ML20203A913

Forwards RAI Re IPEEE Submittal Re Seismic,Fire,High Winds, Floods & Other External Events.Response Requested within 45 Days

ML20203A913
Person / Time
Site:	Crystal River
Issue date:	02/12/1998
From:	Raghavan L NRC (Affiliation Not Assigned)
To:	Richard Anderson FLORIDA POWER CORP.
References
TAC-M83612, NUDOCS 9802240151
Download: ML20203A913 (11)

Text

_ _ _ _ _ _ _ - _ _ _ _ _ _ _

February 12, 1998 Mr. Roy A. Anderson Senior Vice Presloent Nuclear Operation:

Florida Power Corporation ATTN: Manager, Nuclear Licensing Crystal River Energy Complex (SA2A) 15760 W Power Line Street Crystal River, Florida 34428-6708

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE CRYSTAL RIVER 3 NUCLEAR PLANT. INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (TAC NO. M83612)

Dear Mr. Anderson:

Based on our ongoing review of the Crystal River Unit 3 (CR3) Individual Plant Examination of External Events (IPEEE) submittal, the U.S. Nuclear Regulatory Commission (NRC) staff has determined that it requires additionalinformation to complete its review. This request for additional information (RAI) are related to the seismic, fire, high winds, floods and other external events (HFO) analyses in the IPEEE. The enclosure includes details for the RAl. The RAI v!as developed by the NRC staff and our contractors, Brookhaven National and Sandia National Llaboratories. The RAI has also been reviewed by the staff's Senior Review Board consisting of NRC personnel and cor.sultants with probablistic risk assessment expertise in external events.

We request your response within 45 days. Should you have any questions related to this letter or the enclosed RAI, please contact me at (301) 415-1471.

i Sincerely, L. Rahhavk' hen o o$ect da' nager Project Directorate 11-3 Division of Reac$. Projects - 1/II Do:ket No. 50-302

Enclosure:

RAI cc w/ enclosure: See next page hfD/

Distribution Docket File B. Hardin ACRS J. Zwolinski PUBLIC A. Rubin M. Cunningham L. Marsh CR 3 r/f OGC J. Jaudon, Region il To receive a copy of this document, indicate in the box: "C" = Copy without attachment / enclosure

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• T February 12, 1998 Mr. Roy A. Anderson Senior Vice President Nuclear Operations Florida Power Corporation ATTN: Manager, Nuclear Licensing Crysta! River Energy Complex (SA2A) 15760 W Power Lina Street Crystal River, Florida 34428-6708

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE CRYSTAL RIVER 3 NUCLEAR PLANT, INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (TAC NO. M83612)

Dear Mr. Anderson:

Based on our ongoing review of the Crystal River Unit 3 (CR3) Individual Plant Examination of Exterr,al Events (IPEEE)',obmittal, the U.S. Nuclear Regulatory Commission (NRC) staff has determined that it requhs additionalinformation to complete its review. This request for additionalinformatior' (RAI) are related to the seismic, fire, high winds, floods and other extemal events (HFO) analy',es ;n the IPEEE. The enclosure includes details for the RAl. The RAl was developeC by the NRC staff and our contractors, Brookhaven National and Sandia National Llaboratories. The RAI has also been reviewed by the staff's Senior Review Board consisting of NRC personnel and consultants with probablistic risk assessment expertise in extemal events.

We request your response within 45 days. Should you have any questions related to this letter or the enclosed ?Al, please contact me at (301) 415-1471.

Sincerely, L. RaghavN en o olect a' nager Project Directorate Il-3 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation Docket No. 50-302

Enclosure:

RAI cc w/ enclosure: See next page Distribution Docket File B. Hardin ACRS J. Zwolinski PUBLIC A. Rubin M. Cunningham L. Marsh CR-3 r/f OGC J. Jaudon, Region 11 To receive a copy of this document, indicate in the box: "C" = Copy without attachment / enclosure "E" = Copy wJ attachment / enclosure "N" = No copy 0FFICE PDl! 3/ Pet l l PDl!*3/LA lTO PDil 3/D l C. )

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NAME LJaghavan:cw 11 s 8Clayton W,.,,-

FHebdon "sN DAff 02/tl /98 02/'. /98 02/a t/98 02! /98 02/ /98 OFFICIAL RECORD COPY DOCUMENT NAME: G:\\ CRYSTAL \\83612.RAI

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t UNITED STATES s

j NUCL EAR REGULATORY COMMISSION 2

WASHINGTON, D.C. 30666-0001 February 12, 1998 Mr. Roy A. Anderson Senior Vice President Nuclear Operations Florida Power Corporation ATTN: Manager, Nuclear Licensing C ystal River Energy Complex (SA2A) 15760 W Power Line Street Crystal River, Florida 34428-6708

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE CRYSTAL RIVER 3 NUCLEAR PLANT, INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (TAC NO. MB3612)

Dear Mr. Anderson:

Cased on our ongoing review of the Crystal River Unit 3 (CR3) Individual Plant Examination of Extemal Events (IPEEE) submittal, the U.S. Nuclear Regulatory Commission (NRC) staff has determined that it requires additionalinformation to complete its review. This request for l

additionalinformation (RAl) are related to the seismic, fire, high winds, floods and other extemal l

events (HFO) analyses in the IPEEE. The enclosure includes details for the RAl. The RAI was developed by the NRC staff and our contractors, Brookhaven National and Sandia National Llaboratories. The RAI has also been reviewed by the staff's Senior Review Board consisting of NRC personnel and consultants with probablistic risk assessment expertise in extemal events.

We request your response within 45 days. Should you have any questions related to this letter or the enclosed RAl, please contact me at (301) 415-1471.

i Sincerely, 4

L.Raghavan, enior Project Manager Project Directorate ll 3 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation Docket No. 50-302

Enclosure:

RAI cc w/ enclosure: See next page 1

W

Mr. Roy A. Anderson CRYSTAL RIVER UNIT NO. 3 Florida Power Corporation cc:

Chairman Mr. R. Alexander Glenn Board of County Commissioners Corporate Counsel Citrus County Florida Power Corporation 110 North Apopka Avenue MAC-ASA Ivemess, Florida 34450-4245 P.O. Box 14042 St. Petysburg, Florida 33733-4042 Mr. Robert E. Grazio, Director Nuclear Regulatory Affairs (SA2A)

Mr. Charles G. Pardee, Director Florida Power Corporation Nuclear Plant Operations (NA2C)

Crystal River Energy Complex Florida Power Corporation 15760 W. Power Line Street Crystal River Energy Complex Crystal River, Florida 34428-6708 15760 W. Power Line Street Crystal River, Florida 34428-6708 Senior Resident inspector Crystal River Unit 3 Mr. Bruce J. Hickle Director U.S. Nuclear Regulatory Commission Director, Restart (NA2C) 6745 N. Tallahassee Road Florida Power Corporation Crystal River, Florida 34428 Crystal River Energy Complex 15760 W. Power Line Street Mr. John P. Cowan Crystal River, Florida 34428-6708 Vice Presiderit, Nuclear Production (NA2E)

Mr. Robert B. Borsum Florida Power Corporation Framatome Technologies Inc.

Crystal River Energy Complex 1700 Rockville Pike, Suite 525 15760 W. Power Line Street l

Rockville, Maryland 20852 Crystal River, Florida 34428-6708 Mr. B;ll Passetti Mr. James S. Baumstark Office of Radiation Control Director, Quality Programs (SA2C)

Department of Health and Florida Power Corporation Rehabilitative Services Crystal River Energy Complex 1317 Winewood Blvd.

15760 W. Power Line Street Tallahassee, Florida 32399-0700 Crystal River, Florida 34428-6708 Attomey Gene;al Regional Adrninistrator, Region 11 Department of Legal Affairs U.S. Nuclear Regulatory Cornmission The Capitol 61 Forsyth Street, SW., Suite 23T85 Tallahassee, Florida 32304 Atlanta, GA 30303 3415 Mr. Joe Myers, Director Mr. Kerry Landis Division of Emergency Preparedness U.S. Nuclear Regulatory Commission Department of Community Affairs 61 Forsyth Street, SW., Suite 23T85 2740 Centerview Drive At!snta, GA 30303-3415 Tallahassee, Florida 32399-2100 l

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9 CRYSTAL RIVER UNIT 3 IPEEE Requests for Additional Information A.

Selsmic

1. For a reduced scope seismic IPEEE (applicable to Crystal River 3), Generic Letter 88-20, Supplement 4 requests that a walkdown be conducted to verify the seismic design basis and to identify and resolve potential seismic interactions for all items included in the safe shutdown equipment list (SSOL). The SSEL is to be compiled based on the selection of a preferred and attemate safe shutdown path, one of which considers a selsmically induced small break LOCA.

Please identify, by specific reference to applicable USI A 46 program results, how the objectives of the selsmic IPEEE have been addressed for Crystal River 3.

2. In a safety evaluation report dated February 9,1995, the staff concluded that the approach to achieve and maintain hot shutdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during a seismic event (seismic-induced transient) was acceptable for purposes of resolving Unresolved Safety Issue A-46. A selsrr.le-induced small break LOC 6. is normally considered in IPEEE esatuations.

Was a selsmic Induced small break LOCA consideredin the Crystal River A 46 orIPEEE seismic evaluation? If not, please provide thejustification for omitting this evaluation. If this event was considered, please provide the walkdown findings and list of those components assocInted with the hot shutdown success paths. Also, please provide success pcth logic diagrams that illustrate the basis for their selection.

3. Please provide a IIst of components that were (would be) considc: edin the containment performance walkdown for Crystal River, describe the basis for developing this list, and provide the walkdown results.

4. Please provide a discussion of the process for addressing seismic fire interaction concems at Crystal River. List all walkdown related findings pertaining to potential seismic. fire Interactions involving IPEEE components.

5. ForanyInstances of selsmic interactions concems related to block walls that were noted by the selsmic review team, please provide calculations that demonstrate adequate seismic capability of these walls.

Enclosure

8.

Fire

1. The heat 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 l

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 inversely related to the amour,t of heat req. sired 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 unroslistically 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 with a forced ventilation rate of 1130 cfm; 2) the FIVE recommended damage l

temperature for qualified cable is 700'F for qualified cable and 450'F for unquallT.ed cable; and,

3) the SFPE curve in the figure is generated from a correlation provided in the Society for Fire l

Protection Engineers Handbook [1.1].

Based on evidence provided by a 1982 paper by Cooper et al. [1.2], the EPRI Fire PRA i

implementation Guide recommends a heat loss factor of 0.94 for fires with durations greater than i

five minutes and 0.85 for " exposure fires away from a wall and quickiy developing hot gas layers."

l However, as a general statement, this appears to be a mis nterpretation of the results. Reference

[1.2], which documents the results of multi-compartment fire experiments, states that the higher heat loss factors are essociated with the movement of the hot gas layer from the buming i

compartment to adjacent, cooler compartments. Earlier in the experiments, where the hot gas layer is limited to the burning compartment, ReTerence [1.2) 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. In summary, (a) hot gas layer predictions are very sensitive to the assumed value of the heat loss facto;, md (b) large heat loss factors cannot be justified for single-room scenarios based on the information referenced in the EPRI Fire FRA /mplementation Guide, i

For each scer.ario where the hot gas layer temperature was calculated, please specify the heat loss factor value used in the analysis, in light of the pieceding discussion, please either:

a) Justify the value used and discuss its effect on the identification of fire vulnerabilities, or i

j b) repeat the analysis using a more justifiable value and provide the resulting change in j

scenario contribution to core damage frequency l

2

-l

_ _ - _ _ _. _=.

l' Time Temperature Curves 900 m,

1 SM e H.F = 0.70

+ H.F = 0.85 E 600.

+ H.F = 0.94 x H.F = 0.99 j

g 400.

300.

l l

200 -

100-o m ix x x

• x

• x :* :* * * *

• bkk$$b$$$k0$$$$

j Time (s)

Figure 1: Sensitivity of the hot gas layer temperature predictions to the assumed heat loss factor

References:

1.1.

'SFPE Handbook of Fire Protection Engineering,2nd Edition

• P.J. DiNenno, et al, eds.,

National Fire Protection Association,1995, p 3 -140.

1.2.

L.Y. Cooper, M. Harkieroad, J. Quintierm, W. Rinkinen, "An Experimental Study of Upper Het Layer Stratification in Full-Scale Multiroom Fire Scenarios," Joumal of Heat Transfer, v. 104, 7419 (November 1982).

2. In the EPRI Fire *RA implementation Guide, test results for the control cabinet heat release rate have been misinterpreted and have been inappropriately extrapolated. Cabinet heat release rates as low as 65 Btu /see are used in the Guide. In contrast, experimental work has developed heat release rates ranging from 23 to 1171 Blu/sec.

Considering the range of heat release rates that could be applicable to different control cabinet fires, and to ensure that cabinet fire areas are not prematurely screened out of the analysis, a heat release rate in the mid-range of the currently available experimental data (e.g.,550 Btu /sec) should be used for the analysis. !

l

t Discuss the heat release rates used in your assessment cf control cabinet fires. Please provide a discussion of changes in the IPEEE fire assessmene results If it is assumed that the heat release from a cabinet fire Is increased to $50 BtWs.

3.

The guidance for fire studies, such as that provided in the FIVE methodology, describes procedures for determining boundaries of multi-compartment fires. The guidance is clear on the need to initially include previously screened fire compartments. Fire-induced failure of barriers in screened fire zones may result in risk significant failures by allowing fire propagation into risk-significant zones. Such failures may occur in both active and passive barriers. Fire zones with the potential for barrier failure include those with high combustible loading, unrated fire barriers, or no fire protection. High hazard areas, such as those in the Turbine Buildir g, are typically of interest in this regard.

From the discussions of multi-compartment fire scenarios (Section 4.5.7) provided, it is not clear that such propagating fires have been considered. It does not appear that previously screened fire compartments have been included in the analysis. The contribution to risk from mult. compartment fire scenarios, including the previously screened fire zones, needs to be ast.essed for risk significance, and quantified in the final result if it is fourid to be significant.

Please reexamine the multi-compartment fire scenario definItlon, including ti,e previously screened compartments, to encure that all significant fire scenarios have been evaluated.

Determine the contribution to core damage frequency from any new fire scsnarloc Identitled.

4.

NUREG 1407, Section 4.2 and Appendix C, ar:d GL 8S-20, Supplement 4, request that documentation be submitted with the IDEEE submittal with regard to the 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 potentialimprovements 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 MCR) might lead to potential control systems vulneraoilities. 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 Iesolution of this issue include:

(a) Electricalindependence of the remote shutdown control systeins: The primary concem of control systems interactions occurs at plants that do not provide independent remete shutdown control systems. The electricalindependence of the remote shutdown panel and the evaluation of the level of indication ond control of remote shutdown control and monitoring circuits need to be assessed..

(b) Loss of control equipment or power before transfer: The potential for loss of control power for certain control circuits as a result of hot shorts and/or blown fuses before transferring control from the MCR to remote shutdown locations needs to be assessed.

(c) 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 fite-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.

(d) Totalloss of system function: The potential for totalloss of system function as a result of fire.

Induced redundan' component failures or electrical distribution system (power source) failure needs to be addressed.

Please describe your remote shutdown cepability, including the nature and location of the shutdown station (s), as well as the types of control actions which can be taken from the rennte panel (s). Describe how yourpmcedures provide for transfer of control to the remote i

station (s). Provide an evaluation of whetherloss of contrelpower due to hot shorts and/or

\\

blown fuses could occur prior to transferring control to the remote shutdown location and identify the risk contribution of these types of fallures (if these failures are screened, please provide the basis for the screening). Finally, provide en evaluation of whether spurious actuation of components as a result of fire-Induced cable faults, hot shorts, or component failures could lead to component damage, a LOCA, or en Interfacing systems LOCA prior to taking control from the remote shutdown panel (considering both spurious starting and running of pumps as well as the spurious repositioning of valves).

5. From the submittalit can be inferred that hot shorts have not been conLdered as a failure mode for control or instrumentation cables, in particular, hot short considerations should include the treatment of conductor to-conductor shorts within a given cable. Hot shorts in control cables can simulate the closing of control switches leading, for example, to the repositioning of valves, spurious operation of motors and pumps, or the shutdown of operating equipment. These types of faults might, for example, lead to a LOCA, diversion of flow within various plant systes, deadheading and failure of important pumns, premature or undesirable switching of pe ; suction sources, or undesirable equipment weratior A For MCR abandonment scenarios, such spurious operations and actions may not be indicated at the remote shutdown panel (s), may not be directly recoverable from remote shutdown locations, or may lead to the loss of remote shutdown capabihty (e.g. through loss of RSP power sources). In instrumentation circuits hot shorts may cause misleading plant readings potentiallyleading to inappropriate control actions or generation of actuation signals for emergency safeguard features.

Discuas to what extent these issues have been consideredin the IPEEE. If they have not been considered, please provide an assessment of how Inclusion of potential hot shorts would Impact the quantification of fire risk scenerlos In the IPEEE.

l

.s.

L4

6. In the analysis of containment performance it was stated that 'the CR 3 fire zones with the largest contribution to the overall core damage risk (i.e., those listed in Table 5.5-4) were examined to see if any might border on the containment, thereby possibly causing a bum through of an electrical penetration, a seal on a containment hatch, or some other containment penetration." (Table 5.5-4 lists generic manual suppression time lines. It is assumed that the submittalintended to refer to Table 5.5-7.) The resulting assessment is considered incomplete. It was the intent of NUREG 1407 that the IPEEE evaluate the potential for unique, fire-induced challenges to containment isolation and containment bypass scenarios. The licensee's approach ignores the effect of het shorts on containment isolation valves or the potential for a fire induced interfacing LOCA. Such scenarios would not be limited to areas adjacent to containment.

Please resvaluate the potential for fire Induced degradation of containment performance in the areas discussed above. Identify any unique vulnerabilities to fire-Induced events, as compared to those identified in the IPE Internal events analysis.

7. Fires in the main control room (MCR) are potentially risk significant because they can cause l&C failures (e g., loss of signals or spurious signals) for multiple redundant divisions, and because they can force control room abandonment. Although data from two experiments conceming the timing of smoke-induced, forced control room abandonment are available [7.1), the analysis based on the cata must properly consider the differences in configuration between the experiments and the actual control room being evaluated for fire risk. In particular, the experimental configuration included l

placement of smoke detectors inside the cabinet in which the fire originated, as well as an open l

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 control room fire abandonment [7.2). In addition, another study raises questions about control room habitability due to room air temperature concems [7.3).

Please provide the detalled assumptions (including the assumed fire frequency, any frequency reduction factors, and the probability of abandonment) used in analyzing fires In the main controlroom (MCR) andjustifications for these assumptions. In particular, if the probability of abandonment is based on a probability distribution for the time required to suppress the fire, pleasefustify the parametric form of the distribution and specify the data used to quantify the distribution parameters.

References:

7.1 J. Chavez, et al.,"An Experimental Investigation of Intemally ignited Fires in Nuclear Power Plant Cabinets, Part Il-Room Effects Tests," NUREG/CR-4527/V2, October 1988.

7.2 J. Lambright, et al.,"A Review of Fire PRA Requantification Studies Reported in NSAC/181,"

prepared for the United States Nuclear Regulatory Commission, April 1994.

7.3 J. Usher and J. Boccio, " Fire Environment Determination in the LcSalle Nuclear Power Plant Control Room," NUREG/CR 5037, prepared for the United States Nuclear Regulatory Commission, October 1987. ~

8. The investigation of Jources of fixed combustibles including hydrogen gas storage vessels or piping (e.g., hydrogen pipe lines and the poter'tial effects of their rupture) is included in the EPRI FIVE methodology (page 10.6 4) as a part of the walkdown verification evaluation. Although hydrogen tanks were identified as fixed ignition sources, '.he potential effects of hydrogen line ruptures were not discussed. Leaks or breaks in hydrogen supply piping could result in the accumulation of a combustible mixture of air and hydrogen in vital areas, resulting in a fire and/or explosion that could damage vital safety related systems in the plant.

Please provide the basis for screening out hydrogen IIne ruptures, or Identify any unique vulnerabilltles due to such IInes and means forprotecting against fires assocInted with their fallure.

9. The Intermediate Building contains equipment required for safe shutdown. The only area in the intermediate Building which was not reported to be screened out was the turb!ne driven emergency feedwater (EFW) pump, Penetration Area, Fan Room (Fire Zone IB 95-200C).

Please provide the basis for screening of fire areas In the intermediate Building.

C. High Winds, Floods, and Other External Events (HFOs)

1. Section 5.2.3 of NUREG 1407 states that a plant walkdown should be performed concentrating on outdoor faciltties that could be affected by high winds, onsite storage of hazardous materials, and offsite developments. The submittat does not indicate whether any plant walkdowns were performed in the evaluation of HFO (e.g., to identify if there were any cases where failure of Non-Category 1 structures during a storm could significantly damage Category 1 structures).

Please describe any walkdowns that were performed, and brlefly summarize the results.

2. Section 5.2.4 of NUREG-1407 states that in performing the IPEEE, the licensee should identify any significant changes in plant equipment or procedures that have occurred since the issuance of the plant operating license. The submittal did not indicate whether or not there hEd been any such changes that could impact on the results of the IPEEE HFO evaluation.

l Pleese confirm whether or not significant changes have occurred that could affect the HFO assessment, and, if appilcable, discuss theirimpact.

3.

As noted in NUREG-1407, Section 2.4, the latest piobable maximum precipitation criteria publiNd by the National Weather Service call for higher rainfall intensities over shorter time intervm and smaller areas than have previously been considered; this could result in higher site flooding levels, and greater roof ponding levels.

Please assess the effects of applying these now criteria to Crystal River. Additional Information is given in Generic Letter 89 22 (attached).

7-