ML20212D596

Safety Evaluation Supporting Licensee IPEEE Process.Plant Has Met Intent of Suppl 4 to GL 88-20

ML20212D596
Person / Time
Site:	Point Beach
Issue date:	09/15/1999
From:	NRC (Affiliation Not Assigned)
To:
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ML20212D580	List: ML20212D577 ML20212D596 ML20212D641 ML20212D678
References
GL-88-20, NUDOCS 9909240019
Download: ML20212D596 (10)

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,,, wAsNINotoN, o.c. sosewoos g, . . . . . /g STAFF EVALUATION REPORT ON INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS SUBMITTAL FOR POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2

1. INTRODUCTION On June 28,1991, the NRC issued Generic Letter (GL) 88-20, Supplement 4 (with NUREG-1407, Procedural and Submittal Guidance) requesting all licensees to perform individual plant examinations of external events (IPEEE) to identify plant-specific vulnerabilities to severe accidents and to report the results to the Commission together with any licenser . determined improvements and corrective actions. The licensee for Point Beach Nuclear Plaat, Units 1 and 2, Wisconsin Electric Power Company (WE), submitted its response to the NRC in June 1995.

A " Step 1" review was performed which examined the IPEEE submittal and the licensee's response to a staff request for additionalinformation (RAI). This review focused on the completeness and " reasonableness" of the IPEEE, considering the design and operation of Point Beach. On the basis of the Step 1 review and further review by a senior review board (SRB), the NRC staff concluded that the aspects of high winds, floods, transportation, and other extemal events were adequately r.ddressed. However, the seismic and fire IPEEE needed a supplemental review because of specific concems related to the seismic analyses (e.g., seismic capacity estimates, resolution of reported anomalies and outliers, and relay chatter issue) and fire analyses (e.g., potential for loss of offsite power due to fire, modeling of oil-filled transformers in the cable spreading room, and the analysis of hot shorts). (For a detailed discussion of these concems, see Section 3 of the Technical Evaluation Report (TER)

(Attachment 1). The SRB consisted of NRC staff from the Office of Nuclear Regulatory Research (RES) and the Office of Nuclear Reactor Regulation (NRR), and RES consultants (Sandia National Laboratories) (SNL), with probabilistic risk assessment expertise for extemal

) events.

in March 1998, the NRC staff sent a supplemental RAI to the licensee that focused on the s'eismic and fire concems identified during the Step 1 review. The licensee submitted its response to this RAI in July 1998 and a follow-up response in December 1998. Based on the review of the licensee's response to this RAI, EnergyResearch, Inc. (ERI), completed the -

supplement to its TER in March 1999. The review findings are summarized in the evaluation section below. Details of the ERl's findings are presented in the TER and its supplemental TER, which are attached to this SER.

In accordance with Supplement 4 to GL 88-20, the licensee provided information to address the resolution of Unresolved Safety issue (USI) A-45, " Shutdown Decay Heat Removal g2g0g0 [ ENCLOSURE p PDR

• - c 2

Requirements," Generic Safety issue (GSI) 57," Effects of Fire Protection System Actuation on Safety-Related Equipment," GSI-103,"Desicn for Probable Maximum Precipitation," GSI-131,

" Potential Seismic Interaction involving the Movable in-Core Flux Mapping System Used in Westinghouse Plants," and the NUREG/CR-5088 Fire Risk Scoping Study (FRSS) issues that were explicitly requested in Supplement 4 to GL 88-20 and its associated guidance in NUREG-1407. The licensee did not propose to resolve any additional USts or GSis as part of the Point Beach IPEEE.

II. EVALUATION The Point Beach Nuclear Plant consists of a dual-unit, Westinghouse 2-loop pressurized water reactor (PWR), with a large dry containment. The seismic category I structures were designed to a seismic acceleration level of 0.12g peak ground acceleration (PGA) with a spectral shape conforming to a Housner spectrum. The licensee performed a seismic probabilistic risk assessment (PRA), including a qualitative and quantitative containment performance analysis for Point Beach, Units 1 and 2. The licensee's seismic PRA uses the existing Individual Plant l

Examination of Intemal Events (IPE) level 1 logic models and the level 2 containment event tree I model for quantifying seismic risk. The seismic PRA approach uses a surrogate element to

. represent the seismic failure effects of all components that were screened out at a high confidence oflow probability of failure (HCLPF) value of 0.3g PGA. Simplified fragility calculations and detailed fragility calculations were performed for components that were not screened out. Since Point Beach is a USl A-46 plant, the resolution of outliers and the walkdowns were performed in coordination with evaluation and walkdowns under Point Beach's USl A-46 program.

The licensee's fire IPEEE analysis was based on the Electric Power Research Institute's (EPRI's) fire , induced vulnerability evaluation (FIVE) method. However, the licensee stated that the data was modified using EPRI's Fire PRA implementation Guide to reduce the conservatism in the FIVE fire propagation analysis. The licensee's fire analysis made extensive use of the plant information gathered for Appendix R compliance. The licensee used the IPE model to establish the possibility of core damage resulting from a fire-initiated event. The conditional core damage probability was based on the equipment and systems unaffected by the fire. Human actions considered in the IPE plant model were included in the fire impact assessment. The IPE human error probabilities were modified to take into account the additional stress that could be caused by the occurrence of a fire.

For other external events, the licensee used the progressive screening approach as described in NUREG-1407 to evaluate high winds, floods, transportation, and nearby facility accidents.

The licensee reviewed and updated the results of a Point Beach PRA performed previously for USl A-45," Shutdown Decay Heat Removal Requirements"(under NRC Task Action Plan A-45), l based on screening, bounding, and some probabilistic calculations. Among high winds, floods, '

and other extemal events, the licensee analyzed tornadoes and external floods further using quantitative bounding and PRA evaluations. Historical data were used for determining straight wind, tomado, and external flood frequencies. Some site-specific data were used for the analyses of aircraft crashes, land transportation accidents, and nearby facility events. No formal PRA or bounding analysis was performed for transportation and nearby facility accidents. These events were screened out due to their low frequency of occurrence.

i

1 4 3

Core Damane Freauency and Seismic Canadtv Estimates The licensee estimated a seismic core damage frequency (CDF) of 1.4E-5/ reactor year (RY) using the 1989 seismic hazard curves developed by the Electric Power Research Institute, and i a CDF of 1.3 E-5/RY using the 1993 seismic hazard curves developed by Lawrence Livermore  !

. National Laboratory. The licensee also estimated the plant seismic' capacity, in terms of high confidence of low probability of failure (HCLPF) value, as 0.16g PGA with operator actions and random (nonseismic) failures included, or 0.25g PGA with these actions and failures excluded.

The licensee estimated a fire CDF of 5.1E-5/RY in its IPEEE submittal. The licensee estimated

~ that the CDF due to internal events is about 1.2E-4/RY, including internal flooding.

The licensee estimated that the CDF due to external floods is about 2.8E-6/RY and the CDF due to high winds is 3.4E-7/RY. Other extemal events (e.g., transportation and nearby facility accidents) were considered as risk insignificant based on their low hazard frequencies.

Dominant Contributors The licensee reported that the seismic CDF is dominated by seismically induced sequences (

, such as failure of cable trays inside the cable spreading room (62%), failure of cable trays outside the cable spreading room (7%), and failure of a surrogate element (16%), as discussed below. The licensee used an approach of a surrogate element to represent the seismic failure effects of all components (e.g., soils, building / structures, reactor vessel) that were screened out at a HCLPF value of 0.3g PGA. Although the contribution of the seismic sequence involving the surrogate element appeared to be exaggerated at Point Beach (i.e., the combined CDF contribution of those sequences involving soils, building / structures, and the reactor vessel as an individual component, based on the conventional PRA approach, would be far less than that of the sequence represented by the surrogate element), the ranking of the dominant sequences is not changed (see Section 2.1.10 of the attached TER). The inside cable spreading room

. sequence consisted of the seismically induced failure of cable trays leading to loss of indication and/or control combined with failure to shut down the plant remotely. The outside cable spreading room sequence consisted of the seismically induced failure of cable trays leading to loss of power to all essential equipment. The dominant seismic contributors to the estimated CDF are operator actions (e.g., failure to shut down the plant from the remote shutdown panel, failure to provide service water backup to auxiliary feedwater (AFW) pump suction), seismic faults that lead directly to core damage (e.g., failures of cable trays, surrogate element (e.g.,

failure of soils, building structures), and failures of critical equipment (e.g., transformers,480V load centers, level transmitter for condensate storage tank (CST)).

The fire CDF is dominated by fires in the control room, the cable spreading room, the AFW pump room, the gas turbine room, the vital and nonvital switchgpar rooms, the diesel generator -

rooms, and the monitor tank room. All fire event sequences were quantified assuming all equipment / cables in the area would fail by the fire.' However, no fire in a single area or compartment (by itself), except the cable spreading room, would have the potential of directly

- causing core damage. In addition to the damage caused by the fire in these areas and compartments, other failures (e.g., a reactor coolant pump (RCP) seal loss-of-coolant accident,

. operator failure to establish RCP seal injection via the refueling water storage tank, failure of }

the safety relief valves), in combination with a fire, contributed to the calculated fire CDF.

o 1 i

4 The external flood CDF is dominated by flooding as a result of the rising water level of Lake Michigan, combined wih wind wave effects and water runup, wt'ich would result in

' flooding of the turbine build;ng and leads to loss of the ability to remove decay heat. The dominant tomado/ wind-induced core damage sequences were determined to involve failures of the diesel generator exhaust stacks, which lead to failure of both diesel generators.

The licensee's IPEEE analyses appear to have examined the significant initiating events and dominant accident sequences at the plant.

Containment Performance The licensee performed both qualitative and quant %tive assessments of the containment performance under seismic conditions at Poir.t i: leach. The qualitative assessment examined containment safeguards systems significant to large early releas,c, including containment integrity, mechanical penetrations, containment isolation, and containment cooling. The

)

j quantitative assessment focused on bypass and containment isolation failure sequences using

)

the Point Beach IPE plant damage states, release categories, and containment response '

analysis. The licensee estimated the frequency of early release to be 2.1E-6/RY, and the

- dominant contributors were failures of containment isolation due to the loss of either AC or DC power and failures of the containment low and intermediate range pressure transmitters due to

. the impact of the seismically induced collapse of the adjacent block walls.

The licensee's evaluation of containment performance with regard to fire hazard focused on assessing the potential for containment isolation or bypass failure. The licensee identified and evaluated areas that could potentially impact the function of containment isolation. The licensee concluded that fire is not expected to result in any significant containment isolation failure or bypass.

The licensee's containment perform.ance analyses for seismic and internal fire events appear to have considered important containment performance issues and are consistent with the intent of Supplement 4 to GL 88-20. j Generic Safety lasues A's part of the IPEEE, a set of generic and unresolved safety issues (USl A-45, GSI-131, )

GSI-103, GSI 57, and the SNL FRSS lasues) were specifically identified in Supplement 4 to GL 88-20 and its associated guidance in NUREG-1407 as needing to be addressed in the

'IPEEE. The NRC staff's evaluation of these issues is provided below.

-1. USl A-45," Shutdown Decay Heat Removal Requirements" The licensee has chosen to subsume its resolution of USl A45 in the USI A-46 program.

However, the licensee provided a brief discussion of decay heat removal (DHR) capabilities following a seismic event in its IPEEE submittal. The seismic PRA modeled the systems available for DHR, and the CDF contributions due to failures of DHR systems

.and support systems were tabulated in the IPEEE submittal. The licensee credited the -

AFW system for providing DHR capability under fire conditions and provided the analysis to address the USl A-45 issue in the IPEEE submittal. The NRC staff finds that the i

. 4 5

licensee's USI A-45 evaluation is consistent witn the guidance provided in Section 6.3.3.1 of NUREG-1407 and, therefore, the NRC staff considers this issue resolved.

2. GSI-131, " Potential Seismic interaction involving the Movable in-Core Flux Mapping System Used in Westinghouse Plants" Even though this issue is not directly applicable to Point Beach because the flux mapping cart is not movable, as stated in the submittal, the licensee provided a discussion on the seismic capability of the cart. The licensee pointed out that the Point Beach cart is identical to the Kewaunee Nuclear Power Plant flux mapping cart, which was analyzed under the Kewaunee IPEEE. Since the seismic demands for both plants are quite similar, and the Kewaunee IPEEE result showed that the cart possesses adequate seismic capability, the licensee concluded that the Point Beach cart also has adequate seismic capability, in addition, the licensee pointed out that two lateral struts, added to the cart for lateral restraints by a previous modification, will increase its capability under seismic conditions. The NRC staff considers this issue resolved for Point Beach on the basis of the information provided in the IPEEE and the TER.

3. GSt-103, " Design for Probable Maximum Precipitation" As part of the IPEEE flood analysis, the licensee reviewed and updated the results of the USl A-45 study, including the assessment of GSI-103 (local precipitation in Section 5.2.5 and roof ponding in Section 5.2.6 of the IPEEE submittal), and concluded that the i frequency of Probable Maximum Precipitation (PMP) events is below 1E-6/RY and has insignificant impact on Point Beach. The NRC staff finds that the licensee's GSI-103 evaluation is consistent with the guidance provided in Section 6.2.2.3 of NUREG-1407 and, therefore, the NRC staff considers this issue resolved.

4. Fire Risk Scoping Study issues The licensee has explicitly addressed the SNL FRSS issues (Section 4.8 of the IPEEE submittal). The NRC staff finds that the licensee's evaluations are consistent with the guidance provided in FIVE, which was accepted by the NRC staff and, therefore, the NRC staff considers these issues resolved.

5. GSI-57," Effects of Fire Protection System Actuation on Safety-Related Equipment" Although the licensee's IPEEE submittal did not explicitly discuss GSI 57, the information provided in the submittal (Section 4.8) addressing seismic-fire interactions and total environment equipment survival is related to this issue. The NRC staff finds that the licensee's evaluation is consistent with the guidance provided in NUREG-1407 and, therefore, the NRC staff considers this issue resolved.

in addition to those safety issues discussed above that were explicitly requested in

-Supplement 4 to GL 88-20, four generic safety issues were not specifically identified as issues that should be resolved under the IPEEE program; thus, they were not explicitly discussed in Supplement 4 to GL 88-20 or NUREG-1407. However, subsequent to the issuance of the

. generic letter, the NRC evaluated the scope and the specific information as requested in the 4

r y -

6 generic letter and the associated guidance, and concluded that the plant-specific analyses being requested in the IPEEE program could also be used, through a satisfactory IPEEE submittal review, to resolve the extemal event aspects of these four safety issues. The '

following discussions summarize the NRC staff's evaluation of these safety issues at Point Beach.

1. GSI-147," Fire-induced Altemate Shutdown / Control Room Panel interactions" The licensee's IPEEE submittal contains information (Section 4.8.5 of the Point Beach IPEEE submittal) addressing this issue. The licensee reviewed the circuits associated with Point Beach altemate shutdown capability and did not discover any control system j interaction concems. The hot short concems (e.g., the possibility of inadvertent opening of l a power-operated relief valve of an atmospheric steam dump valve) were discussed in its response to an NRC staff RAI (Section 2.4.1 of the attached TER). Based on the results

! of the IPEEE submittal review, the NRC staff considers that the licensee's process is capable of identifying potential vulnerabilities associated with this issue. On the basis that no vulnerability associated with this issue was identified in the IPEEE submittal, the NRC staff considers this issue resolved for Point Beach.

l 2. GSI-148," Smoke Control and Manual Fire-Fighting Effectiveness" i

The licensee addressed this issue in Section 4.8.3 of its IPEEE submittal. The licensee

provided a discussion of the Point Beach's fire protection program and its associated fire brigade training program related to this issue. Based on the results of the IPEEE submittal review, the NRC staff considers that the licensee's process is capable of identifying potential vulnerabilities associated with this issue. On the basis that no vulnerability associated with this issue was identified in the IPEEE submittal, the NRC staff considers this issue resolved for Point Beach.

3.  !

GSI-156, " Systematic Evaluation Program (SEP)" '

I

The licensee's IPEEE submittal contains information directly addressing the following extemal-events-related SEP issues

settlement of foundations and buried equipment (Sections 3.1.2.1,3.1.3.10, and 3.1.4.2); dam integrity and site flooding (Section 5.2.2);

site hydrology and ability to withstand floods (Section 5.2); industrial hazards (Section 5.3);

tomado missiles (Section 5.1) and severe weather effects on structures (Sections 5.1 and

~ 5.2). Although the licensee's IPEEE submittal did not contain information explicitly addresting the other SEP issues (i.e., design codes, criteria, and load combinations; and seismic design of structures, systems, and components), a conclusion may be drawn that these issues were implicitly addressed as part of the licensee's IPEEE process at l Point Beach because the seismic PRA has taken the as-built, as-operated conditions into l consideration in its structural response'and component fragility analyses, and no potential L seismic vulnerabilities were identified. Based on the results of the IPEEE submittal review, i l the NRC staff considers that the licensee's process is capable of identifying potential l vulnerabilities associated with this issue. On the basis that no vulnerability associated with t

this issue was identified in the IPEEE submittal, the NRC staff considers this issue resolved for Point Beach.

7

14. GSI-172, " Multiple System Responses Program (MSRP)"

The licensee's IPEEE submittal contains information directly addressing the following

. extemal events-related MSRP issues: (1) effects of fire protection system actuation on non-safety-related and safety-related equipment (Sections 4.8.1 and 4.8.4), (2) seismically induced fire suppression system actuation (Section 4.8.1), (3) seismically induced fires (Sections 3.1.3 and 4.8.1), (4) effects of hydrogen line rupture (Sections 4.12 and 5.3.3),

(5) the IPEEE-related aspects of common cause failures related to human errors (Sections 3.1.3, 3.1.5, 4.6.2, 4.8.4, 5.1.5, and 5.1.6), (6) non-safety-related control system / safety-related system dependencies (Sections 3.1.2,3.1.3,3.1.4, and 4.8.1),

. (7) effects of flooding and/or moisture intrusion on non-safety-related and safety-related equipment (Sections 3.1.3 and 4.8.1), (8) seismically induced spatial / functional interactions (Sections 3.1.2,3.1.3,3.1.4, and 4.8.1), (g) seismically induced flooding (Sections 3.1.3 and 4.8.1), (10) seismically induced relay chatter (Sections 3.1.2 and 3.1.4), and (11) evaluation of earthquake magnitude greater than safe shutdown earthquake (Section 3).

l Based on the results of the IPEEE submittal review, the NRC staff considers that the licensee's process is capable ofidentifying potential extemal events-related vulnerabilities I

associated with GSI-172. On the basis that no potential vulnerability associated with this

.ssue was identified in the IPEEE submittal, the NRC staff considered the IPEEE-related aspects of this issue resolved for Point Beach.

Unious Plant Features. Potential Vulnerabilities. and imorovements The licensee did not identify any unique safety features at the plant.

The licensee's definition of vulnerability associated with extemal events is the same as that used in their IPE. The licensee's criteria for determining if any plant vulnerability exists are: a CDF goal of less than 1E-4/RY and.a goal for frequency of large fission product release of less than 1E-6/RY. Although the IPEEE did not identify any vulnerabilities associated with extemal events at Point Beach, the licensee has implemented certain plant modifications and proposed to take certain actions related to relay and anchorage issues. These licensee-identified improvements and proposed actions are listed below:

High Winds e The diesel generator stacks were modified to accommodate higher winds.

Fire o- The control logic for automatic start of the motor-driven AFW pumps was modified such

- that a fire in the nonsafeguards switchgear will not prevent the motor-driven pumps from starting, e' The procedure for mitigating fires in the control room and cable spreading room was revised to include verification of closure of an additional containment isolation valve.

4

8 Seismic

e. Numerous equipment anchorages (e.g., transformers, a reactor trip breaker cabinet, inverters and battery chargers) were fixed for seismic concerns (see SER on USl A-46 resolution, dated July 7,1998).

e The seismic capacity concems of certain tank anchorages were either reanalyzed or

. upgraded by modification (see SER on USI A-46 resolution).

'e The supports for certain cable and conduit raceways in the cable spreading room were modified to increase their seismic capacities (see SER on USl A-46 resolution).

e Certain relays, identified as outliers during an IPEEE/A-46 walkdown, will be reevaluated and resolved by the fall of 1999, except those pertaining to the diesel generators. The licensee plans to replace the controls for the diesel generators by the year 2002 (see SER on USl A-46 resolution).

In addition, the licensee stated that they intend to use the Point Beach IPEEE as a decision-making tool in many aspects of engineering support and plant operations as long as the tool adds value (Section 8 of the IPEEE submittal). The licensee also stated that a clear understanding of the assumptions, limitations, uncertainties is required in order to avoid

- misinterpretation of the IPEEE results. Although the implications of using surrogate element in

. the seismic PRA (e.g., limitations, uncertainties) were not discussed in the submittal (see Section 3.1 of Attachment 2), the impact on finding the dominant sequences at Point Beach may not be very significant. As discussed previously in the section on dominant contributors, the ranking of the dominant seismic sequences would not be significantly changed. Therefore, the NRC staff believes that, as a whole, the licensee has gained some significant insights related to potential vulnerabilities to severe accidents through the IPEEE.

Ill. CONCLUSION On the basis of the above findings, the NRC staff notes that (1) the licensee's IPEEE is complete with regard to the information requested by Supplement 4 to GL 88-20 (and associated guidance in NUREG-1407), and (2) the IPEEE results are reasonable given the Point Beach design, operation, and history Therefore, the NRC staff concludes that the licensee's IPEEE process is capable of identifying the most likely severe accidents and severe accident vulnerabilities, and therefore, that the Point Beach, Units 1 and 2, IPEEE has met the intent of Supplement 4 to GL 88-20 'and the resolution of specific generic safety issues discussed in this SER.

I

9 l It should be noted that the NRC staff focused its review primarily on the licensee's ability to l examine Point Beach for severe accident vulnerabilities. Although certain aspects of the IPEEE

)

l ' were explored in more detail than others, the review is not intended to validate the accuracy of the licensee's detailed findings (or quantification estimates) that stemmed from the .

examination. Therefore, this SER does not constitute NRC approval or endorsement of any l

- lPEEE material for purposes other than those associated with meeting the intent of Supplement 4 to GL 88 20 and resolving the generic safety issues discussed in Section 11 of this SER.

Attachments: 1. Technical Evaluation Report

2. Supplemental Technical Evaluation Report Date: September 15,1999 i

i 4

e i

' M.B. Sellman September 15, 1999 l In addition, the licensee's IPEEE submittal contains some specific information that addresses the external event aspects of GSI-147," Fire-Induced Alternate Shutdown / Control Room Panel Interactions," GSI-148, " Smoke Control and Manual Fire-Fighting Effectiveness," GSI 156,

" Systematic Evaluation Program (SEP)," and GSI-172," Multiple System Responses Program."

The specific information associated with each issue is identified and discussed in the enclosed SER. Based on the review of the information contained in the submittal, the NRC staff  !

considered that the licensee's process is capable of identifying potential vulnerabilities  !

associated with these issues. On the basis that no vulnerabilities associated the external event {

aspects of these issues were identified at Point Beach, the NRC staff considers these issues resolvec, for this plant site.

l This completes the NRC staff's efforts in regard to TAC Nos. M83661 and M83662. The NRC staff appreciates your efforts in regard to this matter. If you have any questions concerning the staff's review of your IPEEE submittal, please contact me at (301) 4151355.

Sincerely, Original signed by: l Beth A. Wetzel, Senior Project Manager, Section 1 Project Directorate 111 Division of Licensing Project Management Office of Nuclear Reactor Regulation

Enclosure:

Staff Evaluation Report cc w/ encl: See next page DISTRIBUTION Docket File PD31 Reading PUBLIC OGC ACRS GCook RHernan SBlack JChen, RES CWoods, RES DCoe RLanksbury, Rlli DOCUMENT NAME: g:\pdiii-1\ptbeach\ltr83661 To receive a copy of this document, indicate in the box: "C" = Copy without attachment / enclosure "E" = Copy with attachment / enclosure "N" = No copy OFFICE PM:PD31 _ lE LA:PD31 .

E SC:PD}1 , E NAME BWetzel /MM RBouling h% '

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' ' i /M99 DATE 9 // f/99 *\ /LT99 OFFICIAL RECORD COPY

4 4 tre p* *, UNITED STATES E

NUCLEAR REGULATORY COMMISSION E WASHINGTON, D.C. Sma "1

\ *****

/ September 15, 1999 Mr. Michael B. Sellman '

Senior Vice President and Chief Nuclear Officer Wisconsin Electric Power Compnny 231 Vbst Michigan Street Milmukee, WI 53201

SUBJECT:

POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 - REVIEW OF INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE) SUBMITTAL (TAC NOS. M83661 AND M83662)

Dear Mr. Sellman:

On June 30,1995, Wisconsin Electric Power Company (the licensee) submitted its response to Generic Letter (GL) 88-20, Supplement 4, which requested all licensees to perform individual plant examinations of external events (IPEEE) (1) to identify plant-specific vulnerabilities to severe accidents, and (2) to report the results to the Nuclear Regulatory Commission (NRC),

together with any licensee-determined improvements and corrective actions. The NRC staff's review findings on the Point Beach Nuclear Plant, Units 1 and 2, IPEEE submittal are summarized in the enclosed Staff Evaluation Report (SER). Attached to the SER is a Technical  !

Evaluation Report (TER) and its supplement prepared by the NRC's contractor, Energy )

Research Inc. (ERI).

The NRC staff performed a " Step 1" review, which examined the IPEEE submittal and the I licensee's response to a staff request for additional information (RAI). This review focused on the completeness and " reasonableness" of the IPEEE considering the design and operation of

. Point Beach. On the basis of this review and further review by a senior review board (SRB), the NRC staff concluded that the aspects of high winds, floods, transportation, and other external events were adequately addressed. The SRB consisted of NRC staff from the Office of Nuclear Regulatory Research (RES) and the Office of Nuclear Reactor Regulation (NRR), and .

RES consultants (Sandia National Laboratories), with probabilistic risk rtsessment expertise for external events. The seismic and fire IPEEE needed a supplemente; twiew because of specific concerns related to the seismic analyses (e.g., seismic cap 20ty estimates, resolution of reported anomalies and outliers, and relay chatter issue) and fire ana;yses (e.g., potential for loss of offsite power due to fire, modeling of oil-filled transformers in tne cable spreading room, and the analysis of hot shorts). For a detailed discussion of these concerns, see Section 3 of the TER (Attachment 1 to the SER).

In March 1998, the NRC staff sent a supplemental request for additionalinformation to the licensee that focused on the seismic and fire concerns identified during the Step 1 review. The licensee submitted its response to this RAI in July 1998 and a followup response in December 1998. Based on the review of the licensee's response to this RAI, ERI completed the supplement to its TER in March 1999.

~@

M.B. Sellman 2-The licensou estimated a seismic core damage frequency (CDF) of 1.4E 5/ reactor year (RY) using the 1989 seismic hazard curves developed by the Electric Power Research Institute (EPRI), and a CDF of 1.3 E-5/RY using the 1993 seismic hazard curves developed by .

Lawrence Livermore National Laboratory. The licensee also estimated the plant seismic capacity, in terms of high confidence of low probability of failure (HCLPF) value, as 0.16g PGA with operator actions and random (nonseismic) failures included, or 0.25g PGA with these actions and failures excluded.

The licenses estimated a fire CDF of about 5.1E 5/RY in its IPEEE submittal. The licensee also estimated that the CDF due to intomal events is about 1.2E-4/RY, including internal flooding.

For other extemal events, the licensee used the progressive screening approach as described in NUREG-1407 to evaluate high winds, floods, transportation, and nearby facility accidents.

The licensee reviewed and updated the results of a Point Beach probabilistic risk assessment (PRA) performed for Unresolved Safety issue (USl) A-45," Shutdown Decay Heat Removal Requirements" (NRC Task Action Plan A-45), based on screening, bounding, and some probabilistic calculations. The licensee analyzed tornadoes and extemal floods further using quantitative bounding and PRA evaluations. Historical data were used for determining straight

. wind, tornado, and external flood frequencies. Some site specific data were used for the analyses of aircraft crashes, land transportation accidents, and nearby facility events. No

. formal PRA or bounding analysis was performed for transportation and nearby facility accidents. These events were screened out due to their low frequency of occurrence.

The licensee's definition of vulnerability associated with external events is the same as that used in their IPE. The licensee's criteria for determining if any plant vulnerability exists are (1) a CDF goal of less than 1E-4/RY and (2) a goal for frequency of large fission product release of less than 1E-6/RY. Although the IPEEE did not identify any vulnerabilities associated with extemal events at Point Beach, the licensee has implemented certain plant modifications and proposed to take certain actions. These licensee identified improvements and proposed actions are to: (1) modify the diesel generator stacks (high wind-related), (2) modify control

- logic for motor-driven AFW pumps and change fire mitigating procedure (fire-related), and (3) upgrade certain anchorages for equipment and tanks, modify certain cable tray supports, and reevaluate and replace certain relays (seismic-related).

The licensee has addressed the following generic safety issues (GSis): GSI 57," Effects of Fire Protection System Actuation on Safety-Related Equipment," GSI 103, " Design for Probable

. Maximum Precipitation (PMP)," and GSI 131 " Potential Seismic Interaction involving the Movable in-Core Flux Mapping System Used in Westinghouse Plants." The licensee also addressed USl A-45, " Shutdown Decay Heat Removal Requirements," and the Fire Risk Scoping Study (FRSS) issues that were explicitly requested in Supplement 4 to GL 88 20 and its associated guidance in NUREG-1407.

On the basis of the Step 1 and supplemental RAI reviews, the NRC staff concludes that the licensee's IPEEE process is capable of identifying the most likely severe accidents and severe accident vulnerabilities and, therefore, that the Point Beach IPEEE has met the intent of Generic Letter 88-20, Supplement 4.

4

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• I l

M.B. Sellman 3 In addition, the licensee's IPEEE submittal contains some specific information that addresses the external event aspects of GSI 147," Fire-induced Alternate Shutdown / Control Room Panel Interactions," GSI-148, " Smoke Control and Manual Fire-Fighting Effectiveness," GSI 156,

" Systematic Evaluation Program (SEP)," and GSI-172, " Multiple System Responses Program."

The specific information associated with each issue is identified and discussed in the enclosed SER. Based on the review of the information contained in the submittal, the NRC staff considered that the licensee's process is capable of identifying potential vulnerabilities associated with these issues. On the basis that no vulnerabilities associated the external event aspects of these issues were identified at Point Beach, the NRC staff considers these issues resolved for this plant site.

This completes the NRC staff's efforts in regard to TAC Nos. M83661 and M83662. The NRC staff appreciates your efforts in regard to this matter, if you have any questions concerning the staff's review of your IPEEE submittal, please contact me at (301) 415-1355.

Sincerely, bi 0 l Beth A. Wetzel, Senior Project Manager, Section 1 Project Directorate lli i Division of Licensing Project Management Office of Nuclear Reactor Regulation

)

Enclosure:

Staff Evaluation Report  ;

cc w/ encl: See next page  ;

i

o .

Wisconsin Electric Power Company Point Beach Nuclear Plant cc: Units 1 and 2 Mr. John H. O'Neill, Jr.

Shaw, Pittman, Potts & Trowbridge Ms. Sarah Jenkins 2300 N Street, NW Electric Division Washington, DC 20037-1128 Public Service Commission of Wisconsin P.O. Box 7854 Mr. Richard R. Grigg Madison, Wisconsin 53707-7854 President and Chief Operating Officer

]

Wisconsin Electric Power Company )

231 West Michigan Street Milwaukee, Wisconsin 53201 Mr. Mark E. Reddemann Site Vice President Point Beach Nuclear Plant Wisconsin Electric Power Company 6610 Nuclear Road Two Rivers, Wisconsin 54241 Mr. Ken Duveneck Town Chairman Town of Two Creeks 13017 State Highway 42 Mishicot, Wisconsin 54228 i

Chairman l Public Service Commission I of Wisconsin P.O. Box 7854 Madison, Wisconsin 53707-7854 Regional Administrator, Region 111 U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, illinois 60532-4351 Resident inspector's Office U.S. Nuclear Regulatory Commission 6612 Nuclear Road .

Two Rivers, Wisconsin 54241 October 1998

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STAFF EVALUATION REPORT ON INL'lVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS SUBMITTAL FOR POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 j i

L 1. INTRODUCTION On June 28,1991, the NRC issued Generic Letter (GL) 88-20, Supplement 4 (with NUREG-  !

1407, Procedural and Submittal Guidance) requesting all licensees to perform individual plant examinations of extemal events (IPEEE) to identify plant-specific vulnerabilities to severe )

accidents and to report the results to the Commission together with any licensee-determined )

- improvements and corrective actions. l oe licensee for Point Beach Nuclear Plant, Units 1 and 2, Wisconsin Electric Power Company (WE), submitted its response to the NRC in June 1995.

A

• Step 1" review was ' performed which examined the IPEEE submittal and the licensee's response to a staff request for additional information (RAI). This review focused on the completeness and

• reasonableness" of the IPEEE, considering the design and operation of Point Beach. On the basis of the Step 1 review and further review by a senior review board (SRB), the NRC staff concluded that the aspects of high winds, floods, transportation, and other extemal events were adequately addressed. However, the seismic and fire IPEEE needed a supplemental review because of specific concems related to the seismic analyses (e.g., seismic-capacity. estimates, resolution of reported anomalies and outliers, and relay chatter issue) and fire analyses (e.g., potential for loss of offsite power due to fire, modeling of oil-filled transformers in the cable spreading room, and the analysis of hot shorts). (For a detailed discussion of these concems, see Section 3 of the Technical Evaluation Report (TER)

(Attachment 1). The SRB consisted of NRC staff from the Office of Nuclear Regulatory Research (RES) and the Office of Nuclear Reactor Regulation (NRR), and RES consultants (Sandia National Laboratories) (SNL), with probabilistic risk assessment expertise for extemal events, in March 1998, the NRC staff sent a supplemental RAI to the licensee that focused on the seismic and fire concems identified during the Step 1 review. The licensee submitted its response to this RAI in July 1998 and a follow-up response in December 1998. Based on the review of the licensee's response to this RAI, Energy Research, Inc. (ERI), completed the

- supplement to its TER in March 1999. The review findings are summarized in the evaluation section below Details of the ERl's findings are presented in the TER and its supplemental TER, which are attached to this SER. ,

in accordance with Supplement 4 to GL 88-20, the licensee provided information to address the I resolution of Unresolved Safety issue (USI) A-45, " Shutdown Decay Heat Removal -

ENCLOSURE

Requirements," Generic Safety issue (GSI) 57,

• Effects of Fire Protection System Actuation on Safety-Related Equipment,* GSI-103, " Design for Probable Maximum Precipitation," GSI-131,

. Potential Seismic Interaction involving the Movable in-Core Flux Mapping System Used in Westinghouse Plants," and the NUREG/CR-5088 Fire Risk Scoping Study (FRSS) issues that were explicitly requested in Supplement 4 to GL 88-20 and its associated guidance in NUREG-1407. The licensee did not propose to resolve any additional USIs or GSIs as part of the Point Beach IPEEE.

11. EVALUATION The Point Beach Nuclear Plant consists of a dual-unit, Westinghouse 2-loop pressurized water reactor (PWR), with a large dry containment._ The seismic category I structures were designed to a seismic acceleration level of 0.12g peak ground acceleration (PGA) with a spectral shape conforming to a Housner spectrum. The licensee performed a seismic probabilistic risk assessment (PRA), including a qualitative and quantitative containment performance analysis for Point Beach, Units 1 and 2. The licensee's seismic PRA uses the existing Individual Plant Examination of Intamal Events (IPE) level 1 logic models and the level 2 containment event tree model for quantifying seismic risk. The seismic PRA approach uses a surrogate element to

- represent the seismic failure effects of all components that were screened out at a high

- confidence of Ic,w probability of failure (HCLPF) value of 0.3g PGA. Simplified fragility calculations and detailed fragility calculations were performed for components that were not screened out. Since Point Beach is a USl A-46 plant, the resolution of outliers and the walkdowns were performed in coordination with evaluation and walkdowns under Point Beach's USl A-46 program.

The licensee's fire IPEEE analysis 'was based on the Electric Power Research Institute's (EPRl's) fire-induced vulnerability evaluation (FIVE) method. However, the licensee stated that the data was modified using EPRl's Fire PRA implementation Guide to reduce the conseNatism in the FIVE fire propagation analysis. The licensee's fire analysis made extensive use of the plant information gathered for Appendix R compliance. The licensee used the IPE model to establish the possibility of core damage resulting from a fire-initiated event. The conditional core damage probability was based on the equipment and systems unaffected by the fire. Human actions considered in the IPE plant model were included in the fire impact assessment. The IPE human error probabilities were modified to take into account the

> additional stress that could be caused by the occurrence of a fire.

. For other extemal events, the licensee used the progressive screening approach as described in NUREG-1407 to evaluate high winds, floods, transportation, and nearby facility accidents.

The licensee reviewed and updated the results of a Point Beach PRA performed previously for

, USl A-45," Shutdown Decay Heat Removal Requirements"(under NRC Task Action Plan A-45),

based on screening, bounding, and some probabilistic calculations. Among high winds, floods, and other extemal events, the licensee analyzed tornadoes and external floods further using quantitative bounding and PRA evaluations. Historical data were used for determining straight wind, tomado, and extemal flood frequencies. Some site-specific data were used for the analyses of aircraft crashes, land transportation accidents, and nearby facility events. No formal PRA or bounding analysis was performed for transportation and nearby facility accklents. These events were screened out due to their low frequency of occurrence.

I.

, 4 f-3 Core Damane Freauency and Seismic C=__nacity Estimates .

The licensee estimated a seismic core damage frequency (CDF) of 1.4E-5/ reactor year (RY) using the 1989 seismic hazard curves developed by the Electric Power Research Institute, and a CDF of 1.3 E 5/RY using the 1993 seismic hazard curves developed by Lawrence Livermore National Laboratory. The licensee also estimated the plant seismic capacity, in terms of high confidence of low probability of failure (HCLPF) value, as 0.16g PGA with operator actions and random (nonseismic) failures included, or 0.25g PGA with these actions and failures excluded.

The licensee estimated a fire CDF of 5.1E-5/RY in its IPEEE submittal. The licensee estimated that the CDF due to internal events is about 1.2E-4/RY, including internal flooding.

The licensee estimated that the CDF due to external floods is about 2.8E-6/RY and the CDF due to high winds is 3.4E 7/RY. Other external events (e.g., transportation and nearby facility accidents) were considered as risk insignificant based on their low hazard frequencies.

Dominant Contributors

~ The licensee reported that the seismic CDF is dominated by seismically induced sequences such as failure of cable trays inside the cable spreading room (62%), failure of cable trays outside the cable spreading room (7%), and failure of a surrogate element (16%), as discussed below. The licensee used an approach of a surrogate element to represent the seismic failure effects of all components (e.g., soils, building / structures, reactor vessel) that were screened out at a HCLPF value of 0.3g PGA. Although the contribution of the seismic sequence involving the surrogate element appeared to be exaggerated at Point Beach (i.e., the combined CDF.

contribution of those sequences involving soils, building / structures, and the reactor vessel as an individual component, based on the conventional PRA approach, would be far less than that of the sequence repruented by the surrogate element), the ranking of the dominant sequences is not changed (see 3ection 2.1.10 of De attached TER). The inside cable spreading room sequence consisted of the seismically induced failure of cable trays leading to loss of indication and/or control combined with failure to shut down the plant remotely. The outside cable spreading room sequence consisted of the seismically induced failure of cable trays leading to loss of power to all essential equipment. The dominant seismic contributors to the estimated CDF are operator actions (e.g., failure to shut down the plant from the remote shutdown panel, failure to provide service water backup to auxiliary feedwater (AFW) pump suction), seismic faults that lead directly to core damage (e.g., failures of cable trays, surrogate element (e.g.,

failure of soils, building structures), and failures of critical equipment (e.g., transformers,480V

- load centers, level transmitter for condensate storage tank (CST)).

The fire CDF is dominated by fires in the control room, the cable spreading room, the AFW pump room, the gas turbine room, the vital and nonvital switchgear rooms, the diesel generator rooms, and the monitor tank room. All fire event sequences were quantified assuming all equipment / cables in the area would fail by the fire. However, no fire in a single area or compartment (by itself), except the cable spreading room, would have the potential of directly

' causing core damage, in addition to the damage caused by the fire in these areas and compartments, other failures (e.g., a reactor coolant pump (RCP) seal loss of-coolant accident, operator failure to establish RCP seal injection via the refueling water storage tank, failure of the safety relief valves), in combination with a fire, contributed to the calculated fire CDF.

4 The extemal flood CDF is dominated by flooding as a result of the rising water level of Lake Michigan, combined with wind wave effects and water runup, which would result in

. flooding of the turbine building and leads to loss of the ability to remove decay heat. The dominant tornado / wind-induced core damage sequences were determined to involve failures of the diesel generator exhaust stacks, which lead to failure of both diesel generators.

The licensee's IPEEE analyses appear to have examined the significant initiating events and

' dominant accident sequences at the plant.

Containment Performance

- The licensee performed both qualitative and quantitative assessments of the containment performance under seismic conditions at Point Beach. The qualitative assessment examined containment safeguards systems significant to large early release, including containment integrity, mechanical penetrations, containment isolation, and containment cooling. The quantitative assessment focused on bypass and containment isolation failure sequences using the Point Beach IPE plant damage states, release categories, and containment response analysis. The licensee estimated the frequency of early release to be 2.1E-6/RY, and the dominant contributors were failures of containment isolation due to the loss of either AC or DC power and failures of the containment low and intermediate range pressure transmitters due to

the impact of the seismically induced collapse of the adjacent block walls.

The licensee's evaluation of containment performance with regard to fire hazard focused on assessing the potential for containment isolation or bypass failure. The licensee identified and evaluated areas that could potentially impact the function of containment isolation.' The licensee concluded that fire is not expected to result in any significant containment isolation ,

failure or bypass. I

)

The licensee's containment performance analyses for seismic and intemal fire events appear to have considered important containment performance issues and are consistent with the intent of Supplement 4 to GL 88-20.

Generic Safety Issuer As part of this IPEEE, a set of generic and unresolved safety issues (USl A-45, GSI-131, GSI-103, GSI-57, and the SNL FRSS issues) were specifically identified in Supplement 4 to GL 88-20 and its associated guidance in NUREG-1407 as needing to be addressed in the

- lPEEE. The NRC staffs evaluation of these issues is provided below.

1. USl A-45, " Shutdown Decay Heat Removal Requirements" The licensee has chosen to subsume its resolution of USI A45 in the USI A-46 program.

However, the licensee provided a brief discussion of decay heat removal (DHR) capabilities following a seismic event in its IPEEE submittal. The seismic PRA modeled the systems available for DHR, and the CDF contributions due to failures of DHR systems and support systems were tabulated in the IPEEE submittal. The licensee credited the AFW system for providing DHR capability under fire conditions and provided the analysis to address the USl A-45 issue in the IPEEE submittal. The NRC staff finds that the l

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l licensee's USl A-45 evaluation is consistent with the guidance provided in Section 6.3.3.1 of NUREG 1407 and, therefore, the NRC staff considers this issue resolved.

2. GSI-131, " Potential Seismic Interaction involving the Movable In-Core Flux Mapping System Used in Westinghouse Plants" Even though this issue is not directly applicable to Point Beach because the flux mapping cart is not movable, as stated in the submittal, the licensee provided a discussion on the seismic capability of the cart. The licensee pointed out that the Point Beach cart is identical to the Kewaunee Nuclear Power Plant flux mapping cart, which was analyzed under the Kewaunee IPEEE. Since the seismic demands for both plants are quite similar, and the Kewaunee IPEEE result showed that the cart possesses adequate seismic i capability, the licensee concluded that the Point Beach cart also has adequate seismic capability, in addition, the licensee pointed out that two lateral struts, added to the cart for lateral restraints by a previous modification, will increase its capability under seismic conditions. The NRC staff considers this issue resolved for Point Beach on the basis of the information provided in the IPEEE and the TER.

{

3. GSi-103," Design for Probable Maximum Precipitation" l As part of the IPEEE flood analysis, the licensee reviewed and updated the results of the USI A-45 study, including the assessment of GSI-103 (local precipitation in Section 5.2.5 and roof ponding in Section 5.2.6 of the IPEEE submittal), and concluded that the .

l . frequency of Probable Maximum Precipitation (PMP) events is below 1E-6/RY and has ,

! Insignificant impact on Point Beach. The NRC staff finds that the licensee's GSI-103 '

evaluation is consistent with the guidance provided in Section 6.2.2.3 of NUREG-1407 and, therefore, the NRC staff considers this issue resolved.

4. Fire Risk Scoping Study issues The licensee has explicitly addressed the SNL FRSS issues (Section 4.8 of the IPEEE I

submittal). The NRC staff finds that the licensee's evaluations are consistent with the guidance provided in FIVE, which was accepted by the NRC staff and, therefore, the NRC staff considers these issues resolved. i

5. GSI-57," Effects of Fire Protection System Actuation on Safety-Related Equipment"

, Although the licensee's IPEEE submittal did not explicitly discuss GSI-57, the information provided in the submittal (Section 4.8) addressing seismic-fire interactions and total environment equipment survival is related to this issue. The NRC staff finds that the licensee's evaluation is consistent with the guidance provideci in NUREG-1407 and, therefora, the NRC staff considers this issue resolved.

In addition to those safety issues discussed above that were explicitly requested in i Supplement 4 to GL 88-20, four generic safety issues were not specifically identified as issues that should be resolved under the IPEEE program; thus, they were not explicitly discussed in Supplement 4 to GL 88-20 or NUREG-1407. However, subsequent to the issuance of the I generic letter, the NRC evaluated the scope and the specific information as requested in the

i I

6 generic letter and the associated guidance, and concluded that the plant-specific analyses being requested in the IPEEE program could also be used, through a satisfactory IPEEE submittal review, to resolve the extemal event aspects of these four safety issues. The following discussions summarize the NRC staff's evaluation of these safety issues at Point Beach.

i

1. GSI-147,

• Fire-Induced Alternate Shutdown / Control Room Panel Interactions" The licensee's IPEEE submittal contains information (Section 4.8.5 of the Point Beach IPEEE submittal) addressing this issue. The licensee reviewed the circuits associated with Point Beach attemate shutdown capability and did not discover any control system interaction concerns. The hot short concems (e.g., the possibility of inadvertent opening of a power-operated relief valve of an atmospheric steam dump valve) were discussed in its response to an NRC staff RAI (Section 2.4.1 of the attached TER). Based on the results of the IPEEE submittal review, the NRC staff considers that the licensee's process is ,

capable of identifying potential vulnerabilities associated with this issue. On the basis that I no vulnerability associated with this issue was identified in the IPEEE submittal, the NRC staff considers this issue resolved for Point Beach.

2. GSI-148, " Smoke Control and Manual Fire-Fighting Effectiveness" The licensee addressed this issue in Section 4.8.3 of its IPEEE submittal. The licensee provided a discussion of the Point Beach's fire protection program and its associated fire brigade training program related to this issue. Based on the results of the IPEEE submittal <

review, the NRC staff considers that the licensee's process is capable of identifying potential vulnerabilities associated with this issue. On the basis that no vulnerability associated with this issue was identified in the IPEEE submittal, the NRC staff considers this issue resolved for Point Beach.

3. GSI-156, " Systematic Evaluation Program (SEP)"

l The licensee's IPEEE submittal contains information directly addressing the following external-events-related SEP issues: sett'ement of foundations and buried equipment ,

(Sections 3.1.2.1,3.1.3.10, and 3.1.4.2); dam integrity and site flooding (Section 5.2.2); i site hydrology and ability to withstand floods (Section 5.2); industrial hazards (Section 5.3); l tomado missiles (Section 5.1) and severe weather effects on structures (Sections 5.1 and  ;

5.2). Although the licensee's IPEEE submittal did not contain information explicitly I addressing the other SEP issues (i.e., design codes, criteria, and load combinations; and seismic design of structures, systems, and components), a conclusion may be drawn that these issues were implicitly addressed as part of the licensee's IPEEE process at Point Beach because the seismic PRA has taken the as-built, as-operated conditions into consideration in its structural response and component fragility analyses, and no potential seismic vulnerabilities were identified. Based on the results of the IPEEE submittal review, the NRC staff considers that the licensee's process is capable of identifying potential vulnerabilities associated with this issue. 'On the basi,s that no vulnerability associated with 1 this issue was identified in the IPEEE submittal, the NRC staff considers this issue '

resolved for Point Beach.

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4. GSI-172, " Multiple System Responses Program (MSRP)"

The licensee's IPEEE submittal contains information directly addressing the following )

external events-related MSRP issues: (1) effects of fire protection system actuation on 1 non safety-related and safety-related equipment (Sections 4.8.1 and 4.8.4) (2) seismically induced fire suppression system actuation (Section 4.8.1), (3) seismically induced fires

- (Sections 3.1.3 and 4.8.1), (4) effects of hydrogen line rupture (Sections 4.12 and 5.3.3),

(5) the IPEEE-related aspects of common cause failures related to human errors

. (Sections 3.1.3, 3.1.5, 4.6.2,4.8.4, 5.1.5, and 5.1.6), (6) non-safety-related control system / safety-related system dependencies (Sections 3.1.2,3.1.3,3.1.4, and 4.8.1), 1 (7) effects of flooding and/or moisture intrusion on non-safety-related and safety-related equipment (Sections 3.1.3 and 4.8.1), (8) seismically induced spatial / functional interactions (Sections 3.1.2, 3.1.3, 3.1.4, and 4.8.1), (9) seismically induced flooding (Sections 3.1.3 and 4.8.1), (10) seismically induced relay chatter (Sections 3.1.2 and 3.1.4), and (11)

. evaluation of earthquake magnitude greater than safe shutdown earthquake (Section 3).  ;

Based on the results of the IPEEE submittal review, the NRC staff considers that the licensee's process is capable of identifying potential extemal events-related vulnerabilities associated with GSI-172. On the basis that no potential vulnerability associated with this issue was identified in the IPEEE submittal, the NRC staff considered the IPEEE-related aspects of this issue resolved for Point Beach.

Unious Plant Features. Potential Vulnerabilities. and imorovements; The licensee did not identify any unique safety features at the plant.

The licensee's definition of vulnerability associated with external events is the same as that used in their IPE. The licensee's criteria for determining if any plant vulnerability exists are: a CDF goal of less than 1E-4/RY and a goal for frequency of large fission product release of less than 1E-6/RY. Although the IPEEE did not identify any vulnerabilities associated with external events at Point Beach, the licensee has implemented certain plant modifications and proposed to take certain actions related to relay and anchorage issues. These licensee-identified improvements and proposed actions are listed below:

High Winds e The diesel generator stacks were modified to accommodate higher winds. i Fire eL The controllogic for automatic start of the motor-driven AFW pumps was modified such that a fire in the nonsafeguards switchgear will not prevent the motor-driven pumps from starting.

e The procedure for mitigating fires in the control room and cable spreading room was revised to include verification of closure of an additional containment isolation valve.

8 Seismic e Numerous equipment anchorages (e.g., transformers, a reactor trip breaker cabinet, inverters and battery chargers) were fixed for seismic concems (see SER on USl A-46 resolution, dated July 7,1998),

o The seismic capacity concems of certain tank anchorages were either reanalyzed or upgraded by modification (see SFR on USl A-46 resolution).

e The supports for certain cable and conduit raceways in the cable spreading room were modified to increase their seismic capacities (see SER on USl A-46 resolution).

• Certain relays, identified as outliers during an IPEEE/A-46 walkdown, will be reevaluated and resolved by t5e fall of 1999, except those pertaining to the diesel generators. The licensee plans to replace the controls for the diesel generators by the year 2002 (see SER on USI A-46 resolution).

In addition, the licensee stated that they intend to use the Point Beach IPEEE as a decision-making tool in many aspects of engineering support and plant operations as long as the tool adds value (Section 8 of the IPEEE submittal). The licensee also stated that a clear understanding of the assumptions, limitations, uncertainties is required in order to avoid misinterpretation of the IPEEE results. Although the implications of using surrogate element in the seismic PRA (e.g., limitations, uncertainties) were not discussed in the submittal (see Section 3.1 of Attachment 2), the impact on finding the dominant sequences at Point Beach may not be very significant. As discussed previously in the section on dominant contributors, the ranking of the dominant seismic sequences would not be significantly changed. Therefore, the NRC staff believes that, as a whole, the licensee has gained some significant insights related to potential vulnerabilities to severe accidents through the IPEEE.

Ill. CONCLUSION On the basis of the above findings, the NRC staff notes that (1) the licensee's IPEEE is complete with regard to the information requested by Supplement 4 to GL 88-20 (and associated guidance in NUREG-1407), and (2) the IPEEE results are reasonable given the Point Beach design, operation, and history. Therefore, the NRC staff concludes that the licensee's IPEEE process is capable of identifying the most likely severe accidents and severe accident vulnerabilities, and therefore, that the Point Beach, Units 1 and 2, IPEEE has met the intent of Supplement 4 to GL 88-20 and the resolution of specific generic safety issues discussed in this SER.

i

9 it should be noted that the NRC staff focused its review primarily on the licensee's ability to examine Point Beach for severe accident vulnerabilities. Although certain aspects of the IPEEE were explored in more detail than others, the review is not intended to validate the accuracy of the licensee's detailed findings (or quantification estimates) that stemmed from the examination. Therefore, this SER does not constitute NRC approval or endorsement of any IPEEE material for purposes other tha., those associated with meeting the intent of Supplement 4 to GL 88-20 and resolving the generic safety issues discussed in Section 11 of this SER.

Attachments: 1. Technical Evaluation Report

2. Supplemental Technical Evaluation Report Date: September 15,1999 S

l