Text

License Amendment Request for Extension of the Essential Services Chilled Water System Allowed Outage Time and Removal of an Expired Note from Technical Specifications
	ML19049A027
Person / Time
Site:	Harris
Issue date:	02/18/2019
From:	Hamilton T; Duke Energy Progress
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RA-19-0007
	Download: ML19049A027 (180)
	v • d • e

Tanya M. Hamilton Vice President Harris Nuclear Plant 5413 Shearon Harris Rd New Hill, NC 27562-9300 10 CFR 50.90 February 18, 2019 Serial: RA-19-0007 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400 Renewed License No. NPF-63

Subject:

License Amendment Request for Extension of the Essential Services Chilled Water System Allowed Outage Time and Removal of an Expired Note from Technical Specifications Technical Specifications Sections:

3.1.2.4, Charging Pumps - Operating 3.5.2, ECCS Subsystems - Tavg Greater Than or Equal To 350°F 3.6.2.1, Containment Spray System 3.6.2.2, Spray Additive System 3.6.2.3, Containment Cooling System" 3.7.1.2, Auxiliary Feedwater System 3.7.3, Component Cooling Water System 3.7.4, "Emergency Service Water System" 3.7.6, Control Room Emergency Filtration System 3.7.7, Reactor Auxiliary Building (RAB) Emergency Exhaust System 3.7.13, Essential Services Chilled Water System 3.8.1.1, AC Sources - Operating Ladies and Gentlemen:

In accordance with the provisions of 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy),

hereby requests a revision to the Technical Specifications (TS) for the Shearon Harris Nuclear Power Plant, Unit 1 (HNP). The proposed license amendment revises TS 3.1.2.4, TS 3.5.2, TS 3.6.2.1, TS 3.6.2.3, TS 3.7.4, and TS 3.7.13 to permit one train of the Essential Services Chilled Water System (ESCWS) to be inoperable for up to 7 days. This change will allow for extended maintenance activities on the ESCWS and air handlers supported by the ESCWS for equipment reliability. In addition, this proposed amendment removes an expired note previously added to numerous TS sections by implementation of License Amendment 153. By letter dated September 16, 2016, the NRC issued License Amendment 153 and a safety evaluation report for temporary TS changes to support the replacement of the A Emergency Service Water (ESW) pump (Agencywide Documents Access and Management System Accession No. ML16253A059). Implementation of the A ESW pump replacement was completed on

U.S. Nuclear Regulatory Commission Page 2 of 3 Serial RA-19-0007 September 29, 2016. The note was added to allow temporary changes to TS 3.1.2.4, TS 3.5.2, TS 3.6.2.1, TS 3.6.2.2, TS 3.6.2.3, TS 3.7.1.2, TS 3.7.3, TS 3.7.4, TS 3.7.6, TS 3.7.7, TS 3.7.13, and TS 3.8.1.1. This license amendment request (LAR) proposes to remove the temporary change to TS described in the NRC letter dated September 16, 2016.

Attachment 1 of this LAR provides Duke Energy's evaluation of the proposed changes.

Attachment 2 provides a copy of the proposed TS changes. Attachment 3 provides a copy of the proposed TS Bases changes. Compensatory measures that will be taken as described in Attachment 1 for the AOT extension are identified in Attachment 3. Attachment 4 provides the following contents to support information provided in Attachment 1: Attachment 4 of plant programs procedure PLP-114, Relocated Technical Specifications and Design Basis Requirements; a table of areas served by Engineered Safety Feature (ESF) fan coolers in the Reactor Auxiliary Building (RAB); and ESCWS Flow Diagrams. Attachment 5 provides the probabilistic risk assessment (PRA) analysis and calculation detail for the proposed allowed outage time (AOT) extension.

Duke Energy requests NRC review and approval of this LAR within one year of its acceptance date. The amendment shall be implemented within 90 days following approval.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated North Carolina State Official.

This letter does not contain any regulatory commitments.

Should you have any questions regarding this submittal, please contact Art Zaremba, Fleet Licensing Manager, at (980) 373-2062.

I declare under penalty of perjury that the foregoing is true and correct. Executed on F ebruary 18, 2019.

Sincerely, Tanya M. Hamilton Attachments:

1. Evaluation of Proposed Changes

2. Proposed Technical Specifications Changes

3. Proposed Technical Specifications Bases Changes

4. Attachment 4 of PLP-114, Relocated Technical Specifications and Design Basis Requirements; Table of Areas Served by ESF Fan Coolers in the RAB; and ESCWS Flow Diagrams

5. ESCWS Extended AOT LAR PRA Input

U.S. Nuclear Regulatory Commission Page 2 of 3 Serial RA-19-0007 September 29, 2016. The note was added to allow temporary changes to TS 3.1.2.4, TS 3.5.2, TS 3.6.2.1, TS 3.6.2.2, TS 3.6.2.3, TS 3.7.1.2, TS 3.7.3, TS 3.7.4, TS 3.7.6, TS 3.7.7, TS 3.7.13, and TS 3.8.1.1. This license amendment request (LAR) proposes to remove the temporary change to TS described in the NRC letter dated September 16, 2016. of this LAR provides Duke Energy's evaluation of the proposed changes. provides a copy of the proposed TS changes. Attachment 3 provides a copy of the proposed TS Bases changes. Compensatory measures that will be taken as described in for the AOT extension are identified in Attachment 3. Attachment 4 provides the following contents to support information provided in Attachment 1: Attachment 4 of plant programs procedure PLP-114, Relocated Technical Specifications and Design Basis Requirements; a table of areas served by Engineered Safety Feature (ESF) fan coolers in the Reactor Auxiliary Building (RAB); and ESCWS Flow Diagrams. Attachment 5 provides the probabilistic risk assessment (PRA) analysis and calculation detail for the proposed allowed outage time (AOT) extension.

Duke Energy requests NRC review and approval of this LAR within one year of its acceptance date. The amendment shall be implemented within 90 days following approval.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated North Carolina State Official.

This letter does not contain any regulatory commitments.

Should you have any questions regarding this submittal, please contact Art Zaremba, Fleet Licensing Manager, at (980) 373-2062.

I declare under penalty of perjury that the foregoing is true and correct. Executed on February 18, 2019.

Sincerely, Tanya M. Hamilton Attachments:

1. Evaluation of Proposed Changes

2. Proposed Technical Specifications Changes

3. Proposed Technical Specifications Bases Changes

4. Attachment 4 of PLP-114, Relocated Technical Specifications and Design Basis Requirements; Table of Areas Served by ESF Fan Coolers in the RAB; and ESCWS Flow Diagrams

5. ESCWS Extended AOT LAR PRA Input

U.S. Nuclear Regulatory Commission Page 3 of 3 Serial RA-19-0007 cc: J. Zeiler, NRC Senior Resident Inspector, HNP W. L. Cox, III, Section Chief, N.C. DHSR M. Barillas, NRC Project Manager, HNP NRC Regional Administrator, Region II

U.S. Nuclear Regulatory Commission Serial RA-19-0007 SERIAL RA-19-0007 ATTACHMENT 1 EVALUATION OF PROPOSED CHANGES SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63

U.S. Nuclear Regulatory Commission Page 1 of 29 Serial RA-19-0007 Evaluation of Proposed Changes 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 System Design and Operation 2.2 Current Technical Specification Requirements 2.3 Reason for the Proposed Changes 2.4 Description of the Proposed Changes

3.0 TECHNICAL EVALUATION

3.1 Technical Specification Systems Affected by the Proposed Changes 3.2 Room Heatup Analysis 3.3 Additional System for Safe Shutdown 3.4 Risk Analysis for the Proposed Changes 3.5 Defense-in-Depth Considerations 3.6 Assumptions and Compensatory Measures 3.7 Compliance with Current Regulations 3.8 Evaluation of Safety Margins 3.9 Configuration Risk Management 3.10 Conclusions

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedents 4.3 Significant Hazards Consideration 4.4 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

S

6.0 REFERENCES

U.S. Nuclear Regulatory Commission Page 2 of 29 Serial RA-19-0007 1.0

SUMMARY

DESCRIPTION In accordance with the provisions of 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy),

proposes a change to the Shearon Harris Nuclear Power Plant, Unit 1 (HNP), Technical Specifications (TS) to permit one train of the Essential Services Chilled Water System (ESCWS) to be inoperable for up to 7 days to allow for extended maintenance activities on the ESCWS and air handlers supported by the ESCWS for equipment reliability.

The proposed seven-day allowed outage time (AOT) is based on findings of the probabilistic risk assessment (PRA) analysis and calculation provided in Attachment 5. This license amendment request (LAR) concludes that extending the AOT to 7 days provides plant operational flexibility while simultaneously reducing overall plant risk. This reduction is because the risk incurred by having an ESCWS train unavailable for a longer time at power will be substantially offset by the benefits associated with completing maintenance on the ESCWS and the air handlers supported by the ESCWS for equipment reliability.

In addition, this proposed amendment removes an expired note which was previously added to numerous TS sections by implementation of License Amendment 153 at HNP. The note was added to allow temporary changes to TS 3.1.2.4, Charging Pumps - Operating, TS 3.5.2, ECCS [Emergency Core Cooling Systems] Subsystems - Tavg Greater Than or Equal To 350°F, TS 3.6.2.1, Containment Spray System, TS 3.6.2.2, Spray Additive System, TS 3.6.2.3, Containment Cooling System, TS 3.7.1.2, Auxiliary Feedwater System, TS 3.7.3, Component Cooling Water System, TS 3.7.4, Emergency Service Water System, TS 3.7.6, Control Room Emergency Filtration System, TS 3.7.7, Reactor Auxiliary Building (RAB)

Emergency Exhaust System, TS 3.7.13, Essential Services Chilled Water System, and TS 3.8.1.1, AC Sources - Operating. By letter dated September 16, 2016, the NRC issued License Amendment 153 for HNP and a safety evaluation report for the temporary TS changes to support the replacement of the A Emergency Service Water (ESW) pump (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16253A059).

Implementation of the A ESW pump replacement was completed on September 29, 2016. This LAR proposes to remove the temporary changes to TS described in the NRC letter dated September 16, 2016.

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation Essential Services Chilled Water System From Section 9.2.8 of the HNP Final Safety Analysis Report (FSAR), "Essential Services Chilled Water System," the objective of the ESCWS is to provide chilled water to the cooling coils of air handling units for the following systems:

Control Room Air Conditioning System

RAB Engineered Safety Feature (ESF) Equipment Cooling System

RAB Switchgear Rooms Ventilation System

RAB Electrical Equipment Protection Rooms Ventilation System

RAB Non-Nuclear Safety (NNS)-Ventilation Systems

Fuel Handling Building Spent Fuel Pool Pump Room Ventilation System

U.S. Nuclear Regulatory Commission Page 3 of 29 Serial RA-19-0007 The ESCWS is designed to meet the following design bases:

The system supplies a nominal 44-degree Fahrenheit (F) chilled water to the cooling coils in the air handling units.

The system is designed with sufficient redundancy to meet the single failure criteria.

The system is designed to meet Safety Class 3 and Seismic Category I requirements.

The system is designed to provide accessibility for adjustments and periodic inspections and testing of the principal system components.

The ESCWS consists of two 100 percent capacity subsystems (one is normally operating and one is normally in standby). Each subsystem consists of a package water chiller, a chilled water pump, an expansion tank, a makeup tank, a chemical addition tank, service water recirculation pump, and an independent piping system. Each package water chiller is comprised of a compressor, condenser, flow control device, and an evaporator. The chiller refrigeration cycle chills the chilled water and rejects heat to the Service Water System through the condenser unit.

The condenser unit is supplied with cooling water from the Service Water System during normal and emergency plant operation. There is a condenser recirculating pump that recirculates Service Water to maintain the minimum flow rate through the chiller condenser tubes to minimize biological fouling. The chiller water pump circulates chilled water to the cooling coils of air handlers supported by the ESCWS.

The evaporator is a four-pass shell and tube type heat exchanger (chilled water on tube side) that transfers the heat from the chilled water to the refrigerant. Refrigerant enters the heat exchanger as low temperature, low pressure, and exits as high temperature, low pressure. Heat input from the chilled water converts the refrigerant to a low-pressure, superheated vapor. Low-pressure, superheated refrigerant vapor is directed to the compressor suction through baffle plates. Each compressor is lubricated by oil that is contained in an oil reservoir, pumped to an oil cooler that is cooled by chilled water, then flows into an emergency reservoir that ports oil to the speed changer and bearings of the compressor.

Compressor motor-operated prerotation vanes control chiller load between 10 percent and 100 percent. Pitch of vanes is changed to vary suction flow to the compressor as needed based on the heat load on the ESCWS. As heat load increases, the chilled water temperature increases above the nominal 44 degrees F. Compressor prerotation vanes rotate toward open, which increases the flow of refrigerant vapor through the compressor. Increased vapor flow provides additional cooling capacity, increasing load on the chiller unit. If heat load decreases, prerotation vanes rotate toward shut, reducing load on the chiller unit. The vanes will continue to shut until chiller load is reduced to 10 percent. At that point, the vanes have reached their minimum operational position. If chilled water temperature continues to lower with the prerotation vanes shut to their minimum 10 percent load position, the hot gas bypass valve will start opening. The hot gas bypass valve bypasses high temperature refrigerant from the condenser to the evaporator, in effect creating an additional load. The hot gas bypass valve modulates open to allow control at a chiller load less than 10 percent.

Refrigerant is compressed to high pressure vapor with a corresponding temperature increase.

Refrigerant vapor is discharged from the compressor to the condenser where it becomes a liquid at high pressure, giving up heat to service water. The condenser is a two-pass shell and tube type heat exchanger (service water on tube side). Liquid refrigerant then enters a flow control device (restricting orifice) where pressure is reduced to near the boiling point. The liquid flows into the evaporator where it boils and vaporizes, removing heat from the chilled water.

U.S. Nuclear Regulatory Commission Page 4 of 29 Serial RA-19-0007 The expansion tank accommodates system volume changes, maintains positive pressure in the piping loop and provides a means for adding makeup water to the system. Makeup water to the system is fed from the Demineralized Water System during normal operation. During post-accident conditions, the makeup water is fed from the Service Water System. The water level in the tank is automatically maintained by a level switch. A chemical addition tank located in the system provides the necessary chemicals to prevent corrosion and scale buildup in the system.

Chemical addition is manual when it is required by periodic water analysis tests. The makeup tank originally served as a pressurized expansion tank. The makeup tank was subsequently converted to a water-solid tank in the normal makeup water flow path when atmospheric expansion tanks were installed. The makeup tank no longer has any specific function other than to serve as part of the ESCWS pressure boundary.

The ESCWS is automatically started upon receipt of a Safety Injection Actuation Signal (SIAS).

Non-essential portions of the ESCWS are automatically isolated from the essential portions upon receipt of a SIAS. In the event of a failure in a single train of the ESCWS during an accident, a redundant 100% capacity system would still be available. Upon receipt of a SIAS, the demineralized water supply to the chillers will be isolated using redundant solenoid operated valves arranged in series. The supply will then be provided from the Service Water System. The source of water supply to the condenser section of the ESCWS is from the Service Water System during normal and emergency plant conditions.

In the event of loss of offsite power, all active components such as valve operators, water chiller motors, chilled water pumps, controls and instrumentation will be supplied with power from the emergency diesel generators (EDGs). Each subsystem is powered from a different emergency bus. Upon loss of offsite power, the ESCWS chillers and chilled water pumps are automatically sequenced to reduce starting power requirements from the standby EDG. Each chiller is furnished with a compressor starter, operational and safety controls, interlocks and other controls for local and remote operation.

Each air handler that receives chilled water flow from the ESCWS has a temperature control valve (TCV). When a fan is in service, chilled water flow is directed through its associated cooling coils. When the fan is secured, the TCV repositions to shut off or to bypass the chilled water flow to the cooling coils. Attachment 4 of this LAR includes ESCWS condenser flow diagrams for the A Train and B Train portions of the system and the ESCWS heating, ventilation, and air conditioning (HVAC) distribution flow diagrams for the A Train and B Train portions of the system. Attachment 4 of this LAR also includes Attachment 4 of plant programs procedure PLP-114, Relocated Technical Specification and Design Basis Requirements, which contains the operational temperature limits for various areas of the plant. The area temperature limits are established to ensure that environmentally qualified equipment will not be exposed to temperatures beyond that to which they were originally qualified. The consequences of exceeding the area temperature limits are that extended exposure to elevated temperatures could contribute to equipment degradation and cause the degradation to exceed the rate assumed by the HNP Environmental Qualification (EQ) Program. The temperature limits are applicable whenever equipment in the area is required to be functional. Area temperatures are verified to be within limits every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Control Room Air Conditioning System The Control Room Air Conditioning System (CRACS) provides heating, ventilation, cooling, filtration, air intake and exhaust isolation for the Control Room Envelope (CRE) during normal

U.S. Nuclear Regulatory Commission Page 5 of 29 Serial RA-19-0007 operation and during a design basis accident. Air handling units 'AH-15 1A-SA' and 'AH-15 1B-SB' provide HVAC to the Control Room area. The CRACS is designed to maintain the Control Room at a design temperature of 75 degrees F dry bulb and maximum relative humidity not to exceed 50 percent, assuring personal comfort as well as a suitable environment for continuous operation of controls and instrumentation. The air in the Control Room is cooled by a cooling coil that receives chilled water from the ESCWS. When heating is required, the air is heated by an electric heating coil to maintain the design space temperature identified above.The CRACS is designed to detect the introduction of radioactive material into the Control Room and automatically isolate all air intakes and exhausts upon a high radiation signal or SIAS and to remove airborne radioactivity from the Control Room to the extent that dose to the Control Room operator following a design basis accident does not exceed the limit specified in General Design Criteria (GDC) 19. In addition, the RAB Normal Ventilation System will be secured, and the RAB Emergency Exhaust System (RABEES) will be started. The RAB Normal Ventilation System must be secured to preclude the possibility of postulated system failures from impacting the ability of the CRE to maintain a positive pressure of greater than 1/8-inch water gauge relative to adjacent areas. When the RAB Normal Ventilation System is secured, the RABEES is initiated to maintain the potentially contaminated areas of the RAB at sub-atmospheric pressure to limit outleakage and to remove radon gas from the RAB.

During normal operation, the CRACS operates in a recirculation mode with the Control Room Emergency Filtration System (CREFS) de-energized. The outside makeup air mixes with the returned air before it is conditioned by the air handling units. The Control Room is maintained at a slightly positive pressure with respect to the adjacent area so that the air from other sources entering the Control Room is minimized. The pressurization of the Control Room is maintained automatically by modulating exhaust fan dampers.

The CREFS provides a protected environment from which occupants can control the plant following an uncontrolled release of radioactivity, hazardous chemicals, or smoke. The system consists of two independent, redundant trains that recirculate and filter the air in the CRE and a CRE boundary that limits the inleakage of unfiltered air. The CREFS is an emergency system, parts of which may also operate during normal unit operation in the standby mode of operation.

Actuation of the CREFS places the system in the emergency mode (i.e., isolation with recirculation mode) of operation. Actuation of the system closes the unfiltered outside air intake and unfiltered exhaust dampers, and aligns the system for recirculation of the air within the CRE through redundant trains of high efficiency particulate air (HEPA) and charcoal filters. The emergency mode also allows for pressurization and filtered ventilation of the air supply to the CRE.

RAB ESF Equipment Cooling System The RAB ESF Equipment Cooling System is designed to serve areas in the RAB that contain equipment essential for safe shutdown and to maintain space temperatures at or below an electrical equipment design temperature of 104 degrees F for all areas containing essential equipment which operate during safe shutdown. The ESF fan coolers start on a temperature rise due to start-up of the essential equipment contained in the corresponding areas. The RAB ESF Equipment Cooling System consists of cooling systems for various ESF equipment areas and the Steam Tunnel Ventilation System. The cooling systems consist of factory-fabricated air handling units. Each unit consists of a fan section, a cooling coil section and a filter section.

Chilled water for the cooling coils is supplied from the ESCWS. Air is drawn from each fan

U.S. Nuclear Regulatory Commission Page 6 of 29 Serial RA-19-0007 coolers surrounding area and discharged to the space it serves. Upon receipt of an SIAS or for an undervoltage condition upon a loss of offsite power, the fan coolers in this system are sequenced on-line by sequencer panels. The Steam Tunnel Ventilation System is not supported by the ESCWS.

There are numerous equipment areas at HNP that may be supplied by either 100% redundant A or B Train ventilation units. Generic Letter 80-30 (Reference 1) describes the philosophy of relaxing single failure criteria while in a TS limiting condition for operation. Generic Letter 80-30 states, By and large, the single failure criterion is preserved by specifying Limiting Conditions for Operation (LCO) that require all redundant components of safety related systems to be OPERABLE. When the required redundancy is not maintained, either due to equipment failure or maintenance outage, action is required, within a specified time, to change the operating mode of the plant to place it in a safe condition. The specified time to take action, usually called the out-of-service time, is a temporary relaxation of the single failure criterion, which, consistent with overall system reliability considerations, provides a limited time to fix equipment or otherwise make it operable. Based on this guidance provided by Generic Letter 80-30, areas that can be supplied by either 100% redundant A or B Train ventilation units do not have to meet single failure criterion provided a TS LCO action condition exists (for example, the ESCWS TS). Equipment located in common areas can be considered operable if only one train of the ESCWS is available provided the TS LCO action is entered for the inoperable train of the ESCWS. Attachment 4 includes a table that identifies the various areas in the RAB that are served by ESF fan coolers and identifies each air handling unit and its location, identifies whether redundant train cooling capability exists and, if not, the affected systems TS LCO entry as necessary. This table also identifies the safety related credited start signal(s) (SIAS or undervoltage condition upon a loss of offsite power) for each air handling unit.

RAB Switchgear Rooms Ventilation System The RAB Switchgear Rooms Ventilation System (RABSRVS) is designed to maintain a controlled environment in all served areas to assure suitable operating conditions for plant personnel and continuous operation of vital systems and equipment for indoor space temperatures that are identified in the FSAR Table 9.4.0-1, to maintain airflow from areas of low potential contamination to areas of progressively higher potential contamination, and to exhaust sufficient air from the battery rooms to prevent the accumulation of combustible concentrations of hydrogen. The areas served by the RABSRVS include Battery Rooms, heating and ventilation (H&V) Equipment Rooms, Rod Control Cabinets Room, the Auxiliary Control Panel Room, and Process Instrumentation Control (PIC) Rooms in the RAB on the 286 elevation.

There are two switchgear rooms in the RAB with independent air conditioning systems.

Switchgear Room A air conditioning system is connected to safety channel A, and Switchgear Room B air conditioning system is connected to safety channel B. Air handling units 'AH-12 1A-SA' and 'AH-12 1B-SA' provide HVAC to the Switchgear Room A. Air handling units 'AH-13 1A-SB' and 'AH-13 1B-SB' provide HVAC to the Switchgear Room B. The chilled water for cooling coils of these air handling units is supplied from the ESCWS. The design for Switchgear Room ventilation provides for two parallel fans powered by an associated safety bus with capability of chilled water cooling from either the A or B Train ESCWS. Either cooling coil (supplied by A or B Train ESCWS) may be used to maintain operability of supported components. The AH-12 1A-SA and the AH-13 1A-SB fans (emergency fans) have the capability to automatically start on a SIAS whereas the AH-12 1B-SA and AH-13 1B-SB (normal fans) do not receive an automatic ESF actuation system signal to start.

U.S. Nuclear Regulatory Commission Page 7 of 29 Serial RA-19-0007 Each RABSRVS, consists of a missile protected air intake, medium efficiency filter, electric heating coil, two 100 percent redundant chilled water cooling coils connected in series and two 100 percent redundant centrifugal fans arranged in parallel. Each fan is provided with a motorized inlet damper and a gravity discharge damper to prevent air recirculation through the idle fan. One fan is normally operating and the other fan is on standby. The outside air intake valves are de-energized closed to ensure that the switchgear room will not become pressurized.

This pressurization could adversely impact the required pressurization of the Control Room to all adjacent areas. Air is supplied to the areas served through a sheet metal ductwork distribution system. The Auxiliary Control Panel Room, which is normally ventilated by the Switchgear Room B system, has a provision for a redundant ventilation from the Switchgear Room A system.

RAB Electrical Equipment Protection Room Ventilation System The RAB Electrical Equipment Protection Room Ventilation System (RABEEPRVS) is designed to maintain suitable ambient conditions for personnel comfort and safety. The system maintains area temperatures to assure proper operation of vital systems and equipment. The RABEEPRVS consists of two 100 percent capacity subsystems in parallel (one in operation and one in standby). Each subsystem is powered by its respective safety channel.

Each ventilation supply subsystem consists of a motorized inlet damper, medium efficiency filter, chilled water cooling coil, supply fan, gravity damper and electric heating coil. The conditioned air is supplied to the areas served through a sheet metal ductwork distribution system. Both ventilation supply subsystems are connected to a common missile protected outside air intake with tornado damper and two motorized butterfly isolation valves in series.

The exhaust system consists of two redundant, 100 percent capacity fans. Each fan is provided with a back-draft discharge damper to prevent air recirculation through an idle fan. Air handling units 'AH-16 1A-SA' and 'AH-16 1B-SB' provide HVAC to the PIC Room 305' elevation. These air handling units also cool the Solid State Protection System that includes the Engineered Safety Features Actuation System and the Reactor Protection System. Repair Shop Spaces and the Auxiliary Relay Cabinet Room are also cooled by this system. The cooling coil in each unit is supplied with chilled water from the ESCWS.

RAB Normal Ventilation System The RAB Normal Ventilation System (RABNVS) is designed to provide normal ventilation for areas containing equipment essential for safe shutdown in the RAB, including the Chemical and Volume Control System (CVCS) chiller area, 480-volt auxiliary bus area, areas containing non-essential equipment, surrounding access aisles and RAB stairways and H&V equipment rooms.

The RABNVS is supplied by outdoor air. The RABNVS is designed to maintain space temperatures as indicated in FSAR, Table 9.4.0-1, during normal plant operation.

The RABNVS discharges are monitored to detect and control the release of airborne radioactivity. The normal supply and exhaust systems are designed with sufficient redundancy to ensure continuous reliable performance during normal plant operations. The supply system is provided with two 100 percent capacity redundant operating fans (one operating and one standby); the exhaust system is provided with four 25 percent capacity operating fans. The containment pre-entry purge exhaust unit serves as a standby unit for RAB Normal Exhaust System. The RABNVS maintains air flow from areas of low potential radioactivity to areas of progressively higher potential radioactivity and will isolate service to selected post-accident,

U.S. Nuclear Regulatory Commission Page 8 of 29 Serial RA-19-0007 potentially contaminated areas upon receipt of a SIAS or a Control Room Isolation Signal, to enable the RAB Emergency Exhaust System to maintain these areas below atmospheric pressure.

RAB NNS-Ventilation System The RAB NNS-Ventilation System is designed to provide normal ventilation to maintain a controlled environment suitable for plant personnel and continuous operation of systems and equipment in areas not served by the RABNVS. This system is automatically isolated from the essential portions of the ESCWS upon receipt of a SIAS. This system maintains airflow from areas of low potential radioactivity to areas of progressively higher potential radioactivity. Each RAB NNS-Ventilation Subsystem consists of an outside air intake plenum, medium efficiency filter, electric heating coil, chilled water cooling coil and centrifugal supply and return fans.

Cooling coils in these units receive chilled water from the ESCWS.

Spent Fuel Pump Room Ventilation System The Spent Fuel Pump Room Ventilation System is designed to provide cooling for mechanical equipment for protection of equipment motors and to provide cooling for the Emergency Exhaust System for protection of the fan motors during a fuel handling accident.

The system consists of two redundant, 100 percent capacity air handling units (one operating, and one standby). Each air handling unit includes medium efficiency filters, a chilled water cooling coil and a centrifugal fan. The chilled water to the cooling coil of the air handling unit is provided from the ESCWS. Air handling units 'AH-17 1-4A-SA' and 'AH-17 1-4B-SB' provide HVAC to the Spent Fuel Pump Room located in the Fuel Handling Building. These air handling units supply cooled air to spent fuel pool pumps and heat exchangers area, Motor Control Center (MCC) room, H&V equipment area, and emergency filtration area during normal and emergency conditions. In the event of loss of offsite power, the air handling units of the Spent Fuel Pool Pump Room Ventilation System may be powered by the EDG system. There are no TS requirements associated with this system.

The proposed change to remove an expired note from TS 3.1.2.4, TS 3.5.2, TS 3.6.2.1, TS 3.6.2.2, TS 3.6.2.3, TS 3.7.1.2, TS 3.7.3, TS 3.7.4, TS 3.7.6, TS 3.7.7, TS 3.7.13, and TS 3.8.1.1 is administrative and non-technical in nature. This change does not involve any system design or operational requirements.

2.2 Current Technical Specification Requirements TS 3.7.13, Essential Services Chilled Water System, requires at least two independent ESCWS loops to be operable in modes 1-4. With only one ESCWS operable, at least two loops are to be restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the plant must be in hot standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in cold shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

TS 3.1.2.4, Charging Pumps - Operating, requires at least two charging/safety injection pumps to be operable in modes 1-3. With only one charging/safety injection pump operable, at least two charging/safety injection pumps are to be restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the plant must be in hot standby and borated to a shutdown margin as specified in the core operating limits report at 200°F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; at least two charging/safety injection pumps are to be restored to operable status within the next 7 days or the plant must be in hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

U.S. Nuclear Regulatory Commission Page 9 of 29 Serial RA-19-0007 TS 3.5.2, ECCS Subsystems - Tavg Greater Than or Equal To 350°F, requires two independent ECCS subsystems to be operable in modes 1-3. With one ECCS subsystem inoperable, the inoperable subsystem is to be restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the plant must be in at least hot standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in hot shutdown within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

TS 3.6.2.1, Containment Spray System, requires two independent Containment Spray Systems to be operable in modes 1-4. With one Containment Spray System inoperable, the inoperable Spray System is to be restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the plant must be in at least hot standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and if the inoperable Spray System is not restored within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the plant is to be in cold shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

TS 3.6.2.3, Containment Cooling System," requires that four containment fan coolers to be operable in modes 1-4. With one train of the required containment fan coolers inoperable and one train of the Containment Spray System inoperable, the inoperable Spray System is to be restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the plant must be in at least hot standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in cold shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The inoperable train of containment fan coolers must be restored to operable status within 7 days of initial loss or the plant must be in hot standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in cold shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

TS 3.7.4, "Emergency Service Water System," requires at least two independent emergency service water loops to be operable in modes 1-4. With only one emergency service water loop operable, at least two loops are to be restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the plant must be in hot standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in cold shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The proposed change to remove an expired note from TS 3.1.2.4, TS 3.5.2, TS 3.6.2.1, TS 3.6.2.2, TS 3.6.2.3, TS 3.7.1.2, TS 3.7.3, TS 3.7.4, TS 3.7.6, TS 3.7.7, TS 3.7.13, and TS 3.8.1.1 is administrative and non-technical in nature. The TS sections that contain the expired note were addressed in Duke Energy's October 29, 2015, LAR (ADAMS Accession No. ML15302A542), as well as in the License Amendment 153 Safety Evaluation Report, dated September 16, 2016 (ADAMS Accession No. ML16253A059), for this LAR.

2.3 Reason for the Proposed Changes There are maintenance activities for the ESCWS and air handlers supported by the ESCWS that require the system to be unavailable for greater than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . If it is determined that a chiller compressor issue is present that requires refrigerant replacement, a 7-day maintenance window is needed. Conditions such as compressor oil system leakage that cannot be isolated, excessive refrigerant leakage, hot gas bypass valve excessive leak-by, or evaporator/condenser tube leaks are examples of issues that may require a refrigerant replacement. Other maintenance activities such as opening and cleaning the ESCWS condenser tubes due to a service water fouling event, replacement of the compressor motor due to an electrical fault, or refurbishment of an air handling unit that is supported by the ESCWS, may necessitate a 7-day maintenance window. Maintenance activities that are associated with identifying and correcting actual or potential degraded conditions extend to all supporting functions for the conduct of these activities. From review of maintenance activities completed on the ESCWS over the past

U.S. Nuclear Regulatory Commission Page 10 of 29 Serial RA-19-0007 5 years, there have been a few instances where the maintenance duration on the system (not including the time period of post-maintenance testing) lasted between 41-57 hours of the allowed 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Therefore, the AOT extension will provide improved flexibility in completing maintenance activities. The 7-day AOT may also be used for installation of planned ESCWS modification-related maintenance activities for equipment reliability.

The note shown in Section 2.4 of this attachment was added to allow temporary changes to TS 3.1.2.4, TS 3.5.2, TS 3.6.2.1, TS 3.6.2.2, TS 3.6.2.3, TS 3.7.1.2, TS 3.7.3, TS 3.7.4, TS 3.7.6, TS 3.7.7, TS 3.7.13, and TS 3.8.1.1. The note allowed these TS systems to be inoperable for a total of 14 days only to allow for the implementation of design improvements on the A ESW pump prior to October 29, 2016. The temporary extension of TS was taken and is expired.

2.4 Description of the Proposed Changes

1. The proposed license amendment revises TS 3.1.2.4, TS 3.5.2, TS 3.6.2.1, TS 3.6.2.3, TS 3.7.4 that will apply for B Train AOT extensions only, and TS 3.7.13. While the LAR is for maintenance on the ESCWS and air handlers supported by the ESCWS, the impact of this support system on other TS systems associated with the inoperable train is accounted for in the proposed TS changes.

The HNP TS require two ESCWS trains to be operable. Under the proposed change, an inoperable train of ESCWS must be restored to operable status within 7 days. In the condition with any one of these trains inoperable, the remaining operable train is adequate to provide chilled water to the cooling coils of air handling units that support the operable Charging Pump area, the ECCS System area, the Containment Spray System area, the Containment Cooling System area, and the Emergency Service Water System (ESWS) area. However, the overall reliability is reduced because a single failure to the operable train could result in a loss of function. The compensatory measures described in Section 3.6 of this attachment will manage risk during the proposed AOT. The PRA analysis and calculation contained in Attachment 5 conclude that the impact on plant risk is acceptable.

The proposed change revises the following HNP TS Action Statements:

Action Statement for TS 3.1.2.4 Action Statement A for TS 3.5.2 Action Statement for TS 3.6.2.1 Action Statement C for TS 3.6.2.3 Action Statement for TS 3.7.4 Action Statement for TS 3.7.13 The proposed change to the TS Action Statements listed above adds a note similar to the following that states:

One Train of [Applicable TS or TS System] is allowed to be inoperable for a total of 7 days to allow for maintenance on the Essential Services Chilled Water System and air handlers supported by the Essential Services Chilled Water System. Prior to exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the compensatory measures described in TS Bases 3.7.13 and HNP LAR correspondence letter RA-19-0007 shall be implemented.

U.S. Nuclear Regulatory Commission Page 11 of 29 Serial RA-19-0007 The TS Action Statements affected by this LAR are applicable in modes 1-4, with the exception of Action Statement for TS 3.1.2.4 and Action Statement A for TS 3.5.2, which are applicable in modes 1-3.

The proposed change to the TS Bases for TS 3.7.13 includes a description of the note added to TS for the ESCWS AOT extension and the compensatory measures that will be in place prior to exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

2. The proposed change to remove an expired note revises the following HNP TS Action Statements:

Action Statement for TS 3.1.2.4 Action Statement A for TS 3.5.2 Action Statement for TS 3.6.2.1 Action Statement for TS 3.6.2.2 Action Statement C for TS 3.6.2.3 Action Statement A for TS 3.7.1.2 Action Statement for TS 3.7.3 Action Statement for TS 3.7.4 Action Statement A.1 for TS 3.7.6 Action Statement A for TS 3.7.7 Action Statement for TS 3.7.13 Action Statement B.3 for TS 3.8.1.1 The proposed change to the TS Action Statements listed above is to remove the expired note that states:

[Applicable TS A Train or TS System] is allowed to be inoperable for a total of 14 days only to allow for the implementation of design improvements on the A Train ESW pump. The 14 days will be taken one time no later than October 29, 2016. During the period in which the A Train ESW pump supply from the Auxiliary Reservoir or Main Reservoir is not available, Normal Service Water (NSW) will remain available and in service to supply the A Train ESW equipment loads until the system is ready for post maintenance testing. Allowance of the extended Completion Time is contingent on meeting the Compensatory Measures and Conditions described in HNP LAR submittal correspondence letter HNP-16-056.

The TS Action Statements affected by this LAR are applicable in modes 1-4, with the exception of Action Statement for TS 3.1.2.4, Action Statement A for TS 3.5.2, and Action Statement A for TS 3.7.1.2, which are applicable in modes 1-3.

The TS Bases for TS 3.7.4, Emergency Service Water System, includes a description of the expired note and a list of conditions required for the 14-day AOT. The description of the note and list of conditions are no longer applicable to this TS Bases section since the note has expired.

The marked-up TS pages illustrating the proposed changes described above are provided in . The proposed TS Bases pages illustrating the proposed changes described above are provided in Attachment 3.

U.S. Nuclear Regulatory Commission Page 12 of 29 Serial RA-19-0007

3.0 TECHNICAL EVALUATION

3.1 Technical Specification Systems Affected by the Proposed Changes The TS changes for the systems affected by the proposed AOT extension are identified in Section 2.4 of this attachment. The following information provides a summary of the impact of the proposed TS changes on the operation of the Charging Pumps, ECCS, Containment Spray System, Containment Cooling System, and the ESWS due to the requested 7-day AOT extension for ESCWS maintenance.

Emergency Core Cooling System (ECCS) and Charging Pumps:

The ECCS provides shutdown capability by means of boron injection for the following accident conditions: a loss of coolant accident (LOCA) including a pipe break or a spurious relief or safety valve opening in the Reactor Coolant System (RCS) which would result in a discharge larger than that which could be made up by the normal makeup system; a rupture of a control rod drive mechanism causing a rod cluster control assembly ejection accident; a steam or feedwater system break accident including a pipe break or a spurious power operated relief or safety valve opening in the secondary steam system, which would result in an uncontrolled steam release or a loss of feedwater; and a steam generator tube failure. The system is designed to tolerate a single active failure (injection phase) or a single active or passive failure (recirculation phase). The capabilities are accomplished by a combination of suitable redundancy, instrumentation for indication and/or alarm of abnormal conditions, and relief valves to protect piping and components against malfunctions. The ECCS can meet its minimum required performance level with onsite or offsite electrical power. The ECCS consists of the centrifugal charging pumps, residual heat removal pumps, accumulators, a boron injection tank, residual heat removal heat exchangers, a refueling water storage tank, along with associated piping, valves, instrumentation, and other related equipment. Each train is powered from a separate ESF bus.

The ESCWS provides cooling to air handling units for the residual heat removal pump areas and the charging pump areas. During the proposed 7-day ESCWS AOT with one train of the ESCWS inoperable, the remaining operable ESCWS train is adequate to provide chilled water to the cooling coils of air handling units that support the operable train residual heat removal pump area and the charging pump area. The inoperable train residual heat removal pump and charging pump are not rendered unavailable as a result of the AOT entry.

Containment Spray System and the Containment Cooling System:

The Containment Spray System is designed to remove heat and fission products from a post-accident containment atmosphere by spraying borated sodium hydroxide solution into the containment. The Containment Spray System consists of two separate trains of equal capacity, each capable of meeting the system design basis spray coverage. Each train includes one containment spray pump, spray headers, nozzles, valves, piping, a containment recirculation sump, a cavitating venturi, and an eductor. The refueling water storage tank and the containment spray additive tank are common to both Containment Spray loops. Each train is powered from a separate ESF bus.

The Containment Fan Coolers ensure that adequate heat removal capacity is available when operated in conjunction with the Containment Spray System during post-LOCA conditions. The Containment Spray System and the Containment Fan Coolers are redundant to each other in

U.S. Nuclear Regulatory Commission Page 13 of 29 Serial RA-19-0007 providing post-accident cooling of the containment atmosphere. However, the Containment Spray System also provides a mechanism for removing iodine from the containment atmosphere.

The ESCWS provides cooling to air handling units for the containment spray pumps. The ESCWS also provides cooling to air handling units for the MCC area that supports the Containment Cooling System. During the proposed 7-day ESCWS AOT with one train of the ESCWS inoperable, the remaining operable ESCWS train is adequate to provide chilled water to the cooling coils of air handling units that support the operable train Containment Spray System and the Containment Cooling System. The inoperable train Containment Spray System and Containment Cooling System are not rendered unavailable as a result of the AOT entry.

Service Water System (SWS):

The Service Water system consists of two normal service water (NSW) pumps, two ESW pumps, two ESW booster pumps, associated piping, valves, and instrumentation. During unit start-up, shutdown, and normal operation, service water requirements are met by one of the NSW pumps. The NSW pump in operation takes suction from the circulating water cooling tower basin. The heated service water is returned to the cooling tower through the circulating water return pipes. The ultimate heat sink for HNP utilizes two alternate sources of cooling water: the Auxiliary Reservoir and the Main Reservoir. Under emergency conditions, the service water supply is switched from the cooling tower to the emergency service water pumps with preferred suction from the Auxiliary Reservoir through the ESW Intake Channel. The Main Reservoir serves as a backup supply of water for the Auxiliary Reservoir. Water from both the Main and Auxiliary Reservoirs passes through concrete intake structures. Each structure consists of bays separated by concrete walls. The ESW pumps are in dedicated bays in the ESW and Cooling Tower Make-up Water Intake Structure. The ESW pumps are not affected by the proposed AOT extension.

The ESW booster pump is provided to ensure that cooling water pressure inside the containment fan cooler units is higher than containment pressure during a LOCA. This prevents leakage of containment radioactivity into the ESW system. An orifice downstream of the fan cooler units provides increased system resistance during booster pump operation. The booster pump is placed in service by a SIAS. Start of the booster pump causes the orifice to be placed into service by closing the orifice bypass valve. Flow bypasses the booster pump and orifice during normal plant operation. The ESCWS provides cooling to the air handling units for the A Train and B Train ESW booster pumps. The A Train ESW booster pump area may be cooled by either of two air handling units that are powered by the A Train and B Train respectively.

The B Train ESW booster pump area may be cooled by AH-8 1X-SB, which is powered by the B Train power supply. There is no impact to the A Train ESW system with the proposed ESCWS AOT since an air handler unit for this area may be powered by either A or B Train.

When the proposed 7-day ESCWS AOT is entered on the B Train of ESCWS, the B Train ESW booster pump will be inoperable and the operable (A Train) ESCWS train is adequate to provide chilled water to the cooling coils of air handling units that support the operable train ESW booster pump. For the B Train ESCWS AOT entries, the inoperable B Train ESW booster pump is not rendered unavailable as a result of the AOT entry.

The proposed changes to remove expired notes from TS 3.1.2.4, TS 3.5.2, TS 3.6.2.1, TS 3.6.2.2, TS 3.6.2.3, TS 3.7.1.2, TS 3.7.3, TS 3.7.4, TS 3.7.6, TS 3.7.7, TS 3.7.13, and TS 3.8.1.1 are administrative non-technical changes which remove temporary TS changes added to

U.S. Nuclear Regulatory Commission Page 14 of 29 Serial RA-19-0007 support the A ESW pump replacement. These temporary requirements are no longer necessary since the A ESW pump replacement is complete. Upon approval and implementation of the proposed changes, the HNP TS will no longer contain the note that is currently expired. This proposed change implements an administrative non-technical change, and given the above, additional technical evaluation of the administrative non-technical changes proposed in this LAR is not necessary.

3.2 Room Heatup Analysis A room heatup analysis has been completed for HNP using the GOTHIC computer code. This analysis documents expected area temperatures in the RAB when the HVAC systems are not functioning. The areas investigated are located on the 190, 236, 261, 286, and 305 elevations of the RAB. The 19 areas initially reviewed include the Residual Heat Removal system /

Containment Spray rooms, Charging Safety Injection Pump (CSIP) rooms, Component Cooling Water (CCW) system area, Boric Acid Pumps area, Secondary Waste Sample Tank & Pump area, HVAC Chiller areas, Switchgear Rooms, Battery Rooms, Transfer Panel areas, PIC Room, Auxiliary Relay Cabinet Room, and the Main Control Room. Room heat up following the loss of HVAC was evaluated. This analysis conservatively screens rooms located in the RAB for unacceptable temperatures resulting from a loss of HVAC event. The initial screening determined the maximum temperature in each area investigated within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following a loss of HVAC event. The resulting temperatures were evaluated for acceptability based upon industry standards for equipment temperature limits, equipment manufacturer qualification packages, and equipment qualification temperature values. The initial screening conservatively assumes a loss of HVAC in all rooms modeled. Screening was performed using GOTHIC parameter volumes, thermal conductors, and heater components. The highest heat loads provided in the HVAC calculations are utilized, which are typically conservative.

The results of the initial screening in the heatup analysis determined that most areas analyzed will not exceed unacceptable temperature values within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following a loss of HVAC. The following areas were determined to potentially warrant operator action to ensure temperatures are maintained within acceptable values:

CSIP Room Areas: The room heatup analysis evaluated the effectiveness of an operator action to open the CSIP Room door and install a portable fan in the CSIP room door, when room cooling was lost, to keep the CSIP room temperature within acceptable limits for pump operation. The portable fan is powered by a permanent non-safety power supply. Based on this evaluation and development of plant procedures to implement the action to put a fan in the CSIP Room door upon loss of room cooling, the HNP PRA model was revised to include this operator action.

The Switchgear Rooms: The Switchgear Room cooling loads during the summer months are low enough that without operator actions, the equipment would continue to meet the PRA success criteria. Winter loads are greater and operator actions are required. When the ESCWS chillers and air handling unit fans are not available during winter operation, electric heating coil unit loading of transformers in the B Switchgear Room causes the acceptance criteria to be exceeded. However, operator actions are available to aide in maintaining Switchgear room temperatures within acceptable limits, such as placing ventilation in the smoke purge mode of operation. The smoke purge operation of this ventilation system relies upon non-safety power. The opening of room doors is another means of maintaining

U.S. Nuclear Regulatory Commission Page 15 of 29 Serial RA-19-0007 acceptable room temperatures; however, this is not the preferred method of cooling for this area. As a result of this analysis, the loss of room cooling to the Switchgear Rooms has been modeled and included in the HNP PRA Model.

Auxiliary Relay Cabinet Room: The maximum temperature determined in the Auxiliary Relay Cabinet Room within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following a loss of HVAC indicates that operator action may be needed to maintain acceptable temperatures. Upon a loss of HVAC to the Auxiliary Relay Cabinet Room, an operator may place the ventilation system in the smoke purge operation mode, if needed, that will allow outside air entry into the room. The smoke purge operation of this ventilation system relies upon non-safety power. The opening of room doors is another means of maintaining acceptable room temperatures; however, this is not the preferred method of cooling for this area. The Auxiliary Relay Cabinet Room is not included in the HNP PRA Model based upon the results of the heatup analysis.

Special consideration was given to failure of the Control Room Complex HVAC, since increased temperatures in relay and instrumentation cabinets could result in failure of these components and affect control of the plant. The Control Room Complex is continuously manned, so HVAC malfunctions would be immediately noticed, either by redundant alarms or by physically noticing the change in temperature, and investigated. In the event of an HVAC malfunction, compensatory actions such as placing temporary fans, opening cabinet doors, taking manual control of components or using local indications would occur. Based on this consideration, a loss of HVAC to the Control Room Complex is not expected to significantly contribute to the overall risk of severe accidents.

To summarize, due to the design of the HNP, with very large equipment spaces, the PRA model only requires the ESCWS chillers for the CSIP Rooms and the Switchgear Rooms. Even for those cases, operators may place the ventilation system in the smoke purge operation mode to allow air entry if needed upon a loss of both ESCWS Chillers. The installation of spot coolers in certain areas for personnel comfort may be also completed when both of the ESCWS Chillers are unavailable.

3.3 Additional System for Safe Shutdown:

The Alternate Seal Injection (ASI) System is an independent, automatically-actuated back-up system for seal injection on the reactor coolant pumps (RCPs) that is not reliant on normal plant electrical or cooling systems. The ASI System also provides defense in depth for Station Blackout (SBO) coping, since the loss of normal seal injection coupled with loss of CCW seal cooling during a SBO will result in seal failure and consequential RCS leakage. The ASI System provides seal cooling which prevents RCP seal failure and provides a means of Reactor Coolant makeup. This significantly improves the ability of HNP to provide RCS cooling and inventory control during a SBO event. The ASI System is supported by the Dedicated Shutdown Diesel Generator (DSDG) System. The DSDG System is composed of a generator and its associated engine and control system, as well as the supporting sub-systems of instrumentation and valves. This system does not meet all the requirements to be considered an emergency alternating current (AC) power source since it does not supply all the loads needed for safe shutdown of the plant. Instead, it provides an independent 480-volt electrical power source for the ASI System and other important loads to augment HNPs ability to achieve safe shutdown of the plant in the event of a fire or SBO.

U.S. Nuclear Regulatory Commission Page 16 of 29 Serial RA-19-0007 In order to prevent damage to RCP seals, the ASI System is designed to be fully initiated in approximately two minutes and 45 seconds from detection of loss of RCP seal coolant flow. The time delay between detection of a loss of normal power (and resulting loss of power to the ASI System) and re-energizing of the MCC by the DSDG at minimum operating parameters is approximately 25 seconds for normal operation. However, the DSDG is programmed to crank the engine up to three times (seven seconds on, seven seconds off each time), which means it could take up to 60 seconds to begin supplying MCC loads. Thus, the DSDG System is designed to properly support the ASI System when a loss of normal power is postulated concurrently with a loss of normal RCP seal injection cooling.

The primary function of the DSDG System is to provide emergency backup power to the ASI system, which is a backup to the normal reactor coolant pump seal injection system, and to provide power for charging emergency batteries in the event of a loss of offsite power or any other interruption of the ASI Systems normal feed. The DSDG System allows the ASI System to actuate and operate independently of existing plant power, and is initiated upon loss of the normal power source to the ASI System.

The DSDG System is designed to automatically start the DSDG upon loss of normal AC power.

The system is also designed to automatically connect and supply loads, as well as stop the DSDG and re-transfer to normal plant power after it is restored. The DSDG System is able to provide an alternate 480 volts alternating current (VAC) feed for a maximum 400 kilowatts (kW) of load for a minimum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Prior to exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the proposed 7-day ESCWS AOT on either train, action will be taken to ensure no discretionary maintenance or discretionary testing is planned for the remaining duration on the ASI system or the DSDG system.

3.4 Risk Analysis for the Proposed Changes The evaluation for the proposed AOT extension consisted of a review of the impacted plant systems and their safety functions. Duke Energy has quantitatively and qualitatively assessed the risk impact on the affected safety functions. There are no systems, structures, or components (SSCs) that will change status due to the changes. No new accidents or transients will be introduced by the proposed changes. No physical changes are being made to any of the systems affected by the AOT extension. The function and operation of these systems will remain the same, as described in the FSAR. Protective measures will be taken to ensure that unanticipated compromises to system redundancy, independence, and diversity will not occur during maintenance activities. The impact of the proposed change on safety margins was also considered. Extending the AOT to 7 days for one inoperable train of the ESCWS does not impact any assumptions or inputs in the FSAR.

The PRA analysis and calculation for the proposed AOT are presented in Attachment 5. Plant risk impacts were assessed quantitatively using the internal events, internal flooding, high winds and fire PRA models. Additionally, plant risk was assessed qualitatively for external flooding and seismically-induced events. The results show that the risk significance from extending the proposed AOT for an inoperable ESCWS train from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 7 days is small and within limits provided by Regulatory Guide (RG)1.174 (Reference 2) and RG 1.177 (Reference 3). Risk-informed improvements to TS are intended to maintain or improve safety while reducing unnecessary burden, and to bring TS into congruence with the Commissions other risk-

U.S. Nuclear Regulatory Commission Page 17 of 29 Serial RA-19-0007 informed regulatory requirements, in particular the risk assessment and management requirements of 10 CFR 50.65(a)(4).

3.5 Defense-in-Depth Considerations Duke Energy work management procedure guidance requires both a deterministic and probabilistic evaluation of risk for the performance of all maintenance activities. This procedure guidance uses the Level 1 probabilistic safety analysis model to evaluate the impact of maintenance activities on core damage frequency. The Maintenance Rule Program and system unavailability limits will control the frequency of entry into the proposed AOT. The proposed AOT will be implemented when the plant is in modes 1-4 and will not be performed on A and B ESCWS trains simultaneously.

Defense-in-Depth Principles:

In addition to the TS, the Work Management Program and the associated procedures and programs that implement the Maintenance Rule Program under 10 CFR 50.65(a)(4) provide for controls and assessments to preclude the possibility of simultaneous planned outages of redundant trains and ensure system reliability. This proposed LAR meets the defense-in-depth principles described in RG 1.177 and RG 1.174. The following elements, as identified in RG 1.174, Section 2.1.1.2, have been evaluated. The impact of the proposed change on these elements is as follows:

1. Preserve a reasonable balance among the layers of defense (i.e., minimizing challenges to the plant, preventing any events from progressing to core damage, containing the radioactive source term, and emergency preparedness).

Prevention of core damage depends on the ability to continuously remove decay heat after an initiating event. During the extended AOT of 7 days, if a design basis accident occurred, the operable ESCWS train remains available to cool areas of plant equipment that are needed to mitigate the event.

Due to the design of the HNP, with very large equipment spaces, the PRA model only requires the ESCWS chillers for the CSIP Rooms and the Switchgear Rooms, as described in Section 3.2 of this attachment. The heatup analysis described in Section 3.2 of this attachment conservatively screens rooms located in the RAB for unacceptable temperatures resulting from a loss of HVAC event. The initial screening determined the maximum temperature in each area investigated within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following a loss of HVAC event. The resulting temperatures were evaluated for acceptability based upon industry standards for equipment temperature limits, equipment manufacturer qualification packages, and equipment qualification temperature values. The initial screening conservatively assumes a loss of HVAC in all rooms modeled.

Screening was performed using GOTHIC parameter volumes, thermal conductors, and heater components. The highest heat loads provided in the HVAC calculations are utilized, which are typically conservative.

The results of the initial screening in the heatup analysis determined that most areas analyzed will not exceed acceptable temperature values within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following a loss of HVAC. There are a few areas that were determined to potentially warrant operator action to ensure temperatures are maintained within acceptable values, as described in Section 3.2 of this attachment. Prior to exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from the time of TS 3.7.13 LCO entry, operator actions

U.S. Nuclear Regulatory Commission Page 18 of 29 Serial RA-19-0007 for the CSIP area cooling, Auxiliary Relay Cabinet Room cooling, and Switchgear Room cooling following a loss of HVAC will be briefed with Operations, as described in Section 3.6 of this attachment. The fan used for the CSIP area cooling will be pre-staged and verified to be functional.

The capability to implement operator actions for the CSIP area cooling, Switchgear Room cooling, and Auxiliary Relay Cabinet Room cooling, as described in Section 3.2 of this attachment, provide defense in depth. These defense-in-depth measures are aimed at ensuring availability of the CSIP and maintaining temperature limits established for the Switchgear Rooms and the Auxiliary Relay Cabinet Room if a failure of the operable ESCWS train were to occur. These defense-in-depth measures have been established to prevent core damage, containment damage, and to preserve consequence mitigation. The ability of Duke Energy staff to respond to an emergency at HNP is not impacted by this change.

2. Preserve adequate capability of design features without an overreliance on programmatic activities as compensatory measures.

The proposed license amendment involves a change to the AOT with one train of the ESCWS operable in modes 1-4. During the timeframe of the proposed AOT on one train of the ESCWS, the opposite train of the ESCWS will remain operable and capable of performing necessary safety functions, consistent with accident analysis assumptions. The AOT on the system that is currently allowed by the HNP TS for a time period of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months will be extended to 7 days. The safety analysis acceptance criteria stated in the FSAR are not impacted by this change. The proposed change will not allow plant operation in a configuration outside the design basis. The only programmatic features are those associated with risk management actions described in Section 3.6 of this attachment. These compensatory measures provide a qualitative risk impact to the PRA analysis and calculation for this LAR; no quantitative credit was taken in the PRA analysis for any of the proposed compensatory measures.

Due to the design of HNP, with very large equipment spaces, the PRA model only requires the ESCWS chillers for the CSIP Rooms and the Switchgear Rooms, as described in Section 3.2 of this attachment. The heatup analysis described in Section 3.2 of this attachment conservatively screens rooms located in the RAB for unacceptable temperatures resulting from a loss of HVAC event. The results of the initial screening in the heatup analysis determined that most areas analyzed will not exceed unacceptable temperature values within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following a loss of HVAC. There are a few areas that were determined to potentially warrant operator action to ensure temperatures are maintained within acceptable values, as described in Section 3.2 of this attachment. Prior to exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from the time of TS 3.7.13 LCO entry, operator actions for the CSIP area cooling, Auxiliary Relay Cabinet Room cooling, and Switchgear Room cooling following a loss of HVAC will be briefed with Operations, as described in Section 3.6 of this attachment.

The fan used for the CSIP area cooling will be pre-staged and verified to be functional. The defense-in-depth measure for the CSIP area cooling relies upon proceduralized actions that have been in place since 2007. Operators are familiar with this defense-in-depth measure and routinely use this guidance during certain maintenance activities to limit unavailability of a CSIP.

Additionally, prior to exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the proposed AOT, action will be taken to ensure no discretionary maintenance or discretionary testing is planned for the remaining duration on the

U.S. Nuclear Regulatory Commission Page 19 of 29 Serial RA-19-0007 Motor-Driven and Turbine-Driven Auxiliary Feedwater (AFW) Pumps, the ESW System, the NSW System, EDGs, the ASI System, and the DSDG System. A posted protected equipment list will be in effect for the CSIP rooms, the Switchgear Rooms, and the opposite train (operable)

ESCW chiller. The Fire Protection tracking log will be reviewed for fire hazards and fire impairments. Transient combustibles and hot work in fire risk-sensitive areas will be limited.

Restrictions on work activities will be in place that involve components that if lost or failed could result in a direct plant trip or transient. The compensatory measures described in Section 3.6 of this attachment are intended to reduce the potential of risk-significant configurations, however are not overly relied upon in the PRA analysis for the proposed licensing amendment.

3. Preserve system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty.

During normal operation, the RABNVS and the ESCWS are designed to provide ventilation for areas containing equipment essential for safe shutdown in the RAB, including the CVCS chiller area, 480-volt auxiliary bus area, areas containing non-essential equipment, surrounding access aisles and RAB stairways and H&V equipment rooms, as described in Section 2.1 of this attachment. Numerous areas cooled by fans that rely on the ESCWS contain redundant fans as identified in Section 2.1 of this attachment and in the table of areas served by ESF fan coolers in the RAB that is contained in Attachment 4 of this LAR.

During the timeframe of the proposed AOT on one train of the ESCWS, the opposite train of the ESCWS will remain completely operable and capable of performing the necessary safety functions, consistent with accident analysis assumptions. Additionally, prior to exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the proposed AOT, action will be taken to ensure no discretionary maintenance or discretionary testing is planned for the remaining duration on the Motor-Driven and Turbine-Driven AFW Pumps, the ESW System, the NSW System, EDGs, the ASI System, and the DSDG System. A posted protected equipment list will be in effect for the CSIP rooms, the Switchgear Rooms, and the opposite train (operable) ESCW chiller, as identified in Section 3.6 of this attachment. The PRA analysis for this LAR indicates that the proposed AOT extension provides acceptable system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty.

4. Preserve adequate defense against potential common-cause failures (CCFs).

The compensatory measures described in Section 3.6 assure the availability of independent, redundant, and diverse means of accomplishing critical safety functions during the proposed AOT duration. There is no change in failure mechanisms associated with the ESCWS as a result of the AOT change from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 7 days.

5. Maintain multiple fission product barriers.

The proposed ESCWS AOT change does not directly impact any of the three fission product barriers (Fuel Cladding, Reactor Coolant System, Containment Building) or otherwise cause their degradation. With the ability to continuously remove decay heat, the proposed change does not affect the fuel cladding. The reactor coolant pressure boundary and Containment

U.S. Nuclear Regulatory Commission Page 20 of 29 Serial RA-19-0007 Building are not challenged by this LAR. Independence of barriers is not degraded because the proposed TS AOT extension has no impact on the physical barriers.

6. Preserve sufficient defense against human errors.

Duke Energy will provide oversight and support for emergent issues. Defense-in-depth measures include operator actions for the CSIP area cooling, Auxiliary Relay Cabinet Room cooling, and Switchgear Room cooling following a loss of HVAC to these areas. Prior to exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from the time of TS 3.7.13 LCO entry, these actions will be briefed with Operations. The fan used for the CSIP area cooling will be pre-staged and verified to be functional at this time also. Additionally, prior to exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from the time of TS 3.7.13 LCO entry, equipment will be protected and administrative controls will be in place to support the compensatory measures described in Section 3.6 of this attachment. Pre-job briefs will be conducted prior to and during the evolution to reinforce good human performance behaviors and barriers that reduce risk. The opposite train of all associated TS will be protected during the AOT. Duke Energy fleet staff will be available to support plant staff with resolution of issues during the proposed AOT.

7. Continue to meet the intent of the plants design criteria.

This activity does not modify the plant design or the design criteria applied to SSCs during the licensing process. Prior to exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the proposed AOT, Duke Energy will implement compensatory measures that will include the limitation of discretionary maintenance or testing on equipment as described in Section 3.6. These compensatory measures are intended to manage risk during the proposed AOT. The ESCWS is designed to meet the requirements of GDC 2, 44, 45, and 46. The ESCWS is designed to be operated in the proposed manner. Additional details regarding compliance with the General Design Criteria are provided in Section 4.1 of this attachment.

3.6 Assumptions and Compensatory Measures The assumptions used for the PRA analysis and calculation described in Attachment 5 include compensatory measures that will be utilized during the proposed AOT, which are listed below.

These compensatory measures provide a qualitative risk impact to the calculation results; no quantitative credit was taken in the PRA analysis for any of the proposed compensatory measures. These compensatory measures will be implemented prior to exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from the time of TS 3.7.13 LCO entry. These compensatory measures are included in the proposed TS Bases changes shown in Attachment 3 of this LAR.

1. The following equipment and the corresponding power supplies will be posted protected:

Air handling units for the operable CSIP areas:

AH-9A (CSIP 1A-SA Area), AH-9B (CSIP 1B-SB Area), or AH-10 (CSIP 1C-SAB Area)

Air handling units for the Switchgear Rooms with operable equipment:

AH-12 1A-SA and AH-12 1B-SA supply Switchgear Room A; AH-13 1A-SB and AH-13 1B-SB supply Switchgear Room B

Operable ESCWS chiller and operable chilled water pump

U.S. Nuclear Regulatory Commission Page 21 of 29 Serial RA-19-0007

2. The Fire Protection tracking log will be reviewed for fire hazards and fire impairments.

Transient combustibles and hot work in these fire risk-sensitive areas will be limited:

Fire compartment FC25 - RAB HVAC Room (MCC 1A21-SA, MCC 1A31-SA)

Fire compartments FC34 and FC35 - Switchgear Room A and Switchgear Room B

Fire compartment FC41 - Turbine Building (Zone 1-G-261 - 6.9 kV Switchgear)

Fire compartment FC54 - Transformer Yard

3. Restrictions on work activities will be in place that involve components that if lost or failed could result in a direct plant trip or transient.

4. Operator actions for the CSIP area cooling, Switchgear Room cooling, and Auxiliary Relay Cabinet Room cooling, if needed following a loss of HVAC, will be briefed with Operations.

The fan used for the CSIP area cooling will be pre-staged and verified to be functional.

5. Discretionary maintenance or discretionary testing on equipment that support the following systems will be avoided for the remaining duration of the TS 3.7.13 LCO entry:

ESCW System (operable train)

Motor-Driven and Turbine-Driven AFW Pumps

ESW System and NSW System

EDGs

ASI System and DSDG System 3.7 Compliance with Current Regulations This LAR itself does not propose to deviate from existing regulatory requirements. Compliance with existing regulations is maintained by the proposed change to the plant's TS requirements.

Additional details may be found in Section 4 of this LAR.

3.8 Evaluation of Safety Margins The design and operation of the ESCWS is not altered by this LAR. The AOT on the system currently allowed by the HNP TS for a time period of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months will be extended to 7 days. The safety analysis acceptance criteria stated in the FSAR are not impacted by this change. The proposed change will not allow plant operation in a configuration outside the design basis. The requirements regarding the ESCWS credited in the accident analysis will remain the same. As such, it can be concluded that safety margins are not impacted by the proposed change.

The proposed change involves an AOT extension of the current TS listed in Section 2.4 of this attachment. The systems that are affected during a particular ESCWS outage time period are all associated with the train that corresponds to the affected ESCWS train, leaving one train of safety equipment fully operable and capable of performing its safety functions. Preserving the operability of one ESCWS train during the 7-day AOT will maintain the balance among the prevention of core damage, prevention of containment failure, and consequence mitigation.

The HNP PRA model is sufficiently robust and suitable for use in risk-informed processes such as for regulatory decision making. The peer reviews that have been conducted and the resolution of findings from those reviews demonstrate that the internal events, internal flooding, high winds, and fire models of the PRA have been performed in a technically correct manner.

The assumptions and approximations used in development of the PRA have also been reviewed and are appropriate for their application. Duke Energy procedures are in place for

U.S. Nuclear Regulatory Commission Page 22 of 29 Serial RA-19-0007 controlling and updating the models, when appropriate, and for assuring that the models represent the as-built, as-operated plant. The conclusion, therefore, is that the HNP PRA models are acceptable for use as the basis for risk-informed applications, including assessment of proposed TS amendments. Attachment 5 contains the PRA analysis and calculation completed for the proposed ESCWS AOT extension.

RG 1.177, Section 2.4, provides the following acceptance guidelines for evaluating the risk associated with the revised completion time:

The licensee has demonstrated that the TS completion time (CT) change has only a small quantitative impact on plant risk. An incremental conditional core damage probability (ICCDP) of less than 1.0x10-6 and an incremental conditional large early release probability (ICLERP) of less than 1.0x10-7 are considered small for a single TS condition entry.

The calculation results described in Attachment 5 of this LAR conclude that any time the 7-day AOT is entered, the ICCDP is 1.7 x10-8 and the ICLERP is 3.0 x10-10.

RG 1.174 provides guidance on delta core damage frequency (CDF) and delta large early release frequency (LERF) values. The 7-day AOT entry has a delta CDF of 8.7 x10-7 per year and a delta LERF of 1.6 x10-8 per year. These delta CDF and delta LERF values are considered to represent a very small risk increase, as presented in Figures 4 and 5 of RG 1.174. Therefore, these metrics satisfy the risk guidelines of RG 1.174 and RG 1.177 and represent an insignificant impact on average annual plant risk.

3.9 Configuration Risk Management 10 CFR 50.65 (a)(4) 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. Duke Energy has work management and execution procedures that are in place to ensure that risk-significant plant configurations are avoided. These documents are used to address the Maintenance Rule requirements, including the on-line (and off-line) maintenance policy requirement to control the safety impact of combinations of equipment removed from service.

The proposed LAR will not result in any changes to the current configuration risk management program. Duke Energy manages this process using a blended (i.e., quantitative and qualitative) risk assessment approach with its Electronic Risk Assessment Tool (ERAT). The Phoenix software program is used to analyze plant risk in both real time ('Operator Screen' mode) as well as a look-ahead of plant configurations over a specified period of time ('Scheduler Screen' mode). Prior to entering the proposed 7-day AOT, operators will review the plant schedule to identify and correct any significant potential risk impacts occurring during the AOT. During the AOT, risk will be monitored in real time and any emergent risk configurations will be addressed appropriately.

Additionally, prior to planned work execution, scheduling personnel must consider the effects of severe weather and grid instabilities on plant operations. This qualitative evaluation is inherent of the duties of Work Management. Responses to actual plant risk due to severe weather or grid instabilities are programmatically incorporated into applicable plant emergency or response procedures.

U.S. Nuclear Regulatory Commission Page 23 of 29 Serial RA-19-0007 The key safety significant systems impacted by this proposed LAR are currently included in the Maintenance Rule program and, as such, availability and reliability performance criteria have been established to assure that they perform adequately.

3.10 Conclusions The results of the justification described above provide assurance that the systems and equipment required to safely shut down the plant and mitigate the effects of a design basis accident will remain capable of performing their safety functions, with the established assumptions and compensatory measures in place for the proposed AOT. The proposed TS AOT extension is consistent with NRC guidance and meets the principles of current regulations, defense-in-depth philosophy, and maintains sufficient safety margins.

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 NRC 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.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria General Design Criteria, as outlined in 10 CFR 50, Appendix A, were considered for the proposed amendment. Duke Energy will maintain the ability to meet GDC 2, 44, 45, and 46, which are applicable to the ESCWS design, with the proposed licensing amendment.

Additionally, Duke Energy will maintain the ability to meet GDC 35, 36, and 37, which are applicable to ECCS design, with the proposed licensing amendment. The applicable GDCs considered are described below:

GDC-2: Design bases for protection against natural phenomena Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflect: (1) Appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be performed.

GDC-35: Emergency core cooling A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite

U.S. Nuclear Regulatory Commission Page 24 of 29 Serial RA-19-0007 electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

GDC-36: Inspection of emergency core cooling system The emergency core cooling system shall be designed to permit appropriate periodic inspection of important components, such as spray rings in the reactor pressure vessel, water injection nozzles, and piping, to assure the integrity and capability of the system.

GDC-37: Testing of emergency core cooling system The emergency core cooling system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the active components of the system, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system.

GDC-44: Cooling Water A system to transfer heat from structures, systems, and components important to safety, to an ultimate heat sink shall be provided. The system safety function shall be to transfer the combined heat load of these structures, systems, and components under normal operating and accident conditions. Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

GDC-45: Inspection of cooling water system The cooling water system shall be designed to permit appropriate periodic inspection of important components, such as heat exchangers and piping, to assure the integrity and capability of the system.

GDC-46: Testing of cooling water system The cooling water system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and the performance of the active components of the system, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation for reactor shutdown and for loss-of-coolant accidents, including operation of applicable portions of the protection system and the transfer between normal and emergency power sources.

RG 1.174 (Reference 2) describes a risk-informed approach, acceptable to the NRC, for assessing the nature and impact of proposed permanent licensing-basis changes by considering engineering issues and applying risk insights. This RG also provides risk-acceptance guidelines for evaluating the results of such evaluations.

U.S. Nuclear Regulatory Commission Page 25 of 29 Serial RA-19-0007 RG 1.177 (Reference 3) describes an acceptable risk-informed approach specifically for assessing proposed TS changes in AOTs. This RG describes a three-tiered approach for licensees to evaluate the risk associated with proposed TS AOT changes.

Tier 1 of RG 1 .177 assesses the risk impact of the proposed change in accordance with acceptance guidelines consistent with the Commission's Safety Goal Policy Statement, as described in RG 1.177. Tier 1 assesses the impact on operational plant risk based on the change in CDF and change in LERF. It also evaluates plant risk while equipment covered by the proposed AOT is out of service, as represented by ICCDP and ICLERP.

Tier 2 of RG 1.177 is the identification of potentially high-risk configurations that could result if equipment, in addition to that associated with the proposed license amendment, is taken out of service simultaneously, or if other risk-significant operational factors, such as concurrent system or equipment testing, are also involved. The purpose of this evaluation is to ensure that appropriate restrictions on dominant risk-significant configurations associated with the change are in place.

Tier 3 of RG 1.177 requires the licensee to provide assurance of compliance with 10 CFR 50.65(a)(4) to ensure the risk impact of taking equipment out of service is appropriately assessed and managed.

RG 1.200 (Reference 4) describes an acceptable approach for determining whether the quality of the PRA model, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA model can be used in regulatory decision making for light-water reactors.

NUREG-1855 (Reference 5) provides guidance on how to treat uncertainties associated with PRA in risk-informed decisionmaking. NUREG-1855 focuses on epistemic uncertainty and the guidance provided includes acceptable methods of identifying and characterizing the different types of epistemic uncertainty and the ways that those uncertainties are treated. In accordance with NUREG-1855, sensitivities were performed as needed to verify the key sources of uncertainty for the proposed ESCWS AOT extension. The evaluation of uncertainties for the proposed ESCWS AOT extension is addressed in Attachment 5 of this submittal.

As stated previously, although a train of the ESCWS will be inoperable during the proposed 7-day AOT, the equipment it supports will remain in its normal ESF actuation system configuration and will be functional. The opposite train of the ESCWS will remain operable. NRC Generic Letter 80-30 states that the specified time to take action when an LCO is not met is a temporary relaxation of the single failure criterion since the completion time provides a limited time to fix equipment or otherwise make it operable.

There are no permanent changes to the design of the ESCWS or its supported systems involved with this LAR. The evaluations provided within this proposed amendment confirm that the plant will continue to comply with the applicable design criteria. Additionally, prior to exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the proposed 7-day ESCWS AOT on either train, action will be taken to ensure no discretionary maintenance or discretionary testing is planned for the remaining duration on the ESCW System (operable train), Motor-Driven and Turbine-Driven AFW Pumps, ESW system, EDGs, ASI system, DSDG system, or the NSW system.

U.S. Nuclear Regulatory Commission Page 26 of 29 Serial RA-19-0007 In this configuration, the operable train will respond as designed during design basis events.

The requested period of 7 days to complete the required actions of the affected TS is reasonable considering the redundant capabilities of the above systems, the defense-in-depth measures that will be available, and compensatory measures that will be in place as discussed in Section 3.6 of this attachment.

The proposed change to remove an expired note from TS 3.1.2.4, TS 3.5.2, TS 3.6.2.1, TS 3.6.2.2, TS 3.6.2.3, TS 3.7.1.2, TS 3.7.3, TS 3.7.4, TS 3.7.6, TS 3.7.7, TS 3.7.13, and TS 3.8.1.1 is administrative and non-technical in nature. Upon approval and implementation of this proposed change, the HNP TS will continue to comply with the applicable regulatory requirements and criteria discussed in the regulatory evaluation associated with Duke Energy's October 29, 2015, LAR (ADAMS Accession No. ML15302A542), as well as the requirements described in the License Amendment 153 Safety Evaluation Report, dated September 16, 2016 (ADAMS Accession No. ML16253A059), for this LAR. Therefore, additional discussion of the applicable regulatory requirements and criteria is not required.

4.2 Precedents While no exact precedent for an ESCWS AOT extension was identified for the proposed LAR, two license amendments were identified that involve permanent AOT changes for TS systems from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 7 days. Crystal River Unit 3 submitted a license amendment request to extend the AOT from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 7 days for an inoperable low-pressure injection (LPI) train, reactor building spray (RBS) train, decay heat closed cycle cooling water train, and decay heat seawater train. This license amendment was approved by the NRC on April 30, 2008 (ADAMS Accession No. ML081060231). Oconee Nuclear Station, Units 1, 2, and 3, submitted a license amendment request to extend the AOT from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 7 days for an inoperable LPI train. This license amendment was approved by the NRC on June 18, 2003 (ADAMS Accession No. ML031690273). Each of these license amendments reference a Babcock and Wilcox Topical Report BAW-2295A, Revision 1. The results of the analysis in the topical report show that the risk significance from extending the completion time for an inoperable LPI train or an inoperable RBS system from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 7 days was small and within RG 1.174 and RG 1.177 guidance.

The results of the analysis in this ESCWS AOT extension proposed LAR present a similar conclusion.

No precedent letters were identified for the removal of the expired note from TS 3.1.2.4, TS 3.5.2, TS 3.6.2.1, TS 3.6.2.2, TS 3.6.2.3, TS 3.7.1.2, TS 3.7.3, TS 3.7.4, TS 3.7.6, TS 3.7.7, TS 3.7.13, and TS 3.8.1.1.

4.3 Significant Hazards Consideration Pursuant to 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy), proposes a license amendment request (LAR) for the Technical Specifications (TS) for Shearon Harris Nuclear Power Plant, Unit 1 (HNP). The proposed LAR revises HNP TS 3.7.13, Essential Services Chilled Water System, and associated TS Sections for systems supported by the Essential Services Chilled Water System (ESCWS) that includes TS 3.1.2.4, Charging Pumps -

Operating, TS 3.5.2, ECCS [Emergency Core Cooling Systems] Subsystems - Tavg Greater Than or Equal To 350°F, TS 3.6.2.1, Containment Spray System, TS 3.6.2.3, Containment Cooling System, and TS 3.7.4, "Emergency Service Water System," for B Train ESCWS inoperability only, based upon the impact to the B Emergency Service Water (ESW) booster

U.S. Nuclear Regulatory Commission Page 27 of 29 Serial RA-19-0007 pump operability. This change is to allow for maintenance activities on the ESCWS and air handlers supported by the ESCWS for equipment reliability. The proposed amendment will permit one train of the ESCWS to be inoperable for a total of 7 days.

In addition, this proposed amendment removes an expired note from TS 3.1.2.4, TS 3.5.2, TS 3.6.2.1, TS 3.6.2.2, TS 3.6.2.3, TS 3.7.1.2, TS 3.7.3, TS 3.7.4, TS 3.7.6, TS 3.7.7, TS 3.7.13, and TS 3.8.1.1, which was previously added to permit temporary TS changes for replacement of the A ESW pump replacement. The A ESW pump replacement was completed on September 29, 2016. These changes are administrative and non-technical in nature.

Duke Energy has evaluated whether or not a significant hazards consideration is warranted with the proposed amendment by addressing the three criterion set forth in 10 CFR 50.92(c) as described below:

(1) Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

The operable train of the ESCWS and supported equipment will remain fully operable during the 7-day allowed outage time. The unavailable train of the ESCWS and supported equipment function as accident mitigators. The removal of a train of the ESCWS from service for a limited period of time does not affect any accident initiator and therefore cannot change the probability of an accident. The proposed change has been evaluated to assess the impact on systems affected and the upon design basis safety functions.

The activities covered by this LAR also include defense-in-depth compensatory measures. There will be no effect on the analysis of any accident or the progression of the accident since the operable ESCWS train is capable of serving 100 percent of all the required heat loads. As such, there is no impact on consequence mitigation for any transient or accident.

The proposed changes to TS 3.1.2.4, TS 3.5.2, TS 3.6.2.1, TS 3.6.2.2, TS 3.6.2.3, TS 3.7.1.2, TS 3.7.3, TS 3.7.4, TS 3.7.6, TS 3.7.7, TS 3.7.13, and TS 3.8.1.1 that remove an expired note are administrative, non-technical changes which remove temporary TS requirements added as part of the HNP License Amendment 153 issued on September 16, 2016 (Agencywide Documents Access and Management System Accession No. ML16253A059), that are currently obsolete.

As a result, operation of the facility in accordance with the proposed changes will not significantly increase the consequences of accidents previously evaluated.

(2) Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed amendment is 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 ESCWS and its supported TS systems that includes Charging Pumps, ECCS subsystems, Containment Spray System, Containment Cooling System, and the Emergency Service Water System, B Train. The requested change does not involve the addition or removal of any plant system, structure, or component.

U.S. Nuclear Regulatory Commission Page 28 of 29 Serial RA-19-0007 The proposed TS changes do not affect the basic design, operation, or function of any of the systems associated with the TS impacted by the amendment. Implementation of the proposed amendment will not create the possibility of a new or different kind of accident from that previously evaluated.

The proposed changes to TS 3.1.2.4, TS 3.5.2, TS 3.6.2.1, TS 3.6.2.2, TS 3.6.2.3, TS 3.7.1.2, TS 3.7.3, TS 3.7.4, TS 3.7.6, TS 3.7.7, TS 3.7.13, and TS 3.8.1.1 that remove an expired note are administrative, non-technical changes which remove temporary TS requirements added as part of the HNP License Amendment 153 issued on September 16, 2016, that are currently obsolete.

In conclusion, this proposed LAR does not impact any plant systems that are accident initiators and does not impact any safety analysis. Therefore, operation of the facility in accordance with the proposed changes will not create the possibility of a new or different kind of accident from any previously evaluated.

(3) Does the proposed amendment involve a significant reduction in a margin of safety?

The margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident condition.

These barriers include the fuel cladding, the reactor coolant system, and the containment system. The performance of the fuel cladding, reactor coolant, and containment systems will not be impacted by the proposed LAR.

Additionally, the proposed amendment does not involve a change in the operation of the plant. The activity only extends the amount of time a train of the ESCWS is allowed to be inoperable to complete maintenance for equipment reliability. The incremental conditional core damage probability (ICCDP) and incremental conditional large early release probability (ICLERP) calculated for the 7-day AOT are within the limits presented in Regulatory Guides 1.174 and 1.177.

The proposed changes to TS 3.1.2.4, TS 3.5.2, TS 3.6.2.1, TS 3.6.2.2, TS 3.6.2.3, TS 3.7.1.2, TS 3.7.3, TS 3.7.4, TS 3.7.6, TS 3.7.7, TS 3.7.13, and TS 3.8.1.1 that remove an expired note are administrative, non-technical changes which remove temporary TS requirements added as part of the HNP License Amendment 153 issued on September 16, 2016, that are currently obsolete.

Therefore, operation of the facility in accordance with the proposed changes will not involve a significant reduction in the margin of safety.

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

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

U.S. Nuclear Regulatory Commission Page 29 of 29 Serial RA-19-0007 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.

5.0 ENVIRONMENTAL CONSIDERATION

S Duke Energy has determined that the proposed amendment would change a requirement with respect to use of a facility component located within the restricted area, as defined in 10 CFR

20. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released onsite, 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. NRC, Generic Letter 80-30, Clarification of the Term "Operable" as It Applies to Single Failure Criterion for Safety Systems Required by TS, dated April 10, 1980

2. NRC, Regulatory Guide 1.174, Revision 3, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,"

dated January 2018 (ADAMS Accession No. ML17317A256)

3. NRC, Regulatory Guide 1.177, Revision 1, An Approach For Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications, dated May 2011 (ADAMS Accession No. ML100910008)

4. NRC, Regulatory Guide 1.200, Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," dated March 2009 (ADAMS Accession No. ML090410014)

5. NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, dated March 2017 (ADAMS Accession No. ML17062A466)

U.S. Nuclear Regulatory Commission Serial RA-19-0007 SERIAL RA-19-0007 ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATIONS CHANGES SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63

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INSERT A

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

One charging/safety injection pump train is allowed to be inoperable for a total of 7 days to allow for maintenance on the Essential Services Chilled Water System and air handlers supported by the Essential Services Chilled Water System. Prior to exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the compensatory measures described in TS Bases 3.7.13 and HNP LAR correspondence letter RA-19-0007 shall be implemented.

EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - Tavg GREATER THAN OR EQUAL TO 350°F LIMITING CONDITION FOR OPERATION 3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:

a. One OPERABLE Charging/safety injection pump,

b. One OPERABLE RHR heat exchanger,

c. One OPERABLE RHR pump, and

d. An OPERABLE flow path capable of taking suction from the refueling water storage tank on a Safety Injection signal and, upon being manually aligned, transferring suction to the containment sump during the recirculation phase of operation.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s* or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

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

The A Train ECCS subsystem is allowed to be inoperable for a total of 14 days only to allow for the implementation of design improvements on the A Train ESW pump. The 14 days will be taken one time no later than October 29, 2016. During the period in which the A Train ESW pump supply from the Auxiliary Reservoir or Main Reservoir is not available, Normal Service Water will remain available and in service to supply the A Train ESW equipment loads until the system is ready for post maintenance testing. Allowance of the extended Completion Time is contingent on meeting the Compensatory Measures and Conditions described in HNP LAR submittal correspondence letter HNP-16-056.

Add 'INSERT B' SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a. At the frequency specified in the Surveillance Frequency Control Program by:

1. Verifying that the following valves are in the indicated positions with the control power disconnect switch in the "OFF" position, and the valve control switch in the "PULL TO LOCK" position:

SHEARON HARRIS - UNIT 1 3/4 5-3 Amendment No. 154

INSERT B

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

One ECCS subsystem train is allowed to be inoperable for a total of 7 days to allow for maintenance on the Essential Services Chilled Water System and air handlers supported by the Essential Services Chilled Water System. Prior to exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the compensatory measures described in TS Bases 3.7.13 and HNP LAR correspondence letter RA-19-0007 shall be implemented.

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INSERT C

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

- One Containment Spray System train is allowed to be inoperable for a total of 7 days to allow for maintenance on the Essential Services Chilled Water System and air handlers supported by the Essential Services Chilled Water System. Prior to exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the compensatory measures described in TS Bases 3.7.13 and HNP LAR correspondence letter RA-19-0007 shall be implemented.

CONTAINMENT SYSTEMS SPRAY ADDITIVE SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.2 The Spray Additive System shall be OPERABLE with:

a. A Spray Additive Tank containing a volume of between 3268 and 3768 gallons of between 27 and 29 weight % of NaOH solution, and

b. Two spray additive eductors each capable of adding NaOH solution from the chemical additive tank to a Containment Spray System pump flow.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With the Spray Additive System inoperable, restore the system to OPERABLE status within 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s* or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the Spray Additive System to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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

The Spray Additive System is allowed to be inoperable for a total of 14 days only to allow for the implementation of design improvements on the A Train ESW pump. The 14 days will be taken one time no later than October 29, 2016. During the period in which the A Train ESW pump supply from the Auxiliary Reservoir or Main Reservoir is not available, Normal Service Water will remain available and in service to supply the A Train ESW equipment loads until the system is ready for post maintenance testing. Allowance of the extended Completion Time is contingent on meeting the Compensatory Measures and Conditions described in HNP LAR submittal correspondence letter HNP-16-056.

SURVEILLANCE REQUIREMENTS 4.6.2.2 The Spray Additive System shall be demonstrated OPERABLE:

a. At the frequency specified in the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position;

b. At the frequency specified in the Surveillance Frequency Control Program by:

1. Verifying the contained solution volume in the tank, and

2. Verifying the concentration of the NaOH solution by chemical analysis.

c. At the frequency specified in the Surveillance Frequency Control Program by verifying that each automatic valve in the flow path actuates to its correct position on a containment spray or containment isolation phase A test signal as applicable; and

d. At the frequency specified in the Surveillance Frequency Control Program by verifying each eductor flow rate is between 17.2 and 22.2 gpm, using the RWST as the test source containing at least 436,000 gallons of water.

SHEARON HARRIS - UNIT 1 3/4 6-12 Amendment No. 154

CONTAINMENT SYSTEMS CONTAINMENT COOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.3 Four containment fan coolers (AH-1, AH-2, AH-3, and AH-4) shall be OPERABLE with one of two fans in each cooler capable of operation at low speed. Train SA consists of AH-2 and AH-3. Train SB consists of AH-1 and AH- 4.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With one train of the above required containment fan coolers inoperable and both Containment Spray Systems OPERABLE, restore the inoperable train of fan coolers to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With both trains of the above required containment fan coolers inoperable and both Containment Spray Systems OPERABLE, restore at least one train of fan coolers to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Restore both above required trains of fan coolers to OPERABLE status within 7 days of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With one train of the above required containment fan coolers inoperable and one Containment Spray System inoperable, restore the inoperable Spray System to OPERABLE status within 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s* or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Restore the inoperable train of containment fan coolers to OPERABLE status within 7 days of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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

The A Train containment fan coolers and the A Train Containment Spray System are allowed to be inoperable for a total of 14 days only to allow for the implementation of design improvements on the A Train ESW pump. The 14 days will be taken one time no later than October 29, 2016. During the period in which the A Train ESW pump supply from the Auxiliary Reservoir or Main Reservoir is not available, Normal Service Water will remain available and in service to supply the A Train ESW equipment loads until the system is ready for post maintenance testing. Allowance of the extended Completion Time is contingent on meeting the Compensatory Measures and Conditions described in HNP LAR submittal correspondence letter HNP-16-056.

Add 'INSERT D' SURVEILLANCE REQUIREMENTS 4.6.2.3 Each train of containment fan coolers shall be demonstrated OPERABLE:

a. At the frequency specified in the Surveillance Frequency Control Program by:

1. Starting each fan train from the control room, and verifying that each fan train operates for at least 15 minutes, and

2. Verifying a cooling water flow rate, after correction to design basis service water conditions, of greater than or equal to 1300 gpm to each cooler.

b. At the frequency specified in the Surveillance Frequency Control Program by verifying that each fan train starts automatically on a safety injection test signal.

SHEARON HARRIS - UNIT 1 3/4 6-13 Amendment No. 154

INSERT D

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

One train of containment fan coolers and one Containment Spray System train are allowed to be inoperable for a total of 7 days to allow for maintenance on the Essential Services Chilled Water System and air handlers supported by the Essential Services Chilled Water System. Prior to exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the compensatory measures described in TS Bases 3.7.13 and HNP LAR correspondence letter RA-19-0007 shall be implemented.

PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

a. Two motor-driven auxiliary feedwater pumps, each capable of being powered from separate emergency buses, and

b. One steam turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s* or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With two auxiliary feedwater pumps inoperable, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible. (NOTE: LCO 3.0.3 and all other LCO Required Actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status. Following restoration of one AFW train, all applicable LCOs apply based on the time the LCOs initially occurred.)

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

The A Train auxiliary feedwater pump is allowed to be inoperable for a total of 14 days only to allow for the implementation of design improvements on the A Train ESW pump. The 14 days will be taken one time no later than October 29, 2016. During the period in which the A Train ESW pump supply from the Auxiliary Reservoir or the Main Reservoir is not available, Normal Service Water will remain available and in service to supply the A Train ESW equipment loads until the system is ready for post maintenance testing. Allowance of the extended Completion Time is contingent on meeting the Compensatory Measures and Conditions described in HNP LAR submittal correspondence letter HNP-16-056.

SURVEILLANCE REQUIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:

a. At the frequency specified in the Surveillance Frequency Control Program by:

1. Demonstrating that each motor-driven pump satisfies performance requirements by either:

a) Verifying each pump develops a differential pressure that (when temperature - compensated to 70°F) is greater than or equal to 1514 psid at a recirculation flow of greater than or equal to 50 gpm (25 KPPH),

or b) Verifying each pump develops a differential pressure that (when temperature - compensated to 70°F) is greater than or equal to 1259 psid at a flow rate of greater than or equal to 430 gpm (215 KPPH).

SHEARON HARRIS - UNIT 1 3/4 7-4 Amendment No. 154

PLANT SYSTEMS 3/4.7.3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 At least two component cooling water (CCW) pumps*, heat exchangers and essential flow paths shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With only one component cooling water flow path OPERABLE, restore at least two flow paths to OPERABLE status within 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s** or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.3 At least two component cooling water flow paths shall be demonstrated OPERABLE:

a. At the frequency specified in the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position; and

b. At the frequency specified in the Surveillance Frequency Control Program by verifying that:

1. Each automatic valve servicing safety-related equipment or isolating non-safety-related components actuates to its correct position on a Safety Injection test signal, and

2. Each Component Cooling Water System pump required to be OPERABLE starts automatically on a Safety Injection test signal.

3. Each automatic valve serving the gross failed fuel detector and sample system heat exchangers actuates to its correct position on a Low Surge Tank Level test signal.

The breaker for CCW pump 1C-SAB shall not be racked into either power source (SA or SB) unless the breaker from the applicable CCW pump (1A-SA or 1B-SB) is racked out.

- The A Train component cooling water flow path is allowed to be inoperable for a total of 14 days only to allow for the implementation of design improvements on the A Train ESW pump.

The 14 days will be taken one time no later than October 29, 2016. During the period in which the A Train ESW pump supply from the Auxiliary Reservoir or Main Reservoir is not available, Normal Service Water will remain available and in service to supply the A Train ESW equipment loads until the system is ready for post maintenance testing. Allowance of the extended Completion Time is contingent on meeting the Compensatory Measures and Conditions described in HNP LAR submittal correspondence letter HNP-16-056.

SHEARON HARRIS - UNIT 1 3/4 7-11 Amendment No. 154

PLANT SYSTEMS 3/4.7.4 EMERGENCY SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4 At least two independent emergency service water loops shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With only one emergency service water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s* or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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

The A Train emergency service water loop is allowed to be inoperable for a total of 14 days only to allow for the implementation of design improvements on the A Train ESW pump. The 14 days will be taken one time no later than October 29, 2016. During the period in which the A Train ESW pump supply from the Auxiliary Reservoir or Main Reservoir is not available, Normal Service Water will remain available and in service to supply the A Train ESW equipment loads until the system is ready for post maintenance testing. Allowance of the extended Completion Time is contingent on meeting the Compensatory Measures and Conditions described in HNP LAR submittal correspondence letter HNP-16-056.

Add 'INSERT E' SURVEILLANCE REQUIREMENTS 4.7.4 At least two emergency service water loops shall be demonstrated OPERABLE:

a. At the frequency specified in the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position; and

b. At the frequency specified in the Surveillance Frequency Control Program by verifying that:

1. Each automatic valve servicing safety-related equipment or isolating non-safety portions of the system actuates to its correct position on a Safety Injection test signal, and

2. Each emergency service water pump and each emergency service water booster pump starts automatically on a Safety Injection test signal.

SHEARON HARRIS - UNIT 1 3/4 7-12 Amendment No. 154

INSERT E

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

The B Train emergency service water loop is allowed to be inoperable for a total of 7 days to allow for maintenance on the Essential Services Chilled Water System and air handlers supported by the Essential Services Chilled Water System. Prior to exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the compensatory measures described in TS Bases 3.7.13 and HNP LAR correspondence letter RA-19-0007 shall be implemented.

PLANT SYSTEMS 3/4.7.6 CONTROL ROOM EMERGENCY FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6 Two independent Control Room Emergency Filtration System (CREFS) trains shall be OPERABLE.*

APPLICABILITY: a. MODES 1, 2, 3, and 4

b. MODES 5 and 6

c. During movement of irradiated fuel assemblies and movement of loads over spent fuel pools ACTION:

a. MODES 1, 2, 3 and 4:

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

In addition to the Actions below, perform Action c. if applicable.

1. With one CREFS train inoperable for reasons other than an inoperable Control Room Envelope (CRE) boundary, restore the inoperable CREFS train to OPERABLE status within 7 days** or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

2. With one or more CREFS trains inoperable due to inoperable CRE boundary:

a. Initiate action to implement mitigating actions immediately or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ;

b. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , verify mitigating actions ensure CRE occupant radiological exposures will not exceed limits and that CRE occupants are protected from hazardous chemicals and smoke or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ;

c. Restore CRE boundary to OPERABLE within 90 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The control room envelope (CRE) boundary may be opened intermittently under administrative controls.

- The A CREFS train is allowed to be inoperable for a total of 14 days only to allow for the implementation of design improvements on the A Train ESW pump. The 14 days will be taken one time no later than October 29, 2016. During the period in which the A Train ESW pump supply from the Auxiliary Reservoir or Main Reservoir is not available, Normal Service Water will remain available and in service to supply the A Train ESW equipment loads until the system is ready for post maintenance testing. Allowance of the extended Completion Time is contingent on meeting the Compensatory Measures and Conditions described in HNP LAR submittal correspondence letter HNP-16-056.

SHEARON HARRIS - UNIT 1 3/4 7-14 Amendment No. 153

PLANT SYSTEMS 3/4.7.7 REACTOR AUXILIARY BUILDING (RAB) EMERGENCY EXHAUST SYSTEM LIMITING CONDITION FOR OPERATION 3.7.7 Two independent RAB Emergency Exhaust Systems shall be OPERABLE.*

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With one RAB Emergency Exhaust System inoperable, restore the inoperable system to OPERABLE status within 7 days** or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With two RAB Emergency Exhaust Systems inoperable due to an inoperable RAB Emergency Exhaust System boundary, restore the RAB Emergency Exhaust System boundary to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.7 Each RAB Emergency Exhaust System shall be demonstrated OPERABLE:

a. At the frequency specified in the Surveillance Frequency Control Program by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 15 continuous minutes with the heaters operating;

b. At the frequency specified in the Surveillance Frequency Control Program or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following significant painting, fire, or chemical release in any ventilation zone communicating with the system by:

1. Verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the unit flow rate is 6800 cfm +/- 10% during system operation when tested in accordance with ANSI N510-1980;

2. Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, has a methyl iodine penetration of 2.5% when tested at a temperature of 30°C and at a relative humidity of 70% in accordance with ASTM D3803-1989.

The RAB Emergency Exhaust Systems boundary may be opened intermittently under administrative controls.

- The A Train RAB Emergency Exhaust System is allowed to be inoperable for a total of 14 days only to allow for the implementation of design improvements on the A Train ESW pump. The 14 days will be taken one time no later than October 29, 2016. During the period in which the A Train ESW pump supply from the Auxiliary Reservoir or Main Reservoir is not available, Normal Service Water will remain available and in service to supply the A Train ESW equipment loads until the system is ready for post maintenance testing. Allowance of the extended Completion Time is contingent on meeting the Compensatory Measures and Conditions described in HNP LAR submittal correspondence letter HNP-16-056.

SHEARON HARRIS - UNIT 1 3/4 7-17 Amendment No. 156

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ELECTRICAL POWER SYSTEMS A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION ACTION (Continued):

3. Restore the diesel generator to OPERABLE status within 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s** or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; and

4. Verify required feature(s) powered from the OPERABLE diesel generator are OPERABLE. If required feature(s) powered from the OPERABLE diesel generator are discovered to be inoperable at any time while in this condition, restore the required feature(s) to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from discovery of inoperable required feature(s) or declare the redundant required feature(s) powered from the inoperable A.C. source as inoperable.

c. With one offsite circuit and one diesel generator of 3.8.1.1 inoperable:

NOTE: Enter applicable Condition(s) and Required Action(s) of LCO 3/4.8.3, ONSITE POWER DISTRIBUTION - OPERATING, when this condition is entered with no A.C. power to one train.

1. Restore one of the inoperable A.C. sources to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

2. Following restoration of one A.C. source (offsite circuit or diesel generator),

restore the remaining inoperable A.C. source to OPERABLE status pursuant to requirements of either ACTION a or b, based on the time of initial loss of the remaining A.C. source.

- The A diesel generator is allowed to be inoperable for a total of 14 days only to allow for the implementation of design improvements on the A Train ESW pump. The 14 days will be taken one time no later than October 29, 2016. During the period in which the A Train ESW pump from the Auxiliary Reservoir or Main Reservoir is not available, Normal Service Water will remain available and in service to supply the A Train ESW equipment until the system is ready for post maintenance testing. Allowance of the extended Completion Time is contingent on meeting the Compensatory Measures and Conditions described in HNP LAR submittal correspondence HNP-16-056.

SHEARON HARRIS - UNIT 1 3/4 8-2 Amendment No. 153

U.S. Nuclear Regulatory Commission Serial RA-19-0007 SERIAL HNP-19-0007 ATTACHMENT 3 PROPOSED TECHNICAL SPECIFICATIONS BASES CHANGES SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63

3/4.7 PLANT SYSTEMS BASES 3/4.7.3 COMPONENT COOLING WATER SYSTEM The OPERABILITY of the Component Cooling Water System ensures that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analyses.

3/4.7.4 EMERGENCY SERVICE WATER SYSTEM The OPERABILITY of the Emergency Service Water System ensures that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analyses.

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

A one-time change to TS 3.7.4 extends the action statement completion time from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 14 days in order to replace the 'A' ESW pump. This change also affects TS 3.1.2.4, Charging Pumps - Operating, TS 3.5.2, ECCS Subsystems - Tavg Greater Than or Equal To 350°F, TS 3.6.2.1, Containment Spray System, TS 3.6.2.2, Spray Additive System, TS 3.6.2.3, Containment Cooling System, TS 3.7.1.2, Auxiliary Feedwater System, TS 3.7.3, Component Cooling Water System, TS 3.7.4, Emergency Service Water System, TS 3.7.6, Control Room Emergency Filtration System, TS 3.7.7, Reactor Auxiliary Building (RAB)

Emergency Exhaust System, TS 3.7.13, Essential Services Chilled Water System, and TS 3.8.1.1, AC Sources - Operating.

A note similar to the following is placed in each of the above listed TS:

The 'A' Train emergency service water loop is allowed to be inoperable for a total of 14 days only to allow for the implementation of design improvements on the 'A' Train ESW pump. The 14 days will be taken one time no later than October 29, 2016. During the period in which the 'A' Train ESW pump supply from the Auxiliary Reservoir or Main Reservoir is not available, Normal Service Water will remain available and in service to supply the A' Train ESW equipment loads until the system is ready for post maintenance testing. Allowance of the extended Completion Time is contingent on meeting the Compensatory Measures and Conditions described in HNP LAR submittal correspondence letter HNP-16-056.

SHEARON HARRIS - UNIT 1 B 3/4 7-3 Amendment No. 153

3/4.7 PLANT SYSTEMS BASES 3/4.7.4 EMERGENCY SERVICE WATER SYSTEM (Continued)

CONDITIONS ASSOCIATED WITH ONE TIME TS CHANGE 1 Normal Service Water (NSW) will remain available and in service for the duration of the allowed outage time (AOT) to support operation of the A Emergency Diesel Generator if required. OP-155, Diesel Generator Emergency Power System, Section 5.1.2, EDG Control Room Manual Start, step 2 says VERIFY service water flow has been established to the EDG per OP-139. OP-139, Section 5.3, Supplying Both ESW [Emergency Service Water] Headers with NSW/Securing ESW Pump, requires the NSW header in service and the ESW header filled and vented per Section 8.24, which would align Service Water to the EDG.

This condition is an assumption in the risk metric calculations for the AOT.

2 The 'B' Train ESW will remain operable. OWP-SW, Service Water, includes component lineups necessary when an ESW pump is inoperable that provides defense-in-depth for prevention of core damage and containment failure. The lineup steps for time periods when the A ESW pump is inoperable include the lifting of leads to disable the Safety Injection (SI) close signal to service water valve 1SW-39 and service water valve 1SW-276. This allows the breakers to be maintained on and allows expeditious isolation capability in the event of a SW leak in the Reactor Auxiliary Building (RAB). This lineup also defeats the SI signal to service water valve 1SW-276 to maintain it open. As long as service water valves 1SW-274 and 1SW-40 are operable, the B Train ESW header is isolable and operable.

3 In accordance with OMM-001, Operations Administrative Requirements, the following equipment is posted protected by Operations when A ESW pump is unavailable: Switchyard (Breakers 52-1, 52-2, 52-3 and Line Panels 5, 6, and 7), B ESW pump and breaker, B-Train Process Instrumentation Control (PIC) cabinets (PIC 2, 4, 10, 14, and 18), and the A Start-up Transformer.

This condition is an assumption in the risk metric calculations for the AOT.

4 Prior to the AOT entry, the weather forecast will be reviewed for any forecasted weather that could affect the availability of offsite power. The outage will not commence if weather conditions are predicted that could adversely affect the availability of offsite power. WCM-001, On-line Maintenance Risk Management, requires review of the weather forecast prior to the beginning of this maintenance outage.

This condition is an assumption in the risk metric calculations for the AOT.

SHEARON HARRIS - UNIT 1 B 3/4 7-3a Amendment No. 153

3/4.7 PLANT SYSTEMS BASES 3/4.7.4 EMERGENCY SERVICE WATER SYSTEM (Continued)

CONDITIONS ASSOCIATED WITH ONE TIME TS CHANGE 5 The opposite train or critical equipment listed below and supporting components will be posted protected:

x EDGs (both A and B EDGs) x NSW Pumps and power supplies (both A and B NSW Pumps) x Dedicated Shutdown Diesel Generator x Alternate Seal Injection Pump x Turbine Driven Auxiliary Feedwater (AFW) Pump x B ESW Pump Quantitative credit has been taken in the risk metric calculations for this condition.

6 Continuous fire watches in risk critical areas will be instituted on the protected train, which will include the following rooms:

x B Electrical Switchgear Room x B Cable Spread Room x B Battery Room Quantitative credit has been taken in the risk metric calculations for this condition.

7 Restrictions will remain in place on hot work and transient combustibles in the following rooms:

x B Electrical Switchgear Room x B Cable Spread Room x B Battery Room Qualitative Risk Impact.

8 Operators will be briefed on the procedures and guidance for the equipment lineup necessary for the proposed AOT activity.

Quantitative credit has been taken in the risk metric calculations for this condition.

9 Operators will be briefed to improve operator response for ASI System actions.

Quantitative credit has been taken in the risk metric calculations for this condition.

SHEARON HARRIS - UNIT 1 B 3/4 7-3b Amendment No. 153

3/4.7 PLANT SYSTEMS BASES 3/4.7.4 EMERGENCY SERVICE WATER SYSTEM (Continued)

CONDITIONS ASSOCIATED WITH ONE TIME TS CHANGE 10 The B ESW pump discharge pressure transmitter will be calibrated within three months prior to the proposed AOT.

Quantitative credit has been taken in the risk metric calculations for this condition.

11 The B ESW pump discharge strainer differential pressure will be checked when the B ESW pump is in service and a backwash will be completed to verify it is clean within one month prior to the proposed AOT. This will ensure that the strainer is clean and capable of performing its duty during the AOT.

Qualitative Risk Impact.

12 Switchgear Room in Turbine Building 286 will be protected, in order to minimize the risk to NSW power supplies.

Qualitative Risk Impact.

13 Restrictions will be in place on switchyard work or other maintenance and testing that could cause a plant trip for the duration of the AOT. Additionally, the system load dispatcher will be contacted once per day to ensure no significant grid perturbations are expected during the extended AOT.

Qualitative Risk Impact.

14 The FLEX ESW pump will be pre-staged in advance of the AOT entry to allow for connection to the A Train ESW header, to provide alternate cooling to the A EDG in the event of a loss of offsite power (LOOP). Dedicated personnel will be available to make the necessary equipment manipulations such that the A EDG will be started within approximately one hour of the LOOP. The A EDG will be manually started and operations will energize the necessary loads to perform the safety function of decay heat removal in the event of a LOOP.

Quantitative Risk Impact.

SHEARON HARRIS - UNIT 1 B 3/4 7-3c Amendment No. 153

3/4.7 PLANT SYSTEMS BASES 3/4.7.4 EMERGENCY SERVICE WATER SYSTEM (Continued)

CONDITIONS ASSOCIATED WITH ONE TIME TS CHANGE 15 All associated B Train equipment for the Technical Specifications (TS) listed below, which are the only operable trains, are to be protected during the extended AOT.

TS 3.1.2.4, Charging Pumps - Operating TS 3.5.2, ECCS Subsystems - Tavg Greater Than or Equal To 350°F TS 3.6.2.1, Containment Spray System [CSS]

TS 3.6.2.2, Spray Additive System TS 3.6.2.3, Containment Cooling System [CCS]

TS 3.7.1.2, Auxiliary Feedwater [AFW] System TS 3.7.3, Component Cooling Water [CCW] System TS 3.7.4, Emergency Service Water System [ESWS]

TS 3.7.6, Control Room Emergency Filtration System [CREFS]

TS 3.7.7, Reactor Auxiliary Building [RAB] Emergency Exhaust System TS 3.7.13, Essential Services Chilled Water System [ESCWS]

TS 3.8.1.1, AC Sources - Operating 16 The Demineralized Water Storage Tank will be maintained between 29 and 34 feet for the duration of the AOT.

17 The following actions will be taken prior to and during the proposed AOT as described:

x EDG cooling flow will be verified prior to the AOT entry.

x B EDG loading and operational check will be completed prior to the AOT entry.

x B ESW pump operational check will be completed prior to the AOT entry.

x Proceduralized EDG inspections and checks will be performed daily for reliability during the AOT, which are normally completed weekly.

x Freeze protection equipment as required and ventilation in the intake buildings will be verified as functional prior to the AOT.

x Position of low head safety injection recirculation to Refueling Water Storage Tank isolation valves, 1SI-448 and 1SI-331, will be verified prior to the AOT, in addition to other SW valves that will support the clearance for the pump replacement.

SHEARON HARRIS - UNIT 1 B 3/4 7-3d Amendment No. 153

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INSERT A

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

The TS 3.7.13 action statement completion time of 7 days is for maintenance on the Essential Services Chilled Water System (ESCWS). Entry into this action statement also affects TS 3.1.2.4, Charging Pumps - Operating, TS 3.5.2, ECCS [Emergency Core Cooling Systems]

Subsystems - Tavg Greater Than or Equal To 350°F, TS 3.6.2.1, Containment Spray System, TS 3.6.2.3, Containment Cooling System, and TS 3.7.4, "Emergency Service Water System,"

for B Train ESCWS inoperability only, based upon the impact to the B Emergency Service Water (ESW) Booster Pump operability. The B Train ESW booster pump area is cooled by AH-8 1X-SB, which is powered by the B Train power supply. There is no impact to the A Train ESW Booster Pump or the A Train ESW System since an air handler unit for this area may be powered by either A or B Train power supplies.

A note similar to the following is placed in each of the above listed TS:

One Train of [Applicable TS or TS System] is allowed to be inoperable for a total of 7 days to allow for maintenance on the Essential Services Chilled Water System and air handlers supported by the Essential Services Chilled Water System. Prior to exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the compensatory measures described in TS Bases 3.7.13 and HNP LAR correspondence letter RA-19-0007 shall be implemented.

COMPENSATORY MEASURES FOR 7-DAY ALLOWED OUTAGE TIME 1 The following equipment and the corresponding power supplies will be posted protected:

x Air handling units for the operable charging safety injection pump (CSIP) areas:

AH-9A (CSIP 1A-SA Area), AH-9B (CSIP 1B-SB Area), or AH-10 (CSIP 1C-SAB Area) x Air handling units for the Switchgear Rooms with operable equipment:

AH-12 1A-SA and AH-12 1B-SA supply Switchgear Room A; AH-13 1A-SB and AH-13 1B-SB supply Switchgear Room B x Operable ESCWS chiller and operable chilled water pump 2 The Fire Protection tracking log will be reviewed for fire hazards and fire impairments.

Transient combustibles and hot work in these fire risk-sensitive areas will be limited:

x Fire compartment FC25: RAB HVAC Room (MCC 1A21-SA, MCC 1A31-SA) x Fire compartments FC34 and FC35: Switchgear Room A and Switchgear Room B x Fire compartment FC41: Turbine Building (Zone 1-G-261 - 6.9 kV Switchgear) x Fire compartment FC54: Transformer Yard 3 Restrictions on work activities will be in place that involve components that if lost or failed could result in a direct plant trip or transient.

4 Operator actions for CSIP area cooling, Switchgear Room cooling, and Auxiliary Relay Cabinet Room cooling, if needed following a loss of HVAC, will be briefed with Operations. The fan used for the CSIP area cooling will be pre-staged and verified to be functional.

INSERT A (continued)

COMPENSATORY MEASURES FOR 7-DAY ALLOWED OUTAGE TIME 5 Discretionary maintenance or discretionary testing on equipment that support the following systems will be avoided for the remaining duration of the TS 3.7.13 LCO entry:

x Essential Services Chilled Water System (operable train) x Motor-Driven and Turbine-Driven Auxiliary Feedwater Pumps x ESW System and Normal Service Water System x Emergency Diesel Generators x Alternate Seal Injection System and the Dedicated Shutdown Diesel Generator

U.S. Nuclear Regulatory Commission Serial RA-19-0007 SERIAL RA-19-0007 ATTACHMENT 4 ATTACHMENT 4 OF PLP-114, RELOCATED TECHNICAL SPECIFICATIONS AND DESIGN BASIS REQUIREMENTS; TABLE OF AREAS SERVED BY ESF FAN COOLERS IN THE RAB; AND ESCWS FLOW DIAGRAMS SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63

Attachment 4 Sheet 1 of 3 Area Temperature Monitoring 1.0 OPERATIONAL REQUIREMENTS 1.1 The temperature of each area shown in Table A shall not be exceeded for more than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or by more than 30°F.

APPLICABILITY: Whenever the equipment in an affected area is required to be functional.

ACTION:

a. With one or more areas exceeding the temperature limit(s) shown in Table A for more than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , prepare within 30 days an evaluation to demonstrate the continued functionality of the affected equipment.

b. With one or more areas exceeding the temperature limit(s) shown in Table A by more than 30°F, prepare an evaluation as required by Action a. above and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either restore the area(s) to within the temperature limit(s) or declare the equipment in the affected area(s) non-functional.

2.0 SURVEILLANCE REQUIREMENTS 2.1 The temperature in each of the areas shown in Table A shall be determined to be within its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

PLP-114 Rev. 028 Page 16 of 36

Attachment 4 Sheet 2 of 3 Area Temperature Monitoring TABLE A MAXIMUM AREA TEMPERATURE LIMIT (°F)

REACTOR AUXILIARY BUILDING

1. Control Room Envelope (El 305') 75

2. Process I&C, Room (El 305') 80

3. Rod Control Cabinets Area (El 305') 104

4. Auxiliary Relay Cabinet Room (El 305')* 80

5. AH-15 Ventilation Room (El 305')* 104

6. A&B Battery Rooms (El 286') 85***

7. A&B Switchgear Rooms (El 286') 88****

8a. Process I&C Room A (El 286')* 85 8b. Process I&C Room B (El 286')* 85

9. Auxiliary Transfer Panel Room (El 286')* 104

10. Auxiliary Control Panel Room (El 286)* 88

11. Main Steam, Feedwater Pipe Tunnel 122

12. SA&SB Electrical Penetration Areas (El 261' & 286') 104

13. E-6 Rooms (El 261')* 104

14. Area with MCC 1A35SA and 1B35SB (El 261') 104

15. HVAC Chillers, Auxiliary FW Piping & Valve Area (El 261') 104

16. CCW Pumps, CCW Hx, Auxiliary FW Pumps Area (El 236') 104

17. 1A-SA, 1B-SB, and 1C-SAB Charging Pump Rooms (El 236') 104**

18. Service Water Booster Pump 1B-SB (El 236') 104

19. Mechanical and Electrical Penetration Areas (El 236') 104

20. Containment Spray Additive Tank, and H&V Equipment Area (El 216') 104

21. Trains A&B Containment Spray Pump, RHR Pump, H&V Equipment Areas (El 190') 104 See Notes on next page PLP-114 Rev. 028 Page 17 of 36

Attachment 4 Sheet 3 of 3 Area Temperature Monitoring TABLE A MAXIMUM AREA TEMPERATURE LIMIT (°F)

FUEL HANDLING BUILDING

22. Trains A&B Emergency Exhaust System Areas (El 261') 104

23. Spent Fuel Pool Cooling Pump Room (El 236) 115.5 WASTE PROCESSING BUILDING

24. H&V Equipment Room (El 236') 104 MISCELLANEOUS

25. Tank Area (El 236') 104

26. Diesel Fuel Oil Storage Building (El 242') 122

27. Emergency Service Water Electrical Equipment Room 116

28. Emergency Service Water Pump Room 122

29. 1A-SA & 1B-SB H&V Equipment Rooms (El 292') 122

30. 1A-SA & 1B-SB H&V Equipment Rooms (El 280') 118

31. 1A-SA & 1B-SB Electrical Rooms (El 261') 116

32. 1A-SA & 1B-SB Diesel Generator Rooms (El 261') 120 Notes:

Areas 4, 5, 8, 9, 10, and 13 were added per PNSC Meeting 92-36.

- An Engineering Disposition has been performed regarding the selection of these setpoints. Reference ED ESR 00-00137 for additional information. This note added per AR# 3744.

- - Battery Room temperature of 85°F was established per EC 402748.

- - - Ambient temperature of up to 104°F in the A & B Switchgear Room is acceptable whenever ventilation is not available during maintenance outage activities. (Ref. EC 278851)

PLP-114 Rev. 028 Page 18 of 36

1 Table of Areas Served by Engineered Safety Feature (ESF) Fan Coolers in the Reactor Auxiliary Building (RAB)

Area HVAC Unit Location Redundant TS LCO entry Safety train Related cooling Start Signal capability Containment spray pump AH-5 1A-SA RAB 190 No TS 3.6.2.1, Containment Spray SIAS and and RHR pump area AH-5 1B-SB elevation System Undervoltage TS 3.5.2, ECCS Subsystems - Tavg Greater Than or Equal To 350°F Emergency Service Water AH-6 1A-SA RAB-236 Yes n/a SIAS and (ESW) Booster Pump SA AH-6 1B-SB elevation Undervoltage area AH-7 1A-SA AH-7 1B-SB Component Cooling Water Pump and Auxiliary Feedwater Pump areas ESW Service Water AH-8 1X-SB RAB-236 No TS 3.7.4, "Emergency Service Water SIAS and Booster Pump SB area elevation System," B train only Undervoltage Charging safety injection AH-9 1A-SA RAB 236 No TS 3.1.2.4,Charging Pumps - SIAS and pump (CSIP) area: CSIP elevation Operating Undervoltage 1A-SA TS 3.5.2, ECCS Subsystems - Tavg Greater Than or Equal To 350°F CSIP 1B-SB area AH-9 1B-SB No TS 3.1.2.4,Charging Pumps - SIAS and Operating Undervoltage TS 3.5.2, ECCS Subsystems - Tavg Greater Than or Equal To 350°F CSIP 1C-SAB area AH-10 1A-SA Yes n/a SIAS and AH-10 1B-SB Undervoltage

2 Table of Areas Served by Engineered Safety Feature (ESF) Fan Coolers in the Reactor Auxiliary Building (RAB)

Area HVAC Unit Location Redundant TS LCO entry Safety train Related cooling Start Signal capability Mechanical Penetration AH-11 1A-SA RAB 236 Yes n/a SIAS and area and Electrical AH-11 1B-SB elevation Undervoltage Penetration area Switchgear A areas: AH-12 1A-SA RAB 286 Yes n/a SIAS and Process Instrumentation elevation Undervoltage Control (PIC) Room A A Battery Room AH-12 1B-SA RAB 286 Yes n/a none Auxiliary Control Panel elevation Room Heating & Ventilation Equipment Room Electrical Penetration areas A Auxiliary Transfer Panel Room Switchgear B areas: AH-13 1A-SB RAB 286 Yes n/a SIAS and PIC Room B elevation Undervoltage B Battery Room Rod Control Cabinet Room AH-13 1B-SB RAB 286 Yes n/a none Auxiliary Control Panel elevation Room Heating & Ventilation Equipment Room Electrical Penetration areas B Auxiliary Transfer Panel Room

3 Table of Areas Served by Engineered Safety Feature (ESF) Fan Coolers in the Reactor Auxiliary Building (RAB)

Area HVAC Unit Location Redundant TS LCO entry Safety train Related cooling Start Signal capability Control Room AH-15 1A-SA RAB 305 Yes n/a SIAS and AH-15 1B-SB elevation Undervoltage PIC Room AH-16 1A-SA RAB 305 Yes n/a SIAS and Auxiliary Relay Cabinet AH-16 1B-SB elevation Undervoltage Room Repair Shop Spaces HVAC Chillers, AFW AH-19 1A-SA RAB 261 Yes n/a SIAS and Piping, and Valve Pump AH-19 1B-SB elevation Undervoltage area AH-20 1A-SA AH-20 1B-SB MCC 1A22-SA for AH-23-1X-SA RAB-236 No TS 3.6.2.3, Containment Cooling SIAS and Containment Cooling elevation System" Undervoltage System Electrical Penetration area AH-24 1X-SA RAB 261 No Temperatures are monitored in SIAS and elevation accordance with PLP-114 limits. Undervoltage Electrical Penetration area AH-25 1X-SB RAB 261 No Temperatures are monitored in SIAS and elevation accordance with PLP-114 limits. Undervoltage H&V Equipment; AH-26 1A-SA RAB 261 No TS 3.7.7, Reactor Auxiliary Building SIAS and RABEES supply power AH-26 1B-SB elevation Emergency Exhaust System (7-day Undervoltage area LCO) impacted but no extension (E-6 Rooms) needed Containment spray tank, AH-28-1A-SA RAB 216 Yes n/a SIAS and boron injection tank and AH-28-1B-SB elevation Undervoltage pump area

4 Table of Areas Served by Engineered Safety Feature (ESF) Fan Coolers in the Reactor Auxiliary Building (RAB)

Area HVAC Unit Location Redundant TS LCO entry Safety train Related cooling Start Signal capability MCC 1B22-SB for AH-29 1X-SB RAB 236 No TS 3.6.2.3, Containment Cooling SIAS and Containment Cooling elevation System" Undervoltage System and Instrument Rack area RAB MCC 1A35-SA area AH-92 1A-SA RAB 286 Yes n/a SIAS and AH-92 1B-SB elevation Undervoltage RAB MCC 1B35-SB area Rod Control Cabinet Room AH-93 1X-SA RAB 305 Yes: AH-13 n/a SIAS and elevation fans Undervoltage provide cooling also

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 A A CAR-2168 G-498S02 B VENT B 3SW-V908SA-1 3

3SW1-934SA-1 3SW-H809SA-1 3SW-R16SA-1 3SW1-811SA-1 PI 7SW1-904-1 9207 A PI PX 9206 3SW1-853SA-1 ED S A

VENT C CS 3SW-S300SA-1 SEE NOTE 3 S C 3SW-H846SA-1 3SW3/4-837SA-1 SS SEE NOTE 7 SS SS P7 (1A-SA) 3SW-V801SA-1 3SW-V810SA-1 SS 3SW-V800SA-1 3SW-B301SA-1 CS CS CS 3SW1-812SA-1 3SW1-807SA-1 SS CS 3SW1-936SA-1 3SW1-852SA-1 3SW-B302SA-1 3SW 8- 800SA- 1 3SW1-851SA-1 3SW-V929SA-1 3SW-V845SA-1 SEE NOTE 7 3SW-V844SA-1 DR DR DR D 3SW6-875SA-1 ED 7SW3/4-813-1 D

7SW3-906-1 FT 3SW11/4-1653SA-1 9209 RELIEF TO OUTSIDE A SA ATMOSPHERE FE 3SW11/2-902SA-1 9207 A 3SW-V902SA-1 3SW 8- 801SA- 1 S 3SW3/4-900SA-1 3SW3/4-905SA-1 7SW11/2-901-1 PG PT 3SW-H900SA-1 3SW-H905SA-1 9425 9209 E 3SW11/2-814SA-1 A A 3SW-H64SA-1 E

3SW-H63SA-1 SA NOTES:

3SW-V805SA-1 3SW-V812SA-1 SA 3SW1-850SA-1 7SW-J9-1 1. FOR GENERAL NOTES AND DETAILS 3SW1-802SA-1 3SW1-118SA-1 7SW11/2-903-1 3SW-V843SA-1 SEE DWG CAR-2168-G-509 S01 3SW-H1009SA-1 ED ED

3SW-H842SA-1 PI 2. FOR LEGEND AND SYMBOLS 3SW3/4-1630SA-1 PI 3SW1-849SA-1 PI 9210 SEE DWG CAR-2168-G-528 S02 9208 3 TI A A CS V PCV 3. TEMPORARY STRAINERS USED TI DR 9210 S SS S A 9209 FOR INITIAL FLUSHING, ARE TO VENT 9208 3SW-V808SA-1 S A 3SW1-854SA-1 A SA BE REMOVED BEFORE START-UP.

CS SS SS S 3SW1-803SA-1 3SW-B300SA-1 SS SS V SS CS EH 4. FOR REFERENCE DWGS SEE F F 3SW-V847SA-1 CS NOTE 7 SW SW CS V CS CONDENSER DWG CAR-2168-G-498 CAR-2165-G-047 3SW12-83SA-1 3SW12-84SA-1 CAR-2165-G-047 (2500 GPM)

(I4) SS 3SW-V850SA-1 CS F.O.

5. HOSES WILL BE USED BETWEEN THE DRAIN (J7) 3SW-S32SA-1 SS TEMP. STRNR.

CS 3SW12-84SA-1 LINES AND EQUIPMENT DRAINS WHEN NECESSARY.

CS 3SW3/4-859SA-1 7SW3/4-861-1 7SW3/4-860-1 6. (*) SYMBOL INDICATES VENDOR SUPPLIEDITEM.

SS 3SW1-810SA-1 3SW-V851SA-1 3 3SW10-938SA-1 3

3SW-V804SA-1 7. VALVE IS STAINLESS STEEL.

3CH3/4-486SA-1 3CX-V90SA-1 COMPRESSOR 3 3SW3/4-862SA-1 3CX1-122SA-1 PI TI

TE TI PI 3CH-R1SA-1 3CX1-145SA-1 VENT 9423 9421 COND. REFRIGERANT 9433 9430 9430 3 3CX-V88SA-1 TRANSFER SYSTEM 3CX-V493SA-1 A A SS A A A 7CH1-1103-1 SA S S S S CS G 3CX1-123SA-1 3CX-V105SA-1 3CH3/4-487SA-1 SS 3CH-V75SA-1 ED G

3CX1-2310SA-1 V CS V 3CH3/4-24SA-1 3CX1-170SA-1 3CH-V425SA-1 CH 3CH-V427SA-1 COOLER 3CH10-91SA-1 CAR-2168-G-498 3CX10-168SA-1 (1298 GPM) 3CH-B5SA-1 (A1) 3CX-B7SA-1 3CX1-256SA-1 ESCW MAKE-UP 3CH2-96SA-1 3CH1-43SA-1 3CH1-95SA-1 3SW-V870SA-1 TANK (1A-SA) 3CX-V162SA-1 3CH-V39SA-1 3CH-V426SA-1 3CH-V76SA-1 3CX1-2311SA-1 3 DR 3CH-V77SA-1 3CX1-282SA-1 3 3 3CX-V492SA-1 DR 3CX1-171SA-1 7CH2-1003-1 3CX-V89SA-1 S 3CH1-97SA-1 H 7CX11/2-1007-1 3CX11/2-120SA-1 3SW-V868SA-1 V 3CX-V106SA-1 3CH-V74SA-1 7CH-V1026-1 H

3CX-V166SA-1 3CX3/4-482SA-1 SEE 3CX-V91SA-1 3CX1-121SA-1 3CX-V362SA-1 SEE FE NOTE 5 7CH11/2-1107-1 FT NOTE 5 9422 FUNNEL 3CX11/2- 119SA- 1 CS 9429 3SW1-863SA-1 A S SS A 3CX3/4-483SA-1 7CH11/2-1104-1 7CH-V1027-1 SA 3CX1-169SA-1 3 V 3CH11/2-99SA-1 7CH-V1001-1 ED 3CX-V104SA-1 3CX-V363SA-1 3CX-R6SA-1 7CH-V1028-1 PI 3CH11/2-98SA-1 3 7CH2-1004-1 9422 7CH1-1034-1 A S WATER CHILLER WC-2 I DR (1A-SA) CHEMICAL INCORPORATED: EC# 301132 I

7CX1-1008-1 3CX3/4-8SA-1 LG ELECTRONICALLY 3 7CX1-1104-1 ADDITION 34 (G10) 752 TONS SIGNED 3CX-V167SA-1 9431 3CX-R2SA-1 TANK 7CX1-1105-1 A 3CX1-281SA-1 (1A-NNS) INCORPORATED: EC# 95241 ELECTRONICALLY 33 (F5,F6,F12,G9,G10,H5) SIGNED 7CX2-1004-1 7DW-H148-1 P4 (1A-SA)

INCORPORATED: EC# 92230 ELECTRONICALLY 7DW1-627-1 3CX1-127SA-1 7CX-V1101-1 32 (C7,E6,E8,E11,E12,E13,F7,F8,F10,F11,F13) 7DWf-638-1 SIGNED BY TANK 7DW1-627-1 3CX1-124SA-1 V MANUFACTURER ELECTRONICALLY 7CX-V1001-1 7CX1-2302-1 31 INCORPORATED: EC# 86759 (C7,C12)

SIGNED 7DW-S4-1 3CX-V2280SA-1 ED 7DW-V513-1 7DW-P39-1 3CX-V2281SB-1 PX J S S S 3CX-V87SA-1 30 INCORPORATED: EC# 79156 (C9,C11)

ELECTRONICALLY SIGNED J

7DW-V514-1 ED DEMIN. WATER SUPPLY 3CX-S1SA-1 CAR-2165-G-049S02 7CX2-1003-1 SEE NOTE 3 ELECTRONICALLY (C4) 7DW-V516-1 29 INCORPORATED: EC# 73410 (F13,17,F18)

SIGNED 7DW-V512-1 PI 3CX-V86SA-1 3CX-V2284SA-1 3CX1-264SA-1 7DW-V515-1 7DW-V517-1 9421 3 DR ELECTRONICALLY A S 7DW-V511-1 7DW-H149-1 3CX-V170SA-1 3 3CX-V85SA-1 28 INCORPORATED: EC# 70557 (H5)

SIGNED D

7DWf-639-1 3CX2-126SA-1 3CX-B5SA-1 ELECTRONICALLY 3CX1-125SA-1 INCORPORATED: EC# 65194 7DW-V518-1 L.O. 27 (D1-D5, E1-E5, F1-F6, G2-G5, H3-H5) SIGNED 3CX1-2455SA-1 CX 7DW1/2-626-1 CAR-2168-G-498 3CX-V93SA-1 3CX10-118SA-1 K 7DW1-628-1 (M1) 26 INCORPORATED: EC# 61384 (E3, F1-F3) & EC# 60748 (C8)

ELECTRONICALLY SIGNED K

MECH HVAC 3CX11/2-119SA-1 25 3-3-05 INCORPORATED: EC# 60417 (F3) JAP RPP CRW INCORPORATED: EC# 58679 24 10-19-04 (F3, G3, I2-I4, J2-J4)

JAP RPP CRW INCORPORATED: ESR 94-00164 23 11-16-00 JWB WFL RPP FOR DLS (E10,E13,F9,F10)

L 22 11-6-96 INCORPORATED IN THIS REVISION:

RPP LLS DLS DLS L ESR-9500717 (D9,D10,E8-E10)

INCORPORATED IN THIS REVISION:

21 1-12-94 LLS RPP MMP WAS PCR-6411 (F13,F15)

INCORPORATED IN THIS REVISION:

20 11-3-93 GRS LLS MMP WAS PCR-2512 FR09 (G5)

INCORPORATED IN THIS REVISION:

19 12-1-92 TRW RPP GOW AMW PCR-5534(J1-J5)

THIS DRAWING HAS BEEN ELECTRONICALLY REDRAWN WITH M 18 11-9-92 DESIGN CHANGE M LLS RPP GOW MMP INCORPORATED: PCR-2512 (E1-E4, F1-F5, G1, G2)

REV DATE DESCRIPTION DWN CHK DPE DPPE THIS DRAWING HAS BEEN PRODUCED AND IS CONTROLLED ON THE NED CAD/E SYSTEM. ANY REVISIONS TO THIS DOCUMENT SHOULD BE DONE USING THE CAD/E SYSTEM TO ASSURE PROPER CONTROL OF THE ELECTRONIC DRAWING DATABASE.

NUCLEAR SAFETY RELATED (Q-LIST)

YES NO PARTIAL F.P.-Q R.W.-Q N N DUKE ENERGY NUCLEAR ENGINEERING DEPARTMENT - RALEIGH, N.C.

PLANT: SCALE:

HARRIS NUCLEAR PROJECT - UNIT 1 NONE TITLE: DWG NO.

HVAC ESSENTIAL SERVICES CHILLED WATER CONDENSER CAR-2168 FLOW DIAGRAM UNIT 1-SA G-498S02 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 A A CAR-2168 G-499S02 B VENT B 3SW-V909SB-1 3SW-V806SB-1 3 3SW1-935SB-1 3SW1-816SB-1 3SW-H816SB-1 3SW-S301SB-1 PI SEE NOTE 3 3SW-R17SB-1 3SW1-820SB-1 9207 7SW1-907-1 B PI PX 9206 3SW-H817SB-1 S B VENT 3SW3/4-838SB-1 C ED SS S 3SW1-821SB-1 C

SS CS CS 3SW-H838SB-1 SS SS CS P7(1B-SB) 3SW-B304SB-1 CS SS CS 3SW1-845SB-1 3SW-B305SB-1 CS SS CS 3SW-V821SB-1 CS SS CS 3SW 8- 804SB- 1 3SW1-937SB-1 3SW-V841SB-1 3SW-V930SB-1 3SW1-844SB-1 DR 3SW-V837SB-1 DR DR 7SW3/4-808-1 D ED D

3SW1-848SB-1 3SW6-876SB-1 RELIEF TO OUTSIDE FT ATMOSPHERE 9209 B

3SW11/4-1654SB-1 3SW11/2-912SB-1 SB 7SW3-910-1 3SW 8- 805SB- 1 FE 3SW-V904SB-1 3SW-H901SB-1 9207

PG PT 3SW3/4-908SB-1 B 7SW11/2-913-1 S 9425 9209 E 3SW-V903SB-1 3SW11/2-822SB-1 B B 3SW-H66SB-1 E 3SW-H815SB-1 3SW3/4-911SB-1 SB SB NOTES:

3SW-V820SB-1 3SW1-847SB-1 3SW1-819SB-1 3SW-H65SB-1 7SW11/2-914-1

1. FOR GENERAL NOTES AND DETAILS 3SW-V840SB-1 7SW-J10-1 SEE DWG CAR-2168-G-509 S01 3SW-V1008SB-1 3SW1-119SB-1 ED ED PI 2. FOR LEGEND AND SYMBOLS 3SW3/4-1629SB-1 PI 3SW-V836SB-1 9210 SEE DWG CAR-2168-G-528 S02 PI 9208 B 3 TI B

3SW12-86SB-1 V 3SW1-843SB-1 3. TEMPORARY STRAINERS USED TI DR 9210 S 3SW1-846SB-1 CS SS FOR INITIAL FLUSHING, ARE TO VENT S 9208 B 3SW-H819SB-1 PCV B S BE REMOVED BEFORE START-UP.

9209 SS S SS V SS SS EH SB 4. FOR REFERENCE DWGS SEE 3SW-V839SB-1 3SW1-823SB-1 B F SW CS CS V CONDENSER CS CS SW DWG CAR-2168-G-499 F CAR-2165-G-047 3SW12-85SB-1 3SW12-86SB-1 CAR-2165-G-047 (J11)

CS (2500 GPM) CS F.O. (J11)

SS 3SW-S33SB-1 CS SS CS 3SW-V848SB-1 SS 3SW3/4-855SB-1 TEMP. STRNR. 3SW-V849SB-1 3SW-B303SB-1 SS 6. (*) SYMBOL INDICATES VENDOR SUPPLIED 7SW3/4-857-1 7SW3/4-858-1 3SW1-818SB-1 3SW3/4-909SB-1 ITEM.

3

3SW-H814SB-1 3CX-V101SB-1 3 3SW10-939SB-1 7. HOSES WILL BE USED BETWEEN THE DRAIN

3 3CX1-130SB-1 COMPRESSOR

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PI TI TE TI PI 3CH-R2SB-1 3CX1-133SB-1 9421 COND. REFRIGERANT 9433 9430 3CX-V100SB-1 9423 9430 3 B B TRANSFER SYSTEM B B 7CH-1-1100-1 CH 3CX-V495SB-1 VENT B S S SB S S CAR-2168-G-499 G 3CX1-134SB-1 3CX-V108SB-1 3CH3/4-483SB-1 3CH-V84SB-1 ED (D17) G V COOLER V 3CX1-174SB-1 3CH3/4-90SB-1 3CH-V423SB-1 3CH-V422SB-1 CH 3CX1-2312SB-1 (1298 GPM) 3CH10-5OSB-1 CAR-2168-G-499 3CX10-172SB-1 (A16)

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3CX-V367SB-1 3CX3/4-484SB-1 3CX11/2- 128SB- 1 3CX-V103SB-1 3CX1-129SB-1 SEE SEE FE 3CX-V364SB-1 NOTE 7 FT NOTE 7 9422 FUNNEL 9429 B

S B 3 7CH11/2-1101-1 7CH11/2-1102-1 7CH-V1031-1 3CX3/4-485SB-1 SB ED 3CX1-173SB-1 3CH11/2-105SB-1 V 3CX-V107SB-1 7CH-V1002-1 3CX11/2-131SB-1 3CX-V365SB-1 3CX-R5SB-1 7CH-V1029-1 PI 3CH11/2-104SB-1 3 7CH1-1006-1 9422 7CH1-1035-1 3

B S WATER CHILLER WC-2 I DR (1B-SB) CHEMICAL I 3CX3/4-10SB-1 7CX1-1101-1 LG 752 TONS ADDITION 3CX-V368SB-1 9431 INCORPORATED: EC# 301132 ELECTRONICALLY B TANK 35 7CX1-1102-1 (G10) SIGNED 7CX1-1013-1 3CX-R4SB-1 1B-NNS 3CX1-448SB-1 INCORPORATED: EC# 406320 ELECTRONICALLY 34 (C7,C10) SIGNED P-4 (1B-SB) BY TANK 7CX-V1100-1 INCORPORATED: EC# 404085 ELECTRONICALLY 7DWf-640-1 3CX1-132SB-1 MANUFACTURER 33 (G8) SIGNED V 7CX-V1004-1 3CX1-135SB-1 7DW-S5-1 ELECTRONICALLY INCORPORATED: EC# 94413 R1 7DW1-623-1 3CX-V2282SA-1 32 (C9) SIGNED 7DW-H150-1 ED 7CX2-1009-1 7DW-V523-1 ED PX INCORPORATED: EC# 94413 ELECTRONICALLY J 7DW-P40-1 3CX-V2283SB-1 3CX-V95SB-1 31 (E6,E8,E11,F7,F8,F11,F12,F13) SIGNED J 7DW-V524-1 S S S 7CX2-1010-1 DEMIN. WATER SUPPLY INCORPORATED: EC# 93478 3CX-S2SB-1 ELECTRONICALLY CAR-2165-G-049S02 30 (F5,F6,F12,L5)

SEE NOTE 3 SIGNED (C6) 7DW-V525-1 3CX-V96SB-1 INCORPORATED: EC# 79156 ELECTRONICALLY 7DW-V522-1 PI 3SW 1- 864SB- 1 3CX-V2285SB-1 29 (C7,D11,E10) 7DW-V526-1 3 9421 SIGNED 3CX1-449SB-1 7DW-H151-1 B S 3 INCORPORATED: EC# 65195 7DW-V521-1 7DW-V527-1 3CX-V98SB-1 ELECTRONICALLY DR 28 (C8) SIGNED D 7DWf-641-1 3CX-V369SB-1 INCORPORATED: EC# 60749 ELECTRONICALLY 27 (C8) 3CX1-136SB-1 SIGNED 3CX-B6SB-1 3CX2- 137SB- 1 7DW-V528-1 7DW1/2-622-1 INCORPORATED: EC# 61385 ELECTRONICALLY 26 K 3CX1-2456SB-1 CX (E1-E3, F2, F3) SIGNED K

MECH HVAC 7DW1-624-1 CAR-2168-G-499 (H16) INCORPORATED: EC# 58681 25 10/25/04 (G3, J2-J5)

JAP RPP CRW 3CX21/2-176SB-1 24 5-10-03 INCORPORATED: EC# 48735 (C9) JAP RPP CHG SS CS CX INCORPORATED: ESR 94-00164 L.O.

23 11-16-00 (E10,E13,F9,F10)

JWB WFL RPP FOR DLS 3CX11/2-128SB-1 CAR-2168-G-499 3CX10-68SB-1 3CX10-68SB-1 3CX-V99SB-1 (K16)

L INCORPORATED: ESR-9500717 L (B9,E8-E9,F3,F7) 22 11-6-96 RPP LLS DLS DLS NON-DESIGN CHANGE: CORRECTED DRAFTING ERROR (E9)

INCORPORATED IN THIS REVISION:

21 1-12-94 LLS RPP MMP WAS PCR-6411 (F13).

INCORPORATED IN THIS REVISION:

20 11-3-93 PCR-2512 FR09 (G5). GRS LLS MMP WAS INCORPORATED IN THIS REVISION:

19 12-1-92 TRW RPP GOW AMW PCR-5534(J1-J5).

M THIS DWG HAS BEEN ELECTRONICALLY MMP M

REDRAWN WITH DESIGN CHANGE.

18 10-15-92 GRS RPP GOW for INCORPORATED:PCR-2512(E2-E4,F1-F5, JGT G1,G2)

REV DATE DESCRIPTION DWN CHK DPE DPPE THIS DRAWING HAS BEEN PRODUCED AND IS CONTROLLED ON THE NED CAD/E SYSTEM. ANY REVISIONS TO THIS DOCUMENT SHOULD BE DONE USING THE CAD/E SYSTEM TO ASSURE PROPER CONTROL OF THE ELECTRONIC DRAWING DATABASE.

NUCLEAR SAFETY RELATED (Q-LIST)

YES NO PARTIAL F.P.-Q R.W.-Q N DUKE ENERGY N

NUCLEAR ENGINEERING DEPARTMENT - RALEIGH, N.C.

PLANT: SCALE:

HARRIS NUCLEAR PROJECT - UNIT 1 NONE TITLE: DWG NO.

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HARRIS NUCLEAR PROJECT - UNIT 1 IscALE'NONE 1 TITLE: IDWG NO.

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MASTER EC# 63160 I ELECTRONICALLY SIGNED AND DATE REV DESCRIPTION APPROVED II THIS DRAWING HA5 BEEN PRODUCED AND IS CONTROLLED O N THE NED CAD/E

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F . P . - Q O R . W . - Q I COOLING COIL. ( 3 8 1 FOR A H - I 2 C O l t(5A)

FO R AH- I3 I CAROLINA POWER & LIGHT COMPANY NUCLEAR ENGINEERING DEPARTMENT - RALEIGH, N.C. I cp&L 1"' HARRIS NUCLEAR PROJECT - UNIT 1 I""': NONE TITLE: DWG NO.

I5 'SUPPLIED BY WATEIZ CHILLER PJC-2(IB-S0) HVAC-ESSENTIAL SERVICES CHILLED FLOW DIAGRAM - U N I T 1-S B WATER-DISTRIBUTION CAR-2 168-G-0499 DETAI L A(07)

U.S. Nuclear Regulatory Commission Serial RA-19-0007 SERIAL RA-19-0007 ATTACHMENT 5 ESCWS EXTENDED AOT LAR PRA INPUT SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63

Appendix A - ESCW Extended AOT LAR Technical Input Summary A-1

A.1 PRA Scope The change in risk associated with the requested 7-day AOT for a single train of the Essential Services Chilled Water system (ESCW) out of service has been evaluated for Harris Nuclear Plant (HNP) in accordance with the guidance of RGs 1.174 and 1.177 (Refs. A.11.1 and A.11.2). Hazard groups were evaluated to determine which sources of risk could affect the decision, and the risk from such hazards was assessed quantitatively and qualitatively using PRA models that have been assessed against the Capability Category II Supporting Requirements (SRs) in the existing PRA standards as well as Reg. Guide 1.200 (Ref. A.11.5).

The Harris PRAs currently model internal events for CDF and LERF, internal flooding, high winds and fire.

These models have been peer reviewed and the impact of the open findings has been evaluated.

Section 2.3.2 of RG 1.177 identifies the NRCs regulatory position on PRA scope, and states, in part:

in some cases, a PRA of sufficient scope may not be available. This will have to be compensated for by qualitative arguments, bounding analyses, or compensatory measures.

This section further states, in part:

The scope of the analysis should include all hazard groupsunless it can be shown that the contribution from specific hazard groups does not affect the decision.

RG 1.174 Section 2.3.1 further clarifies this concept:

A qualitative treatment of the missing modes and hazard groups may be sufficient when the licensee can demonstrate that those risk contributions would not affect the decision; that is, they do not alter the results of the comparison with the acceptance guidelines A.2 PRA Acceptability The Harris Nuclear Plant (HNP) PRA models are sufficiently robust and suitable for use in risk informed processes such as regulatory decision making. The peer reviews that have been conducted and the resolution of findings from those reviews demonstrate that the internal events, internal flooding, fire and high winds models of the PRA have been performed in a technically correct manner. The assumptions and approximations used in development of the PRA have also been reviewed and are appropriate for their application. Duke Energy procedures are in place for controlling and updating the models, when appropriate, and for assuring that the model represents the as-built, as-operated plant (Ref. A.11.6). The conclusion, therefore, is that the HNP PRA models are acceptable to be used as the basis for risk-informed applications including assessment of proposed Technical Specification amendments.

A.2.1 Internal Events 2007 Internal Events Upgrade. The 2007 revision to the internal events model of record (MOR) incorporated findings and observations (F&O) resolutions for the April 2006 HNP PRA Self-Assessment to meet ASME/ANS Internal Events Standard (Revision 1) for Category II compliance. A peer review was performed to support the NFPA 805 license amendment request submittal. Major revisions included expansion of plant-specific data, Human Reliability Analysis (HRA) updates, and addition of new or more detailed heating, ventilation and air conditioning (HVAC) models for CSIP rooms, Switchgear rooms, and A-2

Emergency Service Water (ESW) pump rooms. The model revision also included addition of logic to address fire- induced multiple spurious failures, developed in conjunction with HNP adoption of the NFPA 805 program for fire-induced vulnerabilities. Other general updates to the model included an update to the initiating event frequencies, revision of the station blackout (SBO) induced seal LOCA, and Loss of offsite power (LOOP) recovery. Motor control center modeling was improved to support the NFPA 805 LAR with the required level of detail.

2010 Update. The major change for the 2010 update was the addition of the Alternate Seal Injection -

Dedicated Shutdown Diesel Generator (ASI-DSDG) to the MOR. The installation of the ASI-DSDG modification provided a diverse and redundant power source for alternate seal injection and also to the emergency DC battery chargers, as described in the NFPA 805 LAR. This reduced the effect of the 4-hour coping duration of the batteries by providing a means to supply DC power to the DC busses during SBO.

The LOOP initiator was separated into plant, grid, switchyard and weather induced LOOPs, which allowed the model to apply recovery actions to the higher frequency events (plant and switchyard). Other changes related to de-energizing charging pump discharge header cross-connect valves, adding temporary air compressors, and updates from fire model were added to the 2010 model. This was not an upgrade and a peer review was not required for these revisions.

In 2015, an updated self-assessment of the current model was completed against the requirements of ASME/ANS PRA Standard as endorsed by RG 1.200, Rev. 2. The results of the assessment, provided in Appendices N and O of Ref. A.11.21, demonstrated the model met all supporting requirements at an appropriate capability category (i.e., CC II or higher).

F&O Closeout. In 2017, an independent assessment was performed to review actions taken by Duke Energy to close out the 12 open internal events F&Os (Ref. A.11.22). The assessment was a pilot for the process documented in the draft of Appendix X to NEI 05-04 (Ref. A.11.23). NRC staff observed the pilot closure on-site event held January 31 through February 1, 2017. All finding level internal events F&O dispositions were determined to have been adequately addressed and are now considered CLOSED and no longer relevant to the PRA model.

2017 / 2018 Update. (Ref. A.11.24) The model was updated in 2017 to document the sequence quantification for the revised model-of-record MOR17 supporting the Harris plant. The model also incorporated the credit for FLEX equipment as well as implementing several PRA Tracker items.

The model was subsequently updated in 2018 (MOR18) to incorporate the results from additional dependencies from the HRA analysis as well as updated initiator frequencies. No additional fault tree or data changes were made in this revision. The 2017 and 2018 models are identical (with the exceptions of additional dependencies added to the tree).

Based on these reviews, the HNP internal events PRA meets the requirements of the ASME/ANS PRA Standard as endorsed by RG 1.200, Revision 2, at an appropriate capability category to support the extended AOT T.S. LAR for the ESCW chillers.

A-3

A.2.2 Internal Flooding 2014 Update. The internal flooding portion of the HNP PRA was upgraded in 2014 to meet the requirements of the ASME/ANS PRA Standard (Ref. A.11.4) and RG 1.200, Rev. 2 (Ref. A.11.25). A comprehensive flooding analysis was performed to meet the supporting requirements of the Internal Flooding Probabilistic Risk Assessment (IFPRA) portion of the PRA standard. The most noteworthy changes to the flooding model included the addition of spray effects and high energy line breaks (HELB) and their associated impacts on PRA equipment not previously included. The analysis resulted in the identification and quantification of flood-induced scenarios that were incorporated into the model. A focused peer review of the IFPRA was conducted (Ref. A.11.21, Appendix L) following the guidance of NEI 05-04 (Ref.

A.11.23) to assess the model against the supporting requirements of the ASME/ANS PRA standard. All F&Os from the peer review have been dispositioned and incorporated into the current internal flooding model.

F&O Closeout. In 2017, an independent assessment was performed to review actions taken by Duke Energy to close out the 25 open internal flooding F&Os (A.11.22). The assessment was a pilot for the process documented in the draft of Appendix X to NEI 05-04 (Ref. A.11.23). NRC staff observed the pilot closure on-site event held January 31 through February 1, 2017. 15 of 25 finding level F&O dispositions were determined to have been adequately addressed and are now considered CLOSED and no longer relevant to the PRA model. Of the remaining F&Os, 8 were considered PARTIALLY CLOSED and 2 remain OPEN.

A.2.3 Fire The HNP fire PRA was developed using the guidance provided by NUREG/CR-6850 (Ref. A.11.26) in support of the NFPA 805 fire protection program, and HNP was a pilot plant for implementation of NFPA 805. The fire PRA is built upon the internal events PRA which was modified to capture the effects of fire. In 2008, both a follow-up, partial scope industry peer review and an NRC staff review were conducted on the fire PRA (Ref. A.11.21, Appendices G and J). The follow-up industry peer review compared the fire PRA against the requirements of the ANSI/ANS 58.23-2007 Standard (Ref. A.11.27) in accordance with guidance in NEI 07-12, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines (Ref. A.11.28). All findings have been reviewed and the resolutions were submitted as part of the NFPA 805 LAR. The results of the NRC staff quality review of the Fire PRA are documented in the HNP NFPA 805 Safety Evaluation for transition to a risk-informed, performance-based fire protection program, ADAMS Accession Numbers ML101750602 and ML101750604 (Refs. A.11.29 and A.11.30, respectively). The quality review concluded that the technical adequacy and quality of the HNP PRA is sufficient for the fire risk evaluations that support NFPA 805 fire protection program.

2013 Update. The 2013 revision implemented resolutions for the previously identified conservatisms in the fire model. The main changes involved updating human failure events, dependency analysis, and recovery rule files. Other updates included additional walkdowns to identify fixed and transient ignition sources, crediting of implemented plant modifications, and updates to fire frequency bin numbers to match the newest version of Ref. A.11.26. This was not an upgrade and a peer review was not required for these revisions.

2017 Closeout. Five findings remain open from the original NRC review and one finding from the 2008 peer review (per A.11.31). Additionally, there are 6 SRs met at CC-1 or less with no open findings. Closed findings were reviewed and closed for the Fire PRA model using the process documented in Appendix X A-4

to NEI 05-04, NEI 07-12 and NEI 12-13, Close-out of Facts and Observations (F&Os) (Ref. A.11.46), as accepted by the NRC in a letter dated May 3, 2017 (Ref. A.11.47).

A.2.4 High Winds The HNP High Wind PRA was peer reviewed in 2015 (Ref. A.11.20). Four (4) finding F&Os were generated.

These findings were subsequently dispositioned by the high winds model vendor. However, according to Ref. A.11.39, some resolutions / dispositions to the F&Os provided by the vendors are not yet accepted and approved by Duke Energy. The updated high winds analysis presented in Ref. A.11.32 is intended to address this issue.

A.3 General Assumptions The following assumptions were applied globally to the PRA analysis:

Section 3/4.7.13 of plant Technical Specifications (Ref. A.11.7) requires two loops of the ESCW system to be operable in Modes, 1, 2, 3 and 4. Therefore, this assessment will be performed considering at-power operation only. There will be no assessment for shutdown conditions.

The assumed equipment unavailability is set to the average test and maintenance (T&M) as required per Section 5.7.6(d) of A.11.8. This unavailability pertains to all SSCs except for those manipulated for the conditional case.

The HFEs to open doors and implement portable fans as an alternate means of cooling the switchgear room / CSIP rooms do not significantly impact the risk results since their risk importance (i.e., Fussell-Vesely) is small (i.e., typically 1% or less).

A.4 Common Cause Failure (CCF) Evaluation The logic models for the extended AOT configuration did not require any new common cause events because no new failure modes were added that required a CCF assessment.

A.5 Risk Results The internal events, internal flooding, high winds and fire models were quantified to determine the CDF, LERF, CCDP and CLERP that would result from the approval of the extended AOT T.S. change. Sections A.5.1 - A.5.4 contain the quantitative delta CDF and LERF results for the internal events, internal flooding, fire and high winds hazards. Sections A.5.5 - A.5.7 contain qualitative analyses for external flooding, seismic and shutdown risk. Section A.5.8 contains the quantitative ICCDP / ICLERP results.

A.5.1 Internal Events Analysis The Internal Event CDF and LERF results are shown in Table A.5.1.

A-5

Table A.5.1 - Internal Events Results Case CDF ( / yr) LERF ( / yr)

Base Case 2.86E-06 1.07E-06 AOT Configuration 2.87E-06 1.07E-06 Delta 2.8E-09 1.3E-10 The increases in CDF and LERF are solely due to new accident sequences in which a train of ESCW is removed from service. Removing the availability factor makes the delta CDF 3.0E-09/yr and the delta LERF 1.4E-10/yr.

Importance measures were obtained for the extended AOT case as a way of determining Tier 2 measures.

Results were captured for basic events with a Fussell-Vesely (F-V) value greater than or equal to 0.005, a Risk Achievement Worth (RAW) greater than or equal to 2 and a Risk Reduction Worth (RRW) greater than or equal to 1.005. Tier 2 measures are presented in Section A.7.

A.5.2 Internal Flooding Analysis The Internal Flooding CDF and LERF results are shown in Table A.5.2.

Table A.5.2 - Internal Flooding Results Case CDF ( / yr) LERF ( / yr)

Base Case 6.07E-06 4.69E-07 AOT Configuration 6.07E-06 4.69E-07 Delta 1.8E-09 9.0E-11 The increases in CDF and LERF are solely due to new accident sequences in which a train of ESCW is removed from service. Removing the availability factor makes the delta CDF 2.0E-09/yr and the delta LERF 1.0E-10/yr.

Importance measures were obtained for the extended AOT case as a way of determining Tier 2 measures.

Results were captured for basic events with a Fussell-Vesely (F-V) value greater than or equal to 0.005, a Risk Achievement Worth (RAW) greater than or equal to 2 and a Risk Reduction Worth (RRW) greater than or equal to 1.005. Tier 2 measures are presented in Section A.7.

A.5.3 Fire Analysis The Fire CDF and LERF results are shown in Table A.5.3.

Table A.5.3 - Fire Results Case CDF ( / yr) LERF ( / yr)

Base Case 2.30E-05 4.75E-06 AOT Configuration 2.37E-05 4.76E-06 Delta 7.2E-07 1.2E-08 A-6

The increases in CDF and LERF are solely due to new accident sequences in which a train of ESCW is removed from service. Removing the availability factor makes the delta CDF 8.2E-07/yr and the delta LERF 1.4E-08/yr.

Importance measures were obtained for the extended AOT case as a way of determining Tier 2 measures.

Results were captured for basic events with a Fussell-Vesely (F-V) value greater than or equal to 0.005, a Risk Achievement Worth (RAW) greater than or equal to 2 and a Risk Reduction Worth (RRW) greater than or equal to 1.005. Tier 2 measures are presented in Section A.7.

A.5.4 High Winds Analysis The High Winds CDF and LERF results are shown in Table A.5.4.

Table A.5.4 - High Winds Results Case CDF ( / yr) LERF ( / yr)

Base Case 2.14E-06 2.24E-07 AOT Configuration 2.18E-06 2.26E-07 Delta 3.9E-08 1.7E-09 The increases in CDF and LERF are solely due to new accident sequences in which a train of ESCW is removed from service. Removing the availability factor makes the delta CDF 4.4E-08/yr and the delta LERF 1.9E-09/yr.

Importance measures were obtained for the extended AOT case as a way of determining Tier 2 measures.

Results were captured for basic events with a Fussell-Vesely (F-V) value greater than or equal to 0.005, a Risk Achievement Worth (RAW) greater than or equal to 2 and a Risk Reduction Worth (RRW) greater than or equal to 1.005. Tier 2 measures are presented in Section A.7.

A.5.5 External Flooding Impact For the IPEEE submitted in 1995 (Ref. A.11.9), HNP utilized a screening approach described in NUREG-1407 (Ref. A.11.10) to identify potential vulnerabilities due to external floods. In 2013, HNP completed a Flood Hazard Reevaluation Report in response to NRC 10 CFR 50.54(f) regarding recommendations of the Near-Term Task Force (NTTF) review of insights from the Fukushima Dai-ichi accident (Ref. A.11.11). The results indicated some flood levels determined during the hazard reevaluation exceed the Current Licensing Basis (CLB) flood levels. The increased levels are the result of newer methodologies and not the result of errors within the CLB evaluations. Although some flood levels exceed the CLB flood levels, the increased levels do not exceed the flood protection capabilities and do not impact safety-related equipment. Thus, they do not require a quantitative risk evaluation. External flooding risk is therefore evaluated qualitatively and screened as no impact for this License Amendment Request. Taking a train of the ESCW system out of service, therefore, is assessed to be unaffected by an external flood during an extended AOT.

A.5.6 Seismic Risk Initially, the ESCW system was walked down as part of Harris' Individual Plant Examination for External Events (IPEEE) submittal (Ref. A.11.9). A seismic margins assessment (SMA) approach was used for the assessment which required a walkdown and anchorage calculations.

A-7

There are four air handling units (AHUs) for ESCW on the SMA safe shutdown equipment list (SSEL). The AHUs are located in the Reactor Auxiliary Building (RAB) with two units at elevation 236' and two units at 305'. The AHUs are associated with cooling the charging / safety injection pump rooms, electrical cabinet rooms and the control room. No interaction or maintenance issues were noted during the walkdown. The AHUs were screened by the seismic review team (SRT) based on the walkdown.

Likewise, the SSEL contained two chillers for ESCW. The chillers are located in the (RAB) at elevation 261'.

The chillers provide chilled water to the AHU's. No interaction or maintenance issues were noted during the walkdown. The chillers were also screened by the SRT based on the walkdown, anchorage calculations and the seismic analysis.

In 2014, HNP completed a Seismic Hazard Evaluation and Screening Report in response to NRC 10 CFR 50.54(f) regarding recommendations of the Near-Term Task Force (NTTF) review of insights from the Fukushima Dai-ichi accident (Ref. A.11.12). The results indicated the updated seismic hazard is lower than evaluated in the IPEEE and is not a significant hazard requiring quantitative risk evaluation. The NRC's comparison of the ground motion response spectrum (GMRS) to the safe shutdown earthquake (SSE) used in the IPEEE analysis for HNP is shown in Figure A-1 (Ref. A.11.13). In addition, a graph of the uniform hazard response spectra (UHRS) at 1E-4 and 1E-5, along with the GMRS, were presented as shown in Figure A-1 below (from Ref. A.11.12).

The NRC staff reviewed the information provided by Duke for the reevaluated seismic hazard for the HNP site (Ref. A.11.14). Based on its review, the NRC staff concluded that Duke conducted the hazard reevaluation using present-day methodologies and regulatory guidance, it appropriately characterized the site given the information available, and met the intent of the guidance for determining the reevaluated seismic hazard. Based upon this analysis, the NRC staff concluded that Duke provided an acceptable response to the requested Information items identified in Enclosure 1 to the 50.54(f) letter. Further, the staff concluded Duke's reevaluated seismic hazard is acceptable to address other actions associated with NTTF Recommendation 2.1: "Seismic". In reaching this determination, the NRC staff confirmed Duke's conclusion that its GMRS for the HNP site is bounded by the SSE in the 1 to 15 Hz range and above 40 Hz range, but exceeds the SSE in a portion of the frequency range from approximately 15 to 40 Hz. As such, a seismic risk evaluation and spent fuel pool (SFP) evaluation were not merited; however, a high frequency (HF) confirmation was merited. This does not apply to the ESCW equipment since high frequency reviews under the Fukushima response were related to relays and contactors. In fact, Harris was screened out of the HF review based on minimal exceedance of the GMRS over the SSE (see Figure A Ref. A.11.13).

Given these results, the risk from seismically induced failure of the ESCW is assessed to be a very low probability event and the incremental risk resulting from one train of ESCW being inoperable for a period of 7 days is considered to be acceptable.

A-8

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0. I 0 .1 1 10 100 Spectral frequency, Hz Figure A Harris Nuclear Plant GMRS A.5.7 Shutdown Operations Section 3/4.7.13 of plant Technical Specifications (Ref. A.11.7) requires two loops of the ESCW system to be operable in Modes, 1, 2, 3 and 4. Thus, the impact of the extended AOT will have no negative impact on shutdown risk.

A.5.8 ICCDP / ICLERP for 7-Day AOT The ICCDP and ICLERP for one entry into the T.S. are now computed. First, the delta CDF and LERF computed in the Sections A.5.1 thru A.5.4 above are tabulated below:

Table A.5.8 - Risk Summary, All Hazards CDFAOT LERF AOT CDF Base LERF Base CDF Delta 1 LERF Delta 1 Risk Risk

( I yr.) ( I yr.) ( I yr.) ( I yr.)

( I yr.) ( I yr.)

Internal Events 2.86E-06 l.07E-06 2.87E-06 l.07E-06 3.0E-09 1.4E-10 Internal Flood 6.07E-06 4.69E-07 6.07E-06 4.69E-07 2.0E-09 1.0E-10 Fire 2.30E-05 4.75E-06 2.37E-05 4.76E-06 8.2E-07 1.4E-08 High Winds 2.14E-06 2.24E-07 2.18E-06 2.26E-07 4.4E-08 1.9E-09 Total 3.41E-05 6.SlE-06 3.48E-05 6.53E-06 8.7E-07 1.6E-08 Thus, for a 7-day AOT, the ICCDP and ICLERP are, 1 Availability factors removed A-10

ICCDP = [(CDFAOT Config. - CDFBaseline) x (7 days) / 365 days/yr]

= (8.7E-07) x 7 / 365

= 1.7E-08 ICLERP = [(LERFAOT Config. - LERFBaseline) x (7 days) / 365 days/yr]

= (1.6E-08) x 7 / 365

= 3.0E-10 Therefore, for a 7-day AOT, the delta CDF is less than 1E-06 / yr. and the delta LERF is less than 1E-07 and thus meet the acceptance guidelines of RG 1.174 for a very small risk increase. Per Figures 4 and 5 of RG 1.174, these changes in risk are in Region III which, per the RG, will be considered regardless of whether there is a calculation of the total CDF and LERF (Region III).

Similarly, the ICCDP is less than 1E-6 and the ICLERP is less than 1E-07; therefore, these risk metrics meet the acceptance guidelines of RG. 1.177, Section 2.4.

A.6 PRA Model Configuration and Control Program The HNP PRA Models of Record (MORs) are maintained as controlled documents and are updated on a periodic basis to represent the as-built, as-operated plant. Duke Energy procedures provide the guidance, requirements, and processes for the maintenance, update, and upgrade of the PRA (Ref A.11.6):

a. The process includes a review of plant changes, selected plant procedures, and plant operating data as required, through a chosen freeze date to assess the effect on the PRA model.

b. The PRA model and controlling documents are revised as necessary to incorporate those changes determined to impact the model.

c. The determination of the extent of model changes includes the following:

Accepted industry PRA practices, ground rules, and assumptions consistent with those employed in the ASME/ANS PRA Standard (Refs. A.11.4 & A.11.15),

Current industry practices,

NRC guidance (e.g., RG 1.174 and RG 1.177, Refs. A.11.1 & A.11.2 respectively),

Advances in PRA technology and methodology, and

Changes in external hazard conditions.

For plant changes of small or negligible impact, the model changes can be accumulated and a single revision is performed at an interval consistent with major PRA revisions. The results of each evaluation determine the necessity and timing of incorporation of a particular change into the PRA model. An electronic tracking database (PRA Tracker) is utilized to document pending model changes and updates.

A review of the electronic tracking database was conducted to determine if there were any open medium or high risk impact items that needed to be assessed against the current MORs for this analysis. These items are addressed below:

A-11

Impact Tracking ID Brief Description Disposition (If N/A)

Description EC 291969 adds several pipe This is an internal flooding issue dealing segments for FLEX equipment to the with FLEX equipment piping. Any piping suction and discharge lines of the failures at or near the AFW pumps would H-16-0018 Medium AFWMDPs. This piping and these not impact the ESCW chillers. Therefore, valves are to be considered in the this mod. has no direct impact on this T.S.

Internal Flooding PRA model. LAR application.

EC 291710 adds a fire suppression system to the Diesel Building (Bays This is a plant mod. involving FLEX 2A & 2B) where the FLEX equipment equipment and its impact on the Fire PRA.

will be staged. This involves a dry Any initiation of fire suppression H-16-0023 Medium pipe sprinkler system, a water flow equipment in the Diesel Building would switch and air pressure hi/lo switch. not impact the ESCW chillers. Therefore, Additional doors are to be added as this mod. has no direct impact on this T.S.

well to allow secondary access in LAR application.

case of debris.

Adding this action would not affect the Add operator action to isolate a delta risk since the action would be H-18-0007 Medium ruptured steam generator credited in both the base and AOT cases.

No impact on this application.

A.7 Tier 2 Component Evaluation RG 1.177 (Ref. A.11.2) defines Tier 2 of the NRC staffs three-tiered approach for evaluating the risk associated with proposed TS AOT changes as the identification of potentially high-risk configurations that could exist if equipment, in addition to that associated with the change, were to be taken out of service simultaneously or other risk-significant operational factors, such as concurrent system or equipment testing, were also involved. The objective of this part of the evaluation is to ensure that appropriate restrictions on dominant risk-significant configurations associated with the change are in place.

Duke Energy relies on several methods to limit work on high risk configurations. These methods consist of Technical Specifications (Tech Specs) and Selected Licensee Commitments (SLC), Cycle Schedule, Protected Equipment schemes, and the Electronic Risk Assessment Tool (ERAT).

Tech Specs and SLC specify requirements for SSCs to be operable or functional. Tech Specs and SLC specify a completion time (AOT) for SSCs. Generally, when multiple trains are out of service, the AOT is very short or a shutdown is required. In the case of ESCW, the AOT for Section 3/4.7.13 of the HNP plant Technical Specifications will be increased from from 72 hrs to 7 days.

During the AOT, planned or discretionary maintenance that renders the available ESCW Chiller train inoperable and unavailable is prohibited while in the extended AOT condition. Protected equipment plans will be developed for important SSCs. These plans are maintained by the Operations group. Duke procedure AD-OP-ALL-0201 (Ref. A.11.17) provides guidance for the management of protected equipment.

Duke Energy's online work management practices are described in AD-WC-ALL-0200 (Ref. A.11.16) A key provision of this practice is the use of a Cycle Schedule. "Plant systems are grouped in a rotating cycle of Work Weeks. System groupings are based on Technical Specification requirements, Probabilistic Risk Assessment (PRA) and resource loading."

A-12

Work on those SSCs which is not prohibited by Tech Specs or SLC, the Cycle Schedule, or the Protected Equipment Plan will be managed using the Electronic Risk Assessment Tool (ERAT). As outlined in Duke procedure AD-NF-ALL-0501 (Ref. A.11.18), Duke manages this process using a blended (i.e., quantitative and qualitative) configuration risk assessment approach.

Based on SSCs assumed to be available in the PRA analysis, the operable ESCW chiller train should be maintained in protected train status when the plant is in the ESCW AOT. Additionally, importance metrics were evaluated to determine any other components which the plant should minimize removing from service during the AOT. These importance measures (Fussell-Vesely, Risk Achievement Worth and Risk Reduction Worth) during the AOT provide insights into what equipment should remain available during the extended AOT. The SSCs whose unavailability should be minimized during the AOT, based upon Fussell-Vesely, Risk Achievement Worth and Risk Reduction Worth importance measures, are given in Table A.7 below:

Table A.7 - Tier 2 SSCs SSC Risk Metric(s) Reason Protected in PRA risk Maintaining essential service chilled ESCW System (operating train) assessment water availability IE (CDF)

Motor-Driven and Turbine- IF (CDF / LERF) Provide secondary side heat removal Driven AFW Pumps HW (CDF / LERF) capability Fire (CDF / LERF)

Emergency Service Water IE (CDF) Provide cooling to available ESCW Chiller System HW (CDF / LERF) train and EDGs (if needed)

IE (CDF)

Emergency DGs Provide emergency power (if needed)

HW (CDF / LERF)

Needed for independent, automatically-ASI-DSDG Fire (CDF) actuated back-up RCP seal injection NSW --------------------------- Source of water for ESWS A.8 Tier 3 Evaluation -

Tier 3 of RG 1.177 requires the licensee to provide assurance of compliance with 10 CFR 50.65(a)(4) to ensure the risk impact of taking equipment out of service is appropriately assessed and managed. As outlined in procedure AD-NF-ALL-0501 (Ref. A.11.18), Duke manages this process using a blended (i.e.,

quantitative and qualitative) risk assessment approach with its Electronic Risk Assessment Tool (ERAT).

HNP uses the Equipment Out of Service (EOOS) software program to analyze plant risk in both real time

('Operator Screen' mode) as well as a look-ahead of plant configurations over a specified period of time

('Scheduler Screen' mode). Prior to entering the extended AOT, EOOS operators can review the plant schedule to identify and correct any significant potential risk impacts occurring during the AOT. During the AOT, risk will be monitored in real time and any emergent risk configurations will be addressed appropriately. Duke Energys configuration risk management program requires the implementation of risk management actions to help alleviate risk when risk significant configurations are entered. Thus, plant risk will be effectively managed prior to and during the extended AOT.

A-13

A.9 Compensatory Actions According to the analyses presented in Ref. A.11.19 and due to the design of the Harris plant (with very large equipment spaces), the HNP PRA model only requires the chillers for the Charging/Safety Injection Pump (CSIP) Rooms and the Switchgear Rooms. The loss of a cooler in an HVAC system serving a charging pump room at HNP may result in an unacceptable increase in the room temperature after startup of the charging pump. Furthermore, the Switchgear Rooms' cooling loads during the summer months are low enough such that, without operator action, the equipment would continue to meet the PRA success criterion of 131oF. The Switchgear Rooms' cooling loads in the winter months are higher due to electric heating coil unit loading of the transformers and will require operator action to meet the PRA success criterion. The ability of the operators to apply procedural steps to open doors and provide circulation from adjacent spaces is credited when the chillers are unavailable in the internal events PRA model.

The risk-sensitive areas are therefore the CSIP rooms, the Switchgear Rooms and the opposite train (operable) ESCW chiller (WC-2A or -2B). With this in mind, prior to exceeding the initial 72 hrs. of the AOT, the following compensatory actions are proposed:

Discretionary maintenance or discretionary testing on equipment that support the Tier 2 systems listed above will be avoided for the remaining duration of the TS 3.7.13 (Ref. A.11.7).

The following equipment and the corresponding power supplies will be posted protected:

o Air handling unit for the operable CSIP areas: AH-9A (CSIP 1A-SA Area), AH-9B (CSIP 1B-SB Area), or AH-10 (CSIP 1C-SAB Area) o Air handling units for the Switchgear Rooms with operable equipment: AH-12 1A-SA and AH-12 1B-SA supply Switchgear Room A; AH-13 1A-SB and AH-13 1B-SB supply Switchgear Room B

o Operable ESCWS chiller and operable chilled water pump

The Fire Protection tracking log will be reviewed for fire hazards and fire impairments.

Accordingly, transient combustibles and hot work in these fire risk-sensitive areas will be limited:

o Fire compartment FC25 - RAB HVAC Room (MCC 1A21-SA, MCC 1A31-SA) o Fire compartments FC34 and FC35 - Switchgear Rooms A and B o Fire compartment FC41 - Turbine Building (Zone 1-G-261 - 6.9 kV Switchgear) o Fire compartment FC54 - Transformer Yard

Restrictions on work activities that involve components that if lost or failed could result in a direct plant trip or transient.

Operator actions for the CSIP area cooling, Switchgear Room cooling, and Auxiliary Relay Cabinet Room cooling, if needed, following a loss of HVAC, will be briefed with Operations. The fan used for the CSIP area cooling will be pre-staged and verified to be functional.

A.10 Generic Sources of Modeling Uncertainty -

The generic sources of modeling uncertainties from EPRI Report 1016737 (Ref. A.11.37) have been evaluated for the internal events model. The review of these sources of uncertainties is documented in Table A.10.1.

A-14

Generic sources of uncertainty identified in WCAP-16304-P (Ref. A.11.38) and plant-specific sources of uncertainty (from Ref. A.11.36) for internal events have been evaluated for this application. The review of these sources of uncertainties is documented in Table A.10.2. Further, plant-specific sources of uncertainty for the fire, high winds and internal flooding models (Refs. A.11.42, A.11.32 and A.11.43) are provided in Tables A.10.3, A.10.4 and A.10.5, respectively.

Major assumptions made in the analysis of the proposed ESCW TS change are documented in Table A.10.6.

In accordance with NUREG-1855, Rev. 1 (Ref. A.11.44), sensitivities were performed as needed to verify the key sources of uncertainty. These are discussed in Tables A.10.1 through A.10.6.

A-15

Table A.10.1 Uncertainties Considered from EPRI 1016737 Against Extended AOT LAR Impact Issue Description Issue Characterization NUREG-1855 type Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected of Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44)

Initiating Event Analysis (IE) 1 - Grid stability The LOOP frequency is a Applicable function of several factors LOOP sequences Aleatory No including switchyard design, The proposed ESCW TS change the number and independence does not impact LOOP of offsite power feeds, the local Consequential LOOP frequencies or LOOP recovery power production and Aleatory No uncertainties. In addition, use of sequences consumption environment and generic LOOP data (NUREG/CR-the degree of plant control of 6928) may introduce a minor the local grid and grid conservative bias; however, it is maintenance. reasonable to assume that Three different aspects industry data is generally relate to this issue: applicable to Harris. Thus, this does not introduce a key source 1a. LOOP initiating event LOOP or consequential of model uncertainty.

frequency values and recovery LOOP sequences with Aleatory No Applications activities such as probabilities 1b. Conditional offsite power recovered. NOEDs take into account grid LOOP probability 1c. Availability stability factors such as of dc power to perform weather-related issues; thus, restoration actions. seasonal and regional variations are not considered to be key sources of model uncertainty.

A-16

Table A.10.1 Uncertainties Considered from EPRI 1016737 Against Extended AOT LAR Impact Issue Description Issue Characterization NUREG-1855 type Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected of Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) 2 - Support System Increasing use of plant-specific Support system event Aleatory No Applicable Initiating Events models for support system sequences initiators (e.g., loss of SW, The loss of ESCW is not modeled CCW, or IA, and loss of ac or dc as an initiating event and thus is buses) have led to not a source of model inconsistencies in approaches uncertainty for this application.

across the industry. A number of challenges exist in modeling of support system initiating events:

2a. Treatment of common cause failures 2b. Potential for recovery A-17

Table A.10.1 Uncertainties Considered from EPRI 1016737 Against Extended AOT LAR Impact Issue Description Issue Characterization NUREG-1855 type Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected of Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) 3 - LOCA It is difficult to establish values LOCA sequences Aleatory No Applicable initiating event for events that have never frequencies. occurred or have rarely The PRA uses consensus occurred with a high level of models for quantifying confidence. The choice of probabilities of rare events.

available data sets or use of This is not a key source of specific methodologies in the uncertainty for this application.

determination of LOCA frequencies could impact base model results and some applications.

Accident Sequence Analysis (AS) 4 - Operation of Station Blackout events are Credit for continued Aleatory No Applicable equipment after important contributors to operation of these battery depletion baseline CDF at nearly every US systems in sequences The proposed ESCW TS NPP. In many cases, battery with batteries change does not affect depletion may be assumed to depleted (e.g., long- battery depletion uncertainty.

lead to loss of all system term SBO sequences) In addition, battery depletion capability. Some PRAs have after 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is modeled in credited manual operation of the PRAs. Credit for operation systems that normally require of equipment in accordance dc for successful operation (e.g., with procedural guidance turbine- driven systems such as given battery depletion RCIC and AFW). addresses the limitations brought about by the loss of dc power. Thus, this issue is not a key source of model uncertainty for this application.

A-18

Table A.10.1 Uncertainties Considered from EPRI 1016737 Against Extended AOT LAR Impact Issue Description Issue Characterization NUREG-1855 type Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected of Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) 5 - RCP seal LOCA The assumed timing and Accident sequences N/A N/A Not applicable with the use of treatment - PWRs magnitude of RCP seal LOCAs involving loss of seal industry consensus RCP seal given a loss of seal cooling can cooling model.

have a substantial influence on the risk profile. Application will not affect time or magnitude given a loss of seal cooling. The uncertainty associated with this item would have similar impacts on the base case and the conditional case. Therefore, this item is not a key source of uncertainty for this application.

6 - Recirculation Recirculation pump seal leakage Accident sequences N/A N/A Not Applicable - HNP is a PWR pump seal leakage can lead to loss of the Isolation with long-term use of treatment - Condenser. While recirculation isolation condenser BWRs w/ Isolation pump seal leakage is generally Condensers modeled, there is no consensus approach on the likelihood of such leaks.

A-19

Table A.10.1 Uncertainties Considered from EPRI 1016737 Against Extended AOT LAR Impact Issue Description Issue Characterization NUREG-1855 type Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected of Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44)

Success Criteria (SC) 7 - Impact of Many BWR core cooling systems Loss of containment N/A N/A Not Applicable - HNP is a PWR containment utilize the suppression pool as a heat removal venting on core water source. Venting of scenarios with cooling system containment as a decay heat containment venting NPSH removal mechanism can successful substantially reduce NPSH, even lead to flashing of the pool. The treatment of such scenarios varies across BWR PRAs.

A-20

Table A.10.1 Uncertainties Considered from EPRI 1016737 Against Extended AOT LAR Impact Issue Description Issue Characterization NUREG-1855 type Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected of Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) 8 - Core cooling Loss of containment heat Long term loss of decay Epistemic No Applicable success following removal leading to long-term heat removal sequences containment containment over- Application has no impact on loss failure or venting pressurization and failure can of containment heat removal through non hard be a significant contributor in uncertainties. The uncertainty pipe vent paths some PRAs. associated with this item would Consideration of the have similar impacts on the base containment failure mode case and the conditional case.

might result in additional Therefore, this item is not a key mechanical failures of credited source of uncertainty for this systems. Containment venting application.

through soft ducts or containment failure can result in loss of core cooling due to environmental impacts on equipment in the reactor/auxiliary building, loss of NPSH on ECCS pumps, steam binding of ECCS pumps, or damage to injection piping or valves. There is no definitive reference on the proper treatment of these issues.

A-21

Table A.10.1 Uncertainties Considered from EPRI 1016737 Against Extended AOT LAR Impact Issue Description Issue Characterization NUREG-1855 type Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected of Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) 9 - Room heat- up Loss of HVAC can result in Dependency on HVAC Epistemic Yes Applicable calculations room temperatures exceeding for system modeling equipment qualification limits. and timing of According to the analyses presented in Reference A.11.19 Treatment of HVAC accident progressions and due to the design of the Harris requirements varies across the and associated plant (with very large equipment industry and often varies within success criteria. spaces), the HNP PRA model only a PRA. There are two aspects to requires the chillers for the this issue. One involves Charging/Safety Injection Pump whether the SSCs affected by (CSIP) Rooms and the Switchgear loss of HVAC are assumed to Rooms. In addition, the ability of fail (i.e., there is uncertainty in operators to apply procedural the fragility of the steps to open doors and provide circulation from adjacent spaces is components). The other credited in the PRA models.

involves how the rate of room Furthermore, the Switchgear heat- up is calculated and the Rooms' cooling loads during the assumed timing of the failure. summer months are low enough such that, without operator action, the equipment would continue to meet the PRA success criterion.

The Switchgear Rooms' cooling loads in the winter months are higher due to electric heating coil unit loading of the transformers and will require operator action to meet the PRA success criterion. A sensitivity was performed assuming the operators fail to take action to provide emergency cooling during the winter months.

As a result, an examination of the delta risk cut sets indicated a CDF increase >1E-06/yr and a LERF increase >1E-08/yr.

Therefore, with the primary areas A-22

requiring cooling and the proper operator actions identified, a loss of HVAC is recognized as a key source of uncertainty for this application. The compensatory actions address this by having operators briefed on the proceduralized actions for recovering a loss of HVAC to the CSIP Rooms, Switchgear Rooms and the Auxiliary Relay Cabinet Room prior to exceeding the initial 72 hrs. of the AOT. In addition, the fan used for the CSIP area cooling will be pre-staged and verified to be functional.

A-23

Table A.10.1 Uncertainties Considered from EPRI 1016737 Against Extended AOT LAR Impact Issue Description Issue Characterization NUREG-1855 type Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected of Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) 10 - Battery life Determination of battery Epistemic No Applicable calculations depletion time(s) and the associated accident According to the analyses sequence timing and presented in Reference A.11.19 Station Blackout events are related success criteria. and due to the design of the important contributors to Harris plant (with very large baseline CDF at nearly every US equipment spaces), the HNP NPP. Battery life is an important PRA only models the chillers for factor in assessing a plants the Charging/Safety Injection ability to cope with an SBO. Pump (CSIP) Rooms and the Many plants only have Design Switchgear Rooms and is not Basis calculations for battery required to be modeled for life. Other plants have very cooling of the battery rooms.

plant/condition specific calculations of battery life. The batteries are assumed to Failing to fully credit battery deplete following loss of their capability can overstate risks, chargers since they are not and mask other potential designed to operate for a 24-hr contributors and insights. mission time. As such, battery Realistically assessing battery depletion after 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is modeled life can be complex. in the PRAs. The dependency of batteries on their chargers is treated realistically. Thus, this issue is not a key source of model uncertainty for this application.

A-24

Table A.10.1 Uncertainties Considered from EPRI 1016737 Against Extended AOT LAR Impact Issue Description Issue Characterization NUREG-1855 type Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected of Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) 11 - Number of PWR EOPs direct opening of all System logic modeling Epistemic No Applicable PORVs required for PORVs to reduce RCS pressure representing success bleed and feed - for initiation of bleed and feed criterion and accident This application does not affect the PWRs cooling. Some plants have sequence timing for success criteria for opening of performed plant-specific performance of bleed and PORVs in feed and bleed cooling analysis that demonstrate that feed and sequences scenarios. The uncertainty less than all PORVs may be involving success or associated with this item would sufficient, depending on ECCS failure of feed and bleed. have similar impacts on the base characteristics & initiation case and the conditional case.

timing. Therefore, this item is not a key source of uncertainty for this application.

A-25

Table A.10.1 Uncertainties Considered from EPRI 1016737 Against Extended AOT LAR Impact Issue Description Issue Characterization NUREG-1855 type Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected of Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) 12 - Containment All PWRs are improving Recirculation from sump Epistemic No Applicable sump / strainer ECCS sump management (PWRs) or from the performance practices, including suppression pool (BWRs) The containment sumps could fail installation of new sump system modeling and during the recirculation phase of strainers at most plants. sequences involving operations due to clogging. This is injection from these addressed in the RHR system All BWRs have improved their sources (Note that the model (Ref. A.11.33, Appendix A.2).

suppression pool strainers to modeling should be Nevertheless, this is a concern for reduce the potential for relatively LOCA sequences and not for ESCW.

plugging. However, there is not straightforward, the The uncertainty associated with a consistent method for the uncertainty is related to this item would have similar treatment of suppression pool the methods or impacts on the base case and the strainer performance. references used to conditional case. A sensitivity of determine the likelihood the delta risk cut sets assuming of plugging the sump sump clogging indicated an strainer and common increase in CDF of 8E-07/yr and an cause failure by blockage increase in LERF of 1E-08/yr.

of the strainers.) Therefore, this item is not a key source of uncertainty for this application.

A-26

Table A.10.1 Uncertainties Considered from EPRI 1016737 Against Extended AOT LAR Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) 13 - Impact of failure of Certain scenarios can lead to Success criterion for Epistemic No Applicable pressure relief RCS/RPV pressure transients prevention of RPV requiring pressure relief. overpressure (Note Application has no impact on the Usually, there is sufficient that uncertainty potential for SSCs providing capacity to accommodate the exists in both the pressure relief to fail, and does not pressure transient. However, determination of the affect the success criteria for in some scenarios, failure of global CCF values prevention of RPV overpressure.

adequate pressure relief can that may lead to RPV The uncertainty associated with be a consideration. Various overpressure and this item would have similar assumptions can be taken on what is done with impacts on the base case and the the impact of inadequate the subsequent RPV conditional case. Therefore, this pressure relief. overpressure item is not a key source of sequence modeling.) uncertainty for this application.

Systems Analysis (SY) 14 - Operability of Due to the scope of PRAs, System and accident Epistemic Yes Applicable equipment in beyond scenarios may arise where sequence modeling of According to the analyses presented in design basis equipment is exposed to available systems and Reference A.11.19 and due to the environments beyond design basis required support design of the Harris plant (with very environments (w/o room systems large equipment spaces), the HNP PRA cooling, w/o component only models the chillers for the cooling, w/ deadheading, in Charging/Safety Injection Pump (CSIP) the presence of an un-isolated Rooms and the Switchgear Rooms.

Upon a total loss of HVAC, the ability LOCA in the area, etc.). of operators to apply procedural steps to open doors and provide circulation from adjacent spaces as needed is credited. A sensitivity was performed assuming the operators fail to take action to provide emergency cooling to these areas. An examination of the delta risk cut sets indicated a CDF increase of >1E-06 and a LERF increase of >1E-08. Keep in mind however, this is for a total loss of HVAC only and A-27

does not involve LOCAs or steam line break scenarios in these areas.

Nevertheless, this is considered to be a key source of uncertainty.

Table A.10.1 Uncertainties Considered from EPRI 1016737 Against Extended AOT LAR Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44)

Human Reliability Analysis (HR) 15 - Credit For ERO Most PRAs do not give much, if System or accident Epistemic No Applicable any credit, for initiation of the sequence modeling with Emergency Response incorporation of HFEs and Although the Harris PRA and Organization (ERO), including HEP value determination associated HRA analysis do not actions included in plant-specific in both the Level 1 and take credit for ERO support and SAMGs and the new B5b Level 2 models staff per se, they do credit the use mitigation strategies. The of FLEX in the internal events additional resources and model. However, scenarios capabilities brought to bear via involving FLEX do not involve the ERO can be substantial, ESCW. A sensitivity was performed especially for long-term events. with FLEX removed from the model which resulted in a very small increase in total delta risk.

Therefore, this is not a key source of uncertainty for this application.

Internal Flooding (IF) 16 - Piping failure mode One of the most important, and Likelihood and Epistemic No Applicable uncertain, inputs to an internal characterization of flooding analysis is the internal flooding sources As shown in the internal flooding frequency of floods of various and internal flood event delta risk determination, this magnitudes (e.g., small, large, sequences and the application is not sensitive to pipe catastrophic) from various sources (e.g., clean water, timing associated with breaks. This is not a key source of untreated water, salt water, human actions involved uncertainty for this application.

etc.). in flooding mitigation.

EPRI has developed some data, but the NRC has not formally endorsed its use.

A-28

Table A.10.1 Uncertainties Considered from EPRI 1016737 Against Extended AOT LAR Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44)

LERF Analysis (LE) 17 - Core melt arrest Typically, the treatment of LERF / Level 2 Epistemic No Applicable in-vessel core melt arrest in-vessel has containment event tree been limited. However, sequences This application does not affect the recent NRC work has phenomenological uncertainty indicated that there may be present in the treatment of core more potential than melt arrest. The uncertainty previously credited. An associated with this item would example is credit for CRD in have similar impacts on the base BWRs. case and the conditional case.

Therefore, this item is not a key source of uncertainty for this application.

18 - Thermally induced NRC analytical models and LERF / Level 2 Epistemic No Applicable failure of hot leg/SG research findings continue to containment event tree tubes - PWRs show that a thermally sequences As described in EPRI TR-1016737, induced steam generator a TISGTR is assumed to occur in tube rupture (TISGTR) is cases with RCS pressure high, a more probable than steam generator depressurized predicted by the industry. and an RCP running. This approach There is a need to come to will have an equal impact on the agreement with NRC on the base and conditional cases. This thermal hydraulics modeling issue is not a key source of model of TI SGTR. uncertainty for this application.

A-29

Table A.10.1 Uncertainties Considered from EPRI 1016737 Against Extended AOT LAR Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) 19 - Vessel failure The progression of core melt LERF / Level 2 Epistemic No Applicable mode to the point of vessel failure containment event tree remains uncertain. Some sequences This application does not affect the codes (MELCOR) predict that uncertainty in core melt even vessels with lower head progression. The uncertainty penetrations will remain associated with this item would intact until the water has have similar impacts on the base evaporated from above the case and the conditional case.

relocated core debris. Other Therefore, this item is not a key codes (MAAP), predict that source of uncertainty for this lower head penetrations application.

might fail early. The failure mode of the vessel and associate timing can impact LERF binning, and may influence HPME characteristics especially for some BWRs and PWR ice condenser plants).

20 - Ex-vessel cooling The lower vessel head of LERF / Level 2 Epistemic No Applicable of lower head some plants may be containment event tree submerged in water prior to sequences This application does not affect the relocation of core debris the feasibility of, or analysis of, ex-to the lower head. This vessel cooling of lower head. Thus, presents the potential for the the uncertainty associated with core debris to be retained in- this item would have similar vessel by ex-vessel cooling. impacts on the base case and the This is a complex analysis conditional case. Therefore, this impacted by insulation, item is not a key source of vessel design and degree of uncertainty for this application.

submergence.

A-30

Table A.10.1 Uncertainties Considered from EPRI 1016737 Against Extended AOT LAR Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) 21 - Core debris contact In some plants, core debris LERF / Level 2 Epistemic No Applicable with containment can come in contact with the containment event tree containment shell (e.g., some sequences This application does not affect the BWR Mark Is, some PWRs impact of molten core debris on including free- standing steel containment integrity. The containments). uncertainty associated with this Molten core debris can item would have similar impacts challenge the integrity of the on the base case and the containment boundary. Some conditional case. Therefore, this analyses have demonstrated item is not a key source of that core debris can be cooled uncertainty for this application.

by overlying water pools.

22 - ISLOCA IE ISLOCA is often a significant ISLOCA initiating event Aleatory No Applicable Frequency contributor to LERF. One key sequences Determination input to the ISLOCA analysis This application has no impact on are the assumptions related the ISLOCA analysis. The delta cut to common cause failure of sets associated with the base case isolation valves between the and the conditional case were RCS/RPV and low pressure examined. There were no piping. There is no consensus contributions from ISLOCAs.

approach to the data or Therefore, this item is not a key treatment of this issue. source of uncertainty for this Additionally, given an application.

overpressure condition in low pressure piping, there is uncertainty surrounding the failure mode of the piping.

A-31

Table A.10.1 Uncertainties Considered from EPRI 1016737 Against Extended AOT LAR Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) 23 - Treatment of The amount of hydrogen Level 2 containment Epistemic No Applicable Hydrogen combustion burned, the rate at which it is event tree sequences in BWR Mark III and generated and burned, the The potential for hydrogen PWR ice condenser pressure reduction mitigation combustion in containment is plants credited by the suppression addressed realistically, considering pool, ice condenser, plant design. In addition, HNP is structures, etc. can have a not an ice condenser plant.

significant impact on the Therefore, this issue is not a source accident sequence of model uncertainty for this progression development. application.

24 - Basis for HEPs There is not a consistent System or accident Epistemic No Applicable method for the treatment of sequence modeling with pre-initiator and post- incorporation of HFEs The HEPs to start the chilled water system (auto-start failure) and for initiator human errors. and HEP value operators to implement alternate However, human failures determination means of cooling to the CSIP and events are typically significant switchgear rooms (i.e., portable contributors to CDF and LERF. fans) have a small contribution (i.e., Fussell-Vesely) to the risk results.

Existing HFEs were developed in accordance with industry-accepted methods and have been peer reviewed and found acceptable. Any uncertainty in human error probabilities would have similar impacts on the base case and on the conditional case. Thus, this topic does not present a key source of uncertainty for this application.

A-32

Table A.10.1 Uncertainties Considered from EPRI 1016737 Against Extended AOT LAR Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) 25 - Treatment of HFE There is not a consistent Quantification of Epistemic No Applicable dependencies method for the treatment of dependent human potentially dependent post- errors The HEP to start the chilled initiator human errors. SPAR water system (auto-start models do not generally failure) is not a key source of include dependencies. uncertainty for this application.

Hence, its dependency on other HEPs is not a key source of uncertainty as well.

HFE dependency is assessed in the model using industry-accepted methods. This application is not sensitive to the method of assessing dependency of HFEs.

Thus, this topic is not a key source of uncertainty for this application.

26 - Intra-system Common cause failures have CCF data values and Epistemic No Applicable common cause events been shown to be important associated system contributors in PRAs. As limited model representations Common cause failures involving plant-specific data is available, the ESCW chillers are not generic common cause factors dominant contributors to risk.

are commonly used. The delta cut sets associated with Sometimes, plant-specific the base case and the conditional evidence can indicate that the case were examined. There were generic values are only very small contributions inappropriate. (~2%) from common cause events. Therefore, this is not a key source of uncertainty.

A-33

Table A.10.2 Internal Events Model Uncertainty Evaluation for HNP ESCW TS AOT Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) 1 - Treatment of break WCAP-16304-P identifies the LOCA sequences Aleatory No Applicable.

location break location for a LOCA as a key source of uncertainty. LOCA break locations are not From a deterministic impacted by this application. The viewpoint, the location of a uncertainty associated with this break (particularly the smaller item would have similar impacts breaks), is significant for the on the base case and the following reasons: conditional case. Therefore, this item is not a key source of (1) It determines the liquid level uncertainty for this application.

to which the RV can be refilled (2) It may affect the amount of injection flow reaching the vessel (3) It influences the extent of the inventory loss Breaks can occur at any location around the typical RCS.

While this is an obvious statement, many analyses used for setting success criteria are based on the limiting break location, i.e., at the bottom of the cold leg.

A-34

Table A.10.2 Internal Events Model Uncertainty Evaluation for HNP ESCW TS AOT Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) 2 - Approach to WCAP-16304-P identifies LOCA sequences Epistemic No Applicable.

selecting success success criteria as a source of criteria uncertainty in LOCA analyses. The LOCA success criteria are based on plant-specific A review of LOCA models procedures and thermal-hydraulic within PRA models indicates analyses performed using MAAP that the success criteria (Reference A.11.34). Since the assignment encompasses success criteria for the LOCA many assumptions. Success analysis is based on realistic plant criteria are those elements parameters and operating that must be in place such procedures, this issue is not that when a plant challenge considered a key source of model occurs, pre-planned measures uncertainty for this application.

can be executed to bring the plant to a safe shutdown state.

3 - Sump blockage WCAP-16304-P identifies LOCA sequences Epistemic No Applicable.

sump blockage as a source of uncertainty in LOCA analyses, The containment sumps could fail since modeling sump blockage during the recirculation phase of is outside the scope of current operations due to clogging. This is industry design efforts. addressed in the RHR system model (Ref. A.11.33, Appendix A.2).

Nevertheless, this is a concern for LOCA sequences and not for ESCW.

A sensitivity was performed with the sump clogged. As a result, there was only a very modest increase in the delta risk. Therefore, this item is not a key source of uncertainty for this application.

A-35

Table A.10.2 Internal Events Model Uncertainty Evaluation for HNP ESCW TS AOT Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) 4 - Boric acid WCAP-16304-P identifies boric LOCA sequences Epistemic No Applicable.

precipitation acid precipitation as an unknown source of The proposed ESCW TS change uncertainty in LOCA analyses. would not add or reduce any Simplification is typically uncertainty related to boric acid made based on ECCS pump precipitation. In addition, boric and valve reliability. acid precipitation is discussed in Uncertainty is in where the the CSIP system notebook (Ref.

boric-acid precipitate forms, A.11.33, App. A.1). To avoid this e.g., upper plenum or fuel problem, at approximately 6.5 bundles. hours after an accident requiring safety injection, the operator will manually shift the CSIP discharge flowpath from the RCS cold legs to the hot legs by operating the header isolation valves.

Thereafter, hot and cold leg injection is alternated every 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . This topic is not a key source of uncertainty for this application.

A-36

Table A.10.2 Internal Events Model Uncertainty Evaluation for HNP ESCW TS AOT Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for this Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45)

Application (Ref. A.11.44) 5 - HVAC performance WCAP-16304-P identifies HVAC LOCA sequences Epistemic Yes Applicable performance as an unknown According to the analyses presented in source of uncertainty in LOCA Reference A.11.19 and due to the analyses. design of the Harris plant (with very large equipment spaces), the HNP PRA Loss of HVAC may impact model only requires the chillers for the operation of ECCS pump Charging/Safety Injection Pump (CSIP) motors. SSC sensitivity to HVAC Rooms and the Switchgear Rooms. In addition, the ability of operators to performance is a function of apply procedural steps to open doors PWR class, geographic location, and provide circulation from adjacent and season. spaces is credited in the PRA models.

Furthermore, the Switchgear Rooms' cooling loads during the summer months are low enough such that, without operator action, the equipment would continue to meet the PRA success criterion. The Switchgear Rooms' cooling loads in the winter months are higher due to electric heating coil unit loading of the transformers and will require operator action to meet the PRA success criterion. A sensitivity was performed assuming the operators fail to take action to provide emergency cooling during the winter months. As a result, an examination of the delta risk cut sets indicated a CDF increase >1E-06/yr and a LERF increase >1E-08/yr.

Therefore, with the primary areas requiring cooling and the proper operator actions identified, a loss of HVAC is recognized as a key source of uncertainty for this application. The compensatory actions address this by having operators A-37

Table A.10.2 Internal Events Model Uncertainty Evaluation for HNP ESCW TS AOT Impact briefed on the proceduralized actions for recovering a loss of HVAC to the CSIP Rooms, Switchgear Rooms and the Auxiliary Relay Cabinet Room prior to exceeding the initial 72 hrs. of the AOT.

In addition, the fan used for the CSIP area cooling will be pre-staged and verified to be functional.

A-38

Table A.10.2 Internal Events Model Uncertainty Evaluation for HNP ESCW TS AOT Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for this Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45)

Application (Ref. A.11.44) 6 - Component Cooling WCAP-16304-P identifies CCW LOCA sequences Epistemic No Applicable.

water performance performance as an unknown source of uncertainty in LOCA The requirement for sump cooling analyses. during the recirculation mode of Failure of CCW can prevent core cooling is modeled in the PRA sump cooling and result in long (Reference A.11.35). The term containment failure with uncertainty associated with this consequent flashing of the sum item would have similar impacts on with resultant failure of ECCS the base case and the conditional pumps. case. Therefore, this item is not a key source of uncertainty for this application.

7 - Mini recirculation WCAP-16304-P identifies mini LOCA sequences Epistemic No Applicable.

valves recirculation lines as an unknown source of uncertainty The proposed ESCW TS change does in LOCA analyses. not impact operation of the mini-Pumps can operate dead- recirculation valves. In addition, the headed for a finite time. Many safety injection system analysis plants assume that dead headed includes the recirculation miniflow operation immediately results in lines within the model (Ref. A.11.33, component failure. For plants App. A.1). The uncertainty that do credit dead headed associated with this item would pump operation, the credit is have similar impacts on the base typically limited to times of between five and 30 minutes. case and the conditional case.

The size-range of LOCAs Therefore, this item is not a key affected by this failure varies. source of uncertainty for this application.

A-39

Table A.10.2 Internal Events Model Uncertainty Evaluation for HNP ESCW TS AOT Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for this Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45)

Application (Ref. A.11.44) 8 - Pressurizer PORVs A LOCA typically results in a Phase LOCA / SGTR modeling; Epistemic No Applicable.

are the only available A and Phase B isolation. The HRA (controllable) means of Phase B isolation results in the The proposed ESCW TS change does loss of instrument air and depressurizing the RCS not impact the pressurizer PORVs component cooling water to the following an initiating nor does it impact the pressurizer containment. This, in turn, results event. in the unavailability of the sprays. In addition, according to Ref.

pressurizer spray valves and the A.11.36, the inclusion of pressurizer reactor coolant pumps needed to spray to mitigate SGTR events depressurize the containment potentially decreases the CDF by using the pressurizer sprays. approximately 0.3%. (It should be Although the pressurizer sprays noted that utilization of pressurizer are unavailable for LOCA events, sprays is not always successful they are still available for steam depending on the non-condensable generator tube rupture events.

gases present in the system which For SGTRs, the pressurizer sprays are the preferred method of tends to negate the effectiveness of depressurizing the RCS. Not spray.)

including the pressurizer sprays in the SGTR sequences reduces the Therefore, this topic is not a key reliability of the RCS source of uncertainty for this depressurization and increases application.

the likelihood of creating a LOCA because of a stuck open pressurizer PORV.

A-40

Table A.10.2 Internal Events Model Uncertainty Evaluation for HNP ESCW TS AOT Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for this Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45)

Application (Ref. A.11.44) 9 - Realignment of the The swing CSIP can be LOCA modeling; HRA Epistemic No Applicable.

swing CSIP cannot be realigned in about 25 credited for emergency minutes, (unless it was The proposed ESCW TS change boration or for a loss of previously aligned to the impacts the modeling of the CSIPs RCP seal cooling. same train as the failed only in the sense that it provides pump then it would be cooling to the Charging / Safety available then 10 minutes). Injection Pump (CSIP) Rooms.

This is too long to be useful Hence, realignment of the swing for emergency boration and CSIP is not a key source of for loss of seal cooling (25- uncertainty for this application.

minute case). Of course, the action would only be applied in sequences caused by a loss of both normally aligned CSIP pumps.

A-41

Table A.10.2 Internal Events Model Uncertainty Evaluation for HNP ESCW TS AOT Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for this Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45)

Application (Ref. A.11.44) 10 - CCF of CSIPs to With realignment of the LOCA modeling; HRA Epistemic No Applicable.

start can only occur swing CSIP now taking about after an undervoltage (10 minutes if aligned with The proposed ESCW TS change on the 6.9 kV bus the failed pump and power) impacts the modeling of the CSIPs corresponding to the otherwise 25 minutes, it is only in the sense that it provides operating pump. not unreasonable to cooling to the Charging / Safety consider situations in which Injection Pump (CSIP) Rooms.

the swing CSIP is substituted Hence, the issue of creating a CCF for a failed (normally grouping of the swing CSIP and operating) CSIP. Although the standby CSIP is not a key the swing CSIP and the source of uncertainty for this standby CSIP may not be application.

started simultaneously, they might be called upon to start within a 25 minutes of each other. In this case, a CCF grouping of the swing CSIP and the standby CSIP should be considered. However the CCF factor for a pump that is normally not aligned or running would not be expected to have a running failure in a short period of time. Thus the CCF factor if used would be much smaller than the normally running pumps.

A-42

Table A.10.2 Internal Events Model Uncertainty Evaluation for HNP ESCW TS AOT Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) 11 - Based on Administrative controls and RHR modeling Epistemic No Applicable.

administrative post-entry inspections are controls limiting the expected to prevent any The proposed ESCW TS change amount of loose loose materials from being does not impact modeling of the items allowed into left in containment. RHR system or the containment containment and on However, while debris sump. It also does not impact the inspections associated with a LOCA itself Foreign Materials Exclusion (FME) conducted after is not explicitly addressed, program. As indicated for Item # 3 containment entries the conservative modeling of above, a sensitivity was (per the Technical sump clogging performed with the sump Specifications), sump accommodates the clogged. As a result, there was clogging due to probability of this debris only a very modest increase in the debris intrusion on clogging the containment delta risk. Therefore, this item is the sump screens is sumps. not a key source of uncertainty for expected to be a this application.

non-significant contributor to RHR system unavailability.

Clogging of the sumps is conservatively addressed in the model.

A-43

Table A.10.2 Internal Events Model Uncertainty Evaluation for HNP ESCW TS AOT Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) 12 - Procedurally This is based on an assumption Small LOCA events; Aleatory No Applicable.

required operator that the volume of the N2 in HRA actions to close the the accumulators is not The proposed ESCW TS change three accumulator sufficient to cause blockage of does not impact modeling of the discharge valves and an RCS leg. This also assumes accumulators nor does it impact lock the breakers no further leakage from the small LOCA events.

open in order to N2 system into the prevent injecting accumulators (i.e., the Therefore, this topic is not nitrogen (N2) into normally closed valves considered to be a key source of the RCS are not supplying the N2 do not leak). uncertainty for this application.

assumed to be This is not an issue for large required in the PSA.

and medium LOCAs where any N2 is likely to be swept out of the break. However, for small LOCAs or transients in which the RCS must be depressurized to get to shut down conditions, the insertion of N2 into the RCS could become problematic.

The operator action should be retained in these sequences.

13 - The turbine- This is a conservative LOOP recovery Aleatory No Applicable.

driven auxiliary assumption used to simplify models; HRA feedwater pump the model.

(TDAFWP) is Figure D.18, p. 67 in Appendix conservatively D shows SG dry out times assumed to ranging from 43 to 56 immediately fail if no minutes (with and without flow is available to A-44

Table A.10.2 Internal Events Model Uncertainty Evaluation for HNP ESCW TS AOT Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) steam generators B RCPs operating, respectively). The proposed ESCW TS change and C (i.e., the steam Use of the TDAFWP would does not impact steam generator flow available as the shorten these dry out times dry out. Furthermore, per Ref.

generator dries out is somewhat, but its use could A.11.36, taking credit for neglected). still provide considerable providing flow to SG A with failed cooldown and supply to SGs B and C would only depressurization of the extend the time for requiring feed secondary and RCS before and bleed cooling by there is insufficient steam to approximately 15 minutes. Hence, operate the TDAFWP. changing this assumption would not provide a major effect on the HRA for implementing feed and bleed and it is therefore not important to the analysis of SG dry out time (Ref. A.11.35).

Therefore, this topic is not considered to be a key source of uncertainty for this application.

14 - ESW is unlikely Check valve 1SW-50 and the NSW / ESW system Aleatory No Applicable.

to be failed by the discharge MOV for the in- modeling NSW failing to isolate service NSW pump would have Per Ref. A.11.36, backleakage when an NSW pump to fail open for a diversion through 1SW-50 has been is unavailable. path to be created should the measured and determined to be pump become unavailable. insignificant when compared to This was believed to be normal ESW pump flow of 20,000 statistically insignificant. gpm. ESW loads require approximately 15,000 gpm.

However, check valve 1SW-50 is not checked for Therefore, this topic is not backleakage; therefore, it has considered to be a key source of a 40-year exposure time.

uncertainty for this application.

Thus, the probability of a A-45

Table A.10.2 Internal Events Model Uncertainty Evaluation for HNP ESCW TS AOT Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) diversion path created by backleakage through 1SW-50 and a failed open MOV is not statistically insignificant.

15 - Loss of water in Because there is no safety- CCW system Aleatory No Applicable.

any part of the CCW related makeup to the CCW modeling; HRA surge tank fails the system/surge tank, a loss of The proposed ESCW TS change entire CCW system inventory from the surge tank does not impact the CCW system.

due to a loss of CCW was assumed to result in In addition, per Ref. A.11.36, the inventory. system failure. probability of CCW surge tank failure is small enough so that it is The CCW surge tank contains not included in a CDF truncated at a baffle that extends to 40 1.0E-11 or lower. Other piping percent of the tank's height.

ruptures in the CCW system are As noted in Appendix A.10, modeled such that affected train Sect. A.10.1.2.2, 3rd is failed and the plant trips.

paragraph, p. 5, the baffle Recovery is available using either ensures that ". . . a single the unaffected train or the swing passive failure in the CCW pump.

system will not result in a loss of suction to both trains, Therefore, this topic is not assuming the two trains are considered to be a key source of separated by the isolation uncertainty for this application.

valves." Thus, the potential for one train continuing to operate following a loss of water from the tank should be included in the model.

16 - The parallel A geometric average was used IA system modeling Aleatory No Applicable.

paths for IA to the to calculate the average turbine building exposure time for each IA header are not train.

tested, so an average This modeling approach is not A-46

Table A.10.2 Internal Events Model Uncertainty Evaluation for HNP ESCW TS AOT Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) exposure time of 5.7 consistent with industry The proposed ESCW TS change months applies, practices. The typical does not impact the IA system.

representing an approach is to use flags to There are no system interfaces average of a 40-year identify the in-service and between ESCW and IA.

plant life exposure standby paths. A 20-year time and a 24-hour exposure time would then be Therefore, this topic is not exposure time. used for components in the considered to be a key source of standby path. uncertainty for this application.

17 - It is not The operator failure Loss of AC power Aleatory No Applicable.

necessary to model probability is assumed to be system modeling the operator failing "much less likely" than other The proposed ESCW TS change to close the breaker failure modes. does not impact the AC power for buses 1A1 and system. Also, per Ref. A.11.36, Component failures 1B1 from the control this is not an area of uncertainty.

associated with the loss of room. This is more of a documentation power to buses 1A1 and 1B1 issue and it should be revised for are on the order of E-5 or E-6.

clarity.

Failure of an operator to close the breakers would be Therefore, this topic is not expected to be on the order considered to be a key source of of E-3. Thus, omission of the uncertainty for this application.

operator error based on insignificant probability is not correct.

A-47

Table A.10.2 Internal Events Model Uncertainty Evaluation for HNP ESCW TS AOT Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) 18 - The CCFs for The method used to determine the CCW system modeling Epistemic No Applicable.

CCW pumps must be common cause factor for the CCW derived using data pumps using Table C.19, while The proposed ESCW TS change necessary, has a number of arbitrary from other pumps or questionable elements. First, an does not impact the CCW system.

because of the assumption is made that the pumps HVAC maintains temperature scarcity of CCF data used in the development of the CCF conditions within the range on CCW pumps. factor are similar to the Harris CCW required for operation of the CCW pumps (although there is a "factor" included in the analysis for pumps. Due to the size of the applicability). The variety of these room containing the three pumps, pumps, from RHR to service water to excessive temperatures are not salt water, makes it unlikely that more expected in the event of a failure than a few are close enough in design to draw meaningful conclusions. of HVAC. Therefore, loss of HVAC Further, three of the eleven events is not modeled for CCW.

come from one system at one plant (clearly there were design and/or Therefore, this topic is not operational issues there that may not relate very well to other sites). considered to be a key source of Secondly, the development of factors uncertainty for this application.

for applicability to the Harris CCW system and to an initiating event is based completely on engineering judgment of the analysts. Third, it is not clear whether systems engineers or operators had any input into the development of these factors. Fourth, it is not clear how the values for the "Totals" and "Adjustment Factor" in Table C.20 were derived. Lastly, there is no information about the uncertainty associated with the development of these factors. There is no way of being able to tell the difference between the CCF factor developed using this method and the CCF developed for a component with a large amount of operating history.

Given the number of questions a A-48

Table A.10.2 Internal Events Model Uncertainty Evaluation for HNP ESCW TS AOT Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) sensitivity analysis on the CCW CCFs would be useful.

19 - For HNP the HNP and only one other Success criteria; HRA Epistemic No Applicable.

additional makeup Westinghouse 3 loop plant source using the low- considered use of RCS The proposed ESCW TS change pressure injection depressurization and use of does not impact the ability to system is considered. the LHSI as a makeup source depressurize the steam for SGTR sequences. The basis generators to use low pressure for the other W3LP plants not injection.

using LHSI is not provided, but would be of interest. Therefore, this topic is not considered to be a key source of While depressurization of the uncertainty for this application.

RCS is an option, not much is said about performing this action while trying to manage a tube rupture in one of the steam generators. It is not clear whether the depressurization would help or complicate the process of isolating the ruptured steam generator (discussion of the depressurization is presented without any reference to potential effects on the ruptured steam generator).

A-49

Table A.10.2 Internal Events Model Uncertainty Evaluation for HNP ESCW TS AOT Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) 20 - ATWS scenarios The assumption that for LOCA Transient analysis Epistemic No Applicable.

are evaluated for sequences the reactor is shut non-LOCA transients, down by the boron provided The proposed ESCW TS change which are assumed by the SI systems is does not impact non-LOCA to be mitigated by inconsistent with WCAP- transients or ATWS mitigation emergency boration 15955, as was stated in scenarios. ESCW is not from the flow from Section 4.8 of Section 4. systematically related to the the safety injection performance of ATWS mitigation This is a source of uncertainty systems. strategies.

and model limitations. WCAP-15955 does state the ATWS is Therefore, this topic is not considered. However, it does considered to be a key source of not discuss the effect of uncertainty for this application.

boration due to safety injection or if the RWST boron levels would be adequate for an ATWS mitigation. Additionally, it is clear that if at least some rods are inserted along with boron injection, the ATWS is mitigated. Thus, the normal ATWS probabilities do not apply. This issue could be an area of further investigation.

21 - Loss of Off-site The screening method has LOOP; high winds Aleatory No Applicable.

power frequency been used in the past and analysis uses a screening found acceptable. The proposed ESCW TS change method and also does not impact loss of offsite The lack of hurricane data assumes that power or high winds scenarios.

and due to the requirement hurricanes will not to shut down before wind affect Harris Plant Therefore, this topic is not speeds increase to hurricane considered to be a key source of strength and the number and A-50

Table A.10.2 Internal Events Model Uncertainty Evaluation for HNP ESCW TS AOT Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) different directions the off- uncertainty for this application.

site lines go to.

22 - The PORV failure This is a frequency based Fire modeling Aleatory No Applicable.

to reclose following upon generic data, no plant opening demand specific data for failure to Per Ref. A.11.36, this could be a reclose. significant contributor to CDF for fires. The proposed ESCW TS change does not impact the fire analysis or any plant scenarios requiring the PORV to reclose after opening.

Therefore, this topic is not considered to be a key source of uncertainty for this application.

23 - MAAP analysis This is based upon MAAP Success criteria; HRA Epistemic No Applicable.

indicates that the analysis and EOP actions.

plant can cooldown Even though this action is to LPI/RHR considered important per Ref.

conditions with a RCP A.11.36, the proposed ESCW TS seal LOCA without change does not impact plant HPI, if an aggressive cooldown to RHR conditions.

cooldown is used and started early enough Therefore, this topic is not considered to be a key source of uncertainty for this application.

24 - The HNP PSA This is an Industry consensus LOCA analysis Epistemic No Applicable.

model uses the WOG model.

2000 RCP seal failure model and it has assumed RCP seal leakage every time A-51

Table A.10.2 Internal Events Model Uncertainty Evaluation for HNP ESCW TS AOT Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44) both Seal Injection The proposed ESCW TS change and Thermal Barrier does not directly impact RCP seal cooling are lost. LOCA analysis; however, ESCW impacts the CSIPs only in the sense that it provides cooling to the Charging / Safety Injection Pump (CSIP) Rooms. The ability of operators to apply procedural steps to open doors and provide circulation from adjacent spaces is credited when the chillers are unavailable. Further, RCP seal cooling is an industry consensus model.

Therefore, this topic is not considered to be a key source of uncertainty for this application.

25 - Generic data and This has been the position Data Epistemic No Applicable.

component of the PSA model since the boundaries are start of the industry The proposed ESCW TS change assumed to be does not impact generic data and consistent component boundaries.

Therefore, this topic is not considered to be a key source of uncertainty for this application.

26 - Assumptions This is a new system and the System modeling Epistemic No Applicable.

regarding operation, EC is the only available testing frequency, resource.

etc. for the Alternate Seal Injection (ASI) and Dedicated A-52

Table A.10.2 Internal Events Model Uncertainty Evaluation for HNP ESCW TS AOT Impact Issue Description Issue Characterization NUREG-1855 type of Key Source of Applicability and Resolution for Topic Discussion of Issue Part of Model Affected Uncertainty Uncertainty (Yes/No)

(Ref. A.11.45) this Application (Ref. A.11.44)

Shutdown Diesel Per Ref. A.11.36, this item was Generator (DSDG) retained as an area of uncertainty systems were made until further documentation is due to information available; however, the proposed not being available. ESCW TS change does not impact any of the systems mentioned.

Therefore, this topic is not considered to be a key source of uncertainty for this application.

A-53

Table A.10.3 Fire Model Uncertainty Evaluation for HNP ESCW TS AOT Impact (from Ref. A.11.42)

NUREG-1855 type Key Source of Applicability and Resolution Category Item of Uncertainty Uncertainty (Yes/No) Conservatism (Ref. A.11.45) for this Application (Ref. A.11.44)

Bounding values from Epistemic No Yes.

1 - Source HRR Conservatism has already NUREG/CR-6850 were typically been applied. For cutsets used for the 98th percentile It does not seem likely involving the ESCW Chillers, file based on the HRR case. For that the actual source the delta risk will be a limited number of sources configurations could essentially the same since (cabinets) these values were support the default HRRs the base case and AOT will adjusted based on fire be both be affected.

modeling insights. Transient Therefore, no impact on this HRRs were also adjusted down application.

in areas with stricter transient controls.

Closed cabinet treatment for Epistemic No Yes 2 - Source HRR Conservatism has already MCCs. HNP assumes MCCs are been applied. For cutsets closed sources, however The data for the involving the ESCW Chillers, guidance indicates that 480 V guidance is the delta risk will be MCCs can experience energetic interpreted essentially the same since faults which can create conservatively. At the base case and AOT will openings to support fire best, only a small be both be affected.

growth. portion of the MCC fires Therefore, no impact on this would lead to an open application.

cabinet situation. A 0.1 probability was applied to account for this.

No credit is given for the Epistemic No Yes 3 - Source HRR Conservatism has already incipient / smoldering stages of profile been applied. For cutsets fire growth. Allowing for these involving the ESCW Chillers, phases will provide more the delta risk will be time for manual essentially the same since suppression credit.

the base case and AOT will be both be affected.

Therefore, no impact on this application.

A-54

Table A.10.3 Fire Model Uncertainty Evaluation for HNP ESCW TS AOT Impact (from Ref. A.11.42)

NUREG-1855 type Key Source of Applicability and Resolution Category Item of Uncertainty Uncertainty (Yes/No) Conservatism (Ref. A.11.45) for this Application (Ref. A.11.44)

Most fires use a 12-minute Epistemic No Yes 4 - Source HRR Conservatism has already ramp to peak HRR, 8 minutes profile been applied. For cutsets at peak, and 19 minutes decay. Consideration of other involving the ESCW Chillers, No methodology for factors will likely shorten the delta risk will be consideration of combustible duration or reduce peak essentially the same since loading and other factors is HRR for most scenarios.

the base case and AOT will provided.

be both be affected.

Therefore, no impact on this application.

Use multi-point fires is not Epistemic No Yes 5 - Target Conservatism has already based on target importance Selection been applied. For cutsets More than two points or involving the ESCW Chillers, more the delta risk will be targeted selection of the essentially the same since two points may reduce the base case and AOT will CDDP.

be both be affected.

Therefore, no impact on this application.

Scenarios with the FM target Epistemic No Yes 6 - Target Conservatism has already set only uses single point fires.

Selection been applied. For cutsets Using multi-point fires involving the ESCW Chillers, may reduce CCDPs the delta risk will be essentially the same since the base case and AOT will be both be affected.

Therefore, no impact on this application.

A-55

Table A.10.3 Fire Model Uncertainty Evaluation for HNP ESCW TS AOT Impact (from Ref. A.11.42)

NUREG-1855 type Key Source of Applicability and Resolution Category Item of Uncertainty Uncertainty (Yes/No) Conservatism (Ref. A.11.45) for this Application (Ref. A.11.44)

ZOI for target damage is based Epistemic No Yes 7 - Target Conservatism has already on 400°F due to Kerite cable.

Selection been applied. For cutsets 650°F may be more Reduced target sets due involving the ESCW Chillers, appropriate to smaller ZOIs should the delta risk will be reduce CCDPs.

essentially the same since the base case and AOT will be both be affected.

Therefore, no impact on this application.

Target damage is based on Epistemic No Yes 8 - Damage Time Conservatism has already 400°F due to Kerite cable.

been applied. For cutsets 650°F may be more A higher damage involving the ESCW Chillers, appropriate threshold will provide the delta risk will be more time for essentially the same since suppression.

the base case and AOT will be both be affected.

Therefore, no impact on this application.

Epistemic No Yes 9 - Damage Time Target damage does not credit Conservatism has already conduit been applied. For cutsets More time should be involving the ESCW Chillers, available for suppression the delta risk will be due to the use of essentially the same since conduit.

the base case and AOT will be both be affected.

Therefore, no impact on this application.

A-56

Table A.10.3 Fire Model Uncertainty Evaluation for HNP ESCW TS AOT Impact (from Ref. A.11.42)

NUREG-1855 type Key Source of Applicability and Resolution Category Item of Uncertainty Uncertainty (Yes/No) Conservatism (Ref. A.11.45) for this Application (Ref. A.11.44)

Epistemic No Yes 10 - Time to HGL See target selection and Conservatism has already Extending the time to damage time items been applied. For cutsets the first tray igniting will involving the ESCW Chillers, extend the time to HGL the delta risk will be formation and provide essentially the same since more time for the base case and AOT will suppression.

be both be affected.

Therefore, no impact on this application.

The cable tray growth model in 11 - Time to HGL Epistemic No Yes Conservatism has already NUREG/CR-6850 has limited been applied. For cutsets applicability and appears to be involving the ESCW Chillers, conservative when applied the delta risk will be outside the limits.

essentially the same since the base case and AOT will be both be affected.

Therefore, no impact on this application.

Epistemic No Indeterminate 12 - Time to HGL Fire spread within a cable tray Even though Ref. A.11.42 is offset by decay. has deemed this as Can vary depending of indeterminate, decay is not factors involved.

credited in the fire model.

This would therefore introduce some conservatism in the model.

Hence, no impact on this application.

A-57

Table A.10.3 Fire Model Uncertainty Evaluation for HNP ESCW TS AOT Impact (from Ref. A.11.42)

NUREG-1855 type Key Source of Applicability and Resolution Category Item of Uncertainty Uncertainty (Yes/No) Conservatism (Ref. A.11.45) for this Application (Ref. A.11.44)

NUREG/CR-6850 methodology Epistemic No Yes 13 - Non- Conservatism has already subtracts brigade response Suppression been applied. For cutsets time from the timeline for The suppression data involving the ESCW Chillers, suppression appear to contain the the delta risk will be brigade response; essentially the same since therefore, the NSP the base case and AOT will results are high.

be both be affected.

Therefore, no impact on this application.

HNP estimate actual brigade Epistemic No Yes/No 14 - Non- A sensitivity was performed response times to be 50% of Suppression with the scenario event drill times Drill times do not include frequencies (SEFs) adjusted all factors representative to the frequency of that fire of actual fire response occurring without being times. This has no suppressed (i.e., worst case).

impact if brigade is The delta CDF went from included in suppression 8.10E-07 to 8.59E-07 and curves.

the delta LERF went from 1.20E-08 to 1.24E-08.

Therefore, the brigade response times do not present a significant impact for this application.

A-58

Table A.10.3 Fire Model Uncertainty Evaluation for HNP ESCW TS AOT Impact (from Ref. A.11.42)

NUREG-1855 type Key Source of Applicability and Resolution Category Item of Uncertainty Uncertainty (Yes/No) Conservatism (Ref. A.11.45) for this Application (Ref. A.11.44)

Epistemic No Yes 15 - Non- Incipient Detection Credit Conservatism has already Suppression been applied. For cutsets It can be conservatively involving the ESCW Chillers, assumed that 90% of the the delta risk will be fires will be prevented essentially the same since during the incipient the base case and AOT will phase and thereby be both be affected.

prevent any damage Therefore, no impact on this beyond the ignition site.

application.

The actual success percentage is expected to be much higher.

Epistemic No Yes 16 - Power dependencies for Conservatism has already Quantification spurious operation been applied. For cutsets In some cases, it appears involving the ESCW Chillers, that the power may not the delta risk will be be available to support essentially the same since some fire-induced the base case and AOT will spurious events.

be both be affected.

Therefore, no impact on this application.

Epistemic No Yes 17 - HRA Screening Conservatism has already Quantification been applied. For cutsets No credit is given to involving the ESCW Chillers, OMAs [Operator Manual the delta risk will be Actions] and most ex-essentially the same since control room actions.

the base case and AOT will be both be affected.

Therefore, no impact on this application.

A-59

Table A.10.4 High Winds Model Uncertainty Evaluation for HNP ESCW TS AOT Impact (from Ref. A.11.32)

NUREG-1855 type of Key Source of Uncertainty Description Uncertainty Uncertainty (Yes/No) Uncertainty Characterization (Ref. A.11.44) (Ref. A.11.45) 1 - High wind initiating events were Epistemic No Full integration can introduce conservatisms with determined on the basis of discrete both the high-confidence and low-confidence intervals along the mean high wind hazard curves. Using the mean hazard curve hazard curve. Full integration of the therefore produces a more reasonable result.

high wind hazard curve with the Further, using full integration should not affect fragility curves was not performed. delta risk calculations since conservatisms introduced on the base case would also be introduced on the AOT case. This uncertainty therefore has no impact on this application.

2 - The hazard curves are based on a Epistemic No Any conservatisms or non-conservatisms combination of high wind data and introduced by this methodology would affect both expert opinion. the base case and the AOT case. This uncertainty therefore has no impact on this application.

3 - The quantification engine applies to Epistemic No Even though the point estimate could be a min-cut, upper bound approximation overestimated, this should not affect delta risk of the point estimate based on the calculations since conservatisms introduced on generated cut sets. At higher wind the base case would also be introduced on the speed intervals, the fragility values AOT case. This uncertainty therefore has no approach 1.0 and min-cut, upper bound impact on this application.

estimation is not sufficient to provide accurate results. The point estimate could be overestimated.

A-60

Table A.10.4 High Winds Model Uncertainty Evaluation for HNP ESCW TS AOT Impact (from Ref. A.11.32)

NUREG-1855 type of Key Source of Uncertainty Description Uncertainty Uncertainty (Yes/No) Uncertainty Characterization (Ref. A.11.44) (Ref. A.11.45) 4 - The fragilities of some equipment Epistemic No Any conservatisms or non-conservatisms were quantified on the assumption of introduced by using state of knowledge versus mutual independence of the high wind mutual independence would affect both the base impact. There may be some state of case and the AOT case. This uncertainty therefore knowledge correlation issues that has no impact on this application.

could impact this assumption.

5 - The CAFTA computer code and suite Epistemic No This is acceptable. No impact on this application.

of codes is used to quantify the PRA model. Therefore, the limitations and assumptions associated with the CAFTA suite of codes applies to the HNP HWPRA quantification process.

6 - Some SSCs are vulnerable to Epistemic No As stated, this conservatism has an insignificant crimping, such as the SRV stacks and impact on the CDF and LERF results. No impact on the PORVs stacks, due to high wind this application.

missile damage. In this PRA, the tornado missile fragilities used are based on a hit probability, which is conservative compared to modeling crimping. This conservatism has an insignificant impact on the CDF and LERF results because the probability of an SSC hit is relatively low and the components are not risk significant to the plant response.

A-61

Table A.10.5 Internal Flooding Model Uncertainty Evaluation for HNP ESCW TS AOT Impact (from Ref. A.11.43)

NUREG-1855 type Key Source of Assumption Is Impact on Description of Uncertainty Uncertainty (Yes/No) Basis No. (Ref. A.11.45)

Model Small?

(Ref. A.11.44)

Blocked HRAs were only Very likely operator would be able to 1 Epistemic No Yes considered for flood and HELB complete actions. No actions involving events. Spray events were electrical equipment were credited in assumed not to result in spray areas. No impact on this analysis.

conditions that would prevent operator actions from being performed.

Unless otherwise defined, the Pipe lengths for flood scenarios 2 Epistemic No Yes flooding frequency for spray with significant contribution to flood risk events was based on the were re-evaluated and actual lengths minimum value for the flood were used. No impact on this analysis.

frequency of 40 feet of pipe or the minimum pipe length in an area.

Propagation pathways were not Flood scenarios were reviewed and 3 Epistemic No Yes considered for creating a blocked there are multiple paths between areas.

path which would prevent an No impact on this analysis.

HRA recovery from being viable.

That is, blocked HRAs are only considered for the compartment that is experiencing the flood except for flood compartments FLC17b and FLC17i.

In situations in which the flood Flood scenarios with large impacts were 4 Epistemic No Yes scenario had differing reviewed and specific trains and lengths consequences depending on the of pipe were determined. No impact on train of the flooding source and this analysis.

there was no split in the fluid system frequency based on train, the frequency was partitioned equally between the trains.

A-62

Table A.10.5 Internal Flooding Model Uncertainty Evaluation for HNP ESCW TS AOT Impact (from Ref. A.11.43)

NUREG-1855 type Key Source of Assumption Is Impact on Description of Uncertainty Uncertainty (Yes/No) Basis No. (Ref. A.11.45)

Model Small?

(Ref. A.11.44)

Truncation of model results was 5 Truncation of Sequences Epistemic No Yes evaluated and met the change in CDF and LERF truncation limits No impact on this analysis.

Complementary terms are used for large 6 Rare event approximation Epistemic No Yes HEP probabilities to reduce over predicting risk No impact on this analysis.

One top model used. Mincut upper 7 Cutset merging Epistemic No Yes bound approach used. Combined cutsets are the result of different scenarios. No impact on this analysis.

Application of the state-of The uncertainty parameters are 8 Aleatory and No Yes knowledge correlation incorporated into the data tables and Epistemic evaluated using industry standard software. No impact on this analysis.

A-63

Table A.10.6 Key Assumptions for HNP ESCW TS AOT LAR Impact NUREG-1855 type of Key Source of Uncertainty Assumption Description Uncertainty Assumption Characterization (Yes/No) (Ref. A.11.45)

(Ref. A.11.44) 1 - Section 3/4.7.13 of plant Technical Epistemic No The intent of the proposed ESCW TS AOT change is to Specifications requires two loops of the extend the LCO for Modes 1, 2, 3, and 4. Thus, this ESCW system to be operable in Modes, 1, assumption is valid for the application. Furthermore, 2, 3 and 4. Therefore, this assessment will the PRA models supporting this application are at-be performed considering these modes of power models and, as such, would produce bounding /

operation only. There will be no conservative results for lower modes.

assessment for shutdown conditions.

2 - The assumed equipment unavailability Epistemic No This assumption merely complies with the procedural is set to the average test and requirements of AD-NF-NGO-0500 for providing PRA maintenance (T&M) as required per support for risk-informed applications.

Section 5.7.6(d) of Ref. A.11.8. This unavailability pertains to all SSCs except for those manipulated for the conditional case.

Epistemic Yes The F-Vs for the operator actions to provide an 3 - The HFEs to open doors and alternate means of cooling to the switchgear room /

implement portable fans as an alternate CSIP rooms is less than 1% for CDF and LERF (in the AOT means of cooling the switchgear room / conditional case) with a ESCW chiller out of service. This CSIP rooms were initially assumed to not indicates the recovery actions are highly reliable and significantly impact the risk results; successful. However, a sensitivity analysis in accordance however, a subsequent sensitivity with NUREG-1855, Rev. 1, in which the operators fail to analysis in accordance with NUREG-1855, provide alternate cooling prior to exceeding the Rev. 1 (Ref. A.11.44) indicates this is a key respective maximum allowable room temperatures source of uncertainty. rendered a delta CDF > 1E-06. Thus, this assumption is considered to be a key source of uncertainty.

A-64

A.11 REFERENCES A.11.1 USNRC Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3.

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Technical Specifications, Revision 1.

A.11.3 ASME, Internal Events PRA Standard, ASME RA-Sc-2007.

A.11.4 ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME and the American Nuclear Society, February 2009.

A.11.5 Regulatory Guide 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, USNRC, March 2009.

A.11.6 AD-NF-NGO-0502, Probabilistic Risk Assessment (PRA) Model Technical Adequacy, Rev. 2 A.11.7 3/4.7.13 of HNP Plant Technical Specifications, Amendment 154 A.11.8 AD-NF-NGO-0500, Corporate PRA Support for Emergent Issues and Risk-Informed Applications, Rev. 2 A.11.9 Shearon Harris Nuclear Power Plant, Unit No. 1, "Individual Plant Examination for External Events (IPEEE) Submittal," Carolina Power & Light Company, Docket No. 50-400/License No. NPF-63, June 1995 A.11.10 NUREG-1407, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events for Severe Accident Vulnerabilities, June 1991 A.11.11 HNP-13-031, "Flooding Hazard Reevaluation Report", US NRC, 3/12/2013 A.11.12 Duke Energy letter, "Seismic Hazard Evaluation and Screening Report, Shearon Harris Nuclear Power Plant, Unit 1", Docket No. 50-400, March 2014 A.11.13 USNRC, Support Document for Screening and Prioritization Results Regarding Seismic Hazard Re-Evaluations for Operating Reactors in the Central and Eastern United States, ADAMS Accession No. ML1413A126, May 2014 A.11.14 USNRC, Shearon Harris Nuclear Plant - Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights From the Fukushima Dai-ichi Accident (TAC NO. MF3952), December 2015 A.11.15 ASME/ANS RA-Sb-2013, Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, American Society of Mechanical Engineers, New York, NY, September 2013.

A.11.16 AD-WC-ALL-0200, On-Line Work Management, Rev. 13 A.11.17 AD-OP-ALL-201, Protected Equipment, Rev. 4 A.11.18 AD-NF-ALL-0501, Electronic Risk Assessment Tool (ERAT), Rev. 1 A.11.19 HNP-F/PSA-0058, "Appendix J - Room Heatup Analysis", Rev. 2 A-65

A.11.20 PWROG-15056-NP, Focused High Winds PRA Peer Review for Shearon Harris Nuclear Power Plant, Rev. 0 A.11.21 HNP-F/PSA-0069, "HNP - PSA Model Peer Review Resolution", Rev. 3 A.11.22 ABS Consulting, Harris Nuclear Plant, PRA Finding Level Fact and Observation Technical Review, Report No. R-3857458-2026, March 2017 A.11.23 NEI 05-04, Rev. 2, Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard A.11.24 HNP-F/PSA-0086, PRA Model Sequence Quantification, Rev. 3 A.11.25 HNP-F/PSA-0104, HNP Internal Flooding Scenario Consolidation, Rev. 0 A.11.26 NUREG / CR-6850 Final Report, Fire PRA Methodology for Nuclear Power Facilities, Vol. 2 A.11.27 ANSI / ANS 58.23-2007, Fire PRA Methodology A.11.28 NEI 07-12, Rev. 1, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines A.11.29 Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment Regarding Adoption of National Fire Protection Association Standard 805, "Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants" (TAC NO. MD8807), ADAMS Accession Numbers ML101750602, June 28, 2010 A.11.30 Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 133 to Renewed Facility Operating License No. NPF-63 Transition to a Risk-informed, Performance-based Fire Protection Program in accordance with 10 CFR 50.48(c) Carolina Power & Light Company, Shearon Harris Nuclear Power Plant, Unit No. 1 Docket No. 50-400 ADAMS Accession Numbers ML101750604, July 22, 2010 A.11.31 EPM, Inc., F&O Closeout by Independent Assessment of the Harris Nuclear Plant Fire PRA Model, Report No. R2919-002-001, Rev. 1, October 2017 A.11.32 HNP-F/PSA-0099, "HNP High Wind Probabilistic Risk Assessment (HWPRA): Plant Response Model", Rev. 0 A.11.33 HNP-F/PSA-0065, HNP PRA - System Notebooks, Rev. 8 A.11.34 HNP-F/PSA-0054, HNP PRA - Appendix F - Thermal-Hydraulic Analyses, Rev. 2 A.11.35 HNP-F/PSA-0052, HNP PRA - Appendix D - Success Criteria, Rev. 3 A.11.36 HNP-F/PSA-0080, HNP PRA - Appendix U - Assumptions and Uncertainty, Rev. 1 A.11.37 EPRI Report 1016737, Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments, December 2008 A.11.38 WCAP-16304-P, Strategy for Identifying and Treating Modeling Uncertainties in PRA Models:

Issues Concerning LOCA and LOOP, Revision 0 A.11.39 HNP High Winds PRA Peer Review - Resolutions and Comment Dispositioning A.11.40 EPRI TR-1019259, Fire Probabilistic Risk Assessment Methods Enhancements: Supplement 1 to NUREG/CR-6850 and EPRI 1011989, December 2009 A.11.41 NUREG-2180, Determining the Effectiveness, Limitations, and Operator Response for Very Early Warning Fire Detection Systems in Nuclear Facilities (DELORES-VEWFIRE), December 2016 A-66

A.11.42 HNP-F/PSA-0079, "Harris Fire PRA - Quantification", Rev. 3 A.11.43 HNP-F/PSA-0095, HNP Internal Flooding Quantification, Rev. 1 A.11.44 NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, Rev. 1, March 2017.

A.11.45 EPRI Report 1013491, Guideline for the Treatment of Uncertainty in Risk-Informed Applications, October 2006.

A.11.46 NEI Letter to NRC, Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (F&Os), March 27, 2017, (ADAMS Accession Number ML17086A431)

A.11.47 NRC Letter to Mr. Greg Krueger (NEI), U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 7-12, and 12-13, Close Out of Facts and Observations (F&Os), May 3, 2017, (ADAMS Accession Number ML17079A427)

A-67

Appendix B - Harris F&O Dispositions B-1

B.1 Internal Events, CDF All finding level F&O dispositions were determined to have been adequately addressed and are now considered CLOSED and no longer relevant to the PRA model (Ref. B.1).

B.2 Internal Flooding Per an independent review of open internal flooding F&O items (Ref. B.1), 27 were closed, 8 were partially closed and 2 remain open. These are discussed on the following pages:

B-2

HNP Internal Flooding F&O Disposition F&O ID: Associated SRs: Peer Review CC Assessment:

1-7 IFSN-A2 Finding (SR is Met)

==

Description:==

Flood alarms are identified in the HRA analyses presented in Table 7-2 of HNP-F/PSA-0094. However, the alarms are not specifically identified nor the alarms correlated to the flood source that causes the flooding event. Identification of alarms that are expected for each flood source that could release fluid in each area is required by the SR.

Resolution:

Per the suggested resolution an additional column has been added to Table 7-2 of HNP-F/PSA-0094 in order to list the specific alarms that might be available to indicate floods or leaks in the compartment.

Table 7-2 was revised to list the specific alarms or indications of leaks or flooding per compartment as well as the specific alarms to aid in flood identification in the area.

Independent Review Assessment:

Status: Partially Closed.

Basis: Table 7-2 of HNP-F/PSA-0094 lists alarms and indications that can be used to identify the flooding conditions in each of the flood compartments. However, the alarms and indications listed in Table 7-2 may not be always sufficient or clear (with the exception of Fire Water system, Chilled Water System, CCW, Circulating Water system, CVCS, SW, etc.) for use to identify the specific flood sources that cause the flooding conditions. SR IFSN-A2 requires the identification of flood alarms for each flood source and each flood area.

Recommendations: Provide additional information in Table 7-2 or in a new table to permit the identification of the flood source system given a flooding condition. The specific indications and alarms identified need to correlate with the specific flood sources for the identification of flood source system.

Evaluation of F&O impact on proposed application:

The specific alarms that might be available to indicate floods or leaks in a specific compartment have been added which results in this Supporting Requirement being MET. Documentation was revised to list the alarms or indications of leaks or flooding per compartment as well as the specific alarms to aid in flood identification in a particular area.

The F&O closure team suggested, however, that the documentation might not be sufficient or clear (for a subset of systems) to identify the specific source that caused a flood. Duke Energy disagrees with the closure teams suggestion. HNPs Ops procedures are symptom based diagnostic procedures that are not tied to specific sources, and the indicators and alarms help the operator diagnose the location and source of a flood. Dominant sources have relevant alarms identified. There is no direct correlation between specific indications and alarms to specific flood sources. No further analysis is required for this extended T.S. AOT LAR application.

B-3

HNP Internal Flooding F&O Disposition F&O ID: Associated SRs: Peer Review CC Assessment:

1-9 IFSN-A4 Finding (Not Met)

==

Description:==

Flow through floor drains is calculated and documented in Table 6-9 of HNP-F/PSA-0091. However, it appears that flow is incorrectly calculated for situations when multiple floor drains are connected to a common drain line. The calculations shown in HNP-F-PSA-0091 show a capacity per floor drain and the total capacity in each flood area is the average capacity per drain multiplied by the number of floor drains.

However, no discussion of how multiple drains are connected to common drain lines is provided. When multiple drains flow through a common drain line, the flow from each successive drain greatly reduces the flow from each drain in the system.

Resolution:

All floor drains above the 190 elevation drain to the Floor Drain Transfer Tank (FDTT) on the 190 level through a series of common drain pipes and risers. The total capacity of the drains for a particular flood compartment will be limited by the common drain line/riser for that compartment, so the drain flow calculations have been revised.

The locations of the floor drains, drain lines, and risers are shown in the revised Attachment 4 of HNP-F/PSA-0091. The equations used to recalculate drain flow are provided, and a calculation of flow through a typical series of floor drains connected to a common drain line has been performed.

The revised drain flow calculation demonstrated that the common drain line/riser has excess capacity to remove water from multiple floor drains for spray scenarios (<100 gpm) in a given flood compartment.

The common drain line, however, does not have sufficient capacity to provide beneficial removal of water for larger flood scenarios. This conclusion about capacity from the typical model is applicable to all flood compartments, so detailed modeling by flood compartment of multiple, similar configurations of a complex drain system was not performed.

IFSN-A4 says to, ESTIMATE the capacity of the drains[and] ACCOUNT for these factors in estimating flood volumes and SSC impacts from flooding. The capacity of the drains has been estimated and their ability to mitigate flood effects has been included in the scenarios, where applicable, thus satisfying this F&O. The propagation analysis documented in HNP-F/PSA-0092 includes removal of water by the floor drains for spray scenarios but does not credit removal of water by the floor drains for other scenarios.

Section 6.3.3 and Attachment A of HNP-F-PSA-0091 have been updated accordingly to include the revised analysis.

Independent Review Assessment:

Status: Partially Closed.

B-4

Basis: Section 6.3.6 of and Attachment 4 to Calculation HNP-F/PSA-0091 document the revised analysis of the drainage system in RAB. Based on this analysis for RAB, for spray events resulting in a flow rate of less than 100 gpm, the resulting flood is within the capacity of the drain system and will not result in submergence of SSCs in the flood originating compartment. For scenarios other than sprays, no credit is taken in the flood propagation analysis for beneficial removal of water from a flood compartment through the floor drains. For buildings other than RAB, however, drain analysis was not performed and no qualitative evaluation was documented. In particular, upper elevations in the Turbine Building (TB) could potentially flow downward to the basement and caused additional damage to PRA equipment in the TB basement (e.g., condensate pumps, etc.).

Recommendations: Perform drain analysis for buildings other than RAB (e.g., upper elevations in TB, etc.). Flood submergence scenarios should be considered due to flood water flow through the drain system e.g., in flood compartments containing the sumps or the Floor Drain Transfer Tank cubicle on the 190 elevation in RAB), including backflow through the drain line. More detailed discussion of the evaluation for each building should be documented.

Evaluation of F&O impact on proposed application:

The analysis of the floor drainage system was revised for the Reactor Auxiliary Building (RAB), and the supporting requirement was evaluated to be Met for the RAB by the F&O Closure team. The RAB contains most of the IF-PRA risk. Other buildings (such as the Turbine Building or Diesel Generator Building) were not assessed at the time, as inclusion of the drain propagation analysis would not provide any meaningful risk insights.

No further analysis is required for this extended T.S. AOT LAR application.

B-5

HNP Internal Flooding F&O Disposition F&O ID: Associated SRs: Peer Review CC Assessment:

1-16 IFSO-A4 Finding (SR is Met)

==

Description:==

Flooding events caused by human induced actions such as overfilling of tanks, flow diversion etc., are not addressed. Considerations of such events is required by the SR.

Resolution:

Plant level pipe break data on floods caused by human-induced maintenance errors and generic best estimates of associated plant level flood frequencies are already included in Revision 3 of the EPRI pipe failure rate report (EPRI TR 3002000079). This includes human errors such as overfilling of tanks and flow diversion that result in floods. Section 7 of EPRI TR 3002000079 provides tables and estimates of plant-level flood frequencies to support the estimation of flood initiating event frequencies caused by these maintenance errors. It is important to note that this does not include human errors resulting in pressure boundary failures since they are already included in direct failures involving failure of the pressure boundary caused by degradation mechanisms, loading conditions, and human error. Following guidance on the use of generic plant level maintenance-induced flood frequencies to support IFPRAs as described in Section 7 of the EPRI pipe failure rate report, Section 6.8.3 of HNP-F/PSA-0093 Revision 000 has already addressed computation of flood frequencies by HNP flood compartment and fluid system that are associated with flooding events caused by human-induced actions. Furthermore, to complement these generic frequencies, HNP Operating Experiences (OE) have been reviewed for maintenance-induced flood events and documented in Section 6.8.1 of HNP-F/PSA-0093 Revision 000.

Independent Review Assessment:

Status: Partially Closed.

Basis: Maintenance-induced flooding frequencies by system and by flood compartment are evaluated in Section 6.8.3 of HNP-F/PSA-0093. It appears that the apportionment of the maintenance-induced flood frequencies by system to individual flood compartment is not performed in a manner consistent with the characteristics of the maintenance-induced flooding since it was done by the fraction of the system pipe length located in each flood compartment (although it follows exactly the guidance provided in EPRI Report 3002000079).

Maintenance-induced flooding scenarios are modeled in Sections 7.3.4 and 7.4.2 (as well as Attachment

9) of HNP-F/PSA-0092 for CCW heat exchangers and ESCW chillers in Flood Compartments FLC17b (RAB Elevation 236) and FLC18a (RAB Elevation 261), respectively. Insufficient description is provided for the screening process used to select the maintenance-induced flooding scenarios included in the HNP IFPRA model.

With no proper justification, the maintenance-induced flooding frequencies apportioned to flood compartments other than the above two compartments were not accounted for in the IFPRA model.

B-6

Since the frequency of maintenance induced flooding was derived from actual industry events, the frequencies apportioned to the flood compartments not selected for flood scenario modeling cannot be discarded unless it can be demonstrated that no open maintenance (including both PM and CM) can be performed on the subject fluid system during power operation.

Recommendations: Provide more thorough description of the screening process used to select the maintenance-induced flooding scenarios included in the IFPRA model. Check with EPRI for additional guidance regarding the basis for the method recommended for the apportionment of the frequency to the individual flood areas and the intended approach to the selection and modeling of the maintenance induced flood scenarios in selected flood areas.

Evaluation of F&O impact on proposed application:

Plant level pipe break data on floods caused by human-induced maintenance errors and generic best estimates of associated plant level flood frequencies are included per Revision 3 of the EPRI pipe failure rate report, EPRI TR 3002000079 (Ref. B.9). This includes human errors such as overfilling of tanks and flow diversion that result in floods. Human errors resulting in pressure boundary failures are included in direct failures involving failure of the pressure boundary caused by degradation mechanisms, loading conditions and human error. To complement the generic data, HNP Operating Experience (OE) was reviewed for maintenance-induced flood events and documented in the IFPRA analysis. No further analysis is required for this extended T.S. AOT LAR application.

B-7

HNP Internal Flooding F&O Disposition F&O ID: Associated SRs: Peer Review CC Assessment:

1-18 IFSN-B3 Finding (Not Met)

==

Description:==

The assessment of door failure heights is evaluated in HNP-F-PSA-0092, section 6. The analysis of doors is based entirely on assumptions. However, these assumptions are not listed in Section 5 of the document. The standard requires that assumptions be listed and characterized.

Resolution:

Door failure assumptions have been revisited based on a civil calculation, HNP-C/RAB-1008 Rev. 0. This calculation demonstrates the pressure a standard door adjacent to the Main Control Room can withstand to be at least 1.5 psig away from the doorframe with a safety factor of 4. This pressure loading was applied to a flooding scenario and new door failure heights were calculated. This is available in Section 6.1 of HNP-F/PSA-0092.

Previous assumptions regarding door failure heights have been deleted or reworded.

Independent Review Assessment:

Status: Open.

Basis: Civil Calculation HNP-C/RAB-1008, Rev. 0 provides a Harris-specific analysis that indicates a standard 3X7 tornado door can withstand a sustained pressure of 1.5 psig away from the doorframe with a safety factor of 4. Based on this pressure loading, it was estimated that the door failure differential flood height is at least 6.5 feet (note that the estimated door failure differential flood height at Fort Calhoun was even higher). However, the critical failure modes evaluated in Civil Calculation HNP-C/RAB-1008, Rev. 0 only include failures of door frame, door latch, door hinge plate, and door hinge pin. The analysis did not consider warping of door resulting in failure to latch. For fire doors, the warping failure mode may be more vulnerable than the other failure modes, based on the analysis of fire door manufacturer test data for another U.S. nuclear plant.

Also, the evaluation performed in Civil Calculation HNP-C/RAB-1008, Rev. 0 is for tornado door which is considered to be stronger than the standard fire doors and non-fire rated normal egress doors. As such, the door failure criterion of 6.5 feet of differential flood height should not be applied to the fire doors and normal egress doors.

It is not clear if this door failure differential flood height was applied to the RAB doors. If yes, it is inappropriate. If no, the use of the criteria of 1 foot/3 feet mentioned in the EPRI IFPRA guidance report appears to be too conservative for the RAB fire doors.

Recommendations: Re-examine the specific criteria used for the door failures in HNP IFPRA and ensure that a more realistic criterion is used.

B-8

Evaluation of F&O impact on proposed application:

Assessment of HNP-specific door failures has been incorporated into the internal flooding PRA model and the documentation has been updated. No further analysis is required for this extended T.S. AOT LAR application.

B-9

HNP Internal Flooding F&O Disposition F&O ID: Associated SRs: Peer Review CC Assessment:

1-19 IFQU-A5 Finding (Met)

==

Description:==

SR HR-G4 requires that the analyses be based on realistic estimates of the time to receive cues. The analyses used an assumption of 5 minutes to receive cues and assumed that service low pressure alarms would be received. Experience shows that only for extremely large breaks would low pressure alarms be received and no analyses were seen that justified use of low pressure alarms for the HNP flood scenarios.

No evaluation of the time to receive drain and sump alarms was provided. The basis for timing of the events analyzed was a scenario evaluated in the FSAR and that timing may not be applicable to the scenarios evaluated in the HNP IF PRA.

Resolution:

The HRA calculation (HNP-F/PSA-0094) has been revised to include a table that states the specific alarms to indicate floods in each flood area (Table 7-2). The HRA calculation has also been revised to include a table (Table 7-2) that documents the analysis of the RAB sump level alarms and the expected time to alarm for spray events as well as flood events in the respective flood area. The sumps are identified in attachment 6 of the HNP Internal Flooding Areas and Sources Calculation (HNP-F/PSA-0091). A discussion about the flood drain alarms for spray events are documented in F&Os 1-10 and 2-3.

The new information has been incorporated into the HRA calculator for validation of timing and scenario development per the suggested resolution.

Independent Review Assessment:

Status: Partially Closed Basis: Analysis of RAB sump level alarms was documented in Table 7-4 of Calculation HNP-F/PSA-0094 for a spray event with a leak rate of 100 gpm and a flood event with a break flow of 2,000 gpm. However, the timings of the low pressure and high flow alarms are not addressed (i.e., no evaluation was found). The sump level alarms will support the identification of a flooding condition. However, it is not sufficient to support the identification of the specific flood source. No basis is provided to justify that 5 minutes are sufficient to diagnose the flood source and make decision on how to isolate the break.

Recommendations: Either address the timings of receipt of the low pressure alarms and high flow alarms for the different flooding scenarios analyzed, or justify that identification of the flood source by the equipment/auxiliary operator and decision by the MCR operators on how to isolate the break can be accomplished within 5 minutes.

Evaluation of F&O impact on proposed application:

B-10

The HRA calculation has been revised to include the specific alarms that indicate floods in each flood area. Documentation of analysis of the RAB sump level alarms has been added, and the expected time for floor drain alarms from spray events in each flood area is included. The new information was incorporated into the HRA timing and scenario development per the suggested resolution. No further analysis is required for this extended T.S. AOT LAR application.

B-11

HNP Internal Flooding F&O Disposition F&O ID: Associated SRs: Peer Review CC Assessment:

2-3 IFSN-A3 Finding (Met)

==

Description:==

While Attachments 1-4 of HNP-F-PSA-0094 identifies the automatic and manual actions that have the ability to terminate or contain propagation for the four events requiring HRA, the documentation does not include similar actions for the remaining sources and areas. Identification of the actions is required by the SR.

Resolution:

Section 7.2 of HNP-F/PSA-0094 has been modified to include the following information:

All floor drains and equipment drains above the 190 elevation drain to the Floor Drain Transfer Tank (FDTT) or the Equipment Drain Tank (EDT) respectively. When these tanks reach a high level set-point they are automatically pumped to the Floor Drain Tank (FDT) and Waste Hold-up Tanks (WHT) respectively. There are also Hi-Hi FDTT and EDTT Level Alarms that would prompt operators to take manual actions if the automatic features failed (APP-105 1-1 and APP-105 1-3 respectively). Although this action will not keep up with the higher flow-rates expected from a flood or major flood this action will aid in containing the propagation of flood waters to the extent of the drain system and the capacity of the transfer pumps.

Although these drains are not credited in the HNP internal flooding analysis it still demonstrates an automatic action that would be used to contain propagation as stated by the peer review team to help satisfy this comment. Once the FDTs and WHTs are 85% capacity operators will receive an alarm that should prompt them to manually align the pumps to additional tanks to aid in mitigating the propagation of flood waters. There are sumps in the RAB 190, Service Water Tunnel (216) and the RAB 236 elevations that will automatically pump down thus aiding in the mitigation of flood water accumulation.

These sumps are identified in table 7-4 which also displays their respective volume, alarm and calculated time to alarm. Additional manual actions are documented in Table 7-2 of HNP-F/PSA-0094. This table has been modified to include a column that corresponds to each flood compartment that states the manual actions to isolate or mitigate flood propagation. Procedural guidance is provided to direct operators to manually mitigate the accumulation of flood waters in step 3.10.g of AOP-022 which states:

"EVALUATE opening doors to adjacent non-critical areas to limit rise in water level at the break location.

These automatic and manual flood mitigation actions have been discussed and confirmed with Operations and documented in the HRA Calculator.

Independent Review Assessment:

Status: Partially Closed Basis: Section 7.2 of Calculation HNPF/ PSA-0094 describes the automatic actions by the sump pumps as well as the manual operator actions to align the pumps to additional tanks. In addition, Table 7-2 of HNP-B-12

F/PSA-0094 identifies the manual operator actions that can be implemented to mitigate the flooding condition and propagation in the affected flood compartments. However, no manual action (e.g., break isolation) is identified for many of the flood compartments. Most of the manual actions identified are opening doors to noncritical areas. In Table 7-2, no considerations were given to isolation of the ruptured or leaking piping system by closing specific MOVs or manual valves. Nevertheless, isolation actions are modeled for many of the flood scenarios. They are just not listed in Table 7-2.

Recommendations: Document manual break isolation actions such that all proper operator responses are identified in Table 7-2 for each flood source in each flood compartment.

Evaluation of F&O impact on proposed application:

Documentation has been added to describe the automatic actions by the sump pumps as well as the manual operator actions to align the pumps to additional tanks. In addition, the manual operator actions that can be implemented to mitigate the flooding condition and propagation in the affected flood compartments have been identified. No further analysis is required for this extended T.S. AOT LAR application.

B-13

HNP Internal Flooding F&O Disposition F&O ID: Associated SRs: Peer Review CC Assessment:

2-4 IFSN-A6 Finding (Not Met)

FEV-A5 Finding (Met)

==

Description:==

Not all flood failure mechanisms are considered in the susceptibility of components to flood-induced failures. HELBs alone can result in high humidity and temperature which in turn will result in fire sprinkler discharge. Assessment of these failure mechanisms is required by RG 1.200.

Resolution:

An analysis of high energy line breaks (HELBs) has been performed, and a new appendix describing the analysis has been added to the HNP-F/PSA-0091 calculation. The accident scenarios have been updated to include HELBs and the resulting effects. Jet impingement, pipe whip, high temperature and high humidity effects have been considered.

Independent Review Assessment:

Status: Partially Closed Basis: Attachment 10 to Calculation HNPF/PSA-0091, Revision 1 provides the evaluation of such flood failure mechanisms as jet impingement, pipe whip, high temperature, high humidity, compartment pressurization, etc. that may result from the high energy line breaks (HELB). A criterion of 20 feet (for pipes with inner diameter less than 24) or 10D (for pipes with inner diameter greater than 24) was used to determine whether an SSC or fire protection sprinkler would be impacted by the effects of HELB.

While the criteria of 20 feet/10D is adequate for the analysis of jet impingement and pipe whip, there is no analysis documented to demonstrate that the effects of high humidity and high temperature resulting from failure of high energy piping would not propagate beyond 20 feet/10D causing SSCs failures.

According to the HNP PRA staff, the only flood compartment in which not all PRA equipment is failed by a HELB scenario is a large room in the RAB, in which the 20 feet/10D zone of influence (ZOI) was applied.

The temperature as a function of time in RAB at Elevation 261 after a MSLB in the steam tunnel (with door D10 to RAB open) was analyzed. The results indicate that, near the sprinkler header, the ceiling temperature reached is unlikely to activate the sprinklers. And, the peak temperature in the immediate proximity of Instrument Racks A1-R33 and A1-R22 (located directly outside of Door D10) would experience the direct effects of the steam plume coming through Door D10. Relative humidity in the area near Instrument Rack A1-R33 (El. 263.25), which is bounding, reaches 100% for more than 20 minutes.

Relative humidity values near the chillers and HVAC equipment peak at 100%. The high energy lines in the RAB includes the steam supply line to the TDAFW pump and the charging lines. Although the steam lines for the TDAFW pump pass through RAB 236 elevation, the steam isolation valves located in the steam tunnel are normally closed during power operation, except during the TDAFW pump test. As such, this area is only exposed to the potential of a high energy line break during the TDAFW pump test.

The HNP IFPRA needs to verify that no PRA equipment would be impacted by high humidity or high temperature beyond the 20 feet/10D ZO, even for the rupture of the TDAFW pump steam supply line.

B-14

Recommendations: Provide analysis to demonstrate that the effects of high temperature and high humidity beyond 20 feet/10D would not cause additional PRA component damage.

Evaluation of F&O impact on proposed application:

An analysis of high energy line breaks (HELBs) has been performed, and a new appendix describing the analysis has been added to the IFPRA documentation. The accident scenarios have been updated to include HELBs and the resulting effects. Jet impingement, pipe whip, high temperature and high humidity effects have been considered. No further analysis is required for this extended T.S. AOT LAR application.

B-15

HNP Internal Flooding F&O Disposition F&O ID: Associated SRs: Peer Review CC Assessment:

2-8 IFEV-A7 Finding (CC I/II)

==

Description:==

While a great number of maintenance induced flooding frequencies were calculated, no evidence could be found that they were ever included in the model. The value for each of these events is significant when compared to the pipe break frequency values used in the same areas. Therefore, consideration of maintenance-induced events could have a significant effect on the overall results.

Resolution:

In communications with Operations personnel, it was determined that the only maintenance-induced flooding events that would occur in Mode 1 are located in the RAB 236 and 261 elevations (FLC17b and FLC18a, respectively). Specifically, they are the CCW heat exchangers and the ESCW chillers. These two flood compartments decision trees were altered to include Maintenance-Induced as a failure mechanism and scenarios were developed for them. This can be found documented in Sections 7.3.4 and 7.4.2 of HNPF/PSA-0092 as well as Attachment 9 of the same calc.

Independent Review Assessment:

Status: Open Basis: Maintenance-induced flooding scenarios are modeled in Sections 7.3.4 and 7.4.2 (as well as ) for CCW heat exchangers and ESCW chillers in Flood Compartments FLC17b (RAB Elevation 236) and FLC18a (RAB Elevation 261), respectively. Insufficient detailed description is provided for the screening process used to select the maintenance-induced flooding scenarios included in the IFPRA model. During the onsite resolution review, it was indicated by the HNP Operations that open PM will not be performed on the CCW heat exchangers and ESCW chillers during power operation.

Since the frequency of maintenance induced flooding is derived from actual industry events, the frequencies apportioned to the flood compartments not selected for flood scenario modeling cannot be discarded unless it can be demonstrated that no open maintenance (including both PM and CM) can be performed on the subject fluid system during power operation.

Recommendations: Provide more thorough description of the screening process used to select the maintenance-induced flooding scenarios included in the IFPRA model. Proper treatment and modeling of the maintenance-induced flooding frequencies should be considered.

Evaluation of F&O impact on proposed application:

In communications with Operations personnel, it was determined that the only maintenance-induced flooding events that could occur in Mode 1 are the CCW heat exchangers and the ESCW chillers. These B-16

two flood compartments decision trees were modified to include Maintenance-Induced flooding as a failure mechanism and scenarios were developed.

Additional documentation needs to be added on how Duke selected the maintenance-induced flooding scenarios and needs to assess if the maintenance-induced flooding frequency was apportioned properly.

This is a documentation issue and will have no impact on extended T.S. AOT LAR application.

B-17

HNP Internal Flooding F&O Disposition F&O ID: Associated SRs: Peer Review CC Assessment:

2-11 IFQU-A7 Finding (Met)

==

Description:==

The FRANX software was used to quantify the HNP internal flooding model which utilizes the fault tree linking approach. SR QU-A2 of Section 2.2-7 states that the frequencies of individual sequences need to be estimated for CDF and this was not done for internal flooding.

Resolution:

CDF results are now reported by event tree sequence in Revision 1 of HNP-F/PSA-0095.

Independent Review Assessment:

Status: Partially Closed Basis: Top CDF/LERF cutsets are presented in Table 5.1-1/5.2-1 and Attachments L/M of Calculation HNP-F/PSA-0095. The quantified CDF/LERF results of the top contributing flooding scenarios are given in Tables 5.1-2/5.2-2. Complete listing of the quantified CDF/LERF results for flooding scenarios are provided in Attachments J/K to Calculation HNP-F/PSA-0095.

Based on Duke PRA staff, FRANX includes calculation for accident sequences for LERF, but not for CDF.

Figures 5.6.1 and 5.6.2 show CDF by what is labeled as the sequence type, which are actually by IE, not sequence. In any event, estimates of the accident sequences are not included in the documentation.

Recommendations: Provide documentation of the quantified accident sequences for flooding scenarios.

Evaluation of F&O impact on proposed application:

Top CDF and LERF quantification results have been reported per the Standard and the IFPRA documentation has been updated. This is a documentation issue only and there is no impact on this extended T.S. AOT LAR application.

B-18

HNP Internal Flooding F&O Disposition F&O ID: Associated SRs: Peer Review CC Assessment:

2-12 IFQU-A7 Finding (Met)

==

Description:==

The FRANX software was used to quantify the HNP internal flooding model which utilizes the fault tree linking approach. The FRANX model is configured to apply recovery actions. A truncation of 1E-08 was applied for the CCDP which is considered sufficiently low to capture an appropriate number of cutsets to calculate an accurate CDF. The flooding model was quantified similarly to the internal events model which included the removal of cutsets with mutually exclusive events. Section 4.7.2 of HFPF-PSA-0095 states that the new HEPs associated with flooding were assumed to be independent of any other HEP in a scenario, however QU-C2 in Section 2.27 states that dependency between HEPs in a cutset or sequence must be assessed.

Resolution:

The HNP dependency analysis is now included in of Revision 1 of HNPF/PSA-0094, HNP Internal Flooding HRA Calculation. This dependency analysis is located in Section 7.7 which states:

The top combinations of operator actions identified in cutset reviews all had a CCDP of 1.0 therefore any operator actions beyond the securing of the flood were not dependent. The remaining operator action combination cutsets were analyzed and determined to be of such low value (i.e. E-8) that their impact on results were negligible. This is because the time between the necessary actions to be performed were long term (essentially hours) and thus the dependency was determined to be non-existent. The Internal events dependency values are addressed in the initial version of the IFPRA HRA calculation. Some initiating event operator actions were removed from combinations of actions. This is because OPER-D64 was inappropriately used in the combination determination for several other combinations, namely OPER-T58, OPER-T59, OPERQ17, OPER-Q18, OPER-Q21, OPER-Q24 and OPER-Q25. The internal events operator actions have been reviewed and appropriately penalized based on the available cues and the timing of actions with relation to the internal flooding event and the associated actions. The applied penalties are detailed in section 7.1. The expected actions related to flooding events are captured in table 7-7. This table lists the typical internal events operator actions as they relate to the flooding scenarios and evaluates those actions during the flood event as well as the associated penalties.

Independent Review Assessment:

Status: Partially Closed Basis: Section 7.7 of HNP-F/PSA-0094 indicates that there is no dependency between the flood mitigation actions and the subsequent operator actions carried over from the internal events PRA since the time between these actions are sufficiently long (essentially hours). However, a specific combination-by-combination evaluation of the dependency should be provided to demonstrate that indeed there is insufficient dependency between these two groups of operator actions.

B-19

Recommendations: Provide documentation of the specific combination-by-combination evaluation of the dependency to demonstrate that indeed there is insufficient dependency between these two groups of operator actions.

Evaluation of F&O impact on proposed application:

The HNP dependency analysis has been included in the IFPRA documentation. The documentation states that there is no dependency between the flood mitigation actions and the subsequent operator actions carried over from the internal events PRA since the time between these actions are sufficiently long. This is a documentation issue only and there is no further analysis is required for this extended T.S. AOT LAR application.

B-20

B.3 Fire There are five (5) findings that remain open from the original NRC review conducted in 2008 to support the NFPA 805 pilot process (Ref. B.2, APPENDIX G). Also in 2008, a partial scope follow-on peer review was conducted against the ANSI/ANS 58.23-2007 Standard requirements (Ref. B.2, APPENDIX J). Of the 16 findings from this review, resolutions to fifteen (15) were successfully addressed and were closed per an independent assessment against NEI 05-04/07 - 12/12-06 Appendix X (Ref. B.3). Thus, only 1 of these findings remains open.

Additionally, there are 2 SRs met at CC-1 with no open findings (FSS-D7, FSS-D9) - both items were dispositioned and accepted in the NFPA 805 application.

B-21

Table B.3 - Disposition of Open Fire F&Os Finding Supporting Capability Description Disposition for ESCW CT TS Number Requirement(s) Category (CC)

Supporting Requirement FSS-F3 FSS-F3-01 FSS-F3 CC I remained largely unchanged from ANSI/ANS-58.23-2007, for which ASME/ANS RA-S- ASME/ANS RA- Finding FSS-F3 was initiated, to 2007 (draft) Sa-2009 ASME/ANS RA-Sa-2009, for which the Capability Category I was determined.

The current analysis does not address this requirement of the standard. CC-I requires a Capability Category I was based on the qualitative assessment of the risk associated qualitative assessment of exposed with the selected fire scenarios (i.e., scenarios structural steel which is documented associated with fire induced failure of structural as Attachment 8 to Duke calc. HNP-steel structures). No clear scenario description F/PSA-0079, Rev. 3 (Ref. B.4).

is currently available. It is recommended that However, Attachment 1 of EC 409388, the scenarios in the turbine building are Rev. 0, subsequently documented a described from the point of view of fire PRA quantitative assessment of exposed scenarios. For a CC-I, the qualitative scenario structural steel that is description should include an ignition source, sufficient to meet Capability Category possible targets, impacts to the plant operation II/III.

(e.g. turbine trip, reactor trip, etc), and how the reactor will be shut down after the event.

Inclusion of the quantitative impact from the structural steel failure analysis in the FPRA would be expected to have an equal effect on the base case and the AOT case. There is no net impact to the application.

Supporting Requirements HRA-C1 and HRA-C1-3 HRA-C1 CC I/II/III HR-G1 remained largely unchanged from ASME/ANS RA-S-2007 (draft) for B-22

Table B.3 - Disposition of Open Fire F&Os Finding Supporting Capability Description Disposition for ESCW CT TS Number Requirement(s) Category (CC)

ASME/ANS RA-S- ANSI/ANS- HR-G1 was incorporated by reference. The which Finding HRA-C1-1 was initiated 2007 (draft) 58.23-2007 approach to determining which HEPs are to ANSI/ANS-58.23-2007 for which the developed using a detailed analysis does not Capability Category I/II/III was conform to the standard definition of determined. For ASME/ANS RA-Sa-significant for capability category II. Given the 2009, Supporting Requirement HRA-C1 fact that the model is still in development, this was assigned Capability Categories of I, is understandable. II, and III, but Support Requirement HR-G1 remained largely unchanged.

Capability Category II was determined for HRA-C1.

Tables 61 and 62 of Duke calc. HNP-F/PSA-0079, Rev. 3 (Ref. B.4), list significant operator actions having a F-V greater than 0.005 or RAW greater than 2, respectively. Section 7.1.3 of Duke calc. HNP-F/PSA-0075, Rev. 2 (Ref. B.5), describes the selection of HFEs for detailed analysis. Based on established criteria (e.g., inadequate instrumentation or short time window), some significant HFEs were not selected for detailed analysis and were instead conservatively assumed to be failed or left at a screening value.

However, the significant operator actions that were selected for detailed analysis are sufficient to provide the B-23

Table B.3 - Disposition of Open Fire F&Os Finding Supporting Capability Description Disposition for ESCW CT TS Number Requirement(s) Category (CC) risk insights for the ESCW T.S. AOT application.

There is no impact to the application.

Supporting Requirements HRA-C1 and HRA-C1-6 HRA-C1 CC I/II/III HR-G6 was incorporated by reference. It is too HR-G6 remained largely unchanged early in the process for this supporting from ASME/ANS RA-S-2007 (draft) for ASME/ANS RA-S- ANSI/ANS- requirement to have been achieved which Finding HRA-C1-6 was initiated 2007 (draft) 58.23-2007 satisfactorily, since only a few HFEs have been to ANSI/ANS-58.23-2007 for which the developed in detail. Capability Category I/II/III was determined. For ASME/ANS RA-Sa-2009, Supporting Requirement HRA-C1 was assigned Capability Categories of I, II, and III, but SR HR-G6 remained largely unchanged. Capability Category II was determined for HRA-C1.

Plant-specific and scenario-specific influences on human performance were addressed by a well-defined and self-consistent process, as described in Section 7.1.3 of Duke calc.

HNP-F/PSA-0075, Rev. 2 (Ref. B.5). This ensured the results were logical and consistent with inputs and method of analysis.

There is no impact to the application.

B-24

Table B.3 - Disposition of Open Fire F&Os Finding Supporting Capability Description Disposition for ESCW CT TS Number Requirement(s) Category (CC)

Supporting Requirement FQ-E1 and FQ-E1-2 FQ-E1 NOT MET The definition of significant contributor in the the Supporting Requirements for HLR-PRA standard includes the idea of summing, in QU-D and HLR-LE-F remained largely ASME/ANS RA-S- ANSI/ANS- rank order, the fire sequences and considering unchanged from ASME/ANS RA-S-2007 (draft) 58.23-2007 any in the top 95%, or any that individually 2007 (draft), for which Finding FQ-E1-2 contribute 1% or more, as significant. This was initiated, to ANSI/ANS-58.23-determination has not been made for fire CDF 2007, for which the NOT MET was or LERF. Harris does not appear to use the determined, to ASME/ANS RA-Sa-definition as provided in the PRA standard. 2009.

This SR continues to be NOT MET. This is a documentation-only issue and does not affect quantification of risk.

There is no impact to the application.

Supporting Requirement FQ-F1 and FQ-F1-1 FQ-F1 CC I/II/III the Supporting Requirements for HLR-QU-F and HLR-LE-G remained largely ASME/ANS RA-S- ASME/ANS RA- unchanged from ASME/ANS RA-S-2007 (draft) Sa-2009 2007 (draft), for which Finding FQ-F1-1 QU-F2 - Several of the recommended was initiated, to ASME/ANS RA-Sa-documentation requirements are not in place, 2009, for which the Capability specifically items b, e, f, g, i, j, m. Category I/II/III was determined.

Duke calc. HNP-F/PSA-0079, Rev. 3 (Ref. B.4), documents the majority of the typical documentation requirements:

B-25

Table B.3 - Disposition of Open Fire F&Os Finding Supporting Capability Description Disposition for ESCW CT TS Number Requirement(s) Category (CC) b) Attachment 32 documents records of the cut set review process.

e) Section 6.0 documents the total plant CDF and contribution from the different initiating events, however accident sequences were not individually documented.

f) Accident sequences were not individually documented.

g) Table 62 documents equipment and human actions with RAW > 2.0. In addition, Section 6.4 includes insights which make note of particular credit taken to mitigate potentially-dominant accidents.

i) Section 7.0 documents the uncertainty distribution for the total CDF.

j) Tables 61 and 62 documents importance measure results.

m) Section 3.0 documents the use of qualified software and controlled electronic input files. Section 5.5 documents the process the development of the FRANX input files an operation of FRANX. Section 10.0 documents the controlled electronic output files. This is a documentation-only issue.

B-26

Table B.3 - Disposition of Open Fire F&Os Finding Supporting Capability Description Disposition for ESCW CT TS Number Requirement(s) Category (CC)

There is no impact to the application.

Supporting Requirement FQ-F1 and FQ-F1-2 FQ-F1 CC I/II/III QU-F3 - There is currently no record of the Supporting Requirements for HLR-significant contributors to fire CDF. QU-F and HLR-LE-G remained largely ASME/ANS RA-S- ASME/ANS RA- unchanged from ASME/ANS RA-S-2007 (draft) Sa-2009 2007 (draft), for which Finding FQ-F1-2 was initiated, to ASME/ANS RA-Sa-2009, for which the Capability Category I/II/III was determined.

Section 6.0 of Duke calc. HNP-F/PSA-0079, Rev. 3 (Ref. B.4), documents the significant contributors to CDF, however accident sequences were not individually documented. This is a documentation-only issue.

There is no impact to the application.

B-27

B.4 High Winds Four (4) High Winds PRA Finding F&Os were generated during the focused peer review in 2015 (Ref. B.6). These findings were subsequently dispositioned by the high winds model vendor.

However, according to Ref. B.7, several resolutions / dispositions to the F&Os provided by the vendors are not yet accepted and approved by Duke Energy. As such, all Finding F&Os remain OPEN. The updated high winds analysis presented in Ref. B.8 is intended to address this issue.

B-28

HNP High Wind F&O Disposition F&O ID: Associated SRs: Peer Review CC Assessment:

Observation #: 1 WFR-B2 Finding

==

Description:==

DOCUMENT the process used in the wind fragility analysis. For example, this documentation typically includes a description of:

(a) the methodologies used to quantify the high-wind fragilities of SSCs, together with key assumptions.

(b) a detailed list of SSC fragility values that includes the method of analysis, the dominant failure mode(s), the sources of information, and the location of each SSC.

(c) the basis for the screening out of any generic high-capacity SSCs.

Basis For Significance:

The documentation can be enhanced by addressing the following items:

a. The anchorage of the Dedicated Shutdown Diesel Generator (DSDG) was not evaluated (Volume IV),

the evaluation indicates the overturning wind speed of 202 mph, therefore the anchorage evaluation or discussion on the screening is required. Also the DSDG target area can be enhanced by correcting the documentation on the effective target areas and exclusion of the extra surfaces.

b. It was indicated that it is conservative to assume the SSCs inside Category-I structures will not fail until the structure fails (and the structures are screened based on their design criteria). However the peer review team believes that the venting in the structures needs to be evaluated; the passage of tornado may cause rapid pressure drop, resulting in escape of the air from the building; if the exit is not rapid enough, it causes change in the internal pressure. This may cause failures including different SSCs or block walls before failure of the structure. Discussions with ARA engineers indicated that the screening was done during the walkdown based on the size of the vents and their location in respect to the wind direction. A brief discussion in the documentation to justify the rationale of what was done will enhance the documentation.

c. The startup transformer (SUT) was excluded from the wind fragility, based on the overturning analysis.

The weight of the transformer was not available for the analysis. A better justification on use of the transformer weight will improve the documentation.

d. The anchorage evaluation of the Air Compressor and Air Receivers (Volume I), was evaluated for the bolts in tension. Discussions with ARA engineers indicated that the shear was considered in the capacity of the anchors, the discussion about the shear failure of the anchor bolts will enhance the documentation.

e. A discussion of why and how the 80/20 factor between the ASCE/NBCC codes was determined would be helpful. It is clear that these weighting factors are assigned by engineering judgment, the peer review team were interested to understand the reasoning behind this selection.

f. Volume III Attachment 5 of the HNP-F/PSA-0098 calc has tables and graphs which are titled "Hit Probability for Test Data" and "Damage Probability for Test Data". Upon discussion with the HNP HW PRA team members, the PRA does not differentiate at all between these two and the suggestion is to B-29

improve the documentation so that these terms are explained in a clearer fashion in both the fragility and plant response documentation.

Resolution:

Improve the documentation on the identified items.

Vendor and Duke Resolution Disposition:

Vendor Proposed Approach:

1a) (Response to PR Item 1a) - Information on the anchorage of the DSDG was requested, but not made available during the project. As such, the resulting fragility does not consider anchorage of the DSDG. If additional information is provided, the wind pressure fragility of the DSDG is re-evaluated and discussion of the anchorage of the DSDG is added to Volume IV, Section 8.10 - this results in either an updated wind pressure fragility for the DSDG or screening of the DSDG from the wind pressure fragility analysis.

1b) (Response to PR Item 1a) - The exposed area of the DSDG in Volume III, Table 8-6 is modified to not include the area of the bottom surface of the DSDG which is not exposed to missile impact -- this results in an updated missile fragility for the DSDG.

1c) (Response to PR Item 1b) - Agree that Atmospheric Pressure Change (APC) loads is propagated through vent openings in Category I structures. However, UFSAR Section 3.3 requires that these buildings be designed for APC loading. Since the structures were designed with vents, it follows that the interior walls and components of such buildings would have to be designed to resist the APC loading as well. The vented Category I Structures housing SSCs at HNP include the ESW Intake and Screening structures, and the Emergency Diesel Generator Building. During the walkdown and review of plant drawings for these buildings, no evidence of conditions that would negate this assumption, such as masonry walls or unvented SSCs, was found. This discussion is added to Volume I, Section 3.2, Item 1.a and to list of assumptions in Volume I, Section 5.

1d) (Response to PR Item 1c) - The following discussion is added after the second sentence of the first paragraph of Volume I, Attachment 3, Section A3.3: ARA (2014) documents the wind pressure and missile fragility analyses conducted in support of the HW-PRA for the McGuire Nuclear Station. This analysis included two transformers of similar size to the HNP Startup Transformers.

1e) (Response to PR Item 1d) - The text in Volume I, Attachment 3, Sections A3.1 and A3.2 incorrectly refers to the tensile strength of the bolts. The phrase tensile strength should be replaced with the phrase design value. The design value presented in the reference includes the effects of both tension and shear on the anchor bolts. The phrase is updated and an explanation of the phrase design value is included in the evaluation of the air compressor and receiver anchor bolts in Volume I, Attachment 3, Sections A3.1 and A3.2.

1f) (Response to PR Item 1e) - The application of the 80/20 weighting between the ASCE and NBCC values for internal pressure coefficients is an engineering judgment based weighting. Also note that uncertainty in these internal pressure coefficients are treated as a part of the uncertainty analysis. Additional discussion of these weighting factors and reasoning behind their selection is added to Volume IV, Section 3.2.2, Item 5.a.iii.

B-30

1g) (Response to PR Item 1f) - The HW PRA does differentiate between missile hit and damage probabilities. However, it uses EITHER the missile hit OR damage probability for any given target/SSC.

The discussion of failure modes included in MFCalc added for the resolution of WFR-A1 Observation 1 to Volume III, Section 3.1 makes this clear.

Duke Disposition of Vendor Proposed Approach:

1a) - Observation RESOLVED 1b) - Observation RESOLVED 1c) - Observation NOT RESOLVED. It is felt that a portion of the proposed response does not adequately address the issue.

1d) - Observation RESOLVED 1e) - Observation RESOLVED 1f) - Observation RESOLVED 1g) - Observation NOT RESOLVED. It is felt that a portion of the proposed response does not adequately address the issue.

Evaluation of F&O impact on proposed application:

These issues are primarily documentation only and would not impact the final results. There is no impact to the HNP ESCW AOT LAR submittal.

B-31

HNP High Wind F&O Disposition F&O ID: Associated SRs: Peer Review CC Assessment:

Observation #: 2 WFR-B2 Finding

==

Description:==

DOCUMENT the process used in the wind fragility analysis. For example, this documentation typically includes a description of:

(a) the methodologies used to quantify the high-wind fragilities of SSCs, together with key assumptions.

(b) a detailed list of SSC fragility values that includes the method of analysis, the dominant failure mode(s), the sources of information, and the location of each SSC.

(c) the basis for the screening out of any generic high-capacity SSCs.

Basis for Significance:

1- The following items are noted in regard to the MFCalc program:

a. MFCalc data was generated from the regression analysis of the available TORMIS data. The analysis used plant specific missile surveys, drawings and calculations, however some generic data from the three surveyed sites in the TORMIS analysis was also used in the MFCalc analysis. While the lack of site specific attributes may influence fragility results, it is not believed to affect the risk results in a significant manner. If you have an application for which the HW results are critical, then I would spend the effort in increasing the detail on site specific attributes. If not, would not spend the effort.

b. The discussion of the missile count can be enhanced. For instance, a clarification on why vehicles and trees are not included in the count could be discussed in more detail.

c. The total number of missiles listed (specifically in the tables in Attachment 2 of Volume 3) are not fully documented for each zone. MFCalc does not require the type of the missile, but the reviewers were interested in knowing the nature of the missiles at different zones. For instance, during the walkdown, the security fence was noted on top of the Jersey Barriers, these posts were not embedded and were bolted into the Jersey Barrie, these bolts may fail in shear and may results in more number of missiles, however it was not clear to the peer review team whether this was considered or not.

d. The documentation of the selection of particular TORMIS parameters that were used in MFCalc could be made clearer. The parameters that best fit to the TORMIS data and selection criteria on certain parameters versus excluding certain other parameters are not fully documented.

e. Missile Treatment - requires an explanation in a clearer fashion, why Step 2 of the 3-step process discussed on page35 of 52 (top paragraph) of the HNP-F/PSA-0098 Volume III is not used in generating the MFCalc process.

f. Use of 600-foot radius in MFCalc -the criteria of 600 feet for the missile survey radius is not fully documented and why any other choice (e.g., 300 or 1200 feet) would not have altered the fragility results appreciably.

Resolution:

Improve the documentation on the identified items.

B-32

Vendor and Duke Resolution Disposition:

Vendor Proposed Approach:

2a) (Response to PR Item 2a) - Generic missile population data from other sites were NOT included in the HNP missile fragility analysis. All missile data developed to produce missile fragilities using MFCalc were developed based on the walkdown of the site conducted in August 2014. As discussed in Volume III, Section 3.1, MFCalc is a statistical missile fragility model that is based on missile fragilities developed using TORMIS for other nuclear power plants. We believe that the peer review team is concluding that generic data is used because the numbers of tree and vehicle missiles were not included in the fitting parameters for MFCalc. The parameters chosen when developing the data fits for MFCalc are based on engineering judgment and past experience as explained throughout Volume III, Section 3.2 and are shown to produce reasonable estimates of missile fragility in Volume III, Section 8.3.3. The discussion of the fitting parameters used for the development of the MFCalc data fits in Volume III, Section 3.2 has been edited to clarify the discussion above.

2b) (Response to PR Item 2b) - The sixth paragraph of Volume III, Section 3.1 indicates that vehicle and tree missiles are not included in the fitting parameters for the development of the MFCalc. The fifth sentence of this paragraph has been updated to read For example, tree and vehicle missiles are not used as fitting parameters because we know from previous experience these types of missiles are generally located several hundred feet from critical SSCs and have very small contributions to the overall fragility of the SSCs considered in the analyses.

2c) (Response to PR Item 2c) - The total numbers of missiles by missile type were not recorded for each missile source during the missile survey because this level of detail is not required for an MFCalc analysis.

That being said, the data tables presented in Volume III, Attachment 2 are potentially misleading and are revised to report only the missile data collected and remove the reference to missiles by missile type.

2d) (Response to PR Item 2c) - Also, as discussed during the peer review, the security fence sections mounted to jersey barriers around the plant were included in the missile survey counts.

2e), 2f) (Response to PR Item 2d) - Volume III, Section 3.1 documents the fitting parameters that were ultimately selected for use in the MFCalc data fits. This discussion has been edited to clarify the parameters chosen and reasons for not including some parameters, such as vehicle and tree missiles. As indicated in the text, selection of the fitting parameters is based largely on our past experience with both TORMIS and statistical missile fragility analyses. Further, the data fits chosen are shown to produce reasonable estimates of missile fragility in Volume III, Section 8.3.3.

2g) (Response to PR Item 2e) - Wind speed intensity is included as a separate fitting parameter in MFCalc.

Application of the wind speed intensity missile population factors discussed in step 2 of the process would result in accounting for wind speed twice in the fitting parameters. The following text is added to paragraph 7 of Volume III, Section 3.1:

We note that in TORMIS analyses, the number of missiles available from missile source structures is varied by wind speed to reflect damage states of the buildings producing the missiles. However, since wind speed is considered as a separate fitting parameter, the number of missiles used to compile the n300 and n600 parameters is based on the total missile population rather than the missile population by wind speed intensity. This is necessary to prevent accounting for wind speed in two separate fitting parameters.

B-33

Additionally, a reference to this discussion is be added to the discussion of the structure missile estimation method discussed in Volume III, Section 8.1.5.

2h) (Response to PR Item 2f) - MFCalc actually uses two different radii (300 ft and 600 ft) for compilation of missile statistics as input as discussed in paragraph 6 of Volume III, Section 3.1. This discussion explains that the selection of these radii is based on engineering judgment and experience in developing missile statistical models from TORMIS data and provides a reference to a previous statistical analysis. Further, Volume III, Section 8.3.3 presents the goodness of fit plots and discussion of the comparisons between model results and the training data.

Duke Disposition of Vendor Proposed Approach:

2a) - Observation NOT RESOLVED. It is felt that a portion of the proposed response does not adequately address the issue.

2b) - Observation NOT RESOLVED. It is not clear where the "clarification" agreed in the action plan is.

2c) - Observation RESOLVED 2d) - Observation RESOLVED 2e), 2f) - Observation NOT RESOLVED. It is not clear where the "clarification" agreed in the action plan is.

2g) - Observation NOT RESOLVED. Documentation of the resolution is not correct.

2h) - Observation NOT RESOLVED. It is felt that a portion of the proposed response does not adequately address the issue.

Evaluation of F&O impact on proposed application:

These issues are primarily documentation only and would not impact the final results. There is no impact to the HNP ESCW AOT LAR submittal.

B-34

HNP High Wind F&O Disposition F&O ID: Associated SRs: Peer Review CC Assessment:

Observation #: 1 WPR-A5 Finding

==

Description:==

The supporting requirement states that in the HRA, additional stresses that can increase the likelihood of human errors or inattention, compared to the likelihood assigned in the internal events HRA need to be justified.

Assumption 3 states that impacts on operator actions that directly involve establishing/maintaining vessel injection or decay heat removal are assumed to be negligible for high winds initiators because these are immediate actions. These are not immediate actions and as such must be considered for increased probability of failure. Assumption 3 is not used in the analysis.

Assumption 5 states that actions that must be performed within 5 minutes and involve reactivity control and injection are of high priority. These actions are performed regularly in the simulator and therefore, are not impacted by the high wind initiators. This is true for reactivity control, but not injection.

Justification for this assumption is inadequate.

Initially, there was significant disagreement between the utility and the peer reviewer whether operator actions for decay heat removal and inventory control should be increased or not. After several discussions on the topic and review of the HRA analysis, it was determined that only immediate operator actions related to reactivity control were not increased in the PRA model. This made sense since reactivity is one of the first items checked and manual trip of the reactor or emergency boration would be implemented within the first few minutes. The following resolution was agreeable to both parties.

Basis for Significance:

Assumptions 3 and 5 as written do not examine the additional stresses that can increase the likelihood of human errors or inattention for decay heat removal or inventory control, compared to the likelihood assigned in the internal events HRA when decay heat removal and inventory control are undertaken in non-high wind event accident sequences. No justification is provided for not increasing the likelihood of these human errors for decay heat removal or inventory control.

Resolution:

In PSA-0100, Assumption 3 is not used. Should be deleted since it creates confusion. Assumption 5 should be revised to say reactivity control only since that is the only way it is used in the PRA. The assumptions look reasonable for manual reactor trip since this is an immediate and obvious response.

Vendor and Duke Resolution Disposition:

Vendor Proposed Approach:

In HNP-F/PSA-0100, deleted Assumption 3 and revised Assumption 5 to only say reactivity control.

B-35

Duke Disposition of Vendor Proposed Approach:

Observation RESOLVED Evaluation of F&O impact on proposed application:

These issues are primarily documentation only and would not impact the final results. There is no impact to the HNP ESCW AOT LAR submittal.

B-36

HNP High Wind F&O Disposition F&O ID: Associated SRs: Peer Review CC Assessment:

Observation #: 2 WPR-C1 Finding

==

Description:==

SR WPR-C1 requires that the analyses be documented in a manner that facilitates PRA applications, upgrades and peer review.

Several occurrences were found that made review difficult or were actual documentation errors. These include:

Table 6-20 and Table 6-21 present some operator actions with the designation (e.g., OPER-26) without a description of the operator action. In most cases the description can be found in another table, but this makes review somewhat cumbersome.

Very little description is provided as to how the HWPRA utilizes the internal events model. No mention is made that the general transient event tree is used without modification. At a minimum, a reference to the internal events accident sequence analysis and the general transient event tree should be provided.

Errors were found in Table 8-1, the 1.200 ASME Capability Cross

Reference:

WPR-A2 (event trees/fault trees) points to Section 6.4.2. Section 6.4.2 discusses significant seq., and cutsets.

WPR-A11 (recovery) points to Section 6.3.6.3. There is no 6.3.6.3. Recovery is addressed in 6.3.7.3.

WPR-B2 (uncertainties) points to Section 6.3.7. Section 6.3.7 discusses recovery.

WPR-C3 (uncertainty) points to Section 6.3.7. Uncertainty is in 6.3.8.

Description of CDF Cutsets 2 and 6 of Table (6-17) indicate DSDG fail to run, whereas the BE is DSDG missile strike.

Section 6.5.2 (Significant cutsets) indicates the top 10 cutsets are dominated by failure of both EDGs (CCF or FTR) along with failure of DSDG (missile strike or test & maint.). From the top 10 cutsets, DSDG fails due to missile (4), FTR (3) and T&M (1). Section 7.0 (Conclusions and Recommendation) states that the significant contributors to CDF and LERF are wind induced loop, with a secondary failure of the DSDG of secondary importance.

Basis for Significance:

Although the plant model analysis appears to be of good quality, documentation errors or inadequacies, at a minimum make review and understanding more difficult, and could result in misinterpretation of findings and insights.

The Conclusions and Recommendations section is vital for the end user of this analysis to apply results and insights to plant modifications and operations.

Resolution:

Correct typos and misstatements.

B-37

Update the Conclusions and Recommendations to more accurately present the calculated results and to be consistent with statements in prior sections of the document.

Vendor and Duke Resolution Disposition:

Vendor Proposed Approach:

None at this time. (These are documentation issues.)

Duke Disposition of Vendor Proposed Approach:

Observation is NOT RESOLVED. It is not clear what corrections are made and what the revisions are.

Evaluation of F&O impact on proposed application:

These issues are primarily documentation only and would not impact the final results. There is no impact to the HNP ESCW AOT LAR submittal.

B-38

REFERENCES B.1 ABS Consulting, Harris Nuclear Plant, PRA Finding Level Fact and Observation Technical Review, Report No. R-3857458-2026, March 2017 B.2 HNP-F/PSA-0069, "HNP - PSA Model Peer Review Resolution", Rev. 3 B.3 EPM, F&O Closeout by Independent Assessment of the Harris Nuclear Plant (HNP) Fire PRA Model Against the ASME PRA Standards Requirements to Meet NEI 07-12 Appendix X, Rev. 1, October 2017 B.4 HNP-F/PSA-0079, "Harris Fire PRA - Quantification", Rev. 3 B.5 HNP-F/PSA-0075, "Harris Fire PRA - Human Reliability Analysis", Rev. 2 B.6 PWROG-15056-NP, Focused High Winds PRA Peer Review for Shearon Harris Nuclear Power Plant, Rev. 0 B.7 HNP High Winds PRA Peer Review - Resolutions and Comment Dispositioning B.8 HNP-F/PSA-0099, "HNP High Wind Probabilistic Risk Assessment (HWPRA): Plant Response Model",

Rev. 0 B.9 EPRI TR 3002000079, Pipe Rupture Frequencies for Internal Flooding Probabilistic Risk Assessments; Rev. 3 B-39

RA-19-0007, License Amendment Request for Extension of the Essential Services Chilled Water System Allowed Outage Time and Removal of an Expired Note from Technical Specifications

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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

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RA-19-0007, License Amendment Request for Extension of the Essential Services Chilled Water System Allowed Outage Time and Removal of an Expired Note from Technical Specifications

Text

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

Reference:

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