ML17191B020

Forwards RAI Re IPEEE for Plant,Units 2 & 3,per 971230 Submittal.Response Requested within 60 Days of Receipt of Ltr

ML17191B020
Person / Time
Site:	Dresden
Issue date:	12/14/1998
From:	Rossbach L NRC (Affiliation Not Assigned)
To:	Kingsley O COMMONWEALTH EDISON CO.
References
TAC-M83616, TAC-M83617, NUDOCS 9812160195
Download: ML17191B020 (14)

Text

Mr. Oliver D. Kingsley, President Nuclear Generation Group

• Commonwealth Edison Company Executive Towers West Ill 1400 Opus Place, Suite 500 Downers Grove, IL 60515 December 14, 1998

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE) FOR DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 (TAC NOS. M83616 AND M83617)

Dear Mr. Kingsley:

By letter dated December 30,

• 1997, Commonwealth Edison Company (Com Ed) submitted the results of the Individual Plant Examination of External Events (IPEEE) for Dresden Nuclear Power Station, Units 2 and 3. The staff has identified additional information that is needed in order for the staff to complete their review of this submittal. The additional information needed relates to the IPEEE analyses of seismic and fire events. We do not have any questions related to high winds, floods, or other external events. This request for additional information (RAI) was reviewed by the "Senior Review Board" comprised of NRC staff and consultants with probabilistic risk assessment expertise in external events. These questions were discussed with members of your staff on November 17, 1998. Please provide the information requested within 60 days of receipt of this letter. Should your staff have any questions about this request, please contact me at (301) 415-2863.

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UNITED STATES NUCLEAR REGULATORY C.OMMISSION Mr. Oliver D. Kingsley, President Nuclear Generation Group Commonwealth Edison Company Executive Towers West Ill 1400 Opus Place, Suite 500 Downers Grove, IL 60515 WASHINGTON, D.C. 20555-0001 uece111ber 14, 1998

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING TH.E INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE) FOR DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 (TAC NOS. M83616 AND M83617)

Dear Mr. Kingsley:

By letter dated December 30, 1997, Commonwealth Edison Company (ComEd) submitted the results of the Individual Plarit Examination of External Events (IPEEE) for Dresden Nuclear Power Station, Units 2 and 3. The staff has identified additional information that is needed in order for the staff to complete their review of this submittal. The additional information needed relates to the IPEEE analyses of seismic and fire events. We do not have any questions related to high winds, floods, or other external events. This request for additional information (RAI) was reviewed by the "Senior Review Board" comprised of NRC staff and consultants with probabilistic risk assessment expertise in external events. These question$ were discussed with members of your staff on November 17, 1998. Please provide the information requested within 60 days of receipt of this letter. Should your staff have any questions about this request, please contact me at (301) 415-2863.

Docket Nos. 50-237 and 50-249

Enclosure:

RAI cc w/encl: See next page Sincerely, Lawrence W. Rossbach, Project Manager Project Directorate 111-2

• Division of Reactor Projects - Ill/IV Office of Nuclear Reactor Regulation

0. Kingsley Commonwealth Edison Company cc:

Michael I. Miller, Esquire Sidley and Austin One First National Plaza Chicago, Illinois 60603 Commonwealth Edison Company Site Vice President - Dresden 6500 N. Dresden Road Morris, Illinois 60450-9765 Commonwealth Edison Company Dresden Station Manager 6500 N. Dresden Road Morris, Illinois 60450-9765 U.S. Nuclear Regulatory Commission Dresden Resident Inspectors Office 6500 N: Dresden Road Morris, Illinois 60450-9766 Regional Administrator U.S. NRC, Region Ill 801 Warrenville Road Lisle, Illinois 60532-4351 Illinois Department of Nuclear Safety Office of Nuclear Facility Safety 1035 Outer Park Drive Springfield, Illinois 62704 Chairman Grundy County Board Administration Building 1320 Union Street Morris, Illinois 60450 Document Control Desk-Licensing Commonwealth Edison Company 1400 Opus Place, Suite 400 Downers Grove, Illinois 60515 Commonwealth Edison Company Reg~ Assurance Supervisor - Dresden 6500 N. Dresden Road Morris, Illinois 60450-9765 Dresden Nuclear Power Station Units 2 and 3 Mr. David Helwig Senior Vice President Commonwealth Edison Company Executive Towers West Ill 1400 Opus Place, Suite 900 Downers Grove, IL 60515 Mr. Gene H. Stanley PWR's Vice President Commonwealth Edison Company Executive Towers West Ill 1400 Opus Place, Suite 900 Downers Grove, IL 60515 Mr. Steve Perry BWR's Vice Piresident Commonwealth Edison Company.

Executive Towers West Ill 1400 Opus Place, Suite 900 Downers Grove, IL 60515 Mr. Dennis Farrar Regulatory Services Manager Commonwealth Edison Company Executive Towe rs West II I 1400 Opus Place, Suite 500 Downers Grove, IL 60515 Ms. Irene Johnson, L.icensing Director Nuclear Regulatory Services Commonwealth Edison Company Executive Towers West Ill 1400 Opus Place, Suite 500 Downers Grove, IL 60515 Mr. Michael J. Wallace Senior Vice President Commonwealth Edison Company Executive Towers West Ill 1400 Opus Place, Suite 900 Downers Grove, IL 60515

DRESDEN UNITS 2 AND 3 REQUEST FOR ADDITIONAL INFORMATION ON IPEEE SUBMITTAL Seismic

1.

In Section 3.4 (cribhouse masonry walls), Section 3.6 (IPEEE-only relays), and Section 3.8 (Open Items Pending Resolution) of the submittal, ComEd identifies items that had not been evaluated and had been classified as outliers, for tracking purposes.

Consequently, the IPEEE submittal is incomplete, and the plant high confidence, low probability of failure (HCLPF) is undetermined at this time.

In addition, the submittal is somewhat confusing with respect to disposition of (1) identified items with HCLPF capacity< 0.2g peak ground acceleration (PGA) (e.g., cable tray supports), which are designated "potential design basis issuesn, and (2) items with HCLPF capacity> 0.2g but< 0.3g PGA. These items are listed on p.1-3 (and again on p.3-4) of the submittal.*

Please provide the following information in order to establish the plant HCLPF:

(a)

Results of the evaluation for the cribhouse masonry walls, IPEEE-only relays, and "Open Items Pending Resolutionn.

(b)

What; if any, plant improvements have been implemented or are scheduled (please provide schedule) for the items listed in (a) above.

(c)

HCLPF capacities, after any improvements, for the items listed in (a) above.

(d)

Tabulation of the HCLPF capacities for items listed on p.1-3 of the submittal, following any implemented or scheduled plant improvements. For scheduled improvements, please provide the implementation schedule.

2.

The submittal does not provide numerical details of the seismic loading utilized for the IPEEE nor the original design basis seismic loading. To evaluate the appropriateness of the reported HCLPF capacities, review of this numeri~I data is necessary.

In addition, ComEd states, without providing a quantitative basis, that equipment screened during the A-46 program is assigned a HCLPF capacity of at least 0.3g PGA.

Please provide the following information to expedite completion of the review:

(a}.

The ground and in-structure response spectra used in the IPEEE.

l (b)

The corresponding safe shutdown earthquake (SSE) design basis ground and in-structure response spectra.

2 (c)

If different from (b), the A-46 in-structure response spectra.

(d)

A quantitative basis for assigning a HCLPF capacity of at least 0.3g PGA for equipment screened in the A-46 program.

3 The re-evaluation of masonry walls for IPEEE (discussed in Section 3.4.3 of the submittal) utilizes 7% damping, compared to 2% damping in the I.E.Bulletin 80-11 evaluation. An increase in the allowable mortar tensile stress from 23 psi to 32 psi.is not a sufficient technical basis for this increase in damping, because only an extremely low level of response is achieved in both analyses. Scaling should be limited to the ratio of allowable tensile stress in the mortar.

Please provide the following additional information concerning the masonry wall evaluations for IPEEE:

(a)

The HCLPF calculations for the two (2) masonry walls with the lowest capacities, for 7% damping and for 2% damping. Please discuss the implications of using 2% damping on the HCLPF capacities in Table 3.1.

(b)

HCLPF capacities for the cribhouse masonry walls, based on 2% damping. (As noted in Question 1, the evaluation of the cribhouse masonry walls was left as an open item in the submittal.

4.

Decay heat removal is achieved in the success paths by the use of the Low Pressure Coolant Injection (LPCI) system in the torus cooling mode with the Containment Cooling Service Water (CCSW) system providing cooling water to the LPCI heat exchangers.

The pumps of the CCSW system take suction from Bay 13 of the crib house.

Regarding the availability of an ultimate heat sink, it is stated in the submittal (Section 3.4.2, page 3-20) that "The NRC safety evaluation for SEP Topic 11-4.E concluded that the plant is designed so that it can be safely shut down in the event of failure of the Dresden dam and the loss of the pool impounded by it. Part of the basis for this conclusion was that there is enough water impounded in the intake and discharge canals below their high point elevations to allow a safe shutdown of Dresden Units 2 and

3. Based on the SEP evaluation, the failure of the dam or dike will not impact the ability to safely shut down Unit 2 and 3", and that "Upon dam failure, the water level in Bay 13 is maintained by the screen wash refuse pit pumps. This requires inserting stop logs and some valving."

The above statements are not sufficient to indicate that there is sufficient cooling water to support the success paths identified in the IPEEE. The systems available for a safe shutdown in the IPEEE may not be the same as those used in the safety evaluation for SEP Topic 11-4.E. As discussed above, the only systems included in the IPEEE for DHR are the LPCI in the torus cooling mode and the CCSW system. DHR by isolation condenser or shut down cooling is not available. The service water system and the fire protection system are also not available in the IPEEE.

3 (a)

Please discuss in more detail the effect of dam failure on the success paths.

Please provide the analysis results that show that the water remaining in the intake and discharge canals is sufficient to support the operation of the cooling water pumps and the cooling needs for all units after dam failure. Please include in the discussion the potential for (e.g., the HCLPF values) and the effects of failure of other structures that may affect water availability, in particular the walls of the intake and discharge canals.

(b)

Please discuss the operator actions required to assure adequate cooling.

Please include in the discussion the systems required (e.g., the availability of the

• screen wash refuse pit pumps and associated components after a seismic margin* earthquake, SME), the time allowed for operator actions, the procedures for operator actions, and the consideration of the adverse effect of the seismic event on operator actions.

(c)

In a recent letter to the NRC (J. M. Heffley to USNRC letter dated September 9, 1998 regarding Failure of the Dresden Lock & Dam), ComEd states that "to satisfy Generic Letter 87-02 program requirements and enhance plant safety using the Seismic Qualification Utility Group (SQUG) methodology, ComEd searched for a method to supply make-up water to the shell of the isolation condensers through piping and components that are seismically qualified or that can be seismically verified using the Generic Implementation Procedure (GIP) of the SQUG program." If such a plant modification is implemented, this will add a success path for decay heat removal, but torus cooling may still be needed for the small LOCA case required in the IPEEE. Please include in your response the status of the above-mentioned program and its effect on IPEEE results.

It is noted that only equipment in the Success Path Equipment List (SPEL) should be considered as available in the above discussions (e.g., service water and shutdown cooling may not be available) and that concurrent demand of cooling water from all units needs to be considered (e.g., may need the operation of more thari one CCSW pumps).

4 Fire

1.

The automatic suppression failure analysis used reliability values from the Fire Induced Vulnerability Evaluation (FIVE) methodology. This data is acceptable for systems that have been designed, installed, and maintained in accordance with appropriate industry standards, such as those published by National Fire Protection Agency (NFPA).

Please verify that automatic fire suppression systems at Dresden meet NFPA standards.

2.

Fire severity factors were used in the analyses of many fire compartments. At issue is the use of severity factors in scenarios where fire suppression was explicitly credited.

The severity factors used appear to be determined, in part, by eliminating fires successfully suppressed from contributing to the fire frequency. Thus, the potential for a large fire that the severity factor represents, is dependent upon the success of fire suppression. There appears to be a significant possibility that the use of a fire severity factor when fire suppression is explicitly modeled credits suppression efforts twice.

Please describe the instances in the Dresden fire assessment in which automatic fire suppression was credited explicitly in conjunction with the use of a fire severity factor.

For each case explain why such credit does not constitute redundant credit for suppression. Please reanalyze the core damage frequency (CDF) contribution from each scenario where redundant credit for suppression is identified.

3.

It appears that the Dresden IPEEE fire analysis assumed that the plant cables are not IEEE-383 qualified. However, the Dresden fire analysis assumed a cable ignition temperature of 932°F (see page 4-16 of the submittal) and cites the EPRI Fi;e PRA

. Implementation Guide (FPRAIG) as the basi_s for this value. This value is significantly optimistic in comparison to piloted ignition temperatures observed in tests by Sandia National Laboratories (SNL) (Ref. NUREG/CR-5546). The SNL tests show that the piloted ignition temperature for cables. will be as low or lower than the thermal damage threshold; hence, use of a piloted ignition temperature of no greater than 425°F would be appropriate for unqualified cable. The assumed temperature of 932°F may have resulted in the optimistic treatment of cable fire growth behavior.

If a cable ignition temperature of 932°F was used, please describe the fire scenarios, associated cables, and analysis results for those cases in which it was applied. Include a specific basis for the assumption that the cable$ at Dresden are consistent with this temperature. Altemative/y, provide an assessment of the impact on the analysis results (CDF) if it is assumed that the flammability and/or appropriate non-qualified cable damage properties, including a piloted ignition temperature of 42S'F, as appropriate.

4.

In computing the extent of fire propagation and equipment damage for a given scenario, it is important that experimental results not be used out of context. Inappropriate use of experimental results (e.g., employing propagation times specific to a particular cable tray separation to fires involving cable trays with lesser separation) can lead to improper assessments of scenario importance. In one case [R.1], rather than performing fire model calculation and using the results, experimental data from a test performed to

5 model cable.tray fire propagation in the absence of an exposure fire was used to model cable response to an exposure fire, which led to over *an order of magnitude reduction in predicted fire-induced core damage frequency.

The submittal apparently assumes a fixed fire spread geometry (35°) for at least one cable tray scenario and fixed propagation delay times between the involvement of subsequent stack trays in the fire. The submittal does not provide a basis for expecting the results of limited experimental observation to be reproduced in the plant fire scenarios.

For each fire scenario in which experimental data were used to estimate the rate and extent of fire propagation, please describe the scenario and how the experimental results were used in the analysis. In those cases where the analysis that was used is found to be unjustified, analyze the scenario using FIVE (or a similar methodology) and provide the results (equipment damaged) of these calculations. Indicate which experimental results were used and how they were utilized in the reanalysis, and justify the applicability of these experimental results to the scenario being analyzed. The.

discussion on results applicability should compare the geometries, ignition sources, fuel type and loadings, ventilation characteristics, and compartment characteristics of the experimental setup( s) with those of the scenario of interest.

5.

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 determines the amount of heat available to the hot gas layer.

A larger heat loss factor means that a larger amount of heat (due to a more severe fire, a longer burning 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 R.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 1 Om x 5m x 4m compartment with a forced ventilation rate of 1130 cfm; 2) the FIVE-recommended damage temperature for qualified cable is 700°F for qualified cable and 450°F for unqualified cable; and, 3) the SFPE curve in the figure is generated from a correlation provided in the Society for Fire Protection Engineers (SFPE) Handbook [R.2].

Based on evidence provided by a 1982 paper by Cooper et al. [R.3], the EPRI Fire PRA

• implementation Guide recommends a heat loss factor of 0.94 for fires with durations greater than five minutes and 0.85 for "exposure fires away from a wall and quickly developing hot gas layers." However, as a general statement, this appears to be a.

misinterpretation of the results. Reference R.3, 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 R.3 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.

6 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 PRA Implementation Guide.

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For each multi-compartment or single area scenario analyzed where the hot gas layer temperature was calculated, please describe the scenario and specify 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 (such as the 0. 7 value recommended by the FIVE methodology). Please provide the resulting changes in scenario contributions to core damage frequency.

6.

The EPRI Fire PRA Implementation Guide methodology for evaluating the effectiveness of suppression efforts treats manual recovery of automatic suppression systems as being independent of subsequent manual efforts to suppress the fire. This assumption is optimistic, since the fire conditions (e.g., heat, smoke) that lead to the failure of recovery efforts can also influence the effectiveness of later suppression efforts. Such an approach, therefore, can overlook plant-specific vulnerabilities.

It is important that all relevant factors be considered in an evaluation of the effectiveness of fire suppression. These factors include: (a) the delay between ignition and

7 detector/suppression system actuation (which is specific to the configuration being analyzed); (b) the time-to-damage for the critical component(s) (which is specific to the fuel type and loading as well as to the configuration being modeled); (c) the response time of the fire brigade (which is plant-specific and fire-location-specific); (d) the time required by the fire brigade to diagnose that automatic suppression has *failed and to take manual action to recover the automatic suppression system; and, (e) performance shaping factors (PSF) affecting fire brigade actions. These PSFs could include factors such as perseverance (persistent efforts made to recover a failed automatic suppression system), smoke obscuration, and impaired communications [R.1].

Finally,.it should be noted that the Nuclear Regulatory Commission (NRC) staff's evaluation of the F.IVE methodology [R.4] specifically stated that licensees need to assess the effectiveness of manual fire-fighting teams by using plant-specific data from fire brigade training to determine the response time of the fire fighters.

• Please identify those scenarios for which credit is taken for both manual recovery of automatic suppression systems and manual suppression of the fires (if manual recovery efforts are unsuccessful), and please indicate the plant equipment that may be affected by the fires. In the analysis of these scenarios, how are dependencies between manual actions. treated? Please justify the treatment, considering the expected fire environment, the recovery actions required, and the manual fire suppression actions required.

7.

The treatment of rnanual suppression appears to be. derived from curves that indicate manual suppression success as a function of fire-fighting time. The submittal does not provide a basis for a quantitative assessment of manua*I suppression effectiveness at Dresden. An acceptable approach to the assessment of manual suppression success compares the damage time to the time required for suppression. The suppression time includes the time to detect the fire, the brigade response time, fire assessment time, and the extinguishment time.

Please provide a comparison of the manual suppression time to the damage time for those compartments where manual suppression was credited. Include in this assessment any adjustments resulting from responses to questions above addressing ignition and damage temperatures, propagation delay assumptions, or model

. parameters.

8.

Assumptions concerning the effectiveness of unrated fire barriers can have a major impact on the screening of multi-compartment fires. The potential for fire barrier failure due to fires in high-hazard areas (e.g., large spills of oil or other liquid fuel, oil-filled transformers, large turbine fires) can also be important a)

Section 4. 7.3.3.2 implies that unrated fire barriers have been credited in the fire study's multi-compartment analysis. Please discuss the impact of eliminating the credit for such barriers in the multi-compartment analyses. (In the analyses, a damage temperature of 425°F for unqualified cable should be used. If a higher temperature is used, such as the 70CJ'F referenced in Section 4. 7.3.3.2 of the submittal, please provide justification.)

8 b)

From the discussions provided for multi-compartment fire scenarios, it can not be determined that the impact resulting from all barriers in high hazard fire areas failing has been considered. Please evaluate the effect of such barrier failures and describe the resulting CDF contributions from the associated scenarios.

9.

Section 4.2 and Appendix C of NUREG-1407, and GL 88-20, Supplement 4 [R.6],

request that documentation be submitted with the IPEEE submittal with regard to the Fire Risk Scoping Study (FRSS) [R. 7] 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 potential cpntributors 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 panels, and shutdown systems. Specific areas that have been identified as requiring attention in the resolution of this issue include:

(a)

Electrical jndependence 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 panels and the evaluation of the level of indication and 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 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 panels, 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)

Total loss of system function: The potential for total loss of system function as a result of fire-induced redundant component failures or electrical distribution system (power source) failure needs to be described.

9 Please provide a description of the control and instrumentation functions that are provided at each remote shutdown station. For each such function indicate whether or not it can*be isolated from damage in the main control room. Has the IPEEE identified or considered any scenarios that might not be mitigated by the remote stations? Provide an evaluation of the reliability of the remote shutdown stations 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 component damage (including damage to MOVs per Information Notice 92-18), a LOCA or interfacing system LOCA, prior to taking control from the remote shutdown panels, spurious starting and running of pumps, and repositioning of valves. Describe how your procedures provide for transfer of control to the remote station(s). Provide an evaluation of whether loss of control power due to hot shorts and/or blown fuses could occur prior to transferring control to the remote shutdown locations and identify the risk contribution of these types of failures (if these failures are screened, please provide the basis for the screening).

10.

The submittal indicates that some fire areas contain elements of both units. The general concern is that the CDFs resulting from fires that impact both units could be significant.

Except for.LOOP, fires that could affect both units were not considered.

For multi-unit sites, there are three issues of potential interest and, for Dresden, a specific concern. 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-important systems/components for each unit that are housed in each such area, (2) for each area identified 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 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 such 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. In the event of a dual unit

. LOOP at Dresden, the submittal states only that "operator actions incorporate the added risk." The fire analysis should include the unavailability of the all 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).

10 Please provide the following relevant information regarding this issue: (1) indicate whether any fire response safe shutdown procedures call for unit cross-connects, (2) the operator actions associated with these procedures, and (3) if any such cross-connects are required, 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 involving 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.

11.

As a result of the seismic/fire walkdown, a hydrogen seal oil control panel and a turbine

• generator hydrogen monitor were found to be unanchored or inadequately anchored.

Hydrogen lines are routed through these cabinets so the potential for a gas release exists. The submittal did not assess the potential risk associated with these lines.

• Please evaluate the risk associated with a seismic/fire event due to inadequate seismic anchoring of the hydrogen seal oil control panel and a turbine generator hydrogen monitor. Alternatively, describe any existing systems or procedures which could mitigate the impact of such hydrogen system failure during a seismic event. Discuss the results, and any plant modifications that might reduce the potential risk as appropriate.

12.

The Dresden procedures for the use of alternate shutdown in the event of main control room abandonment following a fire seems to indicate that the procedures call for reliance on an emergency diesel generator (EDG) whether or not off-site power remains available. Hence, there is a potential that a station blackout (SBO) could result if the controls for the diesel are impacted by fire, the load sequencer fails to perform its function properly, and/or if manual recovery of the diesel generator is compromised by an*

inability to isolate either the connected load sequencers or the damaged control circuits.*

Describe how this aspect of the Dresden altemate shutdown procedures was considered in the IPEEE analysis of fires leading to main control room abandonment and reliance on remote shutdown. If it has not been addressed, provide an assessment of the risk significance of potential SBO scenarios associated with remote shutdown.

Fire RAI References R.1 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.

• ~.

11 R.2 P. J. DiNenno, et al, eds., "SFPE Handbook of Fire Protection Engineering," 2nd Edition, National Fire Protection Association, p. 3-140, 1995.

R.3. 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.

RA A. Thadani, "NRC Staff Evaluation Report on Revised NUMARC/EPRI Fire Vulnerability Evaluation (FIVE) Methodology,", U.S. Nuclear Regulatory Commission, August 21, 1991 (letter to W. Rasin, NUMARC, with enclosure, "Staff Evaluation of the Fire Vulnerability Evaluation (FIVE) Methodology for Use in the IPEEE").

R.5 "Individual Plant Examination of External Events for Seismic, Fire, High Winds/Tornadoes, External Floods, Transportation Accidents; Dresden Nuclear Power

. Station Units 2 and 3," Final Report, Vol. 1, ComEd, December 30, 1997.

R.6 USN RC, "Individual Plant Examination of External Events for Severe Accident Vulnerabilities - 10 CFR §50.54(f)," Generic Letter 88-20, Supplement 4, April 1991.

R.7 J. Lambright, et al., "Fire Risk Scoping Study: Investigation of Nuclear Power Plant Fire Risk, Including Previously Unaddressed Issues," NUREG/CR-5508, prepared for the United States Nuclear Regulatory Commission, January 1989.