ML20236F047
| ML20236F047 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 06/26/1998 |
| From: | Dromerick A NRC (Affiliation Not Assigned) |
| To: | Cruse C BALTIMORE GAS & ELECTRIC CO. |
| References | |
| TAC-M83603, TAC-M83604, NUDOCS 9807020003 | |
| Download: ML20236F047 (13) | |
Text
,
m 26, M 8 Mr. CharlIs H. Cruse f
Vice Frzsid:nt-Nuclear Energy
(
Biltim:re G:s cnd Electnc Comp:ny Calvert Cliffs Nuclor Power Plant 1657 Calvert Cliffs Parkway Lusby, MD 20657-4702
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 AND 2 - lPEEE SUBMITTAL (TAC NOS. M83603
)
AND M83SO4) f
Dear Mr. Cruse:
Based on our ongoing review of the Calvert Cliffs 1 and 2 individual Plant Examination of Extemal Events (IPEEE) submittal of August 28,1998, we have developed the enclosed request for additional information (RAl). The RAls are related to the IPEEE analyses in the areas related to seismic and fire events. The RAls in the seismic and fire areas were developed by our
. contractors, Brookhaven and Sandia National Laboratories, respectively. There are no RAls
- related to high winds, floods, and other extemal events.
We request that Baltimore Gas and Electric Company provide its response within 60 days of receipt of this letter. -if you have any questions regarding this matter, please contact me on (301) 415-3473.
Sincerely,
)
Original Signed by:
)
i Alexander W. Dromerick, Senior Project Manager Project Directorate I-1 Division of Reactor Projects - 1/II l
Office of Nuclear Reactor Regulation i
Docket Nos. 50-317
/
and 50-318 g
Enclosure:
Request for Additional Information cc w/ encl: See next page 1)li)/
DISTRIBUTION:
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Mr. Charles H. Cruse June 26, 1998 Vice Pr;sid:nt - Nucl:ar Energy Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Lusby, MD 20657-4702 I
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON CALVERT CLIFFS NUCLEAR PO'NER PLANT, UNIT NOS.1 AND 2 - IPEEE SUBMITTAL (TAC NOS. M83603 AND M83604)
Dear Mr. Cruse:
Based on our ongoing review of the Calvert Cliffs 1 and 2 individual Plant Examination of Extemal Events (IPEEE) submittal of August 28,1998, we have developed the enclosed request for additionalinformation (RAl). The RAls are related to the IPEEE analyses in the areas related to seisn.ic and fire events. The RAls in the seismic and fire areas were developed by our contractors, Brookhaven and Sandia Natir.nal Laboratories, respectively. There are no RAls related to high winds, floods, and other extemal events.
I We request that Baltimore Gas and Electric Company provide its response within 60 days of receipt of this letter, if you have any questions regarding this matter, please contact me on (301) 415-3473.
Sincerely, Original Signed by:
i A!axander W. Dromerick, Senior Project Manager Project Directorate 1-1 Division of Reactor Projects - t/11 I
Office of Nuclear Reactor Regulation Docket Nos. 50-317 and 50-318
Enclosure:
Request for Additional Information cc w/ encl: See next page DISTRIBUTION:
Docket File OGC PUBLIC ACRS PDI-1 R/F C. Hehl, Region 1 J. Zwolinski (A)
S. Bajwa S. Little A. Dromerick A. Rubin, T-10E50 DOCUMENT NAME:G:\\CC1-2\\C1283603.RAI n.
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j NUCLEAR REO;)LATORY COMMISSION f
WASHINGTON, D.C. 20006 4001 p%
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June 26, 1998 l
Mr. Charies H. Cruse Vice President-Nuclear Energy Baltimore Gas and Electric Company l
1 Calvert Cliffs Nuclear Power Plant
[
1650 Calvert Cliffs Parkway Lusby, MD 20657-4702
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 AND 2 - IPEEE SUBMITTAL (TAC NOS. M83603 AND M83604) i
Dear Mr. Cruse:
Based on our ongoing review of the Calvert Cliffs 1 and 2 Individual Plant Examination of Extemal Events (IPEEE) submittal of August 28,1998, we have developed the enclosed request for additional information (RAI). The RAls are related to the IPEEE analyses in the areas related to seismic and fire events. The RAls in the seismic and fire areas were developed by our l
contractors, Brookhaven and Sandia National Laboratories, respectively. There are no RAls l
related to high winds, flooos, and other extemal events.
l We request that Baltimore Gas and Electric Company provide its response within 60 days of receipt of this letter. If you have any questions regarding this matter, please contact me on (301) j 415-3473.
l l
Sincerely, U
i Alex der W. Dromerick, Senior Project Manager Project Directorate 1-1 l
Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation f
Docket Nos. 50-317 and 50-318
Enclosure:
Request for Additional Information cc w/ encl: See next page 1
l l
e, I
Mr. Charles H. Cruse Baltimore Gas & Electric Company Calvert Cliffs Nuclear Power Plant cc:
President Mr. Joseph H. Walter, Chief Engineer Calvert County Board of Public Service Commission of Commissioners Maryland 175 Main Street Engineering Division Prince Frederick, MD 20678 6 St. Paul Centre Baltimore, MD 21202-6806 James P. Bennett, Esquire Counsel Kristen A. Burger, Esquire Baltimore Gas and Electric Company Maryland People's Counsel 4
P.O. Box 1475 6 St. Paul Centre Baltimore, MD 21203 Suite 2102 Baltimore, MD 21202-1631 Jay E. Silberg, Esquire Shaw, Pittman, Potts, and Trowbridge Patricia T. Bimie, Esquire 2300 N Street, NW Co-Director Washington, DC 20037 Maryland Safe Energy Coalition P.O. Box 33111 Mr. Bruce S. Montgomery, Director Baltimore, MD 21218 -
NRM Calvert Cliffs Nuclear Power Plant Mr. Loren F. Donatell 1650 Calvert Cliffs Parkway NRC Technical Training Center Lusby, MD 20657-4702 5700 Brainerd Road Chattanooga, TN 37411-4017 Resident inspector U.S. Nuclear Regulatory i
Commission P.O. Box 287 St. Leonard, MD 20685 i
Mr. Richard I. McLean, Manager Nuclear Programs Power Plant Research Program Maryland Dept. of Natural Resources Tawes State Office Building, B3
. Annapolis, MD 21401 Regional Administrator, Region l U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406
s
+
CALVERT CLIFFS NUCLEAR POWER PLANT. UNIT NOS.1 AND 2 REQUEST FOR ADDITIONAL INFORMATION ON IPEEE SUBMITTAL Seismic 1
- 1. The Calved Cliffs Nuclear Power Plant (CCNPP) IPEEE analysis shows that sequences with failure of all but one emeigency diesel generator (EDG) and resulting spurious safety system actuation are dominant. It seems that de operator action of cross-tying vital buses to backup power before battery depletion is an important determinant of the risk contribution of such sequences.
a) Please describe the operator actions needed to accomplish such vital bus cross-tying, including whether such actions are proceduralized, whether it calls for the shedding of certain loadsis required, whether such actions are local or control room activated, and whether there are accessibility problems.
b) Please also provide a discussion related to the impact on core damage frequency if no credit is given to this operator action.
- 2. In the submittal, no information is provided regarding the evaluation procedures for component and equipment fragilities.' To allow completion of the IPEEE review, please provide detailed calculations and results for the following components:
l Control Room HVAC Control Panels 1NB108,2NB408 (median =0.66g).
Refueling Water Storage Tank (RWST) (median =0.48g).
Condensate Storage Tank (CST) (median =0.96g).
Diesel Generator 1 A.
- 3. In Section 3.2 of the submittal, it is stateil that "During tho walkdown the equipment was evaluated based on the screening c:riteria presented in EPRI NP-6041, "A Methodology for l
Assessment of Nuclear Power Plant Seismic Margin." However, the screening criteria in EPRI NP-6041 are based on the NUREG!CR-0098 spectra, while the Lawrence Livermore National Laboratory (LLNL) uniform hazard spectrum (UHS), which has a much lower spectral amplification factor, was used in the IPEEE. Please provide the rationale used for this screening criterion. In padicular, describe the application of this screenine to the following components (see Table 2-4 of EPRI NP-6041):
- NSSS supports.
- Control rod drive housings and mechanisms.
- Active valves.
- Batteries and racks
- Vertical pumps.
- 4. There is, no discussion of the instrument air system in the submittal. This system was found to be risk important at some plants. As no dependency matrix is given, please describe this j
system, its dependencies and the role of instrument air in the plant (including t::ontcinment l
isolation) Please discuss how this system was modeled in the PRA. Does this system I
impact any containment performance izsues? If it does, please describe the impact and its effect on containment perforrrance.
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describe this system, its dependencies and the role of instrument air in the plant (including containment isolation). Please discuss how this system was modeled in the PRA. Does this system impact any containment performance issues? If it does, please describe the impact and its effect on containment performance.
5.
Correlated failures can be important in seismic analysis in that they defeat redundancy of equipment. It is not stated in the submittalif any correlated failures were modeled (for example, in case of the condensate storage tanks). Please clarify the treatment of correlated failures.
High Winds, Floods, and Other External Events There are no RAls in these areas.
2
Fire 1.
NUREG-1407, Section 4.2 and Appendix C, and GL 88-20, Supplement 4, request that
- documentation be submitted with the iPEEE submittal with regard to the Fire Risk Scoping Study (FRSS) issues, including the basis and assumptions used to address these issues, and a discussion of the findings and conclusions. NUREG-1407 also reouests that evaluation results and potential improvements be specifically highlighted.
Control system interactions involving a combination of fire-induced failures and high probability random equipment failures were identified in the FRSS as potential contributors to fire risk.
The issue of control systems interactions is associated primarily with the potential that a fire in the plant (e.g., the main control room [MCR]) might lead to potential control systems vulnerabilities. Given a fire in the plant, the likely sources of control systems interactions could happen between the control room, the remote shutdown panel, and shutdown systems. Specific areas that have been identified as requiring attention in the resolution of this issue include:
(a)
Electricalindependence of the remote shutdown control systems: The primary concern of control systems interactions occurs at plants that do not provide
' independent remote shutdown control systems. The electrical independence of the remote shutdown panel and the evaluation of the level of indication and control of remote shutdown control and monitoring circuits need to be assessed.
(b)
Spurious actuation of components leading to component damage, loss-of-coolant accident (LOCA), or interfacing ' systems LOCA: The spurious actuation of one or more safety-related to safe-shutdown-related components as a result of fire-induced cable faults, hot shorts, or component failures leading to component damage, LOCA, or interfacing systems LOCA, prior to taking control from the remote. shutdown panel, needs to be assessed. This assessment also needs to include the spurious starting and running of pumps as well as the spurious repositioning of valves.
Also, pege 4-67, Section 4.8.5, of the submittal states that a licensee analysis had -
" confirmed that safe plant shutdown can be achieved by using available controls and indications located on the Auxiliary Shutdown Panel" and that "certain safe shutdown components are provided with local / remote switching capabilities to provide local control under certain fire scenarios." However, the submittal provided no discussion of the control and instrumentation functions that are available at the Auxiliary Shutdown Panel (ASP), those functions that can or can not be locally isolated, nor the nature of the fire scenarios that can or can not be mitigated using the ASP.
Please provide a description of the control and instrumentation functions that are provided on the Auxiliary Shutdown Panel. For each such function indicate whether or not it can be isolated from damage in the main control room. Describe in detail the Appendix R analysis that is cited as having demonstrated the adequecy of this remote
- shutdown capability. In particular, describe the fire / damage scenarios that were 3
1
considered in the Appendix R analysis and discuss the correspondence between those scenarios and the scenarios identified in the IPEEE that might require use of the ASP.
Has the IPEEE identified or considered any scenarios that might not be mitigated by the ASP 7 Provide an evaluation of the reliability of the ASP that includes consideration of spurious actuations that might result from fire-induced cable faults, hot shorts, or component failures. Include in this evaluation the potential for such faults to lead to I
component damage (including damage to MOVs per information Notice 92-18), a LOCA or interfacing system LOCA, spurious starting and running of pumps, and repositioning of valves.
.2.
The analysis of the main control room (MCR) has assumed a conditional probability of MCR abandonment given a fire of 3.4E-3 based on a suppression model taken from the EPRI Fire PRA Implementation Guide (FPRAIG). However, this abandonment model only applies to situations that include an optimally placed smoke detector within the initiating panel. The use of the 3.4E-3 conditional abandonment probability has n6t been adequately justified and should include specific consideration of the test conditions as compared to the configuration at Calvert Cliffs. Further, the analysis has independently applied two separate severity factors; namely, a 6/9 reduction for manually suppressed fires versus the total number of MCR fires, and a 1/3 reduction for severe versus minor fires. All three of these factors inherently consider the same phenomena; namely, the likelihood that a fire would be suppressed before control room abandonment would be required. Hence, independent appl; cation of all three factors appears to represent " double counting" for the likelihood of suppression.
Based on the above discussion, please reassess the conditional probability of control room abandonment, given a fire, that includes specific consideration of the fire detection features provided at Calvert Cliffs. Also, reassess the CDF contribution due to MCR fires assuming that only one factor is applied to a given sequence (i.e., either apply a single severity factor or a single non-suppression probability to the analysis but not both).
3.
The hes loss factor is defined as the fraction of energy released by a fire that is transferred to the enclosure boundaries. This is a key parameter in the prediction of component damage, as it determines the amount of heat available to the hot gas layer.
In Fire-induced Vulnerability Evaluation (FIVE), the heat loss factor is modeled as being i
inversely related to the amount of heat required to cause a given temperature rise.
Thus, for example, a larger heat loss factor means that a larger amount of heat (due to a more severe fire, a longer buming time, or both) is needed to cause a given temperature rise. It can be seen that if the value assumed for the heat loss factor is unrealistically high, fire scenarios can be improperly screened out. Figure 1 provides a representative example of how hot gas layer temperature predictions can change assuming different heat loss factors. Note that: 1) the curves are computed for a 1000 kW fire in a 10m x Sm x 4m compartment w;th a forced ventilation rate of 1130 cfm; 2) the FlVE-recommended damage temperature fer qualified cable is 700*F for qualifed cable and 450'F for unqualified cable; and,3) the SFPE curve in the figure is generated i;
from a correlation provided in the Society for Fire Protection Engineers Handbook [3.1].
4 L'
1 I
Time Temperature curves l
900 e
s
_e 800 SFE
_e H.F = 0.70
+ H.F = 0.85 600-
+ H.F = 0.94 i
j a _ FLF = 0.99 l
g.
400.
E~
a 200
. Ax xxxm.x.xx:xg**xmexxmx"N]
100 0$kk$2$$$kh$$$$
Time (s)
Figure 1 Sensitivity of the hot gas layer temperature preciictions to the assumed heat loss factor Based on evidence provided by a 1982 paper by Cooper et al. [3.2], the EPRl Fire PRA l
Implementation Guide recommends a heat loss factor of 0.94 for fires with duraticns I
greater than five minutes and 0.85 for " exposure fires away from a wall and quickly l
developing hot gas layers." However, as a general statement, this appears to be a l
misinterpretation of the results. Reference [3.2], which documents the results of multi-compartment fire experiments, states that the higher heat loss factors are associated with the movement of the hot gas layer from the burning compartment to adjacent, cooler compartments. Earlier in the experiments, where the hot gas layer is limited to the burning compartment, Reference [3.2] reports much lower heat loss factors (on the order of 0.51 to 0.74). These lower heat loss factors are more appropriate when analyzing a single compartment fire.
l l
l in summary, (a) hot gas layer predictions are very sensitive to the assumed value of the
{
heat loss factor; and (b) large heat loss factors cannot be justified for single-room scenarios based on the information referenced in the EPRI Fire FRA Implementation i
Guide.
l The Calvert Cliffs IPEEE fire study uses a heat loss factors of 0.85 based on Inform.ation in the FPRAIG. Hot gas layer (HGL) effects were apparently considered in detailed l
1 5
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=
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compartment analyses. For each scenario where the hot gas layer temperature was calculated, please verify the heat loss factor value used in the analysis. In light of the preceding discussion, please either: a) Justify the value used and discuss its effect on the identification of fire vulnerabilities, or b) repeat the analysis using a more justifiable value and provide the resulting change in scenario contribution to core damage frequency.
3.1 P. J. DiNenno, et al, eds., "SFPE Handbook of Fire Protection Engineering," 2nd Edition, National Fire Protection Association, p. 3-140,1995.
3.2 L. Y. Cooper, M. Harkleroad, J. Quintiere, W. Rinkinen, "An Experimental Study of Upper Hot Layer Stratification in Full-Scale Multiroom Fire Scenarios," ASME Journal of Heat Transfer, 104,741-749, November 1982.
4.
In the EPRI Fire PRA Implementation Guide, test results for the control cabinet heat i
release rate (HRR) have been misinterpreted and have been inappropriately I
extrapolated. Cabinet HRRs as low as 65 BTU /s are recommended in the Guide in contrast, experimental work has demonstrated cabinet HRRs ranging from 23 to 1171 BTU /s.
In the Calvert Cliffs assessment it would appear that different cabinet HRR have been applied based on the nature of the panels ranging from 65 to 270 BTU /s. While the use of different HRR values for different panels is fully appropriate, it would appear that the.
cited values have not been used consistently in the analysis, and that they do not encompass the HRRs measured during testing. In particular, on page 4-1-13 of the submittal, it is noted that the main control panels in the MCR are of an open configuration, and yet a HRR of only 65 BTU /s has been assumed. In addition, it is stated that more extensive damage would be realized should the panel fire intensity reach 282 BTU /s. Given the available test data, it would appear that a significant potential exists for MCR panel fires that are substantially more intense than those assumed in the study.
l Please provide an assessment of the changes in the IPEEE fire assessment results if it is assurned that a MCR panel fire might reach a peak fire intensity of at least 550 BTU /s.
5.
Fires that could affect both units were not considered. The submittal indicates that some fire areas contain elements of both units. For multi-unit sites, there are several '
issues of potential interest. Hence, please answer the following:
(a)
A fire in a shared area might cause a simultaneous trip demand for more than one unit. This may considerably complicate the response of operators to the fire event, and may create conflicting demands on plant systems which are shared
. between units. Please provide the following information regarding this issue: (1) identify all fire areas that are shared between units and the potentially risk importar,t systems / components for each unit that are housed in each such area, (2) for each area iderMied in (1), provide an assessment of the associated multi-unit fire risk, (3) for the special case of control rooms, assess the likelihood of a 6
i j
a fire or smoke-induced evacuation with subsequent shutdown of both units from remote shutdown panels, and (4) provide an assessment of the risk contribution of any se:h multi-unit scenario.
-(b)
At some sites, the safe shutdown path for a given unit may call for cross-connects to a sister unit in the event of certain fires. Hence, the fire analysis should include the unavailability of the cross-connected equipment due to outages at the sister unit (e.g., routine in-service maintenance outages and/or the potential that normally available equipment may be unavailable during extended or refueling outages at the sister unit). Please provide the following information regarding this issue: (1) indicate whether any fire response safe shutdown procedures call for unit cross-connects, and (2) if any such cross-connects are required, determine the impact on fire risk if the total unavailability of the sister unit equipment is included in the assessment.
(c)
Propagation of fire, smoke, and suppressants between fire zones containing equipment for one unit to fire zones containing equipment for the other unit also can result in multi-unit scenarios. Hence, the fire assessment for each unit should include analyses of scenarios addressing propagation of smoke, fire and suppressants to and from fire zones containing equipment for the other unit.
From the information in the submittal, it is not clear if these types of scenarios are possible. Please provide an assessment of the risk contribution of any such multi-unit scenarios.
l d)
Specific to Calvert Cliffs is the fact that panel fires that require main control room (MCR) evacuation can lead to a self-induced station blackout (SISBO). This would impact both units simultaneously because the MCR is shared; however, the analysis has considered only the impact of a SISBO on a single unit. Please evaluate and discuss the core damage contribution associated with fires that might lead to MCR evacuation for both units simul *neously, including the 4
potential for,a dual unit SISBO event.
6.
The Calvert Cliffs IPEEE assessment has assumed that ' fires would not propagate beyond the contents of the cabinet itself" if the cabinet is " enclosed on all four vertical sides as well as the top and bottom" (see, for example, pp. 4-31 and 4-E-11). This assumption is consistent with the EPRI Fire FRA Implementation Guide, but can be optimistic for high-voltage cabinets (480V or higher). For high-voltage cabinets an explosive breakdown of the electrical conductors may breach the integrity of the cabinet and allow fire to spread to combustibles located above the cabinet. For example, switchgear fires at Yankee-Rowe in 1984 and Oconee Unit 1 in 1989 both resulted in fire damage outside the cubicles.
Provide an assessment of the impact on the IPEEE results if it is assumed that fires j-involving closed electrical panels of 480V or higher might propagate beyond the cabinet l_
of origin.
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7.
On page 4-40 of the submittal it is stated that " cable-to-cable failures were not considered in the analysis." Cable-to-cable hot shorts have the potential to lead to the application of disabling voltages to lower voltage circuits, and the application of power to non energized circuits. The potential risk contribution of such faults would be influenced by factors such as the mixing, within a raceway, of high and low voltage power cables, and power cables with instrument and control cables.
. Discuss the licensee practice with regard to the mixing within a raceway of power, control, and instrumentation cables. For those areas that are identified in Table 4.6.5c
" Fire Modeled Compartments" of the submittal, identify raceways where multiple power cables are routed or where power cables are mixed in a common raceway either with lower voltage power cables or with instrument and control cables. For these raceways, provide an assessment of the potential risk contribution (CDF) that might result if cable-to-cable hot shorts are assumed possible.
l 8.
It appears that the Calvert Cliffs IPEEE fire analysis has assumed that the plant cables l
are either IEEE-383 qualified or that they are equivalent to IEEE-383 qualified cables l
(see, for example, the discussion on pp. 4-33 and 4-I-13). Given the age of the two Calvert Cliffs units, this assumption appears optimistic and was not substantiated. The l
lEEE-383 standard is primarily a severe accident equipment qualification standard that also includes a flame spread test. In a fire context, it might erroneously be assumed i
that only the flame spread pass / fail status of a cable is of interest. In practice, the recommended cable damage thresholds reflect the more robust thermal performance that might be assumed for IEEE-383 qualified cables based on the demonstrated severe accident performance as compared to unqualified cables whose thermal performance has not been demonstrated (as per FIVE, the assumed damage thresholds are 700'F 2
2
' and 1 BTU /ft /s for qualified cables and 425'F and 0.5 BTU /ft /s for unqualified cables).
Hence, the assumptions of cables being equivalent to IEEE-383 cables should include consideration of both the implied flammability properties (ignition temperature, rate of flame spread, and likelihood of self-ignited fires) and thermal damage thresholds.
Please provide a specific basis for the assumption that the cables at Calvert Cliffs are either IEEE-383 qualified or equivalent to IEEE-383 qualified cables. Include in the response specific consideration of both the flammability properties (ignition, flame spread, and likelihood of self-ignited fires) and the thermal damage properties.
Attematively, provide an assessment of the impact on the analysis results (CDF) if it is assumed that the cables are not IEEE-383 equivalent and the flammability and/or appropriate non-qualified cables damage properties are used.
9.
Section 4-E of the submittal documents an analysis of the cable spreading room (CSR).
In the analysis of fires involving the "1C40 Panels" on page 4-E-12 it is stated for "the unit protection panels (A through F)" that "there are no barriers or metal extensions between panel sections." Nonetheless, the analysis has assumed that a fire could only damage a section of this panel"one panel wide (about three feet)" based on an assumption that "it is unlikely that a fire in these panels would spread or propagate in a horizontal plane." This assumption has not been substantiated, appears optimistic, and has a significant impact on the analysis. Given the lack of any physical barriers to fire 8
spread, it appears prudent to assume that under some circumstances, more widespread fire growth and damage is possible. Presumably, while the likelihood of more severe fires might be low in comparison to the scenarios that were considered, the conditional core damage probabilities (CCDPs) may also be substantially higher, hence, a sigr.ificant potential risk contributor may have been neglected.
Reassess the CDF contribution of fires involving the 1C40 panels in the cable spreading room if it is assumed that a fire might propagate between panels and describe the resulting analysis in detail. Also, quantify the CCDPs that result if one assumes damage to any one panel, any two adjacent panels, any three adjacent panels, and to the entire set of unit protection panels (A through F).
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