ML20248L110
| ML20248L110 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 06/03/1998 |
| From: | Hall J NRC (Affiliation Not Assigned) |
| To: | Horn G NEBRASKA PUBLIC POWER DISTRICT |
| References | |
| TAC-M83611, NUDOCS 9806100436 | |
| Download: ML20248L110 (13) | |
Text
_ _ _ _ - - _ _ _ _ _ _ _
Mr. G. R. Hom.
June 3, 1998 8
Sr. Vice President of Energy Supply Nebraska Public Power District.
~141415th Street Columbus, NE 68601 SUBJECT!
REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE INDIVIDUAL-PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE) FOR THE COOPER NUCLEAR STATION (TAC NO. M83611)
Dear Mr. Hom:
Based on our ongoing review of the Individual Plant Examination of Extemal Events (lPEEE) submittal dated October 30,1996, for the Cooper Nuclear Station (CNS), the staff has developed l
the enclosed request for additional information (RAl). The RAI is related to the seismic analysis, fire analysis, and high winds, floods, and other extemal events (HFO) analysis for the station.
You are requested to provide a response to the enclosed RAI within 60 days of the receipt of this letter. If there are any questions concoming the enclosure, please contact me by phone at 301-415-1336, or by e-mail at Jrh@nrc. gov on the Intemet.
Sincerely, OpIGINAL SIGNED BY:
James R. Hall. Senior Project Manager Project Directorate IV-1 Div!sion of Reactor Projects lil/IV Cffice of Nuclear Reactor Regulation Docket No. 50-298
Enclosure:
Request for Additional Information cc w/ encl: See next page
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WASHINGTON, D.C. 30666-c001 June 3, 1998 Mr. G. R. Hom Sr. Vice President of Energy Supply Nebraska Public Power District 141415th Street Columbus, NE 68601
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE) FOR THE COOPER NUCLEAR STATION (TAC NO. M83611)
Dear Mr. Hom:
Based on our ongoing review of the Individual Plant Examination of Extemal Events (IPEEE) submittal dated October 30,1996, for the Cooper Nuclear Station (CNS), the staff has developed the enclosed request for additional information (RAI). The RAI is related to the seismic analysis, fire analysis, and high winds, floods, and other extemal events (HFO) analysis for the station.
You are requested to provide a response to the enclosed RAI within 60 days of the receipt of this letter. If there are any questions conceming the enclosure, please contact me by phone at 301-415-1336, or by e-mail at Jrh@nrc. gov on the intemet.
Sincerely, ea /l. W James R. Hall, Senior Project Manager Project Directorate IV-1 Division of Reactor Projects lil/IV Office of Nuclear Reactor Regulation Docket No. 50-298
Enclosure:
Request for AdditionalInformation cc w/ encl: See next page 1
l
Mr. G. R. Hom Nebraska Public Power District Cooper Nuclear Station cc:
Mr. John R McPhail, General Counsel Lincoln Electric System Nebraska Public Power District ATTN: Mr. Ron Stoddard P. O. Box 499 1040 0 Stmet Columbus, NE 68602-0499 Box 80869 Lincoln, NE 68501 Nebraska Public Power District ATTN: Mr. J. H. Swailes MidAmerican Energy Vice President of Nuclear Energy ATTN: Dr. William D. Leech, Manager-Nuclear P. O. Box 98 907 Walnut Street Brownville, NE 68321 P. O. Box 657 Des Moines, IA 50303-0657 Randolph Wood, Director Nebraska Department of Environmental Nebraska Public Power District Control ATTN: Mr. B. L. Houston, Nuclear P. O. Box 98922 Licensing & Safety Manager Lincoln, NE 68509-8922 P. O. Box 98 Brownville, NE 68321 Mr. Larry Bohlken, Chairman Nemaha County Board of Commissioners Nemaha County Courthouse 1824 N Street Aubum, NE 68305 Senior Resident inspector U.S. Nuclear Regulatory Commission P. O. Box 218 Brownville, NE 68321 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Ariington,TX 76011 Ms. Cheryl Rogers, LLRW Program Manager Division of Radiological Health Nebraska Department of Health 301 Centennial Mall, South P. O. Box 95007 Lincoln, NE 68509-5007 Mr. Ronald A. Kucers, Department Director of Intergovemmental Cooperation Department of Natural Resources P.O. Box 176 1
Jefferson City, MO 65102 1
REQUEST FOR ADDITIONAL INFORMATION (RAll RELATED TO THE LICENSEE'S IPEEE SUBMITTAL COOPER NUCLEAR STATION This request for additional information (RAI) involves the submittal from the licensee dated October 30,1996. A list of references is provided at the end of this RAl. The separate technical areas of the RAI are listed below.
A. Fire events 1.
The IPEEE submittal indicated that hot short (HS) failures were considered in the assessment; however, it can not be determined to what extent the licensee has considered HS as a failure mode for control or instrumentation cables, in particular, HS considerations should include the treatment of conductor-to-conductor shorts within a system of two or more cables. HSs in control cables can simulate the closing of control switches leading to the repositioning of valves, spurious operation of motors and pumps, or the shutdown of operating equipment. These types of faults might lead to a loss-of-coolant accident (LOCA), diversion of flow within various plant systems, deadheading and failure of important pumps, premature or undesirable switching of pump suction sources, or undesirable equipment operations. For main control room (MCR) abandonment scenarios, such spurious operations and actions may not be indicated at the remote shutdown panel (s), may not be directly recoverab'c from remote shutdown locations, or may lead to the loss of remote shutdown capability (e.g., through loss of power sources to the remote safe shutdown panel). In instnamentation circuits, HSs may cause misleading plant readings potentially leading to inappropriate control actions or generation of actuation signals for emergency safeguard features.
Pages 4-3 and 4-6 in the licensee's submittal address the treatment of HSs in the CNS fire risk assessment. From this brief discussion, it is not clear to what extent HSs were included in the fire analysis. In particular, the potential for fire-induced LOCAs (e.g.,
through spurious opening of safety re"9f valves) and inte,&;ng system LOCAs is not discussed in the submittal. In addition, the potential for HSs in control room abandonment scenarios was not specifically addressed.
Discuss how the above HS issues (that is, the impact of HS-induced failures on safety systems or functions) have been considered in the CNS IPEEE. If they have not been considered, provide an assessment of how the inclusion of potential HS failures would impact the quantification of fire risk scenarios in the CNS IPEEE.
2.
Fires in the MCR are potentially risk-significant because they can cause instrumentation and control (l&C) failures (e.g., loss of signals or spurious signals) for redundant i
divisions, and because they can force MCR abandonment. Although data from two i
experiments concoming the timing of smoke-induced, forced MCR abandonment are available [ Reference 1), the data must be carefully interpreted, and the analysis must properly consider the differences in configuration between the exp:,nments and the actual MCR being evaluated for fire risk. In particular, the experimental configuration included placement of smoke detectors inside the cabinet in which the fire originated, as well as ENCLOSURE
2 an open cabinet door for that cabinet. in one case, failure to account for these configuration differences led to more than an order of magnitude underestimate in the conditional probability of forced MCR abandonment [ Reference 2). In addition, another study raises questions about MCR habitability due to room air temperature concoms
[ Reference 3].
Provide the detailed assumptions (including the assumed fire frequency, any frequency reduction factors, and the probability of MCR abandonment) used in analyzing thw MCR and the basis for these assumptions, in particular, if the probability of MCR abandonment is based on a probability distribution for the time required to suppress the fire, provide a justification of the basis for the selection of the parametric form of the probability distnbution and specify the data used in quantifying the distribution parameters.
3.
NUREG-1407, Section 4.2 and Appendix C.3, and Generic Letter (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 requests 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 potentiel 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 MCR) might lead to potential control systems vulnerabilities.
Given a fire at the CNS, control systems interactions could occur between the MCR, the remote shutdown panel (RSP), and other systems required for safe shutdown. Specific areas (sub issues) that have been identified as requiring attention in the resolution of this issue include:
3.1 Electricalindependence of the remote control systems: The primary concem of control systems interactions occurs at plants bat do not provide independent remoto (any location other than the MCR) control systems. The electrical independence of the RSP and the evaluation of the level of indication and control of remote control and monitoring circuits (e.g., water level control, reactor pressure) need to be assessed.
3.2 Loss of control equipment or power before transfer: The potential for loss of control power for certain control circuits as a result of HSs and/or blown fuses before transferring control from the MCR to remote shutdown locations (the RSP or any other location) needs to be assessed.
3.3 Spurious actuation of canponents leading to component damage, a LOCA, or ren interfacing systems LOCA: The spurious actuation of one or more safe-shutdown-related components as a result of fire-induced cable faults, HSs, or component failures leading to component damage, LOCA, or interfacing systems LOCA, prior to taking control from the RSP or any other location, 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.
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3 3.4 Total loss of system function: The potential for total loss of system function as a
'asult of fire-induced redundant component failures or electrical distribution system (power source) failures needs to be addressed.
Provide an evaluation of whether loss of contml power due to HSs and/or blown fuses J
could occur prior to transfening control to the remote shutdown locations and identify the core damage frequency (CDF) contribution of these types of failures, if these failures are screened for a fire area, please provide the basis for the screening. In addition, provide an evaluation of whether spurious actuation of components as a result of fire-induced individual cable faults, HS, or component failures could lead to loss of system function, a LOCA, or an interfacing systems LCCA prior to taking control from the remote shutdown locations (considering both' spurious starting and running of pumps as well as the spurious repositioning of valves).
' 4.
The iPEEE submittal indicated that fires initiated in control cabinets would be confined to the cabinet but evaluated the potential for damage to cabling above the cabinets by assuming a fire located on top of the cabinet. The heat release rates (HRRs) for cabinet fires, an iraportant heat transfer parameter in determining the potential for damage, are not specded in the submittal, in the Electric Power Research Institute (EPRI) Fire PRA implemordation Guide, test results for the control cabinet HRRs have been misinterpreted and have been inappropriately extrapolated. Cabinet HRRs as low as 65 BTU /sec are used in the Guide. In contrast, experimental work has developed HRRs ranging from 23 to 1171 BTU /sec.
Considering the range of HRRs that could be applicable to different control cabinet fires, and to ensure that cabinet fire areas are not prematurely screened out of the fire risk analysis, a HRR value in the mid-range of the currently available experimental data (e.g.,
550 BTU /sec) is considered to be appropriate and should be used for the analysis.
Discuss the HRRs used in the CNS IPEEE fire assessment of control cabinet fires and the changes in the fire assessment results if it is assumed that the HRR value for a cabinet fire is increased to 550 Btu /sec.
5.
The IPEEE fire assessment assumed that fires initiated in power distribution panels bested in the cable spreading room (CSR) would be retained in the cabinets. Although this assumption was used for all cabinet fires in the IPEEE submittal, the potential for damage to overhead cables was evaluated except for the cabinets in the CSR. In addition, it is not clear from the submittal whether there is a potential for fire propagation between adjacent panels. The EPRI Fire PRA Implementation Guide assumes that fire propagation to adjacent cabinets cannot occur, if the cabinets are separated by a double wall with an air gap or if the cabinet in which the fire originates has an open top. This can be an optimistic assumption for high-voltage cabinets since an explosive breakdown of the electrical conductors may breach the integrity of the cabinet and allow fire to propagate to combustibles located above the cabinet. For example, switchgear fires at Yankee-Rowe in 1984 and Oconee Unit 1 in 198g both resulted in fire damage outside 3
- the cubicles.
i s
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4 Provide the basis for the assumption that all fires in the electrical cabinets in the CSR will be retained in the CSR cabinets and a quantitative assessment of the impact of fire propagation from fires in these cabinets.
6.
Fire severity factors (SFs) were used in the analysis of many unscreened fire compartments in the CNS IPEEE. In the case of oil spiHs, values for SFs were apparently obtained through a limited data analysis as recommerided by the EPRI Fire PRA implementation Guide. The SFs for other fire types appear to have been subjectively defined, but no basis was provided. The SFs were also used in fire scenarios where fire suppression was credited. Since the potential for a large fire could be dependent upon one or more fire suppression agents, there appears to be a significant possibility that the use of a fire SF when fire suppression is modeled for a fire scenario (which determines a fire type and a fire size) may be inappropriate (a potential situation of double credits).
For the fire scenarios where both fire suppression and SFs were credited, if any, discuss the appropriateness of crediting both. Also, various SFs used in the IPEEE are found to be lower than typically observed SF estimates. Provide the basis for the lower estimates of SFs used in the IPEEE.
7.
The Fire-Induced Vulnerability Examination (FlVE) methodology requires that propagation through fire barriers that do not meet the fire compartment interactions analysis (FCIA) criteria be considered in the analysis. The IPEEE submittal does not always indicate if fire spread through unscreened compartment boundaries was accounted for in the IPEEE. Specifically, in the quantitative screening assessment of compartments with unscreened boundaries, it is unclear if components in the adjoining compartment (s) were damaged by the fire in a given compartment.
Discuss the impact of fire propagation through the unscreened boundaries on the quantitative screening assessment performed for the following fire compartments: 1 F, 28,2D,4A/4C/4D,58,2E, and 13D. If propagation-induced component damage through these boundaries was not considered, discuss the impact of inter-compartment propagation on the fire-induced CDF for applicable compartments.
8.
The unreliability estimates for automatic fire detection system (AFDS) and the suppression systems (SS) provided in the FIVE methodology were used in the IPEEE submittal. Section 4.5 of the IPEEE indicates that use of this data is appropriate because the AFDS and the SS at CNS are designed and maintained in accordance with appropriate industry standards, such as the National Fire Protection Association (NFPA).
However, an USNRC inspection [ Reference 4] has indicated that not all AFDS and the SS at CNS meet NFPA standards.
Discurs the impact of the above non-conformance to NFPA standards on the unreliability estimates of the AFDS and the SS and the resulting CDF for the fire compartments in the Reactor Building containing the nonconforming AFDS and the SS. Altematively, provide a basis for the lower unreliability estimates in the IPEEE of the AFDS and the SS which are based on NFPA standards.
5 9.
In general, the fire risk associated with a given compartment is composed of contributions from fixed and transient ignition sources. Neglect of either contribution can lead to an underestimate of the compartment's risk and, in some cases, to improper screening of fire scenarios. Further, the presence of transient combustibles can also impact the potential for fire propagation and component damage. The IPEEE appears to have I
eliminated transient ignition sources in some areas based on observations during the walkdowns. In addition, the fire risk assessment appears to have only considered the amounts, types, and locations of transient combustibles identified during the plant walkdowns. The assessment of the transient fire risk based on a one-time observation is questionable.
Provide an evaluation of the fire risk in unscreened fire compartments from transient g
combustibles and ignition sources taking into account the ignition sources that would be allowed in the compartments. This evaluation should include the amount of transient combustibles that may be present in the compartments at various times, including the potential for violation of transient combustible controls which have occurred at the CNS (Reference 4).
10.
The IPEEE submittal does not indicate whether cables in the CNS are institute of Electrical and Electronic Engineers (IEEE)-383 qualified cables. The submittal lists damage criteria for both qualified and non-qualified cables implying that both exist at the CNS; however, insufficient information was provided to determine if unqualified cable fires were included in the initiating event (IE) frequencies used in the quantitative screening. Furthermore, the fire modeling did not include unqualified cable-initiated fires.
Discuss if qualified IEEE-383 cables are currently used at the CNS and if fires initiated by unqualified cables were included in the fire IE frequencies used in the screening fire assessment. If unqualified cables were excluded from the fire IE frequencies, justify this exclusion and discuss the impact of the exclusion on the fire-induced CDF. Provide the types of cables treated as targets in the detailed fire assessments performed for each of the unscreened compartments.
11.
The heat loss factor (HLF) 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 (HGL). In FIVE, the HLF is modeled as being inversely related to the amount of heat required to cause a given temperature rise. Thus, for example, a larger HLF 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 HLF is unrealistically high, fire scenarios can be improperly screened out. Figure 1 of Reference 5 provides a representative example of how HGL temperature predictions can change assuming different HLFs. Please note that: 1) the curves are computed for a 1000 kW fire in a 10m x Sm x 4m compartment with a forced ventilation rate of 1130 cfm; I
- 2) the FIVE-recommended damage temperature is 700'F for qualified cable and 450*F for unqualified cable; and,3) the Society for Fire Protection Engineers (SFPE) curve in Figure 1 is generated from a correlation provided in the SFPE Handbook [ Reference 5).
s
6
. Based on evidence provided by a 1982 paper by Cooper et al. [ Reference 6], the EPRI Fire PRA implementation Guide recommends a HLF of 0.g4 for fires with duratens greater than five minutes and 0.85 for " exposure fires away from a wall and quicidy developing HGLs." However, as a general statement, this appears to be a misinterpretation of the results. Reference 6, which documents the results of multi-compartment fire experiments, stated that the higher HLFs are associated with the movement of the HGL from the buming compartment to adjacent, cooler compartments.
Eariier in the experiments, where the HGL is limited to the buming compartment, Reference 6 reports much lower HLFs (on the order of 0.51 to 0.74).
These lower HLFs are more appropriate when analyzing a single compartment fire. in summary, (a) HGL predictions are very sensitive to the assumed value of the HLF; and (b) large HLFs cannot be justified for single-room scenarios based on the information referenced in the EPRI Fire PRA implementation Guide.
For each fire scenario where the HGL temperature was calculated, specify the HLF value used in the analysis, in light of the preceding discussion, either: (a) Justify the value used and discuss its effect on the identification of fire-induced vulnerabilities, or (b) repeat the analysis using a more justifiable value and provide the resulting change in scenario contribution to CDF.
12.
The IPEEE submittal indicated that non-Appendix R (Title 10 CFR, Part 50) systems were credited in protecting against fires in the upper elevations of the Reactor Building (RB) without an explicit examination of the associated cabling. This was done based on discussions with plant personnel that the non-Appendix R equipment (identified as balance-of-plant equipment) is located in the Turt>ine Building (TB) and intake structure.
The IPEEE submittal indicated that the non-Appendix R systems credited in the CNS IPEEE consisted of offsite power system and the main condenser system. Operation of the main condenser as a decay heat removal system requires operation of many support systems including cooling water systems, non-Class IE power source, and instrument air.
In addition, successful operation also requires that the main steam isolation valves (MSIV) remain open.
Provide the basis that applicable support systems required for operation of the main condenser would not be impacted by fires in the upper RB. In particular, verify that fires in the RB compartments would not impact systems whose failure would result in an MSIV closure signal, and would not affect CNS operator ability to reopen the MSIVs, if closed, due to fires.
13.
The IPEEE fire assessment of the control room indicated that the analysis of fires in two panels were treated uniquely. One control panel (9-3) has no intemal barriers, and fires l
that were postulated were assumed to impact only portions (or sections) of the panel.
Another panel (Board C) contains partial intomal barriers that were assumed to be effective in preventing fire propagation from one section to another section in the panel.
Crediting partial barriers and/or its equivalent as part of prevention of fire growth within l
the panel is questionable in a fire risk analysis.
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Provide an assessment of the CDF from fires in these two panels that could propagate throughout the panel.
B. Seismic events 1.
System analysis for Safe Shutdown Equipment List (SSEL) development is discussed in Section 3.1.2 of the IPEEE submittal. For success path and system selection, EPRI NP-l 6041-SL states that "in general, the selected path for performing the safety funchons to shutdown the reactor will be the one consisting of the front-line systems (and their I
necessary support systems) that were provided as a 'first line of defense', and designed to respond automatically (at least in the short time during and after the seismic margin earthquake to the types of transients and/or accidents that might be induced by a margin earthquake." Based on this industry criterion, the high pressure injection systems (e.g.,
the high pressure coolant injection (HPCI) system and the reactor core isolation cooling (RCIC) system) seem to provide a choice for coolant injection to the vessel. However, the above listed high pressure injection systems are not included in the CNS SSEL, and low pressure coolant systems (e.g., the low pressure coolant injection (LPCI) system and the core spray (CS) system) are used for both success paths. Manual reactor pressure vessel (RPV) depressurization is therefore required for both paths, and consequently, the demands on the depressurization system and operator actions are significant.
RPV depressurization is provided at CNS by the operation of the safety relief valves j
(SRVs) using pneumatic power provided by accumulators. Since the plant
-l nitrogen / instrument air systems, which supply the accumulators, are not included in_the SSEL, accumulators are the only pneumatic power source for the SRVs. Hence, their performance (applicable current design requirements and unreliability per demand
~
considerations) for transient and small LOCA events under seismic margin earthquake (SME) conditions for a mission time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> needs to be evaluated and discussed.
According to the IPEEE submittal, successful RPV depressurization requires the operation of three SRVs, and the probability of " operator failure to depressurize with 3 SRVs"is 4.2E-2. Failure to depressurize the RPV in a timely fashion may therefore be a significant contributor to failing to place the plant under a safe shutdown condition following a seismic margin earthquake (SME) event.
Based on the above considerations:
a.
Discuss how the EPRI NP-6041-SL criterion was applied for selecting appropriate systems for the CNS SSEL, quoted above, in the CNS IPEEE discuss'on. Provide also the basis for not including the HPCI system and the RCIC system in the CNS SSEL (other than the unreliability considerations of these systems based on industry exponence).
b.
Based on CNS procedures, describe the expected operator actions following an SME event. Discuss in detail the operator actions and their failure probabilities following an SME event for RPV depressurization, including the ability of the SRV accumulators to provide pneumatic power sufficient for SRV operation for a mission time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
e 8
c.
List all major equipment for the HPCI system and the RCIC system and desenbe any weak links in these systems following an SME event.
2.
Subtrut better copies of Figures 3.1.2 and 3.1.3 of the IPEEE submittal, to facilitate timely completion of the staffs review of the seismic portion of the IPEEE.
3.
What are the findings of the resolution of USl A-46, and the IPEEE-identifed weak equipment, if any, which have a seismic capacity below the 0.3g peak ground l
acceleration (pga) review level earthquake (RLE)? Discuss the estimated overall high i
confidence of low probability of failure (HCLPF) capacity of the weakest plant equipment based on completion of the USl A-46 resolution program and the CNS IPEEE program.
C. Hiah winds. floods. and other extemal events (HFOs) 1.
Section 5.4.1 of the CNS IPEEE provided a discussion on aircraft hazards from nearby airports and its impact on plant safety. It is the staffs understanding that, as par 1 of this evaluation, the licensee has made use of old air aviation data (e.g.,1974 data). NUREG-1407 guidance requested that the IPfdE should make use of the recent historical aviation data and significant changes in air traffic problems, if any, as part of the aircraft
- crash event frequency estimation process. Discuss whether the IPEEE has made use of recent data (i.e., data for the period between the OL date of 01/18/1974 and 12/1990) on significant changes in air traffic pattom (e.g., increase in number of landings and takeoffs per year at nearby airports).
2.
Section 5.5.1 provides a discussion on safety problems related to lightning hazards at the CNS. Discuss the rrsle and importance of station batteries and any other emergency power equipment needed in shutting down the CNS during severe lightning events.
References:
1.
J. Chavez, et al., "An Experimental Investigation of Intemally Ignited Fires in Nuclear Power Plant Cabinets, Part Il-Room Effects Tests," NUREG/CR-4527N2, October 1988.
2.
J. Lambright, et al.,"A Review of Fire PRA Requentification Studies Reported in NSAC/181," prepared for the United States Nuclear Regulatory Commission, April 1994.
3.
J. Usher and J. Boccio, " Fire Environment Determination in the LaSalle Nuclear Power Plant Control Room," NUREG/CR-5037, prepared for the United States Nuclear Regulatory Commission, October 1987.
4.
" Cooper Nuclear Station NRC Inspection Report 50-298/96-25," United States Nuclear Regulatory Commission, March 1997.
5.
P.J. DiNenno, et al, eds., "SFPE Handbook of Fire Protection Engineering," 2nd Edition, National Fire Protection Association, p. 3-140,1995.
l 9
6.
L. Y. Cooper, M. Harkleroad, J. Quintiere, W. Rinkinen, "An Experimental Study of Upper Hot layer Stratification in Full-Scale Multi room Fire Scenarios," ASME Joumal of Heat Transfer, j,M, 741-749, November 1982.
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