IR 05000322/1984046

Insp Rept 50-322/84-46 on 841203-07.Deviation Noted:Fire Doors Degraded Due to Security Mods,Structural Steel Fireproofing Damaged in Charcoal Filter & Chiller Room & Lack of Emergency Lighting in Specific Locations

ML20140E193
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	12/19/1984
From:	Anderson C, Pullani S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20140E184	List: IR 05000322/1984046 ML20140E178 ML20140E187
References
50-322-84-46, NUDOCS 8501100615
Download: ML20140E193 (27)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /84-46 Docket N License N CPPR-95 Priority Category B Licensee: Long Island Lighting Company 175 East Old Country Road Hicksville, New York 11801 Facility Name: Shoreham Nuclear Power Station

' Inspection At: Shoreham, New York Inspection Conducted: December 3-7, 1984 Inspectors: pulleur ~ / 2 -/ 9- 8i S. V. Pdlla3 V Fire Protection Engineer date Also p'articipating and contributing to the report were:

A. Fresco, Mechanical Systems Specialist, BNL D. Kubicki, Chemical Engineering Branch, NRR H. Thomas, Electri al System Specialist, BNL

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Approved by: / .

LtAaco /L /

C. J. Mders41, Chief, date Plant Systems Section Inspection Summary:

Inspection on December 3-7, 1984 (Inspection Report 50-322/84-46)

Areas Inspected: Special, announced team inspection of emergency lighting and the. safe shutdown capability of the plant in the event of a fire. The inspec-tion involved 170 inspector hours on-site and 66 inspector hours in-office by the team consisting of 4 inspector Results: No violations were identified. Eight deviations were identified in two areas. Eleven items remained unresolved at the end of the inspectio !

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DETAILS 1.0 Persons Contacted 1.1 Long Island Lighting Company '(LILCO)

• L. Britt, Supervisor - Nuclear Licensing

• J. Carney, Corporate Fire Protection Engineer H. Carter, Operations Engineer

• E. Dean, Assistant Operations Engineer R. Diem, Nuclear Station Operator

• M. Giannattasio, Senior Project . Engineer - Electrical

• R. Grunseich, Supervisor - Nuclear Licensing R. Gutmann, Fire Protection Program Manager J. Guttieri, Nuclear Assistant Station Operator J. Haverly, Supervisor -_ Systems Engineering (Impe11 Corporation)

J. Johnson, Equipment Operator

• Kibinal, Director - Quality Assurance, Safety & Compliance W. Laovis, Nuclear Assistant Station Operator N. Lents, Licensing R. Loper, Manager - Operations Staff

• B. McCaffrey, Manager - Nuclear Licensing & Regulatory Affairs P. Miller, Watch Supervisor

• A. Muller, QC Division Manager

• R. Paccione, Nuclear Systems Supervisor P. Quinan, Fire Protection Supervisor

• G. Rhoads, Compliance Engineer (Impe11 Corporation)

• Steiger, Plant Manager M. Vasely, Nuclear Engineer (Systems)

• J. Wynne, Compliance Engineer

• E. Youngling, Manager - Nuclear Engineering Department 1.2 Stone and Webster Engineering Corporation (S&W)

• R. Gauthier, Lead Power Engineer

• J. Murphy, Licensing Engineer

• Poltrino, Lead Control Engineer 1.3 Nuclear Regulatory Commission (NRC)

• C. Anderson, Chief, Plant Systems Section ,

• P. Eselgroth, Senior Resident Inspector l

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• Denotes those present at the exit meetin .0 Purpose This inspection was to ascertain that the licensee is in conformance with his previous commitments with respect to the emergency lighting and with l

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respect to the safe shutdown capability of the plant in the event of a fire. .0 Background

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10 CFR 50.48 and Appendix R of 10 CFR 50 became effective on February 17, 1981 for plants licensed prior to January 1,1979. .For

. plants. licensed or to be licensed after January 1, 1979 (Shoreham falls

under.this category), 10 CFR 50.48 and Appendix R are invoked by the licensing process which includes a review of the Fire Protection Program for conformance with the Standard Review Plan (NUREG-0800), Section 9.5.1, dated July 1981 or its previous version, BTP APCSB 9.5-1 including its Appendix A. Shoreham was reviewed against the latter documen The review of the licensee's Fire Protection Program is documented in the i Safety Evaluation Report (SER) dated April 1981 and its supplements 1 through 4. Various licensee commitments are documented in the SER, its supplements, and several licensee submittals. These commitments were used by the team as bases for the inspectio .0 . Correspondence '

'All correspondence on the subject, between the licensee and the NRC, was reviewed by the inspection team in preparation for the site visit.

Attachment 1 to this report is a listing of the correspondence reviewed.

j 5.0 Post-Fire Safe Shutdown Capability

, The licensee's post-fire safe shutdown analysis of the fire protection provided for safe shutdown equipment is presented in three submittal By letter dated May _ 21, 1981, the licensee provided a comparison of the plant

' design with Appendix R. The licensee also'provided a separation analysi ;of cables within the reactor building by letter dated February 10, 1981

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' and analysis of shutdown circuits outside the reactor buildingLby letter dated July '0, 198 The licensee's post-fire safe shutdown analysis demonstrated that systems

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needed for hot shutdown and cold shutdown are redundant and that one of the redundant systems needed for safe shutdown would be free of fire damage, by providing separation, fire barriers, and/or alternative shutdown capabilit '

J 5.1 Systems Required for Safe Shutdown The safe shutdown analysis considered components, cabling, and

support equipment for systems needed to shut down. Thus, in the event of a fire, at least one train of systems free of fire damage s-would be available to achieve and maintain hot shutdown and to proceed to cold shutdown. For hot shutdown,.at least one of-the following shutdown systems would be available: (1) the Reactor Core Isolation Cooling System, (2) the High Pressure Coolant Injection System, and

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(3) a combination of the Pressure Relief System (safety / relief valves),

the Core Spray System and Residual Heat Removal (RHR) syste For cold shutdown, an appropriate portion of the RHR system would be available. Attachment 2 to this report provides additional details of .the logic used for the shutdow !

Support systems and equipment are also required for safe shutdown. complate list of safe shutdown systems and equipment is given in the liceasee'* Cable Separation Analysis Report (CSAR), Section 4.3.

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5.2 Safe Shutdown capability for Various Fire Areas For equipment located in the primary containment, no fire protection features are provided because the containment atmosphere will be inerted during normal operatio For equipment located in the reactor building (secondary containment),

the licensee provided a cable separation analysis which divided the reactor building into overlapping 45 degree segments. The licensee assumed that all components, the cables and raceways,-in a given segment were lost due to a fire; yet demonstrated that the capability to shut down still existed. .NRC reviewed the cable separation analysis and concluded in Supplement 1 to the SER, Section 9.5.6, that it isoan acceptable method of demonstrating that adequate separation exists between the redundant trains. Additionally, the licensee has committed (by July 10, 1981 letter) to verify that the

"as-built" design has a minimum 20-foot separation between redundant safety-related component For equipment in areas outside the reactor building, the licensee has identified seven areas which contain cables for redundant shutdown'

equipment: the relay room, the control room, the diesel generator-rooms, the emergency switchgear room, the fuel oil pumphouse rooms, the screenwell, and the HVAC roo In the diesel generator rooms, the emergency switchgear room, the

fuel oil pumphouse rooms, and the screenwell, redundant equipment is separated by 3-hour fire-rated barriers. Cabling to this equipment is contained in underground ducts. In the event that fire disables redundant equipment in the HVAC room, control room, or relay room, a remote shutdown panel is provided in the reactor building'(see Supplement 1 to the SER, Section 9.5.6).

5.3 Remote Shutdown Capability Sections 7.4.1.4, 7.5.1.4, and 7.5.1.5 of the Final Safety Analysis

, Report (FSAR) describe the remote shutdown panel's design an o capabilit By letter dated May 21, 1981, the licensee addressed l

'Section III.L of Appendix R. The design objective of the remote i shutdown panel is to achieve and maintain cold shutdown in the event i

of a fire disabling the relay room, the control room or HVAC room; '

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'l The RCIC system, safety / relief valves and one division of the RHR system can be controlled from the remote shutdowr. panel to achieve cold shutdow The design objective of the remote shutdown panel is to conform with

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the performance goals outlined in Section III.L. Reactivity control will be accomplished by a manual scram before the operator leaves the control room. The RCIC system will provide reactor coolant makeup and the RHR system and the safety relief valves will be used for reactor heat removal. Reactor water level, reactor pressure, sup-pression pool water level and temperature, and drywell pressure and

, temperature are among instrumentation available at the remote shutdown panel to provide direct reading of process _ variables. The remote shutdown panel also includes instrumentation and control of support

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functions needed for the shutdown equipment. Transfer switches are installed on this panel to electrically isolate the shutdown circuits from the Control Room / Relay Room and to provide a redundant fused control circuit. Procedures for use of the remote shutdown panel include sequencing of equipment and operator action .0 Inspection Methodology The inspection team examined the licensee's capabilities for separating and protecting equipment, cabling and associated circuits necessary to achieve and maintain hot and cold shutdown conditions. This inspection sampled. selected fire areas which the licensee had identified as being in conformance with BTP APCSB 9.5-1/ Appendix The following functional requirements were reviewed for achieving and maintaining hot and cold shutdown:

• Reactivity control

Pressure-control

• Reactor coolant makeup

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• Decay heat removal

• Support systems

• Process monitoring The inspection team also examined the licensee's capability to achieve and maintain hot shutdown and the capability to bring the plant to cold shutdown condition in the event of a fire in areas where remote shutdown capability is provided. The examination included a review of the draw-ings for the remote shutdown capability and review of the procedures for achieving the remote shutdown. ' Drawings were reviewed to verify electrical-independence from the areas of concer Procedures were reviewed for general content and feasibilit Also inspected were fire detection and suppression systems and the degree of physica1' separation between redundant trains of Safe Shutdown Systems

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(SSSs). The team review included an evaluation of the susceptibility of r ,_ r - , . _ . . . - , . - _ , . , _ , .-...m- 4... ,.,_,.. ._ _._-...____.~_,.+--r- -

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6 a the SSSs for damage from fire suppression activities or from the rupture

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or inadvertent operation of fire suppression system The inspection team examined the licensee's fire protection features i

provided to maintain one train of equipment needed for safe shutdown i free of fire damage. . Included in the scope of this-effort were fire

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area boundaries, including walls, floors and ceilings, and fire protec-tion of openings such as fire doors, fire dampers, and penetration seal The inspection team also examined the emergency lighting for areas of the plant necessary for safe shutdown.

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7.0 Inspection of Protection Provided to Safe Shutdown Systems 7.1 Protection in Various Fire Areas

, The team reviewed the protection provided to SSSs in selected fire

areas for compliance with BTP APCSB 9.5-1/ Appendix R. The following L, fire areas were inspected:

• Battery Room A

• Battery Room B

+- Emergency Switchgear Room 101'

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Emergency Switchgear Room 102

• Emergency Switchgear Room 103

Diesel Generator Room 101'

Diesel Generator Room 102

Diesel Generator Room 103 The cables of redundant divisions in the above areas are completely enclosed by a three-hour rated wal Safe shutdown cables of a

~ division do not cross over and mix with-the cables of its redundant

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divisio The team did not identify any unacceptable conditions except as

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follows: '

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' Backup Information'for CSAR Backup information required to evaluate the licensee's Cable'Separa-tion Analysis Report was not available at the site for the team to'. <

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review:at the time of the inspection. This information is required:

to' confirm. the cable separation in secondary containment. .The licen- *

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see stated that such information can be-generated from the computer program stored in Stone & Webster's office ~at Boston and can be made

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available to NR This is an unresolved item, pending receipt and review of the infor-mation by NRC (50-322/84-46-01). '

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7.2 Safe Shutdown Procedures 7.2.1 Procedure Review

The team reviewed the following safe shutdown procedures: '

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SP 29.022.01, Shutdown from Outside Control Room, Revision 4 \

SP 23.133.01, Remote Shutdown Control System, Revision 5

SP 29.010.01, Emergency Shutdown, Revision 4 i

The scope of review was to ascertain that the shutdown could be attained in a safe and orderly manner, to determine the level of difficulty involved in operating equipment, and to verify that ,,

there was no dependence on repairs for achieving hot shutdow N For purpose of the review, a repair would include installing electrical or pneumatic jumpers, wires or fuses to perform an action required for hot shutdow The team did not identify any unacceptable conditions except as follows: No Reference to the Possibility of a Loss of Offsite Power Condition for Shutdown from Outside the Control Room The primary procedure, SP 29.022.01, does not currently provide any directions to the. operator in the event that off-site power is lost. The licensee has initiated a Station Procedure Change Notice (SPCN) 84-1656 to correct this condition. The team reviewed the SPCN and found it acceptable. This item is close Minimum Actions to be Performed in the Control Room Not Identified In the immediate actions of the primary procedure, reference is made to the Emergency Shu.tdown Procedure, SP 29.010.0 The latter procedure is designed for. control room operation only and does not specify the minimun, operations to be taken prior to evacuating the control roo The licensee agreed to resolve this by relocating.a note in the primary procedure,.SP 29.022.01, that as much of the Emergency Shutdown Procedure as possible should be performed; from-the control room prior to evacuatio This change is also under SPCN 84-1656 previously mentione The team reviewed the SPCN and found that this item is satisfactorily resolved. This item is close <

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7.2.2 Procedure Walk-Through The team walked through selected portions'of the procedures to determine that shutdown could be attained in an orderly and timely fashion. The team did not identify any unacceptable conditions except as follows:

Locations of Certain Remote Shutdown Components Not Specified Certain operations must be performed locally when utilizing the Remote Shutdown Panel. The specific locations of the various valves and components is not indicated in the procedure The licensee has agreed to revise the procedures. Pending issuance of the revised procedures, this remains as an

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unresolved item (50-322/84-46-01).

7.3 Protection for Associated Circuits Appendix R,Section III.G, requires that protection be provided for associated circuits that could prevent operation or cause maloperation of redundant trains of systems necessary for safe shutdow s The circuits of concern are generally associated.with safe shutdown circuits in one of three ways:

Common bus concern

• Spurious signals concern

Common enclosure concern The associated circuits were evaluated by the team 'for common bus, spurious signal, and common enclosure concern Power, control, and instrumentation circuits were examined for potential problems. A sampling basis was used in making the examination,'since many circuits were involved and a determination of cable routing took considerable tim . Connon Bus Concern q,

The common bus concern may be found in circuits,; either safety related or non-safety related, where there is~a common power source with shutdown equipment and the power source is not electrically protected from the circuit of concer The team examined, on a sampling basis, 4160V, 480V and 125V DC bus protective relay coordination. The team also examinaJ, on a sampling basis, the protection for specific instrumentation, controls, and power circuits, including the coordination of fuses and circuit breakers. The licensee plans to perform relay setting at aoproximately 18 month intervals. No unacceptable conditions were identifie ,

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7.3.2 Spurious Signals Concern .

The spurious signal concern is made up of 2 items:

False motor, control, and instrument indications can occur such as those encountered during 1975 Browns Ferry fir These could be caused by fire initiated grounds, short or open circuit *

Spurious operation of safety related or non-safety related components can occur that would adversely affect shutdown capability (e.g., RHR/RCS isolation valves).

The team examined, on a sampling ba. sis, the following areas to ascertain that no spurious signal concern exists:

Current transformer secondaries

High/ low pressure interface a

General fire instigated spurious signals No unacceptable conditions were identified except as follows: Lack of Comprehensive Analysis for High/ Low Pressure Interface The licensee did not provide a complete analysis, as specified by NRC Generic Letter 81-12 (see also Paragraph 7.4.m of this report), for the high/ low pressure interface concern. A partial analysis of this type in the licensee's CSAR addressed only two such valves providing a high/ low pressure interfac These were Valve IE11 *MOV047 (Inboard RHR Isolation Valve) and Valve IEll *MOV048 (Outboard RHR Isolation Valve), as being spuriously actuated due to a fire in the secondary containment. These two valves pro-vide redundant isolation for the high/ low pressure interface between the Reactor Coolant and RHR system Since spurious actuation of only one valve at a time is to be considered for the analysis, the team did not identify. any concern in this case. However, there could be other valves which pro-vide similar high/ low pressure interfaces. The licensee did not perform a comprehensive analysis for all possibilitie This is an unresolved ittm, pending completion of license analysis and its review by NRC (50-322/84-46-03). Lack of Comprehensive Analysis for General Fire Instigated Spurious Signals The general fire instigated spurious signals concern, as specified by NRC Generic Letter 81-12, was not completely analyzed by the licensee (see also Paragraph 7.4.m of this report) to ascertain any effects for a fire which occurs

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anywhere in the plan Particularly, no analysis was made to determine spurious actions due to a fire in the control room / relay roo During the inspection, it was determined that the diesel '

generator feeder breaker d.c. control power could be damaged )

as the result of a fire in the control room and therefore '

'the feeder breaker may not close automatically. The licen-1ee subsequently elected to close the feeder breaker manually in case of such, eventuality. It is recessary to operate this feeder breaker to energize the service water pumps which are necessary for cooling--the diesels and the suppression pool (through RHR heat exchangers). The time limit for providing cooling for the diesels and the sup-pression pool determines how soon the diesel generator feeder breaker must be manually closed. The licensee did not have an analysis showing an estimate of this time limi In the absence of this analysis, the team could not assess whether the manual closing of the feeder breaker could be accomplished in a timely manner to support the cooling requirement ,

In summary, the licensee did not provide an analysis for L ' the above time limit nor did they provide a comprehensive l

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analysis for general fire sinstigated spurious signal This is an unresolved item,' pending completion of licensee analysis and-its review by NRC (50-322/84-46-04).

7.3.3 Common Enclosure Concer ..

The commo.n enclosure concern may be found when redundant circuits are' routed together in a raceway or enclosure and they are not electrically protected or_ when fire can destroy both circuits due to inadequate fire barrier per.etration A number of circuits, selected on a sampling basis, were examined for this concern. No unacceptaale' conditions were identified.

!? 7.4- General Fire protection Features L

L The team examined the general fire grotection features in the plant

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provided to maintain one train of safe shutdown equipment free of

. fire damage. Included in the scope of this effort were fire are '

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' boundaries, including walls, floors and ceilings, and fire protection of openings such as fire doors, fire dampers, penetration seals, fire,

~ protection. systems, and other fire protection features. No unaccept-able conditions were identified except as follows:

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k f Spacing of Fire Detectors in Reactor Building The team observed that the derign of the fire detection systems in the Reactor Building does not conform to NFPA 720/E. Specifi-E cally, the number of detectors per square foot of floor space has E not been met; the maximum distance between individual detectors

- is exceeded (120 feet instead of 30 feet); and the location of detectors below ceilings is nonstandard. This represents a L deviation from the applicant's commitment in Paragraph E.1.a of

Revision 1 (June 1982) to the FHAR, enclosure to licensee letter

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to NRR dated August 6,1982, to design the fire detection system

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- in accordance with the above-referenced standards. The net effect of this deficiency is that a fire might not be detected promptly - if at all - by the installed detector This is a deviation from the licensee commitment (50-322/84-46-05).

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[ Routing of Sprinkler System Control Cables in RCIC and HPCI T

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Areas of Reactor Building E

The team observed that the control cables for the preaction '

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sprinkler systems on. elevation 8 feet of the Reactor Building I are routed through the area protected by the sprinkler systems

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systems to the deluge valves). This means that if a fire should i occur, the cables could be damaged preventing the sprinkler

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systems from actuating and suppressing the fire.

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The licensee has stated that the cables associated with the system actuation are routed outside of the area protected.

However, the team was unable to verify this statement because i- -

of lack of support information.

t a This is an unresolved item, pending licensee submittal of the

" support information for the above statement and its review by NRC (50-322/84-46-06).

I Fire Doors Degraded Because of Security Modifications

The team observed that a significant number of fire doors and

frames have been modified for security reasons. Representatives

from Underwriters Laboratories have inspected these doors and

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concluded that certain modifications could have a significant l adverse impact on the fire integrity of the doors. (Refer to i

" U.L. report of 9/6/84.) U.L. has made recommendations to the licensee to upgrade the doors so as to restore their fire e integrity. The licensee has begun to implement some of U.L.'s L recommendations. The implementation program is not complete as

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of this date. Furthermore, some of the U.L. recommendations cannot be implemented because of their practical difficulty.

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es The licensee has not identified any acceptable mitigation for those U.L.. recommendations that will not be implemented. This represents a deviation from a licensee commitment in Paragraph-

D.I.J of Revision 1 to the FHAR to provide fire doors having a resistance rating at least equal to the required rating of the

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barrier (50-322/84-46-07). Diesel Fire Pump Cables in Electric Fire Pump Room

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The team observed that the cables from the diesel fire pump controller and day tank pumps are routed through-the same fire

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area as the electric fire pump. This represents a deviation from the licensee's commitment in Paragraph E.2.c of Revision 1 to the Rt4R to separate the pumps and their associated components by a 3-hour fire wal The effect of this co,dition is that a

single event could cause the loss of all water for fire protectio .

$ This is a deviation from the licensee commitment

(50-322/84-46-08).

No Fire Damper in Duct between HVAC and Chiller Rooms l

3- The team-observed an HVACiopening in the fire wall separating the HVAC equipment room from'the chiller room on elevation 44

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feet. This opening;is not provided_with a fire damper. The-

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FHAR, Revision 1,. Paragraph D.1.j,~ describes the-licensee's

~ commitment to provide an ' adequately rated -fire damper where a ventilation duct penetrates a fire wal .This is a deviation-from the licensee commitment (50-322/84-46-09).

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i Design Concentration of CO, in Battery Rooms and Cable Tunnel During the inspection, the team observed that in certain fire areas'that are protected by an automatic carbon dioxide fire L suppression system, the floor-to-ceiling distance is-large. The

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team was concerned'that the systems would not be able to maintain the design concentration of extinguishing agent for the required-

" soak" time at locations high in the~ ceiling. The licensee provided theiteam with the.results.of.the acceptance tests of

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all carbon dioxide and the.halon fire extinguishing system .

With the excepti.on of Battery Rooms'Afand B and the Cable Tunnel, design concentrations .were. maintained 'throughout the areas fo the' required duration. In the above-referenced rooms,Jthe design density was not maintained'at the highest test point. .This

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represents a deviation;from the licensee's commitment,-to~ design l these systems in accordance with NFPA 12, as contained in Para- l graph E.5 of Revision'1 to the-FHAR. This is-a' deviation.from- '

the licensee commitment (50-322/84-46-10).

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The licensee subsequently has provided justification for this deviation. The team is referring this issue to NRR for

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resolution.

Fire Detectors in Computer Room Located above Suspended Ceiling The team observed that the fire detectors which initiate CO i

discharge in the computer room are located above the suspended

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ceilin The location of these detectors could. prevent timely successful actuation of the system if a fire occurred. This condition represents a deviation from the licensee's commitment,.

to design the carbon dioxide systems in accordance with NFPA 12, as found in Paragraph E.5 of Revision 1 to the FHAR (50-322/84-46-11).

4  ; Damaged Structural Steel Fireproofing

The' team observed that the fireproofing protection ("pyrocrate")

,. for structural steel which forms a part of the ceiling / floor

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assemblies in the charcoal filter room and chiller room on

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elevation 63 feet was damaged. This is a deviation from the

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licensee's commitment, in Paragraph D.1.J of Revision 1 to the FHAR, oto provide 3-hour protection for the ceiling / floor assen-blies'in these areas (50-322/84-46-12). Lack of Fire Hazard Analysis for Control Building Corridors and-Electric Manhole 1 A fire hazard analysis was.not conducted for corridors 'in the

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s control building and for electric manhole number.l. This repre-sents a. departure from Section D.1.b of Appendix'A.to'BTP APCSB 9.5-1. Safety-related. systems are located in these areas. How . ,

ever,.the cceridors contain no active fire l protection. -This

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, represents a departure from Section III.F~of Appendix ,

, This item is unresolved, pending satisfactory resolution of the

issue by the licensee with~NRR (50-322/84-46-13).

I Single Header Feeds Both-Sprinkler and Standpipe Systems at Reactor Building, Elevation 40 Feet The team observed that, in'the Reactor Building on elevation 405 feet,.a single water supply pipe feeds both the sprinkler and

'. standpipe systems. A single failure at.this point could resul 'in the loss of both primary-and secondary fire protection for the entire Reactor Building. This represents a departure from '

-Section_A.4 of, Appendix A to BTP:APCSB 9.5-1 and conflicts.with Revision 1 of the FHA '

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The plant Technical Specification imposes a requirement to

, ' establish a backup water supply within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if such an event occurs - but no backup capability has been identifie This item is unresolved, pending satisfactory resolution of the issue by the licensee with NRR (50-322/84-46-14). Structural Integrity of Cable Tray Penetration Seah The penetration seals were not tested in a configuration repre-sentative of what is found in the plant. Specifically, cable trays, with unprotected supports, penetrate the fire-rated seal If the supports fail under an actual fire, a dynamic load will be produced which could cause the seal to fail. This does not conform to the guidelines of Section D.3.d of Appendix A to BTP APCSB 9.5- This item is unresolved, pending a satisfactory resolution of the issue by the licensee with NRR (50-322/84-46-15). Sizing of Fire Water Storage Capacity It appears that the water storage capacity for fire fighting was not sized on the basis of the largest sprinkler demand plus an allowance of 1000 gallons per minute for manual hose stream This does not conform to the guidance of Section C.2.e of Appendix A to BTP APCSB 9.5-1 and conflicts with the FHA This item is unresolved, pending satisfactory resolution of the issue by the licensee with NRR (50-322/84-46-16). Licensee Response to Generic Letter 81-12 The licensee has not provided specific responses to the guidance in NRC Generic Letter 81-12 and its clarification. The team was

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not able to ascertain whether the licensee had formally been requested by the staff to provide response This item is unresolved, pending resolution of the issue between the licensee and NRR (50-322/84-46-17).

In addition to the above items, the team had concerns regarding following general fire protection features. These concerns were satisfactorily resolve'd after discussion with the licensee as stated below: Installation of Penetration Seals The team was concerned that fire rated penetration-seals were not installed in accordance with NRC guidelines. Specifically, the team was co'ncerned that the seals were not in sufficient

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thickness to achieve the required fire rating. The team was concerned that internal conduit seals were not installed at the required locations; and that all seals were not yet installed.

k However, the licensee provided test results which confirmed that

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the thickness of the sealant material observed was enough to achieve the required' fire rating. The licensee provided speci-fications and drawings which confirmed that the. location of

seals within conduit complies with the guidelines of Section of Appendix A to BTP APCSB 9.5-1. The licensee also provided the team with documentation confirming that the insta11atio'n of fire seals at Shoreham is complete. These responses were

sufficient to address the concern This item is close Cable Trays in Common Fire Wall of Redundant Switchgear Rooins

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The team observed that cables in trays which originated in one'

switchgear room passed through the common fire wall between the-other switchgear rooms. The team was concerned that cables from more than one shutdown division may1be located in a single fire

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area. However, the licensee provided us the drawings which showed that the observed cables were not shutdown relate ~

This item is close Fire Detectors for Relay Room Cabinets The team observed that certain cabinets in the relay room were not proviced with fire detection per NRC guidelines. However, .,

the licensee demonstrated that these. cabinets-(IH21-RK-24- 1 1H21-RK-23; 1H21-RK-21) were not safety related.The absence.of:

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fire detectors is therefore acceptabl This item is close I Electrical Supervision of Fire Protection Water Supply Valves  !

l The team observed that certain valves which control the water-supply for fire protection were not. electrically supervised or 4 locked open as specified by the licensee in Revision 1 to the FHA The valves _were sealed-.cpen instead, whichiis an accept-able alternativ Moreover, the applicant committed to inspect the sealed valves ~

weekly - effective immedtately. This commitment conforms'to the'

guidelines of Section C.3.b of' Appendix A to BTP APCSB 9.5-1 and H

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. to NFPA Standards 13 and 26 and"is'therefore acceptabl d

This item is close ,

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16 Fire Protection for Contaminated Equipment Storage Room The team observed an area on elevation 150 feet of the Reactor Building identified as the Contaminated Equipment Storage Roo This area has'no active.fira protection. The team was concerned that combustible materfalfwould be stored in this unprotected are However, the licensee stated that this area will only be used for storage of noncombustible materials, the area will be posted accordingly, and that plant fire protection inspections.wil assure compliance. The team found this response acceptabl This item is close Electric Fire Pump Controller The team observed that the controller for the electric fire pump is not U.L. liste However, the licensee provided documentation which indicates that the controller conforms to the guidelines of NFPA 20 for electric fire pump controllers which is an acceptable alternative. Based on this response, the team considers.this item close , Alternate Shutdown Capability for Relay Rooms-Based on a review of previous licensee correspondence, the team was concerned that an alternate shutdown capability may not be available for the relay. room if complete loss of function of all shutdown systems in the relay room was assume 'However,- the team subsequently confirmed that an alternate shut-down capability is available which is electrically and physically independent of the relay roo This item is close .0 : Emergency Lighting s

.Eight-hour battery pack emergency lights are required for areas of the plant necessary for safe shutdown. The licensee is committed to install self-contained eight-hour battery pack emergency lighting in all areas of the plant which could be manned to bring the plant to a safe cold shutdown

'and in access and egress routes to and"from all fire areas (see Supple-ment 1 to the SER, Section 9.5.4).

The team examined the plant er ergency lighting system to ascertain the licensee's compliance with the above commitment. The team did not '

identify any unacceptable conditions except as follows:

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17 Nameplate Ratings of Emergency Lighting Battery Packs There are two types of battery packs for emergency lighting evident in the plant: Exide F-100 6VDC packs with 2 bulbs and B-200 12VDC packs with 3 bulb The nameplate rating is 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> based on 4-bulb operatio ~

The licensee performed a test on one unit of each type which indicated

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that voltage decay was very minimal after 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of testing (i.e.,

less than 0.5V for the 6V packs and slightly more than 1.0V for the 12V packs).

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However, since it could not be ascertained that none of the battery packs have 4 bulbs, the licensee will review all of the emergency battery pack lighting to certify that it is capable of 8-hour operatio This remains as an unresolved item, pending the above licensee action and its review by NRC (50-322/84-46-18). No Emergency Lighting Available in Specific Locations In the course of walking through the procedure for Shutdown From Outside the Control Room, SP 29.022.01, several areas were discovered

.which require local operator actions from outside of the Remote Shut-down Panel and do not contain any 8-hour emergency battery pack ~

lightin The specific locations, all in the Reactor Building, are: Elevation 150' North Side - area of the Reactor Building Closed

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Loop Cooling Water (RSCLCW) Pumps and also the Fuel Pool Cooling Pump are . Elevation 78' at the RPV local instrumentation panel . Elevation 63' at RHR Valve Room for the vent valves 01V-3124 and 01V-3.25 associated with RHR valves MOV-47 and MOV-4 . Elevation 40' NW corner near the elevator for the condensate transfer loop fill valve (04V-0016).

In addition, the installed battery pack in the 101 Diesel-Generator Room cannot be aimed at the control panel because of its locatio This is a deviation from the l'censee commitment, to install self-contained 8-hour battery pack etiergency lighting in all areas of the plant which could be manned to bring the plant to a safe cold shutdown condition, as docunented in Supplement I to the SER, Section 9. (50-322/84-46-19).

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The licensee has agreed to review all of the emergency procedures to determine if there are any other areas which may require emergency lighting not now provided. The licensee is considering, as a possible resolution, procedural changes, as opposed to provision of lightin .0 011 Collection System for Reactor Coolant Pump An oil collection system for reactor coolant pumps is required if the containment is not inerted during normal operation. As Shoreham contain-ment is inerted during normal operation, the above requirement does not apply to this plant. Therefore, no inspection was performed in this are .0 Quality Assurance During the course of the inspection, the team reviewed several drawings, fire hazard analysis, fire protection modification packages, procedures, and other fire protection documents. The scope of review included veri-fication of their technical adequacy, appropriate reviews, design and procurement controls, and other Quality Assurance requirements for the licensee's fire protection program. Except as noted in the previous sections of this report, the team did not identify any other unacceptable condition .0 Unresolved Items Unresolved items are matters for which more information is required in order to ascertain whether they are acceptable, violations, or deviation Unresolved items are discussed in Sections 7.1, 7.2.2, 7.3.2, 7.4, and i 2.0 Conclusions The significant findings of this inspection are summarized in Attachment 3 and are categorized as follows:

, Deviations from specific licensee commitment . - Departures from.BTp APCSB 9.5-1, Appendix . Departures from 10 CFR 50, Appendix With respect to the findings under Category 1, the licensee immediately instituted acceptable compensatory measures for lack of adequate fire protection features until permanent corrective actions are in plac These measures were documented in licensee letter to NRC, SNRC-1122, Mr. John D. Leonard, Jr., Vice President, Nuclear Operations, LILCO, to Dr. Thomas R. Murley, Regional Administrator, NRC Region I, dated December 7, 1984 and confirmed by NRC Confirmatory Action Letter CAL 84-25, Dr. Murley to Mr. Leonard, Jr. , dated December 7,198 l

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With respect to the findings under all the categories, a meeting between the licensee and NRC is planned for the week of January 14, 1985, to

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discuss resolution of the issue .0 Exit Interview

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The inspection. team met with the licensee representatives, denoted in Paragraph 1, at the conclusion of the inspection on December 7. 1984. The team leader summarized the scope and findings of the inspection at that tim The team leader also confirmed with the licensee that the documents reviewed by the team did not contain any proprietary information. The licensee agreed-that the inspection report may be placed in the Public Document Room without prior licensee review for proprietary information (10 CFR 2.790).

At no time during this inspection was written material provided to the licensee by the tea ,

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ATTACHMENT 1-LIST OF CORRESPONDENCE January 31, 1978 -

Letter from Licensee to NRC - Information Relating to FHA June 27, 1978 -

Letter from NRC to Licensee - Additional Questions for-the Licensee

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-August 7, 1978 -

Letter from NRC to. Licensee - Additional Requests for Information Pertaining to Electrical Issues-December 11, 1978 -

Letter from Licensee to NRC - Responses to Questions September 24, 1980 -

Lettar from Licensee - Appendix R Comparison April 1981 -

Shoreham Safety Evaluation Report (SER)

May 21, 1981 -

Letter from Licensee to NRC - Appendix R Comparison July 10, 1981 -

Letter from Licensee to NRC - Understanding ~of Resolution Reached with NRC Staff September 25, 1981 -

Letter from Licenseeoto NRC - Regarding Training September 1981 -

Shoreham Safety Evaluation Report, Supplement No. 1 October 13, 1981 -

Letter from Licensee to NRC - Responses to SER Concerns February 1982 -

Shoreham Safety Evaluation Report, Supplement No. 2 August 6, 1982 - - Letter from Licensee to 'NRC - Enclosing FHAR, Revision 1 February 1983 -

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Shoreham Safety Evaluation Report, Supplement No. 3 April 13, 1983~ -

Cable Separation Analysis June-21, 1983 -

Letter from Licensee to NRC, Regarding SER Confirmatory Item 62 on Remote Shutdown Panel September 1983 -

Shoreham Safety Evaluation Report, Supplement No. 4 i

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PROTECTION SEQUENCE FOR SHUTDOWN i SHOREHAM NUCLEAR POWER STATION-UNIT 1 l-CA8LE !EPARATION ANALYSIS REPORT l i

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ATTACHMENT 3 SUMMARY OF SIGNIFICANT FINDINGS FROM SHOREHAM FIRE PROTECTION INSPECTION 84-46 Deviation / For Details Item N Summary of Finding Departure From Refer to Par . Deviations from Specific l.icensee Commitments:

84-46-05 Spacing of Fire Detectors in Reactor Building

~ Spacing of fire detectors not per NFPA FHAR, Rev. 1, 7. D/E as committed in FHAR Para. E. Routing of Sprinkler System Control Cables in RCIC and HPCI Areas of Building i

Reactor Building, RCIC and HPCI Area: FHAR, Rev. 1, 7. ..

, sprinkler system control cables routed Para. E. through the areas they are designed to protect. The system will net actuate for a fire in the area (Note: Licensee subsequently stated that the cables are outside the area The item is unresolved, pending confirmation.)

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Attachment 3 2 Deviation / For Details Item N Summary of Finding Departure From Refer to Par Fire Doers Degraded Because of Security Modifications Degraded fire doors because of security FHAR, Rev.1, 7. modifications. Fire rating not per Para. D. commitmen Diesel Fire Pump Cables in Electric Fire-Pump Room Cables for the diesel fire pump are FHAR, Rev.1, 7. routed through the electric fire pump Para. E. room. This voids the 3-hour separa-tion between the pumps, committed in FRA No Fire Dampers in Duct Between HVAC and Chiller Rooms

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HVAC Room-Chiller Room: No fire dampers FHAR, Rev. 1,- -7. provided in the HVAC opening as stated Para. D. in FHA Design Concentration of CO, in Battery Rooms and Cable Tunnel Battery Rooms A & 8 and Cable Tunnal: FHAR, Rev.1, 7. Acceptance Test results indicate that Para. design concentration (per NFPA 12) fer C0: is not achievable.

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Attachment 3 3 Deviation / For Details Item N Summary of Finding Departure From Refer to Par Fire Detectors in Computer Room Located above Suspended Ceiling Computer Room: Fire detectors which FHAR, Rev. 1, 7. initiate CO 2are located above false Para. ceiling and as such, could prevent successful initiation (not per NFPA-12).

84-46-12 Damaged Structural Steel Fireproofing Structural steel fireproofing FHAR, Rev. 1, 7. ("pyrocrete") damaged in charcoal filter Para. D. room and chiller, room. Does not conform to provide 3-hour protection for structural steel.

84-46-19 No Emergency Lighting Available i Specific locations Lack of emergency lighting in certain SSER 1, - areas required for safe shutdown for Section 9. control room fir . Departures from BTP APCSB 9.5-1, Appendix A:

84-46-13 Lack of Fire Hazard Analysis for Control Butiding Corridors and Electric Manhole 1 A fire hazard analysis was not con- BTP APCSB 9.5-1, 7. ducted for Control Building corridors Appendix A, and Electric Manhole 1 Sec. D.1.b'

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~ Attachment 3 4 Deviation / For Details Item N Summary of Finding Departure From Refer to Par Single Header Feeds Both Sprinkler and Standpipe Systems at Reactor Butiding, Elevation 40 Feet Reactor Building, Elevation 40 feet: BTP APCSB 9.5-1, 7. a single header feeds both sprinkler Appendix A, and standpipe systems. A single pipe Sec. break will disable both primary and secondary fire protection (not per the BTP).

84-46-15 Structural Integrity of Cable Tray Penetration Seals Cable tray penetration seal may be BTP APCSB 9.5-1, 7. damaged by imposition of dynamic load Appendix A, imposed on the seal by failure of Sec. D. " unprotected" cable tray support in'

case of a fire. Does not conform to the BT Sizing of Fire Water Storage Capacity

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Fire water storage capacity may not be BTP APCSB 9.5-1, 7. sized on the basis of largest sprinkler Appendix A, demand plus an allowance of 1000 gpm Sec. C. for manual hose systems. This is not

.per the BTP or the FHAR, pages 1-20, para. E.

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.t,o Attachment 3 5 Deviation / For Details Ites N Summary of Finding Departure From Refer to Par . Departures from 10 CFR 50,

~ Appendix R:

.84-46-17 Licensee Resoonse to Generic Letter 81-12 The licensee has not provided specific Appendix R, 7. response to the guidance in Generic Sec. III.G.3 &

Letter 81-12 and its clarificatio Sec. II The team is not able to ascertain if (via GL 81-12)

the. staff had formally requested such-respons Lack of Comprehensive Analysis for High/ Low Pressure Interface The licensee did not have a Appendix'R, 7.3. comprehensive' analysi Sec. III.G.3 &

Sec. II (via GL 81-12)

84-46-04 Lack of Comprehensive Analysis for General Fire Instigated Spurious Signals The licensee did not have a Appendix R, 7.3. comprehensive analysi Sec. III.G.3 &.

Sec. II (via GL 81-12)-

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,o Attachment 3 6 Deviation / For Details Item N Summary of Finding Departure From Refer to Par Backup Information for CSAR The backup information to evaluate Appendix R, 7. the licensee Cable Separation Analysis Sec. III. Report (CSAR) was not available at the site for the team's revie Locations of Certain Remote Shutdown Components Not Spec 1*ied The locations are not specified in the Appendix R, ).2. , Safe Shutdown procedure (SP 29.010.01). Sec. III.G.3 &

If specified, it will facilitate an Sec._II orderly and timely shutdow ' Nameplate Rating of Emergency Lighting Battery Packs The licensee to confirm that the Appendix R,. battery packs are of_8-hour capacit Sec. II P l

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