ML20206U703

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Rev 1 to Ran 97-031, Main CR Fire Analysis for IPEEE Section 4-1
ML20206U703
Person / Time
Site: Calvert Cliffs  
Issue date: 05/18/1999
From:
BALTIMORE GAS & ELECTRIC CO.
To:
Shared Package
ML20206U708 List:
References
RAN-97-031, RAN-97-031-R01, RAN-97-31, RAN-97-31-R1, NUDOCS 9905250299
Download: ML20206U703 (28)
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ATTACHMENT f1)

REVISED MAIN CONTROL ROOM FIRE ANALYSIS IPEEE SECTION 4-1 REVISION 1 9905250299 990518 PDR ADOCK 05000317 P

PDR Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant May 18,1999

i Calvttt Cliffs Nuclear Power Plmt Intern 11 Fire Analysis individual Plant Examination Externd Ennts A405,

Main Control Room (CR)

Location:

45' Auxiliary Building Fire Area:

24 CDF:

2.45E-5 The Control Room houses benchboard control boards and miscellaneous vertical control boards for both Unit I and Unit 2. The room contains numerous vertical floor mounted electrical cabinets aligned in rows. The majority of the cabinets are open steel cabinets forming a horseshoe along the West wall.

Some of the panels are enclosed and are at a height to accommodate handswitches and other operator controls. Office equipment, including CRTs, copiers, and fax machine do not have sufficiently large motors to offer a significant ignition source.

The Control Room is an enclosed rectangular structure, approximately fifty-five feet by ninety feet by 3

twenty-two feet for a total volume of over 100,000 ft, having a three-hour minimum fire rating. Fire detection consists ofionization detectors strategically located directly above the main control board. One foot beneath the ceiling are seismically qualified metal egg crate panels that serve to protect the operators from falling glass and debris from broken light fixtures.

Fire Analysis Results The Control Room panels are grouped and represented by thirty-three fire initiating events shown in table 4-I-l below.

Table 4-1-1 Control Room PanelIgnition impacts knitiator Fmquency

{ Panel.

4 Desonption

~ Functionalimpact' iCDFs N

!(INIT)';

A405F1 1.54E-4 1C01 Main Generator and Switchyard Control Board OP 9.22E-7 1CO2 Turbine Control Board 1C29 CB Gen Metering & Misc 1CO2 Main Generator and Switchyard Control Board 1C01 Turbine Control Board 1C23 Plant Oscillograph 1C29 CB Gen Metering & Misc A405F2 1.54E-4 1C15A Reactor Protective System Channels A QZ*, IA*,

5.01 E-7 IB*, SA*, SB*

1C15B Reactor Protective System Channels B Conservatively, RS* and 1C15C Reactor Protective System Channels C PV*

1C15D Reactor Protective System Channels D BGE 4-1-1 RAN 97-031 Rev.1

Cdvert Cliffs Nuclear Power Plant intemalFire Analysis Individual Plant Examin: tion Extemal Events Table 4-I-1 Control Room Panel Ignition impacts (Cont'd)

. Initiator Frequency Panel Desenption -

Functionallmpact

'CDF

- (INIT).

A405F3 2.57E-5 1C35 Feedwater Regulation Control System 11 MP, MN, LF 9.51 E-8 1C36 Feedwater Regulation Control System 12 A405F4 2.0$E-4 1C24A DC Power Control Board HH 7.64E-7 1C24B Fire Protection Control Board 1C26 Vibration Monitor System Control Board 1C24B Fire Protection Control Board 1C39 Miscellaneous Station Recorder Panel 2C39 Miscellaneous Station Recorder Panet A405F5 7.70E-5 1C28 Technical Support Center Isolation Panel QZ*, RR, IA*,lB*

2.85E-7 1C31 Reactor Regulation Channel X Control Board 1C32 Reactor Regulation Channel Y Control Board A405F6 1.28E-4 1C06fr Reactor Coolant Control Board QZ*,00, KX, RS (DSS 6.08E-7 1C25A-D RPS Power Supply Cabinet (A-D) fails), RR, R1, PS, PV*,

1C39 Miscellaneous Station Recorder Panet AQ, SL*, CV, OT, SA, SB, HH, IA, IB, QZ 1C06bk Reactor Coolant Control Board 1C25A-D RPS Power Supply Cabinet (A-D) 1C06 Reactor Coolant Control Board A405F7 5.13E 5 1C07 Chemical and Volume Control Board HH, TB, AQ, SL*, CV, OT 1.97E-7 1C39 Miscellaneous Station Recorder Panet A405F8 1.03E 4 1C08fr/bk Engineering Safeguards Control Board AA*, AB*, QQ, Kl OT, 3.85E-7 1C09 Engineering Safeguards Control Board MV, HA, HB, HW, EA, l

CS, CT, SR A405F9 2.05E 4 1C09fr/bk Engineering Safeguards Control Board AA*, AB*, HZ,00, KI, 7.72E-7 1C08 Engineering Safeguards Control Board KL, OT, MV, VM, HA, 1C10 Engineering Safeguards Control Board HB, VS, HW, EA, EB, CS, CT, SG*, SH*, SR 1C10fr/bk Engineering Safeguards Control Board 1C09 Engineering Safeguards Control Board BGE 4-1-2 RAN 97-031 Rev.1

Calvert Cliffs Wuclear Power Pir_nt internalFire Analysis Individual Plant Examinnion Externet Events Table 4-I-1 Control Room Panel Ignition impacts (Cont'd)

Initiator Frequency Panel'

. Description Functional impact '

CDF E

A405FA 5.13E-5 1C13 SW, SRW, and CCW Control Board GG, HS*, NR*,11,12, S1, 6 52E-7 1C15X Reactor Protection System isolation Cabinet S2,KX,KY,KZ,VC,FC, 1C24B Fire Protection Control Board Fo, WY, HH 1C26 Vibration Monitor System Control Board A405FB 5.13E-5 1C24B Fire Protection Control Board GG, HS*, HH, QZ, NR*,

6.72E-7 1C13 SW, SRW, and CCW Control Board 11,12, S1, S2, KX, KY, 1C15C Reactor Protective System Channel C Reactor KZ, RS*, VC, SP, SL*,

1C15D Protective System Channel D lA*, IB*, FO, SA*, SB*,

1C24A DC Power Control Board WY 1C26 Vibration Monitor System Control Board A405FC 5.13E-5 2C07 Chemical and Volume Control Board U-2 MFW LOST 1.88E-7 2C39 Miscellaneous Station Recorder Panet HH A405FD 7.70E-5 2C06bk Reactor Coolant Control Board F9, HH 2.84E-7 2C06fr Reactor Coolant Control Board 2C25A-D RPS Power Supply Cabinets (A-D) 2C39 Miscellaneous Station Recorder Panel A405FE 2.05E-4 2C08fr/bk Engineering Safeguards Control Board U-2 MFW LOST 7.55E-7 2C09 Engineering Safeguards Control Board AD*, AC*

2C09fr/bk Engineenng Safeguards Control Board 2C08 Engineering Safeguards Control Board i

2C10 Engineering Safeguards Control Board f

A405FM 1.54E-4 2CO3fr Condensate and Feedwater Control Board U-2 MFW LOST 5.03E-7 2C35 Feedwater Regulation Control System 22 2C36 Feedwater Regulation Control System 22 2CO3bk Condensate and Feedwater Control Board 2C35 Feedwater Regulation Control System 21 I

2C36 Feedwater Regulation Control System 22 l

4' 2C38 Unit 2 Feed Pump Turbine Control Board BGE 4-I-3 RAN 97-031 Rev,1

Calvert Clifts Nucletr Power Plant Internal Fire Analysis Individual Plant Examin: tion External Events Table 4-1-1 Control Room Panel Ignition impacts (Cont'd)

Initiator =

Frequency

' Panel Description Functionalimpact CDF (INIT) -

A405FN 1.42E-3 1C15X Reactor Protection System isolation Cabinet NONE 4.55E-6 j

1C22A-G Fz i 54onitoring and Meteorological Boards 1C23 Plant oscillograph 1C28A TSCC Computer Multiplexer

=

1C30A Unit 1 Loose Parts Monitoring System 1C308 Unit 2 Loose Parts Monitoring System 1C33 Waste Processing System Panel 1C38 SGFP 11 & 12 Signal Processor Cabinet 1C60-1 Plant Computer Cabinet 1C60-2 Plant Computer Cabinet 2C02 Turbine Control Board 2C15A-D Reactor Protection System Channels (A-D) 2C15X Reactor Protection System isolation Cabinet 2C23 Vibration Monitor Control Board 2C25A-D RPS Channel A/B/C/D Power Supply Cabinet 2C26 Vibration Monitor Panel Control Board 2C28 Technical Support Center isolation Panel 2C28A TSCC Computer Multiplexer 2C31 Reactor Regulation System Channel X 2C32 Reactor Regulation System Channel Y 2C34 HVAC System Control Panel 2C60-1 Plant Computer Cabinet 2C60-2 Plant Computer Cabinet BGE 4-1 4 RAN 97-031 Rev.1

Calvert Clifts Nuclear Power Plant Internal Fire Analysis Individual Plant Exnninnion Extemal Events Table 4-1-1 Control Room Panel Ignition impacts (cont'd) initiator Frequency Panel-Description Functionalimpacti CDF (INIT) '

Fl1C03 7.70E-5 1C03fr Condensate ard Feedwater Control Board QQ, T1, MC, RI, BV, DW, 1.15E-6 1C35 Feedwater Regulation Control System 11 DV, BS, MN, FT, MS, 1C36 Feedwater Regulation Control System 12 FN*, HX*, UQ*, HU, TG, F9, MH, F3*, LF, OT, MP 1C03bk Condensate and Feedwater Control Board F11C04 1.03E-4 1C04 Aux Feedwater and Computer Control Board QZ*, QQ, MC, RR, Rl, 2 81E-6 1C05 Reactivity Control Board PS, PV*, FT*, FN, FH, 1C31 Reactor Regulation System Channel X F7*, HX*, TF, TG, F9, 1C32 Reactor Regulation System Channel Y MT, MH*, F3*, LF, OT 1C28 Technical Support Center isolation Panel 1C05 Reactivity Control Board 1C04 Aux Feedwater and Computer Control Board F11C17 5.13E 5 1C17 4KV & 480VAC System Normal Control Board H5, AE, AF, DM*,

1.91 E-7 VC, RQ, LF F11C18 5.13E-5 1C18 13KV and 4KV Essential Control Board QC, OE, GE, H5, AA, 1.59E-6 QQ DM*, VC Fl1C19 5.13E-5 1C19 13KV and 4KV Essential Control Board GF, GG, H5, AB, AD 2.17E-6 F11C20 5.13E 5 1C20 13KV and 4KV Essential Control Board OD, OF, GH, H5, GJ, AC, 3.86E-7 QQ,DM*

F11C34 5.13E 5 1C34 HVAC System Control Board HH, QQ, SL*, LF, VM, 1.68E-7 V1,V2,V5

]

Fl2005 1.03E-4 2C04 Aux Feedwater and Computer Control Board F9 3.35E-7 2C05 Reactivity Control Board 2C28 Technical Support Center isolation Panel 2C31 Reactor Regulation System Channel X 2C32 Reactor Regulation System Channel Y 2C05 Reactivity Controi Boaro 2C04 Aux Feedwater and Computer Control Board Fl2C09 1.03E-4 2C10fr/bk Engineering Safeguards Control Board AD*, AC*

3.84E-7 2C09 Engineering Safeguards Control Board F12C13 5.13E-5 2C13 SRW/ Misc Services Control Panel GF, GH, NS, F9 1.91 E-7 2C15X Reactor Protection System Isolation Cabinet 2C24B Radiation Monitoring Panel 2C26 Vibration Monitor System Control Board BGE 4-1-5 RAN 97-031 Rev.1

Cdvert Cliffs Nuclear Power Plant intemalFire Analysis Individud Plant Exarnination External Events Table 4-I-1 Control Room Panel Ignition impacts (cont'd)

Initiator -

Frequency Panel-Description '

Functionalimpact CDF E

Fl2C17 5.13E-5 2C17 4kV/480VAC System Normal Control Board Unit 2 MFW lost, DM*

1.91 E-7 FIC18A 5.13E-5 1C18A EDG 1A 4KV and 480VAC Control Board GE,GJ,QQ 5.23E-7 FIC18B 5.13E-5 iC18B EDG 1B 4KV and 480VAC Control Board GG,GJ,00 4.43E-7 FIC19C 5.13E-5 1C19C EDG OC Control Board GJ,QQ 1.68E-7 FIC20A 5.13E-5 1C20A 2A DG Control Board GF,GJ,00 5.23E-7 I

FIC208 5.13E-5 1C20B 2B DG Control Poard GG, GJ, QQ 4.43E-7 FIC24A 1.03E-4 2C24A Saltwater Isolation Panel Si, S2, GW, GZ 6.24E-7 2C248 Radiation Monitoring Panel 2C15A-D Reactor Protection System Channels (A-D) 2C24B Radiation Monitoring Panel 2C24A Saltwater isolation Panel 2C26 Vibration Monitor System Control Board Approach Osenlew The Main Control Room fire risk is assessed by modeling the various consequences associated with a fire

)

in each Control Room panel. This assessment is coded explicitly into the CCFPRA using the following approach:

Evaluate individual Control Room panels according to their physical characteristics e

including wall separation, component layout and density, conductor routing, and any other construction attributes which could impact fire propagation.

Calculate the heat release rate for each panel to use with plume damage and/or radiant damage computations.

Perform a damage assessment for each panel as an initiating source.

1 Determine the ignition frequency for each panel.

BGE 4-I-6 RAN 97-031 Rev. I

Calvert Cliffs Nuclear Power Plant Internal Fire Analysis Individuil Plmt Examination External Events Determine the likelihood of Control Room evacuation. Fifteen minutes is used as the time available for manual suppression. Failure to suppress results in Control Room evacuation.

Determine the likelihood of propagation between panels (or panel groups) given the fire has not been suppressed within the first fifteen minutes. Propagation between panels is assumed to occur on failure of the Fire Brigade to suppress the fire within fifteen minutes after the Control Room has been evacuated.

Evaluate the resulting functional impact for each panel ignition; the evacuation of the Control Room with the loss of a panel (or panel group), and the evacuation of the Control Room with the loss of multiple panels. The risk-assessment associated with the evacuation of the Control Room includes the evaluation of the impact ofimplementing the Appendix R safe shutdown procedure for Control Room evacuation (AOP-9A).

Bin or combine panel fire scenarios based on like impact and adjust the overall ignition frequency accordingly.

Adjacent Rooms There are areas adjacent to the Main Control Room having fire barriers that are not credited for separation. These areas, all contained within Fire Area 24, are listed below:

hai-

---MM$E%

A400 Control Room Vestibule

)

A401 Operations Shift Office A402 Control Room Toilet A403 Janitor Storage A404 Kitchen - Control Room A405 Main Control Room A406 Unit 2 DAS Computer Room A415

' Area behind SW panel A431 Unit 1 DAS Computer Room A432 Tech Support 45' A434 Passage A436 Tech Support Center A437 Tech Support Annex A438 Shift Supervisor's Office A442 Reserve E,attery Room A443 Passage A444 Central Alarm Station BGE 4-I-7 RAN 97-031 Rev. I I

Caben Clitis Nuclear Power Plant Internal Fire Analysis Individual Plant Examination External Events

' Smoke'is not anticipated to migrate into the Control Room from fires in adjacent or outside rooms.

Smoke from a fire in either Unit's Battery Rooms or the Reserve Battery Room will not migrate into the Control Room because there are separate supply and exhaust duct systems and fans, which are isolated from the Control Room by a separate duct system with an exhaust fan. The Control Room / Cable Spreading Room liVAC system does not provide any ventilation to these rooms. All Battery Rooms are under a negative pressure and have normally closed and locked doors which would further prevent smoke migration.

Rooms A401, A402 and A404 share a common supply from the Control Room IIVAC system. They do not share the CR HVAC return but instead have a separate exhaust duct / fan to atmosphere. The supply does not have a smoke damper; however, should a fire occur the separate exhaust will prevent smoke spread into the Control Room. Supply and return ducting is common to the CR/CSR IIVAC system for the Tech Support Center Rooms (A432 and A436) and the Central Alarm Station (A444). The TSC Annex (A437) does not have common return duct. The return air for A437 is pulled into the Computer Alarm Station (CAS) via an air transfer grill in the common wall between room A437 and the CAS. All rooms have smoke detection which will activate Electro Thermal Link controlled dampers. These dampers will close and isolate the rooms from the rest of the llVAC system and prevent smoke migration into the Control Room.

An evaluation of on-site combustibles installed in the CR Vestibule under FCR 89-166 indicated that the combustible loading in this and other areas adjacent to the CR do not challenge the available fire protection of the plant. The barriers in the Vestibule, Shift Supervisor's Office, and Tech Support Center are made of fire retarding gypsum wallboard which conforms to ASTM C36-66 and has a one hour resistive rating. The Toilet area walls are constructed of ceramic tile. The walls in the remaining areas are constructed of concrete and masonry, which provide for a two-hour rated barrier.

These areas do not contain sufficient fixed ignition sources or combustibles to pose a concern for propagation into the Control Room. These areas do contain combustible materials with low flame spread ratings, such as carpeting and office furniture. liowever, there are insignificant or no sources ofignition in these rooms. It is assumed that a fire in these rooms will not propagate into any adjacent compartments.

Control Room Panel Fires ManualFireSuppression Time Cabinets and consoles are subject to damage from two distinct fire hazards:

1. Fire originating within the cabinet; and I
2. Exposure fires involving combustibles in the room area.

BGE 4-I-8 RAN 97-031 Rev.1 i

Calvert Clifrs Nuclear Power Plant Internal Fire Analpis Individual Plant Examination Extemal Events The only ignition sources within an electrical cabinet are those associated with electrical faults. Given experience with fires reported in the EPRI Fire Events Database, damage can be confined to the site of the overload, and the impact is bounded by the random failure of the component itself(whichis already accounted for in the PRA model).

The likelihood of detection and suppression is dependent upon whether or not cabinets are fitted with in-cabinet detection. Heat and smoke detectors are installed in the Control Room but not within the

{

. individual cabinets. The Sandia cabinet fire tests indicate a five-minute time lapse between an in-cabinet fire detector detecting smoke and the time that actual flames were observed. These tests used vertical and benchboard cabinets loaded with unqualified cables ignited using an electrical ignition source. No credit is taken for detection and suppression during this phase since there is no cabinet detection.

The potential for significant damage due to a cabinet fire is very small prior to flame ignition. Ignition may be prevented by de-energizing the faulted component and using manual fire extinguishers. Controls in the cabinets are generally not temperature sensitive and temperature sensitive instruments do not typically control safety-related equipment. Fires within instrumentation cabinets are assumed to result in the loss of the entire cabinet. This assumption is consistent with the guidance in the EPRI Fire PRA Implementation Guide (Implementation Guide)and with experimental test results conducted at Sandia.

All components operated by the cabinet in which the fire originates are assumed to fail (unless there is an alternate control means) if the fire is not suppressed within fifteen minutes.

l The Sandia test results indicated that cabinet fires were self-sustaining and produced sufficient quantities of smoke to cause visual impairment with purge rates as high as fourteen room changes per hour. All of the actual Control Room fires in the EPRI Fire Database were small; however, this may have been the result of early extinguishing. Since there are insufficient tools availableto assess smoke production and historical fire data is not conclusive, it is assumed that any fire is capable of producing sufficient smoke to require evacuation of the Control Room given it is allowed to continue burning for a sufficient length of time.

Eleven Sandia tests are available to provide information on smoke build up. Six tests were conducted for 3

small enclosures (11,016 ft ) with ventilation rates of about fourteen room changes per hour. Only one of these, however, was electrically initiated and indicated visual obscuring within thirteen minutes, where time zero is the point at which smoke was first observed from the cabinet. Five tests were performed in 3

larger enclosures (48,000 ft ) in which two fires were electrically initiated. In both of these cases, the main control board was obscured within 15.5 to 19.5 minutes following observation of smoke. For large enclosure fires, the ventilation system does not appear to substantially affect the rate of smoke build up.

The ventilation rate in one case was one room change per hour, and in the other case eight room changes per hour.

Based on the above discussion, it is concluded that the rate of smoke build up in the Control Room will be marginally slower than observed for the large test enclosure. It is determined that smoke obscuration of the control board will not occur for at least twenty minutes after ignition. Allowing three to five minutes for activation of the area ionization smoke detectors, fifteen to seventeen minutes would be available to extinguish a fire in a cabinet with no in cabinet detection prior to the need for Control Room evacuation. This assumes that operators would detect a fire in three to five minutes. This is believed conservative since actual detection time may be in the two to three minute range.

BGE 4-1-9 RAN 97-031 Rev.1

n.

Calvert Cliffs Nuclear Power Plant Internil Fire Andysis Individual Plant Examination External Events

. The guidance provided in NSAC-181 suggests that approximately fifteen minutes is available for a

, typical, nuclear power plant Control Room before fire induced degradation of the Control Room environment requires evacuation. This fifteen-minute time period _is consistent with the guidance in Appendix H of the Implementation Guide for fire propagation to adjacent control panels.

Appendix M of the Implementation Guide suggests that as much as fifteen to twenty minutes may be available for manual fire suppression in the Control Room depending upon a number of factors, including

. interpretation of the test data, volume of the Control Room, and volume changes per hourby the Control Room HVAC. The Implementation Guide recommends that the size of the room and the volume changes per hour be investigated to verify that the SNL test conclusions reasonably apply to the specific site. The 3

CCNPP Main Control Room is over twice the size of the SNL test facility (approximately 100,000 ft )

with a HVAC system rated at twenty room changes per hour in full outside air mode.

Based on the HCR model derivation in theImplementation Guide, the probability of non-suppression in fifteen minutes with no in-cabinet detection is 3.40x10-3, Fire Suppression There is no automatic fire suppression system installed in this room.

Fire Suppression Induced Equipment Failure Suppression of fires is expected to be accomplished manually through utilization of CO2 fire

. extinguishers.' CO is non-corrosive in the short-term and will not result in any equipment damage.

2 PanelIdentification 1

s' Panels are identified by searching the NUCLEIS database for panel type unique equipment identifiers (UEls) in the Control Room (area 405). The results yielded ninety-one discrete panels.

Panels 1(2)C08,1(2)C09, and 1(2)C10 include a full width front portion (on the horseshoe interior) and a full width rear section, across the aisle and opposite of the front section (on the horseshoe exterior). (See Figure 4-I-1.) Similarly,1(2)C03 and 1(2)C06 include partial width rear sections across the aisle.

Since the front or back panel fire modeling attributes are indistinguishable from other individual panel attributes, it is appropriate to treat a front or back panel as an individual panel. Hence, for example, IC06 appears as IC06fr and IC06bk in the analysis. Note that although treated as separate panels for fire modeling, the functional impact of losing either the front or back panel is treated as losing the entire panel.

Using this approach created 101 ignition sources which are the fire scenarios. A discussion of how this methodology effected heat release rates (HRRs) and ignition frequency follows.

-BGE.

4-1-10 RAN 97-031 Rev.1

Calvert Cliffs Nuclea Power Plant Intemal Fire Analysis Individual Plant Examination External Events Figure 4-I-1 Main Control Room t

I Computer Room North Pipe Way Cable Chase IC IC22 A-O l

IC25 A-D IC60-2

/

CM l IC2G l C6 IC60-1 P

gc39 ICl5X IC31 1C05 ICO6 1C07 1C08 IC09 IC10

\\p

/

IC28 ICIS A-D IC28A.)

3-

[

b Y

ICl?

e IC32 #

\\

1C35 /I ICO3 lC33 Y

]

/

ICIBA ICI8 IC36 #

IC23

-p ICO2 IC18B Operators gg39 I I IC29 -p ICOI IC20B IC20 2C23 _g 2CO2 N

IC20A 2C17 2C36 -t R

2C34 2C35 2CO3 2C32 r

\\

l L

% Main Entrance 2004 2C28A y 2C28 / D

\\

2C15 A-D

/

2005 /2COe 2C07 2008 2C09 2C10 2C31 2C60-2 l

2C39 2C60.g -

2C24A l 2C24B l 2C26 Kitchen & Toilet 2C25 A-D IC30A I

Pipe Way Cable Chase 2C I

Computer Room BGE 4111 RAN 97 031 Rev.1

s Calven Clini Nuclear Power Plant Internd Fire Analysa Individual Plant Examincion Externci Events Ignition Frequency The following event tree shows conceptually how main Control Room (CR) panel fires are modeled:

Suppression Severity to Abandonment Manual (6/9)

Severe (2/6)

Yes (13.4E-3)

A Analyzed 1.27E-2 4.22E-3 initiating Event No (3.4E-3)

B Analyzed Frequency = 1.90E.2 Minor (4/6)

Yes (13.4E-3)

C be included in equipment 8.44E 3 l Screened failure data since the I

Control Room is not challenged and fire is This branch is included in Branch B above since it No (3.4E-3) requires CR abandonment.

D Although the fire is minor it I

is assumed to impact the entire panet.

These fires are screened since suppression was not Self-Extinguished (3/9) required and the Control E Room was not challenged.

6.33E-3 The equipment impact is assumed to be included in the equipment failure data.

In Section D3.1, Control Room Electrical Cabinet Fires, theImplementation Guide states that two fires were self extinguished and six fires were suppressed with pottable extinguishers. No information was provided for the other two events so one is categorized as self-extinguished and the other as requiring manual suppression.

In addition, the grease fire located in a kitchen was screened since this was not an electrical cabinet fire and since CCNPP's Main Control Room has a separate ventilation retum for the kitchen which would prevent smoke from impacting the Control Room. This results in three of nine fires being self-extinguished. For fires which are self extinguished, the question of the likelihood of failing to manually suppress is not applicable as shown above in Bnmch E.

The Implementation Guide is also used to provide the basis for the fact that two of the six non-self extinguished fires are assumed to be severe. It is implied from theImplementation Guide that a severe fire is a fire which damages multiple circuits. Note that Section D3.1 concludes the fraction of severe BGE 4-1-12 RAN 97-031 Rev. I

1 Calvert Clifts Nuclear Power Plant Internal Fire Analysis Individual Plant Examination External Events fires is two of ten, but, by screening the grease fire and removing the three self-extinguished fires, this

, ratio becomes two of six. For the four remaining fire events, it is recognized that a minor fire can still has some likelihood ' f forcing Control Room abandonment if not suppressed. These minor fires which are o

not suppressed are included in Branch B. Therefore, the use of severe and minor fire characterization did -

not result in a reduction in the challenge to Control Room abandonment.

l Although the above event tree shows conceptually the approach to modeling this issue, for simplification,

)

the ' model does ; not explicitly. include the suppression (manual /self extinguished) or severity j

(minor / severe) nodes. Instead the likelihood of conditional abandonmentis increased by a factor of three (i.e.1.02x10-2 = 3

  • 3.4x10-3) and the Control Room fire initiating event frequenciesare reduced by a factor of three. This approach ensures that all fire scenarios which result in the Control Room abandonment are captured and estimates the impact of severe fires where abandonment does not occur to
two significant figures. Branches B and D above result in a total ControlRoom abandonment frequency of 4.307x 10-5, 1

Branch B =

1.27x10-2

  • 2/6

1.27x10-2. 4/6

  • 3.4x10-3 = 2.87x10-5 The same result is obtained by reducing the initiating frequency by three and increasing the suppression before abandonment by a factor of three:

1.27x10-2. 1/3

  • 3 + 3.4x10-3 = 4.307x10-5

=

Branch A above represents the severe scenarios which do not result in abandonment; Branch A =

1.27x10-2. 2/6 * (1 - 3.4x10-3) = 4.208x10-3 The Control Room ignition frequency (determined using the methodology described in Section 4) is:

CRif =

1.90x10-2 /yeat By excluding the self-extinguished fires, the compartment ignition frequency is determined to be that portion associated with manually extinguished fires. The EPRI Fire Events Database for control room

' ignitions determined that out of nine applicable events, six were manually extinguished. Therefore,6/9 of the total ignitions require manual suppression, and:

-MANSUPg =

CRg. 6/g = 1,gox10-2 / year. 6/g 1.27x10-2 / year

=

L BGE.

4-1-13 RAN 97-031 Rev.1

Cdvert Clifts Nuclear Power Plant Internal Fire Analysis Individual Plant Examination External Events

~ Of the six fires requiring manual suppression, only two were severe enough to fail multiple functions

, single, function failures are considered bounded by the random equipment failure rates used in the

(

internal events PRA). Therefore,2/ of the manually suppressed fires are severe and the frequency is:

6 SEVEREPFif =

MANSUPjf

  • 2/6 = 1.27x10-2 / year. 2/6 4.22x10-3 / year

=

The individual panel ignition frequency is based on the relative size of the panel. This approach raises the frequency of the larger panels and lowers frequency for the smaller panels. To accomplish this, panels are categorized into four types with a corresponding weight (the largest panels are assigned a weight of one). Each panel is then classified and assigned a weight based on the table below. The sum of the weights assigned to the 101 Control Room panels is 82.25. The weighting factor for a single full-length benchboard panel is therefore 1/82.25. The weighting factors are also shown in the table below:

Full-width vertical or benchboard

.0 1.22 x10~2 Half-width vertical or benchboard 0.75 9.12x10

Full-height back 0.50 6.08x1 &'

Small sub-unit 0.25 3.04 x10-'

This system of apportioning effectively raised the overall ignition frequency of panels like IC03, which have front and back sections (see Panel Identification), but allowed for an individual fire modeling assessment of each section.

For example, for panel IC03 (with full width front and partial width back sections) is:

1C03frif =

SEVEREPFif

  • 1C03frWElGHTING FACTOR = 4.22x10-3 / year
  • 1.22x10-2 =

5.13x10-6 1C03bkif =

SEVEREPFif

  • 1C03bkWEIGHTING FACTOR = 4.22x10-3 / year
  • 9.12x10-3 =

2.57x10-5 The overall functional loss of IC03 due to ignition is 7.70x10-5 / year. However, the resulting damage scenarios for IC03fr and IC03bk occur at 5.13x10-5 / year and 2.57x10-5 / year respectively.

BGE 4-1-14 RAN 97-031 Rev.1 J

l CCivert Cliffs Nuclear Power Plant Internal Fire Analysis Individual Picnt Examination External Events

. Cabinet Heat Release Rate

' Sandia tested benchboard panels (CR-4527 and CR-4527V2 tests ST4, PCT 6, and Test 23).

.Unfortunately, the limited number of tests conducted do not represent the Calvert CliffsCR panel heat 3

loads or type of cable. The Sandia tests do show that qualified cable HRRs are lower than unqualified cable HRRs, but the qualified cable HRRs range from 82 to 1235kW.

To assess the CR panel HRR more accurately, individual panel cable loading is calculated as follows:

A computer query of the NUCLEIS Cable and Raceway Schedule (CRS) identified all cable schemes to or from each CR panel. For each cable scheme the query also identified the

. cable type (for example, silicon asbestos jacket, XLPE) and the cable diameter (usually used to determine conduit or tray fill).

Using guidance from the Implementation Guide (based on Sandia tests), three seven foot sections of XPE IEEE-38312 AWG three-conductor (12-3) cabling contains approximately 20,000 BTUs. Therefore, the heat load for og seven foot length of 12-3 cable is 20,000/3 Btu.

In addition, the heat load is further adjusted for cable length where the length correction (LC)is:

LC =

Cable Length + 7 (in feet)

The average length of a cable in the CR panels is estimated to be five feet. This assumes cables terminate uniformly along the eight foot high panel (mean length is four feet) and the average horizontal run distance (from the floor penetration to panel side wall) is one foot.

Hence, LC =

5+7

. Using the CRS cable code database, the diameter for 12-3 cable is estimated to be 0.5 inch (based on XLPE cable diameters and field experience).

Assuming an individual cable's heat load is proportional to the cable's cross sectional area, each CR cable scheme heat load is adjusted by comparing its diameter to the 12-3 XPE cable diameter.

Area =

A = nr2 Where radius and diameter :

1/$

r=

2 BGE-4-I-15 RAN 97-031 Rev. I

Calytrt Cliffs Nucle:: Power Plant Internal fire Andysis Individud Plant Examination External Events thus:

A=

x(1/ t)2 = 1/ 342 2

4 Comparing each CR panel cable (ACR) and the 12-3 XPE (AXPE, where $XPE = 0.5 inch) yields a diameter adjustment (DA):

DA =

ACR/AXPE = 1/ R&CR2 /1/ R&XPE2

  • 4CR2/&XPE2 = (&CR &XPE)2 4

/

4

($CR/ 0.5 )2

=

From the Sandia benchboard tests using qualified cables, the cable burn rate (BR) is calculated using:

BR =

Peak HRR + Fuelload Where:

j % M igph i g M M y g g g ? qqqq W M ST4 117,000 82 7.00x10

PCT 6 1,530,000 215 1.40x10d Test 23 I,550.000 1235 7.97x104 A representative overall burn rate is computed by averaging the individual test burn rates.

Note, test ST4 results are not used because the cable bundles were intentionally loosened and not typical of the CR installed cables. Therefore, average burn rate is:

Brave =

(BRPCT6 + BR est23)/2 T

(1.40x10-4 + 7.97x10-4) / 2

=

4.69x10-4 sec-1

=

The overall HRR for an individual cable scheme is therefore:

HRReable = Diameter Adjustment

  • Cable Heat Load
  • Length Correction
  • Burn Rate

= (&CR/ 0.5)2 20,000/3

  • 5/7
  • 4.69x10-4 = Btu /s As an additional conservatism,65 Btu /s is used as a minimum panel HRR.

The panel HRR is the sum of the individual cable heat release rates for all cables to and from the panel.

For panels with separate front and back sections, the CRS database does not distinguish which location the cables terminate to. For these panels (described in PanelIdennfication) a visual inspection is used to estimate a percentage of cables terminated at each section with a corresponding HRR adjustment.

BGE 4-I-16 RAN 97-031 Rev. I

i 1

Calvert Clifrs Nuclear Power Plant Internal Fire Analysis Individu J Plant ExaminItion External Events Fire Propagation Between Panels The propagation between panels evaluation is based on the guidance provided in Appendices H and M of the implementation Guide.

. A walkdown was conducted to assess the panel features. To satisfy the EPRI criteria for no propagation requires that the panel be separated from an adjacent panel by an air gap.' This configuration is refe. red to as " double wall with air gap" in the EPRI document. An air gap is defined in this study as an identifiable separation between panels. such that there is no significant conductive heat transfer mechanism. An additional requirement is that the adjacent panel not contain sensitive electronic equipment such as solid state devices.

The consequences of a postulated fire in panels which satisfy this criterion would be limited to only those pknt systems whose wiring are present within the panel enclosure. If the adjacent panel contains

' sensitive electronics, damage to the adjacent panel is assumed to occur unless cooling is provided or the adjacent panel is not fully enclosed.

For those cases where a partial boundary separates adjacent panels, the likelihood of hot gas layer formation, spread of hot gases to an adjacent panel, arrangement of potential ignition sources and combustible material within the exposing and exposed panels, and the location of potential propagation pathways are qualitatively evaluated. Appendix H of the EPRI document states that fire spread to an adjacent panel is delayed by fifteen minutes even when there is no internal barrier. This suggests that damage to an adjacent panel may occur, but damage to the next further panel will not commence for at least fifteen minutes.

PanelDamageAssessment Each Control Room panel is analyzed a fire initiator, resulting in 101 individual fire scenarios. Panels are analyzed for loss of function due to propagation, hot gas layer effects, and radiant exposure damage, in all cases, all the initiating panel functions are considered lost. In addition, when damaged, all functions of the exposed panel are considered lost. (This is conservative when the exposed panel houses sensitive electronic components.) Panel damage assessment is performed as follows:

Propagation to adjacent panels is evaluated using the approach described above. Panels in

-e the " horseshoe" section are an open back configuration that is not amenable to propagation when double walled. Hence, when adjacent panels are separated by a double wall the exposed panel functions remain intact. Enclosed and closed ventilated cabinets separated by double walls are treated in a like fashion.

The open back panels with single walls are considered as having inefTective barriers.

' Consequently, when adjacent panels are separated by a single wall, all exposed panel functions are failed and grouped as a single initiator.

Enclosed and. closed. ventilated cabinets that are fully separated by single walls are qualitatively evaluated for propagation. This evaluation considers the placement, concentration, and type of cables in the exposing cabinet as well as the resulting HRR. The BGE 4-I-17 RAN 97-03i Rev.1

1 4

Calvert Clim Nuclear Power Plant.

Intemal Fire Analysis

. Individual Plant Examination Exterrd Evt.nts adjacent exposed panel is. evaluated for component type and placement, the potential for convective cooling inside the panel, and any other features that might effect propagation.

.Where propagation occurs all adjacent panel functions are failed and grouped as a single initiator.

Panels damaged as a result of propagation beyond the first adjacent panel are not evaluated for functional loss. This is because the time required for fire spread would exceed the established time criterion for Control Room abandonment and, additionally, include an unsuccessful Fire Brigade response, Hot gas layer damage is unlikely. Control room panels (excluding IC18A, IC18B, IC19C, e

IC20A. and IC20B, which are enclosed double wall cabinets) have open mesh tops with a small angle iron soffit and no other features that would trap hot gas.

The open backs of the main control panels create a radiant exposure case for the panels on e

the opposite side of the interior passageway. The radiant damage range is computed using the EPRI worksheet formula:

Y[(HRR eak

  • Radiant fraction of HRR) / (4x
  • Critical radiant flux)]

[

Range =

p The radiant fraction of HRR is 0.40, that is,40% of the HRR is released as radiant energy.

Critical radiant flux, the threshold for radiant energy impingement that will damage the 2

target,' is 1.0 Btu /s/ft. (EPRI tests [A Study of Denugeability of Electrical Cables in Simulated Fire Environments, NP-1767] show that the critical flux for cable jacket / insulation degradation is typically 20i 4 kW/m2 c.,2 Btu /s/ft ). Therefore, radiant 2

damage distances are computed (using calculated panel HRRs described earlier):

Y[(HRR eak

  • 0.40)/(4x
  • 1.0)] = Y[(Btu /s
  • 0.40)/(4x
  • Btu /s/ft )) = ft 2

Range =

p For adjacent or opposing pancis susceptible to radiant damage all functions are assumed lost.

Conservativejudgment is used when the targets included sensitive electronic components.

Table 4-1-2 shows the panel weight, the weighting factor, heat release rate and damage range, panels i

damaged from an ignition,'and the panel configuration.

BGE 4 I-18 RAN 97-031 Rev. I

1 Calvert Cliffs Nuclear Power Pir.nt Internal Fire Analysis Individad P11nt Examination External Events

)

Table 4-1-2 Control Room Panel Fire Modeling Results Panel Penel Penel Weighting Pook-Damage Damaged Description Weight Factor -

HRR Range Penels -

1C01 MAIN GENERATOR AND 1.00 1.22E-02 468 3'10" 1C02,1C29 SWITCHYARD CONTROL BOARD 1CO2 TURBINE CONTROL BOARD 1.00 1.22E-02 406 3'7" 1C01,1C23 1C03 CONDENSATE AND FEEDWATER 0.50 6.08E-03 104 1'10" 1C03 BK CONTROL BOARD 1C03 CONDENSATE AND FEEDWATER 1.00 1.22E-02 935 5' 5" 1C35,1C36 FR CONTROL BOARD 1C04 AUXILIARY FEEDWATER AND 1.00 1.22E-02 366 3'5" 1C28,1C31, COMPUTER CONTROL BOARD 1C32,1C05 1C05 REACTIVITY CONTROL BOARD 1.00 1.22E-02 593 4'4" 1C04 1C06 REACTOR COOLANT CONTROL 0.50 6.08E-03 80 1' 7" NO OTHER BK BOARD 1C06 REACTOR COOLANT CONTROL 1.00 1.22E-02 720 4' 9" 1C25A, B, C, & D; FR BOARD 1C39 1C07 CHEMICAL AND VOLUME 1.00 1.22E-02 471 3'10" 1C39 CONTROL BOARD 1C08 ENGINEERING SAFEGUARDS 1.00 1.22E-02 203 2' 6" 1C09 BK 1C08 ENGINEERING SAFEGUARDS 1.00 1.22E42 203 2'6" 1C09 FR 1C09 ENGINEERING SAFEGUARDS 1.00 1.22E-02 211 2' 7" 1C08,1C10 BK BGE 4-I-19 RAN 97-031 Rev.1

Calvert ClitTb bclect Power Plant inter:nl Fire Anahsis individuci Picnt Examination External Events i

Table 4-1-2 Control Room Panel Fire Modeling ReSults (Cont'd)

Panet Panel {-

Panel Weighting Peak Damage Damaged Descripuon Weight Factor '

.HRR; Range Panole 1C09 ENGINEERING SAFEGUARDS 1.00 1.22E-02 491 3'11" 1C08,1C10 FR l

i i

1C10 ENGINEERING SAFEGUARDS 1.00 1.22E-02 132 2' 1" 1C09 BK 1C10 ENGINEERING SAFEGUARDS 1.00 1.22E-02 308 3' 2" 1C09 J

FR 1C13 SALT WATER, SERVICE WATER 1.00 1.22E-02 832 5' 2" 1C24B,1C15X, AND COMPONENT COOLING 1C26,1C248 WATER CONTROL BOARD 1C15A REACTOR PROTECTION SYSTEM 0.75 9.12E-03 121 1'12" NO OTHER j

CHANNEL A 1C15B REACTOR PROTECTION SYSTEM 0.75 9.12E-03 89 1' 8" NO OTHER CHANNELB 1C15C REACTOR PROTECTION SYSTEM 0.75 9.12E-03 90 1'8" NO OTHER

)

CHANNEL C 1C15D REACTOR PROTECTION SYSTEM 0.75 9.12E-03 106 1'10" NO OTHER CHANNEL D 1C15X REACTOR PROTECTION SYSTEM 0.75 9.12E-03 65 1' 5" NO OTHER l SOLATION CABINET 1C17 4KV AND 480VAC SYSTEM 1.00 1.22E-02 391 3' 6" NO OTHER NORMAL CONTROL BOARD l

1C18 13KV AND 4KV SYSTEM 1.00 1.22E-02 397 3' 7" NO OTHER ESSENTIAL CONTROL BOARD 1C18A EDG 1A 4KV AND 480VAC 1.00 1.22E-02 100 1' 9" NO OTHER CONTROL BOARD BGE 4-1-20 RAN 97-031 Rev. I

Calvert ClitTs Nuclear Power Plant Internal Fire Analysis Individual Plant Exami.ution External Events Table 4-1-2 Control Room Panel Fire Modeling Results (Cont'd)

Panel

Panel Panel Weighting Peak' Damage Damaged Description Weight Factor.

.HRR.

Range

. Panels 1C18B EDG 1B 4KV AND 480VAC 1.00 1.22E-02 100 1' 9" NO OTHER CONTROL BOARD 1C19 13KV AND 4KV SYSTEM 1.00 1.22E-02 688 4' 8" NO OTHER ESSENTIAL CONTROL BOARD 1C19C EDG OC CONTROL BOARD 1.00 1.22E-02 100 1'9" NO OTHER 1C20 13KV AND 4KV ESSENTIAL 1.00 1.22E-02 315 3'2" NO OTHER CONTROL BOARD 1C20A 2A DG CONTROL BOARD 1.00 1.22E-02 100 1' 9" NO OTHER 1C208 2B DG CONTROL BOARD 1.00 1.22E-02 100 1' 9" NO OTHER I

1C22A RADIATION MONITORING AND 1.00 1.22E-02 70 1' 6" 1C22A-H METEOROLOGICAL CONTROL BOARD 1C22B RADIATION MONITORING AND 1.00 1.22E-02 107 1'10" 1C22A-H METEOROLOGICAL CONTROL BOARD 1C22C RADIATION MONITORING AND 1.00 1.22E-02 65 1' 5" 1C22A-H METEOROLOGICAL CONTROL BOARD 1C22D RADIATION MONITORING AND 1.00 1.22E-02 78 1'7" 1C22A-H METEOROLOGICAL CONTROL BOARD 1C22E RADIATION MONITORING AND 1.00 1.22E-02 69 1'6" 1C22A-H METEOROLOGICAL CONTROL BOARD 1C22F RADIATION MONITORING AND 1.00 1.22E-02 96 f 9" tC77A-H METEOROLOGICAL CONTROL BOARD BGE 4-1-21 RAN 97-031 Rev. I

Calvert Cliffs Eclear Power Plant Internal Fire Analysis Individual Plant Examination External Events Table 4-I-2 Control ROOM Panel Fire MOdeling Results (Cont'd)

Panet Panet.

Panel Weighting

.Posk Damage P+naged' Description Weight Factor HRR_

Range

Panels,

1C22G RADIATION MONITORING AND 1.00 1.22E-02 65 1' 5" 1C22A-H METEOROLOGICAL CONTROL BOARD 1C22H RADIATION MONITORING AND 1.00 1.22E-02 121 1'12" 1C22A-H METEOROLOGICAL CONTROL BOARD 1C23 PLANT OSCILLOGRAPH 1.00 1.22E-02 108 1'10" NO OTHER DAMAGE 1C24A DC POWER CONTROL BOARD 1.00 1.22E-02 262 2'11" 1C24B 1C24B FIRE PROTECTION CONTROL 1.00 1.22E-02 364 3' 5" 1C13,1C15C, BOARD 1C15D 1C25A RPS CHANNEL A POWER SUPPLY 0.25 3.04E-0?

149 2' 2" 1C06,1C25B CABINET 1C25B RPS CHANNEL B POWER SUPPLY 0.25 3.04E-03 119 1'11" 1C25A,1C25C, CABINET 1C06 1C25C RPS CHANNEL C POWER SUPPLY 0.25 3.04E-03 92 1' 9" 1C25B 1C25D, CABINET 1C06 1C25D RPS CHANNEL D POWER SUPPLY 0.25 3.04E-03 92 1' 9" 1C25C,1C06 CABINET 1C26 VIBRATION MONITOR PANEL 1.00 1.22E-02 150 2'2" 1C24B CONTROL BOARD 1C28 TECHNICAL SUPPORT CENTER 0.50 6.08E-03 66 1' 5" NO OTHER ISOLATION PANEL 1C28A TSCC COMPUTER MULTIPLEXER 0.50 6.08E-03 65 1' 5" NO OTHER BGE 4-I-22 RAN 97-031 Rev. I

Cottert Cliffs Nuclear Power Plant Internal Fire Analysis Individual Plant Examination External Events Table 4-I-2 Control Room Panel Fire MOdeling ReSults (Cont'd)

Panel Panel Panel Weighting Peak Damage Damaged

~ Description Wolght

. Factor' HRR

. Range.

Panels l

1C29 CB GEN METERING & MISC 1.00 1.22E-02 208 2' 7" NO OTHER 1

1C30A UNIT 1 LOOSE PARTS 1.00 1.22E-02 65 1' 5" NO OTHER MONITORING SYSTEM i

1C30B UNIT 2 LOOSE PARTS 1.00 1.22E42 65 1' 5" NO OTHER j

MONITORING SYSTEM 1C31 REACTOR REGULATION SYSTEM 0.50 6.08E-03 65 1' 5" NO OTHER CHANNEL X CONTROL BOARD 1C32 REACTOR REGULATION SYSTEM 0.50 6.08E-03 65 1' 5" NO OTHER CHANNEL Y CONTROL BOARD 1C33 WPS PANEL 1.00 1.22E-02 99 1' 9" NO OTHER 1C34 HVAC SYSTEM CONTROL BOARD 1.00 1.22E-02 479 3'11" 1C33 1C35 FEEDWATER REGULATION 0.25 3.04E-03 70 1' 6" NO OTHER

)

CONTROL SYSTEM 11 CABINET i

1C36 FEEDWATER REGULATION 0.25 3.04E-03 66 1' 5" NO OTHER CONTROL SYSTEM 12 CABINET

)

1C38 STEAM GENERATOR FEED WATER 0.25 3.04E-03 65 1' 5" NO OTHER PUMP 11 & 12 SIGNAL PROCESSOR CABINET 1C39 MISCELLANEOUS STATION 1.00 1.22E-02 230 2' 8" NO OTHER RECORDER PANEL 2C02 TURBINE CONTROL BOARD 1.00 1.22E-02 401 3'7" 2C23 BGE 4-I-23 RAN 97-031 Rev.1

l

~

Caltert Cliffs Nuclear Power Pitnt Internal Fire Analysis Individut! Plant Examination External Events Table 4-I-2 Control Room Panel Fire Modeling Results (cont'd)

Panel Panel Panet We6ghting Peak-Damage Damaged Description Weight Factor -

HRR Range Panels 2C03 CONDENSATE AND FEEDWATER 0.50 6.08E-03 111 1'11" 2003 BK CONTROL BOARD 2C03 CONDENSATE AND FEEDWATER 1.00 1.22E-02 1002 5' 8" 2C36,2C35 FR CONTROL BOARD

]

2C04 AUXILIARY FEEDWATER AND 1.00 1.22E-02 338 3'3" 2C28,2C31, COMPUTER CONTROL BOARD 2C32,2005 2C05 REACTIVITY CONTROL BOARD 1.00 1.22E-02 614 4' 5" 2C04 2COS REACTOR COOLANT CONTROL 0.50 6 08E-03 76 1' 7" NO OTHER BK BOARD 2C06 REACTOR COOLANT CONTROL 1.00 1.22E-02 686 4' 8" 2C25A, B, C, & D; FR BOARD 2C39 2007 CHEMICAL AND VOLUME 1.00 1.22E-02 490 3'11" 2C39 CONTROL BOARD 2C08 ENGINEERING SAFEGUARDS 1.00 1.22E-02 180 2' 5" 2C09 BK 2C08 ENGINEERING SAFEGUARDS 1.00 1.22E-02 180 2' 5" 2C09 FR 2C09 ENGINEERING SAFEGUARDS 1.00 1.22E42 183 2' 5" 2C08,2C10 BK 2C09 ENGINEERING SAFEGUARDS 1.00 1.22E-02 426 3' 8" 2008,2C10 FR 2C10 ENGINEERING SAFEGUARDS 1.00 1.22E-02 122 1'12" 2009 BK BGE 4-I-24 RAN 97-031 Rev. I l

i n

1 e

Caltert Clifts Nuclear Power Plant internal Fire Analysis Individual Plant Examination Externe.1 Events Table 4-1-2 Control Room Panel Fire Modeling Results (Cont'd)

Penel Penel Panel Weighting Peelt Damage Damaged

{

. Description Weight Factor HRR

' Range-Panels 2C10 ENGINEERING SAFEGUARDS 1.00 1.22E-02 286 3' 0" 2009 FR 2C13 SRW/ MISC SERVICES CONTROL 1.00 1.22E-02 663 4' 7" 2C26,2C24B, PANEL 2C15X 2C15A REACTOR PROTECTION SYSTEM 0.75 9.12E 03 115 1'11" 2C15B CHANNEL A 2C15B REACTOR PROTECTION SYSTEM 0.75 9.12E-03 85 1' 8" 2C15A,2C15C CHANNELB 2C15C REACTOR PROTECTION SYSTEM O.75 9.12E-03 93 1' 9" 2C158,2C15D CHANNEL C 2C15D REACTOR PROTECTION SYSTEM 0.75 9.12E-03 107 1'10" 2C15C CHANNEL D 2C15X REACTOR PROTECTION SYSTEM 0.25 3.04E-03 65 1' 5" NO DAMAGE ISOLATION CABINET 2C17 4KV/480VAC SYSTEM NORMAL 1.00 1.22E-02 389 3'6" NO DAMAGE CONTROL BOARD 2C23 VIBRATION MONITOR 1.00 1.22E-02 215 2'7" NO.OTHER DAMAGE 2C24A SALTWATER ISOLATION PANEL 1.00 1.22E-02 460 3'10" 2C24B,2C15A-D 2C24B RADIATION MONITORING PANEL 1.00 1.22E-02 132 2'1" 2C24A,2C26 2C25A RPS CHANNEL A POWER SUPPLY 0.25 3.04E-03 151 2'2" 2C06,2C25B CABINET 1

I l

I BGE 4-1-25 RAN 97-031 Rev.1

e Ca.lvert Cliffs Nuclear Power Plant internal Fire Analysis Individu.tl Plant Examinetion External Events Table 4-I-2 Control Room Panel Fire Modeling Results (Cont'd)

Panet

' Panel-Panel Weighting Peak Damage Damaged Description :

Weight Factor HRR.

Range Panele 2C25B RPS CHANNEL B POWER SUPPLY 0.25 3.04E-03 121 1'12" 2C25A,2C25C.

CABINET 2C06 2C25C RPS CHANNEL C POWER SUPPLY 0.25 3.04E-03 92 1' 9" 2C25B,2C25D, CABINET 2006 2C25D RPS CHANNEL D POWER SUPPLY 0.25 3 04E-03 92 1' 9" 2C25C,2C06 CABINET 2C26 VIBRATION MONITOR PANEL 1.00 1.22E-02 86 1'8" NO OTHER CONTROL BOARD 2C28 TECHNICAL SUPPORT CENTER 0.50 6.08E-03 65 1' 5" NO OTHER ISOLATION PANEL 2C28A TSCC COMPUTER MULTIPLEXER 0.53 6.08E-03 65 1' 5" NO OTHER 2C31 REACTOR REGULATION SYSTEM 0.50 6.08E-03 65 1' 5" NO OTHER CHANNEL X CONTROL BOARD 2C32 REACTOR REGULATION SYSTEM 0.50 6.08E-03 65 1' 5" NO OTHER CHANNEL Y CONTROL BOARD 1

2C34 HVAC SYSTEM CONTROL PANEL 1.00 1.22E42 65 1'5" NO OTHER 2C35 FEEDWATER RCS 21 0.25 3.04E-03 84 1' 8" NO OTHER 2C36 FEEDWATER RCS 22 0.25 3.04E-03 84 1' 8" NO OTHER 2C38 UNIT 2 FEED PUMP TURBINE 1.00 1.22E-02 85 1' 8" NO OTHER CONTROL BOARD BGE 4-I-26 RAN 97-031 Rev. I

e Caltert Cliffs Nuclear Power Plant Internal fire Analysis Individu:1 Plznt Examination External Events Table 4-1-2 Control Room Panel Fire Modeling Results (cont'd)

Panel

~ Panel --

. Panel Weighting

. Peak Damage Damaged Description Weight Factor i HRR Range' Panels 2C39 MISCELL.ANEOUS STATION 1.00 1.22E-02 417 3' 8" NO OTHER RECORDER PANEL 1C60-1 PLANT COMPUTER CABINET 1.00 1.22E-02 120 1'11" NO OTHER 1C60-2 PLANT COMPUTER CABINET 1.00 1.22E-02 94 1' 9" NO OTHER 2C60-1 PLANT COMPUTER CABINET 1.00 1.22E-02 111 1'11" NO OTHER 2C60-2 PLANT COMPUTER CABINET 1.00 1.22E-02 96 1' 9" NO OTHER Determination ofPanelFunction andResults The functions for each panel are determined through a review of the cable routing and walkdowns. The cable and raceway schedule (CRS) identifies panels as a "to" or "from" location. Therefore, using the cable database developed for the PRA, applicable functions can be identified. Panel walkdowns and control board pictures confirmed the cable database results and, in some cases, produced additional panel functions.

I When applicable, the individual panel function (s) are mapped to the corresponding model top event (s).

To simplify modeling and reduce the number ofinitiators, the panel fire scenarios having similar impact are binned together. Thus,101 panels result in thirty-two Fire Initiating Events (FI A).

Table 4-1-1 shows, for each FIA, the initiation frequency and resulting core damage contribution. Fire scenarios that comprise each FIA are listed in groups. For each group, the underlined panel identifier signifies the fire scenario initiator with any associated panels damaged from the ignition listed below.

BGE 4-1-27 RAN 97-031 Rev. I 1

j

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