L-19-288, Request to Revise the Perry Nuclear Power Plant (PNPP) Fire Protection Program Licensing Basis

Request to Revise the Perry Nuclear Power Plant (PNPP) Fire Protection Program Licensing Basis

ML19352E549
Person / Time
Site:	Perry
Issue date:	12/18/2019
From:	Payne F FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-19-288
Download: ML19352E549 (25)

Text

{{#Wiki_filter:FENOCTM ~ FirstEnergy Nuclear Operating Company Perry Nuclear Power Plant PO. Box 97 1 O Center Road Perry, Ohio 44081 Frank Payne Vice President 440-280-5382 December 18, 2019 L-19-288 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Perry Nuclear Power Plant Docket No. 50-440, License No. NPF-58 Request to Revise the Perry Nuclear Power Plant {PNPP) Fire Protection Program Licensing Basis Pursuant to 10 CFR 50.90, the FirstEnergy Nuclear Operating Company (FENOC) is submitting a request to revise the PNPP fire protection program licensing basis. The request would abandon in place the general area heat detection system in the drywell. The FENOC evaluation of the proposed change is enclosed. NRC staff approval of the proposed change is requested by December 21, 2020. Once approved, the change shall be implemented within 120 days. There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Phil H. Lashley, Acting Manager - Nuclear Licensing and Regulatory Affairs, at (330) 315-6808. I declare un~er penalty of perjury that the foregoing is true and correct. Executed on December 11_, 2019.

Sin0ily, LiA~

Frank R. Payne I

Enclosure:

Evaluation of a Request for Licensing Action cc: NRC Regional Ill Administrator NRC Resident Inspector N RR Project Manager Executive Director, Ohio Emergency Management Agency, State of Ohio (NRC Liaison) Utility Radiological Safety Board

Evaluation of a Request for Licensing Action Page 1 of 17

Subject:

Request to Revise the Perry Nuclear Power Plant (PNPP) Fire Protection Program (FPP) Licensing Basis. 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 System Design and Operation 2.2 Current Licensing Basis Requirements 2.3 Reason for Proposed Change 2.4 Description of Proposed Change

4.0 REGULATORY EVALUATION

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

6.0 REFERENCES

ATTACHMENT:

1. Simplified Layout Figures of the Containment and Drywell

Evaluation of a Request for Licensing Action Perry Nuclear Power Plant Page 2 of 17 1.0

SUMMARY

DESCRIPTION This evaluation supports a FirstEnergy Nuclear Operating Company (FENOC) request to amend Operating License No. NP-58 for the Perry Nuclear Power Plant (PNPP). The proposed amendment would revise the PNPP Fire Protection Program (FPP) licensing basis by deviating from Branch Technical Position (BTP) CMEB 9.5-1, Guidelines for Fire Protection for Nuclear Power Plants, Section C.6.a, Fire Detection. CMEB 9.5-1, Section C.6.a(1) states: Detection systems should be provided for all areas that contain or present a fire exposure to safety-related equipment. The proposed change to the PNPP FPP is to abandon the general area fire detection system located within the drywell. Since the PNPP design uses heat detectors within the drywell, the amendment request will use the term heat detection system in lieu of fire detection system when referring to the PNPP design. A fire protection evaluation was completed by FENOC to support the proposed deviation. The evaluation determined that the proposed FPP deviation reduces the level of defense in depth associated with the FPP. As a result, it adversely affects the ability to achieve and maintain safe shutdown in the event of a fire. Therefore, Nuclear Regulatory Commission (NRC) approval of the proposed change is required. 2.0 DETAILED DESCRIPTION 2.1 System Design and Operation Drywell and Select Support System Descriptions The drywell is a cylindrical structure, located within the containment, that encloses the reactor pressure vessel. It has an outside diameter of 83 feet and a height of approximately 86 feet with a removable steel head cover approximately 32 feet in diameter, which when removed permits access to the reactor pressure vessel. The lower part of the drywell wall is submerged in the suppression pool. Three rows of circular vents penetrate the drywell wall below the normal suppression pool water level to allow free communication between the drywell and containment suppression pool water volumes. The walls, floor, and ceiling (excluding the removable head cover) of the drywell are constructed of steel and concrete. The drywell wall above the suppression pool to the removable head cover has a three-hour fire rating. Entrances into the drywell consist of a double-doored personnel air lock and an equipment hatch. A shield barrier is provided for each entrance. Wall penetrations have three-hour fire-rated seals, except for the suppression pool vents, which are under water.

Evaluation of a Request for Licensing Action Perry Nuclear Power Plant Page 3 of 17 shows simplified layout figures of the containment and drywell. The area that comprises the inside of the drywell, including the reactor pressure vessel but excluding the area directly beneath the reactor pressure vessel, is labelled as fire zone 1RB-1c. The drywell cooling system operates to provide cooling for the drywell. The system is designed as non-safety-related. The system includes 21 dual-element thermocouple temperature detectors mounted in the system ductwork and in the drywell to monitor the temperature within the drywell. The detectors actuate alarms in the control room if the drywell ambient temperature exceeds the setpoint limits. Also located in the drywell are temperature detectors associated with the containment atmosphere monitoring system (CAMS). The function of CAMS is to detect and to monitor the containment to determine if an accident has occurred. The CAMS temperature-monitoring function in the drywell is designed as safety-related. Since the proposed FPP change is centered on temperature conditions in drywell fire zone 1RB-1c, the CAMS description is limited to that location. There are seven resistance temperature detectors (RTDs) located within drywell fire zone 1RB-1c. Six RTDs are arranged in two divisions, each division having three detectors. The divisions are located on opposite sides of the drywell. Temperature signals from the RTDs are recorded in the control room. Each division has a high average drywell temperature alarm in the control room. The seventh RTD provides indication on a remote shutdown panel recorder, which is located outside of the control room. Drywell Combustible Loading Fire loading contained in fire zone 1RB-1c is low, less than 52,000 British thermal units per square foot (BTU/ft2). Cable insulation, which makes up approximately 38 percent of the total combustible material in the zone, is the largest contributor of combustible material and is distributed throughout the fire zone. Most of the cabling is routed in conduit. The largest concentration of combustible material in the fire zone is in the area around the reactor recirculation pumps, which are located at azimuths 150 degrees and 330 degrees on elevation 583 feet - 6 inches. This combustible material consists of motor winding insulation and lube oil, and makes up approximately 28 percent of the total combustible material in the fire zone. The remainder of the combustible material, including hydrogen igniter insulation, lubricating oil, and motor winding insulation (other than the reactor recirculation pump motors), valve operator grease, and miscellaneous other combustibles, makes up approximately 34 percent of the total combustible material in the zone and is distributed through the zone at different elevations.

Evaluation of a Request for Licensing Action Perry Nuclear Power Plant Page 4 of 17 A total fire loading of less than 52,000 BTU/ft2 corresponds to a burn time of approximately 39 minutes. Fire zone 1RB-1c is inaccessible during normal operation. When the drywell is open, administrative controls are utilized to control transient combustibles and hot work that may occur within the fire zone. Administrative controls ensure the drywell is clear of debris, tools, and other unnecessary materials before the drywell is closed. Drywell Heat Detection and Fire Suppression Description Heat detectors are used in the drywell since the high radiation levels could impact the serviceability and function of smoke detectors. There are two heat detection systems located within fire zone 1RB-1c. One system is used for fire warning and fire suppression system activation associated with the two reactor recirculation pumps due to the concentration of combustible materials around the pumps. The suppression function is provided by a local application type carbon dioxide system. This detection system is not being affected by the proposed FPP change. The second system is for general area heat detection. Rate-compensated heat detectors are used in both of the drywell heat detection systems. Rate-compensated heat detectors remove the time delay that normal set-temperature heat detectors experience. Rate-compensated heat detectors will activate when a fire heats up the surrounding atmosphere rapidly, but the detector is not yet at its setpoint temperature. The rate-compensated heat detector will also activate for fires that heat up the surrounding atmosphere slowly, such that the detector activates at its setpoint temperature. The heat detection system associated with the two reactor recirculation pumps is comprised of eight rate-compensated heat detectors located around each reactor recirculation pump. The heat detectors are arranged in two loops around the top and bottom of each pump. The two loops are sub-divided into two zones, each loop having four detectors. Each zone contains two detectors from each loop. The detectors are located between approximately the 601 feet and 619 feet elevations in the drywell. The cabling is routed in conduit. These detectors not only provide alarm functions, but also provide initiation signals to a carbon dioxide suppression system located near the pumps. It takes one detector from each zone to initiate activation of the carbon dioxide suppression system. Discharge of carbon dioxide will occur once a containment isolation valve is opened by the control room operators. When a heat detector reaches its actuation temperature, an alarm signal is sent to the control room.

Evaluation of a Request for Licensing Action Perry Nuclear Power Plant Page 5 of 17 The general area heat detection system consists of 39 rate-compensated heat detectors. The detectors are arranged into two detection zones. One detection zone consists of 20 detectors and provides coverage for the east drywell area. The other detection zone consists of 19 detectors and provides coverage for the west drywell area. The heat detectors are located above approximately the 599 feet, the 627 feet, and 660 feet elevations in the drywell. The instrument cabling is routed in conduit. If a heat detector setpoint is exceeded, an alarm signal is sent to the control room. In accordance with the off-normal instruction for fire actions, if the alarm was in a reactor recirculation pump area, then control room operators would open the containment isolation valve for the carbon dioxide system to ensure the discharge of carbon dioxide. If the alarm was associated with a drywell general area detector and cannot be reset, the control room operators would monitor drywell temperature values using the CAMS, and monitor the affected area and system or component parameters in the area until access into the drywell could be accomplished or adequate time had elapsed with no abnormal temperature increases. Safe Shutdown The PNPP design incorporates various diverse methods for post-fire reactor shutdown. With the availability of offsite power, reactor shutdown would utilize the reactor feedwater system, the main turbine pressure regulators, and the main condenser. With offsite power unavailable, reactor shutdown could utilize one or a combination of several systems including the high pressure core spray system (HPCS), the reactor core isolation cooling system (RCIC), the low pressure coolant injection systems (LPCI), or the low pressure core spray system (LPCS). For actual post-fire conditions, the control room operators will select and utilize any of these systems as conditions warrant. To achieve and maintain safe shutdown, two independent methods of safe shutdown exist. For each method, shutdown could be achieved within 72 hours, with or without offsite power. Each method provides for three basic shutdown functions: reactor shutdown, depressurization, and core cooling. The methods are referred to as Method A and Method B. These shutdown methods were selected, in part, on the basis that both methods are included in the symptom-oriented emergency procedures. That is, as actual fire conditions develop in the plant, the control room operators will respond to a number of possible symptoms (for example, low reactor vessel water level or high reactor pressure). The operator responses include verifying or initiating the systems included in Method A or Method B. Therefore, the operator will respond to a fire in a manner similar to other accidents that impact the safe shutdown capability. The actions taken for safe shutdown in the event of a fire are accomplished by the same procedures used to operate the credited systems under non-fire conditions.

Evaluation of a Request for Licensing Action Perry Nuclear Power Plant Page 6 of 17 Method A The PNPP design utilizes two main divisions of power supplied from offsite sources or two independent diesel generators. A third division of power, including a diesel generator, is dedicated to the HPCS and its support systems. Method A utilizes systems powered from Division I power sources. For reactor shutdown, Method A utilizes output channels A and C of the reactor protection system (RPS) and the control rod drive system. RPS is a normally energized, deenergized to trip, one out of two taken twice logic system. RPS channels A and D sensor cables are routed with the Division I power and control cables to maintain appropriate separation from RPS channels B and C sensor cables. As a result, if a fire damages the Division I cable runs and impairs the A and D sensor cables, the RPS logic would be satisfied, and the reactor would scram. If a reactor scram is not automatically initiated, operators can perform manual actions to initiate the scram using RPS channels B and C. For depressurization and initial core cooling, Method A utilizes a combination of the automatic depressurization system (ADS) and safety relief valves (SRV), with either RCIC, LPCS, or LPCI loop A. RCIC utilizes an inboard containment isolation valve powered from Division II sources. In several fire zones, power and associated control circuits for this valve do not have adequate separation from the Method B systems, and a fire could disable both Method B and RCIC. However, for these areas, reactor inventory control can be provided by LPCS, which is powered by Division I. For those fire zones where only one of these Method A systems for reactor inventory control will be available for shutdown after a fire, the circuits and components for RCIC are referred to as Method A1, and the LPCS will be referred to as Method A2. For Method A1, depressurization is provided initially by steam discharge to the RCIC system. Reactor coolant inventory will be controlled by the RCIC. As the level is restored, shutdown will proceed by operation of the relief valves to reduce reactor system pressure and temperature until the RCIC cut-off pressure is reached. For Method A2, depressurization is provided initially by the ADS. ADS will depressurize the reactor coolant system to the point where the LPCS or LPCI loop A cut-in pressure is reached. Either LPCS or LPCI loop A will be utilized to restore reactor water level. During the depressurization process, the suppression pool cooling mode of RHR could be initiated to control suppression pool temperature. At approximately 135 pounds per square inch gauge (psig), the shutdown cooling mode of RHR would be initiated, thereby achieving cold shutdown. Extended core cooling (decay heat removal) is provided by either the shutdown cooling mode of RHR or the alternate shutdown cooling path for the vessel through the ADS valves.

Evaluation of a Request for Licensing Action Perry Nuclear Power Plant Page 7 of 17 Method B Method B utilizes systems powered from Division II power sources. For reactor shutdown, Method B will utilize RPS output channels B and D and the control rod drive system. RPS is a normally energized, deenergized to trip, one out of two taken twice logic system. RPS channels B and C sensor cables are routed with the Division II power and control cables to maintain appropriate separation from RPS channels A and D sensor cables. As a result, if a fire damages the Division II cable runs and impairs the B and C sensor cables, the RPS logic would be satisfied, and the reactor would scram. If a reactor scram is not automatically initiated, operators can perform manual actions to initiate the scram using RPS channels A and D. For depressurization and initial core cooling, Method B utilizes a combination of ADS and LPCI loop B or Loop C. ADS depressurizes the reactor coolant system to the LPCI loop B or loop C cut-in pressure. Generally, LPCI loop C is used to restore reactor water level. During the depressurization process as reactor coolant inventory is restored, suppression pool cooling using RHR loop B could be initiated to control suppression pool temperature. As cooldown proceeds, the shutdown cooling mode of RHR would be initiated, thereby achieving cold shutdown. Extended core cooling (decay heat removal) is provided by either the shutdown cooling mode of RHR or the alternate shutdown cooling path for the vessel through the ADS valves. Fire Zone 1RB-1c Equipment for each safe shutdown method located within fire zone 1RB-1c includes, but is not limited to: ADS valves, RHR valve, RCIC valve, and Division I and II power and control cables. The following features ensure that one of the two methods of safe shutdown will be maintained should a fire occur in this zone. The Division II-powered RHR inboard shutdown cooling suction isolation valve is located in this area. The inboard shutdown cooling suction isolation valve is needed for shutdown cooling (cold shutdown) operation of the RHR system. To protect the high-low pressure interface, power to the Division I-powered RHR outboard shutdown cooling suction isolation valve is disconnected in Modes 1, 2, and 3. Neither valve requires protection during hot shutdown. Procedural direction is provided to operate these valves, as conditions require, once pressure has decreased to an acceptable point, to achieve and maintain cold shutdown. For a fire in the vicinity of the inboard valve, alternative shutdown cooling can be

Evaluation of a Request for Licensing Action Perry Nuclear Power Plant Page 8 of 17 accomplished using the ADS valves located approximately 90° around the fire zone from this valve. This component separation is acceptable because of the localized combustibles within the fire zone. At locations where circuits from these components are in close proximity, these circuits are separated by structural features that function as radiant energy shields. The circuits for the RHR loop A pump contain an interlock from the Division II RHR inboard shutdown cooling suction isolation valve. The interlock logic circuit provides a stop signal to the pump if either the inboard or outboard suction valve is closed. Loss of circuit power or logic signal will not stop the pump. The Division II supply valve is needed for cold shutdown only. Therefore, in order to initiate shutdown cooling, the operator would need to isolate the circuit if a spurious signal existed. Guidance to perform this activity is contained in an off-normal operating procedure. Materials necessary for this activity are stored onsite. Where the circuits for the RCIC inboard isolation valve (Method A1) are in close proximity to Method B circuits and components, shutdown could be accomplished using LPCS for reactor inventory control (Method A2). The components and circuits for RCIC are separated from the ADS valves and solenoid operators. The conduits containing the control circuits for the Method A ADS solenoid operators are located in close proximity of the conduits containing circuits for the Method B solenoid operators. The ADS valves are required to support the use of LPCS (Method A2) and LPCI (Method B) for reactor inventory control. However, components and circuits for the RCIC system (Method A1) are separated by more than 20 feet, and RCIC does not depend on the ADS valves for operation. Therefore, RCIC would be available to provide reactor inventory control in the event of a fire resulting in the loss of redundant ADS valve solenoid operators. The control circuits for the Division I ADS and SRV solenoid operators are routed in common conduits in the fire zone. Likewise, the control circuits for the Division II ADS and SRV solenoid operators are routed in common conduits. A fire in the area of the conduits containing either Division I or Division II ADS and SRV solenoid operator circuits could result in the spurious opening of several relief valves. The opening of several relief valves would result in the inability to operate RCIC (Method A1). However, LPCS (Method A2) and LPCI (Method B) do not have safe shutdown equipment or circuits located in this fire zone and would remain available for reactor inventory control. Therefore, for a fire in fire zone 1RB-1c, either Method A or Method B systems would be available depending upon the location of the fire. Separation of circuits and components for the redundant safe shutdown methods through distance or structural features that function as radiant energy shields provide adequate protection for this zone.

Evaluation of a Request for Licensing Action Perry Nuclear Power Plant Page 9 of 17 2.2 Current Licensing Basis Requirements The current fire protection licensing basis requirements are those stated in license condition 2.C.(6) of the Perry Nuclear Power Plant Facility Operating License NPF-58: FENOC shall comply with the following requirements of the fire protection program: FENOC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report, as amended, for the Perry Nuclear Power Plant and as approved in the Safety Evaluation Report (NUREG-0887) dated May 1982 and Supplement Nos. 1 through 10 thereto, subject to the following provisions:

a. FENOC may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

2.3 Reason for Proposed Change The drywell general area heat detection system is degraded and has been declared non-functional. High drywell temperatures have degraded the detector cabling, resulting in electrical grounds. Due to the location of the conduits and heat detectors, repair or replacement is infeasible. 2.4 Description of Proposed Change To resolve this condition, the drywell general area heat detection system will be abandoned in-place with no replacement. This resolution does not abandon components associated with detection and suppression of fires in the vicinity of the two reactor recirculation pumps. As a result, the PNPP FPP will deviate from BTP CMEB 9.5-1, Section C.6.a, which states: Detection systems should be provided for all areas that contain or present a fire exposure to safety-related equipment. The proposed FPP change of abandoning the drywell general area heat detection system in-place, with no replacement, was evaluated by FENOC against the criteria of the license condition. The evaluation determined there is a reduction in the level of defense in depth associated with the FPP. As a result, there is an impact upon the FPP that could adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. Therefore, NRC approval of the proposed change is required.

Evaluation of a Request for Licensing Action Perry Nuclear Power Plant Page 10 of 17

3.0 TECHNICAL EVALUATION

Nuclear power plants use the concept of defense in depth to achieve the required degree of fire safety in nuclear power plants. With respect to the FPP, the defense in depth principle is aimed at achieving an adequate balance in: Prevention, Detection and suppression, and Maintenance of safe shutdown capabilities. Therefore, the technical justification of the proposed PNPP FPP change will be described in relation to these three principles of defense in depth. Prevention Fire zone 1RB-1c comprises the region inside the drywell including the reactor vessel, but excludes the area directly beneath the reactor. Wall penetrations have three-hour fire rated seals, except for the suppression pool vents, which are under water. The drywell wall, above the suppression pool to the removeable reactor head cover, has a three-hour fire rating. As such, an exposure fire from other areas affecting equipment in fire zone 1RB-1c is not considered credible. Fire loading within fire zone 1RB-1c is low (approximately 52,000 BTU/ft2). In the area of the reactor recirculation pumps, a location where large concentration of combustibles exists, a local application carbon dioxide system with dedicated heat detectors for system actuation has been provided. When activated, this system alarms in the control room. Discharge of carbon dioxide will occur once a containment isolation valve is opened by the control room operators. This system will not be affected by the proposed abandonment of the drywell general area heat detectors. Cable insulation is a significant contributor of combustible material throughout the fire zone. However, the majority of this cable is routed in conduit. Therefore, extensive damage from a fire in this fire zone cannot be reasonably expected. Fire zone 1RB-1c has a localized fire load coupled with a high ceiling, large and spread-out volume, and non-combustible walls, ceiling, and floor. For these reasons, this fire zone has a low fire hazard potential. Fire zone 1RB-1c is inaccessible during normal operation. When the drywell is open, administrative procedures are utilized to control transient combustibles and hot work that may occur in the fire zone. Administrative controls are in place to ensure the drywell is clear of debris, tools, and other unnecessary materials before the drywell is closed, or the remaining items have been properly evaluated by engineering. Additionally, if a fire was detected in the drywell during normal full power operation, it is estimated to take approximately six hours to shut down the plant

Evaluation of a Request for Licensing Action Perry Nuclear Power Plant Page 11 of 17 and enter the drywell to investigate. Given the short burn time associated with the combustibles present in the drywell (approximately 39 minutes), the fire would burn out before drywell entry was achieved. This action (shutting down the plant and entering the drywell to investigate) is no different than the present operational response if the drywell general heat detectors were still in service. The proposed FPP change to abandon the drywell general area heat detection system does not make any physical changes to the drywell structure. The combustible loading of the fire zone has not been altered. The ignition sources located in the fire zone have not changed. The administrative procedures used to control transient combustibles, hot work, and removal of debris, tools, and other unnecessary materials before the drywell is closed have been maintained. Given the short burn time associated with the combustibles present in the drywell, the fire would burn out before the drywell entry could be made. Therefore, the prevention aspect of the FPP defense in depth principle has not been affected. Detection and Suppression Detection There are two FPP heat detection systems located within fire zone 1RB-1c. One system is used for fire warning and fire suppression system activation associated with the two reactor recirculation pumps. This detection system is not affected by the proposed FPP change. The second system is for general area heat detection and is used for fire warning. The proposed FPP change will abandon this general area heat detection system in place. The proposed change impacts the PNPP FPP by deviating from the requirements of BTP CMEB 9.5-1, Section C.6.a, which states: Detection systems should be provided for all areas that contain or present a fire exposure to safety-related equipment. A fire protection evaluation was completed by FENOC to support the proposed deviation. The evaluation determined that the proposed FPP deviation reduces the level of defense in depth associated with the FPP. As a result, it adversely affects the ability to achieve and maintain safe shutdown in the event of a fire. Therefore, NRC approval of the proposed change is required. To mitigate the impact on the PNPP FPP of the abandonment of the drywell general area heat detection system, there are the heat detectors located within the drywell around the reactor recirculation pumps, and within the drywell cooling system. Additionally, CAMS temperature detectors are located within the drywell. If ambient temperature in the drywell gets too high, the systems will actuate an alarm in the control room. The current off-normal instruction for fire actions states that if there was a fire in the drywell, the control room operators would monitor the drywell temperature using the CAMS, and monitor the area until access into the drywell could be accomplished or

Evaluation of a Request for Licensing Action Perry Nuclear Power Plant Page 12 of 17 adequate time has elapsed with no abnormal temperature increases. The proposed FPP change will not alter these operator actions. Therefore, though the proposed abandonment of the drywell general area heat detection system has an adverse impact upon the PNPP FPP, the proposed change can be adequately mitigated. Suppression There are no fire suppression systems associated with the drywell general area heat detection system. Hence, the proposed FPP change which abandons the drywell general area heat detection system has no impact upon the suppression aspect of the FPP defense in depth principles. Maintenance of Safe Shutdown Capabilities A safe shutdown analysis evaluated fire zone 1RB-1c against 10 CFR 50, Appendix R, Section III.G, Fire protection of safe shutdown capability. This fire zone contains components and circuits for both Method A and Method B safe shutdown systems. The evaluation concluded that if a fire occurred in this fire zone, either Method A or Method B would be available to achieve and maintain safe shutdown, meeting the intent of 10 CFR 50, Appendix R, Section III.G. The installation of a heat detection system and an automatic fire suppression system (10 CFR 50, Appendix R, Section III.G.2., Item e) is not credited in this fire zone to protect safe shutdown components and circuits. The proposed FPP change to abandon the drywell general area heat detection system does not make any physical changes to the drywell structure. The combustible loading and the ignition sources located in the fire zone have not changed. The administrative procedures used to control transient combustibles, hot work, and material control within the drywell have been maintained. The heat detection system and automatic fire suppression system was not credited in the safe shutdown analysis. As a result, the safe shutdown analysis has not been impacted. Therefore, maintenance of the safe shutdown capability, which is a FPP defense in depth principle, has not been affected.

4.0 REGULATORY EVALUATION

FirstEnergy Nuclear Operating Company (FENOC) proposes to revise the Perry Nuclear Power Plant (PNPP) Fire Protection Program (FPP) licensing basis by deviating from the guidance of Branch Technical Position (BTP) CMEB 9.5-1, Guidelines for Fire Protection for Nuclear Power Plants, Section C.6.a, Fire Detection. The proposed revision to the PNPP FPP will abandon the drywell general area heat detection system.

Evaluation of a Request for Licensing Action Perry Nuclear Power Plant Page 13 of 17 4.1 Applicable Regulatory Requirements/Criteria PNPP Regulatory Background The PNPP construction permit application was docketed on June 22, 1973. On May 1, 1976, the Nuclear Regulatory Commission (NRC) revised Standard Review Plan (SRP), Section 9.5.1, Fire Protection, including BTP APCSB 9.5-1, Guidelines for Fire Protection for Nuclear Power Plants. The SRP revision contained guidelines for NRC staff review of fire protection activities contained in nuclear power plant construction permit applications docketed after July 1, 1976. By letter dated September 30, 1976, the NRC transmitted to the PNPP staff APCSB 9.5-1, Appendix A, Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976, which provided acceptable alternative guidance and criteria to the positions given in SRP 9.5-1. This letter also requested an evaluation of the fire protection program at PNPP. The evaluation was to include a comparison of the fire protection provisions proposed for PNPP with the guidelines contained in Appendix A, which for PNPP are those designated as Application Docketed But Construction Permit Not Received as of 7/1/76. PNPP responded by letter dated September 26, 1977 with the comparison. By letter dated May 12, 1981, the NRC informed the PNPP staff that 10 CFR 50, Appendix R, Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979, will be used by the NRC as guidance for fire protection reviews. In July 1981, the NRC issued a major revision to the SRP for use in the review of new license applications. This SRP revision included a revision to SRP Section 9.5.1, incorporating BTP CMEB 9.5-1, which was an update to the earlier fire protection BTPs. By letter dated October 15, 1981, the NRC requested a comparison of the PNPP FPP to 10 CFR 50, Appendix R, identifying and justifying any deviations to Appendix R. By letter dated April 29, 1982, the PNPP staff responded with the requested comparison. In NUREG-0887, Safety Evaluation Report Related to the Operation of Perry Nuclear Power Plant, Units 1 and 2, (SER) dated May 1982, the NRC stated that BTP CMEB 9.5-1 contained the technical requirements of BTP APCSB 9.5-1, Appendix A, and 10 CFR 50, Appendix R. The NRC stated that based upon licensee commitments and NRC staff review, the PNPP fire detection systems conformed with the guidelines of BTP CMEB 9.5-1, Section C.6.a. As indicated in SER Supplement 8 dated January 1986, the NRC concluded that the PNPP fire protection program, with approved deviations, satisfies BTP CMEB 9.5-1 and 10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants (GDC) 3, Fire protection.

Evaluation of a Request for Licensing Action Perry Nuclear Power Plant Page 14 of 17 Regulatory Guide 1.189 The NRC issued Regulatory Guide (RG) 1.189, Fire Protection for Operating Nuclear Power Plants, Revision 0 in April 2001. The NRC developed the RG to provide a comprehensive fire protection program guidance document and to identify the scope and depth of fire protection programs that the staff would consider acceptable for nuclear plants currently operating as of January 1, 2001. The RG states that plants licensed after January 1, 1979, are subject to the requirements of 10 CFR 50.48(a) only, and must meet the provisions of GDC 3 as specified in their license conditions and as accepted by the NRC in their SERs. These plants are typically reviewed to the guidance of SRP Section 9.5-1. For these plants, where commitments to specific guidelines cannot be met, or alternative approaches are proposed, the differences between the licensees program and the guidelines are documented in deviations. If the licensee has adopted the standard license condition and incorporated the fire protection program in the Final Safety Analysis Report (FSAR), the licensee may make changes to the approved fire protection program without prior approval of the NRC if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire as documented in a safety evaluation. The RG states that if the evaluation finds that there is an adverse affect, the licensee should make modifications to achieve conformance or seek a license amendment from the NRC. 10 CFR 50, Appendix A, General Design Criteria [GDC] for Nuclear Power Plants, Criterion 3, Fire protection GDC 3 states, in part, the following: Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety. The proposed change will abandon the drywell general area heat detection system with no replacement being installed. An evaluation was performed to determine the affect of the proposed change on the PNPP FPP. The combustible loading of the drywell is low and has not been affected. The ignition sources within the drywell have not been altered. The component separation and energy shields used within the drywell to protect safe shutdown components have not been altered. Drywell temperature can still be monitored using the heat detectors of the drywell cooling system or the safety-related temperature monitoring function of CAMS. The time frame to shutdown the plant, should a fire be detected in the drywell, is no different than if the drywell general area heat detectors remained in operation. Credit was not taken for the drywell general area heat detection system in the safe shutdown analysis. The evaluation concluded that if a fire occurred in this area, either Method A or Method B would remain available to achieve and maintain safe shutdown. Therefore, GDC 3 remains satisfied with respect to the proposed change.

Evaluation of a Request for Licensing Action Perry Nuclear Power Plant Page 15 of 17 BTP CMEB 9.5-1 BTP Section 6.a(1) states, in part, the following: Detection systems should be provided for all areas that contain or present a fire exposure to safety-related equipment. The proposed change will abandon the drywell general area heat detection system with no replacement being installed. An evaluation was performed to determine the effect of the proposed BTP deviation on the PNPP FPP. The combustible loading of the drywell is low and has not been affected. The ignition sources within the drywell have not been altered. The component separation and energy shields used within the drywell to protect safe shutdown components have not been altered. Drywell temperature can still be monitored using the heat detectors of the drywell cooling system or the containment atmosphere monitoring system. The time to shutdown the plant, should a fire be detected in the drywell, is no different than if the drywell general area heat detectors remained in operation. Credit was not taken for the drywell general area heat detection system in the safe shutdown analysis. The evaluation concluded that if a fire occurred in this area, a method for safe shutdown would remain available meeting 10 CFR 50, Appendix R, Section III.G.2., items d and f. Since the drywell contains safety-related equipment, the PNPP FPP, with respect to the drywell, will not conform to the BTP. As described in Supplement No. 8, to the Safety Evaluation Report Related to the Operation of Perry Nuclear Power Plant, Units 1 and 2, there are several NRC-approved deviations associated with plant areas not having fire detection. Hence, not having fire detection capabilities within an area at the PNPP is not unique. Though the proposed change deviates from BTP CMEB 9.5-1, the change will adequately satisfy the fire protection defense-in depth principles. 4.2 No Significant Hazards Consideration Analysis FENOC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below.

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

Response: No. The proposed amendment will abandon the drywell general area heat detection system with no replacement being installed. The proposed amendment does not abandon components associated with detection and suppression of fires in the vicinity of the two

Evaluation of a Request for Licensing Action Perry Nuclear Power Plant Page 16 of 17 reactor recirculation pumps. There are no physical changes to structures, systems, or components other than those associated with the abandonment of the general area heat detection system. The combustible loading of the drywell has not been affected. The proposed amendment does not impact the initiators of a fire in the drywell. The component separation and energy shields used within the drywell to protect safe shutdown components have not been altered. Safe shutdown equipment would remain functional in the event of a fire. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No. Other than the activities associated with the abandonment, in-place, of the drywell general area heat detectors, the proposed FPP deviation does not involve any physical changes to any systems, structures, or components. Though the purpose of the abandoned system is to monitor the drywell temperature to determine the possibility of a fire, there are other systems located in the drywell that also monitor drywell temperature that can be used for the same purpose. Other than the abandonment of the drywell general area heat detectors, there is no impact to the design function of any plant component and no change to how the components are operated. Safe shutdown equipment would remain functional in the event of a fire. There are no new failure modes introduced as a result of the proposed amendment. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No. The proposed amendment will abandon the drywell general area heat detection system with no replacement being installed. The proposed amendment does not abandon components associated with detection and suppression of fires in the vicinity of the two reactor recirculation pumps. The ability of other plant structures, systems, or components to perform their designed safety function is unaffected by the proposed change. The proposed amendment does not affect the drywell combustible loading or impact the initiators of a fire in the drywell. Component separation and energy shields used within the drywell to protect safe shutdown components have not been altered. The proposed amendment does not alter any safety analyses, safety limits, limiting safety system settings, or method of operating the plant. The changes do not adversely affect plant operating margins.

Evaluation of a Request for Licensing Action Perry Nuclear Power Plant Page 17 of 17 Therefore, the proposed amendment does not involve a significant reduction in a margin of safety. Based on the above, FENOC concludes that the proposed change 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. 4.3 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. Branch Technical Position CMEB 9.5-1, Guidelines for Fire Protection for Nuclear Power Plants, July 1981.

2. Letter dated May 12, 1981 from Mr. Robert L. Tedesco (NRC) to Mr. Dalwyn R.

Davidson (The Cleveland Electric Illuminating Company), subject Fire Protection Safety Evaluation.

3. NUREG-0887, Supplement No. 8, Safety Evaluation Report Related to the Operation of Perry Nuclear Power Plant, Units 1 and 2, January 1986.

4. Generic Letter 86-10, Implementation of Fire Protection Requirements, April 24, 1986.

5. Regulatory Guide 1.189, Fire Protection for Operating Nuclear Power Plants, Revision 0, April 2001.

6. Perry Nuclear Power Plant Updated Safety Analysis Report Section 9A.4.1.1.3, Fire Zone 1RB-1c.

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