

AEC Criterion 6 specifies: "The reactor core shall be designed to function throughout its design lifetime, without exceeding acceptable fuel damage limits which have been stipulated and justified. The core design, together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can

License Amendment Request Attachment 1 Evaluation of Proposed Changes Page 8 of 16 Reduce HPSW System Design Pressure DPR-44 and DPR-56 be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of all offsite power.

The decay heat removal function of the HPSW in support of the RHR system is not adversely affected by the reduction in HPSW design and operating pressures. ECCS pump function is maintained and RHR decay heat removal is assured to meet its design and functional requirements. As a result, the core design is not adversely impacted.

Therefore, the proposed change will not impact the ability of PBAPS to comply with the requirements of AEC Criterion 6.

AEC Criterion 1O- Containment (Category A)

AEC Criterion 10 specifies: "Containment shall be provided. The containment structure shall be designed to sustain the initial effects of gross equipment failures, such as a large coolant boundary break, without Joss of required integrity and, together with other engineered safety features as may be necessary, to retain for as Jong as the situation requires the functional capability to protect the public."

The capability of the containment cooling function of the HPSW and RHR systems is not adversely affected by the reduction in HPSW design and operating pressures, and is assured to meet their design and functional requirements. As a result, the containment design is not impacted.

Therefore, the proposed change will not impact the ability of PBAPS to comply with the requirements of AEC Criterion 10.

AEC Criterion 16 - Monitoring Reactor Coolant Pressure Boundary (Category 8)

AEC Criterion 16 specifies: 11 Means shall be provided for monitoring the reactor coolant pressure boundary to detect leakage. 11 The HPSW system shall be designed to monitor radioactive releases from a RHR heat exchanger leak. Radiation monitors are provided on the HPSW system. The proposed change will not affect this monitoring capability.

Therefore, the proposed change will not impact the ability of PBAPS to comply with the requirements of AEC Criterion 16.

AEC Criterion 17 - Monitoring Radioactivity Releases (Category 8)

AEC Criterion 17 specifies, "Means shall be provided for monitoring the containment atmosphere, the facility effluent discharge paths, and the facility environs for radioactivity that could be released from normal operations, from anticipated transients, and from accident conditions. "

License Amendment Request Attachment 1 Evaluation of Proposed Changes Page 9 of 16 Reduce HPSW System Design Pressure DPR-44 and DPR-56 The reduction in HPSW design and operating pressure does not affect the capability to monitor reactor coolant pressure boundary (RCPB) leakage. In addition, the existing HPSW system radiation monitoring capability will not be affected by this change.

Therefore, the proposed change will not impact the ability of PBAPS to comply with the requirements AEC Criterion 17.

AEC Criterion 37 - Engineered Safety Features Basis for Design (Category A)

AEC Criterion 37 specifies: "Engineered safety features shall be provided in the facility to back up the safety provided by the core design, the reactor coolant pressure boundary, and their protection systems. As a minimum, such engineered safety features shall be designed to cope with any size reactor coolant pressure boundary break up to and including the circumferential rupture of any pipe in that boundary assuming unobstructed discharge from both ends. 11 The reduction in HPSW design and operating pressures does not affect the capability of the HPSW system to support containment heat removal and core cooling.

Therefore, the proposed change will not impact the ability of PBAPS to comply with the requirements of AEC Criterion 37.

AEC Criterion 38- Reliability and Testability of Engineered Safety Features (Category A)

AEC Criterion 38 specifies: "All engineered safety features shall be designed to provide high functional reliability and ready testability. In determining the suitability of a facility for a proposed site, the degree of reliance upon and acceptance of the inherent and engineered safety afforded by the systems, including engineered safety features, will be influenced by the known and the demonstrated performance capability and reliability of the systems, and by the extent to which the operability of such systems can be tested and inspected where appropriate during the life of the plant."

The HPSW system shall be designed to provide high reliability and testability to conform to this criterion. Reliability and testability of the HPSW system are not affected by a reduction in the design and operating pressures.

Therefore, the proposed change will not impact the ability of PBAPS to comply with the requirements of AEC Criterion 38.

AEC Criterion 41 - Engineered Safety Features Performance Capability (Category A)

AEC Criterion 41 specifies: "Engineered safety features such as emergency core cooling and containment heat removal systems shall provide sufficient performance capability to accommodate partial loss of installed capacity and still fulfill the required safety function.

As a minimum, each engineered safety feature shall provide this required safety function assuming a failure of a single active component.

11

License Amendment Request Attachment 1 Evaluation of Proposed Changes Page 10 of 16 Reduce HPSW System Design Pressure DPR-44 and DPR-56 The HPSW system shall be designed so that it supports the Engineered Safety Features with a single active component failure to conform to this criterion. The proposed change does not affect the existing performance capability of the HPSW system, assuming a single active failure.

Therefore, the proposed change will not impact the ability of PBAPS to comply with the requirements of AEC Criterion 41.

AEC Criterion 42 - Engineered Safety Features Components Capability (Category A)

AEC Criterion 42 specifies: "Engineered safety features shall be designed so that the capability of each component and system to perform its required function is not impaired by the effects of a Joss-of-coolant accident."

The HPSW system shall be protected from the effects of a LOCA to conform to this criterion. The proposed change does not alter the environmental effects of a LOCA or affect the capability of the HPSW system to withstand the effects of a LOCA.

Therefore, the proposed change will not impact the ability of PBAPS to comply with the requirements of AEC Criterion 42.

AEC Criterion 52 - Containment Heat Removal Systems (Category A)

AEC Criterion 52 specifies: "Where active heat removal systems are needed under accident conditions to prevent exceeding containment design pressure, at least two systems, preferably of different principles, each with full capacity, shall be provided."

The HPSW system shall be designed so that no single active component failure can prevent the system from supplying cooling water to support the containment post-accident cooling mode to conform to this criterion. The containment cooling function of the HPSW system is not affected by the reduction in HPSW design and operating pressures. The system flowrate and heat removal rate for design basis events are not changed.

Therefore, the proposed change will not impact the ability of PBAPS to comply with the requirements of AEC Criterion 52.

AEC Criterion 58 - Inspection of Containment Pressure Reducing Systems (Category A)

AEC Criterion 58 specifies: "Design provisions shall be made to facilitate the periodic physical inspection of all important components of the containment pressure-reducing systems, such as, pumps, valves, spray nozzles, torus, and sumps."

The HPSW system shall be designed to permit periodic inspections of important components to conform to this criterion. The reduction in HPSW design and operating pressures does not affect this capability.

License Amendment Request Attachment 1 Evaluation of Proposed Changes Page 11 of 16 Reduce HPSW System Design Pressure DPR-44 and DPR-56 Therefore, the proposed change will not impact the ability of PBAPS to comply with the requirements of AEC Criterion 58.

AEC Criterion 59 - Testing Of Containment Pressure-Reducing Systems Components (Category A)

AEC Criterion 59 specifies: "The containment pressure-reducing systems shall be designed so that active components, such as pumps and valves, can be tested 11 periodically for operability and required functional performance.

The HPSW system shall be designed so that pumps and valves can be tested for operability to conform to this criterion. The reduction in HPSW design and operating pressures does not affect this capability.

Therefore, the proposed change will not impact the ability of PBAPS to comply with the requirements of AEC Criterion 59.

AEC Criterion 61 -Testing Of Operational Sequence of Containment Pressure-Reducing Systems (Category A)

AEC Criterion 61 specifies: "A capability shall be provided to test under conditions as close to the design as practical the full operational sequence that would bring the containment pressure-reducing systems into action, including the transfer to alternate 11 power sources.

The ability to demonstrate operability, test the functional performance, and inspect the active components of containment pressure reducing systems and the containment cooling system is provided. This capability is not affected by the reduction in HPSW design and operating pressures.

Therefore, the proposed change will not impact the ability of PBAPS to comply with the requirements of AEC Criterion 61 .

AEC Criterion 67 - Fuel and Waste Storage Decay Heat (Category 8)

AEC Criterion 67 specifies: "Reliable decay heat removal systems shall be designed to prevent damage to the fuel in storage facilities that could result in radioactivity release to 11 plant operating areas or the public environs.

To conform to this criterion, plant fuel handling and storage facilities preclude accidental criticality and provide sufficient cooling for spent fuel. The reduction in HPSW design and operating pressures does not affect the HPSW system in the performance of its support function to the RHR system to supplement fuel pool cooling. The system flowrate and heat removal rate for design basis events are not changed.

Therefore, the proposed change will not impact the ability of PBAPS to comply with the requirements of AEC Criterion 67.

License Amendment Request Attachment 1 Evaluation of Proposed Changes Page 12 of 16 Reduce HPSW System Design Pressure DPR-44 and DPR-56 AEC Criterion 70- Control of Releases of Radioactivity to the Environment (Category B)

AEC Criterion 70 specifies: "The facility design shall include those means necessary to maintain control over the plant radioactive effluents, whether gaseous, liquid, or solid.

Appropriate holdup capacity shall be provided for retention of gaseous, liquid, or solid effluents, particularly where unfavorable environmental conditions can be expected to require operational limitations upon the release of radioactive effluents to the environment.

In all cases, the design for radioactivity control shall be justified (a) on the basis of 10CFR20 requirements for normal operations and for any transient situation that might reasonably be anticipated to occur and (b) on the basis of 10CFR100 dosage level guidelines for potential reactor accidents of exceedingly low probability of occurrence except that reduction of the recommended dosage levels may be required where high population densities or very large cities can be affected by the radioactive effluents."

The plant radioactive waste control systems (which include the liquid, gaseous, and solid radwaste subsystems) are designed to limit the potential offsite radiation exposure to levels below the limits of 10 CFR 20. The plant engineered safeguards are designed to limit the offsite exposure under the postulated design basis accidents to levels below 10 CFR 50.67 (Accident Source Term).

The existing HPSW system design includes radiation monitors that are provided on the HPSW system inlet and outlet piping of the AHR heat exchangers. With the proposed change, the HPSW design and operating pressures will be below that of the AHR system.

A AHR heat exchanger tube leak would release radioactivity into the HPSW system water, which is discharged to the environment. However, the existing HPSW system radiation monitoring and Control Room annunciation capability will not be affected by the change.

Also, existing administrative controls for response to detection of abnormal radiation conditions will be maintained.

Therefore, the proposed change will not impact the ability of PBAPS to comply with the requirements of AEC Criterion 70.

4.2 Precedent Many operating U.S. BWR power plants are designed such that the service water system operates at a pressure below that of the AHR system. Some of the plants are listed below.

Similar to PBAPS Units 2 and 3, these plants have radiation monitors in the service water outlet from the AHR heat exchangers to alert operators upon detection of radioactivity.

Browns Ferry Units 1, 2, and 3 Clinton Power Station Grand Gulf Nuclear Station La Salle Units 1 and 2 Limerick Units 1 and 2 Nine Mile Point Unit 2 River Bend Station

License Amendment Request Attachment 1 Evaluation of Proposed Changes Page 13 of 16 Reduce HPSW System Design Pressure DPR-44 and DPR-56 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 {Exelon), proposes a change to the Peach Bottom Atomic Power Station {PBAPS) Units 2 and 3 Updated Final Safety Analysis Report {UFSAR) to reduce the design pressure rating of the High Pressure Service Water {HPSW) system.

Exelon has evaluated whether or not a significant hazards consideration is involved with the proposed changes in accordance with the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below.

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

No. The HPSW system does not initiate any accidents discussed in Chapter 14 of the PBAPS, Units 2 and 3 UFSAR. A shutdown cooling

{AHR system) malfunction leading to a moderator temperature decrease could result from mis-operation of the cooling water controls for the AHR heat exchangers, as described in UFSAR Section 14.5.2.4. The resulting temperature decrease causes a slow insertion of positive reactivity into the core. However, the proposed change to the HPSW system design pressure will not affect the initiator for this accident. The proposed reduction of the HPSW system design pressure has been evaluated for effects on system piping and components using appropriate codes and standards. The proposed changes do not introduce any failure mechanisms that would initiate a previously analyzed accident. The HPSW and AHR systems remain capable of performing their UFSAR-described design functions for accident mitigation. Moreover, the design and operability requirements currently addressed by the PBAPS Technical Specifications {TS) are unaffected and the design basis radiological analysis of analyzed accidents is unchanged. Thus, the consequences of analyzed accidents are not increased.

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

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

No. The proposed changes will reduce the design and operating pressure in a portion of the HPSW system. This change will not introduce a new mode of plant operation. The system flowrate and heat removal rate for design basis events are not changed. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the

License Amendment Request Attachment 1 Evaluation of Proposed Changes Page 14 of 16 Reduce HPSW System Design Pressure DPR-44 and DPR-56 proposed changes. All accident analysis criteria continue to be met and there are no adverse effects on any safety-related system.

Therefore, the proposed changes do 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?

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 setpoints for the actuation of equipment relied upon to respond to an event. The reduction in HPSW system design pressure permits continued operation of the HPSW and RHR systems in accordance with the plant safety analysis. The core and containment heat removal functions of the HPSW and RHR systems are not affected. The proposed change does not alter the safety limits or safety analysis assumptions associated with the operation of the plant.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

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.

Based on the above evaluation, Exelon 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.

5.0 ENVIRONMENTAL CONSIDERATION

10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusions or otherwise not requiring environmental review," addresses requirements for submitting environmental assessments as part of licensing actions. 10 CFR 51.22, paragraph (c)(9) states that a categorical exclusion applies for Part 50 license amendments that meet the following criteria:

i. No significant hazards consideration (as defined in 10 CFR 50.92(c));

License Amendment Request Attachment 1 Evaluation of Proposed Changes Page 15 of 16 Reduce HPSW System Design Pressure DPR-44 and DPR-56 ii. No significant change in the types or significant increase in the amounts of any effluents that may be released offsite; and iii. No significant increase in individual or cumulative occupational radiation exposure.

The proposed changes do not involve a significant hazards consideration. The reviews and evaluations performed to support the proposed change concluded that all systems will function as designed, and all performance requirements for these systems have been evaluated and found acceptable. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. Operation of the plant with the proposed change does not involve a significant reduction in a margin of safety.

No significant changes in types or amounts of effluents released into the environment will occur as a result of the HPSW system design pressure reduction. The Pennsylvania Department of Environmental Protection (PDEP) National Pollutant Discharge Elimination System (NPDES) permit provides the effluent limitations and monitoring requirements for wastewater at the site.

There is no significant increase in individual or cumulative occupational radiation exposure with the proposed changes.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22, 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.

6.0 REFERENCES

1. Letter from James Barstow (Exelon Generation Company, LLC) to USNRC, "Proposed Alternative to Utilize Code Case N-513-3, 'Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 1,' at a Higher System Operating Pressure," dated March 28, 2014 (ADAMS Accession No. ML14090A140)

2. Letter from James Barstow (Exelon Generation Company, LLC) to USNRC, "Response to Request for Additional Information - Proposed Alternative to Utilize Code Case N-513-3, 'Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 1,' at a Higher System Operating Pressure," dated February 5, 2015 (ADAMS Accession No. ML15036A487)

3. Letter from Travis L. Tate (U.S. NRC) to Bryan C. Hanson (Exelon Generation Company, LLC), "Peach Bottom Atomic Power Station, Units 2 and 3, and Quad Cities Nuclear Power Station, Units 1 and 2 - Relief from the Requirements of the ASME Code (TAC NOS. MF3799, MF3800, MF3801, AND MF3802)," dated March 19, 2015 (ADAMS Accession No. ML15043A496)

License Amendment Request Attachment 1 Evaluation of Proposed Changes Page 16 of 16 Reduce HPSW System Design Pressure DPR-44 and DPR-56

4. Letter from James Barstow (Exelon Generation Company, LLC) to USNRC, "Relief Requests Associated with the Use of Code Cases N-513-4 and N-513-3 for the Fifth lnservice Inspection Interval," dated March 26, 2018 (ADAMS Accession No. ML180868110)

Attachment 2 Peach Bottom Atomic Power Station Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 Peach Bottom HPSW Flow Diagram

LAA - Reduce HPSW System Design Pressure Attachment 2 Peach Bottom HPSW Flow Diagram Page 1of1 DPR-44 and DPR-56 ILAT SIP RESERVOIR 2972 3972 EMERG COOLING ECW TOWER I I ECCS HX'S ECCS HX'S I I PUMP I UN IT 3

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¢ 0498 DISCHARGE CANAL 2233A 22339 2213 2209 3213 32338 3233A COMBINED SYSTEM FLOWPATH

Attachment 3 Peach Bottom Atomic Power Station Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 HPSW System Diagram Showing Current and Modified HPSW Configuration to Reduce Design Pressure

LAR - Reduce HPSW System Design Pressure Attachment 3 HPSW System Diagram Showing Current and Modified HPSW Configuration to Reduce Design Pressure Page 1 of 2 DPR-44 and DPR-56 Alternate Discharse -

Emergency Cooling Water System Design Pressure 150 psig M02(3)803 M00502A M00502B RHRHX Restricting Orifice(s) - MOD502C HPSWPump 3 ROs per A & D loops lROperB&Cloop RO 2801 Motor MOlG-089 RO 2800 Gate Gate Gate Emergency Cooling Normal Discharge - Conowingo Pond Tower Cells A/B/C Water Level 98.5' to 113' per TS3.7.2 Design Pressure 300pslg Design Pressure 150 psig Conowingo Pond Water Level 98.5' to 113' per Design Pressure 450 pslg TS3.7.2 HPSW Current Configuration - Single Loop

LAA - Reduce HPSW System Design Pressure Attachment 3 HPSW System Diagram Showing Current and Modified HPSW Configuration to Reduce Design Pressure Page 2 of 2 DPA-44 and DPA-56 Alternate Discharge -

Emergency Coolins Water System Design Pressure 150 pslg P.epl:w:c MO 10--089 lf.:lv~ *Nith hi~her M02(3)803 MOOS02A Rcmov(: di! R~strkling 01/flccs, (J RU\.

1n A&D ICNp'.'i, .1 RO iri S&C loops) to MOOS028 Ro.!pl<k'e HPSW Pumµ tv n.."\Jut:e Tt1H RHRHX MOOS02C Manual Motor Check Manual Manual Valve Gate Gate Emergency Coolin&

Normal Discharge

Conowlngo Pond Tower C2ils A/B/C Water l.b.i~ 985' to 113' perTS3.7.2 Design Pressure SO pslg Conowingo Pond Water Design Pressure -zoo ps1g Design Pressure 150 psig level 98.5' to 113' per TS3.7.2 HPSW Modified Configuration to Reduce Design Pressure

Attachment 4 Peach Bottom Atomic Power Station Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 Markup of Proposed UFSAR Pages (For Information Only)

PBAPS UFSAR lpage from 7.12.4 Liquid Process Radiation Monitoring System activity due to activation of added corrosion inhibitors. Changes in the normal radiation level could indicate leaks of radioactive water into the system.

The liquid radwaste system provides for collection of waste liquids through various drainage systems. Because of conductivity not all of the waste liquids can be economically purified by demineralization. Consequently, some liquid containing radioactivity is eventually discharged from the system. The liquid radwaste monitor indicates and records the radiation levels in this discharge.

The emergency service water system provides cooling water to the core standby cooling equipment in case of a loss of off-site power. Changes in the normal radiation level of the emergency service water discharge could indicate leakage in the core standby cooling equipment.

Radiation monitors have been provided on each high-pressure service water intake and discharge of the RHR heat exchangers for Unit 2 and 3. In the event that a heat exchanger leak occurs__in canj1mct j on wj th a revers a 1 of narma 1 heat exchanger dj fferenti a 1 pressnre, these monitors will annunciate, in the control room, the presence of radioactivity in the high-pressure service water system. Two monitors have been provided for each unit. Samples are drawn from either the upstream or downstream piping of the RHR heat exchanger depending on HPSW system function (active or idle) .

This will assure RHR leakage through the heat exchanger is monitored periodically. After HPSW operation, water in the system will drain through the HPSW pump discharge check valve, lowering the water level in the inlet and discharge piping to the RHR heat exchangers. Radiation monitoring will be temporarily suspended during periods when a solid water column in the header does not exist. Without the water column present, the transmission path for contamination does not exist, and continued sampling is not required. This will also protect the sample pumps from running dry.

Leakage from fuel elements stored in the fuel storage pool would be detected by radiation monitors suitably located in each spent fuel pool.

The environmental and power supply design conditions are given in Table 7.12.2.

CHAPTER 07 7.12-10 REV. 26, APRIL 2017

UNIT ? UNIT RUtTOR BL&. RE,1.CTOR. SL6 COOLl!IG W.1.HR COOLING W.i.TtR HHT E"i:tHA.HGERS(?)* HEAT HCH,1.ttGERSl2)*

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HIGH PRESS URE. SER.VICE EMERGENCY SERVICE HIGH PRESSURE SERVICE WAiEi! PUMPS WATER ?UMPS (Z) WATER PUMPS EMERGENCY SERVICE WHER (TOT.o.L 4) (TOo~L 4)

ANO HIGH PRESSURE UNIT 2 COMMON UNIT 3 SERVICE WATER SYSTEMS ftGUll 928 Ro*.IS 04/98

PBAPS UFSAR 10.7 HIGH PRESSURE SERVICE WATER SYSTEM 10.7.1 Safety Objective The safety objective of the high pressure service water system is to provide a reliable supply of cooling water for RHR under post-accident conditions.

10.7.2 Safety Design Basis

1. The high pressure service water system is designed to seismic Class I criteria to withstand the maximum credible earthquake without impairing system function.

2. The high pressure service water system is operable during flood conditions.

3. The high pressure service water system is designed with capacity and redundancy to supply cooling water to the RHRS under post-accident conditions.

4. The high pressure service water system is operable during the loss of offsite power.

10.7.3 Power Generation Objective The power generation objective of the high pressure service water system is to supply cooling water to the RHRS for shutdown cooling and for torus cooling.

10.7.4 Power Generation Design Basis

1. The high pressure service water system supplies a reliable source of cooling water to the RHRS.

2. The high pressure service water system is designed for remote-manual initiation.

~ The high pressure service Hater system inhibits leakage of radioactive material from the RHRS to the environment.

~4. The high pressure service water system provides an additional source of water for post-accident containment flooding by a cross tie between the high pressure service water system and the RHRS.

10.7.5 Description r

Each high pressure service water system consists of four 4,500 gpm pumps installed in parallel in the pump structure (Drawing CHAPTER 10 10.7-1 REV. 26, APRIL 2017

PBAPS UFSAR M-315, Sheets 1 through 4) . Normal water supply to the suction of the pumps is from Conowingo Pond. When the high pressure service water system is operated in conjunction with the emergency heat sink (subsection 10.24, "Emergency Heat Sink"), the suction is from the HPSW pump bay which is fed from emergency cooling tower basin. The pump discharge is manifolded and provided with a normally closed, motor-operated valve separating the four pumps into groups of two. Two parallel headers run from the pump structure to the reactor building. Each header delivers the discharge from two pumps to two RHR heat exchangers also in parallel. Under normal conditions, iJhen the respective loop of HPSW is in operation, the service i1ater pressure on the discharge side of the RHR heat exchanger is maintained positive Hith respect to the RHRS side to inhibit leakage of radioactive material into the environment, In the event of a design basis accident or transient in which additional containment cooling capacity is required, a second HPSW pump can be aligned to a second RHR heat exchanger by opening the cross-tie valve.

Under abnormal operating conditions RHRS pressure could exceed high pressure service water system pressure. An RHR heat exchanger leak under these abnormal conditions would result in radioactive RHR water migrating into the high pressure service water system and into the river. To limit the release of radioactive water to the river from this potential release path, signals from the radiation monitors in the sample system which samples the high pressure service water system upstream and downstream of the RHR heat exchangers initiate an alarm in the control room at a predetermined radiation level.

Flanged connection points are available on the high pressure service water system, downstream of the RHR heat exchangers, to allow for a temporary flow path of the RHR heat exchanger cooling water in the event that the normal flow path becomes unavailable.

This alternative flow path is intended to be routed through secondary containment. Therefore, this flow path may only be used when secondary containment is not required.

An intertie is provided between units 2 and 3 high pressure service water system to provide flexibility. A cross tie to the RHRS provides the capability for primary containment flooding.

The high pressure service water system pumps are vertical multistage turbine type. The pump mounting base is of watertight construction to withstand the hydrostatic pressure at the design flood condition. The pump design data is given in Table 10.7.1.

The high pressure service water system piping and valves are designed as described in Appendix A.

CHAPTER 10 10.7-2 REV. 26, APRIL 2017

PBAPS UFSAR 10.7.6 Safety Evaluation The high pressure service water system pumps are installed in a seismic Class I structure. The system meets seismic Class I criteria and is protected against the design flood level.

Each pump is sized to accommodate the design heat removal capacity of one RHRS heat exchanger. They have adequate head (1) to FRaintain the high pressure service Hater systeFR cooling 1dater at a higher pressure than the RHRS, thus inhibiting the release of radioactive FRaterial to the envirorunent, and (2) to permit operation in conjunction with the emergency heat sink. Further, the pumps have both a normal and a standby power supply. In the event of the loss of offsite power, the pumps are supplied from the diesel generators and manually started as required.

Sufficient redundancy is provided in the number of pumps and power supplies, and in the piping arrangement, so that no single system component failure can prevent the system from supplying cooling water to accommodate the normal shutdown mode and the containment cooling mode. Therefore, core decay heat removal during the shutdown periods, or containment cooling during the post-accident condition, can be maintained.

10.7.7 Inspection and Testing Pumps in the high pressure service water system are proven operable by their use or testing during normal station operations.

Motor operated isolation valves can be tested to assure they are capable of opening and closing by operating manual switches in the control room and observing the position lights. Portions of the high pressure service water system normally closed to flow can be tested to ensure their operability and the integrity of the system.

CHAPTER 10 10.7-3 REV. 26, APRIL 2017

I>BAPS UFSAR TABLE 10.7.1 HIGH PRESSURE SERVICE WATER SYSTEM EQUIPMENT DATA High Pressure Service Water Pumps Quantity 4 Per Unit Type Vertical, Turbine~T~yP..._e~~~~~~

~5,000 gpm at 396 ttl Flow/Head Design Point 4,500 gp~ft I Bhp at Rating < 975 hp~

Speed 1,770 rpm Number of Stages ~~

Pump Design: < 240 psig Shut-Off Head > 30g and < 445 Material:

Bowl/Impeller Cast Carbon Steel or Moly Iron/Bronze or Cast Stainless Steel Discharged Head/Column Carbon Steel/Carbon Steel Line Shaft Stainless Steel Bearings Brass/Bronze/Rubber Motor:

Type Vertical, Induction Horsepower 1,000 hp Voltage/Phase/Frequency 4,160 V/3 Phase/60 Hz CHAPTER 10 10.7-4 REV. 23, APRIL 2011

ATIACHMENTS Peach Bottom Atomic Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 EVALUATION OF PROPOSED CHANGES

Subject:

License Amendment Request to Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, and 3.7.1 for Temporary Extension of Completion Times 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

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

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

License Amendment Request Attachment 5 Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, Page 1 of 17 and 3.7.1 for Temporary Extension of Completion Times DPR-44 and DPR-56 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (Exelon) requests an amendment to Renewed Facility Operating License Nos. DPR-44 and DPR-56 for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3.

The proposed changes will temporarily revise the PBAPS, Units 2 and 3, Technical Specifications {TS) Sections 3.6.2.3 "Residual Heat Removal (AHR) Suppression Pool Cooling," TS 3.6.2.4, "Residual Heat Removal (AHR) Suppression Pool Spray," TS 3.6.2.5, "Residual Heat Removal (AHR) Drywell Spray," and TS 3.7.1, "High Pressure Service Water (HPSW) system," to extend, on a one-time basis, four allowable Completion Times

{CTs) of Required Action A.1from7 days to 10 days, for Unit 2. Also, TS 3.7.1 will be revised to extend, on a one-time basis, two allowable CTs of Required Action B.1 for one HPSW subsystem inoperable, from 7 days to 1O days, for Unit 2. For Unit 3, the proposed change will temporarily revise TS 3.6.2.3, 3.6.2.4, 3.6.2.5, and 3.7.1 to extend, on a one-time basis, three allowable CTs of Required Action A.1 from 7 days to 1O days, and one allowable CT from 7 days to 14 days for the A-C subsystem only. Also, for Unit 3, TS 3.7.1 will be revised to extend, on a one-time basis, one allowable CT of Required Action B.1 from 7 days to 1O days and one allowable CT from 7 days to 14 days for the A-C subsystem only.

A footnote will be added to the affected TS LCOs to indicate that the 7-day CTs for the affected systems may be extended up to four times for Required Action A.1, and up to two times for TS 3.7.1 Required Action B.1, for a period of up to 10 days each, with the exception of an extension up to 14 days for the Unit 3 A-C subsystem, with compensatory measures in effect, to allow for modifications to the HPSW system and repairs to Unit 3 AHR Heat Exchanger 3CE024.

These temporary TS changes are needed to allow sufficient time to perform physical modifications of the Units 2 and 3 HPSW systems to support the proposed reduction of the HPSW design pressure.

A description and evaluation of the proposed changes are provided in this attachment.

Attachment 6 provides a copy of the marked-up TS pages that reflect the proposed changes. Attachment 7 provides the marked-up TS Bases pages (for information only).

Attachment 8 provides a summary of the regulatory commitments made in this submittal.

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation High Pressure Service Water (HPSW) Svstem The HPSW system is described in UFSAR Section 10.7, and in TS Section 3.7.1, and TS Bases B.3.7.1. The safety objective of the HPSW system is to provide a reliable supply of cooling water for AHR under post-accident conditions. The major flow paths of the HPSW

License Amendment Request Attachment 5 Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, Page 2 of 17 and 3.7.1 for Temporary Extension of Completion Times DPR-44 and DPR-56 system consist of two independent parallel flow loops serving each unit, designated Division I and Division II. Each flow loop (i.e., subsystem) contains two 4,500 gpm HPSW pumps which discharge to a common header serving two RHR heat exchangers, connected in parallel. Division I contains HPSW pumps A and C, and Division II contains HPSW pumps Band D. In the normal open loop alignment, the HPSW pumps take suction from the Conowingo Pond through the Service Water Bay and the HPSW subsystems discharge through a common pipe for each unit to the discharge pond. The discharge pipe contains a normally open motor-operated isolation valve to the pond and a pipe connection to the emergency cooling water (ECW) system to provide an alternate discharge in the unlikely event that the Conowingo Dam fails or the pond floods. When the alternate discharge is used, the emergency cooling tower (ECT) serves as a supply to the HPSW pumps through the pump bay in a closed loop alignment.

A divisional cross-tie line connecting the two HPSW subsystems on each unit is provided with a normally closed motor-operated isolation valve (M0-2(3)-32-2(3)344). The divisional cross-tie is provided with redundant power supplies and is required to be operable during Modes 1, 2 and 3 to allow HPSW pumps from one subsystem to supply RHR heat exchangers in the opposite subsystem, if required. A unit cross-tie line with two normally closed manual isolation valves HV-2-32-516A and HV-3-32-516B is also provided between one Unit 2 HPSW subsystem (B/D) and one Unit 3 HPSW subsystem (B/D). The cross-tie lines provide the flexibility to establish alternate flow alignments, if needed, under emergency conditions. A supply connection from the HPSW system to the RHR system, through two normally closed motor-operated valves, is provided from one HPSW subsystem per unit to permit the HPSW system to furnish a backup water supply to RHR for post-accident containment flooding.

Residual Heat Removal (RHR) Svstem The RHR system is described in UFSAR Section 4.8. Associated TS and corresponding TS Bases are: RHR and Low Pressure Coolant Injection (LPCI) mode of the RHR system TS 3.5.1, ECCS - Operating; TS 3.5.2, ECCS - Shutdown; TS 3.6.2.3, RHR Suppression Pool Cooling; TS 3.6.2.4, RHR Suppression Pool Spray; TS 3.6.2.5, RHR Drywell Spray; TS 3.4.7, RHR Shutdown Cooling - Hot Shutdown; TS 3.4.8, RHR Shutdown Cooling - Cold Shutdown.

The safety function of the RHR system is to restore and maintain the coolant inventory in the reactor vessel so that the core is adequately cooled after a LOCA. It also provides cooling for the containment so that condensation of the steam resulting from the blowdown due to the design basis LOCA is ensured.

The major components in the RHR system are four main system RHR pumps and four RHR heat exchangers. Four HPSW pumps for each unit support the heat removal function of the RHR system. The functional components of the RHR system are designed in accordance with seismic Class I criteria.

The RHR system is composed of two loops designated Division I and Division II. Each division includes two RHR pumps and two heat exchangers. The A and C pumps and heat exchangers are in Division I and the Band D pumps and heat exchangers are in

License Amendment Request Attachment 5 Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, Page 3 of 17 and 3.7.1 for Temporary Extension of Completion Times DPR-44 and DPR-56 Division II. Each division's two pumps, two heat exchangers, heat exchanger cross-tie line, and associated piping and valves, is referred to as a loop (i.e., subsystem). The RHR system has two loops for all modes of operation. Each loop is in a separate area of the reactor building to minimize the possibility of a single physical event causing the loss of the entire system.

RHR has multiple modes of operation including shutdown cooling, containment cooling and LPCI. In the shutdown cooling mode, it is capable of completing a normal cooldown to 125°F in about 35 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months (with two shutdown cooling subsystems in service) after the main condenser is no longer available. In this mode, the RHR system pumps reactor coolant from one of the recirculation loops through the RHR heat exchangers where cooling takes place by transferring heat to the HPSW system. It is returned to the reactor vessel via either recirculation loop.

In the containment cooling mode, the RHR system provides a means to cool the containment either through suppression pool cooling or containment spray. The safety related function of suppression pool cooling is to remove reactor core decay heat and sensible heat discharged to the suppression pool after a design basis event or accident in order to maintain the suppression pool temperature within an acceptable limit and containment pressure within an acceptable range. When in the containment cooling mode, the RHR pumps are aligned to pump water from the suppression pool through the heat exchangers. It is then either returned to the suppression pool via the full flow test line or diverted to the spray headers in the drywell and above the suppression pool.

In the LPCI mode, the RHR system operates with the High Pressure Coolant Injection (HPCI), Core Spray and automatic depressurization systems to restore and, if necessary, maintain the coolant inventory in the reactor vessel after a LOCA so that the core is sufficiently cooled to preclude excessive fuel clad temperatures and subsequent energy release due to a metal-water reaction. During LPCI operation, the RHR pumps take suction from the suppression pool and discharge into the core region of the reactor vessel through the recirculation loops.

A normally closed cross-tie line connects the AHR pump discharge headers upstream of the AHR heat exchangers within each loop. When only a single AHR pump is available, additional post-LOCA containment cooling capacity can be achieved by opening the cross-tie valve and aligning AHR flow to both RHR heat exchangers within a loop. This also requires starting a second HPSW pump to provide cooling to the second AHR heat exchanger. Flow control valves upstream of each of the AHR heat exchangers, in conjunction with flow elements with Main Control Room indication at the outlet of each heat exchanger, allow operators to balance flow for containment cooling and prevent RHR pump run out.

2.2 Current Technical Specifications Requirements PBAPS TS LCO 3.6.2.3 requires the operability of two AHR suppression pool cooling subsystems when the reactor is in Modes 1, 2, and 3. TS 3.6.2.3, Condition A, and the associated Required Action A.1 address the inoperability of one subsystem. Specifically, Required Action A.1 requires restoration of one AHR suppression pool cooling subsystem to

License Amendment Request Attachment 5 Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, Page 4 of 17 and 3.7.1 for Temporary Extension of Completion Times DPR-44 and DPR-56 operable status, with a CT of 7 days. If Required Action A.1 cannot be satisfied within the CT, Condition Band associated Required Action B.1 require the reactor to be in Mode 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

PBAPS TS LCO 3.6.2.4 requires the operability of two AHR suppression pool spray subsystems when the reactor is in Modes 1, 2, and 3. TS 3.6.2.4, Condition A, and the associated Required Action A.1 address the inoperability of one subsystem. Specifically, Required Action A.1 requires restoration of one AHR suppression pool spray subsystem to operable status, with a CT of 7 days. If Required Action A.1 cannot be satisfied within the CT, Condition C and associated Required Action C.1 require the reactor to be in Mode 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

PBAPS TS LCO 3.6.2.5 requires the operability of two AHR drywell spray subsystems when the reactor is in Modes 1, 2, and 3. TS 3.6.2.5, Condition A, and the associated Required Action A.1 address the inoperability of one subsystem. Specifically, Required Action A.1 requires restoration of one AHR drywell spray subsystem to operable status, with a CT of 7 days. If Required Action A.1 cannot be satisfied within the CT, Condition C and associated Required Action C.1 require the reactor to be in Mode 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

PBAPS TS LCO 3.7.1 requires the operability of two HPSW subsystems and the HPSW cross-tie when the reactor is in Modes 1, 2, and 3. TS 3.7.1, Condition A, and the associated Required Action A.1 address the inoperability of one subsystem. Specifically, Required Action A.1 requires restoration of one HPSW subsystem to operable status, with a CT of 7 days. If Required Action A.1 cannot be satisfied within the CT, Condition C and associated Required Action C.1 require the reactor to be in Mode 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . TS

3. 7 .1, Condition B, and the associated Required Action B.1 address the inoperability of the HPSW cross-tie line. Specifically, Required Action B.1 requires restoration of the HPSW cross-tie to operable status, with a CT of 7 days. If Required Action B.1 cannot be satisfied within the CT, Condition C and associated Required Action C.1 require the reactor to be in Mode 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.3 Proposed Change The proposed change will temporarily revise the PBAPS, Units 2 and 3, TS Sections 3.6.2.3, "Residual Heat Removal (AHR) Suppression Pool Cooling," TS 3.6.2.4, "Residual Heat Removal (RHR) Suppression Pool Spray," TS 3.6.2.5, "Residual Heat Removal (AHR)

Drywell Spray," TS 3.7.1, "High Pressure Service Water (HPSW) System," as follows:

TS Section 3.6.2.3 will be revised to extend, on a one-time basis, four (4) allowable Completion Times (CTs) of Required Action A.1 for one RHR suppression pool cooling subsystem inoperable, from 7 days to 1O days, for Unit 2; and three (3) CTs of Required Action A.1 for one AHR suppression pool cooling subsystem inoperable, from 7 days to 1O days, for Unit 3, with an additional CT of up to 14 days for the A-C subsystem only, to allow for repairs to Unit 3 AHR Heat Exchanger 3CE024.

TS Section 3.6.2.4 will be revised to extend, on a one-time basis, four (4) allowable CTs of Required Action A.1 for one AHR suppression pool spray subsystem inoperable, from 7 days to 1O days, for Unit 2; and three (3) CTs of Required Action

License Amendment Request Attachment 5 Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, Page 5 of 17 and 3.7.1 for Temporary Extension of Completion Times DPR-44 and DPR-56 A.1 for one AHR suppression pool spray subsystem inoperable, from 7 days to 1O days, for Unit 3, with an additional CT of up to 14 days for the A-C subsystem only, to allow for repairs to Unit 3 AHR Heat Exchanger 3CE024.

TS Section 3.6.2.5 will be revised to extend, on a one-time basis, four (4) allowable CTs of Required Action A.1 for one AHR drywell spray subsystem inoperable, from 7 days to 1O days, for Unit 2; and three (3) CTs of Required Action A.1 for one AHR drywell spray subsystem inoperable, from 7 days to 10 days, for Unit 3, with an additional CT of up to 14 days for the A-C subsystem only, to allow for repairs to Unit 3 AHR Heat Exchanger 3CE024.

TS Section 3.7.1 will be revised to extend, on a one-time basis, four (4) allowable CTs of Required Action A.1 for one HPSW subsystem inoperable, from 7 days to 1O days, for Unit 2; and three (3) CTs of Required Action A.1 for one HPSW subsystem inoperable, from 7 days to 10 days, for Unit 3, with an additional one (1) CT of up to 14 days for the A-C subsystem only, to allow for repairs to Unit 3 AHR Heat Exchanger 3CE024. Also, TS 3.7.1 will be revised to extend, on a one-time basis, two (2) allowable CTs of Required Action 8.1 for one HPSW subsystem inoperable, from 7 days to 1O days, for Unit 2; and one (1) allowable CT of Required Action B.1 for one HPSW subsystem inoperable, from 7 days to 1O days, for Unit 3, with an additional one (1) CT of up to 14 days for the A-C subsystem only, to allow for repairs to Unit 3 AHR Heat Exchanger 3CE024.

A footnote will be added to each of the affected TS LCOs that indicates the 7-day CTs for the affected systems may be extended up to four times for Required Action A.1, and up to two times for TS 3.7.1 Required Action 8.1, for a period of up to 1O days or 14 days for the Unit 3 A-C subsystem only, with compensatory measures in effect, to allow for modifications to the HPSW system and repairs to Unit 3 AHR Heat Exchanger 3CE024.

2.4 Reason for the Proposed Changes The physical modifications required to reduce HPSW system design pressure are expected to be installed with the plant on-line (Mode 1). PBAPS TS 3.7.1 requires that two HPSW subsystems shall be operable during Modes 1, 2 and 3. It has been determined that the current HPSW TS 3.7.1, Required Action A.1 Completion Time of 7 days is not sufficient time to replace both HPSW pumps in a loop, or to replace the M0-2(3)-10-089A/8/C/D valves (M0-089 valves) and remove flow orifices from a single subsystem. Also, the TS 3.7.1 Required Action 8.1 CT for the HPSW cross-tie line inoperable, is not sufficient time to replace M0-089 valves and remove flow orifices from a single subsystem. An extension of the CTs will be required for TS 3.7.1, and for TS 3.6.2.3, 3.6.2.4, and 3.6.2.5, due to their dependency on the HPSW system for operability. Based on evaluation of the installation tasks, it has been determined that 10 days will provide sufficient time, with margin, to perform each of the installation tasks for a HPSW subsystem. Therefore, a temporary CT extension from 7 days to 1O days is requested for each of these TS, with the exception of a CT extension from 7 days to 14 days for the Unit 3 A-C subsystem only, to allow for repairs to the Unit 3 AHR Heat Exchanger 3CE024

License Amendment Request Attachment 5 Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, Page 6 of 17 and 3.7.1 for Temporary Extension of Completion Times DPR-44 and DPR-56 The installation plan will require that the modifications be performed in phases, requiring multiple entries into the LCO action statements. System operability must be maintained with interim configurations during installation. With the implementation plan outlined below, the requirements of TS 3.6.2.3, 3.6.2.4, 3.6.2.5, and 3.7.1 will be met, with subsystem operability restored after each interim LCO step. With the M0-089 valves replaced and orifices removed on a given HPSW subsystem, that subsystem will be operable with either the new low head pumps or with the existing high head pumps. However, a subsystem would not be operable with the new low head pumps and existing M0-089 valves or orifices installed, due to the hydraulic resistance represented by the existing valves/orifices.

The Unit 2/3 HPSW system cross-tie was designed to provide operational flexibility to establish alternate flow alignments, if needed, under emergency conditions. It is normally isolated with locked closed manual valves HV-2-32-516A and HV-3-32-5168. Operation with the Unit 2/3 HPSW cross-tie valves open is not governed by Technical Specifications, but is controlled by the Fire Protection Program. Loss of unit cross-tie functionality during the interim period will not adversely affect plant operation in accordance with the Technical Specifications. However, HPSW operation with the Unit 2/3 cross-tie opened is credited for postulated Appendix R fire events in Fire Areas 47 or 48 where all HPSW pumps on one unit may be lost. In order to maintain the Units 2/3 HPSW system cross-tie functional (via manual valves HV-2-32-516A and HV-3-32-5168), all M0-089 valve and orifice work on both units must be performed prior to pump replacement on either unit.

In addition, the Unit 3 RHR Heat Exchanger 3CE024 currently has a leak that requires repair. Therefore, the repair activity will be performed during the proposed extended LCO completion time for Unit 3 subsystem A-C of up to 14 days, since the repairs will require more than 7 days, and the proposed changes to replace the M0-089A/C valves and remove orifices and the heat exchanger repairs together will require more than 1O days.

Installation Plan: The proposed installation plan is outlined below. The lead unit can be either Unit 2 or Unit 3. Also, the HPSW subsystems may be modified in either order - i.e.,

the 8-D subsystem may be worked prior to the A-C subsystem.

Lead Unit

1st LCO - Replace M0-089A/C valves and remove orifices from A-C subsystem

2nd LCO - Replace M0-0898/D valves and remove orifices from 8-D subsystem Lag Unit

3rd LCO - Replace M0-089A/C valves and remove orifices from A-C subsystem

4th LCO - Replace M0-0898/D valves and remove orifices from 8-D subsystem Lead Unit

5th LCO- Replace both HPSW pumps on one subsystem (A-C)

6th LCO - Replace both HPSW pumps on opposite subsystem (8-D)

License Amendment Request Attachment 5 Revise Technical Specifications 3.5.2.3, 3.5.2.4, 3.5.2.5, Page 7 of 17 and 3.7.1 for Temporary Extension of Completion Times DPR-44 and DPR-55 Lag Unit

7th LCO- Replace both HPSW pumps on one subsystem (A-C)

9th LCO - Replace both HPSW pumps on opposite subsystem (8-D)

For the 1st and 2nct LCOs on the lead unit and the 3rc1 and 4th LCOs on the lag unit, piping and valve modifications will result in inoperability of the associated HPSW subsystem, as well as the HPSW cross-tie line. Therefore, HPSW TS 3.7.1 Required Actions A.1 and 8.1 will be entered. For the 5th and 5th LCOs on the lead unit and the 7th and 8th LCOs on the lag unit, the pump modifications will also result in inoperability of the associated HPSW subsystem. However, the HPSW cross-tie line, which connects downstream of the HPSW pumps, would still be operable. Therefore, HPSW TS 3.7.1 Required Action A.1 will be entered, but not Required Action 8.1. This results in the need for only two extended CTs per unit for Required Action 8.1.

For the interim conditions between 5th and 5th LCO entries and between 7th and 9th LCO entries, where two new low head HPSW pumps are installed in one loop and existing high head pumps remain in the opposite loop on a given unit, an alternative HPSW alignment must be considered. To mitigate a D8A-LOCA, two HPSW pumps are credited for containment cooling on the accident unit. Due to emergency diesel generator (EOG) loading limitations with a postulated failure of one EOG, one HPSW pump from each division on the accident unit is credited, with one pump delivering flow through the divisional cross-tie. In the interim configuration, this means that a high head HPSW pump must be capable of operating in parallel with a new low head pump. In this case, pump flows would be imbalanced, and flow from the high head pump must be throttled in order to prevent the high head pump from operating in a potentially damaging runout condition (too far out on its curve).

For this interim configuration, manual valves in the HPSW divisional cross-tie will be maintained in a throttled intermediate position that will provide the necessary pressure drop to prevent runout of the high head pump.

In summary, the extended HPSW and RHR TS LCO CTs for Required Action A.1 will be entered four (4) times for each unit, and the HPSW TS 3.7.1 Required Action 8.1 CT will be entered two (2) times for each unit.

3.0 TECHNICAL EVALUATION

3.1 Deterministic Evaluation Exelon has evaluated the proposed one-time extension of TS 3.5.2.3, 3.5.2.4, 3.5.2.5, and 3.7.1 Required Action A.1 CTs and TS 3.7.1 Required Action 8.1 from 7 days up to 10 or 14 days deterministically, supplemented with nuclear risk insights.

To ensure that the single failure design criterion is met, LCOs are specified in the plant TS requiring all redundant components of the HPSW and RHR systems to be operable. In the event that a HPSW, RHR suppression pool cooling, RHR suppression pool spray, and RHR drywell spray subsystem are inoperable in Modes 1, 2, and 3, existing TS

License Amendment Request Attachment 5 Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, Page 8 of 17 and 3.7.1 for Temporary Extension of Completion Times DPR-44 and DPR-56 requirements permit continued operation for up to 7 days. In this condition, the remaining HPSW subsystem, RHR suppression pool cooling subsystem, RHR suppression pool spray subsystem, and RHR drywall spray subsystem are adequate to perform their design function. However, the overall reliability is reduced because a single failure in the operable subsystem could result in reduced primary containment cooling capability. The current 7-day TS LCO CT is acceptable in light of the redundant HPSW and RHR suppression pool and drywell cooling capabilities afforded by the operable subsystems and the low probability of a design basis accident occurring during this period. With the proposed 10-day or 14-day CT for one HPSW or RHR subsystem inoperable, there will be an increase in the probability of a design basis accident occurring during the extended CT.

However, the probability is still considered low.

During installation of the HPSW modification on one subsystem with the proposed 10-day or 14-day CT, the opposite HPSW subsystem will be operable with two HPSW pumps.

The opposite RHR subsystem will also be operable with two RHR pumps and two heat exchangers. The TS requirement to have two HPSW pumps operable in each subsystem is based on the plant safety analysis for design basis events that require use of the RHR cross-tie to provide containment heat removal capability using one RHR pump and two cross-tied RHR heat exchangers. Events that credit the RHR cross-tie are: OBA LOCA, small steam line break (SSLB), safety-relief valve transient (SRVT), or loss of normal shutdown cooling, combined with a LOOP and single failure of one EOG. Each of these events has a low probability of occurrence. Operation of one RHR pump with one RHR heat exchanger and one HPSW pump, without use of the RHR cross-tie, is sufficient for mitigating other design basis events, including those having a higher probability of occurrence.

Since a single failure is not considered while the unit is in an LCO Action Statement, the operable redundant equipment is capable of performing its required function and maintaining the plant design basis. Thus, the requested extension of the TS LCO CTs will not alter the UFSAR assumptions relative to the mitigation of a design basis accident or transient.

To provide further risk reduction, and consistent with risk insights discussed below, the evaluated completion time of 14 days was reduced to 10 days, where possible, by optimizing preliminary work planning for modification and maintenance activities. However, it was determined that the 14-day CT extension is still required to perform additional Unit 3 RHR heat exchanger leak repairs.

3.2 Nuclear Risk Insights This section provides Probabilistic Risk Assessment (PRA) information to support a one-time, deterministic LAR for TS 3.6.2.3, 3.6.2.4, 3.6.2.5, and 3.7 .1 CT extensions. 10 CFR 50.65 (a)(4), "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," requires that prior to performing maintenance activities, risk assessments shall be performed to assess and manage the increase in risk that may result from proposed maintenance activities. These requirements are applicable for all plant modes.

The proposed LAR will not result in any changes to the current configuration risk management program. The existing program uses a blended approach of quantitative and

License Amendment Request Attachment 5 Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, Page 9 of 17 and 3.7.1 for Temporary Extension of Completion Times DPA-44 and DPR-56 qualitative evaluation of each configuration assessed. Thus, the overall change in plant risk during maintenance activities is anticipated to be addressed adequately considering the proposed amendment.

This license amendment request is not a risk-informed request and, therefore, a risk evaluation is not required. However, to provide additional information, Exelon is providing risk insights related to the proposed change.

Although this technical analysis is based on a deterministic evaluation, a risk analysis was performed that demonstrated with reasonable assurance that the proposed TS changes are within the current risk acceptance guidelines in RG 1.177 for one-time changes. This ensures that the TS change meets the inten.t of the Incremental Conditional Core Damage Probability (ICCDP) and Incremental Conditional Large Early Release Probability (ICLERP) acceptance guidelines of 1.0E-05 (actual 5E-06 for Unit 2 and 6E-06 for Unit 3) and 1.0E-06 (actual 3E-07 for Unit 2 and 5E-07 for Unit 3) established for compatibility with the ICCDP and ICLERP limits of RG 1.177, which is applicable for configuration changes that require normal work controls. The risk analysis was based on summation of the ICCDP and ICLERPs for the following unavailability configurations without credit for additional proposed Risk Management Actions (RMAs). The analysis was performed for each unit. The unavailability of the HPSW loop and pumps impacts the AHR system via the logic in the PAA model.

HPSW and AHR Loop A (10-day, Unit 2 and 14-day, Unit 3 allowed outage time (AOT))

HPSW and AHR Loop B (10-day AOT)

HPSW Pumps A/C (unavailable in series over the 10-day AOT)

HPSW Pumps BID (unavailable in series over the 10-day AOT)

The identification of the AMAs was derived from a detailed review of the results of the risk assessment. None of the RMAs were credited in the base risk analysis; the identified compensatory actions would further lessen the overall risk incurred during the extended periods.

Compensatory actions that are outlined in Section 3.3 provide additional assurance that the risk of the one-time TS LCO CT extensions will be minimized.

3.3 Compensatory Actions The following compensatory measures will be implemented to support the proposed license amendment request:

Adequate staffing will be maintained onsite to facilitate timely response to unexpected conditions during the extended LCO CTs authorized by the proposed license amendment.

Equipment will be protected in accordance with the Protected Equipment Program, as described in Procedure OP-AA-108-117. Protected equipment actions taken in accordance with this procedure support the Configuration Risk Management

License Amendment Request Attachment 5 Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, Page 10 of 17 and 3.7.1 for Temporary Extension of Completion Times DPR-44 and DPR-56 Program and are classified as RMAs for the purpose of compliance with 10 CFR 50.65(a)(4). In addition to the protected opposite HPSW and AHR subsystems required to be operable by TS, elective maintenance, discretionary maintenance and testing on systems that provide support to the protected subsystems (e.g., Emergency Diesel Generators) will be suspended during the extended LCO CTs authorized by the proposed license amendment. Additionally, the following actions will be taken prior to entry into the proposed configuration:

Proper standby alignment of the opposite HPSW and AHR subsystems will be ensured prior to entry into the extended CTs.

Component testing or maintenance of safety systems in the off-site power systems and important non-safety equipment in the off-site power systems which can increase the likelihood of a plant transient or LOOP, as determined by plant management, will be avoided during the extended LCO CTs authorized by the proposed license amendment.

Discretionary substation maintenance shall not be allowed during the extended LCO CTs authorized by the proposed license amendment.

The HPCI pump, Reactor Core Isolation Cooling (RCIC) pump, and AHR pumps will not be removed from service for elective maintenance activities during the extended LCO CTs authorized by the proposed license amendment.

Weather conditions will be monitored and the HPSW modification work will not be scheduled if severe weather conditions are anticipated.

The "N+ 1" Flex Pump and Flex Generator will be pre-staged inside the site protected area to allow for more rapid deployment in the event of a LOOP with concurrent Emergency Diesel Generator failures. The use of the "N+ 1" equipment allows the required FLEX gear to be retained within the protected building, preserving them for an actual FLEX event. This will be controlled in accordance with exiting plant procedures.

Additionally, during the extended LCO CT authorized by the proposed license amendment, Operations shift crews will be briefed at the beginning of each shift regarding actions in response to a Loss of Offsite Power (LOOP) per applicable plant procedures. Also, all required Fire Risk Management Actions (RMAs) will be performed in accordance with site procedures that fulfill the requirements of 10 CFR 50.65(a)(4).

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.36(c) provides that TS will include Limiting Conditions for Operation (LCOs) which are "the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is

License Amendment Request Attachment 5 Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, Page 11 of 17 and 3.7.1 tor Temporary Extension of Completion Times DPR-44 and DPR-56 not met, the licensee will shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met." The proposed changes involve extensions of the affected Completion Times from 7 days to 10 days and from 7 days to 14 days tor the Unit 3 A-C subsystem only. The LCOs themselves remain unchanged, as do the required remedial actions or shutdown requirements in accordance with 10 CFR 50.36(c). Therefore, the proposed changes are consistent with current regulations.

As stated in Appendix Hof the 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.

AEC Criterion 4 - Sharing of Systems (Category A)

"Reactor facilities shall not share systems or components unless it is shown safety is not impaired by the sharing."

AEC Criterion 6 - Reactor Core Design (Category A)

"The reactor core shall be designed to function throughout its design lifetime, without exceeding acceptable fuel damage limits which have been stipulated and justified. The core design, together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of all offsite power. "

AEC Criterion 1O - Containment (Category A)

"Containment shall be provided. The containment structure shall be designed to sustain the initial effects of gross equipment failures, such as a large coolant boundary break, without loss of required integrity and, together with other engineered safety features as may be necessary, to retain for as long as the situation requires the functional capability to protect the public. "

AEC Criterion 14 - Core Protection Systems (Category 8)

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

AEC Criterion 37 - Engineered Safety Features Basis tor Design (Category A)

"Engineered safety features shall be provided in the facility to back up the safety provided by the core design, the reactor coolant pressure boundary, and their protection systems. As a minimum, such engineered safety features shall be designed to cope with any size reactor

License Amendment Request Attachment 5 Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, Page 12 of 17 and 3. 7 .1 for Temporary Extension of Completion Times DPR-44 and DPR-56 coolant pressure boundary break up to and including the circumferential rupture of any pipe in that boundary assuming unobstructed discharge from both ends. "

AEC Criterion 38- Reliabilitv and Testability of Engineered Safetv Features (Category A)

"All engineered safety features shall be designed to provide high functional reliability and ready testability. In determining the suitability of a facility for a proposed site, the degree of reliance upon and acceptance of the inherent and engineered safety afforded by the systems, including engineered safety features, will be influenced by the known and the demonstrated performance capability and reliability of the systems, and by the extent to which the operability of such systems can be tested and inspected where appropriate during the life of the plant. "

AEC Criterion 41 - Engineered Safetv Features Performance Capabilitv (Category A)

"Engineered safety features such as emergency core cooling and containment heat removal systems shall provide sufficient performance capability to accommodate partial Joss of installed capacity and still fulfill the required safety function. As a minimum, each engineered safety feature shall provide this required safety function assuming a failure of a single active component. "

AEC Criterion 42 - Engineered Safety Features Components Capability (Category A)

"Engineered safety features shall be designed so that the capability of each component and system to perform its required function is not impaired by the effects of a loss-of-coolant accident."

AEC Criterion 44 - Emergency Core Cooling Systems Capabilitv (Category A)

"At least two emergency core cooling systems, preferably of different design principles, each with a capability for accomplishing abundant emergency core cooling, shall be provided. Each emergency core cooling system and the core shall be designed to prevent fuel and clad damage that would interfere with the emergency core cooling function and to limit the clad metal-water reaction to negligible amounts for all sizes of breaks in the reactor coolant pressure boundary, including the double-ended rupture of the largest pipe. The performance of each emergency core cooling system shall be evaluated conservatively in each area of uncertainty. The systems shall not share active components and shall not share other features or components unless it can be demonstrated that (a) the capability of the shared feature or component to perform its required function can be readily ascertained during reactor operation, (b) failure of the shared feature or component does not initiate a loss-of-coolant accident, and (c) capability of the shared feature or component to perform its required function is not impaired by the effects of a Joss-of-coolant accident and is not Jost during the entire period this function is required following the accident."

AEC Criterion 49 - Containment Design Basis (Category A)

"The containment structure, including access openings and penetrations, and any necessary containment heat removal systems shall be designed so that the containment

License Amendment Request Attachment 5 Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, Page 13 of 17 and 3.7.1 for Temporary Extension of Completion Times DPR-44 and DPR-56 structure can accommodate without exceeding the design leakage rate the pressures and temperatures resulting from the largest credible energy release following a loss-of-coolant accident, including a considerable margin for effects from metal water or other chemical reactions that could occur as a consequence of failure of emergency core cooling systems."

AEC Criterion 52 - Containment Heat Removal Systems (Category A)

"Where active heat removal systems are needed under accident conditions to prevent exceeding containment design pressure, at least two systems, preferably of different principles, each with full capacity, shall be provided. "

AEC Criterion 70 - Control of Releases of Radioactivity to the Environment (Category 8)

"The facility design shall include those means necessary to maintain control over the plant radioactive effluents, whether gaseous, liquid, or solid. Appropriate holdup capacity shall be provided for retention of gaseous, liquid, or solid effluents, particularly where unfavorable environmental conditions can be expected to require operational limitations upon the release of radioactive effluents to the environment. In all cases, the design for radioactivity control shall be justified (a) on the basis of 10CFR20 requirements for normal operations and for any transient situation that might reasonably be anticipated to occur and (b) on the basis of 10CFR100 dosage level guidelines for potential reactor accidents of exceedingly low probability of occurrence except that reduction of the recommended dosage levels may be required where high population densities or very large cities can be affected by the radioactive effluents."

The proposed license amendment for a one-time extension of TS LCO CTs from 7 days to 1O days, and from 7 days to 14 days for the Unit 3 A-C subsystem only, does not add or delete any safety-related systems, equipment, or loads, or alter the design or functions of the HPSW or RHR systems. Therefore, based on the evaluation documented above, PBAPS will continue to comply with all applicable regulatory requirements.

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 to the health and safety of the public.

4.2 Precedent This proposed license amendment is similar to amendments submitted by Limerick Generating Station on March 19, 201 O (ADAMS Accession No. ML100810151) and approved by the NRC on July 29, 2011 (ADAMS Accession No. ML111960066). The Limerick amendment temporarily modified their Technical Specifications to allow an extension of the allowed outage time from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 7 days for the Residual Heat Removal system, the Residual Heat Removal Service Water (RHRSW) system, the Emergency Service Water system, and the Diesel Generators.

License Amendment Request Attachment 5 Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, Page 14 of 17 and 3.7.1 for Temporary Extension of Completion Times DPR-44 and DPR-56 This proposed license amendment is also similar to amendments submitted for PWR cooling water systems:

South Texas Project Unit 1: extensive, unplanned repairs were necessary for the 'B' Train Essential Cooling Water (ECW) pump. The STP request for a 7-day extension (up to 14 days) for allowed outage times for ECW and systems supported by ECW was approved by the NRC on January 10, 2005 (ADAMS Accession No. ML050100291 ).

Callaway Nuclear Station: extension of the Technical Specification essential service water (ESW) Completion Times from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 14 days for ESW and systems supported by ESW to replace ESW piping, approved February 24, 2009 (ADAMS Accession No. ML090360533).

4.3 No Significant Hazards Consideration Exelon Generation Company, LLC (Exelon) requests a temporary revision to Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3, Technical Specification (TS)

Sections 3.6.2.3, "Residual Heat Removal (RHR) Suppression Pool Cooling," TS 3.6.2.4, "Residual Heat Removal (RHR) Suppression Pool Spray," TS 3.6.2.5, "Residual Heat Removal (RHR) Drywell Spray," and TS 3.7.1, "High Pressure Service Water (HPSW)

System," to extend, on a one-time basis, four (4) allowable Completion Times (CTs) of Required Action A.1 and two (2) allowable CTs of TS 3.7.1 Required Action B.1 from 7 days to 1O days, for Unit 2; and three (3) allowable CTs of Required Action A.1 and one (1) allowable CT of TS 3.7.1 Required Action 8.1 from 7 days to 1O days, for Unit 3, and one (1) allowable CT of Required Action A.1 and one (1) allowable CT of TS 3.7.1 Required Action B.1 from 7 days to 14 days for the Unit 3 A-C subsystem only.

Exelon has evaluated whether or not a significant hazards consideration is involved with the proposed changes in accordance with the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below.

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

No. The HPSW system does not initiate any accidents discussed in Chapter 14 of the PBAPS Units 2 and 3 UFSAR. The RHR system is not an initiator for any accidents described in the UFSAR, except for a shutdown cooling malfunction

- decreasing temperature. The event is classified as an abnormal operational transient. A shutdown cooling malfunction leading to a moderator temperature decrease could result from misoperation of the cooling water controls for the AHR heat exchangers, as described in UFSAR Section 14.5.2.4. The resulting temperature decrease causes a slow insertion of positive reactivity into the core.

However, the proposed TS changes will not affect the initiator for this accident.

The extension of the time duration that one HPSW or AHR subsystem is out of service has no direct physical impact on the plant. The proposed changes do not introduce any failure mechanisms that would initiate a previously analyzed accident. The HPSW and RHR subsystems are normally in a standby mode while the unit is Mode 1 and are not directly supporting plant operation. Therefore, they

License Amendment Request Attachment 5 Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, Page 15 of 17 and 3.7.1 for Temporary Extension of Completion Times DPR-44 and DPR-56 can have no impact on the plant that would make an accident more likely to occur due to subsystem inoperability. During transients or events which require these subsystems to be operating, there is sufficient capacity in the operable HPSW and RHR subsystems to support plant operation or shutdown. Therefore, failures that are accident initiators will not occur more frequently than previously postulated as a result of the proposed changes. The HPSW and AHR systems remain capable of performing their UFSAR-described design functions. Moreover, the design and operability requirements currently addressed by the PBAPS Technical Specifications (TS) are unaffected. Thus, the consequences of analyzed accidents are not increased.

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

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

No. The proposed TS changes will not create the possibility of a different type of accident since they will only extend the time period that one HPSW, RHR suppression pool cooling, RHR suppression pool spray, and RHR drywell spray subsystem can be out of service. The extension of the time durations has no direct physical impact on the plant and does not create any new accident initiators. The systems involved are accident mitigation systems. This change will not introduce a new mode of plant operation and there is no alteration of the HPSW or AHR system design functions or the ability of these systems to perform their design functions.

Therefore, the proposed changes do 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?

No. The present HPSW and RHR TS CTs were set to ensure that sufficient safety-related equipment is available for response to all accident conditions and that sufficient decay heat removal capability is available for a loss of coolant accident (LOCA) coincident with a loss of offsite power (LOOP), and simultaneous safe shutdown of the other unit. A slight reduction in the margin of safety is incurred during the proposed extended CT due to the increased risk that an event could occur in a 10-day or 14-day period versus a 7-day period. This increased risk is judged to be minimal due to the low probability of an event occurring during the extended CT and since the remaining operable HPSW and RHR equipment is adequate to mitigate the consequences of any accident.

These minimum requirements will be met since implementation of the proposed TS changes will require the operability of one HPSW subsystem, one RHR suppression pool cooling subsystem, one RHR suppression pool spray subsystem, and one RHR drywall spray system be maintained during the 10-day

License Amendment Request Attachment 5 Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, Page 16 of 17 and 3.7.1 for Temporary Extension of Completion Times DPR-44 and DPR-56 and 14-day periods. The proposed amendment does not involve any change to system design functions and does not alter the safety limits or safety analysis assumptions associated with the operation of the plant.

Operations personnel are fully qualified by normal periodic training to respond to and mitigate a Design Basis Accident, including the actions needed to ensure decay heat removal while the plant is in the extended LCO CT described within this submittal. Accordingly, procedures are already in place that address safe plant shutdown and decay heat removal for situations applicable to those in the proposed CT extensions.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

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.

Based on the above evaluation, Exelon 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.

5.0 ENVIRONMENTAL CONSIDERATION

A review 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, or would change an inspection or surveillance requirement.

However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents 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(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. Regulatory Guide 1.160, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Revision 3, dated May 2012 (ADAMS Accession No. ML113610098)

License Amendment Request Attachment 5 Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, Page 17 of 17 and 3. 7 .1 for Temporary Extension of Completion Times DPR-44 and DPR-56

2. Letter to NRG for Limerick Generating Station, Units 1and2- Proposed Changes to Technical Specifications Sections 3.5.1, 3.6.2.3, 3.7.1.1, 3.7.1.2 and 3.8.1.1 to Extend the Allowed Outage Times, dated March 19, 2010 (ADAMS Accession No. ML100810151)

3. Letter from NRG to Limerick Generating Station, Units 1 and 2 - Issuance of Amendments Re: Allowed Outage Time Extensions To Support Residual Heat Removal Service Water Maintenance, dated July 29, 2011 (ADAMS Accession No. ML111960066)

4. Letterfrom NRG to South Texas Project Unit 1 - Issuance of Amendment Concerning One-Time Allowed Outage Time Extension for Train B Essential Cooling Water, dated January 1O, 2005 (ADAMS Accession No. ML050100291 ).

5. Letter from NRG to Callaway Nuclear Station - Issuance of Amendment Re:

One-Time Extension of Completion Time for Train B of the Essential Service Water System Piping Replacement and Alternating Current (AC) Sources, dated February 24, 2009 (ADAMS Accession No. ML090360533)

Attachment 6 Peach Bottom Atomic Power Station Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 Markup of Technical Specifications Pages Unit 2 TS Pages 3.6-27 3.6-29 3.6-30a 3.7-1 Unit 3 TS Pages 3.6-27 3.6-29 3.6-30a 3.7-1

RHR Suppression Pool Cooling 3.6.2.3 3.6 CONTAINMENT SYSTEMS 3.6.2.3 Residual Heat Removal CRHR) Suppression Pool Cooling LCD 3.6.2.3 Two RHR suppression pool cooling subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR suppression A. l Restore RHR 7 daysW pool cooling subsystem suppression pool inoperable. cooling subsystem to OPERABLE status.

B. Required Action and B.l Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A not met.

c. Two RHR suppression C.l Restore one RHR 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months pool cooling suppression pool subsystems inoperable. cooling subsystem to OPERABLE status.

D. Required Action and D.l Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition C AND not met.

D.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

The 7-day Completion Time for one RHR suppression pool cooling subsystem inoperable may be extended to 10 days four (4) times until December 31, 2020 with compensatory measures identified in EGC License Amendment Request letter dated September 28, 2018 established and in effect, to allow for modifications to the HPSW system.

PBAPS UN IT 2 3.6-27 Amendment No. 261

RHR Suppression Pool Spray 3.6.2.4 3.6 CONTAINMENT SYSTEMS 3.6.2.4 Residual Heat Removal CRHR) Suppression Pool Spray LCD 3.6.2.4 Two RHR suppression pool spray subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR suppression A.l Restore RHR 7 days8 pool spray subsystem suppression pool inoperable. spray subsystem to OPERABLE status.

B. Two RHR suppression B.l Restore one RHR 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months pool spray subsystems suppression pool inoperable. spray subsystem to OPERABLE status.

c. Required Action and C.l Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Ti me not met.

The 7-day Completion Time for one RHR suppression pool spray subsystem inoperable may be extended to 10 days four (4) times until December 31, 2020 with compensatory measures identified in EGC License Amendment Request letter dated September 28, 2018 established and in effect, to allow for modifications to the HPSW system .

PBAPS UN IT 2 3.6-29 Amendment No. 261

RHR Drywell Spray 3.6.2.5 3.6 CONTAINMENT SYSTEMS 3.6.2.5 Residual Heat Removal (RHR) Drywell Spray LCO 3.6.2.5 Two RHR drywell spray subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR drywel l spray A.l Restore RHR drywell 7 daysW subsystem inoperable. spray subsystem to OPERABLE status.

B. Two RHR drywel l spray B.1 Restore one RHR 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months subsystems inoperable. drywell spray subsystem to OPERABLE status.

c. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met. AND C.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

The 7-day Completion Time for one RHR drywell spray subsystem inoperable may be extended to 10 days four (4) times until December 31, 2020 with compensatory measures identified in EGC License Amendment Request letter dated September 28, 2018 established and in effect, to allow for modifications to the HPSW system.

PBAPS UN IT 2 3.6-30a Amendment No. 288

HPSW System

3. 7 .1 3.7 PLANT SYSTEMS 3.7.1 High Pressure Service Water CHPSW) System LCO 3.7.1 Two HPSW subsystems and the HPSW cross tie shall be OPERABLE.

APPLICABILITY: MODES l, 2, and 3.

ACT! ONS CONDITION REQUIRED ACT! ON COMPLETION TIME A. One HPSW subsystem -----------NOTE-------------

inoperable. Enter applicable Conditions and Required Actions of LCO 3.4.7, "Residual Heat Removal CRHR) Shutdown Cooling System-Hot Shutdown," for RHR shutdown cooling made inoperable by HPSW System.

A.1 Restore HPSW 7 daysW subsystem to OPERABLE status.

B. HPSW cross tie B.1 Restore HPSW cross 7 days8 inoperable. tie to OPERABLE status.

C. Required Action and C.l Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A or B not met.

(continued)

The 7-day Completion Time for one HPSW subsystem inoperable may be extended to 10 days four (4) times until December 31, 2020 and the 7-day Completion Time for the HPSW cross tie inoperable may be extended to 10 days two (2) times until December 31, 2020 with compensatory measures identified in EGC License Amendment Request letter dated September 28, 2018 established and in effect, to allow for modifications to the HPSW system.

PBAPS UN IT 2 3.7-1 Amendment No. 293

RHR Suppression Pool Cooling 3.6.2.3 3.6 CONTAINMENT SYSTEMS 3.6.2.3 Residual Heat Removal CRHR) Suppression Pool Cooling LCO 3.6.2.3 Two RHR suppression pool cooling subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR suppression A.1 Restore RHR 7 days8 pool cooling subsystem suppression pool inoperable. cooling subsystem to OPERABLE status.

B. Required Action and B.l Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A not met.

C. Two RHR suppression C.1 Restore one RHR 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months pool cooling suppression pool subsystems inoperable. cooling subsystem to OPERABLE status.

D. Required Action and D.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition C AND not met.

D.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

' ,,, ,,,

The 7-day Completion Time for one RHR suppression pool cooling subsystem inoperable may be extended ..

~

to 10 days (3) times and to 14 days one (1) time (A-C subsystem only) until December 31,2020 with .

..

~compensatory measures identified in EGC License Amendment Request letter dated September 28, 2018 ..

~established and in effect, to allow for modifications to the HPSW system and repairs to Unit 3 RHR Heat )

Exchanger 3CE024.

................................... . ................ . ..

PBAPS UN IT 3 3.6-27 Amendment No. 265

RHR Suppression Pool Spray 3.6.2.4 3.6 CONTAINMENT SYSTEMS 3.6.2.4 Residual Heat Removal CRHR) Suppression Pool Spray LCO 3.6.2.4 Two RHR suppression pool spray subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR suppression A.l Restore RHR 7 day/I) pool spray subsystem suppression pool inoperable. spray subsystem to OPERABLE status.

B. Two RHR suppression B.l Restore one RHR 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months pool spray subsystems suppression pool inoperable. spray subsystem to OPERABLE status.

c. Required Action and C.l Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Ti me not met.

The 7-day Completion Time for one RHR suppression pool spray subsystem inoperable may be extended to 1O days three (3) times and 14 days one (1) time (A-C subsystem only) until December 31, 2020 with compensatory measures identified in EGC License Amendment Request letter dated September 28, 2018 established and in effect, to allow for modifications to the HPSW system and repairs to Unit 3 RHR Heat Exchanger 3CE024.

PBAPS UN IT 3 3.6-29 Amendment No. 265

RHR Drywell Spray 3.6.2.5 3.6 CONTAINMENT SYSTEMS 3.6.2.5 Residual Heat Removal CRHR) Drywe11 Spray LCD 3.6.2.5 Two RHR drywell spray subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR drywell spray A.l Restore RHR drywell 7 days8 subsystem inoperable. spray subsystem to OPERABLE status.

B. Two RHR drywel 1 spray B.l Restore one RHR 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months subsystems inoperable. drywell spray subsystem to OPERABLE status.

c. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met. AND C.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

The 7-day Completion Time for one RHR drywell spray subsystem inoperable may be extended to 10 days three (3) times and 14 days one (1) time (A-C subsystem only) until December 31, 2020 with compensatory measures identified in EGC License Amendment Request letter dated September 28, 2018 established and in effect, to allow for modifications to the HPSW system and repairs to Unit 3 RHR Heat Exchanger 3CE024.

PBAPS UN IT 3 3.6-30a Amendment No. 291

HPSW System

3. 7 .1 3.7 PLANT SYSTEMS 3.7.1 High Pressure Service Water (HPSW) System LCO 3.7.1 Two HPSW subsystems and the HPSW cross tie shall be OPERABLE.

APPLICABILITY: MODES l, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One HPSW subsystem - - - - - - - - - - - NOTE - - - - - - - - - - - - -

inoperable. Enter applicable Conditions and Required Actions of LCO 3.4.7, "Residual Heat Removal <RHR) Shutdown Cooling System-Hot Shutdown," for RHR shutdown cooling made inoperable by HPSW System.

A.1 Restore HPSW 7 daysG subsystem to OPERABLE status.

B. HPSW cross tie B.1 Restore HPSW cross 7 daysffi inoperable. tie to OPERABLE status C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A or B not met.

continued)

The 7-day Completion Time for one HPSW subsystem inoperable may be extended to 10 days three (3) times and 14 days one (1) time (A-C subsystem only) until December 31, 2020; and the 7-day Completion Time for the HPSW cross tie inoperable may be extended to 1O days one (1) time and 14 days one (1) time (A-C subsystem only) until December 31, 2020 with compensatory measures identified in EGC License Amendment Request letter dated September 28, 2018 established and in effect, to allow for modifications to the HPSW system and repairs to Unit 3 RHR Heat Exchanger 3CE024.

PBAPS UNIT 3 3.7-1 Amendment No. 296

Attachment 7 Peach Bottom Atomic Power Station Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 Markup of Proposed Technical Specifications Bases Pages (For Information Only)

Unit 2 TS Bases Pages B 3.6-57 B 3.6-58 B 3.6-62 B 3.6-63e B 3.7-4 B 3.7-5 Unit 3 TS Bases Pages B 3.6-57 B 3.6-58 B 3.6-62 B 3.6-63e B 3.7-4 B 3.7-5

....--~~~~~~~~~~~~~---.

RHR Suppression Pool Cooling This page provided for information. B 3. 6. 2. 3 No change to this page.

BASES (continued)

APPLICABLE Reference 1 contains the results of analyses used to predict SAFETY ANALYSES primary containment pressure and temperature following large and small break LOCAs. The intent of the analyses is to demonstrate that the heat removal capacity of the RHR Suppression Pool Cooling System is adequate to maintain the primary containment conditions within design limits. The suppression pool temperature is calculated to remain below the design limit.

The RHR Suppression Pool Cooling System satisfies Criterion 3 of the NRC Policy Statement.

LCD During a DBA, a minimum of one RHR suppression pool cooling subsystem is required to maintain the primary containment peak pressure and temperature below design limits (Ref. 1).

To ensure that these requirements are met, two RHR suppression pool cooling subsystems must be OPERABLE with power from two safety related independent power supplies.

Therefore, in the event of an accident, at least one subsystem is OPERABLE assuming the worst case single active failure. An RHR suppression pool cooling subsystem is OPERABLE when one of the pumps, two heat exchangers in the same RHR subsystem, the associated RHR heat exchanger cross tie line, two HPSW System pumps capable of providing cooling to the two heat exchangers and associated piping, valves, instrumentation, and controls are OPERABLE.

Management of gas voids is important to RHR Suppression Pool Cooling System OPERABILITY.

APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment and cause a heatup and pressurization of primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, the RHR Suppression Pool Cooling System is not required to be OPERABLE in MODE 4 or 5.

ACTIONS With one RHR suppression pool cooling subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this Condition, the remaining RHR suppression pool cooling subsystem is adequate to perform the primary containment cooling function. However, the con inued PBAPS UN IT 2 B 3.6-57 Revision No. 126

RHR Suppression Pool Cooling B 3.6.2.3 BASES ACTIONS A.l (continued)

The Completion Time is d'fi db t (*)f overall reliability is reduced because a single failure in mo i ie ya noe ora the OPERABLE subsystem could result in reduced primary one-timechangethatextends containment cooling capability. The 7 day Completion Time the 7-day Completion Time to is acceptable in light of the redundant RHR suppression pool 10 days four (4) times until cooling capabilities afforded by the OPERABLE subsystem and December 31, 2020 to allow for modifications to the HPSW System. The compensatory LL the low probability of a OBA occurring during this period.

.1 measures identified in EGC License Amendment Request If one RHR suppression pool cooling subsystem is inoperable and letter dated September 28, is not restored to OPERABLE status within the required Completion 2018 must be established and Time, the plant must be brought to a condition in which the overall pl ant risk is minimized. To achi eve this status, the in effect. This change also plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

affects TS 3.6.2.4, 3.6.2.5, Remaining in the Applicability of the LCO is acceptable because and 3.7.1. the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 2) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state. The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

Ll With two RHR suppression pool cooling subsystems inoperable, one subsystem must be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . In this condition, there is a substantial loss of the primary containment pressure and temperature mitigation function. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is based on this loss of function and is considered acceptable due to the low probability of a OBA and because alternative methods to remove heat from primary containment are available.

0.1 and D.2 If the Required Action and associated Completion Time of Condition C cannot be met, the plant must be brought to a MODE in which the LCD does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

(continued)

PBAPS UNIT 2 B 3.6-58 Revision No. 66

RHR Suppression Pool Spray B 3.6.2.4 BASES (continued)

ACTIONS A.......l With one RHR suppression pool spray subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this Condition, the remaining OPERABLE RHR suppression pool spray subsystem is adequate to perform the primary containment bypass leakage mitigation function.

However, the overall reliability is reduced because a single The Completion Time is failure in the OPERABLE subsystem could result in reduced modified by a note(*) for a primary containment bypass mitigation capability. The 7 day one-time change that extends Completion Time was chosen in light of the redundant RHR the 7-day Completion Time to suppression pool spray capabilities afforded by the OPERABLE 10 days four (4) times until subsystem and the low probability of a OBA occurring during December 31, 2020 to allow thi period.

for modifications to the HPSW System. The compensatory measures identified in EGC License Amendment Request With both RHR suppression pool spray subsystems inoperable, letter dated September 28, at least one subsystem must be restored to OPERABLE status 2018 must be established and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . In this Condition, there is a substantial in effect. This change also loss of the primary containment bypass leakage mitigation affects TS 3.6.2.3, 3.6.2.5, function. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is based on this loss and 3.7.1. of function and is considered acceptable due to the low probability of a OBA and because alternative methods to remove heat from primary containment are available.

If the inoperable RHR suppression pool spray subsystem(s) cannot be restored to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the overall plant risk is minimized. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 2) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state. The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

(continued)

PBAPS UN IT 2 B 3.6-62 Revision No. 66

RHR Drywell Spray B 3.6.2.5 BASES (continued)

ACTIONS With one RHR drywell spray subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this Condition, the remaining OPERABLE RHR drywell spray subsystem is adequate to mitigate the effects of a steam line break in the drywell. However, the overall reliability is reduced because a single The Completion Time is failure in the OPERABLE subsystem could result in reduced modified by a note (*) for a ability to mitigate the temperature rise associated with a one-time change that extends steam line break in the drywell, for which drywell sprays the 7-day Completion Time to are credited. The 7 day Completion Time was chosen in light of the redundant RHR drywell spray capabilities 10 days four (4) times until afforded by the OPERABLE subsystem and the low probability December 31, 2020 to allow of a steam line break in the drywell occurring during this for modifications to the HPSW period.

System. The compensatory measures identified in EGC License Amendment Request letter dated September 28, 2018 must be established and With both RHR drywell spray subsystems inoperable, at least in effect. This change also one subsystem must be restored to OPERABLE status within affects TS 3.6.2.3, 3.6.2.4, 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . In this Condition, there is a substantial loss of the ability to mitigate the temperature rise associated with and 3.7.1.

a steam line break in the drywell, for which drywell sprays are credited. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is based on this loss of function and is considered acceptable due to the low probability of a steam line break in the drywell and because alternative methods to remove heat from primary containment are available.

C.1 and C.2 If the inoperable RHR drywell spray subsystem(s) cannot be restored to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCD does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

(continued)

PBAPS UN IT 2 B 3.6-63e Revision No. 126 I

HPSW System B 3.7.1 BASES (continued)

APPLICABILITY In MODES 1, 2, and 3, the HPSW System is required to be OPERABLE to support the OPERABILITY of the RHR System for primary containment cooling CLCO 3.6.2.3, "Residual Heat Removal (RHR)

Suppression Pool Cooling," and LCO 3.6.2.4, "Residual Heat Removal CRHR) Suppression Pool Spray") and decay heat removal (LCD 3.4.7, "Residual Heat Removal CRHR) Shutdown Cooling System

-Hot Shutdown"). The Applicability is therefore consistent with the requirements of these systems.

In MODES 4 and 5, the OPERABILITY requirements of the HPSW System are determined by the systems it supports, and therefore, the requirements are not the same for all facets of operation in MODES 4 and 5. Thus, the LCOs of the RHR shutdown cooling system, which requires portions of the HPSW System to be OPERABLE, will govern HPSW System operation in MODES 4 and 5.

ACTIONS With one HPSW subsystem inoperable, the inoperable HPSW subsystem must be restored to OPERABLE status within 7 days.

With the unit in this condition, the remaining OPERABLE HPSW subsystem is adequate to perform the HPSW heat removal function. However, the overall reliability is reduced because a single failure in the OPERABLE HPSW subsystem could result in loss of HPSW function. The Completion Time is based on the redundant HPSW capabilities afforded by the OPERABLE subsystem and the low probability of an event occurring requiring HPSW The Completion Time is during this period.

modified by a note (*) for a one-time change that extends the 7- The Required Action is modified by a Note indicating that the day Completion Time to 1O applicable Conditions of LCD 3.4.7, be entered and Required days four (4) times until Actions taken if an inoperable HPSW subsystem results in an December 31, 2020 to allow for inoperable RHR shutdown cooling subsystem. This is an modifications to the HPSW exception to LCO 3.0.6 and ensures the proper actions are taken System. The compensatory for these components.

.....

measures identified in EGC License Amendment Request letter dated September 28, 2108 must be established and With an inoperable cross tie line, the HPSW cross tie line must in effect. This change also be restored to an OPERABLE status within 7 days. With an affects TS 3.6.2.3, 3.6.2.4, and inoperable HPSW cross tie line, if no additional failures occur, 3.6.2.5. and two HPSW subsystems are OPERABLE, then the two OPERABLE pumps and flow paths ensure two HPSW pumps are available to (continued)

PBAPS UNIT 2 B 3.7-4 Revision No. 114

HPSW System B 3.7.1 BASES ACTIONS .lL.1 (continued)

The Completion Time is provide adequate heat removal capacity following a design basis modified by a note(*) for a one- accident. However, the over al 1 rel i abi 1i ty is reduced because timechangethatextendsthe7- a single failure in the HPSW system could result in a loss of dayCompletionTimeto10 HPSW System function. Therefore, continued operation is daystwo(2)timesuntil permitted only for a limited time. The Completion Time is December 31 , 202 atoallowfor based on remaining heat removal capacity, and the low modifications to the HPSW probabi 1i ty of a OBA occurring during this period.

System. The compensatory measures identified in EGC Ll License Amendment Request letter dated September 28, If one HPSW subsystem or the HPSW cross tie is inoperable and 2018 must be established and not restored within the provided Completion Time, the plant must in effect. This change also be brought to a condition in which the overall plant risk is affects TS 3.6.2.3, 3.6.2.4, and minimized. To achieve this status, the plant must be brought to 3.6.2.5. at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is

..__ _ _ _ _ _ _ _ _ ___,similar to or lower than the risk in MODE 4 (Ref. 5) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state. The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

With both HPSW subsystems inoperable, the HPSW System is not capable of performing its intended function. At least one subsystem must be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time for restoring one HPSW subsystem to OPERABLE status, is based on the Completion Times provided for the RHR suppression pool cooling and spray functions.

The Required Action is modified by a Note indicating that the applicable Conditions of LCO 3.4.7, be entered and Required Actions taken if an inoperable HPSW subsystem results in an inoperable RHR shutdown cooling subsystem. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components.

(continued)

PBAPS UNIT 2 B 3.7-5 Revision No. 114

RHR Suppression Pool Cooling B 3.6 . 2.3 This page provided for information.

BASES (continued) No change to this page.

APPLICABLE Reference 1 contains the results of analyses used to predict SAFETY ANALYSES primary containment pressure and temperature following large and small break LOCAs. The intent of the analyses is to demonstrate that the heat removal capacity of the RHR Suppression Pool Cooling System is adequate to maintain the primary containment conditions within design limits. The suppression pool temperature is calculated to remain below the design limit.

The RHR Suppression Pool Cooling System satisfies Criterion 3 of the NRC Policy Statement.

LCD During a OBA, a m1n1mum of one RHR suppression pool cooling subsystem is required to maintain the primary containment peak pressure and temperature below design limits (Ref. 1).

To ensure that these requirements are met, two RHR suppression pool cooling subsystems must be OPERABLE with power from two safety related independent power supplies.

Therefore, in the event of an accident, at least one subsystem is OPERABLE assuming the worst case single active failure. An RHR suppression pool cooling subsystem is OPERABLE when one of the pumps, two heat exchangers in the same RHR subsystem, the associated RHR heat exchanger cross tie line, two HPSW System pumps capable of providing cooling to the two heat exchangers and associated piping, valves, instrumentation, and controls are OPERABLE.

Management of gas voids is important to RHR Suppression Pool Cooling System OPERABILITY.

APPLICABILITY In MODES 1, 2, and 3, a OBA could cause a release of radioactive material to primary containment and cause a heatup and pressurization of primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, the RHR Suppression Pool Cooling System is not required to be OPERABLE in MODE 4 or 5.

ACTIONS With one RHR suppression pool cooling subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this Condition, the remaining RHR suppression pool cooling subsystem is adequate to perform the primary containment cooling function. However, the con inued PBAPS UN IT 3 B 3.6-57 Revision No. 128

RHR Suppression Pool Cooling B 3.6.2.3 BASES The Completion Time is modified by a note(*) for a one-time Ll (continued) changethatextendsthe7-day overall reliability is reduced because a single failure in Completion Time to 10 days three the OPERABLE subsystem could result in reduced primary (3)timesandto14daysone(1) containment cooling capability. The 7 day Completion Time time (A-C subsystem only) until is acceptable in 1 i ght of the redundant RHR suppression pool cooling capabilities afforded by the OPERABLE subsystem and Dec~':1b~r 31, 2020 to allow for the 1ow probability of a OBA occurring during this period.

mod1f1cat1ons to the HPSW System and repairs to Unit 3 ;\

RHR Heat Exchanger 3CE024. 6..... ..

The compensatory measures -

identified in EGC License If one RHR suppression pool cooling subsystem is inoperable and is not restored to OPERABLE status within the required AmendmentRequestletterdated Completion Time, the plant must be brought to a condition in September28,2018mustbe which the overall plant risk is minimized. To achieve this established and in effect. This status, the pl ant must be brought to at 1east MOOE 3 within 12 changealsoaffectsTS3.6.2.4, hours. Remaining in the Applicability of the LCD is acceptable because the plant risk in MOOE 3 is similar to or lower than the 3.6.2.5, and 3.7.1. risk in MODE 4 (Ref. 2) and because the time spent in MODE 3 to

...___ _ _ _ _ _ _ _ _ _ __.perform the necessary repairs to restore the system to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state. The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

Ll With two RHR suppression pool cooling subsystems inoperable, one subsystem must be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . In this condition, there is a substantial loss of the primary containment pressure and temperature mitigation function. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is based on this loss of function and is considered acceptable due to the low probability of a OBA and because alternative methods to remove heat from primary containment are available.

D.l and D.2 If the Required Action and associated Completion Time of Condition C cannot be met, the plant must be brought to a MODE in which the LCD does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MOOE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

(continued)

PBAPS UN IT 3 B 3.6-58 Revision No. 67

RHR Suppression Pool Spray B 3.6.2.4 BASES (continued)

ACTIONS A....l With one RHR suppression pool spray subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this Condition, the remaining OPERABLE RHR suppression pool spray subsystem is adequate to perform the primary containment bypass leakage mitigation function.

However, the overall reliability is reduced because a single The Completion Time is failure in the OPERABLE subsystem could result in reduced modified by a note (*) for a one- primary containment bypass mitigation capability. The 7 day time change that extends the 7- Completion Time was chosen in light of the redundant RHR day Completion Time to 1O days suppression pool spray capabilities afforded by the OPERABLE three (3) times and 14 days one subsystem and the low probability of a DBA occurring during (1) time (A-C subsystem only) this perio until December 31, 2020 to allow for modifications to the HPSW System and repairs to Unit 3 RHR Heat Exchanger ith both RHR suppression pool spray subsystems inoperable, 3CE024. The compensatory at least one subsystem must be restored to OPERABLE status measures identified in EGC within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . In this Condition, there is a substantial License Amendment Request loss of the primary containment bypass leakage mitigation letter dated September 28, 2018 function. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is based on this loss must be established and in of function and is considered acceptable due to the low effect. This change also affects probability of a DBA and because alternative methods to remove heat from primary containment are available.

TS 3.6.2.3, 3.6.2.5, and 3.7.1.

Ll If the inoperable RHR suppression pool spray subsystem(s) cannot be restored to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the overall plant risk is minimized. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 2) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state. The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

(continued)

PBAPS UN IT 3 B 3.6-62 Revision No. 67

RHR Drywell Spray B 3.6.2.5 BASES (continued)

ACTIONS With one RHR drywell spray subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this Condition, the remaining OPERABLE RHR drywell spray subsystem is adequate to mitigate the effects of a steam line break in the drywell. However,

=-~,,...-~~~---~~~~~~the overall reliability is reduced because a single The Completion Time is modified failure in the OPERABLE subsystem could result in reduced by a note(*) for a one-time abi 1 i ty to mi ti gate the temperature rise associated with a change that extends the 7-day steam 1 i ne break in the drywel 1, for which drywel 1 sprays CompletionTimeto10days are credited. The 7 day Completion Time was chosen in three (3) times and 14 days one 1 i ght of the redundant RHR drywel 1 spray ca pa bi 1 it i es (1) time (A-C subsystem only) afforded by the OPERABLE subsystem and the 1 ow probabi 1 i ty untilDecember 31 , 202 atoallow of a steam line break in the drywell occurring during this period.

for modifications to the HPSW . . . ._

System and repairs to Unit 3 "'

RHR Heat Exchanger 3CE024.

The compensatory measures identified in EGC License With both RHR drywel 1 spray subsystems inoperable, at 1east Amendment Request letter dated one subsystem must be restored to OPERABLE status within September 28, 2018 must be 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . In this Condition, there is a substantial 1oss of established and in effect. This the abi 1 i ty to mi ti gate the temperature rise associated with change also affects TS 3.6 .2 _3 , a steam 1 i ne break in the drywel 1, for which drywel 1 sprays are credited. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is based on this

~3_.6_._2_.4_,_a_n_d_3_.7_._1 _*~~~~~_. loss of function and is considered acceptable due to the low probability of a steam line break in the drywell and because alternative methods to remove heat from primary containment are available.

C.l and C.2 If the inoperable RHR drywell spray subsystem(s) cannot be restored to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

(continued)

PBAPS UN IT 3 B 3.6-63e Revision No. 128 I

HPSW System B 3.7.1 BASES (continued)

APPLICABILITY In MODES 1, 2, and 3, the HPSW System is required to be OPERABLE to support the OPERABILITY of the RHR System for primary containment cooling CLCO 3.6.2.3, "Residual Heat Removal CRHR) Suppression Pool Cooling," and LCO 3.6.2.4, "Residual Heat Removal ( RHR) Suppression Pool Spray") and decay heat removal (LCO 3.4.7, "Residual Heat Removal CRHR)

Shutdown Cooling System-Hot Shutdown"). The Applicability is therefore consistent with the requirements of these systems.

In MODES 4 and 5, the OPERABILITY requirements of the HPSW System are determined by the systems it supports, and therefore, the requirements are not the same for all facets of operation in MODES 4 and 5. Thus, the LCOs of the RHR shutdown cooling system, which requires portions of the HPSW System to be OPERABLE, will govern HPSW System operation in MODES 4 and 5.

ACTIONS With one HPSW subsystem inoperable, the inoperable HPSW subsystem must be restored to OPERABLE status within 7 days.

With the unit in this condition, the remaining OPERABLE HPSW subsystem is adequate to perform the HPSW heat removal function. However, the overall reliability is reduced because a single failure in the OPERABLE HPSW subsystem could result in loss of HPSW function. The Completion Time The Completion Time is modified is based on the redundant HPSW capabilities afforded by the OPERABLE subsystem and the low probability of an event by a note (*) for a one-time occurring requiring HPSW during this period.

change that extends the 7-day Completion Time to 1O days three The Required Action is modified by a Note indicating that (3) times and 14 days one ( 1) time the applicable Conditions of LCO 3.4.7, be entered and (A-C subsystem only) until Required Actions taken if an inoperable HPSW subsystem December 31, 2020 to allow for results in an inoperable RHR shutdown cooling subsystem.

modifications to the HPSW This is an exception to LCO 3.0.6 and ensures the proper System and repairs to Unit 3 RHR actions are taken for these components.

HeatExchanger3CE024. The compensatory measures identified in EGC License Amendment B.1 Request letter dated September With an inoperable cross tie line, the HPSW cross tie line must 28, 2018 must be established and be restored to an OPERABLE status within 7 days. With an in effect. This change also affects inoperable HPSW cross tie line, if no additional failures occur, TS 3.6.2.3, 3.6.2.4 and 3.6.2.5. and two HPSW subsystems are OPERABLE, then the two OPERABLE pumps and flow paths ensure two HPSW pumps are available to (continued)

PBAPS UN IT 3 B 3.7-4 Revision No. 119

HPSW System B 3.7.1 BASES ACTIONS .lL..l (continued) provide adequate heat removal capacity following a design basis The Completion Time is modified accident. However, the overall reliability is reduced because a by a note (*) for a one-time change single failure in the HPSW System could result in a loss of HPSW that extends the 7-day Completion System function. Therefore, continued operation is permitted Time to 10 days one (1) time and only for a limited time. The Completion Time is based on 14 days one (1) time (A-C remaining heat removal capacity, and the low probability of a subsystem only) until December OBA occurring during this period.

31, 2020 to allow for modifications to the HPSW System and repairs

- ~ ,

to Unit 3 RHR Heat Exchanger Ll 3CE024. The compensatory If one HPSW subsystem or the HPSW cross tie is inoperable and measures identified in EGC not restored within the provided Completion Time, the plant must License Amendment Request be brought to a condition in which the overall plant risk is letter dated September 28, 2018 minimized. To achieve this status, the plant must be brought to must be established and in effect. at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Remaining in the Applicability This change also affects TS of the LCO is acceptable because the plant risk in MODE 3 is 3.6.2.3, 3.6.2.4 and 3.6.2.5. similar to or lower than the risk in MODE 4 (Ref . 5) and because the time spent in MODE 3 to perform the necessary repairs to

...____ _ _ _ _ _ _ _ _ _ ___. restore the system to OPERABLE status wi 11 be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state . The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

With both HPSW subsystems inoperable, the HPSW System is not capable of performing its intended function. At least one subsystem must be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time for restoring one HPSW subsystem to OPERABLE status, is based on the Completion Times provided for the RHR suppression pool cooling and spray functions.

The Required Action is modified by a Note indicating that the applicable Conditions of LCO 3.4.7, be entered and Required Actions taken if an inoperable HPSW subsystem results in an inoperable RHR shutdown cooling subsystem.

This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components.

contin PBAPS UN IT 3 B 3.7-5 Revision No . 119

Attachment 8 Peach Bottom Atomic Power Station Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 Summary of Regulatory Commitments

License Amendment Request Attachment 8 Reduce HPSW System Design Pressure and Revise Page 1of1 Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, and 3.7.1 for Temporary Extension of Completion Times Summary of Regulatory Commitments DPR-44 and DPR-56

SUMMARY

OF REGULATORY COMMITMENTS The following table identifies commitments made in this document. (Any other actions discussed in the submittal represent intended or planned actions. They are described to the NRC for the NRC's information and are not regulatory commitments.)

COMMITMENT TYPE COMMITTED COMMITMENT ONE-TIME DATE OR PROGRAMMATIC ACTION (YES/NO)

"OUTAGE" (YES/NO)

The compensatory measures Prior to Yes* No identified in Section 3.3 of commencing the Attachment 5 of this License applicable Amendment Request will be HPSW implemented during the subsystem extended allowed outage times modifications.

associated with the HPSW subsystem modifications for Units 2 and 3.

This is a one-time change during the performance of the HPSW subsystem modifications for PBAPS, Units 2 and 3.

v t e Letter Sequence Request
EPID:L-2018-LLA-0265, License Amendment Request to Reduce High Pressure Service Water System Design Pressure and Revise Technical Specifications 3.6.2.3, 3.6.2.4, 3.6.2.5, and 3.7.1 for Temporary Extension of ... (Approved, Closed)
Initiation Request, Request Acceptance Administration Questions 16 January 2019 27 February 2019 20 August 2019 Answers 15 February 2019 29 August 2019 Results Approval Other:
MONTHYEAR2019-02-27 [Table View]

ML18275A023

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3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

Subject:

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

SUMMARY

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