ML20091G758

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Comprehensive Review & Evaluation of Nypa:Safe Shutdown Capability Reassessment 10CFR50,App R
ML20091G758
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 05/31/1995
From: Sullivan K
BROOKHAVEN NATIONAL LABORATORY
To:
Office of Nuclear Reactor Regulation
Shared Package
ML20091G765 List:
References
TER-4745-T26-5, TER-4745-T26-5-95, NUDOCS 9507270222
Download: ML20091G758 (26)
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' TER-4745/T26-5-95 l

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. I COMPREHENSIVE REVIEW AND EVALUATION OF THE NEW YORK POWER AUTHORITY REPORT

" SAFE SHUTDOWN CAPABILITY REASSESSMENT 10CFR50 APPENDIX R J.A. FITZPATRICK NUCLEAR POWER PLANT" Kenneth Sullivan Engineeririg Technology Division Brookhaven National Laboratory Upton, New York 11973 Prepared for:

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, DC 20555 May 1995 IEnl S

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Enclosure w

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4 CONTENTS Section Ijilt Paae No.

EXECUTIVE

SUMMARY

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1 R EVI EW CRITE R I A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2 POST-FIRE SAFE SHUTDOWN CAPABILITY ............ 2 2.1 Separation of Safe Shutdown Functions . . . . . . . . . . 2 2.2 Post-Fire Safe Shutdown Methodology-General Plant Areas . . . . . . . . . . . . . . . . . . . . . . . . 2 2.3 Safe Shutdown Capability . . . . . . . . . . . . . . . . . . . . 3 2.4 Alternate Shutdown Capability . . . . . . . . . . . . . . . . . 8

3. ASSOCI ATED CIRCUlTS . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3.1 Common Power Supply Associated Circuit Concern ................................ 10 3.2 Common Enclosure Associated Circuit Concern ................................ 10 3.3 Spurious Signal Concern . . . . . . . . . . . . . . . . . . . . . 10 3.4 High/ Low Pressure Interfaces . . . . . . . . . . . . . . . . . . 11 4 CORRECTIVE ACTIONS 4.1 LPCI Alternate Power Supply Circuit Modification ............................. 11 4.2 Upgrade of CST t.evelInstrument Loop Power Supply ............................ 12 4.3 Isolation of Main Steam isolation Valve .................................. 12 4.4 isolation of Safety Relief Valves . . . . . . . . . . . . . . . . 12 4.5 Safety Related RHRSW/ESW Pump R oom Ventilation . . . . . . . . . . . . . . . . . . . . . . . . . . 12 4.6 Isolation of RHR Valves for Reactor Building Fire . . . . . . . . . . . . . . . . . . . . . , . . . . . . . 12 4.7 Reroute / Wrap Safe Shutdown Cables . . . . . . . . . . . . 13 4.8 Emergency Service Water Pump Isolation . . . . . . . . . 13 4.9 Reactor Head Vent Valves . . . . . . . . . . . . . . . . . . . . 13 4.10 Containment Spray Valves ................... 14 4.11 Relocation of Fire Protection Panel 7 6 CO 2-Pf 4 L- 8 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 4.12 HPCI Valve Circuit Modification . . . . . . . . . . . . . . . . 14 4.13 RCIC Valve Control Circuit Modification .......... 14 iii

Table of Contents (Cont'd)

Section Ih[a pg, g 4.14 Installation of Additional Emergency _

Lighting Units ............................ 15 4.15 Relocation of Fire Protection Panel 76CO2-PNL-4 ............................ 15 4.16 Torque Switch Rewiring for Alternate Shutdown MOVs . . . . . . . . . . . . . . . . . . . . . . . . . . 15 4.17 Alternate Appendix R Reactor Vessel Level Instrumentation . . . . . . . . . . . . . . . . . . . . . . . 15 4.18 Ensure Availability of Motor Control Center and Unit Coolers and Re-route Cables for a Fire in Fire Area 1 A and Vil . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 4.19 Conclusion .............................. 16 iv

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6-t EXECUTIVE

SUMMARY

t By letter dated October 26,1992 the New York Power Authority (NYPA) forwarded to NRC a report titled " Safe Shutdown Capability Reassessment 10CFR50 Appendix R." This report documents the results of a reassessment of the post fire safe shutdown capability of the J.A. FitzPatrick Nuclear Power Plant. At the request of the NRC Office of_ Nuclear Reactor Regulation, the Engineering Technology Division of Brookhaven National Laboratory (BNL) performed a comprehensive review of the Licensee's post-fire safe shutdown  :

methodology and analysis of associated circuits, as described in the above referenced submittal. This report provides the results of the evaluation. The review concentrated on Post-Fire Safe Shutdown capability, Associated Circuits and the Licensee's separation analysis methodology.

I NYPA has developed four shutdown methods (Method 1, 2, 3, and 4) capable of  ;

bringing the plant to a cold shutdown condition in the event of fire. The' capability of each ,

shutdown method to meet the post fire; safe shutdown performance goals of Appendix R -

was evaluated.

With the exception of Method 3, the methods proposed by the Licensee to accomplish safe shutdown are acceptable. Method 3 relies on the use of low-pressure injection systems (ADS /LPCI or ADS /CS) from the control room in the event of fire in areas not requiring an altamative shutdown capability, Shutdown Method 3 proposes the use of low pressure ,

injection systems (ADS /LPCI or ADS / Core Spray) as a means of achieving safe shutdown ,

conditions in the event of fire in five (5) areas (Fire Areas IX, X, XI, XVil, and XVill). These areas are described in the Licensee's revised analysis as satisfying the separation / protection requirements of Section Ill.G.2 of Appendix R. The use of low pressure injection systems represents a change from the Licensee's previously approved methodology which only credited the use of low pressure injection systems as an alternative shutdown capability in one area (Fire Ares XV, Torus).  ;

The Licensee's proposed reliance on the use of low pressure injection systems to achieve safe shutdown conditions in the event of fire in areas not requiring altemative shutdown does not satisfy Appendix R Section Ill.G.1 to the extent that its use will not allow hot shutdown conditions to be maintained. Additionally, this approach does not satisfy the shutdown system performance criteria of Section Ill.L of the regulation. As discussed in Section V of Information Notice 84-09, the shutdown system performance criteria of Section lil.L are also applicable to non-altamative shutdown systems. As a result of this finding, it is recommended that the Licensee ensure the availability of a high pressure injection system (i.e., RCIC or HPCI) in the event of fire in Fire Areas IX, X, XI, XVil, and XVill, or seek an exemption from the regulation. Details of this issue are discussed in Section 2.3.2 of this report.

The Licensee's Appendix R separation analysis methodology was found to be satisfactory. The Licensee's enalysis was performed systematically and was based upon detailed review of the FitzPatrick as-built condition and proposed modifications. The safe shutdown separation analysis was performed using an interactive data base management system to collect information on a fire area basis. The data base was computer processed v

through various constraints and iterations to determine the systems, components, cables, manual actions, procedures, and modifications required to demonstrate achievement of a safe shutdown capability for each plant fire area.

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1. REVIEW CRITERIA The criteria used in reviewing the Licensee's submittal are contained in the following documents: ,

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1. " Fire Protection Program for Operating Nuclear Power Plants," 10CFR50 Appendix R, l (45 FR76611, November 19,1980, and 46 FR 44735, September 8,1981).
2. Generic Letter 81-12, dated February 20,1981 l
3. NRC Memorandum To: D.G. Eisenhut From: R.J. Mattson,

SUBJECT:

" Fire Protection Rule Appendix R" dated March 22,1982 (Clarification of Generic letter 81-12)

4. NRC Memorandum TO: R.H. Vollmer, From: R.H. Mattson,

SUBJECT:

" Position Paper on Allowable Repairs for Alternative Shutdown and the Appendix R Requirements for Time Required to Achieve Cold Shutdown," dated July 21,1982.

5. Generic Letter 83-33, dated October 19,1983
6. NRC lE Information Notice 84-09, " Lessons learned from NRC inspections of Fire Protection Safe Shutdown Systems"
7. NRC IE Information Notice 85-09, " Isolation Transfer Switches and Post-Fire Safe Shutdown Capability"
8. Generic Letter 86-10, " Implementation of Fire Protection Requirements," April 24, 1986 1

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2. POST-FIRE SAFE SHUTDOWN CAPABILITY 2.1 Senaration of Safe Shutdown Functions Where components of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside the containment, the Licensee has provided one of the following three means of ensuring that one train of safe

' shutdown equipment remains free of fire damage: (1) Separation of equipment, cabling and associated circuits of redundant safa shutdown systems by a fire barrier having a 3-hour fire rating: (2) Separation of equipment, cabling and associated circuits of redundant safe 1 shutdown systems by a horizontal distance of more than 20 feet free of intervening i combustibles or fire hazards (in addition, automatic fire detection and suppression systems are installed in such areas); and (3) Separation of equipment, cabling, and associated circuits of redundant safe shutdown systems by a fire barrier having a 1-hour fire rating (in addition, automatic fire detection and suppression systems are installed in this area).  !

The Licensee's criteria for providing fire protection for safe shutdown functions '

satisfies Section lil.G of Appendix R, and, therefore, is acceptable.  ;

2.2 Post-Fire Safe Shutdown Methodoloav - General Plant Areas 2.2.1 Analysis Methodology The Licensee's methodology for assessing compliarice with the separation / protection  !

requirements of Section Ill.G of Appendix R consisted of: (1) determining the required plant functions (e.g., reactivity control, decay heat removal, etc.) necessary to achieve and maintain safe shutdown conditions; (2) developing safe shutdown logics that describe the various methods available to accomplish the required shutdown functions; (3) developing an interactive safe shutdown data base management system that identifies locations of equipment, routing of cables, power sources, power and control cables for shutdown related equipment and sort this information on a fire area basis: (4) identifying one or more safe shutdown methods for each plant fire area; (5) relocating cables and equipment, providing fire barriers and fire detection and suppression systems so as to meet the separation / protection requirements of Section Ill.G. of Appendix R.

The Licensee's post fire safe shutdown analysis methodology conforms to the requirements of Appendix R to 10 CFR 50, and is, therefore, acceptable.

2.3 Safe Shutdown Canability Safe shutdown conditions are achieved when the reactor is subcritical, the reactor coolant inventory is above the top of the core and decay heat is being removed at a rate that is approximately equal to its generation. The Licensee's safe shutdown analysis demonstrates that redundancy exists for systems needed for hot and cold shutdown, and has developed four shutdown methods capable of achieving safe shutdown conditions in the 2

event of a fire. The four methods are: (1) HPCI operated from the control room, (2) RCIC operated from the control room, (3) CS or LPCI operated from the control room, and (4) LPCI operated from the alternative control stations.

Shutdown of the reactor and initial reactivity control are accomplished by control rod insertion (scram) from the control room or by altemate means outside the control room.

Reactor coolant inventory is provided by isolation of the RCS and the use of either high or low pressure injection systems. Decay Heat Removal is initially accomplished by the RHR system in the suppression pool cooling mode. The alternate shutdown coolirq mode of RHR (water solid) provides long term core cooling necessary to achieve and maintain cold shutdown conditions.

2.3.1 Evaluation of Post Fire Safe Shutdown Systems For post fire safe shutdown, Appendix R to 10CFR50 provides the following performance goals as criteria for achieving and maintaining safe shutdown conditions:

  • Reactivity Control: capable of achieving and maintaining cold shutdown reactivity conditions. (For FitzPatrick cold shutdown is defined as a plant condition in which the reactor is subcritical and the reactor coolant temperature is less than 212* F).
  • Decay Heat Removal: capable of removing decay heat and provide sufficient capability to allow the transition from hot to cold shutdown conditions
  • Process Monitoring: capable of providing direct readings of the process variables necessary to perform and control the above functions
  • Support Functions: capable of providing the process cooling, lubrication, etc.,

necessary to permit the operation of the equipment used for the above safe shutdown functions In accomplishing the shutdown performance goals outlined above, the equipment and systems used to achieve and maintain hot shutdown conditions should remain free of fire damage and capable of maintaining such conditions for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, using offsite or on site emergency power.

2.3.1.1 Reactivity Control Function Reactivity control is accomplished by insertion of the control rods, either automatically by the Reactor Protection System or manually by operator action to initiate a reactor trip (scram). A manual scram may be accomplished from the control room or by alternate means out' side the control room by opening RPS trip breakers in the relay room, opening the output 3

breakers for the RPS Motor Generator Sets in the East and West Electric Bays or by isolating and venting the CRD instrument air header. Verification of control rod insertion is also available outside the control room at the Hydraulic Control Unit Scram Valves.

2.3.1.2 Reactor Coolant Makeup Control Function Reactor Coolant Makeup will be achieved by isolation of the Reactor Coolant System (RCS) and the use of either a high or low pressure injection system to control coolant level in the reactor. The High Pressure Coolant injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems provide water injection at high pressure using steam turbine-driven pumps that are normally aligned to draw suction from the Condensate Storage Tank (CST). The Licensee states that the CST is of sufficient capacity to replace water lost from the reactor vessel for over eight hours. If required, operators may align the HPCI or RCIC pumps to the suppression pool.

For fire events that may render the normally preferred high pressure injection systems unavailable, the analysis credits the use of the Automatic Depressurization System (ADS) valves in conjuncticn with Core Spray (CS) or Low Pressure Coolant injection (LPCI) mode of the Residual Heat Removal (RHR) system. With this approach, the ADS valves are manually actuated to rapidly depressurize the reactor to a point below the shut-off head of one pump of the Core Spray or RHR systein (LPCI mode). The CS or LPCI pumps take suction from the suppression pool. The Licensee's proposed use of low pressure injection systems in areas not requiring an alternate shutdown capability is discussed further in Section 2.3.2 of this report.

To ensure RCS inventory and pressure control, the integrity of the RCS pressure boundary must be maintained. The Licensee states that maintaining RCS pressure boundary integrity will be eccomplished by the Reactor Vessel isolation /High Low Pressure interface System (RVIS/High Low). This system, which was specifically created by the Licensee to address Appendix R High/ Low pressure interface concerns, provides the capability to isolate the main steam isolation valves and other valves connected to either the reactor vessel or reactor recirculation piping. Operation of the RVIS/High Low system is required regardless of the system being used to provide reactor coolant makeup.

2.3.1.3 Reactor Coolant Pressure Control Function Prior to a controlled cooldown and depressurization, the isolated reactor vessel is protected from overpressurization by eleven (11) relief valves that mechanically self-actuate when pressure is greater than their setpoints. To provide a controlled depressurization of the

  • reactor vessel (maximum 100'F/hr cooldown rate), each of the eleven relief valves is provided with remote-manual operating capability.

2.3.1.4 Decay Heat Removal In the event of a reactor trip coincident with a loss of off-site power, decay heat will initially be removed by natural circulation within the reactor and mechanical operation of the Relief Valves (RVs). Steam discharged from the RVs is condensed in the suppression pool.

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Cooling of the suppression pool will be accomplished by the RHR system in the suppression pool cooling mode. The Licensee states that separate analyses performed by General Electric (GE), demonstrate that one RHR heat exchanger loop, in the suppression pool cooling mode,  !

is sufficient to maintain suppression pool temperatures within acceptable limits. These i analyses include the case where low pressure injection systems are used to accomplish safe l shutdown. _

2.3.1.5 Process Motstoring ,

The process monitoring capability provided to accomplish post-fire safe shutdown includes the following instrumentation:

  • Reactor Pressure
  • Reactor Water Level
  • Suppression Pool Temperature
  • Suppression Pool Level
  • Drywell Temperature
  • Drywell Pressure The Licensee states that sufficient diagnostic instrumentation (e.g., flow and  !

pressure) for the HPCI, RHR, RCIC, and CS systems is available to monitor system  !

performance.  !

The instrumentation provided by the Licensee satisfies the requirements of Appendix R for post-fire safe shutdown, and is, therefore, acceptable.  !

2.3.1.6 Support Functions ,

i Systems and equipment available to support post fire safe shutdown include:

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  • AC Emergency Power System -includes EDGs, EDG support components and i distribution system switchgear  ;
  • 125 V DC Emergency Power System  ;

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With the exception of the Licensee's proposed use of low pressure injection systems as a means of accomplishing reactor coolant Makeup function in the event of fire in areas not requiring an alternative shutdown capability, the systems identified by the Licensee for  ;

achieving and maintaining safe shutdown in the event of fire are acceptable.

2.3.2 Use of Low Pressure injection Systems in the event of fire in areas not requiring an alternative shutdown capability (Fire Areas IX, X, XI, XVil and XVill) c 5 I l

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Discussion:

Section Ill.G.1.s of Appendix R requires one train of systems necessary to achieve and

  • maintain hot shutdown conditions to remain free of fire damage. Additionally, the safe shutdown system performance criteria specified in Section ill.L.1 and Ill.L.2 of Appendix R to 10CFR50 require, in part, that during the post-fire shutdown the reactor coolant system

., process variables remain within those predicted for a loss of normal a.c. power and that the reactor coolant makeup function be capable of maintaining the reactor coolant level above the top of the core.

By letter dated February 13,1984, the NRC staff issued Information Notice 84-09 (IN 84-09), " Lessons Learned from NRC Inspections of Fire Protection Safe Shutdown Systems (10 CFR 50 Appendix R)." The stated purpose of IN 84-09 was to provide guidance to power reactor facilities conducting analyses or making modifications to implement the requirements of Appendix R. With regard to safe shutdown systems and components identified in the fire hazard analysis or associated documentation,Section V of IN 84-09 states, in part: "The systems and equipment needed for post-fire safe shutdown are those systems necessary to perform the shutdown functions defined in section Ill.L of Appendix l R...The acceptance criterion for systems performing these functions is also defined in Section Ill.L...These guidelines apply to the systems needed to satisfy both Section Ill.G and Ill.L of Appendix R."

Based on the staff guidance provided in IN 84-09, the safe shutdown performance criteria of Section Ill.L appear to be applicable to all fire areas, including those for which an altemate shutdown capability is not required (i.e. fire areas that satisfy the redundant train separation criteria of Section Ill.G.2 of Appendix R).

The Licensee proposes the use of ADS /LPCI or ADS /CS as a means of achieving and i maintaining post-fire safe shutdown conditions in fire areas which do not require an altemate '

shutdown capability. However, this approach will not allow hot shutdown conditions to be maintained, and will result in a short-term uncovery of the core. Therefore, its use does not appear to satisfy Sections Ill.G or L of the regulation.

Evaluation:

The Licensee's revised (1992) analysis proposes the use of low pressure injection systems (ADS /LPCI or ADS / Core Spray) as a means of achieving safe shutdown conditions in the event of fire in five (5) areas not requiring an altemative shutdown capability.

Specifically, these areas include:

Fire Area IX, Reactor Building East Side El. 272',

Fire Area X, Reactor Building West Side El.272',

Fire Area XI, South Cable Tunnel El 286',

Fire Area XVil,iteactor Building East Crescent Area, and Fire Area XVill, Reactor Building West Crescent Area.

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This represents a change from the Licensee's previously approved methodology which only credited the use of low pressure injection systems as an attemative shutdown capability and in the event of fire in Fire Ares XV (Torus). The use of low pressure injection systems i in lieu of the normally preferred high pressure systems (i.e., HPCI or RCIC) in the event of fire

. in these areas has boon previously evaluated and approved by NRC in separate exemptions granted on a case-by-case basis. Specifically, by letters dated April 26,1983, September 15,1986, and September 10,1992, the NRC granted exemptions approving tfie use of ADS /LPCI as an alternate shutdown capability in the event of a fire in Fire Area Vil (Control '

Room, Relay Room and Cable Spreading Room), Fire Area ID (North Cable Tunnel) and Fire '

Area XVI (Battery Rooms Corridor Area).

By letter dated May 18,1994 the staff forwarded a Request for Additional Information (RAI) to the Licensee seeking additional information and clarification with regard to the Licensee's proposed methodology. In its response, dated July 22,1994, the Licensee states that it'its view the safe shutdown system performance requirements specified in Section lil.L apply only when altemate or dedicated shutdown systems are used to achieve and maintain cold shutdown. The Licensee's position appears to be based on its determination that since ADS /LPCI or ADS / Core Spray are not considered to be an altemate shutdown capability, the five additional areas in which it proposes to credit their use do not need to satisfy the system performance criteria specified in Section Ill.L of Appendix R. Based on this interpretation, the Licensee concludes that an exemption from the performance requirements of Section Ill.L is not required. The use of low-pressure injection systems (ADS /LPCI or ADS /CS) is not a preferred means of achieving safe shutdown conditions in a BWR. In its response to the staff's RAI, the Licensee concurs with this position and states that this approach (i.e.

' ADS /LPCI or ADS / Core Spray) will only be used when all other means of shutting down the reaMor are not available or when the use of high pressure systems must be avoided. While not the preferred approach, ADS /LPCI and ADS /CS do provide an approved capability for shutting down the reactor and, as discussed in NRC Memorandum From: L. S. Rubenstein, To: R. J. Mattson, dated December 3,1982, have been accepted by the staff for use as an altamative shutdown canability (emphasis added). It is important to note that the staff has only categorically accepted the use of low pressure injection systems without an exemption '

as a means of providing an altemative shutdown capability. The basis for this acceptance rests,in part, with the fact that Section Ill.G.3 of Appendix R requires fire detection and fixed fire suppression systems to be installed in all areas requiring an attemative shutdown capability. The additional fire safety features provided for these areas serve to limit the probability of fire growth and damage thereby reducing the plant's reliance on the altamate capability to accomplish safe shutdown conditions. Under the regulations, areas of the plant which do not require an altemate shutdown capability may not be provided with an equivalent level of fire protection.

Conclusion:

The provisions of Sections Ill.L.1 and lil.L.2 of Appendix R require that systems relied on to achieve safe shutdown be capable of maintaining the reactor coolant level above the _

top of the core. NRC staff interpretations and guidance, as presented in Information Notice 84-09 and NRC Memorandum From: L. S. Rubenstein, To: R. J. Mattson, dated December 3, 1982, indicate that the performance criteria of Section Ill.L are applicable to systems 7

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provided to satisfy Section Ill.G of Appendix R (i.e. non-altemate shutdown systems) and the use of low pressure injection systems (ADS /LPCI or ADS / Core Spray) without an exemption 4 from Section Ill.L is generally accepted only as an alternative shutdown capability.

Additionally, from a review of historical fire protection licensing issues related to the J. A.

FitzPatrick post-fire safe shutdown capability, it appears that in cases where the Licensee had previously proposed the use of low pressure injection systems it had recognized that the system performance limitations did not satisfy specified performance criteria, and requested an exemption.

The Licensee's stated position with regard to the applicability of shutdown system performance criteria contained in Section lil.L of the regulation appears to contradict NRC staff position presented in Section V of Information Notice 84 09.

Based on the above, it is recommended that the Licensee change the shutdown method for the 5 areas in question, or seek an exemption from the safe shutdown performance criteria of Sections Ill.L.1 and Ill.L.2 of Appendix R for Fire Areas IX, X, XI, XVil and XVill. The exemption process will enable the staff to fully evaluate the Licensee's proposed approach against the specific fire hazards and protection features provided for each area in a manner that is consistent with previous licensing actions.

2.4 Alternate Shutdown Canability 2.4.1 Areas Requiring Alternate Shutdown Capability A fire occurring in the following areas has the potential to prevent shutdown from the control room:

  • Fire Area Vil; consisting of the Control Room, Cable Spreading Room and Relay Room
  • Fire Area ID; North Cable Tunnel at 286' El.
  • Fire Area XVl; Battery Rooms Corridor Area The Licensee has provided an alternate shutdown capability that is independent of the main control room for the above areas. The alternate shutdown system utilizes existing plant systems and equipment as identified in Section 2.2, in conjunction with ten (10) remote and auxiliary control stations from which post fire safe shutdown can be accomplished. The alternative shutdown capability includes a Remote Shutdown Panel, designated 25-RSP, and five altamate shutdown panels, designated 25-ASP-1, 2, 3,4, and 5. The doors of each

, panel are key-locked and are provided with anti-tampering switches which alarm in the control room when the doors are opened. In addition the altemative shutdown capability relies on the use of the following four local control stations:

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(1) The Automatic Depressurization System (ADS) relief valve control panel (02 ADS-71) located adjacent to the Remote Shutdown Panel (25-RSP). This

. panel provides local control capability for eleven safety relief valves (2) The local control panels for Emergency Diesel Generator (EDG) B and D located in the Diesel Generator Room and in the Emergency Switchgear R_oom, in addition to providing electricalisolation of the EDGs from the control room, these panels provide local control, indication and metering capabilities for the B and D EDGs and the 4.16kV Emergency Bus Breakers.

(3) The reactor building vent and cooling panel located near 25 ASP-1, which contains isolation switches for the Division B Crescent Area Coolers, and (4) Instrumentation rack 25-6 located opposite the Remote Shutdown Panel 25-RSP.

Where necessary, the alternate shutdown systems include transfer / isolation switches to provide electricalisolation of safe shutdown components from the fire affected areas. The Isolation / Transfer switch design complies with the operability guidelines for alternate shutdown systems outlined in Information Notice (lN) 85-09.

The Licensee states that communications necessary to coordinate operator activities outside the control. room are available and emergency lighting units, having an 8-hour rating, are provided to enable operators to perform required activities.

The Licensee states that actions required to achieve stable hot shutdown. conditions can be accomplished within the first thirty (30) minutes following control room evacuation by six operating personnel (five operators and an Engineer on Shift). Within thirty (30) minutes from the initiation of the fire event, required safe shutdown systems and components will be available such that reactor vessel depressurization and RHR/LPClinjection can be performed.

The Licensee states that the extended operator action time of 30 minutes has been evaluated in an analysis performed by GE which concluded that the thirty minute operator action time does not pose any threat to the fuel cladding integrity or the suppression pool integrity. The adequacy of the GE analysis has been reviewed previously by NRC in evaluations dated April 26,1983 and September 15,1986 and found to be acceptable.

'2.4.3 Safe Shutdown Procedures and Manpower For alternate shutdown from outside the main control room the Licensee has developed plant abnormal operating procedure AOP-43, Plant Shutdown from Outside the Control Room. The procedure contains the steps necessary to implement ADS /LPCI method of shutdown from remote and auxiliary shutdown panels outside the main control room. The Licensee states that sufficient plant staf f and time are available to accomplish safe shutdown.

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2.4.4 Repairs No repair activities are required to achieve hot shutdown conditions. With the exception of establishing emergency ventilation fans for the Battery Room and Emergency Service Water Pump Rooms discussed below, no repair activities are required to achieve cold shutdown conditions. _

A fire in the Battery Room Corridor, Station Battery Charger Room A, Station Battery Room A, Station Battery Charger Room B, Station Battery Room B, Control Room, Relay Room, or Cable Spreading Room, may cause the loss of Station Battery Room and Charger Room ventilation equipment. In the event of fire in these areas the Licensee has developed abnormal operating procedure AOP-58, Station Battery Room Emergency Ventilation, to provide operator guidance for establishing an emergency ventilation capability through the use of portatJe fans powered from a mobile diesel generator. The restoration of ventilation for these areas is not an immediate operator action. The Licensee states that the establishment of emergency ventilation within two hours is sufficient to ensure continued operability of the charger and battery.

The restoration of Battery Room ventilation is not an immediate operator action and all required activities are govemed by written procedures. On this basis, the Licensee's method of providing an emergency ventilation capability for the Station Battery Room in the event of fire in the Battery Room Corridor, Station Battery Charger Room A, Station Battery Room A, Station Battery Charger Room B, Station Battery Room B, Control Room, Relay  :

Room, or Cable Spreading Room, is acceptable.

A fire in cartain areas of the plant may render the Safety-Related Pump Room Ventilation System inoperable. Fire damage to this system could impact the operability < t RHR and ESW SW pumps. The establishment of area ventilation is described by the Licensee ,

in its procedures as a long term action item to accomplish safe shutdown. In its 1992 submittal, the Licensee statas that a plant modification has been implemented to assure area cooling of one train of RHRSW/ESW pumps during a postulated fire. Required modifications will be completed by the end of the next refueling outage. In the interim, the Licensee has implemented pocedures (AOP-28, Section 10) which provide direction for operators to use portable fans powered by a mobile diesel generator as a temporary compensatory measure.

Pending the Licensee's completion of proposed modificcions necessary to achieve long-term compliance, the proposed interim measures of providing emergency ventilation to the Safety Fielated Pump Room are acceptable.

The altemative shutdown capability provided for a fire in Fire Area Vil, consisting of the Control Room, Cable Spreading Room and Relay Room: Fire Area :D, North Cable Tunnel at 286' El.: and Fire Area XVI, Battery Rooms Corridor Area, meets the requirements of Appendix R and is, therefore, acceptable.

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3. ASSOCIATED CIRCulTS 3.1 Common Power Sunniv Associated Circuit Concem I 1

1 The common power supply associated circuit concern arises when equipment required for safe shutdown shares a common power source (e.g., switchgear, MCC, circuit breaker or fuse parml) with non-safe shutdown equipment and fire-induced electrical faults in the non-essential loads will cause a loss of the power source due to inadequate fire protection i features (i.e. protection per Section Ill.G of Appendix R) or circuit protective device l coordination. Proper coordination of electrical protection devices ensures that the protective l device located nearest the fault will operate prior to any protective device located upstream i of a required power source.

i As part of the conduct of its safe shutdown separation analysis, the Licensee identified all required electrical power sources. Protection for the associated circuit common  :

power supply was then demonstrated by the performance of circuit coordination studies.

These studies confirmed that Proper coordination per Appendix R exists and that required power supplies are adequately protected. Where circuits were identified as not being properly coordinated they were included as required circuits in the Appendix R separation analysis.

The Licensee has also evaluated for the effects of fire-induced high-impedance faults on power cables associated with required power supplies. If it was determined that the total load current resulting from such faults was greater than the trip setting of the main supply breaker, a high-impedance fault concern was identified and compensatory actions (i.e.,

shedding of non-essential loads) are taken to isolate the faulted cables.

7 The Licensee's method of protection for the common power supply associated circuit concern satisfies Appendix R requirements and is, therefore, acceptable.

3.2 Common Enclosure Associated Circuit Concem:

The common enclosure associated circuit concern occurs when non-safe shutdown i circuits are routed together with cables of required equipment and they are not provided with a suitable level of electrical protection, or fire can destroy both circuits due to inadequate fire protection features.

The Licensee states that all electrical distribution equipment and cabling is provided  ;

with suitably sized electrical fault protective devices which provide the necessary degree of '

protection frorn electrical fault and overload conditions. Additionally, the Licensee states that fire stops are placed between safe shutdown raceways where a fire could propagate and -

damage redundant divisions.

The Licensee's method of protection for the common enclosure associated circuit concem satisfies Appendix R requirements, and is, therefore, acceptable. .

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- 3.3 Sourious Sianal Concem i

As part of its Appendix R safe shutdown separation analysis the Licensee analyzed '

safe shutdown and associated circuit cables for potential spurious signal concerns. When cables of equipment whose spur'>us operation could affect safe shutdown were identified they were then included as required cables into the Appendix R separation analysis. The identified cables and equipment were then treated in the same manner as active arid passive  ;

mein and support components required to achieve and maintain safe shutdown of the reactor .

in the event of fire. l The Licensee's method of protection for the spurious signal associated circuit concern satisfies Appendix R requirements, and is, therefore, acceptable.

3.4 Hioh/ Low Pressure interfaces .

A highMow pressure interface consists of the boundary between the high pressure reactor coolant system and any low pressure system piping. In the event of a fire-initiated spurious opening of valves which comprise the highMow pressure interface boundary, a flow path would develop into the low pressure system piping resulting in an unisolable loss of coolant accident (LOCA).

The following are highMow pressure interfaces: RHR Shutdown Cooling isolation Valves and the RHR Steam Condensing isolation and Control Valves. The means for' preventing the spurious operation of these valves is as follows: Spurious operation of RHR Shutdown Cooling isolation Valve interface is precluded by pre-fire strategy to ensure that the -

remote disconnect switch for one valve 10MOV-18) is locked open during normal power i operation. Spurious operation of the RHR Steam Condensing Isolation Valve interface is precluded by pre fire strategy to ensure that the feedbreaker of the affected valves is locked open during normal power operation. It should also be noted that in addition to the pre-fire strategy of de-energizing' the highAnw pressure interface components during power operation. I the Licensee has been granted an exemption (reference: NRC SER dated April 26,1983) from the requirement to provide protection against simultaneous three-phase AC and two-wire DC circuit faults.  !

As part of its analysis, the Licensee has identified the following reactor coolant system ,

boundary valves whose fire-induced spurious operation, while not causing a breach of a  !

highAow pressure interface, would cause a significant loss of inventory: HPCI Steam Line i isolation Valves, RCIC Steam Line Isolation Valves, Main Steam Line Isolation Valves, Main Steam Line Drain isolation Valves, Main Steam Relief Valves, and RV Head Vents. The i

Licensee has evaluated the potential for spurious operation of those valves on a fire area basis. Where necessary, plant modifications have been implemented to ensure the isolation j of these valves.

The Licensee's method of protection for High/ Low pressure interfaces is in accordance with Appendix R requirements and Generic Letter 86-10, and is, therefore acceptable. j 12 i

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4. CORRECTIVE ACTIONS During its post-fire safe shutdown reverification and separation analysis the Licensee identified conditions of non-compliance with Appendix R and has initiated permanent corrective actions necessary for their resolution. Conformance measures include the following modifications:

4.1 LPCI Altemate Power Sunolv Circuit Modification This modification provides a control scheme which will enable plant operators in the control room, to isolate the LPCl injection valve independent power supplies and connect a maintenance bypass (Alternate Feed) from another safety related emergency MCC in the same safety division to the valve bus. ,

1 4.2 Unorade of CST Level Inst.ument Loon Power Sunolv l The Condensate Storage Tank (CST) levelinstruc.entation loop is powered from a non-UPS backed power source. In the event of -a loss of offsite power, CST level indication would be lost. To ensure continuous CST level indication, this modification provides an uninterruptable source for CST level instrumentation.

4.3 Isolation of Main Steam Isolation Valves The existing MSIV isolation capability did not satisfy Appendix R requirements.

Specifically, there was no provision to isolate the AC and DC solenoid control circuits for the MSIVs at the at Auxiliary Shutdown Panel 25 ASP-1. Therefore, a fire-induced hot short resulting from fire in the control room, could keep either the AC or DC solenoids, or both, from de-energizing, thereby preventing closure of the MSIVs.

  • This modification installs four dedicated Isolation Switch / indication Light Modules (one module for each of the four MSIVs) on a new Auxiliary Shutdown Panel (25 ASP-4) located ,

outside the control room at elevetion 300' of the Administration Building. The AC and DC coil circuits and indicating light circuitry are isolable from 25 ASP-4.  :

4.4 isolation of Safety Relief Valves There are eleven DC operated Safety Relief Valves (SRVs), of which seven Automatic Depressurization System (ADS) valves are automatically controlled by relay logic circuits. l The remaining four SRVs are manually controlled. For each valve, one of the two solenoids j is operable from the control room. The other solenoid is operated from the Local SRV Control l Panel (02 ADS-071) located at the 300' elevation of the Reactor Building. The solenoids are  !

powered from redundant DC power sources. In the event of fire requiring control room evacuation, all eleven SRVs can be operated manually at the Local SRV Control Panel.  ;

However, there was no provis on for isolating the SRV solenoids from the control room. A l Control Room or Reactor Building fire could induce a hot short and spuriously open these valves irrespective of the position of control switches located in the Control Room.

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This modification installs a dedicated isolation switch for each of the eleven SRVs in

- a new Auxiliary Shutdown Panel (25 ASP-5) located outside the control room on the 300' olevation of the Administration Building. Thus. upon isolation, a control room fire will not impact circuitry required for SRV operation. ,

4.5 Safety Related RHRSW/ESW Pumo Room Ventilation To assure area cooling of one train of RHR/ESW Pumps Ming a postulated fire, this modification will modify the Safety Related Pump Room Ventilation System. It should be  ;

noted that the Licensee has been granted a temporary exemption with respect to the Safety  !

Related Pump Room Ventilation System (Reference NRC evaluation dated September 10, ,

1992). ,

4.6 Isolation of RHR Valves for Reactor Rd!dino Fire To preclude the spurious operation of eight RHR Valves (10MOV-16A,10MOV-16-B, -

10MOV-25A,10MOV-27A,10MOV-258,10 MOV-278,10MOV-66A, and 10MOV-6GB) ar, a result of fire in the Reactor Building, this modification provides eight (8) key-locked selector switches with indicating lights that will enable the control room operator to bypass (manually override) logic and control for the RHR valves. The key-locked switches are two-position (normal and bypass) devices with the key removable in the normal position. Additionally, as ,

part of this modification, cable 1RHRBBC120 which is required to support the operation of  :

RHR valve 10MOV-168, will be rerouted outside the area (Fire Ares X) where operation of 10MOV-168 is required.  !

4.7 Reroute /Wran Safe Shutdown Cables  !

The Licensee's revised safe shutdown separation analysis identified the need to re-route four cables (IDMSBBKO15,1FPSNNC233,1FPSNNC235, AND 1RHRDBHOO4) and provide a one-hour fire rated wrap for power cable 1 ABVBBKO55. Specif!:: ally, this modification inc!udes the following four activities: .

r (1) Rerouting of cable IDMSBBKO15 out of Fire Areas IA, Vll, and XVI, to protect  !

the reactor isolation capability by ensuring that valves 23MOV-60 and 23MOV- l 77 can be maintained in the closed position in the event of fire in these areas, (2) Failure of either 1FPSNNC233 or 1FPSNNC235 as a result of fire in Fire Area l' ll could cause the Train A (71HO5) Switchgear Room CO2 System to spuriously actuate. Since Fire Area 11 relies on A-Train power for safe shutdown, this modification will reroute these cables outside the affected fire area (Fire Area ll).

(3) RHR Pump power cable 1RHRDBHOO4 is re-routed outside Fire Area IA to l ensure the availability of RHR Pump 10P-3D in the event of fire in this area. '

(4) To ensure the availability of Battery Room B Ventilation, the raceway containing cable 1 ABVBBKOO5 will be wrapped in Fire Area IC.  ;

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, I 4.8 Emeraency Service Water Pumo isolation i The Emergency Service Water Pump (46P-28) is fed from 600V Switchgear 71L26.

The breaker cloamg circuit for this pump is monitored by a loss of power relay located in Aux Relay Cabinet AR-68, via cable 1ESWBBC098. In the event of fire in Fire Area Vil (Control Room Cable Spreading Room or Relay Room), a short could develop in this cable, which could cause a loss of breaker control from Auxiliary Shutdown Panel 25 ASP-3 after Control Room Evacuation.

This modification ensures the operation of ESW Pump 46-28 from the Auxiliary Shutdown Panel 25 ASP-3 by providing an isolation capability for the relay circuit which enables the operator to override a possible short in cable 1ESWBBC098.

4.9 Reactor Head Vent Valves Reactor Head Vent valves (02AOV-17) and (02AOV-18) are in series and could spuriously open in the event of a fire in the following fire areas: Fire Area Vil (Contml Room, Cable Spreading Room and Relay Room), Fire Area IC (West Cable Tunnel: El. 260'), and Fire Area X (Reactor Building; El. 272'). At least one cf these valves must remain closed to '

maintain pressure integrity of the reactor.

To ensure that at least one valve will remain closed in the event of a Control Room fire  !

(Fire Aron Vil), this modification uses a spare contact in an existing isolation switch located i on remote shutdown panel 25RSP, to provide controlisolation capability for valve 02AOV-17.

Additionally, the control circuits for both valves (02AOV-17 and 02AOV-18) are separated by j installing a new cable for valve 02AOV-18. This cable will be routed through different fire  ;

areas and a different penetration than those accessed by the cable run for 02AOV-17. j 4.10 Containment Sorav Valves  !

Containment Sprav Valves (10MOV-2SB and 10MOV-318) are in series and may j spuriously open in the event of a fire in Fire Area Vil (Control Room, Cable Spreading Room ,

and Relay Room). These valves must remain closed to maintain proper operation of the RHR system.

To ensure that at least one of the two series Containment Spray Valves will remain closed in the event of a control room fire, this modification provides isolation capability snd alternate controis for valve 10MOV-26B on Auxiliary Shutdown Panel 25 ASP-3.

4.11 Relocation of Fire Protection Panel 76CO2 PNL-8 Panel 76CO2-PNL-8 octuates the CO2 system for the North Emergency Switchgear Room and closes the emergency ventilation dampers for the room. The North Emergency Switchgear Room provides poveer for Division B safe shutdown systems. Panel 76CO2 PNL-7 acteates the CO2 system for the South Emergency Switchgear Room and closes the emergency ventilation dampers for the room. The South Emergency Switchgear Room provides' power for Division A safe shutdown systems. l l

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o Equipment powered from Safe Shutdown Division B is used in the event of fire in the Screenwell Area. However, a fire in the Screenwell may impact either or both of thqse panels. Since Safe Shutdown Division B equipment is used in the event of fire in either the South Emergency Switchgear Room or the Screenwell area, this modification relocates fire protection panel 76CO2-PNL-8 to the Turbine Building, where Safe Shutdown Division A equipment will be relied on to achieve safe shutdown. i 4.12 HPCI Valve Circuit Modification Steam supply line isolation is necessary to shutdown the HPCI turbine. A fire in Fire .

Ares IX (Reactor Budding at El. 272') may cause cable damage which would prevent steam i supply line isolation. In addition, fire damage to cables associated with HPCI valve logic '

circuits could cause spurious operation of valves that may prevent HPCI isolation or HPCI turbine operation.

To ortsure the capability to isolate the HPCI turbine steam supply line during a plant I fire, affected cables were rerouted and control circuits for the HPCI trip solenoid valve were  !

modified. To ensure proper operation of HPCI valves (23MOV-15 and 23-MOV-16) HPCI valve logic bypass switches are provided for each valve. The bypass switch will allow the operator to bypass the HPCI logic and control valve operation in the event of fire.

4.13 BCIC Valve Control Circuit Modification A fire in Fire Areas 11, VI, Vill, and X may disable RCIC control, prevent the operation of the RCIC speed controller, or prevent the ability to isolate the RCIC steam line or operate the RCIC turbine.

A fire in Fire Area 11 could impact the existing control fuses for the RCIC control circuitry, thereby disabling RCIC. To ensure that a fire in these areas will not disable RCIC control this modification adds separate fuses that are wired upstream of the existing RCIC control circuit fuses.

A fire in Fire Areas ll, VI, and Vill may impact cables associated with the RPV low level water permissive to the RCIC system. This would prevent RCIC from initiating. This modification provides a RCIC manual initiation pushbutton switch, to onsure a fire will not prevent the operation of RCIC due to a loss of low RPV level permissive logic signals.

A fire in Fire Area X could preclude the ability to isolate the RCIC steamline. Isolation of the RCIC steamline is necessary to support the operatha of the LPCl/ ADS shutdown method. To ensure that a fire will not prevent RCIC steamline isolation, this modification re-routes the affected cable out of Fire Areas X and 11, the trip circuitry for RCIC valve 13HOV-1 is rewired to a separate power supply, and a bypass switch capability has been provided for the RCIC logic valve 13MOV-16.

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  • 4.14 Installation of Additional Emeroency Liahtina Units This modification resolves the emergency lighting inadequacies identified during the  ;

NRC Appendix R inspection and during the reanalysis of Appendix R compliance. The plant areas covered within the scope of this modification are the Heater Bay, Electrical Bay, Administration Building and Emergency Diesel Generator Building. _

4.15 Relocation of Fire Protection Panel 76CO2-PNL-4 f i

Panel 76CO2-PNL-4 actuates the CO2 system and closes the emergency ventilation i dampers for the Relay Room, an alternate shutdown fire area. The panel is located in the -

Administration Building Hallway and control cables for the Relay Room ventilation dampers  ;

are routed through the South Cable Run Room. A fire in the Administration Building Hallway i or the South Cable Run Room could actuate the CO2 system, close the Relay Room dampers i and de energire the air handling units, thereby preventing adequate ventilation to the Relay }'

Room.

i This modification relocates fire protection panel 76CO2-PNL-4 and it associated pressure switch from the Administration Building Hallway to the Control Room Ventilation  !

Room located in Fire Area Vil/ Zone CR-1 at the 300' elevation. Some cable routing for panel 1 76CO2-PNL 4 will remain in the South Cable Run Room. Howevar, due to changes in the functions of these cables, a fire in this area will no longer impact the Relay Room ventilation ,

system.

4.16 Torous Switch Rewirina for Alternate Shutdown MOVs This modification was performed in response to NRC Information Notice 92-18 and requires the rewiring of torque switches for 22 motor-operated valves required to operate for alternate shutdown. This modification will prevent the bypassing of the torque and associated limit switches by hot shorts in the valve control circuitry that may occur as a -

result of a Control Room fire.

l 4.17 Alternate Anoendix R Reactor Vessel Level Instrumentation A fire in Fire Areas Vill or IX of the Reactor Building may impact the operability of redundant reactor vessel level transmitters by causing the water contained in the ,

instrumentation tubing to boil, resulting in an erroneous indication in the control room.

To ensure the availability of reactor vessellevelindication, this modification installs an i alternate level indicator in Fire Area X.

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. 4.18 Ensure' Availability of Motor Control Center and Unit Coolers and Re-route Cables for a Fire in Fire Area lA and Vll To ensure the availability of required electrical distribution system equipment (4160V SWGR 71L26,600VMCCs 71MCC-263 and 71MCC-261, and 125VDC panels 71BCB-2B and 71BMCC-4) in the event of fire in Fire Area IA, this modification re-routes the following  !

cables outside of Fire Area lA: (1)600V power feeder cables (1C2EFBLO85 and 1C2EFBLO86) between Switchgear 71L26 and MCC 71MCC-263 and (2) 125VDC feeder cables (1 DMSBBLOO1, I DMSBBLOO2, I DMSBBLOO13, I DMSBBLOO4, I DMSBBLOO5, and IDMSBBLOO6),

in the event of fire in Fire Area Vil (Control Room, Cable Spreading Room and Relay Room) Electric Bay Unit Coolers 67UC-1681 and 67UC-16B2 are required to be operational.

To ensure the availability of this equipment, this modification relocates the combination starters for the Unit Coolers from their existing MCC (71MCC 261) to a new power source (71 MCC-262). Since a fire in Fire Area Vil will have no effect on the operation of 71MCC-262, Electric Bay Unit Cooler operation will no longer be susceptible to loss as a result of fire in this area.

4.19, Conclusion The corrective actions described above provide acceptable methods of resolving the identified conditions of non-compliance, and are, therefore, acceptable, a

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b ATTACHMENT 2 SPlB SALP INPUT Plant Name: James A. Fitzpatrick Nuclear Power Plant SER

Subject:

Evaluation of licensee's safe shutdown reassessment and capability TAC Mo. M84780 Summary of Review / inspection Activities The licensee submitted its safe shutdown reassessment and capability. This evaluation was performed by our contractor Brookhaven National Laboratory (BNL).

Narrative Discussion of Licensee Performance Functional Area The licensee provided sufficient information to perform the subject review.

The technical information and supporting justifications for reassessment of safe shutdown capability provided in their submittal was acceptable.

Principal Contributor: A. Singh Date:

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0 W. Cahill L Our contractor, Brookhaven National Laboratory (BNL) reviewed the submittal and the post-fire safe shutdown methodology and analysis of associated circuits. Enclosed is the Technical Evaluation Report (TER) from BNL. The NRC staff has reviewed the TER and agrees with the BNL conclusions. It is recommended that ye'_ seek an exemption from the requirements of Section III.L of Appendix R for the five additional areas. Alteratively, you can ensure the availability of a high pressure injection system for these areas. The details of this issue are discussed in Section 2.3.2 of the attached TER.

Sincerely, Original signed by:

C. E. Carpenter, Jr., Project Manager Project Directorate I-l Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Docket No. 50-333

Enclosure:

Technical Evaluation Report cc w/ encl: See next page DISTRIBUTION:

Docket File PUBLIC PDI-l Reading SVarga JZwolinski LHarsh Slittle CECarpenter 0GC ACRS (4)

CCowgill, RGN-I DOCUMENT NAME: G:\FITZ\ FIT 84780.LTR Ts receive a copy of this document, indicate in the boa: "C" = Copy without enclosures *E* = Copy with enclosures *N* = No copy 0FFICE LA:PDI-l ol) l PM:PDI-l lAf D:PDI-l 3 l ET l l NAME Slittlc4NV CECarpenter:smm LMarsh I

[b -

DATE 09/:)/95 09/g /95 (dt- 09/ (/95 09/ /95 09/ /95 0FFICIAL RECORD COPY I

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