ML17229B001

Technical Evaluation Rept on 'Submittal-Only' Review of IPEEE at St Lucie Nuclear Plant,Units 1 & 2.

ML17229B001
Person / Time
Site:	Saint Lucie
Issue date:	01/31/1998
From:	KAZARIANS M, MOSLEH A, SEWELL R T ENERGY RESEARCH, INC., External (Affiliation Not Assigned), MARYLAND, UNIV. OF, COLLEGE PARK, MD
To:	NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
Shared Package
ML17229A999	List: ML17229B000 ML17229B001 ML17229B002
References
CON-NRC-04-94-050, CON-NRC-4-94-50 ERI-NRC-95-504, NUDOCS 9902020322
Download: ML17229B001 (103)

Text

ERIINRC 95-504 TCCHNICRL CVRLURTION RCPORT ON TH6"SUBMITTAL-ONLY" 86VlGLU OF TH6 INDIVIDURL PLRNT CXRMINRTION OF tXTGRNRL EVENTS RT ST.LUCIC NUCLGRR PLRNT, UNITS 1 RND 2 FINAL REPORT Completed:

January 1997 Final: January 1998 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 r.:."...Vi902020322'~.990g2~awe)C~

mm~~;,+PDR ADOCV.OSOOOSSS P'PDR ERI/NRC 95-504 TECHNICAL EVALUATION REPORT ON THE"SUBMITTALQNLY" REVIEW OF THE INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS AT ST.LUCIE NUCLEAR PLANT (UNITS I AND 2)FINAL REPORT Completed:

January 1997 Final: January 1998 M.Khatib-Rahbar Principal Investigator Authors: R.T.Sewell, M.Kazarians', A.Mosleh-', and A.S.Kuritzky Energy Research, Inc.P.O.Box 2034 Rockville, Maryland 20847 Work Performed Under the Auspices of the..United States Nuclear Regulatory Commission Office of Nuclear Regula'.ory Research Washington.

D.C.20555 Contract No.04-94-050'azarians and Associates, 425 East Colorado Street, Suite 545.Glendale, CA 91205'niversity of Maryland, Deparunent of Materials and Nuclear Enginceriiig, College Park.MD 20742 I~a p.%fg,~

TABLE OF CONTENTS EXECUTIVE

SUMMARY

PREFACE...ABBREVIATIONS

~~~~~~~~~~~~~~Vl Xll...Xlll INTRODUCTION....

1.1 Plant Characterization 1.2 Overview of the Licensee's IPEEE Process and 1.2.1 Seismic 1.2.2 Fire..1.2.3 HFO Events 1.3 Overview of Review Process and Activities...

1.3.1 Seismic.....................

1.3.2 Fire..1.3.3 HFO Events Iinportant Insights....

~~~~~~~~~~~~~~2 3 4 4 5 6 6 8 8~~10 10 11 11~~~~~~~~~13 14 14 14 14 15 15 15 valuations.....

~~~~16 16~~~~~~~0~.=..-..-17 18 18 19 19~~~~~~~~~~~~~~~~~~~~~~~~~~I~~20 21~~~~~~~~~~21 CONTRACTOR REVIEW FINDINGS 2.1 Seismic.~..2.1.1 Overview and Relevance of the Seismic IPEEE Process.2.1.2 Success Paths and Component List..2.1.3 Non-Seismic Failures and Human Actions 2.1.4 Seismic Input 2.1.5 Structural Responses and Component Demands 2.1.6'creening Criteria..2.1.7 Plant Walkdown Process 2.1.8 Evaluation of Outliers 2.1.9 Relay Chatter Evaluation 2.1.10 Soil Failure Analysis 2.1.11 Containment Performance Analysis..2.1.12 Seismic-Fire Interaction and Seismically Induced Flood E 2.1.13 Treatment of USI A<5 2.1.14 Peer Review Process 2.1.15 Summary Evaluation of Key Insights 22 Fire.....2.2.1 Overview and Relevance of the Fire IPEEE Process...2.2.2 Review of Plant Information and Walkdown 2.2.3 Fire-Induced Initiating Events....2.2.4 Screening of Fire Zones...2.2.5 Fire Hazard Analysis 2.2.6 Fire Growth and Propagation

2.2.7 Evaluation

of Component Fragilities and Failure Modes 2.2.8 Fire Detection and Suppression 2.2.9 Analysis of Plant Systems and Sequences............

Energy Research, Inc.f ERI/NRC 95-504 go I 1 4 2.3 2.4~~~~~~~~~~~~~~~~~~~~~~~\~~~\~~~~~N~~~2.2.10 Core Damage Frequency Evaluation

.2.2.11 Analysis of Containment Performance....

2.2.12 Treatment of Fire Risk Scoping Study Issues 2.2.13 USI A%5 Issue HFO Events 2.3.1 High Winds and Tornadoes 2.3.1.1 General Methodology

..2.3.1.2 Plant-Specific Hazard Data and Licensing Basis 2.3.1.3'ignificant Changes Since Issuance of the Operating License 2.3.1.4 Significant Findings and Plant-Unique Features 2.3.1.5 Hazard Frequency 2.3.1.6 Bounding Analysis...2.3.2 External Flooding...

2.3.2.1 General Methodology

....................

2.3.2.2 Plant-Specific Hazard Data and Licensing Basis 2.3.2.3 Significant Changes Since Issuance of the Operating License 2.3.2.4 Significant Findings and Plant-Unique Features 2.3.2.5 Hazard Frequency 2.3.3 Transportation and Nearby Facility Accidents.....2.3.3.1 General Methodology

....2.3.3.2 Plant-Specific Hazard Data and Licensing Basis 2.3.3.3 Significant Changes Since Issuance of the Operating License..2.3.3.4 Significant Findings and Plant-Unique Features 2.3.3.5 Hazard Frequency 2.3.4 Lightning and Others Generic Safety Issues (GS1-147, GSI-148 and GSI-172).....

2.4.1 GSI-147,"Fire-Induced Alternate Shutdown/Control Panel Interaction 2.4.2 GSI-148,"Smoke Control and Manual Fire Fighting Effectiveness".2.4.3 GS1-156,"Systematic Evaluation Program (SEP)"....2.4.4 GS1-172,"Multiple System Responses Program (MSRP)" 22 22 22 23 24 24 24 25 25 25 25 26 26 26 26 27 27 27 28 28 28 29 29 30 30 31 31 31 31 35 OVERALL EVALUATION AND CONCLUSIONS

....3.1 Seismic.3.2 Fire...3.3 HFO Events 42 IPEEE INSIGHTS, IMPROVEMENTS, AND COMMITMENTS 4.1 Seismic.4.2 Fire...~~~~~~~~4.3 HFO Events...~~~45 45 47 47 5 IPEEE EVALUATION AND DATA

SUMMARY

SHEETS...........

Energy Research, Inc.ERI/NRC 95-504 1

6 REFERENCES................

55 Energy Research, Inc.1V ERI/NRC 95-504 LIST OF TABLES Table 3.1 Comparison of FPL's Site-Specific Seismic IPEEE Program Versus NUREG-1407 Recommended Guidelines for a Reduced-Scope Seismic Evaluation

.40 Table 5.1 External Events Results 49 Table 5.2 SSM Seismic Fragility......50 Table 5.3 PWR Success Path Overview Table.........

~.....'.51 Table 5.4 PWR Seismic Success Paths 52 Table 5.5'WR Accident Sequence Overview Table-For Fire PRA Only..'...........

53'Table 5.6 PWR Accident Sequence Detailed Table-Fire PRA Only 54 Energy Research, Inc.ERI/NRC 95-504 EXECUTIVE

SUMMARY

This technical evaluation report (TER)documents a"submittal-only" review of the individual plant examination of external events (IPEEE)conducted for the St.Lucie Nuclear Plant, Units 1 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'irectly 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 RAIs.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 St.Lucie IPEEE submittal, particularly with respect to the objectives described in Generic Letter (GL)88-20, Supplement No.4, and the guidance presented in NUREG-1407.

Florida Power and Light Company (FPL)is the licensee of St.Lucie Unit 1 (St.Lucie-1)and St.Lucie Unit 2 (St.Lucie-2).The St.Lucie IPEEE submittal considers seismic;fire;and high winds, floods and other (HFO)external initiating events.The St.Lucie IPEEE was performed and reviewed by licensee and contractor personnel.

Licensee's IPEEE Process With respect to the seismic IPEEE, St.Lucie Nuclear Plant is assigned to the reduced-scope seismic review category in NUREG-1407.

FPL elected to implement a site-specific program for conducting the seismic IPEEE of St.Lucie Nuclear Plant.The site-specific program was developed primarily in response to GL 8742 for resolution of Unresolved Shfety Issue (USI)A-46 at Turkey Point, Units 3 and 4, and at St.Lucie Unit 1.St.Lucie Unit 2 is not a USI AA6 plant;nonetheless, the same site-specific approach was proposed for its seismic IPEEE.The site-specific program represents a"scaled-back" approach to USI A-46 resolution.

After meetings and correspondence.

with FPL, the NRC never designated its approval of the site-specific program for IPEEE resolution.

Nonetheless, FPL proceeded with use of its site-specific program as the basis for conducting the seismic IPEEE.The site-specific seismic adequacy evaluations conducted for St.Lucie Units 1 and 2 relied primarily on a plant walkdown that focused on component anchorage capability and the potential for adverse seismic-induced spatial interactions..A safe shutdown equipment list (SSEL)was developed based on a success path that assumes loss of offsite power (LQSP).The submittal does not describe the success path nor does it present a success path logic diagram.The evaluation approach does not explicitly address a small loss of coolant accident (LOCA).All components in the SSEL that had not been previously verified as having adequate seismic capacity were walked down by the seismic review team (SRT)~The seismic review team used its judgment in assessing adequacy of seismic anchorage capacity and in identifying spatial interaction concerns.Components with obviously rugged anchorage were screened out;components with questionable seismic anchorage were Energy Research, Inc.vi ERI/NRC 95-504 ra"j g+*<."~"L-E/pl[r.P.'C~w identified as potential outliers.Spatial interaction concerns were also identified as potential outliers.The potential outliers were analyzed in further detail, in order to make a final outlier designation.

Resolutions were proposed for, each designated outlier.Table 3.1 of this TER'compares the features of FPL's site-specific IPEEE program against the elements of a reduced-scope evaluation that have been recommended in NUREG-1407.

The table indicates that FPL's program addresses only a subset of the recommended items/guidelines.

The most significant differences in the two evaluation approaches are judged to be: a lesser scope of components in the FPL approach;a limited treatment of human actions in the St.Lucie study;and no treatment of containment systems in the FPL program.In addition, the format for documenting the seismic IPEEE has not followed the recommendations of NUREG-1407.

It is important to note that-based on findings of a site audit (which involved an inspection of FPL's Turkey Point Nuclear Plant), and pending follow-up action by the licensee-the NRC has reached closure on USI AA6 for St.Lucie-1.To a significant degree, the NRC's resolution of USI A46 concerns has served as direct basis for formulating corresponding review findings in this TER for similar IPEEE concerns at St.Lucie, Units 1 and 2.For the fire IPEEE, the licensee has conducted an extensive and detailed analysis of fire events at St.Lucie.Appendix R documentation has been used to establish fire-related plant features, as well as fire zones and areas.In addition to safe shutdown equipment defined by Appendix R, equipment modeled in the probabilistic risk assessment (PRA)were included in the fire analysis.To support the fire analysis, the licensee has conducted a walkdown of the facility, using engineers familiar with the plant and with fire analysis.The fire IPEEE freeze date is December 1993;this date has been used as the cut-off date for all documentation describing the plant.A consulting firm with e'xperience in fire risk analysis has assisted FPL analysts in the preparation of the fire analysis.The licensee has used fire-induced vulnerability evaluation (FIVE)methodology and associated fire frequency and failure data to evaluate the fire risk.Simple models have been used to evaluate fire damage and human recovery actions.To keep the analysis simple, none of the analyses presented takes into account the specific fire protection features for a given area, nor the specific operator actions for a fire scenario.For redundant train failure frequency evaluation, the PRA models and data of the plant have been used.The licensee has submitted a plan for Appendix R compliance and has addressed the majority of the issues raised as part of that plan, with supporting analysis.Some fire areas and redundant cables and equipment are protected by Thermo-lag.

The effectiveness of Thermo-lag and its importance at St.Lucie Nuclear Plant are currently under investigation by the licensee.For the HFO IPEEE, the general methodology utilized by the licensee conforms to that presented in NUREG-1407 for the analysis of other external events.The licensee has performed a detailed analysis of high winds, external flooding, and transportation and nearby facility accident hazards.Additionally, the other external events have been evaluated to ensure that there are no hazards unique to the plant.Among these other external events, lightning has been analyzed in greater detail.Key IPEEE Findings From the seismic IPEEE, the principal findings consist of qualitative walkdown insights, and few quantitative findings have been reported.The seismic adequacy evaluation for St.Lucie-1 revealed a number of outliers for which safety enhancements have been proposed in response to USI A-46.In addition, the licensee is undertaking follow-up actions to implement a strict seismic housekeeping program in response to concerns identified by the NRC in its USI A-46 review process.Enhancements for IPEEE-only components (i.'e., components outside the scope of USI AA6, but within the scope of IPEEE)were Energy Research, Inc.vii ERI/NRC 95-504

I t><."pi hA M~~~VW F" not addressed.

For St.Lucie-2, there were no concerns identified for the equipment, provided a walkdown of wall-mounted transformers would be performed, and that such transformers would be secured as necessary.

Again, no enhancements were addressed for IPEEEwnly components.

With respect to fire events, the licensee has reached the overall conclusion that there are no significant fire vulnerabilities at St.Lucie.With the exception of the control room, cable spreading room and the"B" switchgear room, all fire zones and areas were screened out based on a 10 per reactor-year (ry)core damage frequency (CDF)criterion.

The CDFs for control room fires were concluded to be 7.49 F10'/ry and 5.90x10'/ry for Units 1 and 2, respectively.

For the cable spreading rooms, the core damage frequencies were determined to be 6.95 x10'/ry and 5.64x10'/ry for Units 1 and 2, respectively.

For both areas (i.e., control room'and cable spreading room), the licensee cites several conservative assumptions in fire occurrence rate and fire severity, and concludes that these two areas do not pose a vulnerability.

The CDF for a fire in the"B" switchgeai rooai was concluded to be 4.30x10'/ry and 4.48 x10~/ry for Units 1 and 2, respectively.

Fire propagation modeling has been performed for this area, and the licensee has concluded that a fire would not propagate throughout the room.With respect to HFO events, the submittal states that the other external events do not present a significant risk to the plant.This conclusion has been reached without performing a detailed PRA, and no HFO core damage frequency is reported for the plant.The hazard-specific conclusions of the analysis are as follows:-Uf df f<<I f dg fdRI PI (SRP)criteria, or the hazard occurrence frequency was demonstrated to be acceptably low.Unit 2 design was found to conform to the SRP criteria, and as such, it was concluded that high winds/tornadoes do not pose a significant threat to the unit.R.IU'dfgd IURRgt Guide (R.G.)1.59 and SRP criteria, and as such, it was concluded that external floods pose no significant risk of a severe accident.-The St.Lucie Units 1 and 2 designs were determined to conform to SRP criteria, and as such, it was concluded that transportation and nearby facility accidents pose no significant risk of a severe accident.Lucie, and the impact of lightning on plant risk is bounded by the internal events analysis.No potential vulnerabilities with respect to any HFO event were identified.

Generic Issues and Unresolved Safety Issues For seismic events, USI A45 (" Shutdown Decay Heat Removal Requirements")is applicable to St.Lucie Nuclear Plant, but was not addressed directly in the licensee's IPEEE submittal report.The site-specific seismic adequacy evaluation studies performed for St.Lucie-1 and St.Lucie-2 considered a success path that depends on seismic capability nf the auxiliary feedwater (AFW)system;seismic capability of other decay heat removal systems (feed and bleed cooling, and residual heat removal)were n'ot specifically addressed.

The condensate storage tank (CST)was the only component of the AFW system that was actually included in the seismic evaluation; the subminal notes that AFW pumps were previously reviewed Energy Research, Inc.vni ERI/NRC 95-504 for seismic adequacy as part of GL 81-14.In response to an RAI issued by the NRC as part of the USI AW review process, the licensee has indicated that there also exists a seismically qualified path for feed-and-bleed cooling at the plant.As part of the fire IPEEE, the licensee has addressed both Sandia fire risk scoping study issues and USI A45 issues.For both cases, the licensee has dealt with the issues and does not identify any outstanding problem areas.However, the possibility of an earthquake causing a fire was not addressed.

For areas with Thermo-lag, the licensee has checked whether the protection intended by Thermo-lag is necessary to reduce the fire CDF below 10~/ry.For som'e compartments, it has been concluded that even without the presence of Thermo-lag, the CDF can be below 10~/ry.With respect to HFO events, the submittal does not describe any formal analysis of other safety issues.Even though a direct discussion of Generic Issue (GI)-103,"Design for Probable Maximum Precipitation PMP)," was not provided in the submittal, FPL noted that there are no concerns associated with the site wooding levels and roof ponding that could accompany increased (beyond design basis)PMP levels.Some information is also provided in the St.Lucie IPEEE submittal which pertains to generic safety issue (GSI)-147, GSI-148 and GS1-172.Vulnerabilities and Plant Improvements The licensee makes a general conclusion in the IPEEE submittal that there are no vulnerabilities to severe accident risk from external initiators.

However, safety enhancements related to specific external initiators have been identified and proposed for resolution.

For seismic events, the plant-specific seismic adequacy evaluations for St.Lucie Nuclear Plant, Units 1 and 2, have revealed a number of noteworthy seismic findings, including some identified seismic outliers, and have proposed relevant plant improvements as needed.The noted conditions are summarized below: St.Lucie Unit I: During the walkdowns, five anchorages and the bracing of the component cooling water surge tank platform were identified as concerns by the SRT.In addition to these five anchorage concerns, six additional anchorage concerns were identified by FPL for similar components in different equipment trains.Plant improvements were proposed to dispose of these concerns.Three seismic interaction concerns were observed and documented, as were some cases of poor seismic housekeeping.

In response to the NRC's USI A<6 review process, the licensee is implementing a program of strict seismic housekeeping.

~Sr.Lucie Unit 2: Two seismic interaction concerns were observed and documented.

Both of these issues were ultimately evaluated and resolved.A concern was also noted pertaining to whether or not the mounting of some internal coils in an energized transformer was seismically adequate.This concern.was investigated during an outage, and it was found that the mounting was adequate.Energy Research, Inc.ix ERI/NRC 95-504 J~~I 1 r It was also stated in the seismic evaluation that a walkdown of wall transformers needed to be performed, to determine whether or not these transformers would need to be secured.For fire events, even though the licensee has concluded that there are no fire vulnerabilities; nevertheless, it has identified several corrective actions to improve fire safety at the plant.The corrective actions include: An analysis of using the cross-tie between the two units to further increase the availability of power to the affected unit under certain fire scenarios, and a revision of the procedures based on the results of this analysis.Revise the current fire protection procedures to ensure that a roll-up door between non-safety switchgears is kept closed at all times.An analysis of fire scenarios in the"B" switchgear room to reduce the CDF contribution for this area to below 10~/ry, and a revision of the procedures based on the results of this analysis.With respect to HFO events, all potential hazards were dismissed as non-significant risk contributors, without performing a detailed PRA, and no vulnerabilities to severe accidents were identified.

Observations In the seismic IPEEE, the site-specific program for s'eismic adequacy evaluations of St.Lucie Units 1 and 2 addresses only a subset of the elements specified in NUREG-1407 as recommended items that should be considered in the seismic IPEEE of a reduced-scope plant.The evaluations do, nonetheless, address some meaningful IPEEE-related concerns, and have resulted in a small number of plant seismic safety enhancements.

Given the NRC's resolution of related USI AA6 concerns for St.Lucie-l, the following are considered to be the most significant remaining weaknesses of the seismic IPEEE submittal:

1.The SSEL is deficient; r 2.A seismic containment performance assessment was not conducted; 3.The treatment of human actions is deficient; 4.The submittal does not provide adequate documentation of seismic-fire/flood interaction concerns, including component-specific walkdown findings;'.The seismic IPEEE is incomplete with respect to reduced-scope evaluation recommendations found in NUREG-1407; and 6.The seismic IPEEE submittal is not documented in accordance with the format recommended in NUREG-1407, Appendix C.In the fire IPEEE, the licensee has expended a considerable effort in the preparation of the fire analysis, and has presented it in a summary form in its IPEEE submittal.

The IPEEE report complies with the conditions set forth in NUREG-1407.

The licensee has employed a proper methodology and data base for Energy Research, Inc.ERI/NRC 95-504 yl conducting the fire analysis.Based on the data presented, notwithstanding some of the weaknesses of the submittal, it can be concluded that the licensee has conducted a reasonable analysis.The overall results are within the range of conclusions reached in other pressurized water reactor (PWR)fire risk studies.With respect to HFO events, the submittal relies mostly on qualitative reasoning to screen out all such events.In general, the analyses are adequately supported, and follow accepted practice and the overall NUREG-1407 guidelines.

In some cases, however, engineering judgments are made without substantiation.

The most important cases that require additional support are in the area of high winds and tornadoes, particularly with respect to hazard to Unit 1 structures.

Energy Research, Inc.Xi ERI/NRC 95-504 PREFACE The Energy Research, Inc., team members responsible for the present IPEEE review documented herein, include: htnh R.T.Sewell M.Kazarians A;Mosleh M.Khatib-Rahbar, Principal Investigator, Report Review A.S.Kuritzky, IPEEE Review Coordination and Integration R.T.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 staff is acknowledged.

Energy Research, Inc.xn ERI/NRC 95-504 ABBREVIATIONS AFW CCW CDF CE CFR CST DBE DC EPRI ERI FCIA FECR FIVE FPL FRSS FSAR Gl GIP GL GSI HCLPF HFO HVAC ICW IPE'PEEE IRS LLNL LOCA LOSP MCC MFW MLW NRC OL PGA PMH PMP PMS PORV PRA PWR RAB RAI Auxiliary Feedwater Component Cooling Water Core Damage Frequency Combustion Engineering Code of Federal Regulations Condensate Storage Tank Design Basis Earthquake Direct Current Electric Power Research Institute Energy Research, Inc.Fire Compartment Interaction Analysis Florida East Coast Railway Fire Induced Vulnerability Evaluation Method Florida Power and Light Company.Fire Risk Scoping Study Final Safety Analysis Report Generic Issue Generic Implementation Procedure (SQUG)-Generic Letter Generic Safety Issue High Confidence of Low Probability of Failure (Capacity)

High Winds, Floods and Other External Initiators Heating, Ventilation and Air Conditioning Intake Cooling Water Individual Plant Examination Individual Plant Examination of External Events In-Structure Response Spectrum Lawrence Livermore National Laboratory Loss of Coolant Accident Loss of Offsite Power Motor Control Center Main Feedwater Mean Low Water Nuclear Regulatory Commission Operating License Peak Ground Acceleration Probable Maximum Hurricane Probable Maximum Precipitation Probable Maximum Surge Power-Operated Relief Valve Probabilistic Risk Assessment Pressurized Water Reactor Reactor Auxiliary Building Request for Additional Information Energy Research, Inc.xiii ERI/NRC 95-504 RCP RCS RLE SER SI SMA SMM SNL SQUG SRP SRT SSE SSEL SSRAP St.Lucie-1 St.Lucie-2 TER USI Reactor Coolant Pump'Reactor Coolant System Review Level Earthquake Staff Evaluation Report Safety Injection Seismic Margin Assessment Seismic Margin Methodology Sandia National Laboratories Seismic Qualification Utility Group Standard Review Plan Seismic Review Team Safe Shutdown Earthquake Safe Shutdown Equipment List Senior Seismic Review and Advisory Panel St.Lucie Nuclear Plant, Unit I St.Lucie Nuclear Plant, Unit 2 Technical Evaluation Report Unresolved Safety Issue Energy Research, Inc.x]v ERI/NRC 95-504 0 0 4, (5 I~~~1 INTRODUCTION This technical evaluation report (TER)documents the results of the"submittalwnly" review of the individual plant examination of external events (IPEEE)for the St.Lucie Nuclear Plant, Units 1 and 2[1].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 (HFO)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, Florida Power and Light (FPL), 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 (RAIs), evaluation of the licensee responses to these RAIs, and finalization of this TER.The emphasis of this review is on describing the strengths and weaknesses of the IPEEE submittal,'articularly 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 configuration and presents an overview of the licensee's IPEEE process and insights, as well as the review process employed for evaluation of the seismic, fire, and HFO-events sections of the St.Lucie IPEEE.Sections 2.1 to 2.3 of this report present.ERI's findings related to the seismic, fire, and HFO reviews, respectively.

Sections 3.1 to 3.3 summarize ERI's conclusions and recommendations from the seismic, fire, and HFO reviews, 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.

St.Lucie Nuclear Plant is a two-unit nuclear power facility located on Hutchinson Island, about halfway between the towns of Ft.Pierce and Stuart, on the eastern (Atlantic) coast of peninsular Florida.Each of the St.Lucie units is a two-loop Combustion Engineering (CE)pressurized water reactor (PWR), with a rated full-power core thermal output of 2,700 MWt and a net electrical output of 890 MWe.The containment for each unit consists of a steel vessel surrounded by a reinforced-concrete biological shield building;an annular space exists between the steel vessel and the shield building.St.Lucie Unit 1 went into commercial operation during December 1976, and St.Lucie Unit 2 began commercial operation during August 1983.The two units do not physically share any major common areas.Each unit has its own turbine building,~reactor auxiliary building (RAB), and containment building.There is one control room and one cable spreadingroomperunit.

Thesetworoomsaresituatedat theupper elevations oftheRAB.Theauxiliary shutdown panels are located in the"B" switchgear room.From a systems stand point, the two units do not share any major functional-related elements, except for offsite power facilities, a cross-tie for electrical power, and a technical support center which is located next Energy Research, Inc ERI/NRC 95-504 to the Unit 1 control room.Each unit has two diesel generators, an auxiliary feedwater (AFW)system, two motordriven main feedwater (MFW)pumps, high and low pressure safety injection (SI)systems, a component cooling water (CCW)system, and an intake cooling water (ICW)system.Both CCW and ICW systems are needed to prevent reactor coolant pump (RCP)seal failure.The AFW system includes two motordriven pumps, as well.as one steam4riven pump that needs direct current (DC)power to start, and can be operated in a manual mode.The design basis earthquake (DBE)peak gro'und acceleration (PGA)for St.Lucie Nuclear Plant is 0.1g (horizontal and vertical)for the safe shutdown earthquake (SSE).The DBE spectral shapes are different for the two units;Unit 1 was designed for a Housner spectral shape, and Unit 2 was designed for a Regulatory Guide (R.G.)1.60 spectral shape.Category I structures for both units are founded on Category-I fill, underlain by cemented sands and sandy limestones.

For the IPEEE study, a cutoff date of December 1993 was used for establishing plant configuration and operating conditions.

1.2.1 Seismic As documented in NUREG-1407, for seismic IPEEE purposes, St.Lucie is binned into the reduced-scope evaluation category.Rather than implementing a reduced-scope seismic evaluation, FPL has pursued the use of a site-specific program for conducting the seismic IPEEE of St.Lucie Nuclear Plant.This site-specific program was developed primarily for treatment of Unresolved Safety Issue (USI)A%6, and represents a"scaled-back" approach to achieving the objectives of GL 8742[4].The justifications cited by FPL for performing a scaled-back analysis include: (a)very low probability of having an earthquake at the SSE level at FPL's plants;and (b)very low values of potential offsite releases and potential risk reductions given the postulated accident scenarios and seismic hazards.FPL's scaled-back site-specific seismic adequacy program was approved, in concept, by the NRC for the purpose of addressing USI A46.However, once FPL submitted the actual seismic adequacy evaluation study[5], the NRC identified a number of concerns and potential deficiencies with the approach.The NRC's concerns are documented in its staff evaluation report (SER)pertaining to USI A-46 resolution

[6].A site investigation by the NRC was held at FPL's corporate headquarters and at the Turkey Point Nuclear Plant during the week of December 4-8, 1995 to help resolve the concerns noted in the NRC's SER.Many of the NRC concerns were alleviated by way of discussions with the licensee and its consultants; for other concerns, the licensee has agreed to implement corrective actions identified by the NRC.These items are documented in an NRC supplemental safety evaluation report (SSER)[7], wherein the NRC states that closure has been reached on all of the SER open items for both Turkey Point and St.Lucie.With respect to the seismic IPEEE, the NRC had concerns with the use of the FPL site-specific approach as a basis for resolving severe accident vulnerability issues.The NRC never gave its approval of FPL's program for treatment of the seismic IPEEE.Nonetheless, FPL proceeded with use of the site-specific seismic adequacy evaluations for USI A<6 as the basis for conducting the seismic IPEEE.Since the licensee's seismic IPEEE is essentially identical to its USI AA6 seismic adequacy evaluation study, and because many of the recommendations outlined in NUREG-1407 for a reduced-scope IPEEE Energy Research, Inc.ERI/NRC 95-504 l j 1 are achieved if an acceptable USI AA6 evaluation has been performed, the NRC's SER and SSER determines (to a significant degree)that a corresponding review conclusion be made for similar IPEEE concerns.Hence, this TER indicates where a review finding has'been based on NRC's safety evaluation for USI A46.It is important to point out that only St.Lucie-1 is a USI A46 plant.The design basis and seismic qualification employed for St.Lucie-2 are similar to current NRC licensing requirements.

Therefore, a seismic adequacy evaluation was not required for St.Lucie-2 as part of USI A46 resolution.

Still, FPL undertook such an evaluation for addressing the seismic IPEEE.FPL's approach to seismic evaluation relies primarily on plant walkdowns and on the use of seismic review team (SRT)judgment, supplemented with calculations, as needed, for resolving outliers.The walkdowns have addressed the following items: equipment seismic capacity versus demand, equipment construction adequacy, anchorage adequacy, seismic spatial interaction concerns, and seismic housekeeping concerns.The main overall elements of FPL's site-specific seismic adequacy evaluation include: Project planning Selection of the seismic review team Preparatory work prior to walkdown Seismic capability walkdowns Limited seismic margin assessment (SMA)calculation work Resolution of outliers Peer review Documentation FPL's approach to these aspects of the seismic IPEEE process for St.Lucie Nuclear Plant is discussed in Section 2.1.FPL found no seismic vulnerabilities to potential severe accidents, but did report a small number of outliers to be resolved.Additionally, in response to the NRC's USI A46 review process, FPL agreed to resolve an additional concern related to seismic housekeeping procedures.

1.2.2 Fire Overall, the licensee has concluded that there are no significant fire vulnerabilities at St.Lucie.With the exception of the control room, cable spreading room and the"B" Switchgear room, all fire zones and areas were screened out based on a 10 per reactor-year (ry)core damage frequency (CDF)criterion.

The licensee cites several conservative assumptions in fire occurrence rate and fire severity for the control room and cable spreading room, and concludes that these two areas do not pose a vulnerability.

Fire propagation modeling has been performed for the"B" switchgear room to verify that this area does not need to be considered as a vulnerability.

The licensee has addressed the Sandia fire risk scoping study (FRSS)issues and USI A<5 concerns.For both cases, the licensee has dealt with the issues and did not identify any outstanding problem areas.However, the possibility of an earthquake leading to a fire was not addressed.

Energy Research, Inc.ERI/NRC 95-504 For areas with Thermo-lag, the licensee has checked whether the protection intended by Thermo-lag is necessary to reduce the fire CDF to below 10~/ry.For some compartments, it has been concluded that even without the presence of Thermo-lag, the CDF is below 10/ry.1.2.3 HFO Events The general methodology utilized in the study conforms to that presented in NUREG-1407 for the analysis of other external events.It consists of the following steps: 1.Review of plant-specific hazard data and licensing bases.2.Determination of conformance of the plant risk significant structures to the 1975 Standard Review Plan (SRP)[8]criteria.3.Screening of plant structures that'meet the SRP criteria for a specific hazard.4.Determination of the hazard frequency for those structures that do not meet the SRP criteria.5.Performing a bounding analysis if the hazard frequency calculated in Step 4 is found to be high.6.Perform a probabilistic risk assessment (PRA), if necessary.

The licensee has performed a detailed analysis of the high winds, external flooding, and transportation and nearby facility accident hazards.Additionally, the potential for external events has also been evaluated to ensure the that there are no hazards unique to the plant.The objectives for this assessment are consistent with those of GL 88-20, Supplement 4[2].FPL personnel have been directly involved in all aspects of the development, quantification, and documentation of the analysis.The major finding of the analysis is that there are no vulnerabilities to severe accident risk from HFO events.In its qualitative review of the St.Lucie 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.4;its strengths and weaknesses with respect to the staff-the-art; and the robustness of its conclusions.

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 included the following major activities:

Completely examine the IPEEE submittal and related documents Develop a preliminary TER and RAIs Examine responses to the RAIs Finalize this 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.Energy Research, Inc.ERI/NRC 95-504 4g 0 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 have indeed been implemented at St.Lucie.1.3.1 Seismic In conducting the seismic review, ERI generally followed the emphasis and guidelines described in the report, Individual Plant Examinanon of External Events: Review Guidance[9], for review of a seismic margin assessment, and the guidance provided in the NRC report, IPEEE Step I Review Guidance Document[10].In addition, on the basis of the St.Lucie IPEEE submittal, ERI completed data entry tables developed in the Lawrence Livermore National Laboratory (LLNL)document entitled"IPEEE Database Data Entry Sheet Package"[11].In its St.Lucie IPEEE seismic review, ERI examined the following documents:

Sections 1, 2, 3, 4.8, 6, 7, and 8 of the IPEEE submittal[1]~The USI A<6 seismic adequacy evaluation of St.Lucie Unit 1[5]~The seismic adequacy evaluation study of St.Lucie Unit 2[12]~Section 3.7.7 of the individual plant examination (IPE)submittal for St.Lucie Units 1 and 2[13]~The NRC's SER[6]and supplemental SER (SSER)[7]of the USI A-46 submittals for Turkey Point, Units 3 and 4, and St.Lucie Unit 1~The licensee's response[14]to the RAIs generated as part of the initial submittal review The IPEEE submittal[1]itself contains only one page of discussion related to seismic evaluation.

Consideration of the seismic adequacy evaluation studies (References

[5,12])and the NRC's evaluation

[6,7]of the licensee's USI A-46 submittal constituted the most significant element of the present seismic review.The checklist of items identified in Reference[9]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 walkdown effort sufficient to accomplish the objectives of the seismic IPEEE?Was the development of success paths performed in a manner consistent to 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?

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

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

Energy Research, Inc.ERI/NRC 95-504

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

It is important to note that, in a number of instances, IPEEE review findings have been reported on the basis of consistency with related findings in NRC's SER[6]and SSER[7]for USI A46, rather than on the basis of a separate review for IPEEE.1.3.2 Fire During this technical evaluation, ERI reviewed the fire-events portion of the IPEEE for completeness and consistency with past experience.

This review was based on consideration of Sections 1, 2, 4, 6, 7 and 8 of Reference[1], and Section II of Reference[14].In addition, a set of layout drawings[15,16]pertaining to fire protection were available for review.The guidance provided in References

[9,10]was used to formulate the review process and organization of this document.The data entry sheets used in Section 5 hav'e been completed in accordance with Reference fl 1].The process implemented for ERI's review of the fire IPEEE included an examination of the licensee's methodology, data, and results.ERI reviewed the methodology for consistency'with currently accepted and state-of-the-art methods.The data element of a fire IPEEE includes, among others, such items as: Cable routing Fire zone/area partitioning Fire occurrence frequencies Event sequences Fire detection and suppression capabilities For a few fire zones/areas that were deemed important, ERI also verified the logical development of the screening justifications/arguments (especially in the case of fire-zone screening) and the computations for fire occurrence frequencies and CDFs.Rather than perform a completely independent set of calculations, however, the review team used its experience and comparisons of other plants and fire evaluation results, in order to judge the accuracy and completeness of the information provided by the licensee.Special attention was directed to: (1)the screening methodology, because a trend to prematurely screen out potentially significant areas or to inadequately justify screening out an area, has emerged as a common problem among past fire PRAs and IPEEE analyses;and (2)the licensee's assumptions, because the results'f many studies are unduly influenced by assumptions made to simplify or introduce conservatisms.

1.3.3 HFO Events The review process for HFO events closely followed the guidance provided in the report entitled IPEEE Step I Review Guidance Document[10].This process involved examinations of the methodology, the data used, and the results and conclusions derived in the submittal.

Sections 1, 2, 5, 6, 7 and 8 of the IPEEE submittal[1], and licensee responses to RAIs[14], were examined in this HFO-events review.The 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 HFO events,'and on confirming that any analysis of SRP conformance was appropriately executed.In addition, the validity of the licensee's conclusions, in consideration of the results reported in the IPEEE submittal, was assessed.Also, bounding-analysis and PRA results pertaining to frequencies Energy Research, Inc.ERI/NRC 95-504 P r' of occurrence of hazards and estimates of conditional probabilities of failure, were checked for reasonableness.

Review team experience was relied upon to assess the validity of the licensee's evaluation.

Energy Research, Inc.ERI/NRC 95-504 2 CONTRACTOR REVIEW FINDINGS 2.1 5gismz A summary of the licensee's seismic IPEEE process has been described in Section 1.2.Here, 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 Caregory and Revie-Level Eanhquake (RLE)St.Lucie Nuclear Plant is located in an area of low seismicity, on the eastern coast of peninsular Florida.Each of the two St.Lucie units is a two-loop Combustion Engineering PWR.St.Lucie Unit 1 went into commercial operation during December 1976, and is in the seismic qualification utility group (SQUG)/USI A46 program;St.Lucie Unit 2 began commercial operation during August 1983, and is not in the USI AP6 program.The design basis earthquake (DBE)peak ground acceleration (PGA)for St.Lucie Nuclear Plant is O.lg (horizontal and vertical)for the safe shutdown earthquake (SSE).The DBE spectral shapes are different for the two units;Unit 1 was designed for a Housner spectral shape, and Unit 2 was designed for a , Regulatory Guide (R.G.)1.60 spectral shape.Category I structures for both units are founded on Category-I fill, underlain by cemented sands and sandy limestones.

Due to the low seismic hazard at the site, St.Lucie has been designated as a reduced-scope plant in NUREG-1407.

The RLE is equivalent to the SSE.b.Seismic IPEEE Process The licensee has implemented a site-specific seismic adequacy evaluation program based on a methodology it has compiled for executing its USI A-46 resolution program at Turkey Point Units 3 and 4, and at St.Lucie Unit 1.(The NRC has determined, pending apprcpriate follow-up action by the licensee, that USI AM has been adequately resolved for St.Lucie-1 and Turkey Point[6,7].)The licensee claims that its process conforms with the Optional Methodology of Paragraph 3.3 in NUREG-1407.

However, the program was never actually approved by the NRC.c.Review Findings The IPEEE process is not fully consistent with the recommended guidelines of NUREG-1407 for St.Lucie.FPL's seismic programs for St.Lucie Units 1 and 2 address only a portion of the seismic IPEEE elements/concerns for a reduced-scope plant.The IPEEE submittal for St.Lucie Unit 1 is essentially identical to the USI AM submittal.

Hence, the concerns/findings documented by the NRC for USI A<6 are applicable to a number of the aspects of the seismic IPEEE.Nonetheless, the fact that FPL's seismic adequacy evaluation program departs from a complete reduced-scope assessment is viewed to be a significant weakness.The overall seismic IPEEE methodology Energy Research, Inc.ERI/NRC 95-504 j r 4 employed by FPL has only a limited potential to achieve IPEEE objectives, and to assess severe accident vulnerabilities at St.Lucie Nuclear Plant.2.1.2 Success Paths and Component List Success was defined, for purposes of identifying a success path, as the ability to achieve and maintain a hot shutdown condition for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.Loss of offsite power was assumed in choosing the success path.In addition, a design basis earthquake was assumed not to trip the reactor.The primary elements of the chosen success path include: supervisory and control function requirements, requirements of decay heat removal via the AFW system, emergency electrical power requirements, chemical and volume control requirements, and equipment cooling (ultimate heat sink)requirements via the CCW and ICW systems.The submittal states that all active equipment pertaining to the success path were identified in developing a safe shutdown equipment list (SSEL).Some passive components, such as tanks and heat exchangers, were also included in the SSEL.A significant number of components (e.g., AFW pumps)were removed from the SSEL because they had been previously reviewed for seismic adequacy in another program.(Similarly, potential interaction concerns that involved block walls were considered resolved if the walls were previously addressed under IE 80-11[17]).The resulting SSEL defines the set of components considered in plant walkdowns.

FPL's seismic adequacy evaluation does not clearly identify the chosen success path, nor does it present a success path logic diagram.Only one success path was involved in developing the SSEL, and only a limited set of components were identified for each major success-path function.The study did not explicitly address a small-break loss of coolant accident (LOCA)in the development of the success path and SSEL.The SSEL considers active components and a partial list of passive components.

The success criterion used in the FPL study is the ability to achieve and maintain hot shutdown for a time period of only 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, rather than the recommended 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.However, in response to RAIs raised by the NRC in its USI AM review process, FPL indicated that the plant has multiple (albeit non-seismically qualified) water sources that could provide cooling for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.In addition, FPL indicated that the plant has the (seismically qualified) capability o~indefinitely long feed-and-bleed cooling.Thus, the equipment list developed in the FPL study appears to be considerably limited, and considers only a subset of components that should be evaluated in a reduced-scope assessment.

2.1.3 Non-Seismic Failures and Human Actions a.Overall Approach 4 4 The seismic adequacy studies note that a review of operating procedures was performed for St.Lucie Nuclear Plant to verify the equipment list and to identify any equipment which might be required to bring the reactor from 100%power to hot shutdown.Additionally, operating procedures to shut down the reactor, take the reactor to hot shutdown, to respond to reactor trip, and to respond to los-of'offsite power were reviewed.No mention is made of specific non-seismic failures or human actions that might limit the capability of the chosen success path.Energy Research, Inc.ERI/NRC 95-504 ly b.Screening Crireria Random and operator failure rates were not reported;no screening criteria were applied with respect to non-seismic failures and human actions.c.Review Findings According to NUREG-1407, candidate success paths should be screened to insure that impacts of non-seismic failures and human actions will not be controlling factors inhibiting the likelihood of successful hot shutdown.FPL's seismic evaluation has not identified the specific random failures and human actions which might compromise the integrity of the chosen success path.Hence, the licensee's study is inadequate in its treatment of non-seismic failures and human actions, which is thus viewed to be a weakness of the study.2.1.4 Seismic Input Seismic inputs for evaluation studies of St.Lucie, Units 1 and 2, were defined by SSE spectra and other plant-specific design-basis commitments in the final safety analysis report (FSAR).For St.Lucie-l, the SSE is identified by a Housner spectral shape anchored to a PGA level of O.lg.For St.Lucie-2, the SSE is a R.G.1.60[18)shape'anchored to the same PGA value.NUREG-1407 indicates that the SSE ground response spectra should be used to define input to structures, and for computing in-structure response spectra.FPL's seismic adequacy evaluation program uses the SSE spectrum or FSAR in-structure spectra as the basis for defining seismic input for components.

Hence, the licensee's definition and use of seismic input is consistent with the guidelines of NUREG-1407 for a reduced-scope plant.2.1.5 Structural Responses and Component Demands St.Lucie Unit 1 had existing floor response spectra curves for the containment structure and the auxiliary building;St.Lucie Unit 2 had existing floor'esponse spectra curves for all safety-related buildings and structures.

Building models used to generate floor response spectra included translational and rotational springs to account for soil-structure interaction effects.The floor response spectra were used to define demands for many of the SSEL components.

For components where existing floor response spectra were not available for assessing demands (which was the case only for St.Lucie Unit 1), estimates of component demands were made based directly on the SSE spectrum.The approach for assessing such demands (for equipment less than 40 feet above grade)was to: (a)take the peak spectral acceleration from the 5%damped SSE spectrum, (b)multiply this peak value by 1.5 to account for building amplification, and (c)multiply again by a factor of 1.25 for conservatism.

NUREG-1407 indicates that existing FSAR in-structure spectra, based on SSE input and FSAR licensing criteria, may be used for evaluating component demands.In the FPL seismic adequacy studies, FSAR in-structure spectra were used, when available, to establish equipment demands.When in-structure spectra were not available (St.Lucie Unit 1 only), a generally conservative procedure based on scaling the peak SSE spectral acceleration was used to define component demands.The licensee's development of component demands thus appears consistent, to a significant degree, with the guidelines of NUREG-1407 Energy Research, Inc.10 ERI/NRC 95-504

'r 0 for a reduced-scope plant.Additionally, the NRC has accepted this aspect of the licensee's analysis for USI A46 resolution

[7].2.1.6 Screening Criteria Screening for the St.Lucie seismic evaluation studies has not followed the formal procedures described in the generic implementation procedure (GIP)[19]or in Electric Power Research Institute (EPRI)NP-6041[20], as recommended in NUREG-1407.

Rather, the procedures described in Reference[21], the Senior Seismic Review and Advisory Panel (SSRAP)document, have generally been implemented.

Whether GIP, EPRI NP-6041, or other procedures are used for screening, screening caveats must be observed, anchorage capacity checks must be performed, and spatial interaction issues must be appropriately assessed.Additionally, in any screening procedure, SRT judgment plays the major role in component evaluations.

FPL's screening approach has been based primarily on SRT judgment, on comparisons of estimated anchorage capacity versus SSE-consistent demand, and on insights derived by the SSRAP.Although the licensee's approach to screening does not conform precisely to the recommendations of NUREG-1407 for a reduced-scope plant, it is judged to be a reasonable process that substantially achieves the significant intent of component screening.

2.1.7 Plant Walkdown Process a.Preparatory 8'ork A pre-walkdown of the plant was performed to help the seismic review team (SRT)members identify what information and assistance would be needed during the seismic capability walkdown.FPL engineers gathered generic and equipment-specific documentation as deemed necessary by the SRT.In addition, FPL staff familiar with plant systems developed the list of equipment to be walked down.b.Seismic Capability Walkdown Plant walkdowns were conducted by an SRT consisting of three highly experienced walkdown experts.The seismic adequacy evaluation studies have relied heavily on the judgment of these engineers.

During the walkdown, FPL provided staff engineers to help support the SRT members, primarily in obtaining additional plant information that was needed on a case-by-case basis.The actual duration of seismic walkdowns is not mentioned in the documentation.

Four considerations were addressed in the plant walkdown screening effort: (1)equipment seismic capacity versus demand, (2)construction adequacy of equipment, (3)anchorage adequacy, and (4)seismic spatial interaction concerns.The walkdown also made note of concerns related to: (5)general seismic"housekeeping" issues.Each of these aspects.of plant walkdowns and component screening is described briefly below.-This screening item pertains to identification of seismic adequacy problems that could be inherent to specific types of unqualified seismic equipment.

These encompass the types of problems that would be found in a qualification test, Energy Research, Inc.ERI/NRC 95-504 including:

functional problems, internally fragile elements, and inadequate overall structural resistance of a cabinet.The St.Lucie seismic adequacy evaluations treated this item in a generic way based on findings of the SSRAP, as documented in Reference[21].It was demonstrated in the evaluation studies that (for use with respect to equipment having a natural frequency greater than 8 Hz and located less than 40 feet above grade)the SSRAP bounding spectrum envelopes plant SSE spectra over the entire frequency range.It was also demonstrated that (for use with respect to equipment having a natural frequency less than 8 Hz or located more than 40 feet above grade)the SSRAP bounding spectrum multiplied by 1.5 enveloped plant floor response spectra.Since the bounding spectrum represents an experience-based seismic ruggedness threshold for unqualified nuclear power plant equipment, the FPL study concludes that seismic capacity versus demand was judged acceptable for all plant components.

The plant walkdowns, therefore, did not give much attention to this screening item, on a component-by-component basis.-This screening item pertains to identification of seismic adequacy problems that could be attributed to the configuration or manner of construction/installation of the equipment at the plant.Generally speaking, the as-built configuration of equipment can be considered adequate, provided that certain caveats have beenconsidered and satisfied.

FPL reasoned that, due to low seismicity at FPL plant sites, specific caveats did not need to be addressed for each type of equipment.

The seismic evaluation study further noted that SRT members are experts in the area of seismic adequacy of equipment, and that they noted any equipment-specific details that they felt were seismically vulnerable.

I'l III I I I I I'I that are due to non-existent or weak anchorage.

The constructed anchorage configuration can be considered as a caveat to be considered in the evaluation of all components.

It is a special caveat, however, because its treatment usually requires more than just a visual inspection; the expected demand on the anchorage and a numerical estimate of anchorage capacity are often needed to satisfy anchorage caveats.In the seismic adequacy evaluations, SRT judgment was used to screen out"obviously rugged" anchorages.

Otherwise, a numerical estimate of seismic adequacy of anchorage components was obtained and compared against component anchorage demand.Any problems noted with anchorage capacity were designated as potential outliers to be resolved.4.-This screening item pertains to the identification of physical effects that could independently compromise the performance of an otherwise well-installed seismically adequate component.

Such physical effects include: objects impacting equipment in any manner, conduit pull-out due to inadequate flexibility of lines attached to equipment, block wall collapses, etc.During the walkdowns, SRT members looked for, and made note of (on walkdown work sheets), any potential seismic spatial interaction concerns;identified concerns were designated as potential outliers.Energy Research, Inc.12 ERI/NRC 95-504

~g P I/

5,-This walkdown item pertains to situations that, although not leading to failure of an important safety-related component, can exacerbate problems and/or inhibit operator effectiveness following an earthquake.

Any instances of poor seismic housekeeping observed by SRT members were noted and reported to FPL.Among these five walkdown items, primary consideration was given to assessing anchorage adequacy and to identifying seismic spatial interactions.

c.Review Findings NUREG-1407 recommends the use of GIP or EPRI NP-6041 walkdown procedures.

The St.Lucie walkdown has implemented procedures substantially similar to these, perhaps allowing for somewhat greater latitude in the use of expert judgment.Due in large part to the exceptional qualifications of the SRT members, and the NRC's acceptance of the seismic walkdown for USI A<6 resolution, the licensee's walkdown process is'considered to be adequate in identifying outliers among those components that have been included in the scope of walkdowns.

2.1.8 Evaluation

of Outliers a.Overall Approach The seismic adequacy evaluations do not make a clear distinction between"outlier" and"potential outlier." All items not screened out by the SRT were addressed in some manner by FPL.For potential anchorage outliers (i.e., those anchorage concerns screened in by the SRT during plant walkdowns), more-detailed calculations were performed to better determine seismic adequacy.Any component having inadequate/low anchorage capacity was identified as an outlier requiring resolution by FPL.b.High Conjfdence of Low Probabiliry of Failure (HCLPF)Calcularions For St.Lucie Unit 2, HCLPF calculations were performed for many large, flat-bottom tanks.No HCLPF calcu]ations were performed for block walls identified to be a potential interaction problem.(The seismic adequacy evaluations rely on earlier IE 80-11 calculations.)

c.Review Findings For some components that were screened-in at St.Lucie Nuclear Plant, capacity calculations were performed to demonstrate whether or not the component met the FSAR demand (or the conservative approximation to the FSAR demand).For components identified as final outliers, however, the outlier assessment was often readily made (without calculation) due to an obviously deficient condition (e.g., seal weld on one side of 480V Motor Control Center (MCC)1A6, missing welds on MCC portion of 480V Load Center 1A3).For each final outlier noted, FPL proposed a corrective measure and submitted non-conformance resolution documentation.

ERI/NRC 95-504 The licensee's walkdown process is judged to be adequate in identifying outliers among those (limited set of)components that have been included in the scope of walkdowns.

Energy Research, Inc.13

2.1.9 Relay Chatter Evaluation NUREG-1407 indicates that completion of the USI A46 review requirements for relay chatter evaluation will satisfy the IPEEE intent for reduced-scope plants that are also USI A-46 plants.For reduced-scope plants that are not also USI A-46 plants, no relay chatter evaluation is necessary.

The licensee's IPEEE submittal does not mention a relay chatter evaluation for St.Lucie-1~However, during NRC's USI A@6 review for Turkey Point and St.Lucie-l, it was revealed that FPL had assessed bad actor relays, verified mountings of relays, and demonstrated that there were no deleterious effects of chatter of bad actor relays.The NRC accepted the licensee's relay evaluation for USI A-46 resolution, and hence, the NUREG-1407 recommendation for the seismic IPEEE is satisfied for St.Lucie-1.NUREG-1407 does not request a relay evaluation for St.Lucie-2, a non-USI A-46 plant.2.1.10 Soil Failure Analysis NUREG-1407 states that no evaluation of soil failures is required for a reduced-scope plant.Correspondingly, the licensee has not performed such an analysis.2.1.11 Containment Performance Analysis For reduced-scope plants, NUREG-1407 requests that performance of containment and containment systems should be addressed.

Components necessary to achieve successful accident mitigation need to be included in the scope of seismic walkdowns and outlier evaluation.

FPL did not include the containment structures or containment systems in its seismic adequacy evaluations of St.Lucie, Units 1 and 2.Hence, the licensee's seismic adequacy studies of St.Lucie Unit I and St.Lucie Unit 2 are not responsive to the NUREG-1407 request for a containment performance analysis.2.1.12 Seismic-Fire Interaction and Seismically Induced Flood Evaluations a.Evaluarion of Seismic-Fire Inreracrions Section 4.8 of the IPEEE submittal report discusses seismic-fire interactions.

The topic of seismic-fire interactions is one element of the Sandia fire risk scoping study (FRSS)issues.The IPEEE submittal states that all Sandia FRSS issues are more than adequately covered through the St.Lucie Fire Protection Program.In terms of details of the seismic-fire evaluation, however, the submittal indicates only that: "Essentially, the II/I criteria was applied to fire systems whose failure could affect operation of safety-related systems." Section 2.2.12 provides addition comments on FPL's seismic-fire evaluation for St.Lucie.No specific discussions of seismically induced fires, of seismic inadvertent actuation of fire suppression systems, or of seismically induced failure of fire suppression systems were provided in the submittal.

Seismic-Fire Walkdown The submittal does not indicate that a seismic-fire walkdown evaluation was conducted.

Energy Research, Inc.14 ERI/NRC 95-504 c.Seismically Induced Flood Evatuarion No documentation pertaining to evaluation of seismically induced floods was submitted.

d.Review Findings The St.Lucie seismic adequacy evaluation has not fully addressed seismic-fire interactions or seismically induced floods.2.1.13 Treatment of USI A45 A reduced-scope seismic assessment should consider the seismic capability of components necessary for successful decay heat removal, in response to USI A<5 (Decay Heat Removal Requirements).

FPL's seismic IPEEE submittal and seismic adequacy evaluation studies for St.Lucie did not directly document findings for any Generic Issues (GIs)lUSIs other than USI A-46 (for St.Lucie Unit 1).Indirectly, USI AQS was addressed owing to the fact that the success path needed to accomplish one method of decay heat removal (i.e., via the AFW system).However, the AFW pumps were eliminated from the seismic evaluation (because they had been previously examined for seismic adequacy elsewhere), and only the condensate storage tank (CST)was identified as a necessary component in the SSEL.The licensee's seismic adequacy study does not address a meaningful scope of components related to decay heat removal functions.

This weakness stems from the fact (noted in Section 2.1.2 of this TER)that the SSEL is only partially complete.2.1~14 Peer Review Process An independent external peer review was conducted by Dr.Paul Smith for the seismic adequacy evaluation studies of St.Lucie-1 and St.Lucie-2.This peer review identified five additional seismic concerns.FPL engineers also reviewed the seismic studies.A meaningful peer review appears to have been conducted for the limited-scope seismic evaluation studies of St.Lucie-1 and St.Lucie-2.2.1.15 Summary Evaluation of Key Insights Only a subset of components needed to ensure successful shutdown are considered in FPL's equipment list, and hence, the seismic IPEEE process has only a limited potential to reveal vulnerabilities or outliers.However, for those components that have been included in the scope of FPL's seismic adequacy evaluation studies, the process implemented for screening outliers, and for addressing their resolution, is considered to be appropriate and adequate.FPL's seismic adequacy evaluation study has identified a number of outliers (primarily relating to weak anchorage), and has proposed relevant modifications to enhance safety.The NRC has already reviewed these outliers and modifications for St.Lucie Unit 1, as part of USI A-46 resolution.

Additionally, the NRC has conducted a site investigation to identify any vulnerabilities that may require further Energy Research, Inc.ERI/NRC 95-504 analysis/treatment.

As a result, FPL is performing follow-up actions to implement a strict housekeeping program at St.Lucie.No outliers reported by the licensee appear to require further analysis for seismic IPEEE purposes.However, additional outliers may have well been found if the'licensee had expanded the scope of its seismic adequacy evaluations to address IPEEEwnly components and issues.Furthermore, the licensee elected not to conduct a containment performance analysis at St.Lucie Nuclear Plant.Thus, no vulnerabilities affecting containment performance, related to seismic behavior of containment systems (e.g., containment cooling, containment isolation, etc.), nor pertaining to direct seismic failure of the containment structures themselves, were identified.

The St.Lucie Nuclear Plant seismic adequacy evaluation studies are capable of finding only a limited set of seismic-related, severe accident vulnerabilities.

2.2 Bzg A summary of the licensee's fire IPEEE process has been described in Section 1.2.Here, the licensee's fire evaluation is described in detail, and discussion is provided regarding significant observations encou'ntered in the present review.2.2.1 Overview and Relevance of the Fire IPEEE Process a.Method Selected for Fire IPEEE The fire analysis was performed per fire-induced vulnerability evaluation (FIVE)methodology (Reference

[22])in two phases.The first phase was a screening step based primarily on contents of a fire zone or area.In the second phase, the frequency of core damage from a fire in a specific fire zone was estimated using the formulations and data provided in the FIVE methodology.

b.Key Assumptions Used in Performing Fire IPEEE The IPEEE does not provide a separate list of assumptions.

However, the present review has identified the following assumptions which could have a significant impact on the final results: Fire barriers/boundaries were taken to be as good as rated.No discussions are provided as to whether active systems (for example a self closing/normally open fire door)are part of fire barrier definition.

No consideration is given to the possibility of open doors, open ducts, failure of fire dampers, etc.This results in cross-zone fires being judged to have negligible risk.Thermo-lag was assumed to be effective for a select group of fire areas.-.All fires in areas containing safe shutdown equipment were assumed to lead to reactor trip.Cont.inment fires were not analyzed explicitly.

This approach was based on the'observation that most containment fires are related to RCP oil fires, which have been minimized with the improve-ments in the oil collection system.Energy Research;Inc.16 ERI/NRC 95-504

~All remotely operated valves with hand-wheels were assumed to be available for manual operation.

c.Status of Appendix R Modiji canons Appendix R modifications were assumed to be completed.

d.New or Existing PRA The IPEEE is a new study.It uses the results of the already completed IPE[13]and PRA[23]for St.Lucie.2.2.2 Review of Plant Information and Walkdown Walkdown Team Composition Different types of walkdowns have been conducted.

Thermo-lag evaluation walkdowns provided information for those compartments where this insulating material was present.For other areas, walkdowns have been conducted specifically for IPEEE fire analysis.In all cases, the walkdown teams have included FPL engineers, fire protection specialists, and personnel from a consulting firm.The following areas have been reviewed: "A" and"B" safety-related switchgear"A" and"B" electrical penetration rooms The cable spreading rooms The"A" cable loft area Reactor auxiliary building basement hallway and hall areas Heating, ventilation, and air conditioning (HVAC)equipment area Intake cooling pump areas Turbine buildings and related areas Auxiliary feedwater pump areas Inverter room Both units have been visited in the walkdowns.

References

[1]and[14]do not provide any details on how the observations of the walkdown have been recorded, and they do not provide the format or a sample of the records.The licensee cites other inspections and the general familiarity of the IPEEE fire analysis team with the plant as additional basis for the IPEEE fire analysis.However, it is not clear whether the information gathered from these inspections was indeed used in the IPEEE fire analysis.b.Significant Walkdown Findings The IPEEE does not indicate that the walkdown team discovered any new fire vulnerability from the plant visit.From the IPEEE, it can be inferred that the walkdown was used mainly to verify equipment and cable locations, to measure area dimensions, and to confirm combustible loadings.c.Signi/cant Plant Features The following is a list of plant features that are deemed to be important:

Energy Research, Inc.17 ERI/NRC 95-504

1.The main feedwater pumps are electric motor driven.2.Loss of the component cooling water (CCW)system can lead to reactor coolant pump (RCP)seal failure.3.The steam driven AFW pump can be operated manually.4.Thermo-lag is used for separation of redundant trains in some areas of the plant.The IPEEE fire analysis has taken credit for the effectiveness of this material only in a select group of fire areas.2.2.3 Fire-Induced Initiating Events a.Were Iniriaring Events Other than Reactor Trip Considered?

Reactor trip, loss of offsite power, and the possibility of a LOCA via reactor coolant pump seal failure have been considered in the IPEEE.b.Were the Initt'aring Events Analyzed Properly?From the information provided by the licensee[1,14], it can be inferred that, for loss of offsite power and reactor coolant pump seal failure, a thorough analysis has been conducted.

However, it is not clear whether the possibility of hot short failures in control cables, and inadvertent opening of the isolation valves of reactor coolant system (RCS)high and low pressure interfaces, have been considered.

For example, the possibility of a power-operated relief valve (PORV)opening inadvertently has not been addressed explicitly.

2.2.4 Screening of Fire Zones a.Was a Proper Screening Methodology Employed?Screening was properly conducted.

Screening was performed per FIVE methodology.

A list of all fire zones is provided in the IPEEE submittal, along with the specific screening criteria and assumptions.

b.Have the Cable Spreading Room and Control Room Been Screened Out?The cable spreading rooms and control rooms of both units have been included in a detailed analysis, and have not been screened out.C.Were There Any Fire Zones/Areas that Have Been Improperly Screened Out?The IPEEE submittal does not provide sufficient information for proper evaluation of the screening effort.From a general stand point, and when compared with other PWR plants that are in compliance with Appendix R requirements, the results seem to be reasonable.

That is, the fire frequencies for each area, and failure probabilities of the redundant trains, appear to be reasonable.

Manual actions may need to be undertaken to ensure availability of a redundant train.There are no indications as to whether the effects of a specific fire on these actions have been considered in the analysis.Energy Research, Inc.18 ERI/NRC 95-504 2.2.5 Fire Hazard Analysis The IPEEE has used the fire occurrence data provided in Reference[23].The exact approach (i.e., choice of weighting factors, ignition sources, etc.)is not specified.

A plant-specific data base has not been used.Per Reference[14], St.Lucie Nuclear Plant has experienced a lower number of fire events than for the average plant.Therefore, this omission can be considered as, conservative.

2.2.6 Fire Growth and Propagation Fire growth and propagation analysis was accomplished via the formulations provided in FIVE.,The submittal provides a separate appendix where the formulations are described.

The most risk-significant area where these methods have been used is the"B" switchgear room.Reference[14]provides a brief description of this area, but does not provide any information regarding the size and exact location of the transient fuels that have been considered.

The submittal does not provide any details of how the specific fire propagation scenarios were developed, or how transient or other fuels have been positioned within the fire areas.a.Treatment of Cross-Zone Fire Spread and Associated Major Assumprions As part of Phase-I screening, the fire compartment interaction analysis (FCIA)methodology of FIVE has been used.Compartments have been combined to form a larger fire zone that takes into account the possibility of fire spread among compartments.

However, possibility of fire spread through normally open active fire barriers (e.g., roll-up doors and fire dampers)has not been addressed explicitly.

Given the discussions provided in References

[1]and[14], it may be concluded.

that, since the possibility of fire spread has been considered as part of FCIA, this phenomenon is of minimal risk significance.

The potential for fire barrier failure resulting from fire fighting activities was not addressed.

An example of such an'event may include a fire in a compartment with train"A" equipment and cables, while access to this room is via the adjacent train"B" compartment.

b.Assumprions Associated ivith Detecrion and Suppression For the majority of the compartments, the specific fire detection and suppression characteristics of the area were not addressed and analyzed.The IPEEE submittal claims that, except for the control room, no credit was taken for manual fire suppression.

A simple model was utilized for suppression system failure.It was assumed that there is a probability of 0.1 for the fire detection'and suppression systems to fail to stop a fire from damaging a large area.This assumption was applied to six fire areas (both units combined), of which two areas have been screened out;The other four areas consist of the cable spreading rooms and control rooms.The competing phenomenon between fire spread and fire detection/suppression was not modeled.Also, there are no indications for the six fire areas as to whether a detailed analysis of the locations of the critical (associated with safe shutdown)cables and equipment within a compartment was conducted.

Such an analysis is important, since in some cases, it is possible for critical equipment and cables to fail, regardless of suppression system success or failure.'Energy'Research, Inc.19 ERI/NRC 95-504 c.Treatment of Suppression-Induced Damage to Equipment, if Applicable In Reference[I]there is no discussion, for any of the phases of the analysis, of suppression-induced damage (i.e., damage to cables and equipment as a result of the activation of the fire suppression system to extinguish a small fire in the area).However, as part of the discussions on the fire risk scoping study issues raised by Sandia, it is explained that all safety-related areas are equipped with pre-action type sprinkler systems, which minimizes the possibility of water spray onto cabinets and motors.Additionally, in Reference[14[this issue is further discussed for fires in the control room.Cable spreading room fire damage can be mitigated by using the alternate shutdown panel.Thus, it can be concluded that the adverse effects of water spray on e1ectrical equipment are of minimal risk significance.

Computer Codes Used, if Applicable The fire propagation, detection, and suppression analysis has been performed using the formulations provided in Reference[23].The formulations are summarized in the appendix to the IPEEE submittal.

The licensee's analysts have developed a computerized version of the formulations, which has been used in the analysis.2.2.7 Evaluation of Component Fragilities and Failure Modes a.Dejinition of Fire-Induced Failures The submittal provides a short discussion on the fire-induced failures.This discussion addresses the availability of remotely operated valves that have a hand-wheel.

No discussion is provided as to whether or not consideration was given to inadvertent operation of equipment and instrumentation.

b.Method Used to Determine Component Capaciries No criteria is mentioned regarding survival capacities of cables and electrical equipment.

Given that the FIVE methodology was used, Appendix R requirements have been met, and the cables are IEEE 383 qualified, the licensee is expected to have used the proper failure criteria.c.Generic Fragiliries The cables are IEFE 383 qualified.

Therefore, the overall conclusions regarding potential for fire spread and failure are, in general, acceptable.

d.Plant-Specific Fragiliries Plant-specific f'ragilities have not been used.e.Technique Used to Treat Operator Recovery Acrions A simple model (failure probability of 0.1)was employed for human recovery actions in the case of a control room fire or a cab1e spreading room fire.This approach and probability value have been commonly used in other fire risk studies, and are deemed to be sufficiently conservative provided the plant Energy Research, Inc.20 ERI/NRC 95-504 employs a written procedure for using the alternate shutdown system and conducts training drills for this procedure.

For fire events outside the control room or cable spreading room, the human, failure probabilities embedded in the internal events model of the IPE have been used for conditional core damage probability estimation.

Reference f14]provides a discussion of the operator actions that may need to be undertaken from outside the control room.In all cases, it is shown that the fire event cannot affect the effectiveness of the operator from reaching the required plant area, and from conducting the action itself.2.2.8 Fire Detection and Suppression The possibility of fire detection and suppression has been taken into account for six fire areas (both units combined).

As mentioned in Section 2.2.7 above, a simple model was used.A probability of 0.1 for detection and suppression failure was used in the majority of the cases.Specific compartment conditions were not modeled explicitly.

This approach could be optimistic if the critical cables and equipment are located within a small part of the room.In other words, regardless of failure or success of fire detection and suppression, the critical set of cables and equipment may be so close together, that in case of a fire within that specific area, the equipment and cables would be rendered failed by the fire before the suppression system has an opportunity to stop the damage.2.2.9 Analysis of Plant Systems and Sequences a.Key Assumpnons Including Success Criteria and Associated Bases The success criteria were directly taken from the probabilistic risk analysis (PRA)of the plant.b.Event Trees (Funcnonal or Systemic)The IPEEE does not provide any discussion as to the modeling methods used in the PRA.c.'ependency Matrix, ifit Is Different from that for Seismic Events No dependency matrix has been provided.d.Plant-Unique System Dependencies The submittal does not identify any plant-unique system dependencies of relevance to fire risk.e.Shared Systems for Muln-Unit Plant The St.Lucie technical support center is shared between the two units.The effect of this center in recovering from a specific fire event may be significant.

However, the overall effect on the fire risk analysis results is generally minimal.There is a possibility of feeding power from one unit to the other.A cross-tie breaker is available to connect the diesel generator of one unit to the other unit.The possibility of using this feature has been included in the recovery actions of at least six fire scenarios.

Reference[14]indicates that this recovery Energy Research, Inc.21 ERI/NRC 95-504 l

action is taken from the control room, which is isolated from the plant areas where the fire may be present.This recovery action has not been consi'dered for the control room and cable spreading room fires.f.Most Signi/cant Human Anions The IPEEE submittal does not address human actions separately.

In two areas, human actions are mentioned:

in controlling the plant from a fire in the cable spreading room or the control room;and in using the cross-'tie breaker to power one unit from the other unit.2.2.10 Core Damage Frequency Evaluation The licensee has provided two examples of computer outputs from CAFTA that demonstrate the conditional core damage frequencies.

The first 100 minimal cutsets that contribute to the core'damage frequency are also shown.It should be noted that, for these two examples, even with the first 100 cutsets, not more than 80%of the overall frequency is accounted for.No concise set of scenarios constitutes the majority of the core damage frequency.

2.2.11 Analysis of Containment Performance a.Signijicant Containment Performance Insights Containment fires were concluded to be insignificant for St.Lucie.Even though St.Lucie has experienced RCP oil fires in the past, the licensee's conclusion appears to be based on the fact that a large fraction of containment fires were attributed to reactor coolant pump oil leaks, and that the plant has since been equipped with an oil collection system.Containment isolation failure was addressed explicitly in the IPEEE.It was concluded that the probability of isolation failure from a fire is low.The analysis and the conclusions are similar to those reported for other PWRs, and therefore, they are considered reasonable.

b.Plant-Unique Phenomenology Considered Even though it is not discussed specifically in the submittal, it is inferred that the same phenomenology has been used as that in the IPE and PRA.Fire sequences and associated failed equipment were analyzed using the IPE containment event trees.2.2.12 Treatment of Fire Risk Scoping Study Issues a.Assumpnons Used to Address Fire Risk Scoping Study Issues Seismic and fire interaction was addressed through the failure of the fire suppression system and its effects on safety equipment.

It is stated that fire suppression design includes provisions to minimize inadvertent actuation from a seismic event.The possibility of fire occurrence from seismic activity was not addressed in the fire analysis portion of the IPEEE.Energy Research, Inc.22 ERI/NRC 95-504 tg 0 t 0~I l~

2.Fire barriers were assumed to be qualified,per the Appendix R effort.Specific procedures have been cited for inspection and maintenance of the fire doors, fire dampers, fire barriers, and penetration seal assemblies.

3.The IPEEE submittal states that all plant personnel who have unescorted access must undergo fire watch training.In addition, strict training and drills are required for the fire brigade.4.Regarding fire suppression system impact on safety systems, it was argued that all sprinkler systems serving safety areas are pre-action type.This feature, if the system is equipped with an alarm, would minimize the likelihood of inadvertent fire water impact on safety systems.5.Control system interaction was addressed via the use of an alternate shutdown panel and isolation switches.A simple model was used for the operators failing to control the plant from this panel in case of a control room or cable spreading room fire.6.Several procedures have been cited that address different aspects of fires at St.Lucie.A specific procedure exists (although no details are provided)that addresses fire emergencies.

Another procedure addresses control room evacuation.

b.Significant Findings 1.Damage caused by suppression system actuation is not a significant issue.2.The fire brigade undergoes sufficient training, and all personnel who have unescorted access act as fire-watch.

The suppression systems, in safety-related areas, can withstand seismic events, and therefore, seismically induced failure of fire equipment is not a problem.4.Inadvertent actuation of fire suppression is not a problem because charged systems are not located in safety-critical areas.5.Procedures are available that address fire-related issues.6.Potential adverse effects'on plant equipment by combustion products were not addressed.

7.Barrier failures were based on the combustible loading of the area.No consideration was given to the mechanical failure of active barriers (e.g., roll-up doors)..2.2.13 USI A45 Issue St.Lucie was one of the plants evaluated by the NRC for decay heat removal adequacy, in the context of USI AA5.Energy Research, Inc.23 ERI/NRC 95-504 a.Methods of Removing Decay Hear Auxiliary feedwater, main feedwater, and feed and bleed cooling are the methods considered for heat removal during and after a fire event.b.Ability of the Plant to Feed and Bleed St.Lucie has this capability.

c.Credit Taken for Feed and Bleed Credit has been taken for feed and bleed capability.

d.Presence of Thermo-Lag St.Lucie contains Thermo-lag.

The licensee has taken credit for the effectiveness of Thermo-lag in a limited number of areas.This issue is being evaluated by the licensee at this time.2.3 HEQFxenh 2.3.1 High Winds and Tornadoes t 2.3.1.1 General Methodology The IPEEE submittal 1 con ain[]t s analyses for both St.Lucie Units I and 2.St.Lucie-2 began commercial operation in August 1983.All of St.Lucie-2's components and systems required for safe shutdown are located in, or protected by, structures that meet the latest Standard Review Plan (SRP).Thus, the"high winds/tornado"-induced risk to this unit was considered to be insignificant and was qualitatively screened out.St.Lucie-1 began commercial operation in December 1976.The submittal reports that the"high winds/tornado"-induced risk to this unit was considered to be insignificant on the basis that all"safety-related systems and components" are: 1.'ocated within tornado missile protected structures, 2.Provided with missile barriers, 3.Have been shown not to be susceptible to missile impact damage, 4.Have been shown not to adversely affect safety if damaged by a missile, or 5.Have a low probability of missile damage.The above conclusions were reached by performing the first five steps of the general methodology presented in Section 1.1.3.Energy Research, Inc.24 ERI/NRC 95-504 2.3.1.2 Plant-Specific Hazard Data and Licensing Basis The site is periodically a6'ected by the passage of tropical cyclones of various intensities, with the months of September and October having the highest frequency of occurrence.

Tornadoes and waterspouts have been observed throughout the year in that part of Florida.The parameters applicable to the design-basis tornado are: t I~External wind forces from a tornado funnel with a horizontal rotational velocity of 300 mph and a horizontal translational velocity of 60 mph, for a total wind velocity of 360 mph.~A decrease in atmospheric pressure of three (3)psi.~Impact loads from a tornado generated missile.The parameters applicable to the design-basis tornado, used for identification of site-specific meteorological conditions, are in agreement with the requirements of R.G.1.76[24].ASCE Paper No.3269[25]and ANSI A58.1[26]were used to transform the, wind velocity into pressure loadings on structures.

The St.Lucie-2 missile spectrum is based on a tornado zone I site, as identified in R.G.1.76.Identification of applicable regional and site-specific meteorological conditions, and hurricane/tornado wind loading, was performed using the same parameters and procedures as used for Unit 2.The Unit 1 design-basis tornado missile spectrum consists of a 2" x4" x10'ood plank traveling at 360 mph, and a 4000 pound automobile traveling at 50 mph.During the Unit 1 licensing review, the NRC staff requested that the Unit 1 capability to withstand a more extensive missile spectrum be evaluated.

2.3.1.3 Significant Changes Since Issuance of the Operating License The submittal does not catalog any significant changes since the time the plant operating license (OL)was issued 2.3.1.4 Significant Findings and Plant-Unique Features No significant findings are cited in the submittal.

A summary of the walkdown procedures used by the licensee, and the qualifications of the team members performing the walkdown, are not provided in the submittal.

2.3.1.5 Hazard Frequency WASH-1300[27];a Dames and Moore (DAM)study, as presented in the St.Lucie-1 FSAR, Appendix F;and NUREG/CR<710

[28]are referenced in the submittal.

However, the NUREG/CRQ710 values are those which were used for the evaluation.

Energy Research, Inc.25 ERI/NRC 95-504 2.3.1.6 Bounding Analysis Bounding analyses were performed for the diesel oil storage tanks, and the intake cooling water (ICW)system and component cooling water (CCW)system pipes.No specific assumptions have been stated.However, the following key implicit assumptions seem to have been made: 1.The conditional missile impact probability reported'in NUREG/CR4710 is applicable to the targets under consideration in the study.2.Threat from only one missile is credible.The submittal reports that the frequency of damage (tornado strike frequency x missile impact probability) for each one of the structures under consideration is less than 10~/ry, and screens out the contribution of the"tornadoes/high winds" hazard to the plant operational risk at this stage of the analysis.2.3.2 External Flooding 2.3.2.1 General Methodology The methodology consisted of identifying the major events of concern, assessing the potential threat presented by the hazards, and evaluating plant defenses against these hypothetical events.2.3.2.2 Plant-Specific Hazard Data and Licensing Basis The probable maximum hurricane (PMH)surge and probable maximum precipitation (PMP)were considered as the major events of concern to St.Lucie.The hydrologic conditions that will optimize the potential erosion at the St.Lucie site were established by conducting a study of historical looping and stalled hurricanes for the time period of 1900 to 1973.During the probable maximum flood, which results'from the PMH surge, the high water level is 17.2 ft mean low water (MLW).The plant grade is at elevation+18.5 ft.MLW, and minimum entrance elevation to all seismic Category-I buildings is+19.5 ft.MLW.Seismic Category-I structures and safety-related components are protected from the effects of high water level and wave run-up that are associated with PMH conditions, by: 1.Designing structures and components to withstand such effects where functionally required, 2.Positioning of the structures and components such that they are located at sufficient grade to preclude inoperability due to external flooding, and/or 3.Housing them within waterproof structures.

The design-basis probable maximum precipitation for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, used in the analysis, was 24.1 inches, over an area of 10 square miles or less.Energy Research, Inc.26 ERI/NRC 95-504

~7 Lw 4 The roof leaders have been designed for a rainfall intensity of six inches per hour.Short periods of more intense rainfall result in water running off the edges of the roofs, with no adverse effects to safety-related equipment.

No water build-up on the roofs in excess of 2" is possible, except for the shield building dome, which is surrounded by a 1'-6" high parapet.The submittal states that none of the above conditions adversely affects the structures or safety-related equipment.

The threat of damage from the probable maximum flood on streams and rivers was discounted on the basis that there are no such waterways located in close vicinity of the plant.The risk from potential dam failures was also discounted, since no dams are located within the hydrological influence of Hutchinson Island.The risk presented by the probable maximum tsunami flooding was discredited on the basis that: (1)there is no evidence to indicate the existence of potential tsunami generators offshore in the site area;and (2)any possible effects at the site location from tsunami generated from far-fiel sources will be negligible compared to the effect of surges caused by the PMH.2.3.2.3 Significant Changes Since Issuance of the Operating License The submittal does not catalog any significant changes that have occurred since the time of OL issuance.The submittal does note that the licensee considered the effects of latest, increased PMP criteria, and concluded that there are no concerns associated with the site flooding levels or roof ponding that couldaccompany an increased PMP.2.3.2.4 Significant Findings and Plant-Unique Features For Unit 2, the design-basis events for flood protection of safety-related equipment and facilities meet the requirements of R.G.1.59, except that the PMH pertinent to the site is the basis for the computation of the probable maximum surge (PMS).The R.G.1.59 PMS would equal+16.7 ft.MLW, whereas the surge assumed by the St.Lucie FSAR analysis is+17.2 ft.MLW.The flood protection recommendations of R.G.1.102 were followed.The St.Lucie Unit 1 safety evaluation report (SER)was finalized before the SRP was issued.Thus, evaluation of the conformance of Unit 1 to the SRP criteria was made by comparing the Unit 1 hazard and design to that of Unit 2.The comparison indicates that the Unit 1 flood protection is similar to Unit 2, and thus meets the SRP criteria.The submittal does not discuss any walkdowns that were performed during the analysis.2.3.2.5.Hazard Frequency Since the St.Lucie Units 1 and 2 designs were determined to meet the R.G.1.59 and SRP criteria, the flooding hazard was qualitatively screened out, and no hazard frequency was estimated.

Energy Research, Inc.27 ERI/NRC 95-504

~4~g k

2.3.3 Transportation

and Nearby Facility Accidents 2.3.3.1 General Methodology The methodology used for evaluation of transportation and nearby facility accidents consists of the following steps: 1.2.3.4.Review of plant-specific hazard data and licensing bases;Determination of conformance of the plant risk significant structures to the 1975 SRP criteria;Screening of plant structures

'that meet the SRP criteria for a specific hazard;and Determination of the hazard frequency for those structures that do not meet the SRP criteria.2.3.3.2 Plant-Specific Hazard Data and Licensi.Ba".s a.Airpons and Airways There are no major airports within 10 miles of the plant.The nearest major airport with commercial facilities is 48 miles from the plant.However, there are several smaller airports in close vicinity to the plant.b.Warerivays The Atlantic Ocean shipping lanes are about 10 to 15 nautical miles east of the plant, thus, no ship or barge explosion can affect the plant structures.

Barges passing in the Intracoastal Waterway are the other source of potential hazard to the plant.c.Highways The governing explosive and/or flammable event was judged to arise on State Route A1A, which passes about 750 feet east of the diesel oil storage tanks, due to a liquefied propane truck accident.The submittal states that the probability of having a potential accident whose consequence can result in a radionuclide release in excess of 10CFR100 guidelines is significantly less than 10'/yr, based on the calculations in the Unit 2 FSAR.The results in the FSAR are reported to have been validated based on discussion with local authorities and a drive-through of the area.However, since the report of the discussions and the drive-through observations are not documented in the submittal, their conclusions could not be evaluated for this TER.d.RailroadsThe nearest railroad is 2 miles west of the plant.This distance was considered to be sufficient to preclude adverse effects to the plant from accidental explosions on the railroad.Chlorine is the most likely hazardous material to be shipped on the rail cars.The submittal references the Unit 2 FSAR which provides an evaluation demonstrating that the probability of a chlorine release adversely affecting the plant is less than 10~per year.The results in the FSAR analysis are reported to be valid based on discussion with local authorities and a drive-through of the area.However, since the report of the discussions and Energy Research, Inc.28 ERI/NRC 95-504 the drive-through observations are not documented in the submittal, their conclusions could not be evaluated for this TER.e.Toxic Chemical Events Chemicals that are non-volatile or liquids, or that spontaneously combust in air, were not considered to pose a threat to control room habitability.

Also, chemicals for which their potential for ignition constitutes a greater hazard than their toxicity, were eliminated from consideration.

The threat to control room habitability from the toxic chemicals which were not eliminated on the basis of the criteria stated above, were eliminated by a detailed assessment of their atmospheric transport and potential for infiltrating into the control room.Ammonium hydroxide, which is stored on site;carbon dioxide;and chlorine, which is the principal toxic substance transported by the Florida East Coast Railway (FECR), are the major chemicals for which their potential impact on control room habitability was analyzed in detail.The threats posed by ammonium hydroxide and carbon dioxide were dismissed on the basis that the concentration of these chemicals inside the control room remains well below the toxicity limit.The threat posed by the release of cMorine due to a railroad accident was dismissed based on the frequency of the design-basis event.f.Indus(rial Faciliries There are no military bases,"missile installations, chemical plants, hazardous material storage areas or drilling operations within 10 miles of St.Lucie.There are no pipelines within 5 miles of the plant.2.3.3.3 Significant Changes Since Issuance of the Operating License The submittal does not catalog any significant changes since the time of OL issuance.2.3.3.4 a.Airways Significant Findings and Plant-Unique Features The estimated number of operations per year from one of the local airports was found to be greater than the SRP screening value of 144,000.b.Waterways Considering the maximum size of barges passing the plant site;Equation 1 of R.G.1.91,"Evaluation of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plants"[29];and the distance between any safety-related structures and the nearest Intracoastal Waterway shipping channel, the risk of damage from a barge explosion was dismissed.

Energy Research, Inc.29 ERI/NRC 95-504

~g 2.3.3.5 Hazard Frequency a.Airways Since the estimated number of operations per year for a local airport was found to be greater than the SRP screening value, the Sandia National Laboratories (SNL)[30]and SRP methods were used to show that the aircraft crash frequency is below 10~/yr.b.ToxI'c Chemical Events The accidental release of the entire contents of chlorine from a tank car was assumed to be an initiating event for a design-basis accident.The frequency of such an event was calculated according to the following equation: where Pi=annual probability of design-basis event under atmospheric stability Class I,'nvolving the i-th chemical;P-probability of a design basis accident for a mobile source per unit length of travel;M,=D=j annual numbers of trips involving the i-th chemical;annual probability of an atmospheric stability class;the length of road, rail, or river in sector j;FJI wind frequency from sector j to outside air intake of the control room for stability Class 1;and number of wind direction sectors.Using the above formula, the overall probability of an event that may affect control room habitability was determined to be 1.4x10'/yr.

Therefore, the St.Lucie Unit 2 design was determined to either meet the SRP criteria, or have a low hazard frequency.

Due to the proximity of Units 1 and 2, the hazard analysis for Unit 2 was.judged to be applicable for Unit 1.2.3.4 Lightning and Others The submittal presents a discussion of the St.Lucie lightning protection system.Based on a review of the St.Lucie FSARs, the plant's operating history, and NUREG/CR-4710 findings, the submittal concludes Energy Research, Inc.'0 ERI/NRC 95-504 I

that there is no unique plant vulnerability to lightning at St.Lucie, and that the impact of lightning on plant risk is bounded by the internal events analysis.2.4 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 which could develop that may create a number of potential control system vulnerabilities.

Control system i 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 As indicated in the response to Question II-3 in Reference[14], for the possibility of occurrence of loss of offsite power and reactor coolant pump seal failure from a fire," a thorough analysis has been conducted.

However, it is not clear whether this analysis considered the possibility of hot short failures in control cables, and inadvertent opening of the isolation valves of reactor coolant system high and low pressure interfaces.

Since the submittal has followed 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-fighting 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 areas away from the fire Reference[31]identifies possible reduction of manual fire-fighting effectiveness and 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-156,"Systematic Evaluation Program (SEP)" Reference[31]provides the description of each SEP issue stated below, and delineates 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-156 may be found.Energy Research, Inc.31 ERI/NRC 95-504

Settlemenr of Foundarions and Buried Equipment IEI: h bj i fbi SEPi i k fy.systems and components are adequately protected against excessive settlement.

The scope of this issue includes review of subsurface materials and foundations, in order to assess the potential static and seismically induced settlement of all safety-related structures and buried equipment.

Excessive settlement or collapse of foundations could result in failures of structures, interconnecting piping, or control systems, such that the capability to safely shutdown the plant or mitigate the consequences of an accident could be comprised.

This issue, applicable mainly to soil sites, involves two specific concerns:~potential impact of static settlements of foundations and buried equipment where the soil might not have been properly prepared, and seismically induced settlement and potential soil liquefaction following a postulated seismic event.Since static settlements are not believed to be a concern, the focus of this issue (when considering relevant information in IPEEEs)should be on seismically induced settlements and soil liquefaction.

It is anticipated that full-scope seismic IPEEEs will address these concerns, following the guidance in EPRI NP-6041.St.Lucie is a reduced-scope plant, and Category I structures for both units are founded on Category-I fill, underlaid by cemented sands and sandy limestones.

The IPEEE submittal provides no discussion of the potential and effects for seismically induced settlements.

or soil liquefaction.

Information on site geology can be found in Section 2.2 of Reference[5].Dam Integrt'ty and Site Flooding t31: bi i fhi'i i 0 lii f p flooding and to ensure a cooling water supply.The safety functions would normally include remaining stable under all conditions of reservoir operation, controlling seepage to prevent excessive uplifting water pressures or erosion of soil materials, and providing sufficient freeboard and outlet capacity to prevent overtopping.

Therefore, the focus is to assure that adequate safety margins are available under all loading conditions, and uncontrolled releases of retained water are prevented.

The concern of site flooding resulting from non-seismic failure of an upstream dam (i.e., caused by high winds, flooding, and other events)is addressed as part of the SEP issue"site hydrology and ability to withstand floods." The concerns of site flooding resulting from the seismic failure of an upstream dam and loss of the ultimate heat sink caused by the seismically induced failure of a downstream dam should be addressed in the seismic portion of the IPEEE.The guidance for performing such evaluations is provided in Section 7 of EPRI NP-6041.As requested in NUREG-1407, the licensee's IPEEE submittal should provide specific information addressing this issue, if applicable to its plant.Information included for resolution of USI A-45 is also applicable to this concern.The St.Lucie IPEEE submittal states (on page 79)that no dams are located within the hydrological influence of the site location on Hutchinson Island.Energy Research, Inc.32 ERI/NRC 95-504 IW'~

~~~~Site Hydrology and Ability to Withstand Floods t3I.'bi i fUii i<<id ifyh i hd gi h in order to ensure the capability of safety-related structures to withstand flooding, to ensure adequate cooling water supply, and to ensure in-service inspection of water-'control structures.

This issue involves assessing the following:

Hydrologic conditions

-to assure that plant design reflects appropriate hydrologic conditions.

Flooding potential and protection

-to assure that the plant is adequately protected against floods.J Ultimate heat sink-to assure an appropriate supply of cooling water during normal and emergency shutdown.As requested in NUREG-1407, the licensee's IPEEE submittal should provide information addressing these concerns.The concern related to in-service inspection of water-control structures, a compliance issue, is not being covered in.the IPEEE.The St.Lucie IPEEE submittal (Section 5.2)has included a discussion of external floods, including effects of hurricane storm surge (pages 76 to 78)and probable maximum precipitation (pages 78, 79, and 82).Industrial Hazards l3'I: h bi fui'i U<<h i f f structures, systems, and components would not be jeopardized due to accident hazards from nearby facilities.

Such hazards include: shock waves from nearby explosions, releases of hazardous gases or chemicals resulting in fires or explosions, aircraft impacts, and missiles resulting from nearby explosions.

As requested in NUREG-1407, the licensee's IPEEE submittal should provide information addressing this'ssue.The St.Lucie IPEEE submittal (Section 5.3)includes the following information of relevance to this issue: Section 5.3.1 of the submittal identifies nearby transportation routes;Section 5.3.2 discusses nearby industrial facilities; Section 5.3.3 discusses offsite and onsite sources of hazardous materials or explosives; Section 5.3.4 discusses hazard data for airports and airways;Section 5.3.5 discusses hazard data for explosions; and Section 5.3.6 discusses potential toxic chemical events.Tornado Missiles bi<<i'i"'i'Ip 9>>(SEP plants)are adequately protected against tornadoes.

Safety-related structures, systems, and components need to be able to withstand the impact of an appropriate postulated spectrum of tornado-generated missiles.As requested in NUREG-1407, the licensee's IPEEE submittal should provide information addressing this issue.The St.Lucie IPEEE (Section 5.1)has involved an evaluation of tornadoes, including tornado-induced missiles.Detailed information and evaluation of tornado-induced missiles is provided in Section 5.1.1.Energy Research, Inc.33 ERI/NRC 95-504 Severe Weather Effects on Structures

[31]: The objective of this issue is to assure that safety-related structures, systems, and components are designed to function under all severe weather conditions to which they may be exposed.Meteorological phenomena to be considered include: straight wind loads, tornadoes, snow and ice loads, and other phenomena judged to be significant for a particular site.As requested in NUREG-1407, the licensee's IPEEE submittal should provide information specifically addressing high winds and floods.Other severe weather conditions (i.e., snow and ice loads)were determined to have insignificant effects on structures (see Chapter 2 of NUREG-1407).,The St.Lucie IPEEE has included evaluations of high winds (hurricanes, and tornadoes) and external floods.Section 5.1 of the submittal discusses hurricanes and tornadoes, and Section 5.2 of the submittal discusses external floods.Section 5.4 of the submittal includes an evaluation for lightning.

Design Codes, Criteria, and Load Combinarions t3f:Thbi i" fhii i<<h<<ip fy should be designed, fabricated, erected, and tested to quality standards commensurate with their safety function.All structures, classified as Seismic Category I, are required to withstand the appropriate design conditions without impairment of structural integrity or the performance of required safety functions.

Due to the evolutionary nature of design codes and standards, operating plants may have been designed to codes and criteria which differ from those currently used for evaluating new plants.Therefore, the focus of this issue is to assure that plant Category I structures will withstand the appropriate design conditions (i.e., against seismic, high winds, and floods)without impairment of structural integrity or the performa'nce of required safety function.As part of the IPEEE, licensees are expected to perform analyses to identify potential severe accident vulnerabilities associated with external events (i.e., assess the seismic capacities of their plants either by performing seismic PRAs or SMAs).The St.Lucie IPEEE has included an evaluation of potential vulnerabilities associated with external events.The submittal does not systematically identify codes, criteria, and load combinations used in design.However, Sections 2.5, and 3.1 to 3.5 of Reference[5]provide some information related to seismic design of structures and equipment; Section 5 1 of the IPEEE submittal provides information related to wind design of structures; Section 5.2 contains some information related to design conditions for withstanding floods;and Section 5.3 of the submittal provides information on design criteria related to transportation and nearby facility accidents, including explosions.

Seismic Design of Structures, Systems, and Components

[31]: The objective of this SEP issue is to review and evaluate the original seismic design of safety-related structures, systems, and components; to ensure the capability of the plant to withstand the effects of a Safe Shutdown Earthquake (SSE).The St.Lucie IPEEE is based on the seismic adequacy evaluation performed as part of the licensee's resolution of USI A-46 concerns (Reference

[5]).Sections 2.5 and 3 of Reference[5]provide some information related to the seismic design of structures and components, and Section 4 of Reference[5]provides a description of the approach and findings of the seismic adequacy evaluation.

Energy Research, Inc.34 ERI/NRC 95-504 Shutdown Systems and Electrical Instrumentarion and Control Features t3I'.th'l'l h 3 I fpl reliable shutdown using safety-grade equipment.

The issue on electrical instrumentation and control is to assess the functional capabilities of electrical instrumentation and control features of systems required for safe shutdown, including support systems.These systems should be designed, fabricated, installed, and tested to quality standards, and remain functional following external events.In IPEEEs, licensees were requested to address USI A45,"Shutdown Decay Heat Removal (DHR)Requirements," and to identify potential vulnerabilities associated with DHR systems following the occurrence of external events.The resolution of USI A-45 should address these two issues.St.Lucie Nuclear Plant had been used as a case study plant by Sandia National Laboratories for probabilistic evaluation of decay heat removal adequacy, in the context of USI A<5.This issue was addressed as part of the IPEEE, in general, and pertinent information is provided in Sections 4.9 and 5.1 (page 72)of the submittal.

Sections 2.1.13 and 2.2.13 of this TER summarize review findings related to USI A<5, respectively, for seismic events and fire events.2.4.4 GSI-172,"Multiple System Responses Program (MSRP)" Reference[31]provides the description of each MSRP issue stated below, and delineates 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 t33:CCP II It h I I 3 f 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.In IPEEEs, licensees were requested to address only the human errors involving operator recovery actions following the occurrence of external initiating events.A very limited discussion of operator recovery actions, following a seismic event, is provided in Section 4.4 of Reference[5].Section 4.6 of the submittal provides some discussion on the treatment of human recovery actions in the internal fire analysis.Non-Safety-Related Control System/Safety-Related Protecrion System Dependencies 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'PE 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 Energy Research, Inc.35 ERI/NRC 95-504 shutdown and control room panel interactions," GSI-147, for fire events, and"seismically induced spatial and functional interactions" for seismic events.Information provided in the St.Lucie IPEEE submittal pertaining to seismically induced spatial and functional interactions is identified below (under the heading Seismically Induced Sparial and Functional Interacrions), whereas information pertaining to fire-induced alternate shutdown and control panel interactions has already been identified in Section 2.4.1 of this TER.Heat/Smoke/Water Propagarion Effects Pom Fires t3 I: d i f I p i f*.NI I d train could potentially be damaged in one of following ways: Heat, smoke, and water may propagate (e.g., through HVAC ducts or electrical conduit)into a second fire zone, and damage a redundant train of equipment.

A random failure, not related to the fire, could damage a redundant train.Multiple non-safety-related control systems could be damaged by the fire, and their failures could affect safety-related protection 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 addressed in GSI-148.The concern of alternate shutdown/control room interactions (i.e., hot shorts and other items just mentioned) is addressed in GSI-147.Information provided in the St.Lucie IPEEE submittal pertaining to GSI-147 and GSI-148 has already been identified in Sections 2.4.1 and 2.4.2 of this TER.Section 4.8 of the submittal presents some limited information pertinent to this issue.Ejjects of Fire Suppression System Actuarion on Non-Safety-Related and Safety-Related Equipment[31]: Fire suppression system actuation events can have an adverse effect on safety-related components, either through direct contact with suppression agents or through indirect interaction with non-safety related components.

Items 2 and 5 of Section 4.8 of the submittal present some limited information pertinent to this issue sects of Flooding and/or Moisture Intrusion on Non-Safety-Related and Safety-Related Equipment[31]: 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 backflow through parts of the plant drainage system.The IPE process Energy Research, Inc.36 ERI/NRC 95-504 addresses the concerns of moisture intro'ton and internal flooding (i.e., tank and pipe ruptures or backflow through part of the plant drainage sys!;m).The guidance for addressing the concern of external flooding is provided in Chapter 5 of NUREG-I"07, and the concern of actuations of fire suppression systems is provided in Chapter 4 of NUREG-1407.

The following information is provided r:l, vant to this issue: external flooding is discussed in Section 5.2 of the St.Lucie IPEEE submittal, and Items 2 and 5 of Section 4.8 present some limited information concerning inadvertent actuation of fi;e suppression systems.Seismically Induced Spatial and Funct'anal Interacrions

[31]: Seismic e ants have the.potential to cause multiple failures of safety-related systems through spatial and functional in',era:tions.

Some particular sources of concern include: ruptures in small piping that may disable e.<<entiai'plant shutdown systems;direct impact of non-seismically qualified structures, systems, and con ponents that may cause small piping failures;seismic functional interactions of control and safety-reiat";>r>><ection systems via multiple non-safety-related control systems'ailures; and indirect impacts, such as d~~~'eneration, disabling essential plant shutdown systems.As part of the IPEEE, it was specifically reque~'..i that se!smi:ally induced spatial interactions be addressed during plant ivalkdowns.

The guidance for per oiining such walkdowns can be found in EPRI NP-6041.The St.Lucie seismic adequacy evali:ation

{Reference

[5])has included a seismic walkdown which investigated the potential for adverse I hy~i'al interactions.

Relevant information can be found in Section 4.7 (particularly Section 4.7.2.3)of Reference[5].Seismically Induced Fires~zgjpIign~zJ~~

[31]: Sei~~al!y indu'ed fires may cause multiple failures of safety-related systems.The occurrence of a seisnii'vent cnuid create fires in multiple locations, simultaneously degrade fire suppression capability, and prevent mitigation of fire damage to multiple safety-related systems.Seismically induced fires is i>>ne aspect of seismic-fire interaction concerns, which is addressed as part of the Fire Risk Scoping Study{)=RSS)issues.{IPEEE guidance specifically requested licensees to evaluate FRSS issues.)In IPEEEs,~e~mt."a!l~induced fires should be addressed by means of a focused seismic-fire interactions walkdown that follows the guidance of EPRI NP-6041.Section 4.8 of the submittal very br!fi~'i~cu<<es seismic-fire interactions; however, no evaluation of seismically induced fires is provided~part of the St.Lucie IPEEE submittal.

Seismically Induced Fire Suppression System Actuanon[31]: Seis-">>c events can potentially cause multiple fire suppression system actuations which, in turn, may cai'-'-":.ilures of redundant trains of safety-related systems.Analyses currently required by fire protection ieg,ilations 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.Items 2 and 5 of Section 4.8 of the:-;<<';;i-al;resent some limited information pertinent to this issue.2 Energy Research, Inc.37 ERI/NRC 95-504 Seismically Induced Flooding till: W Mlyi d kfl i i illy Iiii iii 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 St.Lucie IPEEE submittal has not included a discussion of seismically induced flooding.Seismically Induced Relay Chatter Wl ih W<<ii~qe, i one of the following conditions:

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

~.be seismically qualified; or~be chatter acceptable.

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 specifically requested licensees to address the issue of relay chatter.As noted in Section 2.1.9 of this TER, the St.Lucie IPEEE submittal does not mention relay chatter evaluation.

However, during NRC's USI AW review of Turkey Point, Units 3 and 4, and St.Lucie, Unit 1, it was revealed that the licensee had assessed bad actor relays, verified mountings of relays, and demonstrated that there were no deleterious effects of chatter of bad actor relays.The NRC accepted the licensee's relay evaluation for USI A<6 resolution.

Evaluanon of Earthquake Magnitudes Greater than the Safe Shutdown Earthquake

[31]: 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 are 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.St.Lucie is designated as a reduced-scope plant in NUREG-1407, and consistent with the relevant guidelines for a reduced-scope plant, the IPEEE has considered seismic input equivalent to the SSE level.Earthquake loads in excess of the SSE have not been considered.

sects of Hydrogen Line Ruptures[31j: Hydrogen is used in electrical generators at nuclear plants to reduce windage losses, an'd 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 Energy Research, Inc 38 ERI/NRC 95-504

,~

~~~1 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.

Section 5.3.3 of the St.Lucie IPEEE submittal identifies compressed hydrogen as a potential explosion , source;however, no discussion pertaining to hydrogen line ruptures is provided in the submittal.

Energy Research, Inc.39 ERI/NRC 95-504

~~

3 OVERALL EVALUATION AND CONCLUSIONS 3.1 SeiSaIlC The approach chosen by the licensee for responding to the seismic IPEEE does not address all relevant issues and concerns for St.Lucie Nuclear Plant, a reduced-scope site.A comparison of major features of the FPL seismic adequacy program with the guidelines for a reduced-scope seismic evaluation, is summarized in Table 3.1 below.As can be seen from this table, the primary deficiencies of the FPL approach are: a significantly lesser scope of components in the FPL approach;a limited treatment of human actions for the St.Lucie studies;and no treatment of containment systems in the FPL program.Table 3.1 Comparison of FPL's Site-Specific Seismic IPEEE Program Versus NUREG-1407 Recommended Guidelines for a Reduced-Sco e Seismic Evaluation Element of IPEEE Evaluation Walkdown Relay Evaluauon Soil Failures Screening Criteria Seismic Input Evaluation of Outliers Non-Seismic Failures and Human Acuons Contairunent Performance Assessment USI ARS GI-131 Reduced-Scope Evaluation Guidelines Scope should include all SSEL active components and passive components (structures, raceways, heat exchangers, tanks.piping.etc.)nccded to ensure complete prcferrcd and alternate success paths.USI AW evaluation for USI AA6 plant;No evaluation for non-USI AM plant.No evaluation is necessary.

SRT judgment;GIP screening guidance;Anchorage check based on SSE spectmm and FSAR in.structure response spectrum gRS).SSE spectrum and FSAR IRS (or new mean plus onc-sigma IRS).GIP provisions for USI AM Items;FSAR requirements for non USI AM items.These should bc qualitatively addressed; success paths are chosen to screen out vulnerability to these items.Walkdown, screening, and outlier evaluation of containment structure and components of containment systems.Walkdown, screening, and cvaluauon of decay heat removal outlicrs.Walkdown, screening, and evaluation of seismic adequacy of flux mapping system.FPL's Site-Specific Seismic Adequacy Program Scope includes SSEL active components and gt)its passive components (tanks, heat exchangers);

component list appears incomplete; sclectcd success paths not identilicd.

USI AA6 treatment of electrical raceways was approved by the NRC[7]Bad actor evaluation for St.Lucia-l, approved by NRC for USI AM[71;No evaluation for St.Lucie-2.No evaluation.

SRT judgment;SSRAP bounding spectrum;Anchorage check based on SSE spectrum and FSAR IRS.SSE spectrum and FSAR IRS.Conservative calculation of capacity versus dcmat;d;demand based on conservative usc of SSE spectrum and FSAR IRS;HCLPF calculations for large Qat.bonomed tanks at St.Lucia-2.Limited qualitauve evaluation of actions associated with success path.No evaluation.

No specific evaluation; only partially addressed in chosen success path.Not applicable to St.Luci>>.In addition, the format for documenting the seismic IPEEE was not well structured, and did not follow the recommendations of NUREG-1407.

Energy Research, Inc.40 ERI/NRC 95-504 Despite these significant deficiencies, the St.Lucie seismic evaluations do, nonetheless, address some meaningful IPEEE-related concerns, and have resulted in a small number of plant seismic safety enhancements.

Furthermore, the NRC has already approved many aspects of the licensee's seismic adequacy evaluation approach for USI AW resolution

[6,7]that pertain also to the seismic IPEEE.Based on this submittal-only review, and in consideration of the NRC's findings for USI A-46, the following items are identified as the primary strengths and weaknesses of the seismic IPEEE submittal for St.Lucie Nuclear Plant: 1.The study implements a meaningful approach for screening and outlier evaluation of the limited set of components it addresses.

2.The use of highly experienced seismic walkdown experts has been consistent with the study's heavy reliance on SRT judgments.

3.A number of outliers have been identified, and meaningful corrective safety enhancements have been proposed.1.The SSEL is deficient.

2.A seismic containment performance assessment was not conducted.

3.The treatment of human actions is deficient.

4.The submittal does not provide adequate documentation of seismic-fire/fiood interaction concerns, including component-specific walkdown findings.5.The seismic IPEEE is incomplete with respect to reduced-scope evaluation recommendations found in NUREG-1407.

6.The seismic IPEEE submittal is not documented in accordance with the format recommended in NUREG-1407, Appendix C 3.2 Egg The licensee has expended considerable effort in the preparation of the St.Lucie fire IPEEE.The IPEEE report complies with the conditions set forth in Reference[3].The licensee has employed a proper methodology and data base for conducting the fire analysis.The FIVE methodology has been used for this purpose.The following are the strengths and weaknesses of the submittal:

, Energy Research, Inc.41 ERI/NRC 95-504 1.The overall presentation is clear and well-organized.

There are taf>les and figures to provide information to support the analysis and the conclusions.

3.F Statecf-the-art methodology and proper data have been used.Based on the data presented, it can be concluded that the licensee has conducted a reasonable analysis.The overall results are within the range of conclusions reached in other PWR fire risk studies.Ecakm~1.The possibility of hot shorts and resulting RCS failure from a fire event has not been addressed explicitly in the IPEEE submittal.

Fire suppression system failure probability may not have been used properly.If a critical set of cables and equipment are within a small region of a compartment, the successful operation of the fire suppression system may not matter.3.Probability of failure of the redundant equipment and models used for arriving at the conditional probability of core damage given a fire scenario have not been explained in sufficient detail.4.Cross-zone fire propagation, where active fire barriers are employed, was not addressed explicitly.

The submittal does not provide sufficient information for the reviewers to be able to verify such aspects of the analysis as: the probability of redundant train failures given a fire, fire-induced initiating events, damage from fire suppression system activation, and fire modeling and hot shorts.6.The submittal does not address the possibility of a seismic event leading to a fire.7.There are several compartments for which the frequency of core damage is slightly less than 10~/ry.These areas, although marginally within the screened-out range, have not been addressed in any detail.Certainly, notwithstanding the above observations, the licensee has gained an important experience from the exercise of analyzing the plant for potential fire vulnerabilities.

3.3 HE~LEzenh In general, the conclusions of the submittal are adequately supported and follow the accepted practice and guidelines of NUREG-1407.

Three categories of HFO events are addressed in some detail: high winds and tornadoes, external flooding, and transportation and nearby facility accidents.

No particular weaknesses were found in the submittal regarding the last two categories.

Energy Research, Inc.42 ERI/NRC 95-504 The following provides a description of the areas in the high-winds/tornadoes analysis that contain conclusions which are difficult to verify: Two of the intake cooling water (ICW)valve operators in the valve pit were identified as being potentially vulnerable to vertical missiles.If one or both valves become inoperable, manual valves can be used to ensure adequate ICW flow.The report concludes that"since the valve operators are located below grade, are physically well separated, have an alternate means available to isolate the tie-lines to the turbine water cooling system, and have redundant systems available," adequate tornado protection has been provided (page 56, last paragraph).

However, the report does not specify: (1)whether a procedure exists, or sufficient time would be available, for performing the cited-manual action;(2)on what basis it is concluded that the valves are"well separated";

and (3)where the alternate components are located.For many structures, the inherent capability of the structure was credited to accommodate a specific hazard-for example, the CCW heat exchangers and piping (page 57, third paragraph).

It appears from the submittal that if a structure is judged to be able to withstand a single missile, then it has the capacity to withstand simultaneous (concurrent) impact by several missiles.On page 58, last paragraph, it is stated that the diesel oil tanks are"sufficiently separated to provide an acceptable level of tornado resistance capability," without providing a basis for this statement.

On page 59, second paragraph, it is stated that"...a commitment has been obtained from a local fuel company,..., to supply fuel oil on a 24-hour emergency basis." However, from the explanation provided in the report it is not clear: (a)(b)that, in the event of diesel oil storage tank unavailability, sufficient onsite fuel oil would be available to operate the diesel generators before the arrival of the offsite fuel oil;and that the assumption that the local fuel company's oil supply, or its means of oil delivery, would not be affected by the same tornado that is postulated to hit the site, is valid.In addition to the above ambiguities, determination of the hazard frequency (starting from page 69)is somewhat confusing, and contains the following potentially optimistic assumptions and suppositions:

~The study references the NUREG-1407 statement that, if the original design does not meet the~b U original design basis is sufficiently low, such that the estimated core damage frequency is less than 10'/ry (page 70).In the submittal, as far as"tornadoes/high winds" hazard is concerned, the above statement is interpreted to mean that, if the contribution to core damage frequency as a result of a tornado-induced damage to any one target is less than 10~/ry, then that target can be excluded from further evaluation.

Using this interpretation, some potential targets are screened based on the low frequency of hazard criterion, and the remaining ones by performing bounding analyses (page 73).This interpretation, however, is optimistic and underestimates the potential risk.To screen a hazard, the cumulative risk induced by that hazard has to be below the screening value, not the Energy Research, Inc.43 ERI/NRC 95-504

~~individual risks from each potential target, since, for example, a tornado can impact more than one target, and will generate more than one missile.On page 71, third paragraph, it is stated that, in evaluation of tornado missiles, in addition to weighing the probability of a certain-intensity-tornado occurring and generating a missile, the following must be considered:

P4=Probability that a missile, if generated, will impact the component; P,=Probability that, if struck, loss of system function occurs;and P,=Probability that an independent single failure occurs in the struck component's redundant counterpart.

The submittal assumes a value of 10'or P4, by referencing a docketed Shearon Harris calculation for missile impact on a service water pump.However, no comparison between the characteristics of the service water pump location and the location of the St.Lucie potential targets was made.Since the P, value is highly location specific, assuming a value of 10'or P, may be optimistic.

The submittal also provides a screening review of other external events that may present a potential severe accident vulnerability at St.Lucie Units I and 2, a summary of which is presented in Table 5-23 of the submittal.

Based on this screening, forest fires are claimed to have a minimal potential impact on the plant, and the impact is considered to be bounded by loss of offsite power.However, the potential impact of smoke generated by the fires on the control room habitability, on equipment, and on loss of clean air and instrument air are not addressed.

In general, the approach appears to be sound.However, a comprehensive screening of all potential external fire sources and their effects has not been documented.

Energy Research, Inc.ERI/NRC 95-504 4 IPEEE INSIGHTS, IMPROVEMENTS)

AND COMMITMENTS 4.1 5ehmz The key seismic IPEEE findings are primarily walkdown related;few quantitative insights have been derived from the seismic evaluations.

Thus, no values for seismic core damage frequency, plant-level fragility capacity nor plant-level HCLPF capacity have been estimated as a result of the seismic IPEEE.The seismic adequacy evaluation for St.Lucie-I revealed a number of outliers for which safety.enhancements have been proposed or implemented in response to USI A46;interaction concerns were also noted.For St.Lucie-2, findings related to interaction concerns, but no outliers were noted.Enhancements for IPEEE-only components,(i.e., components outside the normal scope of USI A46, but within the scope of IPEEE)were not addressed.

In addition, containment performance evaluation and evaluation of human actions were not included as part of the licensee's treatment of seismic IPEEE concerns.The noted conditions, and proposed safety enhancements, are summarized below: Sr.Lucie Unit I During the walkdowns, five anchorages and the component cooling water surge tank platform were identified as concerns by the SRT.In addition to these five anchorage concerns, six additional anchorage concerns were identified by FPL for similar components in different equipment trains.The modifications undertaken for the identified concerns are described below:-The existing anchor bolts are corroded.The repair modification for the tanks involves the removal of all existing corrosion, application of protective coatings, installation of cover plates to enclose each anchor bolt pocket, and application of a filler material to protect the bolts from future corrosion.

I bi b i bkd The anchorage modification involves the addition of supplementary fillet welds along the interior of the cabinet base.-The cabinet is burned-through in areas near existing welds.The modification consists of screwing clip angles to the sides of the cabinet and anchoring the clip angles to the wall behind the cabinet, using expansion anchor bolts.-Fillet welds are missing because the existing embedded support channels are not properly located.The modification consists of installing plates to connect the base of the cabinets to the embedded channels.-Several bracing members required by the original platform design are missing.The modification requires installation of additional structural members, to increase lateral stiffness, and relocation of an instrument air line and three tank drain lines.j-This load center is for the pressurizer heater, and it consists of.three cabinets with weak anchorage.

The modifications include: (a)insuring that the cabinets are Energy Research, Inc.45 ERI/NRC 95-504

~f t adequately connected, so they will act as a single unit under seismic loading;and (b)adding fillet welds at the interior base of load center cabinets to anchor them to the embedded channels.The following seismic interaction concerns at Unit 1 were also noted: Potential interaction involving the glass site tube for the component cooling water surge tank.2.Potential interaction involving a block wall adjacent to the component cooling water surge tank.3.An overhead crane adjacent to the intake cooling water pump should be secured away from the pump.In addition, some cases of poor seismic"housekeeping" were observed and documented.

In response to the NRC's USI A46 review process, the licensee is implementing a program of strict seismic housekeeping.

St.Lucie Unit 2 Two seismic interaction concerns were noted: 1~Possible tipping of a cabinet near safety-related equipment; and 2.Questionable support of a component mounted above safety-related equipment.

Both of these issues were ultimately evaluated and resolved.HCLPF calculations were performed for a number of large storage tanks at Unit 2;these calculations demonstrated the capacities to be beyond the design basis.'concern was also noted pertaining to whether'or not the mounting of some internal coils in an energized transformer was seismically adequate.This concern was investigated during an outage, and it was found'hat the mounting was adequate.It was also stated in the seismic evaluation that a walkdown of wall transformers needed to be performed, to determine whether or not these transformers would need to be secured.In addition, the peer review resulted in the following additional findings which have been addressed:

Secure load center over-cabinet crane/winches, and verify that tool box cart in the switchgear room is either removed or properly secured.Secure or remove l&C locker from control room.Reduce battery rack end gaps on the 2A and 2B batteries.

Implement control-room housekeeping improvements regarding storage of Scott Air Packs and immobilizing an unsecured locker.Energy Research, Inc.-46 ERI/NRC 95-504 4.2 Overall, the licensee has concluded that there are no significant fire vulnerabilities at St.Lucie.With the exception of the control room, cable spreading room, and the"B" Switchgear room, all fire zones and areas were screened out based on 10~/ry core damage frequency criterion.

The core damage frequencies for fires in the control rooms were concluded to be 7.49x10'/ry and 5.90 x 10'/ry for Units I and 2, respectively.

For the cable spreading rooms, the core damage frequencies were evaluated to be 6.95 x 10'/ry and 5.64x 10'/ry for Units 1 and 2, respectively.

For both areas (i.e., control room and cable spreading room), the licensee has cited several conservative assumptions in fire occurrence rate and fire severity, and concluded that these two areas do not pose a vulnerability:

The core damage frequency of a fire in the"B" switchgear room was concluded to be 4.30x10'/ry and 4.48 x 10~/ry for Units 1 and 2, respectively.

Fire propagation modeling has been performed for this area, and the licensee has concluded that fire will not propagate throughout the room.The entire fire IPEEE effort, of course, has provided an excellent opportunity for the licensee's engineers to better learn about the characteristics of the plant, the plant behavior under different fire conditions, and the impact of human actions that are necessary to protect the reactor from any adverse effects.4.3 HEQXmds The IPEEE's overall conclusion regarding this category of external events is that any potential core damage scenario has an extremely low frequency in comparison with the frequency of core damage from other initiators.

As a result, no safety enhancements have been identified, and consequently, no commitments are made that would require tracking by the NRC.Energy Research, Inc.47 ERI/NRC 95-504 5 IPEEE EVALUATION AND DATA SUMMAI(Y SHEETS Completed data entry sheets for the St.Lucie Nuclear Plant IPEEE are provided in Tables 5.1 to 5.6.These tables have been completed in accordance with the descriptions in Reference[11].Table 5.1 lists the overall external events results.Table 5.2 summarizes general seismic data pertaining to the evaluation.

Tables 5.3 provides the Seismic Success Paths Overview Table, and Table 5.4 summarizes sequence information for PWR Seismic Success Paths.Accident sequence information provided in Tables 5.5 and 5.6 for fire events are only partially completed due to lack of sufficient information provided in the submittal[1]and Reference[14].Accident sequence tables are not provided for HFO events, since no PRA analyses were performed for these events.Energy Research, Inc.48 ERI/NRC 95-504 Tahle 5.1 External Events Results Plant Name: Event External Fire Screening 0 CDF Plant HCLPF(g)Notes External Flooding 0 Extrcme Winds 0 Internal Fire Nearby Facility Accidents 0 Seismic Activity Transportation Accidents 0 Others 0 Hazardous chemicals and lightning Scrccning:

S=Plant specific analysis;0=Screened out;SO=Bounding analysis Energy Research, Inc.49 ERI/NRC 95-504 1

Table 5.2 SSM Seismic Fragility Plant Name:~Mph Review Level Earthquake (g): Spectral Shape: 'ig'NUREG4098, NRC Guide 1.60, 10,000 year LLNL median UHS, Site Specific, or other)List components and equipments which do not meet RLE (all components) or with lowest HCLPF (less than 10): Component'ondensate Storage Tank (Unit 2)RWST (Unit 2)Diesel Oil Storage Tank (Unit 2)Boric Acid Makeup Tanks (Units 1 and 2)HCLPF (g)>0.3g (but (0.49g)>0.3g (hut (0.64g)1.47g Seismic Sequence Description Seismic Success Path Description Not all lowest HCLPFs were reported;reduced-scope evaluation.

Energy Research, Inc.50 ERI/NRC 95-504 Table 59 PWR Success Path Overview Table Plant Name:~+ggji Sequence Success Path PDS HCLPF (L)'nit Event T-LOOP Success Supports EAC, CCW, ESW Non-Failed Functions Attributes jttjLatssLjjttjtjsxtn:

one of tbe (oasis)nst si, S2.s3, A, v[.ss), T.Loot', T-Rx, T-Tf, T-hlvrs, T.UHs, T-RcP.T.LNMv.T.LMptv, T.Expttr.T QBoc.T<LBIc, T~T~Rvlloav, TWCI, T-(other), or T-(Support System)(.xs)refers to optional suppkmentaty mater(al.QSISI~sSISBt ht most teo of tbe foaesdeS: AC.ACBUI.ACSV2, ACBU3, AUXC~A VXC3.AUXGI.CCIVI DC, EAC, EDC, ESAS I, ESAS2, ESttr, HVAC I, HVAC2.HVAC3, IA.NIT, OA3, OA4, SA, SIM, StV2, SNr3, S)Vd, VAC (p)aid may be blast).tt~tjjDLK)ttt(3jst)tt ht most three of the foaosrbtf t sINT, sDEp, ssMU, RcssoR.Rcs!NT.RcsDEp, Hpl, HPR, LPI.LpR, cpsi, cpsR clp.YENT(lf s 4th and/or 3th sie neoessary, use the Notes rield)iBBjj3aut At most tbsee of tbe foaotshtf t ATWS, SYPASS, 11L, IN D4GIR, SBO, OR HUM (Vald msy be bien))Redused-Seeps Plane no HCLPP ra pash)et ispotted.Energy Research, Inc.51 ERI/NRC 95-504 Tahle 5.4 PWR Seismic Success Paths Plant Name: C H A L L E N G E S T R A T F.G Y S U C C E S P A'r H PRIMARY INTEGRITY P P R A A C D D P I 2 S R B P P P I P S S 0 R R V V C H L A Iiprc P I I C I A I I PRIMARY INVENTORY-INJECllON A I 2 C H P R H L A P P R R R I A R 2 PRIMARY INVENTORY-

.RECIRC SECONDARY INTHiRITY S S T G G T S A M S I V SECONDARY INVENTORY T S M B Gi F W N I S P A A A F M M W I 2 M 3 C C F S S C I 2 I F C 2 I C C C I I I 2 CONTAINMENT I (I N R F NOTES T.Loop gggggg One of thc folk stg Sl.S2.S3.A.V(xx).T LOOP T RX, T Tr.T A1WS.T UHS.T RCP.T.LNMII.T MFW T EXFW T-SUIOC, TCUIIC.T.SOT)A T-SORY/IORV, T~l.T-(Other).

OR T.(Support System).(.xx)refers to optional suppkmcntary material.Acronym of Support Systems: AC, ACBUI, ACB'II2, ACBU3.AUXC2.AUXC3.AUXC4, CCW.DC.EAC, EDC, ESASI, ESAS2.ESW.IIVACI.HVAC2, HVAC3.IA.NIT, OA3, OA4, SA, STM.SW2, SW3.SW4, VAC I,2,3...How many needed to operate H w Human actkn required T=Must bc thmttkdlcontrollcd For Core Damage Prevention Challenges, show only hanlware whose failure is modckd as contributing to core damage.Energy Research, Inc.I 52 ERI/NRC 95-504 C.Table 5.5 PWR Accident Sequence Overview Table Plant Name: in i i For Fire PRA Only I h f Sequence Unit I-Control Room PDS CDF 7.49 x 10-s/I nit.Event Lost Supports Failed Functions Attributes Unit I-Cable Spreading Room 6:96 x 10'/ry Unit I-B Switchgcar Room 4.30 x 10's/ry T-RX (inferred)

SSMU Unit 2-Control Room 5.90 x 10 s/ry HUM Unit 2-Cable Spreading Room 5.64 x 10'/ry Unit 2-B Switchgcar Room 448x 10s/ry T-RX (inferred)

SSMU, HPI HUM, TIL (infcrrcd)

FA-121/51W 2.67 x 10'/ry T-RX (infcrrcd)

SSMU HUM, TIL (infcrrcd).

FAW 1.34 x 10 s/ry T-RX (inferred)

EP SSMU HUM, TIL (inferred)

TQO!her), or Tgsupport System)(-xx)refers to optional supplementary inataiat.r Idtst SutltRxt<At most two oF the following:

AC, ACBU1, ACBU2, ACBU3, AUXC2.AUXC3, AUXC4, CCW, DC, EAC, EDC, ESAS1, ESAS2, ESW, HVACI, HVAC2, HVAC3, IA.NIT, OA3, OA4, SA, STM, SW2, SW3, SW4, VAC (Fieldmay be blank).~aikd utldsistn:

At most thfeeofthe following:

SINT, SDEP, Sshfu, RC<BOR, RCS-INT, RCS-DEP, HPI, HPR.LPI, LPR.CPSI, CPSR, CIF, VENT(lf a 4th and/or 5th are necessary, use the"Notes" Iield)~luil I:Alatadlhr flh f llaaiaSATASS,BYPASS,TII RI~GTR.SBO,ORIIIMIFi idaaayhahla&I Energy Research, Inc.53 ERI/NRC 95-504 t q t Table 5.6 PWR Accident Sequence Detailed Table Plant Name: For Fire PRA Only PRIMARY INTEGRITY PRIMARY INVENTORY-INJECTION PRlhIARY INVENTORY-RFXIRC SECONDARY INTEGRITY SECONDARY INVENTORY CONTAINMENT SEQUENCE Unit I-Control Room Unit I-Cable Sprctalina Room Unit I~8 Switchgcar Room.Unit 2-Control Room Unit 2-Cable Sprcral ina Room R P S P P P P S A 0 R D R V I V 1 2 2 1 1 2 P R C H A C H P D P I'2 S I?1 L P I A A C I C A I 2 C H H P P R R 1 7 2 2 7 2 AAS R R G 2 S S G A T M T T S 8 I V S G M N A F I F W S W P X X X X X X X X A M I A A C M M S 2 3 I C S 2 F F I C C C I 2 C I I C I R I G F 2 N H fl M X NOTES linit'2-8 Switchacar Room 7 FA.I2 I IS I W FA.O X X I Energy Research, Inc.54 ERI/NRC 95-504 6 REFERENCES"Individual Plant Examination of External Events for St.Lucie Units 1 and 2," Florida Power and Light Company, December 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, June 1991.'."Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors," U.S.Nuclear Regulatory Commission Generic Letter 87-02, February 1987.5."Plant Specific Seismic Adequacy Evaluation of St.Lucie Unit 1 to Resolve Unresolved Safety Issue (USI)AM and Generic Letter 8742," prepared for Florida Power and Light Company by Stevenson and Associates, February 18, 1992.6."Evaluation of St.Lucie Unit 1, and Turkey Point Units 3 and 4, Resolution of Unresolved Safety Issue (USI)AA6, Supplement Number 1 to Generic Letter 87-02," U.S.Nuclear Regulatory Commission, Staff Evaluation Report, June 20, 1995.7."Supplemental Safety Evaluation Report of St.Lucie Unit 1 and Turkey Point Units 3 and 4, Resolution of Unresolved Safety Issue (USI)A-46," U.S.Nuclear Regulatory Commission internal memorandum from Richard H.Wessman, Chief of the Mechanical Engineering Branch, to Herbert N.Berkow, Director for Project Directorate II-2, September 19, 1996.8."Standard Review Plan for the Review of Safety Analysis Report for Nuclear Power Plants, LWR Edition," U.S.Nuclear Regulatory Commission, NUREG-0800 (formerly issued as NUREG-75/087), June 1987.9.R.T.Sewell, et al.,"Individual Plant Examination for External Events: Review Guidance," ERI/NRC 94-501 (Draft), May 1994.10."IPEEE Step 1 Review Guidance Document," U.S.Nuclear Regulatory Commission, June 18, 1992.11.S.C.Lu and A.Boissonnade,"IPEEE Database Data Entry Sheet Package," Lawrence Livermore National Laboratory, December 14, 1993.12."plant Specific Seismic Adequacy Evaluation of St.Lucie Unit 2 to Resolve Generic Letter 88-20, Supplement 4," prepared for Florida Power and Light Company by Stevenson and Associates, January 31, 1995.I I"Individual Plant Examination Report for St.Lucie Units 1 and 2," Florida Power and Light Company, April 1992.Energy Research, Inc..55 ERI/NRC 95-504 I 14."St.Lucie Units 1 and 2, Docket No.50-335 and 389, IPEEE-Request for Adtlitionai Information Response-Generic Letter 88-20, Supplement 4," letter to U.S.Nuclear Regulatory Commission, from D.A.Sager, Florida Power and Light Company, dated January 9, 1996.15.Layout drawings from a fireprotection report for St.Lucie Plant Unit 1;Figure Numbers 9.5A-1 through 9.5A-6, Amendment 12, December 1993.16.17.Layout drawings from a fire hazard analysis report for St.Lucie Plant Unit 2;Figure Numbers 9.5A-I through 9.5A-6, Amendment 8, September 1993."Masonry Wall Design," U.S.Nuclear Regulatory Commission, IE Bulletin 80-11, 1980.18."Design Response Spectra for Seismic Design of Nuclear Power Plants," U.S.Atomic Energy Commission, Regulatory Guide 1.60, Revision 1, December 1973.19."Generic Implementation Procedure (GIP)for Seismic Verification of Nuclear Plant Equipment," Seismic Qualification Utility Group (SQUG), Revision 2, February 14, 1992.20."A Methodology for Assessment of Nuclear Power Plant Seismic Margin," Electric Power Research Institute, EPRI NP-6041-SL, Revision 1, August 1991.21."Use of Seismic Experience Data to Show Ruggedness of Equipment in Nuclear Power Plants (Draft)," Senior Seismic Review Advisory Panel, January 17, 1990.22."Fire Induced Vulnerability Evaluation (FIVE)", Electric Power Research Institute, TR-100370, Revision 1, September 1993.23."St.Lucie Probabilistic Risk Assessment," prepared for Florida Power and Light Company by PLG, Inc., PLG4637, July 1988.24."Design Basis Tornado for Nuclear Power Plants," Regulatory Guide 1.76, April,1974.-

25."Winds Forces on Structures," Transactions of the American Society of Civil Engineers, Vol.126, Part II, ASCE Paper No.3269, 1961.26."Building Code Requirements for Minimum Design Loads in Buildings and Other Structures," American National Standards Institute, ANSI A58.1, Committee A58.1.27."Technical Basis for Interim Regional Tornado Criteria," U.S.Atomic Energy Commission, WASH-1300, May 1974, 28.W.R, Cramond, et al.,"Shutdown Decay Heat Removal Analysis of a Combustion Engineering'-Loop Pressurizer Water Reactor," Sandia National Laboratories, NUREG/CRR710, August 1987.29."Evaluation of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plant Sites," Regulatory Guide 1.91.Energy Research, Inc 56 ERI/NRC 95-504

30."Individual Plant Examination of External Events for Turkey Point Units 3 and 4," Florida Power and Light Company, June 1994.31."Sta6'Guidance of IPEEE Submittal Review on Resolution of Generic or Unresolved Safety Issues (GSI/USI)," U.S.Nuclear Regulatory Commission, August 21, 1997.Energy Research, Inc.57 ERI/NRC 95-504