ML20206D333

SER of Individual Plant Examination of External Events Submittal on Susquehanna Steam Electric Station,Units 1 & 2. Staff Notes That Licensee IPEEE Complete with Regard to Info Requested by Suppl 4 to GL 88-20

ML20206D333
Person / Time
Site:	Susquehanna
Issue date:	04/27/1999
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML17164B026	List: ML20206D333 ML20206D341
References
REF-GTECI-A-45, REF-GTECI-DC, TASK-A-45, TASK-OR GL-88-20, NUDOCS 9905040056
Download: ML20206D333 (8)

Text

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STAFF EVALUATION REPORT OF INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE) SUBMITTAL ON SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2

1.0 INTRODUCTION

On June 28,1991, the NRC issued Generic (GL) Letter 88-20, Supplement 4 (with NUREG- l 1407, Procedural and Submittal Guidance) requesting all licensees to perform individual plant examinations of extemal events (IPEEE) to identify plant-specific vulnerabilities to severe

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

improvements and corrective actions. In a letter dated June 27,1994, the licensee, ,

Pennsylvania Power & Light Company (PP&L), submitted its response to the NRC.

A " Step 1" review was performed by NRC's contractor, Energy Research, Inc. (ERI), to examine the licensee's IPEEE submittal for completeness and reasonableness considering the design and operation of Susquehanna. As a part of the Step 1 review, NRC staff sent a request for additional information (RAI) to the licensee in June 1996, and the licensee responded to the RAIin August 1996. On the basis of the Step 1 review, and further review by a senior review board (SRB), the staff concluded that the aspects of high winds, floods, and transportation and other extemal events were adequately addressed. However, the seismic and fire IPEEEs needed a more detailed, Step 2, review because of specific concerns related j to seismic and fire analyses (e.g., seismic capacity calculations, outlier resolutions, unusually j low fire core damage frequency (CDF) results, compartment screening criterion used in the analysis, and human error analysis). For a more detailed discussion of these concems, see Section 3 of the attached TER. The SRB is comprised of RES and NRR staff and RES consultants (Sandia National Laboratories) with probabilistic risk assessment expertise for )

extemal events.

NRC provided a site audit plan for the Step 2 review to the licensee in Mar h 1998 to facility the licensee's preparation for the site audit. The audit plan focused on those seismic and fire concems identified during the Step 1 review. The site audit was held on August 4-5,1998, and included technical discussions with the licensee's staff, review of the licensee's technical analyses and documentation held at the site, and walkdowns of the plant. The licensee responded to all seismic and fire issues identified in the audit plan by means of verbal responses during the site audit, as well as by means of interactive seismic and fire walkdowns.

In addition, in October 1998, the licensee submitted additional information that had been verbally requested during the audit. ERI completed the supplement to its TER in December 1998. The staff's review findings are summarized in the attached SER, and the details of the contractor's findings are presented in the TER and its supplement which appear in attachments to the SER.

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Enclosure

e 0 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 Requirements," Generic Safety issue (GSI) 57, " Effects of Fire Protection System Actuation on Safety-Related Equipment," the NUREG/CR-5088 Fire Risk Scoping Study issues, and GSI-103, " Design for Probable Maximum Precipitation." These issues 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 GSls as part of the Susquehanna IPEEE.

2.0 EVALUATION The Susquehanna Steam Electric Station (SSES) consists of two General Electric BWR-4 reactor units, each having a Mark-Il containment. The seismic category I structures, except diesel generator "E" building, were designed to a seismic acceleration level of 0.1g PGA (peak ground acceleration) with a spectral shape conforming to Regulatory Guide 1.60, Rev.1. The diesel generator "E" building, which is founded in soil, was designed to a seismic acceleration level of 0.15g PGA. The plant was categorized in NUREG-1407 as a 0.3g focused-scope plant. The licensee performed a seismic margin analysis (SMA) using the Electric Power Research Institute's (EPRI) method to evaluate seismic vulnerabilities. As part of EPRI's seismic margin methodology (SMM), the licensee relied on plant walkdowns that focused on evaluating expected functional performance, component anchorage capability, and the potential for adverse seismic-caused spatial interactions. All accessible components were at least " walked-by," whereas at least one component of a given type was subject to a more detailed walkdown. Detailed walkdowns were performed for Unit-1 equipment; walk-bys were l performed for Unit-2 equipment. (The licensee expected that Unit-2 equipment is just as seismically nJgged as Unit-1 equipment.)  !

The licensee has performed a fire probabilistic risk assessment following the general approach described in NUREG/CR-2300 to evaluate fire risk. The licensee employed several data bases for its multi-stage screening and detailed quantitative analysis of potential fire events, including I

estimates of vessel and containment failure frequencies. The licensee employed the progressive screening assessment as described in NUREG-1407 to evaluate risks due to high winds, extemal floods, and transportation and nearby facility accidents any a qualitative containment performance assessment for the Susquehanna IPEEE. 4 Core Damaae Freauency and Seismic Capacity Estimates The licensee estimated the plant seismic capacity, in terms of high confidence of low probability of failure (HCLPF) value to be 0.21g due to a low seismic capacity of certain i components. However, the licensee stated that these low-capacity components rre either not I strictly required for safe shutdown of the plant or their failures may be rectified through manual recovery actions, and the licensee concluded that the plant HCLPF capacity meets the 0.3g review level earthquake (RLE).

The licensee originally estimated a fire CDF of 1E-9 per refueling cycle (12-18 months), which is significantly smaller than CDF values reported by other comparable studies of similar plants.

The reason for this smaller CDF, as stated in the licensee's response to a RAI, was that the

g small human error probabilities' used in the original Individual Plant Examination (IPE) internal events analysis has led to the small overall CDF estimate for fire events. In response to an NRC RAI, the licensee performed additional analyses of the dominant fire scenarios using new screening criteria and revised human error probabilities. The methodology and results were reviewed during the site audit. As a result of the concerns that were raised with regard to assumptions about independence among events, the licensee subsequently performed a sensitivity analysis on the upper cable spreading rooms, which resulted in an estimated fire CDF of 2.4E-7 per reactor year (RY) for this area. This CDF is three orders of magnitude greater than the CDF originally reported for this area. Although the licensee did not perform additional sensitivitjanalyses to revise the results for all other scenarios, similar results could be anticipated for those scenarios. This may be an indication that the fire CDF estimate at Susquehanna could be three orders of magnitude higher than the originally reported CDF.

The licensee estimated that the CDF due to intemal events is about SE-7/RY.

Based on Susquehanna's conformance to the 1975 Standard Review Plan, the licensee corisidered other extemal events (e.g., external floods and high winds) to be insignificant contributors to severe accidents at the Susquehanna site.

Dominant Contributors For the seismic IPEEE, the licensee developed a success path logic diagram that describes the plant functions needed to achieve and maintain the plant in a safe shutdown condition for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The preferred success path relies on high-pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) for inventory control, whereas the alternate success path relies on the automatic depressurization system (ADS) and residual heat removal (RHR) in the low-pressure coolant injection (LPCI) mode. Both success paths rely on reactor scram for reactivity control. These success paths are consistent with the success paths for a BWR-4 plant as described in EPRI's SMM guide. The HCLPF capacity of components in each success path, a HPCI discharge valve and a RHR suppression pool inlet valve, was estimated to be 0.21g. j The fire CDF is dominated by fires in the northem section of the reactor bu,ilding, a battery charger area in the control structure, the upper and lower cable spreading rooms in the control structure, and the main control room. The important system / equipment contributors to the estimated fila CDF that appear in the top sequences are mostly associated with the loss of cables for high pressure coolant injection, ESW systems, DC channels, and RHR systems.

The licensee's fire analysis contains a significant weak point involving the assumption that the severity of a fire and the probability of failure of fire suppression are independent, which contributed to the unusually low CDFs for all fire scenarios. This assumption does not take into account the possibility of damage before suppression takes place. However, since the ranking of the significant scenarios were based on relative values of CDFs, not the absolute values, and this assumption has been applied consistently across all significant fire scenarios, the impact of these optimistic CDF values on the final outcome (i e., relative ranking of scenarios) may be minimal. Therefore, despite this improper assumption in the fire analysis.

• These estimates were modified in the revised IPE analysis as reflected in the IPE SER Supplement, dated August 11,1998.

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the licensee has identified the significant scenarios and dominant accident sequences at the plant and has not missed any potential fire vulnerabilities.

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Corf.ainment Performance The licensee has performed a qualitative assessment of containment performance under seismic conditions at Susquehanna focusing primarily on identifying failure modes unique to seismic events. The licensee has evaluated the possible seismic impacts on structures and components, including containment hatches and drywell head seals, that provide the containment funcbon and concluded that these structures and components are capable of withstanding the RLE without any adverse effects on containment performance. The licensee also has performed seismic containment walkdowns, including an inspection of the containment penetration seals and containment isolation va!ves. The licensee also has assessed the possible containment failure modes caused by fire, including the effect of hot shorts on isolation valves, and concluded that fire is not expected to result in containment failure modes different than those identified in the IPE for internal events.

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

Genenc Safety issues As a part of the IPEEE, a set of generic and unresolved safety issues (USl A-45, GSI-131, GSI-103, GSI-57, and the Sandia Fire Risk Scoping Study [FRSS) issues) were identified in Supplement 4 to GL 88-20 and its associated guidance in NUREG-1407 as needing to be addressed in the IPEEE. The staff's evaluation of these issues is provided below.

1. USl A-45, " Shutdown Decay Heat Removal Requirements" The licensee's success paths developed for the Susquehanna seismic IPEEE address

' decay heat removal requirements following a seismic margin earthquake via the suppression pool cooling mode of RHR. In the fire area, the decay'fieat removal requirements are very similar to that used for the IPE for intemal events. The staff ]

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

2. GSI-131, " Potential Seismic Interaction involving the Movable in-Core Flux Mapping System Used in Westinghouse Plants" The nuclear steam supply system for each unit at Susquehanna consists of a General Electric BWR, hence, GI-131 is not applicable.

3. . GSI-103, " Design for Probable Maximum Precipitation" The licensee has addressed GSI-103 (Section 5.2.5 of the IPEEE submittal) and concluded that the Probable Maximum Precipitation (PMP) criteria will not have any

-5 impact on Susquehanna. The 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 staff considers this issue resolved.

4. . Fire Risk Scoping Study issues

' The licensee has explicitly addressed the Sandia Fire Risk Scoping Study issues (Section 4.8.2 of the IPEEE submittal). The staff finds that the licensee's evaluation is consistent with the guidance provided in NUREG-1407 and, therefore, the 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 addressing seismic-fire interactions and the total environment equipment survival (Section 4.8.2.1) is the primary focus of this issue.

The staff finds that the licensee's evaluation is consistent with the guidance provirled in NUREG-1407 and, therefore, the staff considers the 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 to be resolved under tne !PEEE 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 requested in the genenc letter and the associated IPEEE 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 staff's evaluations of these safety issues at Susquehanna.

1. GSI-147, " Fire-Induced Attemate Shutdown / Control Room Panel Interactions" The licensee's IPEEE submittal contains information (Section 4.8.2'11 of the Susquehanna IPEEE submittal) addressing this issue. The licensee performed a review following the guidance provided in EPRI TR-100370, " Fire-induced Vulnerability Evaluation (FIVE)," concerning control system interactions, and concluded that all circuitry associated with remote shutdown was electrically independent of the control room. Based on the results of the IPEEE submittal review. the 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 staff considers this issue resolved for Susquehanna.

2. GSI-148, " Smoke Control and Manual Fire-Fighting Effectiveness" The licensee's IPEEE submittal contains information (Section 4.8.2.1) addressing this issue. The licensee performed a review of the Susquehanna's fire protection program and its associated fire brigade training program related to this issue. Based on the

n , u g results of the IPEEE submittal review, the 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 staff considers this issue resolved for Susquehanna.

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

Susquehanna is not a SEP plant, hence, GSI-156 is not applicable.

4. 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 safety and non-safety related equipment (Section 4.8.2.1), (2) seismically induced fire suppression system actuations (Sec. 4.8.2.1 and 4.2.2.2 of the submittal, and response to RAl), (3) heat / smoke / water propagation effects from fire (Sec. 4.8.2.1), (4) seismic-fire interactions (Sec. 4.8.2.1), (5) effects of hydrogen line rupture (Sec. 4.8.2.1), (6) the IPEEE-related aspects of common cause failures related to human errors (Sec.

3.5.2.5), (7) non-safety-related control system / safety-related system dependencies (Sec. 3.4.9, and 3.10.5, and 4.8.2.1), (8) effects of flooding and/or moisture intrusion on non safety related and safety-related equipment (Sec. 4.8.2.1 and 5.2), (9) seismically induced spatial interactions (Sec. 3.4.9 and 3.10.5), (10) seismically induced flooding (Sec. 4.8.2.1 and 4.2.2.2 of the submittal, and response to RAl), (11) seismically induced relay chatter (Sec. 3.5.3.2,3.10, and 3.11.2), and (12) evaluation of earthquake magnitude greater than safe shutdown earthquake (Sec. 3).

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

Uniaue Plant Features. Potential Vulnerabilities. and Improvements '4 The submittal did not identify any unique safety features at the plant. The licensee did not provide a definition of severe accident vulnerability and did not identify any seismic, fire, or i other extemal events-related severe accident vulnerabilities.

In the seismic and fire areas the licensee has implemented some procedural changes and a 0 number of minor equipment enhancements in response to the Susquehanna IPEEE walkdown findmgs. These licensee-identified improvements are listed below.

1. . Small" trolleys," used to assist the removal of breakers from AC and DC switchgear cabinets that are located on the top of some of these cabinets, were removed to prevent adverse seismic spatial interaction.

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2. At several locations inside the control and relay rooms, control cabinets and instrumentation panels were bolted together to prevent adverse interaction and potential relay chatter.

3. The color video CRTs in certain control room cabinets were anchored to the panel internal supports to prevent them from sliding under seismic events.

4. Seismic housekeeping improvements and a seismic housekeeping procedure were implemented at the plant.

5. Seismic awareness training was implemented for the plant staff.

6. Additional seismic restraints (a second ring attached to the wall) were added to hydrogen bottles formally restrained only by a single ring.

7. Proceduralimprovements (e.g., restrictions on combustible materials storage, ban on smoking inside buildings, special covers for barrels, and opening of the floor drain in the lower cable spreading room to allow sprinkler system water to drain) were implemented as a result of fire IPEEE.

There were no modifications or improvements in the HFO area.

3.0 CONCLUSION

S On the basis of the above findings, the staff notes that (1) the licensee's IPEEE is complete with regard to the information requested by Supplement 4 to GL 886 (and associated guidance in NUREG-1407), and (2) the IPEEE results are reasonable given the Susquehanna design, operation, and history. Therefore, the 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 Susquehanna IPEEE has met the intent of Supplement 4 to GL 88-20 and the resolution of specific generic safety issues discussed in this SER.

It should be noted that the staff focused its review primarily on the license ('s ability to examine Susquehanna 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 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 Principal Contributor: J. Chen Date: ' April 27, 1999

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TECHNICAL EVALUATION REPORT ON THE REVIEW OF THE INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE) SUBMITTAL AT SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2 '

AttachmerJ 1 4 --

r ATTACHMENT 1 TECHNICAL EVALUATION REPORT ON THE REVIEW OF THE INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE) SUBMITTAL AT SUSOUEHANNA STEAM ELECTRIC STATION . UNITS 1 AND 2 l

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,c ERl/NRC 95-512 e '

TECHNICAL EVALUATION REPORT ON THE

" SUBMITTAL-ONLY" REVIElu OF THE INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS AT SUSQUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2 FINAL REPORT February 1998

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i Energy Research, Inc.

P.O. Box 2034 Rockville, Maryland 20847-2034 Work Performed Under the Auspices of the United States Nuclear Regulatory Commission Office of Nuclear Regulatory Research Washington, D.C. 20555 Contract No. 04-94-050 t.,,iicwarairra -m

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• ERl/NRC 95-512 TECilNICAL EVALUATION REPORT ON Tile "SUMMITTAleONLY" REVIEW OF Tile INDIYlDUAL PLANT EXAMINATION OF EXTERNAL EVENTS AT SUSQUEllANNA STEAM ELECTRIC STATION, UNITS I AND 2 FINAL REPORT February 1998 M. Khatib-Rahbar Principal Investigator Authors:

R. T. Sewell, M. Kazarians', and J. A. Lambright' Energy Research, Inc, P.O. Box 2034 Rockville, Maryland 20847 k

Work Performed Under the Auspices of the United States Nuclear Regulatory Commission Office of Nuclear Ryulatory Rewarch Washington, D.C. 20555 Contract No. 04-94-050 Kazarians and Associates,425 East Colorado Street, Suite 545, Glendale, CA 91205 3

Formerly of Beta Corporation International, presently with Lambright Technical Associates,9009 Lagrima De Oro Road, NE. Albuquerque, NM 8711t

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TABLE OF CONTENTS EXECUTI VE SU M M ARY . . . . . . . . . . . . . . . . . . . . . . . . . . ....................vi PR E FA C E , . . . . . . . . . . . .. . . . . . . . . . . . . . ..... ........................xiii A BBR EVI ATIONS ~ . . . . . . . . . . . . . . . . . . , . . . . . . . . ... . .............. xiv 1 INTRODUCTION . . . . . . .........................................I 1.1 Plant Characterization ............................ . .........I 1.2 Overview of the Licensee's IPEEE Process and Important Insights . . . . . . . . . . . 2 1.2.1 Seismic . . . . . . . . .................. .......... .....2 1.2.2 Fire . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..............3 1.2.3 HFO Events ....... ............. ............ .....3 1.3 Overview of Review Process and Activities . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.3.1 S eis mi c . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.3.2 Fire..............................................5 1.3.3 HFO Events ..................... ...................5 2 CONTRACTOR REVIEW FINDINGS .................................6 2.1 S eis m i c . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 2.1.1 Overview and Relevance of the Seismic IPEEE Process . . . . . . .. . . . . . 6 2.1.2 Success Paths and Component List . . . . . . . . . . . . . . . . . . . . . . . . . . 6 2.1.3 Non-Seismic Failures and Human Actions . . . . . . . . . , , .. . . . . . . . . 7 2.l A S eismi c i nput . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 2.1.5 Structural Responses and Component Demands . . . . . . . . . . . . . . . . . . 8 2.1.6 Screening Criteria . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . 8 2.1.7 Plant Walkdown Process ................................9

-2.1.8 Evaluation of Outliers . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 2.1.9 Relay Chatter Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 2.1.10 Soil Failure Analysis . . . . . . . ........................... 12 2.1.11 Containment Performance Analysis . . . . . . . .................. 12 2.1.12 Seismic-Fire Interaction and Seismically Induced Flood Evaluations . . . . . 12 2.1.13 Treatment of USI A-45 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 2.1.14 Other Safety issues . . . . . . . . . . . . . . . . . . . ....g.......... 14 2.1.15 Peer Review Process . . . . . . . . . . . . . . . . , . . . . . ...........15 2.1.16 Summary Evaluation of Key Insights . . ..... ................ 15 2.2 Fire . . . . . . . . . . . . . . . . . . . . . . . ......................... 16 2.2.1 Overview and Relevance of the Fire IPEEE Process . . . . . . . . . . . . . . . 16 2.2.2 Review of Plant Information and Walkdown . . . .. ............17 2.2.3 Fire-Induced initiating Events . . . . . ............ ......... 18 2.2.4 Screening of Fire Zones . . . . ..................... .... . 18 2.2.5 Fire Hazard Analysis . . . . .......................... . 19 2.2,6 Fire Growth and Propagation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 2.2.7. Evaluation of Component Fragilities and Failure Modes ........... 21  !

2.2.8 Fire Detection and Suppression . . . . . . . . . . . . . . . . . . . . . . . . . . 22 )

Energy Research, Inc. ii ERl/NRC 95-512

6 u 2.2.9 Analysis of Plant Systems and Sequences .. . . ..... . 22 2.2.10 Fire Scenarios and Core Damage Frequency Evaluation .... ... . 22 2.2.11 Analysis of Containment Performance . . . .. . ...... ... . 23 2.2.12 Treatment of Fire Risk Scoping Study Issues . ............... 23 2.2.13 USI A-451ssue . . . . . .......... ........ ..... .... 24 2.3 HFO Events . .... ........... .......... ... .... .. . 24 l 2.3.1 liigh Winds and Tornadoes . . . . . . . .... ... .. . ... 25 1

2.3.1.1 General Methodology . . ............ ... .. 25 2.3.1.2 Plant. Specific Hazard Data and Licensing Basis . . .. 25 2.3.1.3 Significant Changes Since issuance of the Operating License . . . . .. . ....... . . ... . ... 26 2.3.1.4 Significant Findings and Plant-Unique Features ... 26 2.3.1.5 Hazard Frequency . . . .... .. .... . .. .. 26 2.3.2 External Flooding . . . ................... .. ...... 26 2.3.2.1 General Methodology . . . . . . . . . . . . . . . . . . . . 26 2.3.2.2 Plant-Specific Hazard Data and Licensing Basis ..... . 26 2.3.2.3 Significant Changes Since issuance of the Operating License . . . . . . . . . . . . . . . ..... ...... . 27 2.3.2.4 Significant Findings and Plant-Unique Features . . 27 2.3.2.5 Hazard Frequency . . . . .. . . . . . . . . . . . . . . . .. . 27 2.3.3 Transportation and Nearby Facility Accidents .. . ....... ....... 27 2.3.3.1 General Methodology . . . . . . . . ......... . .. 27 2.3.3.2 Plant-Specific Hazard Data and Licensing Basis .... . 28 2.3.3.3 Significant Changes Since issuance of the Operating Lic ens e . . . . . . . . . . . . . . . . . . . . . . . . . ...... 28 2.3.3.4 Significant Findings and Plant-Unique Features .. ... 28

.2.3.3.5 Hazard Frequency . . . ................... . 28 2.3.4 Other Ev e nts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 2.4 Generic Safety issues (GSI-147, GSI 148, and GSI-172) ................ 29 2.4.1 GSI-147, " Fire-induced Alternate Shutdown / Control Panel Interaction" . 29 2.4.2 GSI-148, " Smoke Control and Manual Fire Fighting Effectiveness" . . . . 29 2.4.3 GSI-172, " Multiple System Responses Program (MSRP)" ..... ... 29 3 OVERALL EVALUATION AND CONCLUSIONS . . . . . . . . . . . . . . . . . . . . . . . . 34 3.1 Seismic . . . . . .. .......... ...... ........

4........ 34 3.2 Fire . . . . . . . ..... . ....... .. ................. ... 35 3.3 HFO Events . ....... ..... .. . ...................... 37 4 IPEEE INSIGHTS, IMPROVEMENTS, AND COMMITMENTS . . . . . .. .. . 38 4.1 Seismic . . . . . . . . .............. .......... ....... .. 38 4.2 Fire . . . . . . . . . . . . .............. ........ . ........ . 40 4.3 H.FO Events . . . . . . . . . . . . .... ... . .. ..... . . ... 40 5 IPEEE EVALUATION AND DATA

SUMMARY

SHEETS . .......,......... 48 Energy Research, Inc. iii ERI/NRC 95-512

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6 REFERENCES. - . . - .. .. . ,, - . .. . 53 i

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f.IST OF TABLES i

Table 4.1 Open issues, and Their Resolutions, as identified in the Susquehanna Seismic IPEEE . .. . . ........ .. .... . .. . . . . . . 41 Table 5.1 External Events Results . . . . . ...... . ....... . .. . . .. 49 Table 5.2 SMM Seismic Fragility ... .... . . .. . ....... ... . 50 Table 5.3 BWR Seismic Success Paths . .. .. ... .. . . .. ... 51 l

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Energy Research, Inc. v ERl/NRC 95-512

gg u EXECUTIVE SUAINIARY This technical evaluation report (TER) documents a " submittal-only" review of the individual plant examination of external events (IPEEE) conducted for the Susquehanna Steam Electric Station (SSES),

Units I and 2. This technical evaluation review was performed by Energy Research, Inc. (ERI) on behalf of the U.S. Nuclear Regulatory Commission (NRC). The submittal-only review process consists of the -

following tasks:

• Examine and evaluate the licensee's IPEEE submittal and directly relevant available documentation.

• Develop requests for additional information (RAls) to supplement or clarify the licensee's IPEEE submittal, as necessary.

Examine and evaluate the licensee's responses to RAls.

Conduct a final assessment of the strengths and weaknesses of the IPEEE submittal, and develop review conclusions.

This TER documents ERI's qualitative assessment of the SSES, Units 1 and 2, IPEEE submittal, particularly with respect to the objectives described in Generie Letter (GL) 88-20, Supplement No. 4, and the guidance presented in NUREG-1407.

The licensee of Susquehanna Steam Electric Station is Pennsylvania Power & Light Company (PP&L).

The Susquehanna IPEEE submittal considers seismic; lire; and high winds, floods and other (HFO) external initiating events. Overall management of the Susquehanna IPEEE was controlled by PP&L personnel (Systems Analysis Group). The IPEEE technical work was performed by PP&L staff with assistance provided by consultants. SSES personnel and on-site engineers provided input and assistance to the IPEEE team. An independent peer review was conducted of the IPEEE analyses by personnel in various PP&L functional organizations / departments.

Licensee's IPEEE Process With respect to the seismic IPEEE, NUREG-1407 has assigned Susquehanna to the focused-scope seismic review category. PP&L elected to implement a focused-scope seismic margii) assessment (SMA),

following the Electric Power Research Institute (EPRI) methodology. The seismic IPEEE relied primarily on plant walkdowns that focused on evaluating expected functional performance, component anchorage capability, and the potential for adverse seismic-caused spatial interactions. All accessible components were at least " walked-by," whereas at least one component of a given type was subject to a detailed walk-down. Detailed walkdowns were performed for Unit-1 equipment; in general, walk-bys were performed for Unit-2 equipment. (The licensee expected Unit-2 equipment to be at least as seismically rugged as corresponding Unit-1 equipment.) The safe shutdown equipment list (SSEL) was not expanded to include components required for successful integrity of containment safeguards (coolers and sprays) to prevent early seismic-related containment failures. Rather, a brief qualitative discussion was provided in the seismic IPEEE submittal pertaining predominantly to direct effects on containment structural integrity.

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The submittal also mentioned that external isolation valves received an SMA walkdown, as part of the evaluation of SSEL components. The submittal also pro' ided a discussion concerning evaluation of relay chatter, and a broad discussion pertaining to analysis or potential soil failures.

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I Probabilistic risk assessment (PRA) methodology was employed for the assessment of fire vulnerabilities,  !

including multi-stage screening and extensive and detailed analysis of potential fire events. Several data bases were employed for screening and quantification. Use was made of the Individual Plant Examination '

(IPE) internal events plant logic model. Several walkdowns of the plant were conducted to support the fire analysis. Screening was based on deterministic defense-in-depth criteria. Per these criteria, if at least one pathway for core heat removal remained available, the relevant fire zone under consideration could be screened out. (Some aspects of this screening process appear to be optimistie;

• r example, some fire zones that c(mtain cables have been screened out based on a combustible loading criteria.)

For liFO initiators, the licensee adopted a general methodology which follows that presented in NUREG-1407, and which includes the following major steps:

• Review Susquehanna-specific hazard data and licensing basis Identify significant changes since issuance of the plant operating license (OL)

Verify that the design meets the 1975 Standard Review Plan (SRP) criteria Document study approach and findings Key IPEEE Findings From the seismic IPEEE, calculations of high confidence of low probability of failure (IICLPF) capacities were performed for unscreened SSEL components, producing ilCLPF values as low as 0.21g. (The IICLPF assessments were made with respect to the NUREG/CR-0098 median,55daraped spectral shape.)

lience, strictly speaking, the plant-level IICLPF capacity of Susquehanna does not exceed 0.21g.

Ilowever, the licensee has stated that the low-capacity items are either not strictly required for the SSEL, or that their failure may be rectified through manual operations. Thus, the licensee concludes that the success paths actually do meet the 0.3g review level earthquake (RLE). The licensee has thus not proposed any improvements to increase the plant liCLPF capacity. The submittal did identify a number ofissues which required implementation of resolution approaches. A few equipment modifications were proposed / implemented, and procedures to improve seismic housekeeping were being considered. These enhancements were addressed primarily with respect to a review of safe shutdown equipment; no modifications were found specifically/ exclusively as a result of containment perforipanee considerations.

From the tire IPEEE, the licensee concluded that there are no significant fire vulnerabilities at Susquehanna. The total fire core damage frequency (CDF), under full power conditions, was estimated to be 10* per cycle. (The term " cycle" was not defined, but it is inferred to represent a refueling cycle.)

This fire CDF result is several orders of magnitude smaller than values obtained from studies of other plants. Based on the review of the IPE submittal and the frequcncies of fire occurrence used in the IPEEE fire analysis, it can be inferred that the application of IPE internal events data for conditional core damage frequencies has led to such a small CDF. The significant fire zones were found to be the control room, the relay rooms, and the battery charger rooms. These rooms contain cables and cabinets from multiple trains of the emergency core cooling system (ECCS). Detailed analysis was conducted, and multiple fire scenarios were considered, for each significant tire zone.

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e v The Susquehanna liFO IPEEE submittal has reported that application of the NUREG-1407 progressive screening approach reveals that high winds and tornadoes, external flooding, and transportation and nearby facility accidents all screen-out due to conformance of the Susquehanna design with the NRC 1975 SRP.

In addition, the submittal reports that no other unique external hazards were found to exist for Susquehanna.

Generic Issues and Unresolved Safety issues The Susquehanna seismic IPEEE submittal mentions the following additional issues (which the licensee considers to be resolved):

Unresolved safety issue (USI) A-17. " Systems Interactions in Nuclear Power Plants"

• USl A-40, " Seismic Design Criteria"

. USI A-45, " Shutdown Decay lieat Removal Requirements" USI A-46, " Verification of Seismic Adequacy of Equipment in Operating Plants" Generic issue (GI)-131, " Potential Seismic Interaction involving the Movable In-Core Flux Mapping System Used in Westinghouse Plants"

• Eastern U.S. Seismicity Issue

• Seismic-Fire concerns in the licensee's treatment of USI A-45, the suppression pool cooling mode (SPCM) of residual heat removal (RiiR) was specifically considered to be available for decay heat removal (D11R) following a seismic margin earthquake (SME). For SPCM operation success, one RilR service water (RilRSW) pump. one RIIR pump, and the essential service water (ESW) system must function for RIIR cooling. All equipment required for operation of at least one division / train of SPCM were included in the SSEL. The submittal notes that, for the Unit-2 return valve (IIV-251F0248), for RilR to the suppression pool, there exists a potential for seismic interaction between the valve operator motor and a nearby hand rail. In addition, the Unit-2 Division-1 AC power (motor control center [MCC] 2B237) for RIIR and RIIRSW valves, used in suppression pool cooling, may interact with heating, ventilation and air conditioning (IIVAC) ducting in close proximity. The submittal discounted the possible effects of these interactions, noting that there would be time for manual valve operation. Thus, the submittal concluded that RIIR-SPCM equipment are expected to survive the SME, and provide adequate depth of DIIR capability.

In the tire IPEEE, the licensee addressed Sandia tire risk scoping study issues and USl A-45 issues, and dealt with all relevant major areas of potential concern. Plant walkdowns were corslueted to verify plant conditions with respect to these issues. Seismic-fire interaction was addressed by seismic evaluation of the potential fire sources and fire propagation computations of possible ignition scenarios. Except for a few cabinets, no areas in the plant were found where inadvertent actuation of suppression systems could lead to safety equipment damage. A few electrical cabinets were determined to require splash guards.

Also, special provisions were considered for draining water from a cable spreading room. Aside from these two items, for both sets of issues, the fire IPEEE did not reveal any additional outstanding problem areas. Specific inspection and testing procedures are in place to verify the integrity of penetration seals, fire barriers, and fire dampers. Regarding USI A-45, heat removal capabilities were addressed in detail via use of the IPE models for CDF evaluation. The presence of Thermo-Lag and other fire-retardant wrappings was mentioned in the IPEEE submittal; however, no credit was given to the protection provided by these materials.

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, v For liFO events, no USis/Gis or other non-IPEEE issues were addressed in the IPEEE submittal, including GI-103, " Probable Maximum Precipitation (PMP)."

Some'information is provided in the Susquehanna IPEEE submittal which pertains to generic safety issue (GSI) 147, GSI 148, ard GSI-172.

Vulnerabilities and Plant Improvements In the cover letter to the IPEEE submittal, the licensee draws the general conclusion that the plant is well designed and capable of withstanding severe external challenges, and that the physical condition and cleanliness of the plant were found to be good. The IPEEE submittal does not indicate that any plant-specific vulnerability (to potential severe accidents from external events) exists at Susquehanna Steam Electric Station, but it does note that opportunities for improvement have been identified, and that

, enrrective actions have been taken.

The seismic evaluation identified several open issues requiring resolution. Table 4.1 of this TER (derived from IPEEE Section 310.6) summarizes these issues and the approaches implemented for their resolution.

In some cases, llCLPF capacities have been determined which do not meet the RLE; but these items were ,

judged not to warrant a plant modification. Other issues wer;: resolved by judgments and/or analyses. )

In addition, some minor fixes and enhancements have been proposed for implementation (or have already been implemented) to resolve other issues. The submittal indicates that only one significant hardware change has actually been implemented (i.e., the removal of trolley cranes on top of electrical cabinets).

The submittal states that the risk significance of the other issue:, is low, but that these issues have been included in the SSES Discrepancy Management Program and will be resolved.

From the fire IPEEE, no fire vulnerabilities were identified. No improvements or commitments were deemed necessary to reduce the fire risk at Susquehanna. Ilowever, related to fire risk scoping study issues, the licensee has found a few electrical cabinets that require splash guaids, and has also concluded that special provisions should be considered for draining water from a cable spreading room.

With respect to HFO events, the licensee claims that no IIFO vulnerability exists at Susquehanna. Hence, no plant modifications were postulated or deemed necessary.

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in general, the Susquehanna seismic IPEEE has addressed most major elements specified in NUREG-1407 as recommended items to be considered for a focused-scope plant. The submittal itself gives a clear description of the seismic evaluation, although the format of documentation does not follow that recommended in NUREG-1407. The identification and implementation of safety enhancements has produced some meaningful insights in response to the objectives of GL d8-20, for a focused-scope plant.

However, this aspect of the seismic IPEEE is judged to be incomplete, since no actions have been proposed to increase plant ilCLPF capacity above the existing level of 0.21g. In addition, some elements of the seismic IPEEE are considered to fall short of providing a complete understanding of potential severe accident behavior.

The most significant weaknesses of the seismic IPEEE submittal are described as follows:

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The submittal's treatment of seismic containment performance was superficial, and did not involve walkdowns or evaluations of components needed for successful early accident mitigation following an SME, as recommended in NUREG-1407.

The licensee's assumption that valves, which may potentially be damaged through adverse seismic-spatial interactions, would be manually operated following the SME, so as to maintain success -

paths, appears tenuous.

The seismic IPEEE has identified deficiencies with respect to (a) adjacent panels and cabinets in close proximity that are not fastened together, (b) unanchored color video CRTs, and (c) liCLPF capacities of four components (two valves, an automatic transfer switch, and a motor control center). Ilowever, meaningful explanation as to why no resolutions have been proposed for these items has not been provided.

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• Anomalous conditions were encountered during the walkdown evaluation of seismic-fire l interactions; however, no meaningful explanation has been provided as to why no resolutions have been proposed for these conditions.

• The submittal did not apply screening criteria to non-seismic failures and human actions, as recommended in NUREG-1407 for an EPRI SMA.

The validity of the licensee's approach for evaluating 11CLPF capacities could not be verified.

The licensee's response to related RAls has suggested that the evaluation of liCLPF capacities for important components may not have been adequately performed, in accordance with established procedures.

This review recommends that the NRC pursue an audit of the licensee's llCLPF calculations for identified important components. j The licensee expended considerable effort in the preparation of the tire analysis portion of the IPEEE. This effort has provided an excellent opportunity for licensee engineers to improve their knowledge of the characteristics of the plant, and of how the plant may likely behave under tire-accident conditions. The l i

licensee used Level-1 PRA methodology to identify potential fire vulnerabilities. The IPE plant impact model was used to identify critical equipment and cables, and to conduct the CDF analysis.

4 The following are judged to be the strengths of the submittal:

The fire analysis portion of the IPEEE is well written. The overall presentation is clear and well organized. Tables and figures provide a considerable amount of supporting information.

• State-of-the-art data were used in the tire analysis, e

• The list of fire-related issues addressed by the licensee was extensive.

As listed below, the tire analysis has some weak points which, collectively, may lead to optimistic results, and raise some concerns over the validity of the final conclusions.

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The total fire CDF was found to be 10*/ cycle. This result is much smaller than values typically obtained for BWR plants, The reason for the small frequency is that the conditional core damage frequency adopted from the JPE model is very small, which in turn is dominated by human actions and recovery activities. Given the screening approach, and the method used for identifying vulnerabilities, several tire scenarios that could have otherwise contributed to risk have been excluded from the CDF analysis.

In the screening protocol, the licensee screened out fire zones based on combustible loading, assuming that cables do not contribute to this factor. For example, there are several cable spreading rooms in this plant which were screened out on the basis of their low combustible loading (which implies transient-fueled fires are not possible). For such areas where a large array of important cables are present, this approach can lead to optimistic conclusions, and hence, potentially mask an important vulnerability.

The bensee did not conduct a tire scenario-specific human error analysis. The human error rates from the IPE model were used, based on the premise that the conditions influencing the operator actions would not be much different from those under intarnal event scenarios. The human action and recovery analyses of the IPE cannot be used for fire induced core damage frequency evaluation without an adjustment for the influences of fire on the opera:ers (e.g., the effect on man-power, control room habitability, control panel alarms and instrumentation icMings, etc.). Ti.e core damage frequency, as reported in the IPEEE submittal, may be optimistic because of this omission.

The remote shutdown panel (RSP) was not modeled based on the premise that control mom evacuation would not be necessary under any fire conditions. This conclusion is contrary to industry practices and lack of analysis of the Remote Shutdown Panel, which is installed 'or mitigating a control room or cable spreading room tire, is a glaring omission in the tire analy4is.

The licensee has assumed that tires originating in a cabinet udll not affect cables and equipment outside the cabinet. This assumption is not valid in genetu, and especially when there exist openings on top of a cabinet.

The licensee has assumed that small motors (<50 hp) do not pose a tire threat or tire ignition source to other materials. Without performing a thorough analysis of every small motor, the licensee cannot categorically assume that the tire risk stemming from these mhtors is insignificant.

• *8 The sen.4itivity of the final results to specific assumptions was not presented in the submittal. The assumption regarding cable chases and valve galleries had a significant impact on the final results.

For liFO events, the licensee followed NUREG-1407 and GL 88-20 guidelines. The progressive screening approach was used. All events were screened based on conformance to the 1975 Standard Review Plan.

This review has concluded that:

The HFO analysis was clearly described. However, some external event studies were simply cited as sources of the evaluations. but were not incorporated into the IPEEE submittal report, thus Energy Research, Inc. xi ERl/NRC 95-512 I

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. u preventing a complete review. These studies, nuaetheless, have been clearly referenced in the submittal.

The screening basis employed (conformance to SRP criteria) was clearly identified and referenced explicitly back to the guidance in NUREG-1407. ,

I Insufficient infbrmation was provided in the liFO IPEEE submittal to verify the stated conclusion j that no site-unique external hazards exist at SSES and that it is appropriate that only the tive {

external hazards suggested in NUREG-1407 are studied in the IPEEE.

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u PREFACE The Energy Research, Inc., team members responsible for the present IPEEE review documented herein, include:

Seismic R. Sewell ike M. Kazarians

{11eh Winsis. Flannis and Other External Events J. Lambright Oversicht and Coordination M. Khatib-Rahbar, Principal Investigator and Report Review A. Kuritzky, IPEEE Review Coordination and Integration R. Sewell, Report Integration Dr. John Lambright of Lambright Technical Associates contributed to the preparation of Section 2.4 following the completion of the draft version of this TER.

This work was performed under the auspices of the United States Nuclear Regulatory Commission, Office of Nuclear Regulatory Research. The continued technical guidance and support of various NRC staffis acknowledged.

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. j AllBREVIATIONS ADS Automatie Depressurization System BWR Boiling Water Reactor CDF Core Damage Frequency CST Condensate Storage Tank .

DilR Decay lleat Removal ECCS Emergency Core Cooling System EOP Emergency Operating Procedure EPRI Electric Power Research Institute ERI Energy Research, Inc.

ESSW Essential Station Service Water ESW Essential Service Water FIVE Fire Induced Vulnerability Evaluation Method FSAR Final Safety Analysis Report GI Generic issue GL Generic Letter GSI Generic Safety issue llCLPF Ifigh Confidence of Low Probability of Failure (capacity)

IIFO liigh-Winds, Floods and Other External Events ifPCI liigh Pressure Coolant injection IIVAC lleating, Ventilation and Air Conditioning I&C Instrumentation and Control IPE Individual Plant Examination IPEEE Individual Plant Examination of External Events IRS In-Structure Response Spectrum -

LLNL Lawrence Livermore National Laboratory LOCA Loss of Coolant Accident LPCI Low Pressure Coolant Injection MCC Motor Control Center MSL Mean Sea Level NRC United States Nuclear Regulatory Commission OBE Operating Basis Earthquake OL Operating License P&lD Piping and Instrumentation Diagram PGA i Peak Ground Acceleration PMF Probable Maximum Flood PMP Probable Maximum Precipitation PP&L Pennsylvania Power & Light Company PRA Probabilistic Risk Assessment PWR Pressurized Water Reactor RAI Request for Additional Information RCIC Reactor Core Isolation Cooling RCS Reactor Coolant System RilR Residual lieat Removal RilRSW Residual lleat Removal Service Water Energy Research, Inc. xiv ERI/NRC 95-512

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RLE Review Level Earthquake RSP Remote Shutdown Panel SB0 Station Blackout SDCM Shutdown Cooling Mode

- SEIS Susquehanna Equipment Information System SEWS Seismic Evaluation Work Sheet SMA Seismic Margin Assessment SME Seismic Margin Earthquake SMM Seismic Margin Methodology SPCM Suppression Pool Cooling Mode SPLD Success Path Logic Diagram SQRT Seismic Qualification Report SQUG Seismic Qualification Utility Group SRP Standard Review Plan SRT Seismic Review Team SSE Safe Shutdown Earthquake SSEL Safe Shutdown Equipment List SSES Susquehanna Steam Electric Station SSI Soil-Structure Interaction TER Technical Evaluation Repon USI Unresolved Safety issue i

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u I INTRODUCTION This technical evaluation report (TER) documents the results of the " submittal-only" review of the individual plant examination of external events (IPEEE) for the Susquehanna Steam Electric Station (SSES), Units I and 2 [lj. This technical evaluation review, conducted by Energy Research, Inc. (ERI),

has considered various external initiators, including seismic events: fires; and high winds, floods, and other (IIFO) external events.

The U.S. Nuclear Regulatory Commission (NRC) objective for this review is to determine the extent to which the IPEEE process used by the licensee, Pennsylvania Power & Light Company (PP&L), meets the intent of Generic Letter (GL) 88-20, Supplement No. 4 [2]. Insights gained from the ERI review of the IPEEE submittal are intended to provide a reliable perspective that assists in making such a determination.

This review involves a qualitative evaluation of the licensee's IPEEE submittal, development of requests for additional information (RAls), evaluation of the licensee responses to these RAls, and finalization of the TER.

l The emphasis of this review is on describing the strengths and weaknesses of the IPEEE submittal, panicularly in reference to the guidelines established in NUREG-1407 [3]. Numerical results are verified for reasonableness, not for accuracy; however, when encountered, numerical inconsistencies are reported. .

This TER complies with the requirements of NRC's contractor task order for an IPEEE submittal-only review.

The remainder of this section of the TER describes the plant contiguration and presents an overview of l the licensee's IPEEE process and insights, as well as the review process employed for evaluation of the seismic, fire, and ilFO events sections of the Susquehanna IPEEE submittal. Sections 2.1 to 2.3 of this q report present ERl's findings related to the seismic, fire, and liFO reviews, respectively. Sections 3.1 {

to 3.3 summarize ERI's conclusions and recommendations from the seismic, tire, and liFO reviews, I respectively. Section 4 summarizes the IPEEE insights, improvements, and licensee commitments.

Section 5 includes completed IPEEE data summary and entry sheets. Finally, Section 6 provides a list of references.

1.1 Plant Charneterization Susquehanna Steam Electric Station consists of two General Electric BWR-4 reactor units, each having a Mark-ll containment. Unit I achieved commercial operation in June 1983, a41 Unit 2 commenced commercial operation in February 1985. The net electrical output for each unit is 1,050 MWe. Secondary containment (the reactor building) for each unit consists of a reinforced-concrete shear-wall structure, up to the elevation of the operating tioor, with a steel-framed superstructure and metal roof system above.

The plant is located in Salem Township, Luzerne County, Pennsylvania, approximately 50 miles northwest of Allentown, PA and 70 miles northeast ofliarrisburg, PA. The plant site consists of 1,075 acres located on a terrace above the flood plain of the Susquehanna River (which is about 4000 ft to the east of the plant).

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e l A portion of the plant electric power system is shared between the two reactor units. Each unit includes several cable spreading rooms. The plant has submitted Appendix-R related documentation and has completed the related plant modifications.

The safe shutdown earthquake (SSE) for Susquehanna Steam Electric Station is characterized by a 0.lg )

peak ground acceleration (PGA) for horizontal motion. For all structures except the diesel generator 'E' building, the design spectral shape is derived from a Newmark-type spectrum; for the diesel generator 'E' building, the design spectral shape is derived from a Regulatory Guide (R.G.) 1.60 spectrum. The operating basis earthquake (OBE) is equal to one-half of the SSE. Vertical design motion is equal to the j

horizontal design motion for the diesel generator 'E' building, and two-thirds of the horizontal design i motion for all other structures. All Seismic Category-I structures are founded on rock, with the exceptions of the essential station service water (ESSW) pumphouse and the spray pond.

, The Susquehanna IPEEE submittal does not specify a freeze date for modeling of plant configuration and operation. The study was started in December 1991, and documentation was completed in June 1994.

1.2 Owrview of the Licensee'sjPEEE Process and Imuortant Insirehts 1.2.1 Seismic The Susquehanna seismic IPEEE is a focused-scope evaluation that implements the Electric Power Research Institute (EPRI) seismic margins assessment (SMA) methodology. The review level earthquake (RLE) for Susquehanna is 0.3g PGA.

For the seismic walkdown,453 safe shutdown equipment list (SSEL) components were identified and addressed. Accessible components were either walked-down or walked-by, with application of seismic margin screening guidelines. Outliers were identified, and their associated high confidence of low probability of failure (ilCLPF) capacities were assessed.

The specific elements of the Susquehanna seismic IPEEE, as described in the submittal report, included:

• Seismic Margin Assessment of SSEL Components

• Soil Liquefaction Analysis

• Relay Chatter Evaluat' m (Bad Actors Only)

Consideration of Non-Seismic Failures and liuman Actions i Containment Performance Evaluation Evaluation of unresolved safety issue (USI) A-45

. Consideration of Seismically Induced Fires SMA Work (ilCLPF Calculations and Relay Chatter Evaluation)

• Resolution of Outliers

• Peer Review Documentation 1

The IPEEE submittal did not specifically identify any seismie vulnerabilities at Susquehanna Steam Electric Station. Several open issues were identified and resolved, or planned to be resolved, through the seismic l

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p-g v IPEEE. These resolutions included one plant hardware change, some minor (housekeeping related) corrective actions, and consideration of procedural implementations.

The submittal noted that Susquehanna was found to be well designed, capable of withstanding severe external challcages, and in good physical condition and state of cleanliness.

The study concluded that the plant success paths pass the 0.3g RLE screening basis. This conclusion was not actually supported by the IPEEE, however, since some outlier SSEL components were determined to have llCLPF capacities ranging from 0.21g to 0.26g.

1.2.2 Fire The licensee concluded that the overall fire core damage frequency (CDF) is 10* per cycle. The term

" cycle" was not defined in the submittal; however, it is inferred to represent a refueling cycle. This value for fire-induced CDF is significantly smaller than values obtained in studies of other plants. The licensee found the control room, relay rooms, and battery charging rooms to be the most significant contributors to the overall CDF from fire events.

The licensee further concluded that there are no significant tire vulnerabilities at Susquehanna.

Probabilistic risk assessment (PRA) methodology was used as the basis for identifying potential vulnerabilities. Multi-stage screening, and detailed analyses, were employed by the licensee for the assessment. Screening was based on deterministic defense-in-depth criteria. Per this criteria, it was considered that, if at least one pathway for core heat removal remained available, a given fire zone could be screened out. Some aspects of this screening process, as it has been applied by the licensee, may be optimistic. For example, some fire zones which contain cables have been screened out based on a combustible loading criteria.

The licensee also addressed Sandia tire risk scoping study issues and USI A-4:i issues. As a result of these considerations, the licensee fbund a few electrical cabinets that required splash guards, and concluded that special provisions should be considered for draining water from a cable spreading room. Aside from these two items, for both set ofissues, the fire IPEEE did not reveal any additional outstanding problem areas.

1.2.3 HFO Events it is unclear whether or not the HFO IPEEE analysis addressed a comprehensive lis) of potential external hazards, in its identification of areas where more detailed analyses were judged to be needed. Initiators considered in the analysis of HFO events included: high winds and tornadoes, external floods, and transportation and nearby facility accidents. These external events were all screened out based upon plant conformance with the 1975 Standard Review Plan (SRP) criteria. No bounding analyses or PRA analym were performed for HFO events. No relevant vulnerabilities were identified, and no plant improvements were made.

I.3 Overview of Review Process and Activities l

In its qualitative review of the Susquehanna IPEEE, ERI focused on the study's completeness in reference to NUREG-1407 guidance; its ability to achieve the intent and objectives of GL 88-20, Supplement No.

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4; its strengths and weaknesses with respect to the state-of-the-art; and the robustness of its condusions.

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This review did not emphasize confirmation of numerical accuracy of submittal results; however, any {

numerical errors that were obvious to the reviewers are noted in the review findings. The review process I included the following major activities:

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• Completely examine the IPEEE and related documents '

Develop a preliminary TER and RAls

• Examine responses to the RAls Finalize the TER and its findings Because these activities were performed in the context of a submittal-only review, ERI did not perform a site visit or an audit of either plant configuration or detailed supporting IPEEE analyses and data.

Consequently, it is important to note that the ERI review team did not verify whether or not the data presented in the IPEEE matches the actual conditions at the plant and whether or not the programs or procedures described by the licensee are indeed implemented at SSES.

1.3.1 Seismic In conducting the seismic review, ERI generally followed the emphasis and guidelines described in the report, Individual Plant Examination of External Events: Review Guidance l4), for review of a seismic margin assessment, and the guidance provided in the NRC report, IPEEE Step / Review Guidance Document l5). In addition, on the basis of the Susquehanna IPEEE submittal, ERI completed data entry tables developed in the Lawrence Livermore National Laboratory (LLNL) document entitled "lPEEE Database Data Entry Sheet Package" l6).

In its SSES IPEEE seismic review, ERI examined the following documents:

Sections 1, 2,3,4.8.2.l(2),6,7, and 8 of the licensee's !PEEE submittal ll)

The licensee's response l7] to the RAls generated as part of the initial submittal review The checklist of items identified in Reference [4] was generally consulted in conducting the seismic review.

Some of the primary considerations in the seismic review have included (among others) the following items:

Were appropriate walkdown procedures implemented, and was the walkdoyn effort sufficient to accomplish the objectives of the seismic IPEEE? j l

Was the development of success paths performed in a manner consistent with prescribed practices? {

Were random and human failures properly considered in such development?

Were component demands assessed in an appropriate manner, using valid seismic motion input and structural response modeling, as applicable? Was screening appropriately conducted?

l Were capacity calculations performed for a meaningful set of components, and are the capacity I results reasonable?

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, w Does the submittal's discussion of qualitative assessments (e.g., containment performance analysis, seismic-fire evaluation) refleet reasonable engineering judgment, and have all relevant concerns 1

been addressed?

lias the seismic IPEEE produced meaningful findings, has the licensee proposed valid plant improvements, and have all seismic risk outliers been addressed?

I 1.3.2 Fire '

During this technical evaluation, ERI reviewed the fire events portion of the IPEEE for completeness and l consistency with past experience. This review was based on consideration of Sections 1, 2, 4, 6, 7, 8, and 9 of Reference ll], and on the licensee responses to fire-related RAls l7]. The guidance provided in References [4] and [5] were used to formulate the review process and the organization of this document.

The data entry sheets used in Section 5 are taken from Reference [6].  !

The process implemented for ERI's review of the tire IPEEE included an exarr .ation of the licensee's j methodology, data, and results. ERI reviewed the methodology for consistency with currently accepted and state-of-the-art methods, paying special attention to the screening methodology and to the approach for estimating the frequency of occurrence of a tire scenario, to ensure that no tire scenarios were prematurely eliminated. The data element of a tire IPEEE includes, among others, such items as:

• Cable routing

• Fire zone / area partitioning

• Fire occurrence frequencies

• Event sequences

• Fire detection and suppression capabilities 1.3.3 IIFO Events The review process for ilFO events closely followed the guidance provided in the report entitled IPEEE Step 1 Rniew Guidance Document [5]. This process involved examinations of the methodology, the data used, and the results and conclusions derived in the submittal. The llFO IPEEE methodology was reviewed for consistency with currently accepted practices and NRC recommended procedures. Special attention was focused on evaluating the adequacy of data used to estimate the frequency of liFO events, and on confirming that any analysis of SRP conformance was appropriately execqted. In addition, the validity of the licensee's conclusions, in consideration of the results reported in the IPEEE submittal, was assessed. Also, in some instances, computations of frequencies of occurrence of hazards, fragility values, and failure probabilities were spot-checked. Review team experience was relied upon to assess the reasonableness of the licensee's evaluation.

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2 CONTRACTOR REVIEW FINDINGS 2.1 Seismic A summary of the licensee's seismic IPEEE process has been described in Section 1.2. Itere, the licensee's seismic evaluation is examined in detail, and discussion is provided regarding significant observations encountered in the present review.

2.1.1 Overview and Relevance of the Seismic IPEEE Process

a. Seismic Review Category and RLE Susquehanna is assigned to the focused-scope seismic review category in NUREG-1407. The RLE is described by the NUREG/CR-0098 [8] median spectral shape anchored to a peak ground acceleration (PGA) value of 0.3g.

b. Seismic IPEEE Process PP&L elected to implement a focused-scope evaluation, following the EPRI seismic margins assessment (SMA) methodology, for conducting the seismic IPEEE of Susquehanna Steam Electric Station.

c. Review Findings in general, the seismic IPEEE addresses, in more or less detail, all major elements of concern for a focused-scope analysis, as identified by NUREG-1407. The study has only partially addressed seismic containment performance issues. An adequate evaluation of relay chatter was included in the study. A comparatively broad treatment of soil failures was also provided in the submittal. IICLPF capacities were calculated or inferred (from design and qualification data) for a number of components. Overall, it can be stated that the Susquehanna seismic IPEEE is generally complete with respect to the breadth of guidelines recommended for a focused-scope SMA that is based on the EPRI methodology. The depth of some elements of the analysis, however, is in question, as will be discussed subsequently.

Susqueluuma Steam Electric Station is a dual-unit boiling water reactor (BWR). Safe shutdown paths need to be developed for both units, and walkdowns need to be performed for SSEL components at both units.

These conditions have been achieved in the seismic IPEEE. Ilowever, Unit I received the more thorough i

analysis, as the more detailed evaluations were generally performed in the component walkdowns of Unit-1 equipment: Unit-2 equipment more often received a less rigorous walk-by.

2.1.2 Success Paths and Component List For the seismic IPEEE of Susquehanna, PP&L developed a success path logic diagram (SPLD) that describes the plant functions needed to achieve and maintain a stable shutdown condition for at least 72 I hours. Reactivity control, reactor coolant system (RCS) pressure control, RCS inventory control, and decay heat removal were the essential " core protection safety functions" addressed in the development of I

success paths. The preferred success path relies on high-pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC), for pressure and inventory control, whereas the alternate success path relies Energy Research, Inc. 6 ERI/NRC 95-512 I

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on the automatic depressurization system ( ADS) and residual heat removal (RiiR) in the low-pressure coolant injection (LPCI) mode. Iloth success paths rely on reactor scram for reactivity control. For decay heat removal, the high-pressure (preterred) path makes use of the suppression pool cooling mode (SPCM) of RilR, whereas the low-pressure (alternate) path makes use of the alternate shutdown cooling mode of RilR. These success paths are relatively consistent with the example paths discussed in EPRI NP-6041 for a 13WR-4 plant. Ilowever, the SPLD includes possible use of isolation condenser / makeup for non-loss -

of coolant accident (LOCA) conditions. In addition, core spray is a suggested alternate to LPCI, and RiiR in shutdown cooling mode (SDCM) is a suggested alternate to RIIR-SPCM. The IPEEE submittal notes that core spray will likely be available following an seismie margin earthquake (SME), and hence, provides additional redundancy for success. liowever, the core spray system was not actually included in developing the SSEL; neither were RiiR-SDCM and isolation condenser / makeup.

The success paths chosen for safe shutdown of Susquehanna, following a seismic margin earthquake, assumed both loss of offsite power conditions and small-break LOCA conditions. The IPE event tree for small-break LOCA events was used in developing the success paths.

Once the success paths were determined, the frontline and support systems that comprise these paths were identified. The submittal presented a list of frontline systems and support that are needed to fully ensure integrity of the various elements of the success paths. System piping and instrumentation diagrams (P&lDs) and emergency operating procedures (EOPs) were consulted to identify specific components and instrumentation that comprised the safe shutdown equipment list. The " rule-of-the box" approach was used in delineating separate components. A list of 453 SSEL components was developed from the systems analysis.

Components pertaining to containment safeguard systems were not included in the SSEL.

Overall, the approach described in the Susquehanna seismic IPEEE for defining success paths and developing the SSEL is consistent with the guidelines presented in NUREG-1407. Ilowever, the study has not developed a containment safeguard equipment list for seismic containment performance assessment.

2.1.3 Non-Seismic Failures and lluman Actions The IPEEE submittal provides a qualitative discussion of the consideration of non-seismic failures and human actions, as they affect the chosen success paths. No formal screening approach was implemented to evaluate non-seismic failures and human errors. j The seismic IPEEE submittal stated that failure probability values are consistent with the screening values used in the Maine Yankee seismic margin review. The submittal noted that non-seismic failures are countered by incorporating equipment with high reliability into each path. IIPCI and RCIC, however, are each single train systems, and (in actuality) have a comparatively low combined reliability (0.0024 failure rate per demand). In addition, as far as operator actions are concerned, the submittal indicated that manual starting of the residual heat removal service water (RIIRSW) pumps is a key action, and that this action is part of the design basis. Minimal remote human actions were expected by the licensee to be required in coping with an SME. liowever, the submittal did not discuss the likely impacts of the earthquake itself on human actions and error rates.

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Consequently, the licensee's treatment of non-seismic failures and human actions falls short of the requested guidelines of NUREG-1407 pertaining to the application of relevant screening criteria.

2.1.4 Seismic Input The seismic IPEEE for Susquehanna used the NUREG/CR-0098 median response spectrum, anchored to a peak ground acceleration (PGA) value of 0.3g, to define the ground-motion input for seismic margin evaluation. All seismic Category-l structures, except the ESSW pumphouse and the spray pond, are founded on rock. Consequently, the seismic input for the ESSW pumphouse and spray pond was defined by a NUREGICR-0098 median soil spectrum; the RLE for all other structures was defined by a NUREG/CR-0098 median rock spectrum.

The seismic input selected for use in the Susquehanna IPEEE is consistent with the relevant guidelines presented in NUREG-1407.

2.1.5 Structural Responses and Component Demands In the seismic IPEEE, structural responses and component demands associated with the seismic margin earthquake were obtained simply, by linearly scaling existing design-basis responses and demands. No new structural models were developed, and no new in-structure response spectra were generated. For example, design-basis in-structure response spectra (IRS) were scaled by the ratio of the RLE spectral acceleration to the SSE at a building's dominant natural frequency, in order to obtain the IRS for seismic margin evaluation (see EPRI NP-6041-SL [9]). The submittal provided a rather detailed description of the structural models and parameters, and of the motion time-histories, used to generate the original (design-basis) structural responses and component demands. A soil-structure-interaction (SSI) model was used for the ESSW pumphouse, and a flexible-base model was used for the reactor building; a fixed-base model was employed for the other structures.

Overall, the development of structural responses and component demands used in the Susquehanna IPEEE is consistent with the relevant guidelines presented in NUREG-1407, 2.1.6 Screening Criteria The Susquehanna seismic IPEEE implemented the EPRI seismic margins methodology, and hence, made use of the appropriate screening tables, criteria and procedures discussed in EPRI p'P-6041, Rev.1 [9].

The SMA screening column applicable to spectral accelerations between the limits of 0.8g and 1.2g was apparently used (Tables 2-3 and 2-4 of EPRI NP-6041, Rev.1). Caveats of the seismic margin screening tables were used in the screening process for stmetures, as well as for electrical and mechanical equipment.

In addition to consideration of caveats related to functional capability, anchorage and spatial interaction concerns were addressed in the walkdown(s),

i The screening criteria were apparently applied in a conservative manner, as several components were not screened out in the walkdown evaluations, but were evaluated subsequent to the walkdown. Nonetheless, <

there is no basis to conclude that the screening has not been performed in a careful and appropriate

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Overall, the screening criteria and procedures used in the Susquehanna seismic IPEEE are consistent wi.h NUREG-1407 focused-scope guidelines and appear to be appropriate for screening SSEL components.

2.1.7 Plant Walkdown Process

a. Preparatory Work Prior to the walkdown itself, plant information was collected and reviewed (including seismic design and qualification data), components in the SSEL were located, seismic evaluation work sheets (SEWSs) and walkdown checklists were obtained, and arrangements were made with plant personnel to access the internals of electrical equipment.

b. Systems and Element Selection Walkdoun A separate systems and element selection walkdown was not performed. The licensee has stated that this aspect of the walkdown process was essentially subsumed into the seismic capability walkdowns.

c. Seismic Capability Walldoun During the seismic walkdowns, the seismic review team (SRT) physically reviewed important attributes of the SSEL componerits, according to procedures described in EPRI NP-6041-SL. The three major aspects of the walkdown included assessments of functional capacity, anchorage adequacy, and potential seismic interactions. General seismic housekeeping issues were also addressed in detail.

A detailed review of one item for each equipment type was performed during the walkdowns. The detailed reviews were generally perfbrmed for Unit-1 equipment; Unit-2 equipment were expected to have seismic capabilities at least equal to corresponding Unit-1 equipment. All of the remaining accessible equipment received a walk-by. The purposes of the walk-by were to confirm that the construction pattern was typical, and to identify any potential interaction concerns that may be unique for each equipment item. A limited sampling of distribution systems (e.g., piping, cable trays, conduit, and heating, ventilation and air conditioning [HVAC] ducting) was included in the walk-bys.

Walkdown findings were documented on walkdown checklists and SEWSs; in addition, notes of the walkdowns were written up, and photographs were taken, as needed, to support the notes.

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d. Subsequent Walkdowns A specific discussion of subsequent walkdowns (if any) that were needed was not provided in the submittal.

e. Treatment ofinaccessible Components The submittal noted specific areas and/or components that were inaccessible due to high radiation / contamination. Documentation reviews were apparently conducted in place of physical walkdowns, for such items.

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f. Walkdown Duration and Training ofSeismic Rniew Team The seismic IPEEE walkdown of Susquehanna was conducted by trained licensee personnel and consultants, and apparently took place over a meaningful duration.

g. Rninv Findings The walkdown process conducted for the seismic IPEEE of Susquehanna appears reasonable and appropriate for a focused-scope evaluation, and is capable of identifying outliers with respect to safe shutdown following a seismic margin earthquake. The submittal's summary presentation of walkdown findings and outlier resolutions, by equipment category, is considered to be a significant strength of the IPEEE report.

2.1.8 Evaluation of Outliers

a. Overall Approach Screening of components was apparently conducted in a reasonably conservative manner, as several components were not screened out. Components that could not be screened were not immediately designated as outliers, although NUREG-1407 generally reserves the term " outlier" for such items. (It appears that the licensee uses the term outlier to denote those components having a computed ilCLPF capacity less than the RLE.) Even so, all non-screened components were assessed in the seismic IPEEE to determine whether or not further consideration was warranted.

In some instances, an estimate of the liCLPF capacity was obtained for non-screened components; in other instances, a review of design and/or qualification packages was used to establish equipment adequacy, without the explicit determination of a precise liCLPF capacity. For the seismic housekeeping outliers, steps to implement corrective actions were taken, without further analysis. One outlier was resolved through a hardware change; two other outliers were identified as deficiencies and included in the SSES Discrepancy Management Program.

h. HCLPF Calculations The IPEEE apparently involved HCLPF calculations for several components, as the submittal notes that many unscreened components were assessed as having IICLPF capacities in excess o the 0.3g RLE. The submittal states that such assessments were based on the original seismic design o however, the validity of such assessments could not be verified in this review. Four components were found to have HCLPF capacities less that the RLE: a 11PCI discharge valve (0.21g liCLPF), a RHR-SPCM suppression pool inlet valve (0.21g HCLPF), an automatic transfer switch (0.25g HCLPF), and a 480 V motor control center (MCC) (0.26g liCLPF). The IICLPF calculations for these components were not provided by the licensee for review.

c. Other Calculations No HCLPF calculations were performed for masonry walls. Rather, for masonry walls, the submittal states that a conservative design approach to wall evaluation was undertaken which indicates that the walls Energy Research, Inc. 10 ERl/NRC 95-512 l

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, m have HCLPF capacities in excess of 0.3g. However, the calculational approach implemented to justify this conclusion was not clearly described.

The submittal notes that there are no above-ground flat-bottomed tanks at the plant. All SSE1 tanks were I either screened in the SMA evaluation, or found to be adequate based on seismie qualificaiica information.

The condensate storage tank (CST) was assumed to fail in an SME, and hence, was not included in the ,

SSEL.

d. Review Findings The seismic IPEEE submittal for Susquehanna Steam Electric Station has performed capacity calculations for unsereened components, and has implemented corrective actions tbr a partial set of the identified outliers. The validity of the licensee's HCLPF assessments and of its treatment of masonry walls could not be verified by this review, 2.1.9 Relay Chatter Evaluation Susquehanna Steam Electric Station is not a USI A-46 plant, and hence, NUREG-1407 recommends only the location and evaluation of low-ruggedness (bad-actor) relays.

The licensee *s seismic IPEEE submittal provided a table which lists locations of bad actor relays in SSEL electrical components. The submittal states that a significant part of the systems analysis work for the SMA included the location and evaluation of the impact of low-ruggedness relays. Reference [7] provided the following additional information pertaining to the process for relay evaluation:

The Susquehanna Equipment Information System (SEIS), an electronic database, was initially used to locate low-ruggedness relays.

• Electrical drawings were also reviewed.

Device listing contained in seismic qualitication report (SQRT) binders were also reviewed.

The seismic walkdown was performed to contirm the existence of low-ruggedness relays and I additional non-bad-actor relays associated with SSEL equipment.

A brief discussion is provided in the submittal as to the potential effects of chatter of the identitled relays; however, it is not clear that this discussion provides a comprehensive treatment. In general, the evaluation simply concludes that relay chatter is acceptable, and does not represent a vulnerability with respect to the ability to safely shut down the plant. The submittal indicates that there are no vulnerable relays that may lead to spurious behavior of the fire suppression system, containment isolation system, etc.

l No actions were proposed by the licensee to replace the identified low-ruggedness relays.

Overall, the licensee's evaluation of low-ruggedness relays appears consistent with NUREG-1407 guidelines. Four locations of low-ruggedness relays were identified. However, the licensee's IPEEE Energy Research, Inc. 1I ERl/NRC 95-512 l

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1 documentation was insuf6cient to verify the licensee's conclusions that spurious behavior of these relays would not pose a threat to safe shutdown.

2.1.10 Soil Failure Analysis The seismic IPEEE included a broad evaluation of soil-related failures. The following issues were f addressed:

Liquefaction potential for the terrace deposits below the spray pond Post-earthquake settlement of the terrace deposits

• Stability of the spray pond slopes Lateral sliding of ESSW pumphouse

• Seismic distortions in buried pipelines The soil-failures evaluation appears to be adequate, and no problem areas were identined.

l Thus, the treatment of soil failures in the Susquehanna seismic IPEEE appears to satisfy the guidelines described in NUREG-1407 for a focused-scope plant.

1 2.1.I1 Containment Performance Analysis )

Only a brief qualitative discussion of seismic impacts on containment performance was provided in the seismic IPEEE submittal. A list of components necessary to ensure successful early containment performance, following an SME, was not developed. No comprehensive walkdowns related to components required specifically for mitigation of early containment failure were apparently performed, although the submittal did mention that external isolation valves received an SMA walkdown as part of the evaluation of SSEL components. The submittal also noted that the containment structure and piping / valves are j expected to survive the SME. l The submittal did not provide any discussion of criteria for developing a containment systems equipment list (including an explicitly stated definition of successful containment perforn. nce). The licensee has claimed that, because core damage will not occur as a result of the SME, no such analyses are needed. ,

This perspective is not consistent with the guidelines of NUREG-1407, and thus, the containment performance assessment in the Susquehanna seismic IPEEE is considered to be deficient.

I 2.1.12 Seismic-Fire Interaction and Seismically Induced Flood Evaluations The IPEEE submittal included consideration of seismic-fire interactions as part of the Sandia fire risk scoping study issues. The submittal primarily discussed assessment of the potential for seismically induced fires. Seismic actuation of fire suppression systems and seismic degradation of fire suppression systems were not specifically addressed in significant detail in the submittal. The evaluation of seismically induced fire potential included a seismic-fire interactions walkdown. Items considered in this evaluation include the potential for: toppling of electrical equipment; failure of diesel fuel and lubricating oil tanks; spillage of lube oil; failure of hydrogen gas bottles; fire ignition from non-safety, non-seismic equipment; and seismic failure of fire water, CO2 and lialon systems.

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and provide adequate depth of DliR capability. Ilence, no corrective actions have been implemented or proposed with respect to improving seismic capability of DHR systems.

Overall, the Susquehanna seismic IPEEE includes a meaningful evaluation of potential vulnerabilities in decay heat removal systems, which addresses concerns relevant to USI A-45. Identined deficiencies

- potentially affecting DHR seismic capability need to be further investigated as candidates for improvement.

For instance, the licensee's claim that the suppression pool inlet valve could be operated manually is tenuous, since seismically induced damage could prevent manual valve operation (or access to the valve),

or operator failure to instigate this manual action could occur.

2.1.14 Other Safety issues The submittal also provided discussions addressing USI A-17, " System Interactions in Nuclear Power Plants"; USI A40, " Seismic Design Criteria"; USI A46, " Verification of Seismic Adequacy of Equipment in Operating Plants"; and the Eastern U.S. Seismicity issue.

a. USI A-17 and USI A-40 Resolution With respect to USI A-17, the submittal simply noted that this issue has been addressed within the IPEEE program, through screening and SMA walkdowns of SSEL items. With respect to USI A-40, the submittal stated that the SSEL for Susquehanna did not include any above-ground tanks, and hence, this issue is resolved by the IPEEE. This TER has not included an evaluation of the licensee's treatment of these issues.

b. USl A-46 Resolution The submittal noted that USI A46 is not applicable to Susquchanna. (USI A-46 relates to plants for which construction permits were docketed prior to 1972, whereas construction permits for SSES were issued in late 1973.)

c. Eastern U.S. Seismicity issue With respect to the Eastern U.S. Seismicity Issue, the submittal stated that the IPEEE provides a resolution to this issue, without any additional analyses or documentation.

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d. Generic Safety issues Some seismic-related information having relevance to Generic Safety Issue (GSI)-172 is provided in the submittal, as discussed in Section 2.4.3 of this TER.

e. Resiew Findings  ;

wThe seismic IPEEE includes discussions concerning USI A-17, USI A-40, and the Eastern U.S. Seismicity  !

Issue, and concludes that these issues are considered to be resolved. The submittal notes that USI A-46 is not applicable to Susquehanna.

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4 2.1.15 Peer Review Process PP&L personnel had a meaningful (although not dominant) role in conducting the seismic IPEEE. The PP&L engineers and consultants who participated in the seismic evaluation received Seismic Qualitication Utility Group (SQUG) training and EPRI seismic Individual Plant Examination (IPE) follow-on training.

The submittal identities all IPEEE participants, their roles in conducting / reviewing the IPEEE, and their relevant qualitications.

The Susquehanna seismic IPEEE received independent internal peer reviews from PP&L personnel in various functional organizations / departments. The submittal does not indicate that outside consultants participated in the review. The following aspects were involved in the seismic peer review:

An independe.~.t equipment walkdown on a representative sample of SSEL equipment

• An independent review of a representative sample of SRT calculations An independent review of the seismic portion of the IPEEE report All seismic peer review comments were discussed with, and resolved, by the SRT members. Tl'e submittal did not discuss the training, experience, or qualitications of the seismic peer reviewers.

It thus appears that a meaningful peer review was conducted of the Susquehanna seismic IPEEE, and that the major comments / concerns of the internal peer review team were addressed.

2.1.16 Summary Evaluation of Key insights The seismic IPEEE did not specifically identify any seismic vulnerabilities at Susquehanna Steam Electric Station. Several open issues were identified and resolved, or planned to be resolved, through the seismic IPEEE. These resolutions included one plant hardware change, some minor (housekeeping related) corrective actions, and consideration of procedural implementations.

The submittal noted that Susquehanna was found to be well designed, capable of withstanding severe external challenges, and in good physical condition and state of cleanliness. 1 The lowest liCLPF capacity of an SSEL component was evaluated at 0.21g (NUREG/CR-0098 median, 5%-damped, soil spectrum anchored to a PGA value of 0.3g), which implies that the plant-level liCLPF j 4

capacity is no greater than 0.21g. (The next highest liCLPF capacity computed foi an SSEL outlier was 0.25g.) Nonetheless, the licensee reported that both success paths meet the 0.3g screening level (RLE).

The submittal did isot provide sufficient justification to support this major finding.

Although opponunities for a number of safety enhancements were evident from the submittal, the licensee has only implemented one significant enhancement. No specific actions have been undertaken for the following concerns: (a) adjacent panels and cabinets in close proximity that are not fastened together; (b) color video CRTs that are not attached to their supports; (c) existence of low-ruggedness relays that affect success paths; (d) low-capacity 11PCI pump discharge valve, due to an interaction problem; (e) low-capacity R11R-SPCM suppression pool inlet valve, due to an interaction problem; (t) low-capacity automatic transfer switch, due to an interaction problem; and (g) low-capacity 480 V MCC, due to an interaction problem.

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3 2.2 Bre A summary of the licensee's fire IPEEE process has been described in Section 1.2. liere, the licensee's fire evaluation is described in detail, and discussion is provided regarding significant observations encountered in the present review.

2.2.1 Overview and Relevance of the Fire IPEEE Process

a. Methodology Selectedfor the Fire IPEEE The fire evaluation was conducted using Level-1 tire PRA methodology. Although core damage frequency (CDF) was the principal basis for many of the decisions made in the screening phases, the screening criteria were essentially deterministic in nature. One screening criterion was based on the availability of a core heat removal pathway, and another criterion was based on the combustible loading of a fire zone.

The licensee claims that the " purpose . . . is not exclusively to calculate a core damage number. The core damage frequency calculation is almost a consequence of the study and not its most valuable result."

Unsereened areas and fire zones were subjected to fire propagation analysis using the COMPBRN lile j computer code. If the results of fire propagation analysis showed that critical damage could not c :ur, the area / fire zone was not considered a vulnerability.' Multi-compartmental tire propagatir i and consequential damage were analyzed in the IPEEE. Extensive plant walkdowns were conducted for different steps of the analyses. The frequencies for fire initiation were obtained from plant-specific and general industry tire experience, using simple statistical and uncertainty analysis methods.

The IPEEE submittal included extensive discussions regarding the methods and data used in conducting the fire analysis. The analysis was performed for Unit 1. Systems, components, and areas common to both units were identified in the course of the analysis, and were incorporated into the results.

b. Key Assumptions Used in Performing the Fire IPEEE Following is a list of key assumptions that were either mentioned by the licensee in the IPEEE submittal or have been inferred by the reviewers:

1. Reactor sub-criticality was assumed to be successful in all cases.

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2. Although the licensee stated that fire-induced LOCA is not possible, hot shorts and spurious valve operation were discussed in the IPEEE submittal.

3. Fire barriers / boundaries were taken to be good as rated.

4. Cables have been IEEE 383 qualified, and therefore, will not initiate a fire.

5. ' A fire within an electrical cabinet will not propagate outside the cabinet.

6. Cable failure will occur at 700* F.

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7. Small motors (< 50 hp) were not considered as potential tire sources.

8. No credit was given to fire detection and suppression capabilities.

9. No credit was given to fire-retardant wrappings or Thermo-Lag.

c. Status ofAppendix R Modifcations Appendix-R modifications have been completed at the plan . The IPEEE submittal used the cable routing identified as part of the Appendix-R effort for executing the fire-zone screening analysis,

d. New or Existing PRA The fire IPEEE was based on a new fire PRA study. The core damage frequency calculations were based on the IPE plant resp (mse model.

2.2.2 Review of Plant Information and Walkdown

a. Walkdown Team Composition The two engineers who prepared the fire IPEEE have conducted six plant walkdowns to verify or investigate plant configuration. No samples of walkdown notes were presented in the submittal.

! In Section 6 of Reference [1], the licensee's involvement in the preparation of the IPEEE was discussed.

i PP&L personnel have been involved in conducting the walkdowns, systems analysis, cable routing, and COMPBRN com[oter runs. A consultant with fire and risk analysis expertise was also involved in the IPEEE preparation.

b. Signifcant Walkdown Findings The scope of the walkdowns included investigation of fixed and transient ignition sources, assessment of l the potential for multi-companment fires, delineation of cable and equipment locations, and evaluation of plant features relevant to Sandia fire risk scoping study issues. Reference [1] did not indicate that the walkdown team has discovered any new fire vulnerability as a resul~ of the plant visits. However, two types of plant modifications were considered based on the walkdown findings: (1) installation of splash guards on top of certain electrical cabinets; and (2) installation of a remotely operated valve on the floor drain of one of the cable spreading rooms.

c. Signifcant Plact Features The following plant features are relevant to the fire IPEEE:

1. The plant consists of two BWR-4 reactors with Mark-Il containments. I

2. Unit I started commercial operation in June 1983, and Unit 2 started operation in February 1985.

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3. A portion of the electric power system is shared between the two units. This fact has been inferred from the IPEEE discussions pertaining to different tire zones.

4. Each plant unit includes several cable spreading rooms.

2.2.3 Fire-Induced Initiating Events

a. Were initiating Ewnts Other than Reactor Trip Considerec!?

The components md initiators modeled in the IPE were used in the fire IPEEE. Although the licensee makes it clear that LOCA events were not considered as pan of the fire analysis, the possibility of spurious ADS opening from a control room fire has been considered. The submittal provided no separate discussion or listing of the initiating events included in the fire-impact model. Ilowever. from the discussions provided for individual tire scenarios, it can be inferred that the licensee has given proper consideration to the possibility of occurrence of various initiating events, including the potential for spurious ADS activation (i.e., a LOCA). Although the list ofiPE components was not expanded to include cables and components that would cause an initiator, from the discussions provided in References [1] and

[7], it can be inferred that the possibility of occurrence of various initiators has been considered properly.

b. Were the Initiating Events Analyzed Properly?

The submittal did not address initiating events as a separate topic. liowever, from the discussions provided in References [1] and [7J, it can be inferred that the possibility of various initiators has been properly considered. It was mentioned in the submittal that occurrence of a fire-induced LOCA was considered to be impossible. Also, one cf the screening criteria was based on reactor trip, as well as other transients. If, for a fire zone, there were na cables that could cause a reactor trip or a plant transient, then the fire zone was screened out. As mentioned above, the possibility of spurious actuation of ADS was considered as part of control room fire analysis.

2.2.4 Screening of Fire Zones

a. Was a Proper Screening Methodology Employect?

Screening was performed in several stages. In the first stage, screening consisted of three steps. In the first step, the failure of all equipment and cables in each fire zone was postul4ed. If there existed sufficient means for core cooling, the fire zone was screened out. In the second step, if a plant transient could not occur for a given fire zone, it was screened out. This method, as it was described in the submittal, can be interpreted as optimistic. For example, there may be areas v anin the plant where a large number of stand-by systems are co-located. A fire in such areas may be a significant contributor to the combined unavailability of those systems. Furthermore, an operator may initiate a reactor trip under such a fire condition.

l The third screening step was based on combustible loading. It was assumed that areas having cables only l would not be susceptible to fires. This assumption ignores the potential for transient ignition sources and combustibles. The approach represents an optimistic basis for screening. It fails to recognize areas with 1

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a high concentration of cables that are critical to safe plant shutdown. This item is particularly important in regard to cable shafts and cable tunnels.

In addition to the initial screening stage, the licensee has screened fire areas and combinations of fire areas (as part of multi-companment fire analysis) as pan of detailed fire modeling. The licensee did not provide a complete list of fire areas ad zones, nor of the equipment / system trains that could potentially be affected from fires in the various companments. Therefore, the review of the screening results has been based solely on the metl odology described in the submittal, as well as on some parts of the final results. The building level screening given in Table 4-15 of the submittal is considered reasonable and within the range of results expected for such BWRs. This conclusion, however, is based on the assumption that there is adequate separation among the service water pump units and their associated cables.

b. Have the Cable Spreading Room and the Control Room Been Screened Out?

The control room has been considered in the fire analysis, and a detailed review of the cabinets and of potential fire impacts has also been included. There are several cable spreading rooms in the plant. These rooms have been screened out at different stages of the analysis. These rooms were screened out based on either combustible loading on on a COMPBRN analysis of fire propagation.

c. Were There Any Fire Zones / Areas that Have Been improperly Screened Out?

This issue, from a methodological standpoint, has been discussed above. Some fire zones have been screened out based on combustible loading level. Such an approach is not proper because it does not take into account the systems serviced by the affected cables. Also, some fire zones have been screened out if a plant transient could not occur. As it is discussed in Section 2.2.4a, this method, as it was described in the submittal, can be interpreted as optimistic.

The submittal did not include a list of fire zones and associated equipment and system trains. Therefore, an independent verification of the screening steps could not be made.

2.2.5 Fire Hazard Analysis The fire initiation data provided in Reference [10] and plant-specific Gre occurrence experience have been used for establishing the fire frequency for each fire zone. Weighting factors have been applied to apportion the overall Ore frequency to a specific Dre zone, based on the contents pf the tire zone. The methodology described in the submittal, for collecting the necessary data and apportioning the frequencies, is deemed to be reasonable. The final results were presented for the majority of the fire zones, and these results are within the range of frequencies expected for such fire zones.

Plant-specific conditions have been used to alter the industry data provided in Reference [10]. Plant-specific fire occurrence data has been used in establishing the frequency of fires in the control room.

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i 2.2.6 Fire Growth and Propagation Fire growth and propagation analysis was conducted using COMPBRN life computerized fire modeling and FIVE worksheets. For those fire zones that did not screen out, ignition sources and targets (i.e., safe j shutdown cables and equipment) were identitled and a geometric model of these items was created. More than 10 fire zones have been analyzed for fire growth and propagation, and in all cases, it is concluded . ,

that, other than the ignition source, additional damage may not take place. For cabinet fires, it is assumed that fires originating in a cabinet will not affect cables and equipment outside the cabinet. Although the submittal mentions electrical cabinet tire tests conducted by Sandia National Laboratories, this assumption is not supported by those tests and it is specifically not valid for cabinets with openings on top.

a. Treatment of Cross-71)ne Fire Spread and Associated Major Assumptions 1

1 The possibility of multiple fire-zone equipment failures has been considered in the IPEEE. Cross-zone propagation was assumed to occur if the fire zone, where the fire originates, contains sufficient j combustibles to fuel a large fire. Also, all 3-hour rated fire barriers were assumed to remain intact. A i reasonable analysis has been conducted to address this issue. The results are within the range expected for such BWR plants. Active fire barrir l e been included in the fire analysis. The doors in areas 1 where safe shutdown cables and equipm. located are normally closed. Deterministic analysis has been conducted to determine whether or - gases may propagate to adjacent compartments if a fire

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damper fails to close properly.

b. Assumptions Associated with Detection ard Suppression On page 4-3, in Section 4.0.2.3 of the IPEEE submittal, it is stated that: "The analysis takes no credit for automatic detection or suppression, nor is any credit given for 'Thermo-Lag' or any other type of fire protective wrapping." However, in Section 4.5 of the submittal, the licensee provided a summary of fire detection and suppression features of the plant and of the reliability of the systems.

c. Treatment of Suppression-Induced Damage to Equipment, if Available '

There was no discussion of suppression-induced damage in any of the phases of the analysis. However, this issue was discussed as part of the licensee's response to the Sandia fire risk scoping study concerns.

From this effort, the licensee has considered installing splash guards on some electrical cabinets and providing special remote actuation capability for draining water from the cable spreadipg rooms. The issue l of fire suppression system impacts on safe shutdown equipment has been addressed as part of the Appendix-R submittal, and fire fighting procedures have been modified to minimize the pos3ibility of such an event. In the fire PRA analysis, total component failure was assumed to occur upon fire impact.

d. Computer Code Used, if Applicable COMPBRN IIIe was used extensively to analyze the possibility of damage to a critical set of components.

Based on the results of these analyses, several fire scenarios have been screened out as insignificant risk contributors. To run COMPBRN, numerous assumptions must be made in regards to location and characteristics of the pilot fire. The licensee has described these assumptions in Section 4.3.2 of the submittal report. From a review of a sample of COMPBRN input files presented in Reference [7], it can I

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1 be concluded that reasonable data and equipment and barrier configuration (i.e., walls, ceiling, ignition {

sources, combustible materials and targets) have been used. Ilowever, in at least two of the four sample l COMPBRN input files, only one combustible source was presented with no combustible or safety related f

targets. The purpose of such program executions is not clear and was not explained by the licensee.

Furthermore, since the submittal indicates that fire detection and suppression were not modeled, the merits of fire propagation modeling are not clear.

2.2.7 Evaluation of Component Fragilities and Failure Modes

a. Depnition of Fire-Induced Failures Fire-induced failures have been properly considered. Reference ll] mentioned spurious actuation of valves and other equipment, but did not provide sufficient detail for this review to verify whether or not proper methodology had been employed. The possibility of flow diversion from a fire affecting cables has been included in the analysis. Failure was deemed to occur from heat impact. Other phenomena from a fire event were deemed to have long-term effects, and would not alter the short-term outcome of the fire event.

b. Method Used to Determine Component Capacities Instrumentation and control (l&C) cabinets were assumed to fail at 325" F. The cable damage threshold criterion was taken to be 700* F. This damage criterion is somewhat optimistic, given that Reference [l1]

suggests 662* F as the damage threshold for IEEE 383 cables. liowever, the licensee has claimed that it has performed some sensitivity analyses where lower failure temperatures have been used. The overall impact of the temperature threshold on the final IPEEE results was expected to be minimal.

c. Generic Fragilities A discussion was provided in the fire IPEEE regarding equipment fragilities. Cable and I&C cabinet fragilities were expressed in terms of a threshold temperature (see above). The possibility of failure from l other fire effects was discounted as being insignificant for the short-term conditions of a tire evern Overall, the discussions provided in Section 4.4 of the submittal are considered to be reasonable and  ;

representative of the current knowledge base with respect to fire effects on components.

d. Plant-Specifc Fragilities i No plant-specific failure fragilities, other than those mentioned above, were discussxt in the submittal.

e. Technique Used to Treat Operator Recovery Actions Recovery actions and human error rates of the IPE have been used. The effect of fire, as a performance shaping factor, has not been considered. The possibility of using the remote shutdown panel (RSP) was not modeled at all. It was argued that there will never be the necessity to evacuate the control room.

From such statements it is inferred that the control room fire analysis is deficient and recovery analysis may have used optimistic human error rates. Ifowever, it is not clear whether the deficiency in control room analysis couid lead to an optimistic or pessimistic result.

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2.2.8 Fire Detection and Suppression Fire detection and suppression was not included in any part of fire IPEEE. Ilowever, a discussion was provided regarding the features and reliability of these systems.

2.2.9 Analysis of Plant Systems and Sequences

a. Key Assumptions including Success Criteria and Associated Bases The success criteria were directly taken from the IPE analysis, and these criteria have not been modified for the fire analysis,

b. Event Trees (Functional or Systemic)

Event trees have not been presented in the fire IPEEE submittal. Therefore, whether they are systemic event trees or functional event trees could not be verified.

c. Dependency Matrix, ifit is Diferentfrom that for Seismic Events No dependency matrix has been provided in the submittal.

d. Plan!-Unique System Dependencies The submittal did not describe any unique system dependencies. However, it is inferred that Units 1 and 2 share portions of some of the plant systems.

e. Shared Systemsfor Multi-Unit Plant The IPEEE submittal did not discuss any shared systems between the two units. Ilowever, it is inferred that there are some dependencies between the two units. These dependencies can be identified from some of the data presented in the IPEEE submittal. It can be concluded that the licensee has explicitly addressed such dependencies in its models, and in plant walkdowns.

.f Most Signifcant Human Actions i

Evaluation of human actions should be an integral part of the fire scenarie quantification. The most significant group of fire scenarios in the IPEEE was associated with the m6n control room and relay rooms. From the disemsions provided in the submittal, it is difficult to glean the most significant human actions. However, as was stated earlier, the possibility of using the RSP has not been considered and the human error rates of the IPE was directly employed.

2.2.10 Fire Scenarios and Core Damage Frequency Evaluation The licensee assessed a fire core damage frequency of 10* per cycle. The IPEEE submittal did not provide an explicit definition for the term "per cycle". It is inferred, however, that this terminology applies to one complete refueling cycle. This result for fire CDF is several orders of magnitude smaller than CDF values Energy Research, Inc. 22 ERI/NRC 95-512

~ - - 1 typically reported in fire PRAs. This CDF is based on those fire scenarios that have not been screened out, in Reference [7], the licensee explained that the small probabilities used in the IPE internal events analysis has led to the small overall CDF estimate for tire events. 4 The IPEEE submittal did not provide sutlicient detail to verify a chain of computations. Numerous tables and figures have been provided in the IPEEE submittal. However, verification of the final result is not possible.

2.2.11 Analysis of Containment Performance

a. Signifcant C<mtainment Performance Insights The submittal included a review of containment performance, as part of tire-initiated accident sequences.

It was concluded that containment performance is the same as that analyzed in the IPE. It is inferred that containment failures were established via the failure of support systems and of containment-related frontline systems.

b. Plant-Unique Phenomenology Considered No containment-related event trees have been used in any of the screening phases, ner in evaluating unscreened fire zones.

2.2.12 Treatment of Fire Risk Scoping Study issues

a. Assumptions Used to Address Fire Risk Scoping Study issues All fire risk scoping study issues have been addressed and resolved in the fire IPEEE. In addition, the licensee addressed the issues raised in GI-57, " Effects of Fire Protection System Actuation on Safety Related Equipment." The licensee has presented a detailed discussion for each issue.

1. Control system interaction was addressed in terms of the models used in the IPE and by consideration of the remote shutdown panel. The possibility of flow diversion away from the reactor vessel was considered in the analysis. The possibility of spurious actuation of valves was deemed to be risk insignificant.

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2. Seismic-fire interaction was addressed in terms of the potential for tire occurrence from the effects of an eanhquake. All electrical cabinets were considered to be well anchored. The potential for a lube oil or a diesel fuel oil fire was analyzed in detail using the program COMPBRN. A walkdown of the plant has verified that there are no significant concerns related to non-safety equipment or non-seismic equipment causing safety-equipment damage, by catching fire. The potential for release from hydrogen bottles has been analyzed using COMPBRN.

3. No credit was taken for fire detection and suppression in the IPEEE. However, the licensee has provided a detailed description of the detecti n and suppression systems. The licensee indicated that there are close to 200 members in the plant fire brigade, and that these members undergo training and drills on a regular basis, exceeding the requirements of regulations.

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e In addition to the description in the submittal, Reference [7] noted that the tire protection system is not designed to withstand the SSE (0.lg PGA), and the SME (0.3g PG A) is expected to fail portions of the fire protection system. Some specille conditions encountered during the fire walkdown were cited, including:

Fire pumps were found to be housed in the non-seismically designed circulating water pumphouse. .

The CO 2supply tank is located in the yard, and is not seismically supported.

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Batteries used to start the diesel-driven fire pump were found not to have spacers between cells, and did not have end stops on the battery racks.

A number of small, unsupported metal cabinets are free to tip over during a seismic event.

The submittal concluded that, although weaknesses in fire suppression equipment were identified, sufficient equipment is expecied to be operable for achieving safe shutdown, because of the small potential for seismically induced fire and because of the incapacity of inadvertent tire suppression system actuation to disable safe shutdown equipment. Seismic-fire interactions at SSES were thus deemed to not be risk significant.

Based on the IPEEE submittal, together with supplementary information provided in Reference [7], the evaluation of seismic-fire interactions appears to have been adequately conducted. However, anomalous conditions were noted in fire protection system equipment, and yet, these were not corrected. In most instances, the anomalous canditions could have been fixed with simple, low-cost improvements. Indeed, a primary purpose of an evaluation for seismic-fire interactions is to identify, evaluate, and rectify such anomalous conditions. It is considered a weakness that relevant plant improvements were nG undertaken.

2.1.13 Treatment of USI A-45 The success paths developed for the Susquehanna seismic IPEEE address decay heat removal requirements, following a seismic margin earthquake, via the suppression pool cooling mode (SPCM) of RHR. The normal shutdown cooling mode of RHR was not considered in either success path. The SSEL was developed assuming decay heat removal (DHR) requirements, under conditions of loss of offsite power and small LOCA, for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following the seismic margin earthquake. The elements of the RHR-SPCM decay heat removal systems have been evaluated as part of the cdnsideration of SSEL components, which included a seismic walkdown. For SPCM operation success, one RHRSW pump, one RHR pump, and the essential service water (ESW) system must function for RHR cooling. All equipment i required for operation of at least one division / train of SPCM have been included in the SSEL. The RHR-SPCM form of DHR relies on the suppression pool as the water source for cooling. Based on the walkdown findings, the submittal noted that, for the Unit-2 return valve (HV-251F024B), needed for RHR to the suppression pool, there exists a potential for seismic interaction between the valve operator motor and a nearby hand rail. In addition, the Unit-2 Division-1 AC power (MCC 2B237) for RHR and RHRSW valves, used in suppression pool cooling, may interact with HVAC ducting in close proximity. The submittal discounts the possible effects of these interactions, noting that there is time for manual valve operation. Thus, the submittal concludes that RHR-SPCM equipment are expected to survive the SME, Energy Research, Inc. 13 ERl/NRC 95-512

4. The potential for inadvertent actuation of the fire suppression system, and its resulting impacts on cabinets and other equipment, have been analyzed by the licensee. Vulnerable conditions have been found. Modifications have been proposed in terms of opening the drains to a cable spreading room and installing splash shields on vulnerable cabinets.

The effects of smoke and other products of combustion were deemed in the IPEEE to be slow acting, and therefore, were not consider x1 to be a risk-significant issue.

5. The cross-zone analysis of the fire IPEEE was cited with respect to adequacy of the fire barriers.

Also, the reliability of the penetrations and fire doors was discussed. Reference was made to a test where the non-Appendix-R rated penetrations and barriers were evaluated and found to be of adequate rating.

b. Significant Findings

1. The fire brigade undergoes sufficierd training.

2. The suppression systems include provisions to minimize spraying of safe shutdown equipment.

3. The possibility of a seismic event leading to tire was judged to be remote.

4. Modifications have been proposed in terms of opening the drains to a cable spreading room and installing splash shields on vulnerable cabinets.

2.2.13 USI A-45 issue

a. Methods ofRemoving Decay Heat The IPEEE analysis made use of the IPE models, which incorporate the entire array of heat removal capabilities of the plant.

b. Presence of 7hermo-Lag Thermo-Lag is used at Susquehanna. However, the licensee has not given any credit in the fire IPEEE l to the effectiveness of this insulating material. i 2.3 HFO Events HFO external events were analyzed by a progressive screening approach, to identify those events with core 4

damage frequency (CDF) contributions judged to be less than 10 per reactor year [1].

The general methodology utilized in the IPEEE HFO analysis has followed that presented in NUREG-1407 for the analysis of other events. The HFO assessment has inch.ded three phases:

1. Hazard analysis

2. Plant response (fragility) analysis Energy Research, Inc. 24 ERI/NRC 95-512 l

3. Risk determination and documentation Guidelines provided in GL 88-20, Supplement 4 [2], NUREG-1407 [3], NUREG/CR-2300 [12], and NUREG/CR-5042, Supplement 2 [13], were referenced in the IPEEE as the basis for completion of Step 1.

Progressive screening has consisted of the following steps:

Reviewing plant-specific hazard data and licensing bases.

Identifying significant changes since the plant operating license (OL) was issued.

Establishing whether or not the plant and facilities designs comply with the 1975 SRP criteria.

Determining whether or not the hazard frequency is acceptably low, if necessary.

• Performing a bounding analysis, if necessary.

Performing a probabilistic risk assessment, if necessary.

The following subsections provide a summary of the analysis performed for each type of hazard. ,

2.3.1 liigh Winds and Tornadoes 2.3.1.1 General Methodology i

An evaluation of high winds was performed in accordance with the recommended progressive screening approach [3]. Documentation of this evaluation was not provided in the IPEEE submittal report (Reference

[11), but was stated as being provided in PP&L Calculation EC-RISK-01, which was not available for the present review.

2.3.1.2 Plant-Specific flazard Data and Licensing Basis The design wind velocity for all structures was considered by the licensee to be 80 mph (at 30 feet above ground) with a 100-year recurrence interval. A gust factor of 1.1 was applied to the design wind velocity. l

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The tangential wind velocity was detined as 300 mph (290 mph for the diesel generator 'E' building),

whereas the translational velocity is 60 mph (70 mph for the diesel generator 'E' building). The same method used for transforming wind velocity into an effective design pressure for the design wind loading was also used for tornado wind loadings. There were, however, two adjustments mage in determining the tornado-generated etYeetive design pressures. The gust factor was considered to be unity, and there were no variations in velocity or velocity pressure considered due to the height above ground.

Tornado-generated missiles, such as wood planking, steel pipe, automobiles, steel rods, or utility poles have been evaluated in the design of the tornado-resistant structures. Table 3.3-2 of the final safety anMysis report (FSAR) identifies tornado wind protected systems with the corresponding tornado-resistant enclosure that protects that system.

No plant-specific hazard data was provided in the submittal. The licensing basis for wind and tornadoes was described in some detail.

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2.3.1.3 Significant Changes Since issuance of the Operating License A site review was performed by the licensee to identify any significant changes since issuance of the operating license that might be affected by the high-winds issue. The addition of new facilities / structures or changes in existing facilities / structures were the only developments identified that could potentially affect the original design conditions. Most, but not all, of these additions / changes have been designed to resist the high-wind loading conditions associated with extreme winds and tornadoes. Those additions / changes not designed to resist the high wind loading conditions were not considered to serve any safety-related function or to be important to the continued safe operation of the plant. Although the potential exists for a portion of these additions / changes to become tornado-generated missiles, it was judged that any such missiles would be enveloped by the existing postulated missiles considered in the design of the safety-related facilities / structures.

2.3.1.4 Significant Findings and Plant-Unique Features No significant findings related to high-wind effects were cited in the submittal. No plant-unique features were noted.

2.3.1.5 llazard Frequency No hazard frequency data was provided or cited in the submittal. Instead, the submittal reported that the SSES design conforms to the 1975 SRP criteria. Use of this NUREG-1407 screening criterion obviates the need for the licensee to assess hazard frequency, since conformance to the SRP is taken to indicate, directly, that :he CDF contribution is less than 10+ per reactor-year. It should be noted, that calculation sheet EC-RISK-01 was not provided by the licensee for the present review.

2.3.2 External Flooding 2.3.2.1 General Methodology An evaluation of external flooding has been performed in accordance with the recommended screening approach of NUREG-1407. Documentation of this evaluation was not provided in Reference [1], but was stated to be included in PP&L Calculation EC-RISK-1024, which was not available for the present review.

2.3.2.2 Plant-Specific Hazard Data and Licensing Basis No plant-specific hazard data for external flooding has been provided or cited in the submittal. Rather, the submittal states that historical data were taken into account in the development of the flooding design basis, in addition, the licensee stated that singular and multiple dam failures were considered for the Susquehanna River and its tributaries.

IIistorical data on the most severe ik)od events on record for this portion of the Susquehanna River in the vicinity of SSES were considered by the licensee during the development of the flood design basis. All individual potential ik)od-producing phenomena and the appropriate combinations of such phenomena were considered in establishing SSES as a " dry" site, secure from the effects of external flooding concerns.

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The probable maximum flood (PMF) water elevation, coincident with wind-generated waves, for the Susquehanna River was defined as 548.0 feet mean sea level (MSL), which is over 120 feet below the site grade elevation of 670.0 feet MSL The Susquehanna River is the only water system adjacent to SSES that could have an impact on site flooding and consequently was the only consideration, except for local runoff, in deriving the PMF-generated water elevation. The guidelines provided in Appendix A of Regulatory Guide 1.59 were followed throughout the SSES PMF analyses. .

The potential for seismically induced dam failures upstream of the SSES plant was investigated to determine if such failures could contribute to a flooding event for the Susquehanna River in the vicinity of SSES. In lieu of the more rigorous considerations defined in Appendix A of Regulatory Guide 1.59, a simplified but more conservative approach was taken which demonstrated a significant margin of safety at SSES for flooding resulting from upstream dam failures. Singular as well as multiple dam failures for the Susquehanna River and its tributaries were evaluated.

No information was provided in the submittal regarding GI-103, " Probable Maximum Precipitation (PMP)."

2.3.2.3 Significant Changes Since issuance of the Operating License 1

The licensee reported that there have been no significant changes since issuance of the operating license

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that would directly affect or increase the potential vulnerabilities due to the external flood design basis. l SSES remains a " dry" site, secure from any adverse effects of external flooding.

2.3.2.4 Significant Findings and Plant-Unique Features No significant findings related to external tiood events were cited in the submittal. No plant-unique features were noted.

2.3.2.5 Hazard Frequency No related plant-specific hazard data were cited in the submittal. Instead, the submittal reported that the SSES design conforms to the 1975 SRP criteria. Use of this NUREG-1407 screening basis obviates the need for the licensee to assess hazard frequency, since conformance to the SRP is taken to indicate, directly, that the CDF contribution is less than 104 per reactor-year. It should be noted that, calculation  !

sheet EC-RISK-1024 was not provided by the licensee for the present review. I 2.3.3 Transportation and Nearby Facility Accidents 2.3.3.1 General Methodology An evaluation of transportation and nearby facility accidents was performed in accordance with the recommended screening approach of NUREG-1407. Documentation of this evaluation was not provided in Reference [1], but was stated to be included in the SSES FSAR.

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2.3.3.2 Plant-Specific Ilazard Data and Licensing Basis The licensee stated that the SSES FSAR provided analyses either to establish that the probability of accidents such as exposure to hazardous chemical releases (sulfur dioxide and ammonia) was less than 10 '

(the level above which the event has to be included in the plant design basis), or that under pessimistic assumptions the consequences of accidents such as explosions, fires, or liquid spills would not adversely affect plant safety, because the nearest safety-related stractures and components of the plant are at a greater distance from the hazard than the damage zone of the hazard. The acceptance criteria of SRP Section 2.2.3 were, therefore, met.

2.3.3.3 Significant Changes Since Isr,a.nce of the Operating License With regard to changes related to transportation and nearby facilities accidents, the submittal identified no significant changes since the time of issuance of the plant operating license.

2.3.3.4 Significant Findings and Plant-Unique Features No significant findings related to transportation and nearby facilities accidents were cited in the submittal.

No plant-unique features were noted.

2.3.3.5 Ilazard Frequency No hazard frequency data was provided or cited in the submittal. Instead, the cubmittal reported that the SSES design conforms to the 1975 SRP criteria. Although the SSES FSAR was not available for the present review, conformance to the SRP is taken to indicate, directly, that the CDF contribution is less than 104 per reactor-year. Use of this NUREG-1407 screening basis obviates the need for the licensee to assess hazard frequency.

2.3.4 Other Events The submittal stated that the effects of other external hazards listed in Section 2 of NUREG 1407 were either included in other analyses (e.g., IPE or station blackout [SBOJ analysis) or are not applicable to the SSES site (e.g., volcanic activity). Therefore, no site-unique external hazards were found to exist at SSES, and thus, the licensee deemed that it was appropriate that only the five hazards suggested in NUREG-1407 be studied in the IPEEE. i As a result, no screening analysis of external events other than for seismic, fire, high winds and tornadoes, external flooding, and transportation and nearby facility accidents was provided in the HFO IPEEE submittal.

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2.4 Generic Safety Issues (GSI-147. GSI-148. and GSI-172) l l 2.4.1 GSI-147, " Fire-induced Alternate Shutdown / Control Panel Interaction" '

GSI-147 addresses the scenario of fire occurring in a plant (e.g., in the control room), and conditions l l which could develop that may create a number of potential control system vulnerabilities. Control system interactions can impact plant risk in the following ways:

• Electrical independence of remote shutdown control systems

• Loss of control power before transfer

• Total loss of system function Spurious actuation of components The licensee considered spurious actuations leading to LOCAs or interfacing system LOCAs. Since the submittal has followat the guidance provided in FIVE concerning control system interactions, all circuitry associated with remote shutdown is assumed to have been found to be electrically independent of the control room.

2.4.2 GSI-148, " Smoke Control and Manual Fire Fighting Effectiveness" GSI-148 addresses the effectiveness of manual fire-fighting in the presence of smoke. Smoke can impact plant risk in the following ways:

By reducing manual fire-tighting effectiveness and causing misdirected suppression efforts

• By damaging or degrading electronic equipment By hampering the operator's ability to safely shutdown the plant By initiating automatic fire protection systems in arer.s away from the fire Reference [14] identities possible reduction of manual fire-fighting effectiveness and causing misdirected suppression efforts as the central issue in GSI-148. Manual fire-fighting was not credited in the analysis.

Thus, the issue of manual fire-fighting effectiveness is not addressed in this TER.

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

Reference [14] provides the description of each MSRP issue stated below, and dilineates the scope of information that may be reported in an IPEEE submittal relevant to each such issue. The objective of this subsection is only to identify the location in the IPEEE submittal where information having potential relevance to GSI-172 may be found.

Common Cause Failures (CCFs) Related to Human Errors Descrintion of the Issue [14]: CCFs resulting from human errors include operator acts of commission or omission that could be initiating events, or could affect redundant safety-related trains needed to mitigate the events. Other human errors that could initiate CCFs include: manufacturing errors in components that affect redundant trains; and installation, maintenance or testing errors that are repeated on redundant trains.

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In IPEEEs, licensees were requested to address only the human errors involving operator recovery actions following the occurrence of external initiating events.

Section 3.5.2.5 of the Susquehanna IPEEE submittal provides some minimal information on operator recovery actione, as a consideration for success path selection in the seismic analysis. For the fire IPEEE analysis, the licensee did not conduct a human error analysis; rather, the human error rates from the IPE model were used.

Non-Safety-Related C<mtrol System / Safety-Related Protection System Dependencies Descrintion of the Issue [14]: Multiple failures in non-safety-related control systems may have an adverse impact on safety-related protection systems, as a result of potential unrecognized dependencies between control and protection systems. The concern is that plant-specific implementation of the regulations regarding separation and independence of control and protection systems may be inadequate. The licensees' IPE process should provide a framework for systematic evaluation of interdependence between safety-related and non-safety-related systems, and should identify potential sources of vulnerabilities. The dependencies between safety-related and non-safety-related systems resulting from external events -- i.e., '

concerns related to spatial and functional interactions - are addressed as part of " fire-induced alternate shutdown and control room panel interactions," GSI-147, for fire events, and " seismically induced spatial and functional interactions" for seismic events.

Informatian provided in the Susquehanna IPEEE submittal pertaining to seismically induced spatial and functional interactions is identified below (under the heading Seismically Induced Spatial and Functional Interactions), whereas information pertaining to fire-induced alternate shutdown and control panel interactions has already been identitied in Section 2.4.1 of this TER.

Heat / Smoke / Water Propagation Effectsfrom Fires Description of the Issue [14]: Fire can damage one train of equipment in one fire zone, while a redundant train could potentially be damaged in one of following ways:

j Heat, smoke, and water may propagats (e.g., through HVAC ducts or electrical conduit) into a second fire zone, and damage a redundsnt train of equipment.

A random failure, not related to the fire, could damage a redundant train. j

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Multiple non-safety-related control systems could be damaged by the tire. and their failures could affect safety-related protecti on equipment for a redundant train in a second zone.

A fire can cause unintended operation of equipment due to hot shorts, open circuits, and shorts to ground.

Consequently, components could be energized or de-energized, valves could fail open or closed, pumps could continue to run or fail to run, and electrical breakers could fail open or closed. The concern of water propagation effects resulting from fire is partially addressed in GI-57, " Effects of Fire Protection System Actuation on Safety-Related Equipment." The concern of smoke propagation effects is aAlressed in GSI-148. The concern of alternate shutdown / control room interactions (i.e., hot shorts and o,..er items just mentioned) is addressed in GSI-147.

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Information provided in the Susquehanna IPEEE submittal pertaining o GSI-147 and GSI-148 has alreaJy been identitied in Sections 2.4.1 and 2.4.2 of this TER. Section 4.8 o#the submittal presents information pertaining to this issue.

Effects of Fire Suppression System Actuation on Non-Safety-Related and Safety-Related Equipment Descrintion of the issue [14): Fire suppression system actuation events can han an adverse effect on safety-related components, either through direct contact with suppression agents or through indirect interaction with non-safety related components. This concern is addressed in GI-57.

Information pertaining to suppression-induced damage to equipment, as well as seismically induced inadvertent actuation of fire suppression systems can be found in Sections 4.8 and 4.2.2.2 o.f the submittal.

Reference [7] provides some brief additional information on seismically induced inadvertent actuation of fire suppression systems. Discussion on GI-57 is also provided in Section 4.8.

E.[fects ofFlooding and/or Moisture intrusion on Non-Safety-Related and Safety-Related Equipment Descrintion of the Issue [14]: Flooding and water intrusion events can affect safety-related equipment either directly or indirectly through flooding or moisture intrusion of multiple trains of non-safety-related equipment. This type of event can result from external flooding events, tank and pipe ruptures, actuations of fire suppression systems, or backtiow through parts of the plant drainage system. The IPE process addresses the concerns of moisture intrusion all internal flooding (i.e., tank and pipe ruptures or backtiow through part of the plant drainage system). The guidance for addressing the concern of external flooding is provided in Chapter 5 of NUREG-1407, and the concern of actuations of fire suppression systems is provided in Chapter 4 of NUREG-1407.

The following information is provided relevant to this issue: the Susquehanna IPEEE submittal discusses external floods in Section 5.2; discussion is provided in Section 4.8 regarding actuations of fire suppression systems; limited discussion of seismically induced inadvertent actuation of fire suppression systems is provided in Sections 4.8 and 4.2.2.2, and in Reference 17); and seismically induced external flooding is mentioned in Section 5.2.

Seismically induced Spatial and Functional Interactions Descrintion of the Issue [14]: Seismic events have the potential to cause multiple failures of safety-related systems through spatial and functional interactions. Some particular sources of concern include: ruptures in small piping that may disable essential plant shutdown systems; direct impact of non-seismically qualified structures, systems, and components that may cause small piping failures; seismic functional interactions of control and safety-relat<1 protection systems via multiple non-safety-related control systems' failures; and indirect impacts, such as dust generation, disabling essential plant shutdown systems. As part of the IPEEE, it was specifically requested that seismically induced spatial interactions be addressed during plant walkdowns. The guidance for performing such walkdowns can be found in EPRI NP-6041.

l The Susquehanna IPEEE has included a seismic walkdown which investigated the potential for adverse physical interactions. The submittal states that EPRI NP-6041-SL guidelines were followed in the seismic walkdowns. Relevant information can be found in Sections 3.10 and 3.4.9 of the submittal.

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t Seismically Induced Fires Descrintion of the Issue [14]: Seismically induced fires may cause multiple failures of safety-related systems. The occurrence of a seismic event could create fires in multiple locations, simultaneously degrade fire suppression capability, and prevent mitigation of fire damage to multiple safety-related systems. Seismically induced tires is one aspect of seismic-tire interaction concerns, which is addressed as part of the Fire Risk Scoping Study (FRSS) issues. (IPEEE guidance specifically requested licensees to evaluate FRSS issues.) In IPEEEs, seismically induced fires should be addressed by means of a focused seismic-tire interactions walkdown that follows the guidance of EPRI NP-6041.

Section 4.8 of the Susquehanna IPEEE submittal provides discussion regarding seismically induced fires.

Seismically Induced Fire Suppression System Actuation Descrintion of the issue [14]: Seismic events can potentially cause multiple fire suppression system actuations which, in turn, may cause failures of redundant trains of safety-related systems. Analyses currently required by fire protection regulations generally only examine inadvertent actuations of fire suppression systems as single, independent events, whereas a seismic event could cause multiple actuations of fire suppression systems in various areas.

Sections 4.8 and 4.2.2.2 of the Susquehanna IPEEE submittal, and Reference [7), provide some limited discussion of seismically induced tire suppression system actuation.

Seismically Induced Flooding Descrintion of the Issue [14]: Seismically induced flooding events can potentially cause multiple failures of safety-related systems. Rupture of small piping could provide flood sources that could potentially affect multiple safety-related components simultaneously. Similarly, non-seismically qualified tanks are a potential flood source of concern. IPEEE guidance specifically requested licensees to address this issue.

The Susquehanna IPEEE submittal does not describe any specific evaluation of seismically induced internal flooding, other than consideration of the potential and effects of seismically induced inadvertent actuation of fire suppression systems (as conveyed in Sections 4.8 and 4.2.2.2 of the submittal, and in Reference

[7]). Section 5.2 briefly mentions the topic of seismically induced external flooding.

4 Seismically Induced Relay Charter Descrintion of the Issue [14): Essential relays must operate during and after an earthquake, and must meet one of the following conditions:

remain functional (i.e., without occurrence of contact chattering);

• be seismically qualified; or

• be chatter acceptable.

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e It is possible that contact chatter of relays not required to operate during seismic events may produce some unanalyzed faulting mode that may affect the operability of equipment required to mitigate the event.

IPEEE guidance specincally requested licensees to address the issue of relay chatter.

The Susquehanna IPEEE submittal provides information pertaining to relay chatter in Sections 3.5.3.2, 3.10, and 3.11.2.

Evaluation of Earthquake Magnitudes Greater than the Safe Shutdoun Earthquake Descriplion of the Issue [14]: The concern of this issue is that adequate margin may not have been included in the design of some safety-related equipment. As part of the IPEEE, all licensees :re expected to identify potential seismic vulnerabilities or assess the seismic capacities of their plants either by performing seismic PRAs or seismic margins assessments (SMAs). The licensee's evaluation for potential vulnerabilities (or unusually low plant seismic capacity) due to seismic events should address this issue. j I

The Susquehanna IPEEE has included a focused-scope seismic margin assessment, as documented in l

Section 3 of the submittal. The seismic input for the analysis is described in Sections 3.7 and 3.8. 1 Effects ofHydrogen Line Ruptures Descrintion of the Issue [14): liydrogen is used in electricai generators at nuclear plants to reduce windage losses, and as a heat transfer agent. It is also used in some tanks (e.g., volume control tanks) as a cover gas. Leaks or breaks in hydrogen supply piping could result in the accumulation of a combustible mixture of air and hydrogen in vital areas, resulting in a fire and/or an explosion that could damage vital safety-related systems in the plants. It should be anticipated that the licensee will treat the hydrogen lines and tanks as potential fixed fire sources as described in EPRI's FIVE guide, assess the effects of hydrogen line and tank ruptures, and report the results in the fire portion of the IPEEE submittal.

The Susquehanna IPEEE submittal provides some limited information regarding the potential of seismically induced ruptures of hydrogen storage and hydrogen piping in Section 4.8.

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3 OVERALL EVALUATION AND CONCLUSIONS 3.1 Seismic The approach chosen by the licensee for conducting the seismic IPEEE of Susquehanna Steam Electric Station essentially addresses the niajor elements specified in NUREG-1407 as recommended items that should be considered for a focused-scope plant. The submittal itself gives a clear description of the seismic evaluation, although the format of documentation does not follow that recommended in NUREG-1407.

The identification and implementation of safety enhancements has produced some meaningful insights in response to the objectives of GL 88-20, for a focused-scope plant, liowever, this aspect of the seismic IPEEE is incomplete, since no actions have been proposed to increase plant IICLPF capacity above the existing level of 0.2ig. In addition, a number of elements of the seismic IPEEE are considered to fall short of providing a complete understanding of potential severe accident behavior.

The focused-scope seismic evaluation for Susquehanna Steam Electric Station has identified a number of issues which required implementation of resolution approaches. A few equipment modifications have been proposed / implemented, and procedures to improve seismic housekeeping have been considered. These safety enhancements have been addressed primarily with respect to a review of safe shutdown equipment; no improvements were found specifically/ exclusively as a result of a containment performance evaluation.

Based on this submittal-only review, the following items are identified as the primary strengths and weaknesses of the seismic IPEEE submittal for Susquehanna Steam Electric Station:

Strengths

1. The study has implemented a relevant, meaningful approach for conducting the seismic IPEEE, and considers, to some degree, all major issues of concern.

2. The study has made significant use of plant personnel (who received training in seismic evaluation methods), thus ensuring that the licensee has become familiar with the evaluation.

3. The submittal is generally clear and well-written (however, it does not follow the recommended format of NUREG-1407).

4. The background on structures, on structural models and parameters, a'pd on relevant work conducted in the past was detailed and useful as a reference point for review.

5. The submittal provided a comparatively broad analysis of soil failures.

6. The detailed description of walkdown findings, HCLPF results, and outlier resolutions, as provided in Section 3.10.6 of the sobmittal, is well presented and facilitates understanding of significant study insights.

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Weaknesses

1. The submittal's treatment of seismic containment performance was superficial, and did not involve walkdowns or evaluations of components needed for successful early accident mitigation following an SME, as recommended in NUREG-1407

2. The licensee's assumption that valves, which may potentially be damaged through adverse seismic-spatial interactions, would be manually operated following the SME, so as to maintain success paths, appears tenuous.

3. The seismic IPEEE has identified deficiencies with respect to (a) adjacent panels and cabinets in close proximity that are not fastened together, (b) unanchored color video CRTs, and (c) IICLPF capacities of four components (two valves, an automatic transfer switch, and a motor control center). liowever, meaningful explanation as to why no resolutions have been proposed for these items has not been provided.

4. Anomalous conditions were encountered during the walkdown evaluation of seismic-fire interactions; however, no meaningful explanation has been provided as to why no resolutions have been proposed for these conditions.

5. The submittal did not apply screening criteria to non-seismic failures and human actions, as recommended in NUREG-1407 for an EPRI SMA.

6. The validity of the licensee's approach for evaluating IICLPF capacities could not be verified.

The licensee's response to related RAls has suggested that the evaluation of 11CLPF capacities for important components may not have been adequately performed, in accordance with established procedures.

The present review recommends that the NRC pursue an audit of the licensee's llCLPF calculations for identified important components.

Despite the foregoing weaknesses, it is clear that the licensee has benefitted substantially from the seismic IPEEE effort, that some meaningful plant improvements have been considered, and that the licensee has improved its understanding of plant behavior in response to potential severe, accidents caused by earthquakes. t 3.2 Etr The licensee has expended considerable effort in the preparation of the fire analysis portion of the IPEEE.

The licensee states that Level-1 PRA methodology has been used to identify potential fire vulnerabilities.

The IPE plant impact model has been used to identify critical equipment and cables, and to conduct the core damage frequency (CDF) analysis.

Based on this submittal-only review, the following areas are identified as the primary strengths and weaknesses of the fire IPEEE submitted for Susquehanna Steam Electric Station:

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Strengths

1. The tire analysis portion of the IPEEE is well written. The overall presentation is clear and well organized. Tables and figures provide a considerable amount of supporting information.

2. State-of-the-art data were used in the fire analysis.

3. The list of fire-related issues addressed by the licensee was extensive.

Weaknesses

1. The total fire CDF was found to be 10'/ cycle. This result is much smaller than values typically obtained for BWR plants. The reason for the small frequency is that the conditional core damage frequency adopted from the IPE model is very small, v hich in turn is dominated by human actions and recovery activities. Given the screening approach, and the method used for identifying vulnerabilities, several fire scenarios that could have otherwise contributed to risk have been excluded from the CDF analysis.

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2. In the screening protocol, the licensee screened out fire zones based on combustible loading, assuming that cables do not contribute to this factor. For example, there are several cable spreading rooms in this plant which were screened out on the basis of their low combustible loading (which implies transient-fueled fires are not possible). For such areas where a large array of important cables are present, this approach can lead to optimistic conclusions, and hence, potentially mask an important vulnerability.

3. The licensee did not conduct a tire scenario-specific human error analysis. The human error rates from the IPE model were used, based on the premise that the conditions influencing the operator actions would not be much different from those under internal event scenarios. The human action and recovery analyses of the IPE cannot be used for fire induced core damage frequency evaluation j without an adjustment for the influences of fire on the operators (e.g., the effect on man-power, I control room habitability, control panel alarms and instrumentation readings, etc.). The core damage frequency, as reported in the IPEEE submittal, may be optimistic because of this omission. l

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4. The remote shutdown panel was not modeled based on the premise that coistrol room evacuation would not be necessary under any fire conditions. This conclusion is contrary to industry practices and lack of analysis of the Remote Shutdown Panel, which is installed for mitigating a control room or cable spreading room fire, is a glaring omission in the fire analysis.

5. The licensee has assumed that fires originating in a cabinet will not affect cables and equipment outside the cabinet. This assumption is not valid in general, and especially when there exist openings on top of a cabinet.

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6. The licensee has assumed that small motors (< 50 hp) do not pose a fire threat or fire ignition I source to other materials. Without performing a thorough analysis of every small motor, the licensee cannot categorically assume that the tire risk stemming from these motors is insignificant.

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7. The sensitivity of the final results to specific assumptions was not presented in the submittal. The assumption regarding cable chases and valve galleries had a significant impact on the final results.

The fire analysis, as is evident from the above discussions, in'luded e several weaknesses that may undermine the final conclusions. Although every weakness identified above may not lead to a significant effect on the final results; collectively, however, they may lead to optimistic results, and raise some concerns over the validity of the final conclusions.

3.3 HFO Esents The liFO IPEEE submittal has generally followed the recommended guidelines and basic steps fbr analyzing and reporting potential accident scenarios due to other external events. As judged from this submittal-only review, the following items are viewed as strengths of the liFO IPEEE submittal for Susquehanna Steam Electric Station (no weaknesses are identified):

1. The IIFO analysis was clearly described, liowever, some external event studies were simply cited as sources of the evaluations, and were not incorporated into the IPEEE submittal report, thus preventing a complete review. These studies were, nonetheless, clearly referenced in the submittal.

2. The screening basis employed (conformance to SRP criteria) was clearly identified, and referenced explicitly back to the guidance in NUREG-1407.

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4 IPEEE INSIGIITS, IMPROVEMENTS, AND CONF allTMENTS -

i 4.i = l The Susquehanna seismic IPEEE computed HCLPF capacities for components that were identi6ed as outliers from the seismic margin walkdown evaluations. HCLPF values of 0.21g and 0.26g were obtained for some SSEL components. (These HCLPF assessments have been made with respect to the NUREG/CR-0098 median,55 damped spectral shape, in accordance with the guidelines of NUREG-1407.) Hence, it can be concluded that the plant-level HCLPF capacity of Susquehanna Steam Electric Station does not exceed 0:21g. However, the licensee states that the low-capacity items are either not strictly required for the SSEL, or that their failure may be rectified through manual operations. Thus, the licensee concludes that the success paths meet the 0.3g RLE, and the licensee has not proposed any plant improvements to increase the plant HCLPF capacity.

A significant number of plant conditions and seismic housekeeping concerns were identified by the licensee as open issues requiring resolution. Table 4.1 at the end of this section (derived from IPEEE Section 3.10.6) summarizes these issues and the approaches implemented for their resolution. In some cases (as just noted), HCLPF capacities were determined which do not meet the RLE; but these items were judged in the IPEEE not to warrant plant modification. Other issues were resolved by judgments and/or analyses.

In addition, some minor fixes and enhancements have been proposed for implementation (or have already been implemented) to resolve other issues. From among the foregoing items, the seismic IPEEE submittal

. has specifically drawn out the following findings / improvements as being most significant:

HCLPF Results that Do Not Meet RLE 0.21g HCLPF for HPCI Pump Discharge Valve HV-155-F006; potential impact with adjacent non-Q valve.

0.21g HCLPF for RHR-SPCM-B Suppression Pool Inlet Valve HV-251-F024B; potential impact with adjacent platform handrail.

0.25g HCLPF for Automatic Transfer Switch OATS 556; potential impact with adjacent HVAC support.

0.26g HCLPF for 480 V Motor Control Center MCC-2B237; potentialimpact with adjacent HVAC duct.

Plant Defciencies (Interaction Concerns) that Have Been Fixed or Are Under Evaluation Small " trolleys' used to assist the removal of breakers from AC and DC switchgear cabinets (and load centers) were located on top of some of these cabinets. Although introduced as part of the original plant construction, these lifting devices were not part of the original equipment qualification. The trolleys were removed shortly after they were identified via EDR 94-018 and SOOR 94-222.

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• At several locations inside the control and relay rooms, the walkdowns identified control cabinets l and instrument panels in close proximity, that are not fastened together. Since these panels were l qualified in a " stand-alone" test configuration, the effect of potential impact Lads on internal l components was not addressed in the existing dynamie qualitication documentation. Dynamic l qualification of internal components has, therefore, been judged to be indeterminate. This condition is currently being tracked and dispositioned under EDR 94-030. .

The SMA walkdowns identified that the color video CRTs in Control Room Panels C651 and C601 are resting on, but are not fastened to, the panel internal supports. These color video CRTs could affect the existing dynamic qualitication of the other internal components during a dynamic event, should they slide off of their internal panel supports. This condition is currently being tracked and dispositioned under EDR 94-039.

Seismic Housekeeping Procedures The following issues have been identified, which are associated with seismic housekeeping and general work practices:

Office-type furnishings have been found in the control room, which could interact with nearby safety-related equipment.

• Housekeeping concerns, which mainly involve placing of transient items in close proximity to safety-related equipment, have been identified.

Equipment with missing or loose screws, missing nuts, and missing or broken latches have been encountered.

Such housekeeping items have been transmitted to the plant for corrective action, and are currently being tracked and dispositioned under SOOR 94-341. These items can be handled readily outside of the plant modification process. As part of the SSES Deficiency Management Program, corree:ive actions to  ;

minimize reemrence will be developed. Consideration will be given by the licensee to the following strategies:

• Review and revise, if necessary, existing procedures, programs, and specifiegtions which deal with location of transient equipment and associated safety impact. 4 Performance of periodic inspections of safety related equipmem, by an appropriate walkdown l

team, in an effort to reduce reoccurrence of similar dynamic interacion concerns.

• Training to help in the improvement of plant statrs seismic awareness. Two training presentations already have been provided at the time of submittal of the IPEEE report. Training of maintenance personnel is currently being pursued. It is expected that increased awareness will minimize recurrence of safety impact and housekeeping problems which could adversely affr,, seismic qualification of equipment.

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i The submittal indicates that only one hardware change has actually been implemented, i.e., the removal of trolley cranes. The submittal states that the risk significance of the other issues is low, but that these issues have been included in the SSES Discrepancy Management Program and will be resolved.

4.2 Elre Overall, the licensee has concluded that there are no significant fire vulnerabilities at Susquehanna Steam Electric Station. The total fire CDF, under full-power conditions, was estimated to be 10* per cycle.

(The term " cycle" is not defined, but it is inferred to represent a refueling cycle.) This result for tire-induced CDF is several orders of magnitude smaller than fire CDF values obtained from studies of other plants.

The significant fire zones were found to be the control room, the relay rooms, and the battery charger rooms. These rooms contain cables and cabinets from multiple trains of the ECCS. Detailed analyses have been conducted for these rooms, and multiple fire scenarios were considered.

The entire fire IPEEE effort has provided an excellent opportunity for licensee engineers to improve their knowledge of the characteristics of the plant, and of how the plant would behave under fire accident conditions.

No improvements or commitments were deemed necessary by the licensee to further reduce the tire risk at Susquehanna. However, related to fire risk scoping study issues, t% licensee found a few electrical cabinets that require splash guards, and also concluded that special provisions should be considered for draining water from a cable spreading room.

4.3 IIFO Events With regard to the IPEEE approach and major findings for HFO events, the submittal reports that the SSES specific hazards and licensing bases for high winds, external floods, and transportation and nearby facility accidents were reviewed. Changes since the time of issuance of the plant operating license were evaluated either specifically for the IPEEE or as part of regular updates of the FSAR. Conformance with the criteria of the 1975 SRP was, therefore, demonstrated. Thus, based on the results of the ser:ening approach suggested in the IPEEE generic letter supplement, no significant hazards were found to exist at SSES from these specific external threats.

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4 l As a result, no major safety enhancements have been identified related to HFO initiators, and consequently, no commitments have been made that would require tracking by the NRC. An assessment of the validity of this conclusion can not currently be made, unless additional information is obtained from the licensee.

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Table 4.1 Open issues, and Their Resolutions. as identified in the Susquehanna Seismie IPEEE WALKDOWN ISSUE RESOLUTION MCCs: The housekeeping items (e.g., interaction concerns associated with latching screws, Latching screws to doors of cubieles of MCCs cubiele doors, transient items, tool inserted in 1-are left open, left loose, missing, or cannot be in gap, and scaffolding attached to MCC's base) refastened because they are stripped, have been transmitted to the plant for corrective action and are currently being tracked and Doors to cubicles are taped shut. dispositioned under SOOR 94-341. Since these items can be handled readily outside of the Large tool box in the immediate vicinity of an modification process, they are not considered to MCC in an area not designated as ' transient be outliers.

storage area'.

The 1/16-in gap between the vertical flVAC Breaker tool insened into a 1-in gap between an duet stiffener and MCC 2B237 was assessed and MCC and an adjacent concrete wall. found to be sufficient for SSE loads, but insufficient for SME loads. This interaction Scaffolding attached to the base anchorage of an concern is considered to be an outlier.

MCC.

An HCLPF value of 0.26g was assessed for the A vertical llVAC duet, located to the east side MCC 2B237 interaction. This condition was of MCC 2B237, includes duct stiffeners which not judged to warrant a plant modification (see p are located approximately 1/16-in aw y from the 3-38 of IPEEE submittal).

MCC. During an SME event, the combined displacement of the HVAC duct and the MCC is more than 1/16-in, and they will collide.

Low-Voltage Switchgear: The housek.eeping item (i.e., interaction concern associated with loose latching screws) has been Each switchgear (load center) was found to have transmitted to the plant for corrective action and a breaker hoist located on top of the circuit is currently being tracked and dispositioned breaker cabinets. The breaker hoists were not under SOOR 94-341. Since this item can be secured in place and were free to move during a handled readily outside of tnp modification dynamic event. process, it is not considered b be an outlier.

The walkdown also identified one seismic The switchgears' breaker hoists were assessed interaction concern associated with loose as being outliers. The breaker hoists can induce latching screws on the front doors of the significant impact loads to the switchgear.

switchgear.

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WALKDOWN ISSUE RESOLUTION Low-Voltage Switchgear (cont 3: EDR 94-018 was written to address the breaker hoists on the switchgear. The breaker hoists were removed. A calculation was performed to evaluate the effect of the remaining steel associated with the breaker hoist assembly on the seismic qualitication of the switchgear. The switchgear's SQRT (seismic qualification report team) binder was updated to refleet the breaker hoist issue.

Medium-Voltage Switchgear: The housekeeping items (e.g., interaction concerns associated with transient items, door Placement of transient equipment / tool boxes in latches, missing bolts, and tools hung from the immediate vicinity of switchgear. switchgear) have been transmitted to the plant for corrective action and are currently being Door latches to individual cabinets are left open, tracked and dispositioned under SOOR 94-341.

left partially open, or over engaged. Since these items can be handled readily outside of the modification process, they are not Bolts missing from the sides of switchgear. considered to be outliers.

Tools hung from the sides of switchgear. GE PVD1IC relays were shown to have adequate functioning capability based on the GE type PVDilC relays are located in the qualitication test reports contained in the SQRT switchgear. binder; no chattering was recorded during testing.

GE type liFA65 relays are located in the switchgear. Based on qualification test reports, GE IIFA65 relays were shown to have adequate dynamic capability, except that chattering was detected in excess of 2 milliseconds. Because the relays are used for alarm function only, the contiguration is considered to be acceptable.

4 Horimntal Pumns: The housekeeping item has been transmitted to the plant for corrective action and is currently The walkdown identified one seismic interaction being tracked and dispositioned under SOOR concern associated with the placement of a 94-341. Since this item can be handled readily transient item in the immediate vicinity of a outside of the modification process, it is not pump. considered to be an outlier.

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e WALKDOWN ISSUE RESOLUTION Vertical Pumns: The maintenance item has been transmitted to the plant for corrective action. Since this item The walkdown identified two pumps having can be handled readily outside of the corroded fasteners. modification process, it is not considered to be -

an outlier.

Fluid-Onerated Valves: The unscreened items were assessed and found to have IICLPF capacities in excess of 0.3g 22 of 52 valves could not be screened out. PGA.

Motor-Onerated Valves: In regard to the interaction concerns, the available gaps between SSEL valves and other Clearances between SSEL Valve liv-155-F006 adjacent items were assessed and found to be and Valve 155018, Valve PSV-15513 and pipe sufficient, with the exception of Valves liF-155-support SP-DBB-120-li2003 are noted to be F006 and FIV-251-F024B for SME loads.

about 1 1/4-in,1/4-in, and 1-in, respectively.

With the exception of the two valves just noted, Clearance between stem of SSEL Valve liv- all unscreened valves were assessed and found 251 F024B and a handrail is noted to be about to have IICLPF capacities in excess of 0.3g 1/2-in. PGA.

Clearance between stem of SSEL Valve IIV- liCLPF values of 0.21g were assessed for the 255-F004 and a structural steel tube support is two valve interactions. These conditions were noted to be about 1-in. not judged to warrant plant modifications (see pp. 3-53 and 3-54 of IPEEE submittal).

27 of 96 valves could not be screened out.

Fans: The unsereened items were assessed and found to have llCLPF capacities in excess of 0.3g 8 of 9 fans could not be screened out. PGA.

Air Ilandlers: IIVAC systems in the reactor building have been judged to not be critical, to safe shutdown Several instances of inadequate clearances paths. Interaction concerns in the control between IIVAC components and adjacent items structure are judged to be acceptable, since they in the reactor building and control structure could only lead to local dents in llVAC ducting.

were noted.

The unscreened IIVAC dampers were assessed 21 of 30 items could not be screened out. and found to have liCLPF capacities in excess of 0.3g PGA.

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Chillers: The unscreened chillers were assessed and I found to have liCLPF capacities in excess of '

4 of 4 items could not be screened out. 0.3g PGA.

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WALKDOWN ISSUE RESOLUTION Load Centers Distribution Panels and Fuse The housekeeping items (e.g., interaction Boxes: concerns associated with transient items, loose or missing fasteners, and unengaged latching Each load center was found to have a breaker mechanisms) have been transmitted to the plant hoist located on top ofits enclosure. for corrective action and are currently being tracked and dispositioned under SOOR 94-341.

Several interaction concerns were noted Since these items can be handled readily outside associated with placement of transient items in of the modification process, they are not the immediate vicinity ofload centers and considered to be outliers.

distribution panels. Some distribution panels and fuse boxes were identified as having loose The load centers' breaker hoists were assessed or missing fasteners or unengaged latching as being outliers. The breaker hoists can induce mechanisms. significant impact loads to the switchgear.

Westinghouse SV relays are located in Panel EDR 94-018 was written to address the breaker OD597, hoists on the load centers. The breaker hoists were removed. A calculation was performed to evaluate the effect of the remaining steel associated with the breaker hoist assembly on the seismic qualification of the load centers.

The load centers' SQRT (seismic qualitication report team) binder was updated to refleet the breaker hoist issue.

Westinghouse SV relays were shown to have adequate functioning capability based on the qualification test reports contained in the SQRT binder; no chattering was recorded during l testing. I Battery Rack Assemblies: A listing of battery racks found to have loc.:e j fasteners has been transmitted to the plant for Battery racks were found to be close to block corrective action. The housekeeping item is walls, currently being tracked and dispositioned under SOOR 94-341. Since these items can be Some battery racks were found to have loose handled readily outside of the modification fasteners. process, they are not considered to be outliers.

For some battery racks, space was found The potential block-wall interaction, and the between the battery and the end rail, unfulfilled battery-space caveat, w ere evaluated and found to have no significance.

Energy Research, Inc. 44 ERI/NRC 95-512

l WALKDOWN ISSUE RESOLUTION Battery Chargers and Inverters: The housekeeping items have been transmitted to the plant for corrective action and are Two seismic interaction concerns were identified currently being tracked and dispositioned under associated with placement of transient items in SOOR 94-341. Since these items can be the immediate vicinity of battery chargers and handled readily outside of the modification inverters, process, they are not considered to be outliers.

Two battery chargers were identified as having either an inoperable latching mechanism or missing panel fasteners.

Instruments on Racks: The pipe / support / rack interaction was evaluated and found to be acceptable, by calculating the {

A potential seismic interaction between the 1-in total dynamic displacements of the rack and the DCB-212 pipe, one of its pipe supports, and 1-in pipe, including one of its supports. 1 Rack 2C005 has been identified.  ;

The housekeeping items (placement of Seismic interaction concerns associated with scaffolding) have been transmitted to the plant 1 I

placement of scaftbiding in the immediate for corrective action and are currently being vicinity of instrument racks were identified. tracked and dispositioned under SOOR 94-341.

Since these items can be handled readily outside 24 of 59 items could not be screened out. of the modification process, they are not considered to be outliers.

The unscreened items were assessed and found to have HCLPF capacities in excess of 0.3g PGA.

Control and Instrumentation Panele and Gaps among adjacent, unfastened, cabinets have Cabinets: been documented as a deficiency in EDR 94-030.

At several locations inside the control and relay rooms, gaps (less than 1/2-in) exist between The housekeeping items (tragsient items, adjacent control cabinets and instrumentation inoperable latching mechanism, and missing panels that are not fastened together. panel fasteners) have been transmitted to the plant for corrective action and are currently Transient items found leaning on, or placed being tracked and dispositioned under SOOR near, control and instrumentation panels.94-341. Since these items can be handled readily outside of the modification process, they Some panels have inoperable latching are not considered to be outliers.

mechanism or missing panel fasteners.

1 Energy Research, Inc. 45 ERl/PRC 95 512 l

WALKDOWN ISSUE RESOLUTION 1

{

Control and Instrumentation Panels and Cabinets The situation with Panel IC680 was judged ta l (cont.L be acceptable.

l l

Top of Panel IC680 is 1/8-in away from a duct- The situation with Panel OC512E-B was mounted balancing damper. evaluated and found to be acceptable.

Transfer Panel OC512E-B is 5/8-in away from a GE CFVB relays were shown to have adequate tube section supporting cable trays. functioning capability based on qualitication test reports and on the low seismic demand at the Panels OC519A,B,C,D house low ruggedness panels' floor locations. These relays are used relays (GE CFVB). only for alarm function.

Color video CRTs in Panels C651 and C601 are The condition with CRTs has been documented resting on, but are not fastened to, the panel as a deficiency in EDR 94-039, and is being internal supports, resolved through the SSES Deficiency Management Program.

Automatic Transfer Switches An HCLPF value of 0.25g was assessed for the automatic transfer switch OATS 556. This Top of automatic transfer switch OATS 556 in condition was not judged to warrant a plant diesel generator building 'E' is less than 1/2-in modification (see p. 3-86 of IPEEE submittal).

from a structural steel channel supporting a HVAC duct. Other unscreened items were assessed and found to have HCLPF capacities in excess of 0.3g 3 of 5 items could not be screened out. PGA.

Filters and Supnression Pool Strainers: All unscreened items were assessed and found to have HCLPF capacities in excess of 0.3g PGA.

20 of 20 items could not be screened out.

Miscellaneous - Solenoid Valves: All unscreened items were assessed and found to have HCLPF capacities in excess of 0.3g PGA.

15 of 37 items could not be screened out. ,

Miscellaneous - CIG Bottles: All unscreened items were assessed and found to have HCLPF capacities in excess of 0.3g PGA.

26 of 26 items could not he screened out.

Energy Research, Inc. 46 ERI/NRC 95-512

0 WALKDOWN ISSUE RESOLUTION Check Valves: The housekeeping item (scaffolding near valve) has been transmitted to the plant foi corrective A seismic interaction concern was noted action and is currently being tracked and pertaining to placement of scaffolding in the dispositioned under SOOR 94-341. Since this .

immediate vicinity of Valve 251-F018D. item can be handled readily outside of the modification process, it is not considered to be 7 of 86 items could not be screened out. an outlier.

All unscreened items were assessed ar.d found to have liCLPF capacities in excess of 0.3g PGA.

i Energy Research, Inc. 47 ERI/NRC 95-512 b

. 1 1

5 IPEEE EVALUATION AND DATA

SUMMARY

SIIEETS Completed data entry sheets for the Susquehanna Steam Electric Str.aon IPEEE are provided in Tables 5.1 to 5.3. These tables have been completed in accordance with the descriptions in Reference [6]. Table 5.1 lists the overall external events results. Table 5.2 summa;izes general seismic data pertaining to the j focused-scope seismic evaluation. Table 5.3 provides the BWR Seismic Success Paths table, which gives I a description of the success paths developed for the focused-scope seismic evaluation. The IPEEE submittal has provided some information regarding the f equences of events leading to tire-induced core i damage; however, all fire-accident scenarios were determined to have a frequency less than 10*/ cycle.

Ilence, no data sununary tables are provided pertaining to the tire evaluation. Also, no PRA or bounding analyses were performed as part of the Susquehanna liFO-events IPEEE: hence, no data summary tables are provided pertaining to evaluation for these external ever ts.

4 Energy Research, Inc. 48 ERI/NRC 95-512

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v 6 REFERENCES

1. "Susquehanna Steam Electric Station Individual Plant Examination of External Events (IPEEE),"

Pennsylvania Power & Light Company, June 27,1994.

2. " Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities -

10CFR50.54(f)," U. S. Nuclear Regulatory Commission Generic Letter 88-20, Supplement 4, June 28,1991.

3. J. T. Chen, et al., " Procedure and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities " U. S. Nuclear Regulatory Commission, NUREG-1407, May 1991.

4. R.T. Sewell, et al., " Individual Plant Examination for External Events: Review Guidance,"

ERI/NRC 94-501, Draft, May 1994

5. "lPEEE Step i Review Guidance Document," U. S. Nuclear Regulatory Commission, June 1992.

6. "lPEEE Database Data Entry Sheet Package for FIN A-0467, Development of Guidance and Screening Criteria for hdividual Plant Examinations on External Events (IPEEE): Task 9, Development ofIPEEE Database," Revised, Lawrence Livermore National Laboratory, December 1993.

7. "Susquehanna Steam Electric Station Response to Request for Additional Information on Individual Plant Examination of External Events (IPEEE) Submittal, Units I and 2 (TAC Nos. M74478 and M744790," letter to NRC Document Control Desk, from R. G. Byram, Pennsylvania Power &

Light Company, August 9,1996.

8. " Development of Criteria for Seismic Review of Selected Nuclear Power Plants," U. S. Nuclear Regulatory Commission, NUREG/CR-0098, May 1978.

9. "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," Revision 1, Electric Power Research Institute, EPRI NP-6041-SL, August 1991.

10. " Fire Event Database for U.S. Nuclear Power Plants," Electric Power Reseircl' Institute, NS AC-178L, Final Report, June 1992.

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

Sandia National Laboratories, April 1994.

12. "PRA Procedures Guide," American Nuclear Society and Institute of Electrical and Electronics Engineers, U.S. Nuclear Regulatory Commission, NUREG/CR-2300, January 1983.

13. "Evaluatica of External Hazards to Nuclear Power Plants in the United States, Other External Events," Lawrence Livermore National Laboratory, NUREG/CR-5042. " eplement 2, February 1989.

Energy Research, Inc. 52 ERI/NRC 95-512

- a _

14. " Staff Guidance ofIPEEE Submittal Review on Resolution of Generic or Unresolved Safety issues (GSIIUSI)," U.S. Nuclear Regulatory Commission, August 21,1997.

I Energy Research. Inc. 53 ERl/NRC 95-512

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1 i

ATTACHMENT 2 SUPPLEMENTAL TECHNICAL EVALUATION REPORT ON THE REVIEW OF THE i

}

INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE) SUBMITTAL AT SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2 l

i

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