ML20079B046

Individual Plant Exam for External Events

ML20079B046
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Site:	Waterford
Issue date:	12/31/1994
From:	ENTERGY OPERATIONS, INC.
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NUDOCS 9501040254
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{{#Wiki_filter:l i i WATERFORD 3 INDIVIDUAL PLANTEXAMINAT10NFOR i EXTERNAL EVENTS l Prepared and Submitted by ENTERGY OPERATIONS, INC. December 1994 9501040254 941229 DR ADOCK 0500J 2

-t s 'l .s TABLE OF CONTENTS 1. Executive Summary 1-1 1.1 Background and Objectives 1 1.2 Plant Familiarization 1-1 1.3 _ Overall Methodology 1-3 - 1.4 Summary ofMajor Findings 1-3 2. Examination Description 2-1 2.1 Introduction 2-1 2.2 Conformance with Generic Letter and Supponing Material 2-1 2.3 GeneralMethodology 2-2 2.4 Information Assembly 2-3 3. Seismic Analysis 3-1 3.0 Methodology Selection 3-1 3.1 Seismic Margins Method (Reduced Scope) 3-1 3.1.1 Review of Plant Information, Screening, and Walkdown 3-2 3.1.2 Systems Analysis 3-18 3.1.3 Analysis of Structure Response 3-27 l 3.1.4 Evaluation of Seismic Capacities of Components and Plant 3-33 3.1.5 Analysis of Containment Performance 3-40 3.2 USI A-45, GI-131, and Other Seismic Safety Issues 3-40 5. High Winds, Floods, and Others 5-1 i 5.1 High Winds 5, i 5.2 Floods 5-4 5.3 Transponation and Nearby Facility Accidents 5-9 i 5.4 Others 14 6. Licensee Participation and Internal Review Team 6-1 6.1 IPEEE Program Organization 6-1 6.2 Composition of Independent _ Review Team 6-2 6.3 - Areas ofReview and Major Comments 6-2 6.4 Resolution of Comments 6-2 7. Plant Improvements and Unique Safety Features 7-1 ii -.-->-a--- ---.--------.-._--------.----.---a-----_

8. Sununary and Conclusions (including proposed resolution 8-1 ofUSIs and GIs) i I t t t I i iii

1. EXECUTIVE

1.1 BACKGROUND

AND OBJECTIVES In November of 1988, Generic Letter 88-20 on Individual Plant Examination (IPE) (see Reference 1-1) was issued by the NRC to address severe accident risk because ofinternal events, including internal floods. Waterford 3 responded to this initiative with a Level 1 and limited scope Level 2 Probabilistic Safety Assessment, the results of which were subndtted as the Waterford 3 IPE in keeping with the requirements of Generic Letter 88-20 and the guidance of NUREG-1335 (see Reference 1-2). In June of 1991, Supplement 4 to Generic Letter 88-20 (see Reference 1-3) was issued requesting each licensee to perform an Individual Plant Examination of External Events (IPEEE) to address the severe accident risk posed by external events. External events include seismic events, internal fires, high winds and tornadoes, external floods, and transportation and nearby facility accidents. The purpose of the IPEEE is similar to that of the IPE: (1) to develop an appreciation of severe accident behavior. (2) to understand the most likely severe accident sequences that could occur. (3) to gain a qualitative understanding of the overall probability of core damage and fission product releases. (4) if necessary, to reduce the overall probability of core damage and fission product releases by modifying hardware and procedures that could help prevent or mitigate severe accidents. With the exception of fire events, the Waterford 3 IPEEE project has been completed and the objectives of the IPEEE have been met. This document summarizes the evaluation of the external events, excluding fire, and constitutes the Waterford 3 IPEEE submittal. (Fire events are treated separately from this document.) The guidance provided in NUREG-1407 (see Reference 1-4) has been used to prepare this document. 1.2 PLANT FAMILIARIZATION The Waterford 3 Nuclear Power Plant is a Combustion Engineering designed pressurized water reactor located on the Mississippi River west of New Orleans, Louisiana. The rated thermal power level is 3410 MWt with a gross electrical output of 1153 MWe. Commercial operation l began in the Fall of 1985. k 1 I Page 1-1 z

The NSSS is designed with two U-tube steam generators and four reactor coolant pumps. Nominal reactor coolant system (RCS) pressure is 2250 psi and average RCS temperature is 5740F at full power. Overpressure protection is provided by two safety reliefvalves on the pressurizer that lift at about 2500 psi and discharge to a quench tank. The Waterford 3 design does not include power operated relief valves. Engineered Safety Features Systems at Waterford 3 are typically divided into two separate and independent trains. Two high pressure safety injection (HPSI) pumps (discharge pressure of about 1400 psi), plus an installed spare pump that can be manually aligned to either train are provided. Each HPSI pump injects into all four RCS cold legs. Realignment of suction for the liPSI pumps to the safety injection sump in containment occurs automatically on a low refueling water storage pool level signal. Four Safety Injection Tanks (SITS) discharge into the RCS at a pressure of 600 psi. Two low pressure safety injection pumps (discharge pressure of about 200 psi) provide a high flow rate of safety injection water at low pressure. These pumps also provide shutdown cooling flow. Feedwater to the steam generators is provided by two steam turbine driven main feedwater pumps. Feedwater flow is automatically runback to about 5% full flow on reactor trip. Emergency Feedwater consists of two motor driven pumps (flow rate of about 350 gpm) and one steam turbine driven pump (flow rate of about 700 gpm). A large volume ofwater (well in excess of a 24 hour supply) is available for emergency feedwater. A non-safety auxiliary feedwater pump that is normally used during reactor startup is available as a backup feedwater source. Condensate pumps can also be used for feedwater if the steam generators are depressurized below the condensate pump discharge pressure. Offsite power from the utility grid comes into the switchyard from two independent transmission lines thmugh two startup transformers. During normal operation, the plant receives power from the main generator through two unit auxiliary transformers. When necessary, onsite AC power is provided by two independent emergency diesel generators. Equipment heat loads are removed by a closed Component Cooling Water (CCW) System which rejects heat to the atmosphere through forced air cooling towers. Evaporative cooling towers provide supplemental heat rejection capability when atmospheric or accident conditions warrant. The containment structure is a large, dry free standing cylindrical steel vessel surrounded by a separate reinforced concrete shield buildint / ny water entering the containment flows directly to the containment sump which quickly overflows to the reactor cavity. Thus, the Waterford 3 reactor cavity is almost always wet. Two containment spray pumps control containment pressure during an accident and remove heat from the containment by pumping sump water through the containment spray / shutdown cooling heat exchangers. Four containment fan coolers also remove heat from the containment atmosphere to the CCW system. Page 1-2

i 1.3 OVERALL METHODOLOGY The methodology employed for the IPEEE was consistent with the guidance in NUREG-1407. For seismic events, the reduced-scope seismic margins analysis (SMA) was used, in accordance with Section 3.2 of the NUREG. The evaluation of other external events (high winds, floods, and transportation and nearby facility accidents) was performed using the screening approach described in Section 5 of the NUREG. The methodology used in performing the IPEEE is described in more detail in Section 2.3 of this report and completely in the sections for each external event (Sections 3 and 5). 1.4

SUMMARY

OF MAJOR FINDINGS 1.4.1 Seismic Waterford 3 used an IPEEE reduced scope Seismic Margins Analysis (SMA) that concentrated on walkdowns to identify potential seismic vulnerabilities for equipment, large tanks, distribution systems, and structures. No seismic vulnerabilities were identified. The walkdowns resulted in no outliers that are operability issues at the plant. However, there were three unresolved issues at the completion of the walkdowns. These issues are not significant to seismic risk and are being made to conform with standard practice in seismic design. The issues, proposed resolution, and schedule follow: Issue Proposed Resolution Schedule Loose items in the Remove or restrain loose items in the Complete a modification Control Room vicinity of safety-related cabinets package by February 15,1995 Station air pipe not Formally evaluate the reasons why Complete by March 30,1995 meeting clearance the existing condition is acceptable. requirements Storage of temporary Revise Transient Combustibles and Complete by April 1,1995 l equipment Designated Storage Areas procedure 1.4.2 High Winds, Floods, and Transportation and Nearby Facility Accidents The IPEEE found no high winds, floods, or off-site industrial facility accidents that significantly alters the Waterford 3 estimate of either the core damage frequency, or the distribution of containment release categories. The IPEEE concludes that the plant is in conformance with the 1975 SRP that penains to high winds, on-site storage of hazardous materials, and off-site developments. Page 1-3

.= l.4.3 Proposed Resolution Of USIs And GIs The stated purpose of Unresolved Safety Issue (USI) A-45 (see Reference 1-5)is to " evaluate the adequacy of current designs to ensure that LWRs do not pose unacceptable risk as a result of DHR [ decay heat removal) system failures." No DHR vulnerabilities were found for seismic, high wind, flood, and nearby facility accident events. Therefore, USI A-45 should be considered resolved for Waterford-3 with respect to these external events. Generic Issue (GI) 131 is not applicable, since Waterford-3 is a Combustion En6 Hng pard Waterford-3 is not a USI A-46 plant, so USI A-46 is not applicable. The issue npaial interaction, however, has been addressed as part of the reduced scope SMA. The Waterford 3 IPEEE has not been used to evaluate any other USIs or GIs. i REFERENCES 1-1. Generic Letter No. 88-20, " Individual Plant Examination for Severe Accident i Vulnerabilities - 10 CFR 50.54(f)," USNRC, November 23,1988. 1-2. NUREG-1335, " Individual Plant Examination: Submittal Guidance," USNRC, August 1989. 1-3. Generic Letter No. 88-20, Supplement 4, " Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f)," USNRC, June 28,1991. l-4. NUREG-1407, " Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," USNRC, June 1991. 1-5. NUREG-1289, " Regulatory and Backfit Analysis: Unresolved Safety Issue A-45, Shutdown Decay Heat Removal Requirements." k Page 1-4

m i 2. EXAMINATION DESCRIPTION

2.1 INTRODUCTION

For Waterford 3, the plant specific examination requested by the IPEEE Generic Letter, Supplement 4, except the fire study, has been carried out. The following sections provide a description of the Waterford 3 IPEEE, presented first from the i perspective of how the examination conforms with the IPE Generic Letter, next from the perspective of technical organization and methodology, and then from the perspective of the information assembled to cany out the examination. 2.2 CONFORMANCE WITII GENERIC LETTER AND SUPPORTING MATERIAL NRC Generic Letter 88-20, Supplement 4, states that five external events should be assessed: seismic events, internal fires, high winds and tornadoes, external floods, and transponation and nearby facility accidents. The Waterford 3 IPEEE has assessed these events, with the exception of fire events. Supplement 4 further specifies the acceptable methodologies: Seismic l Generic Letter 88-20, Supplement 4, specifies: Seismic PSA, NRC Seismic Margins Method (SMM), or EPRI SMA (through the walkdown phase for reduced scope plants). The Waterford 3 IPEEE used the EPRI SMA through the walkdown phase and meets the generic letter requirement. High Winds. Floods. and Transportation and Nearby Facility Accidents Generic Letter 88-20, Supplement 4, describes a screening type approach. The Waterford 3 IPEEE usd the specified screening approach. In perforr g the IPEEE at Waterford 3, the guidance provided by GL 88-20, Supplement 4, and NUREG-1407 was considered. The examination process not only met the Waterford 3 and IPEEE objectives for severe accident assessment, but also provided the framework for effective participation of the Waterford 3 staff. The NRC encouraged utility staff participation in IPEEE preparation. With the exception of the seismic evaluation, all of the IPEEE was performed by Waterford 3 staff. The seismic evaluation utilized the expenise of a recognized seismic consultant. Utility personnel prepared the seismic safe shutdown equipment list, participated in the walkdowns, and provided detailed review and comment to the seismic consultant. This ensures that knowledge and skills gained during the Page 2-1

evaluation would be retained in-house so insights and lessons learned could be incorporated into plant procedures and programs more expeditiously. The Waterford 3 IPEEE was put into an independent peer review. The seismic peer review was performed by seismic experts at the seismic consulting firm who were not involved in the IPEEE evaluation. The evaluation of other events were independently reviewed by Waterford 3 personnel. These reviews ensured the accuracy of the IPEEE process and its results. 2.3 GENERAL METIIODOLOGY 2.3.1 Seismic Events Waterford 3 developed and implemented a program to satisfy requirements of the IPEEE seismic evaluation. The program implemented an IPEEE reduced scope Seismic Margins Analysis (SMA) that concentrated on walkdowns to identify potential seismic vulnerabilities for equipment, large tanks, distribution systems, and structures. The basic requirement for walkdowns was that the equipment, tanks, distribution systems, and structures be able to withstand the design basis Safe Shutdown Earthquake (SSE) at the plant and still provide its safe shutdown function. The SMA uses primarily EPRI report NP-6041-SL as guidance, which is not overly prescriptive but relies on the judgment of an experienced team to meet the basic requirement. A Safe Shutdown Equipment List (SSEL), using safety and non-safety-related components, was selected for achieving and maintaining plant shutdown in accordance with plant operating procedures. The SSEL also included items that are potential seismic-induced fire and seismic-induced flood sources within the plant. Seismic Verification Data Sheets that included each equipment item of the equipment list were developed. These sheets contain walkdown observations as well as screening results. There were three walkdowns performed: the Train "B" on-line walkdown during November of 1993, the Train "A" on-line walkdown during 1994, and the outage walkdown during March 1994. 2.3.2 Iligh Winds, Floods, and Transportation and Nearby Facility Accidents The IPEEE used the screening approach described in Generic Letter 88-20, Supplement 4. Waterford 3 seatched for significant changes in the probability for high winds, floods, or off-site industrial facility accidents. The original licensing action for Waterford 3 considered all manner of winds, floods, and industrial accidents. The focus of this part of the IPEEE submittal is: "does the plant meet acceptance criteria listed in the 1975 SRP in terms of high winds, on-site storage of hazardous materials, and off-site developments?" The originallicensing action for Waterford 3 used the 1975 SRP as a basis for finding Waterford 3 acceptable in terms of external hazards with i Page 2-2 i

I a few exceptions that the NRC reviewed and accepted. The exceptions are technical rather than substantive, e.g., the SRP recommended technique for calculating tornado loading on the shield t building was not appropriate given the shallow dome roof of the shield building. .l The Waterford 3 staff reviewed the assumptions in FSAR Chapter 2 with respect to external event initiating frequency. Whenever that frequency is explicit in the FSAR, it is compared to the IPE Level 1 initiating event frequencies. If the external event frequency were small compared to the related Level 1 initiating event frequency, then the external event has an insignificant affect on our estimate of both the core damage frequency and distribution of containment release categories. .i When the external event initiating frequency is indeterminate, the review of"High Winds, Floods, and Others" describes the plant design features and related basis with respect to external events. Note, the external events are typically a subset of the Level 1 initiating events. For example, l storm damage that causes a loss of off-site power is part of the Level 1 assumption behind the transient initiator T5. } As a further review of plant specific hazard data and licensing bases, the Waterford 3 staff also l qualitatively reviewed the external events postulated in FSAR Chapter 2 against spectacular i events in southeast Louisiana. We sought assurance that the postulated events bounded the spectacular events since initial plant startup. j The final part of the review method re-visited the 10 CFR 50.59s written since initial plant start-i up regarding changes that exposed the plant to new external hazards, e.g., a new hydrogen gas pipeline across the site. The probability and consequences of the new configuration is qualitatively related to either FSAR Chapter 2 assumptions or IPE assumptions. Insignificant changes in this group are those that change neither the estimate of the core damage frequency, nor the distribution of containment release categories. f 2.4 INFORMATION ASSEMBLY At the start of the project, a list of specific information needed to begin the IPEEE was established. A list of some of the information assembled includes: Final Safety Analysis Report Piping and Instrumentation Drawings General Arrangement Drawings System Design Basis Documents Plant Procedures i Station Information Management System (SIMS) Component Database. l I Page 2-3 i

~ l Safe Shutdown Equipment List Controlled drawings and procedures, that are updated after plant modifications, were u' sed by the. l IPEEE team. This ensures that the most recent and accurate information on plant configuration and operation was incorporated. Plant walkdowns were performed for the seismic evaluation to ensure that the evaluation represents the as-built plant. The evaluation ofother external events included a review of plant changes since the original license issuance. Finally, coordination ofIPEEE activities was facilitated by giving 'overall project responsibility and oversight to a single group. A key area of coordination was for seismically induced fires. Here, the seismic analysts worked together with fire protection engineering personnel to ensure that seismic-induced fire interactions were addressed in the evaluation of seismic events. i i l l i i I i l l l 1 i Page 2-4

3. SEISMIC ANALYSIS 3.0 METHODOLOGY SELECTION .Waterford 3 Nuclear Station is classified a reduced scope plant as defined NUREG-1407 based on the low seismicity. Therefore, a seismic review of the plant was performed to the plant's original design basis. This was accomplished by performing a Seismic Margins Assessment (SMA) of the Safe Shutdown Equipment List (SSEL) with plant walkdowns in accordance with the guidelines and procedures documented in Electrical Power Research Institute (EPRI) Report NP-6041-SL. Since Waterford 3 Nuclear Statioa is a reduced scope plant, the original design basis Safe Shutdown Earthquake (SSE) ground response spectra and corresponding in-structure response spectra were used as the Review Level Earthquake (RLE) input for the walkdown and evaluation, as requested by NUREG-1407. No new in-structure response spectra were developed and those described in the Waterford 3 Nuclear Station Final Safety Analysis Report (FSAR) were utilized. Safe shutdown success paths were developed to identify the systems that must function to successfully shutdown and cool the reactor following the occurrence of a SSE. A safe shutdown success path is a string of systems which is used to accomplish all of the required safe shutdown functions. 3.1 SEISMIC MARGINS METHOD (REDUCED SCOPE) In the Commission policy statement on severe accidents in nuclear power plants published August 8,1985 (50 FR 32138), the Commission concluded, based on available information, that existing plants pose no undue risk to the public health and safety and that there is no present basis for immediate action for any regtdatory requirements for these plants. However, the Commission convinced itself, based on NRC and industry experience with plant-specific probabilistic safety assessments (PSAs), of the need for a systematic examination of each existing plant to identify any plant specific vulnerabilities to severe accidents. For the seismic evaluation, two alternative methodologies are acceptable to identify potential seismic vulnerabilities. The first is a Seismic Probabilistic Safety Assessment (SPSA). The second is one of the Seismic Margins Assessments (SMA) described in

References:

"The NRC method" and "The EPRI method" Waterford 3 chose to implement the EPRI Seismic Margins Method option as appropriate for a reduced scope plant as defined in NUREG/1407F to satisfy the IPEEE Seismic evaluation.

The NRC has also anticipated the coordinated handling of the IPEEE with completion of other related issues such as USI A-45, " Shutdown Decay Heat Removal Requirement." The influence of the coordinated issues on the Waterford 3 response to the seismic portion of the IPEEE is included in the SMA for this project. The reduced scope SMA consists of developing a Safe Shutdown Equipment List (SSEL) using the EPRI SMA methodology (see Reference 3.1). The seismic review of the plant used the plant's original design basis. The plant walkdown is the critical element of the SMA. The screening Page 3-1

.Y, ) methods and criteria described in EPRI NP-6041-SL as appropriate for a reduced scope SMA are - part of this report. 3.1.1 Review of Plant Information, Screening, and Walkdown 3.1.1.1 General Plant Description The Waterford-3 site is on the west (right descending) bank of the Mississippi River near Taft, Louisiana in the nonhwest portion of St. Charles Parish. About three miles westward is the eastern boundary of St. John the Baptist Parish. The coordinates for the reactor are 29 59' 42" north latitude, and 90' 28' 16" west longitude. [ The site consists of over 3,000 acres of flat land extending from the Mississippi River to the St. Charles Drainage Canal. The site includes about 7500 feet ofriver frontage. About 3,000 feet back from State Road 18, adjacent to the levee, the Missouri Pacific Railway crosses the width of the property. The plant area is on a raised final grade of +17.5 fl. MSL around the Nuclear Plant Island Structure, and +14.5 ft. MSL around the Turbine Building. Structures housing safety-related equipment are flood protected to elevation +29.25 ft MSL. 3.1.1.2 Site Geology The geologic studies of the site and surrounding area were based on interpretations of geologic literature, geologic maps, topographic maps, remote sensing data, surface mapping, subsurface borings, geophysical reflection and refraction surveys, geophysical logs and laboratory tests. The site and surrounding area lie within the Mississippi River Deltaic plain physiographic province. The deltaic plain has a flat topography near sea level, with extensive areas covered by water, swamp, or marsh. In the site and surrounding area, the physiography is dominated by the present Mississippi River. The site is on the outside or eroding bend of the river, between miles 129 and 130 Above Head of the Passes. At the site, the Mississippi River has a maximum depth of about 110 feet and is 2200 feet wide. l The site is almost entirely upon the natural levee of the Mississippi River. The southwest portion of the property, about two miles southwest of the plant site,is fresh-water swamp adjacent to the natural levee. The surface elevations of the natural levee on the property range between near sea level in the southwestern portion to about 14 ft. MSL near the river, at the base of the man-made, flood-control levee. The crest of the Mississippi River flood-control levee, which is the highest point on the site, is about +30 ft MSL. The lowest elevations on the site occur in the swamp at the southwestern end of the property. In this area, elevations are one to two feet above sea level. 4 1 l Page 3-2

i The geologic structures that exist near the site developed in thick sedimentary sequences. They consist of non-tectonic structures associated with salt and clay mobilization and growth faults associated with sediment instability at the shelf edge. Faulting in the site and surrounding area was thoroughly investigated by analyzing existing subsurface data, including the spontaneous potential and resistivity logs of 151 oil wells; the records of ten deep seismic reflection lines; and the analysis of various published and unpublished geologic maps. Analysis of the electric logs of the oil wells allowed the identification of several continuous (across the site area) and correlatable marker horizons. On the basis of these correlations, several graphic interpretations were derived including locations of the sedimentary stmetures near the site. Geologic sections using selected oil wells show buried stmetures and the attitude of bedding. The structures within five miles of the site include the west flank of the Good Hope salt dome and its associated faulting and seven growth faults that are a portion of the fault trend designated as the Grand Chenier fault system in western Louisiana. The nearest salt dome to the site is the Good Hope dome that is centered about six miles east of the site. It is a piercement type of salt dome buried by 9580 ft. of Miocene and younger sediments. The sediments overlying the dome were uplifted and were faulted by the rising salt mass during the period ofits development. Salt dome uplin ceased during the Miocene epoch (5.5 to 22.5 Million Years Before Present). Active petroleum production is occurring from numerous wells drilled in the uplifted and faulted dome. The excavation for the Waterford-3 seismic Category I structural mat was cut 60 ft deep to approximately elevation -48 ft MSL. The excavation, which exposed the upper several feet of the Pleistocene Prairie formation over an area 380 ft by 267 ft. was mapped in detail. In the excavation, the Prairie formation at foundation level consists of horizontally bedded layers ofsilts and clays. The conditions encountered compare very favorably with the data taken from site borings. Mapping of the excavation disclosed no anomalies or discontinuities that may adversely affect the integrity of the foundation materials. In performing the geologic studies of the site and surrounding areas, various forms of remote sensing data supplement surface and subsurface data. The purpose of this analysis was (1) to aid in the interpretation of depositional history of this portion of the Mississippi River deltaic plain and (2) to determine whether surficial expression exists for any of the deep underlying geologic structure. Laboratory testing determined the engineering properties of the detailed descriptions of the soil types encountered by the site boring program. No zones of alteration or irregular weathering exist in the site area. Over 40,000 ft of mostly unconsolidated sediments lie above the crystalline basement rock beneath the site. No unrelieved residual stresses exist in the unconsolidated foundation m No materials exist at the site that could be unstable due to their mineralogy. Page 3-3

.s 1 1 -]

3.1.13' Nuclear Steam Supply System and Containment Structure The nuclear steam supply system (NSSS) is a pressurized water reactor system designed by j

' Combustion Engineering Incorporated.- The containment structure is ' a free-standing steel. i containment vessel surrounded by a reinforced concrete shield building all designed by Ebasco 1 Services Incorporated. .3.1.1.4 Major Structures i The major structures include the Reactor Auxiliary Building, Fuel Handling Building, Cooling i '~ Towers, and Containment Building. 1 t 3.1.1.5 Prin_qioal Design Criteria Principal stmetures, systems and equipment that may serve either to prevent accidents or to. j mitigate their consequences are designed and are erected in accordance with applicable codes to + withstand the most severe earthquakes, flooding conditions, windstorms, temperature and other deleterious natural phenomena that may occur at the site during the lifetime of the plant. Principal structures, systems and equipment are sized for the design power level of the nuclear supply. system output, i.e.~, 3390 Mwt. i' Redundancy in the reactor protective and safety feature systems means no' single failure of any active component of the system can prevent action necessary to avoid an unsafe condition. The i plant design facilitates inspection and testing of systems and components whose reliability are important to the protection of the public and plant personnel. The seismic Category I structures consist of the following: a) Reactor Building (comprising a free standing steel containment vessel, a containment internal structure and a reinforced concrete Shield. b) Reactor Auxiliary Building I c) Fuel Handling Building d) Component Cooling Water System Structure i All seismic Category I structures are in a common structure, the Nuclear Plant Island Stmeture. It is a rectangular box-like reinforced concrete stmeture 380 ft. long, 267 ft. wide and extending 64.5 fl. below grade. { t Page 3-4

i 3.1.1.5.1 Containment Design Criteria i The Containment System does not utilize a concrete containment. The primary containment is a free standing steel pressure vessel surrounded by a reinforced concrete Shield Building. The Shield Building is a seismic Category I structure. The containment vessel, including all its penetrations, is a low leakage steel shell that can withstand the postulated loss of coolant accident and can confine the postulated release of radioactive material. Systems directly associated with the containment vessel are the Containment Spray System, the Containment Cooling System, and the Containment Isolation System. The containment vessel is a cylindrical steel pressure vessel with hemispherical dome and ellipsoidal bottom. It houses the reactor pressure vessel, the reactor coolant piping, the pressurizer, the quench tank, the reactor coolant pumps, the steam generators, and the safety injection tanks. It is completely enclosed by the reinforced concrete Shield Building. An annular space between the walls and domes of the containment vessel and the concrete Shield Building permits construction operations and in-service inspection. The containment vessel is an independent free standing structure, rigidly fixed at its base near the elevation ofits bottom spring line. The containment vessel rests on a concrete base that was placed after the cylindrical shell and the ellipsoidal bottom were constructed and post weld heat treated. Both the Shield Building and the containment vessel rest on a common foundation mat. With the exception of the concrete placed underneath and near the knuckles at the sides of the vessel, there are no structural ties between the containment vessel and the Shield Building above the foundation slab. Therefore, there is virtually unlimited freedom for differential movement between the containment vessel and the Shield Building above the top of the concrete base at elevation -1.50 ft. MSL. Concrete floor fill was placed above the ellipsoidal shell bottom after the vessel has been post weld heat treated, to anchor the vessel. The cylindrical portion of the steel containment shell has a minimum thickness of 1.903 in. on an inside radius of 70 fl. The polar crane girder support plates are welded to the shellliner at approximately six fl. on center. Except for some miscellaneous platform framing and some minor seismic restraints, no major floor framing or seismic restraint supports are attached to the shell liner. Immediately below the crane girder, a heating and ventilating duct approximately 6'-6" wide x 8'-0" deep, running the entire containment circumference, is structurally supported and attached to the shell by means of welded clips. The containment shell also supponed temporary construction loads from the pedestal cranes. The 1.903 in. minimum shell plate thickness increases to a minimum of four in. adjacent to all penetrations and openings. The inside radius of the hemispherical dome is 70'-15/32 in. with a dome plate of 0.95 in. thick connected to the cylindrical ponion of the shell at the tangent line by means of a full penetration weld. The containment spray piping is attached to the dome by means of welded clips as are the dome inspection walkway and platforms. The Shield Building protects the containment vessel from external missiles. Protection from internal missiles is provided by the primary and secondary shield walls and other containment internal structures. Page 3-5

The function of the containment piping penetration assemblies is to provide for passage of process, service, sampling and/or instmmentation pipe lines into the reactor containment vessel, while maintaining the desired containment integrity and providing a leak-tight seal. The materials used for penetrations, including the personnel access air locks, the equipment access hatch, the piping and duct penetration sleeves and the electrical penetration sleeves conform with the requirements set forth by the ASME Boiler and Pressure Vessel Code. A 14 ft. diameter equipment hatch provides equipment access. This is a welded steel assembly, with a double gasketed flanged and bolted cover. Provision is made to pressure the space between the double gaskets to 44 psig. Two personnel air locks are provided. These are welded steel assemblies. Each lock has two double gasketed doors in series. Provision is made to pressurize the space between the gaskets. The doors mechanically interlock to ensure the one door cannot open until the second door seals. The containment vessel was designed, fabricated, erected, and tested in accordance with the requirements of Section III, Subsection NE of the ASME Code for Class "MC" Components, 1971 Edition, up to and including Summer 1971 Addenda, and Code cases 1431,1454-1 and 1517 as approved by Regulatory Guides 1.84 and 1.85. The design, fabrication and erection of suppons and bracing and similar stmetures not within the scope of the ASME Code conform to the requirements of the AISC specifications, except that the welding, welding procedures and welders' qualifications are in accordance with the ASME Code Section IX. The design, fabrication, erection and testing of the containment vessel shall also conform to the ASME Boiler and Pressure Vessel Code, Section II " Material Specifications," and Section VIII "Unfired Pressure Vessels." The containment vessel is code stamped in accordance with Paragraph NE-8000 of Section III of the ASME Boiler and Pressure Vessel Code. The vessel exhibits a general elastic behavior under accident and earthquake conditions ofloading. No permanent deformations due to primary stresses have been permitted in the design under any condition ofloading. The design of the Containment vessel was based on permissible stresses as set forth in the applicable codes. The structure will safely function within the normal design limits as specified in Section III of the ASME Boiler and Pressure Vessel Code Article NE-3000 " Design" and Regulatory Guide 1.57 (June 1973), Design Limits and Loading Combinations for Metal Primary Reactor Containment Systems Components. The polar crane is for erecting the major nuclear supply system equipment and for servicing and refueling when the plant is in operation. The crane loads and its built up ring girder support are carried by the steel containment vessel cylindrical walls. Restraint framing is provided for all pipes, equipment, electrical trays and heating and ventilating ducts where failure of any of these items could effect the safe shutdown of the reactor. Page 3-6

l 3.1.1.5.2 Design Criteriafor Other Category I Structures 3.1.1.5.2.1 Shield Building The Shield Building, is a reinforced concrete structure constructed as a right cylinder with a shallow dome roof. It has an outside diameter of 154 ft. and a height from base slab to the top of the dome of 249.5 ft. The thickness of the wall is three ft. except at the base (below elevation -18.17 ft. MSL) where it is 10.0 ft thick to provide support for the construction of the containment vessel. A nominal four ft. annular space exists between the interior face of the concrete shield structure walls and the outside face of the steel containment. This space provides the means of collecting and diluting any leakage from the containment vessel following a LOCA. A 4.0 ft. nominal clearance between tl e bottom face of the concrete shield stmeture dome and the top of the steel containment dome allows for access for construction and inspection and to assure freedom of movement of the steel containment. The Shield Building is a free standing structure without any structural ties between it and the containment vessel above the foundation level. A concrete fill in the bottom of the structure supports the steel containment. The Shield Building serves the following functions: a) as a biological shield during normal operation and after any accident within the steel containment up to and including the postulated loss of coolant accident, b) as a low leakage structure following any accident within the steel containment up to and including postulated LOCA, and c) as a shield for the primary steel containment for adverse external environmental conditions due to low temperatures, wind, tornadoes, and external missiles. The Shield Building is designed to seismic Categon I requirements. During normal operation, the Shield Building is maintained at a negative pressure by the Annulus Negative Pressure System. After LOCA, the pressure in the annular space will increase due to thermal energy transfer from the containment vessel. This pressure increase will be vented by the Shield Building Ventilation System. 3.1.1.5.2.2 Reactor Auxiliary Building The Reactor Auxiliary Building is a multistog reinforced concrete structure located immediately south of the Reactor Building. The interior floor constmetion is a beam and girder construction supported by reinforced concrete columns. The building occupies an area approximately 260 ft by 219 ft and extends from the top of the common mat at elevation -35 ft. MSL up to roofievels e: vaging from elevation +46 ft MSL to elevation +106.5 ft MSL. Above the common mat, the pj building is stmeturally separated from the centrally located Reactor Building at all levels. , AT i Nt. (p;! C $Q1 Page 3-7 )

The Reactor Auxiliary Building houses the waste treatment facilities, engineered safeguards 1 systems, switchgear, laboratories, diesel generators and main control room. It further provides protection to the cable and piping penetration areas of the Reactor Building. The building exterior walls, floors, and interior partitions provide plant personnel with the necessary biological radiation shielding, and protects the equipment from adverse atmospheric conditions such as winds, temperature, and missiles. The condensate and refueling water storage pools are an integral part of the building. The Reactor Auxiliary Building is seismic Category I compliant considering the loads and loading combinations for abnormal / extreme environmental conditions. The building is protected against exterior flooding up to elevation +29.25 ft. MSL. 3.1.1.5.2.3 Fuel Handling Building The Fuel Handling Building is a reinforced concrete structure located immediately north of the Reactor Building. It occupies an area approximately 73 ft. by 117 ft. and it extends from the top of the common foundation mat at elevation -35 ft MSL to the rooflevel at elevation +94 ft. MSL. Above the common mat the building is structurally separated from the Reactor Building at all levels. The Fuel Handling Building houses a spent fuel pool, spent fuel pool pumps, spent fuel pool heat exchanger, backup fuel pool heat exchanger, spent fuel pool purification pump and heating and ventilating equipment. The building also provides space for the new fuel vault and decontamination area for spent fuel casks, and miscellaneous equipment. The spent fuel pool is a stainless steel-lined reinfbreed concrete tank structure that provides space for storage of spent fuel, spent fuel casks and miscellaneous items. The Fuel Handling Building exterior walls, floors, and interior partitions provide plant personnel with the necessary biological radiation shielding and protect equipment from the effects of adverse atmospheric conditions such as winds, temperature, missiles, flooding and corrosive environment. The Fuel Handling Building is seismic Category I compliant, considering the loads and loading combinations for abnormal / extreme environmental conditions. 3.1.1.5.2.4 Component Cooling Water System (CCWS) Structure Component Cooling Water System (CCWS) structure comprises two independent sets of dry and wet cooling towers located on the east and west side of the Reactor Building. Each set of dry and wet cooling towers consists of a reinforced concrete box structure with overall dimension 37 ft by 103 fl. and 26 fl. by 57 ft, respectively. An access to the equipment hatch of the Reactor Building is provided in the west CCWS structure. Page 3-8

The cooling towers are supported on the common mat at elevation -35 A. hiSL and extend in height to elevation +29.25 A. MSL. Each dry cooling tower is further subdivided into five reinforced concrete chambers and is equipped with three fans supported on the walls at different levels in each chamber (totaling 15 fans). Each wet cooling tower has two reinforced concrete chambers. The minimum thickness of the walls is two A. with 3.5 R. thick walls supporting the fans. The CCWS structure is seismic Category I compliant, considering the loads and loading combination for abnormal / extreme environmental conditions. 3.1.1.5. 3 Design Criteriafor Category I Systems and Equipment System components important to safety and the containment boundary are classified in accordance with ANSI N18.2, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," 1973, and ANSI N18.2a, " Revision and Addendum to Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," 1975. System safety classifications and design and fabrication requirements meet the intent of Regulatory Guide 1.26, " Quality Group Classifications and Standards for Water, Steam, and Radioactive-Waste-Containing Components of Nuclear Power Plants," June 1975, with a clarification noted for the reactor coolant pump bearing oil and cooling systems. Equipment that is not designed and built to the exact AShiE Code specified in Regulatory 1.26 has been specified and listed in the FSAR. 3.1.1.6 Seismic Design Basis 3.1.1. 6.1 Site Seismicity Epicentral locations for all recorded earthquakes in the central Gulf Coastal Plain, including the hiississippi embayment, which have a reported intensity of about IV-V hiodified hiercalli (hihi) or greater, have been investigated. Historic earthquake data were assembled between latitude 27.5 to 37.3 North and longitude 86 to 96 West. The earthquake data is a compilation of U.S. Department of Commerce reports on U.S. Earthquakes,1928 through 1972 the Earthquake History of the United States revised through 1970 and Preliminary Determination of Epicenters Listing issued by the U.S. Geological Survey and other reports. Ten small earthquakes have occurred within about 200 miles of the site. The earthquakes that have occurred are considered non damaging to equipment and distribution systems properly installed to industrial standards. The equipment and distribution systems at Waterford 3 are designed and installed to much more stringent seismic criteria. Page 3-9

The uniform building code designates the vicinity of the site as Zone 0 (on the map entitled " Map of the United States Showing Zones of Approximate Equal Seismic Probability"). The U.S. Coast and Geodetic Survey indicates Zone 0 as an area of no earthquake damage. 3.1.1.6.2 Seismic input to Structures and Equipment The seismic design was based on the acceleration ground response spectrum curves for the operational basis eanhquake, OBE, and for the Safe Shutdown Eadhquake, SSE. The curves were normalized to 0.05g for the OBE and 0.10g for the SSE. The FSAR commitment for an SSE of 0.10g is the legal minimum specified by 10CFR100 Appendix A. This very conservative surface acceleration is double the maximum acceleration appropriate for the maximum canhquake that has occurred in the site's tectonic province during the past 250 years. A synthetic earthquake record with a maximum acceleration of 0.10g was developed to generate response spectra in the safety-related stmetures at Waterford 3. In simulating the earthquakes, a maximum duration of 20 seconds was used in the model, of which 0 to two seconds is the rising period, two to seven seconds is the constant maximum acceleration period, and seven to 20 seconds is the receding period. These durations mimic the available data on earthquakes in this region. The shape of the response spectra of the simulated earthquake for a single degree of freedom approximates N. M. Newmark's Spectrum Curve as discussed in his paper, " Design Criteria for Nuclear Reactors Subject to Earthquake Hazards," Urbana, Illinois, May 25,1967. The maximum amplification, at two percent critical damping is approximately 3.5, greater than the value shown in the Housner Spectmm of TID 7024, but less than the Newmark value. The design response spectra used in the plant design differ from the design response spectra recommended in NRC Regulatory Guide 1.60, Design Response Spectra for Seismic Design of Nuclear Power Plants, Revisicm 1 December 1973. The regulatog guide response spectra have slightly higher values in general. Use of Regulatory Guide 1.60 permits utilization of damping values indicated in Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Power Plants, October 1973. These damping values are equal or greater than the values utilized for Waterford-3 plant design. By utilizing lower damping values in the Waterford-3 design, as compared to the damping values of Regulatory Guide 1.61, the analysis and design of Waterford-3 compensate for any differences. The following paragraph describes in further detail why this is true. The design response spectra used in the plant design produce an amplification factor of 3.5 at two percent damping (for OBE)in the period range 0.15 to 0.5 seconds. The use of Regulatory Guide 1.60 produces an amplification range of 3.54 to 4.25 for the above damping and period range; however, the use of higher damping value allowed by Regulatory Guide 1.61 (four percent for reinforced concrete structures in OBE) results in an amplification factor range of 2.92 to 3.50 for the above period. A similar correlation exists for the SSE. Thus the design response spectra and g the damping factors used for the plant design provide an adequate and approximately equivalent j basis for seismic design. Page 3-10 I a

) The horizontal design response spectra for the SSE and OBE are applied at the bett om of the 1 ' foundation of the common mat of the Nuclear Plant Island Structure, at the top of the Pleistocene i formation in the free field. There is no data in the region relating to horizontal and vertical acceleration for strong motion earthquakes. A vertical acceleration equal to two-thirds of the horizontal acceleration was used in developing the vertical design response spectra. 3.1.1.7 Seismic Review Team (SRT) The Waterford 3 IPEEE SMA included a joint engineering effort between the Waterford 3 Design Engineering Department staff and the consultant project staff. In addition to the project l management and contract management work associated with the use of consultant resources, Waterford 3 engineers were integrated with the consultant team in all aspects of the work. The principal areas where Waterford 3 engineering participation was included were in development of the SMA success paths, and participation as scismic walkdown team members during the seismic screening walkdowns. The walkdown teams were composed of the following personnej: On-line Walkdown Train B 11/9/93 - 11/11/93 Team 1 - Mr. George G. Thomas - SRT Walkdown Team Leader - Stevenson & Associates Mr. Greg Ferguson - SRT Member - Waterford 3 Engineering Mr. Steven Farkas - SRT Member - Waterford 3 Engineering, Systems Supports Mr. John Burke - SRT Member - Waterford 3 Engineering Team 2 - Mr. Stephen Anagnostis - SRT Walkdown Team Leader - Stevenson and Associates Ms. Maria Rosa Gutierrez - SRT Member - Waterford 3 Engineering Mr. Steven Farkas - SRT Member - Waterford 3 Engineering, Systems Supports Mr. John Burke - SRT Member - Waterford 3 Engineering On-line Walkdown Train A 1/25/94 - 1/31/94 Team 1 - Ms. Maria Rosa Gutierrez - SRT - Walkdown Team Leader - Waterford 3 Engineering Mr. Greg Ferguson - SRT Member - Waterford 3 Engineering Mr. Siddarth Munshi - SRT Member - Waterford 3 Engineering ^ Outage Walkdown 3/14/94 - 3/17/94 Team 1-Mr. George G. Thomas - SRT Walkdown Team Leader - Stevenson and Associates i Ms. Maria Rosa Gutierrez - SRT Member - Waterford 3 Engineering Mr. Greg Ferguson -SRT Member - Waterford 3 Engmeenng Page 3-11

Structures Walkdown 3/21/94 - 3/23/94 Team 1 - Dr. John D. Stevenson - SRT Walkdown Team Leader - Stevenson and Associates Ms. Maria Rosa Gutierrez - SRT Member - Waterford 3 Engineering Mr. John Burke - SRT Member - Waterford 3 Engineering 3.1.1.8 Walkdown Preparation A detailed written project plan was prepared before the walkdown that outlined the scope of each task, the detailed technical approach to the performance of each task, and the interfaces between the team members and Waterford 3 personnel. Before the walkdown, data assembly and evaluation was performed to define a technical baseline for the systems analysis and seismic screening walkdown. The design documentation for structures and components were key to performing the screening walkdown evaluations. In general, the design data ofinterest was construction details (such as bolting) and seismic stress j analyses and laboratory test repons that will form the basis of the reduced scope SMA. This task served the needs of the IPEEE seismic data assembly requirements for electrical and mechanical components, in NUREG 1407 and EPRI NP-6041-SL. Specific documentation assembled and evaluated before and during the walkdowns included: the Waterford 3 Safe Shutdown Paths and Equipment List report prepared for the IPEEE by Waterford 3 plant arrangement drawings sections of the Waterford 3 FSAR relating to the seismic criteria and licensing basis for the plant the ground response spectra for the SSE the floor response spectra and how they were generated a sample of construction details of the anchorage including drawings and specifications a sample of procurement and seismic testing specifications for equipment examples of calculations for seismic and anchorage qualifications design basis documents for the Waterford 3 structures selected evaluations for block walls design calculations for a sample oflarge flat bottom tanks A walkdown plan was developed before the walkdowns and included the criteria to be used for the walkdown. 3.1.1.9 _Screenine Walkdown 3.1.1.9.1 Overall Walkdown Conduct For the SMA at Waterford 3, rigorous statistically based sampling criteria are neither practical nor desirable. The SMA procedures and guidelines used were heavily reliant on thejudgment of Page 3-12

l highly experienced engineers and criteria for sampling in this plant likewise r.re modeled around this judgment. There are two areas of a reduced scope SMA where sampling was applicable and used at Waterford 3. They are; (1) screening of structures and components, and (2) walkdown. Issues that influence the sampling described in this plan are; redundancy provided by multi-train systems, similarity in design and location of redundant trains, treatment of single failures, access to components during walkdowns, and systems interactions potential including fire and internal flood s sources. The sampling approach described below is appropriate for modern plants of Waterford 3's v'mtage. The document review and walkdown verified that uniform practices in accordance with the plant design basis for construction, design and installation were implemented. Therefore, the sampling approach was used throughout the effort. The Waterford 3 walkdown confirmed, as expected, that most items in a given equipment class were either identical or very similar. The plant documentation review and walkdown confirmed that the vast majority of equipment was manufactured, and installed as specified. The screening procedures to be used at Waterford 3 for generic categories of equipment and structures

ontained caveats or inclusion rules that were checked during the walkdown. Because the equipment at Waterford 3 was purchased and installed to similar codes and standards, the SRT screened generic classes of equipment on the basis of their relative ruggedness. The screening sampling size for identical or very similar equipment in a given class for caveats was one or greater. The screening size for very similar equipment in a given class with identical or very similar anchorage was two or greater. The increased sample for anchorage is based on experience at other plants that anchorage installations are not always consistent. This is consistent with the guidance given in Appendix D of EPRI NP 6041-SL. A 100 percent " walk-by" of all equipment on the SSEL was employed to check for unique equipment details and for seismic interactions.

The structures at Waterford 3 were screened generically. The drawings and analysis models were reviewed for details that might indicate seismic vulnerabilities in accordance with the requirements of a reduced scope SMA. The drawing and structural analysis reviews confirmed that consistent good practice in design detail and analysis was utilized at Waterford 3. Therefore, it was not necessary to review more than a small sample of the details of connections, reinforcement bar placement, constmetion joints, etc., to make the judgments on screening. Distribution systems installed in bulk such as piping, cable trays, HVAC ducting, electrical conduit and instrument lines were screened generically after completion of a walkdown with verification that the distribution systems meet the inclusion rules. It was confirmed that the design and installation practices at Waterford 3 are consistent, therefore the screeningjudgment was based upon a review of the general specifications and drawings for a single run of each generic class of distribution system. As expected the review of the general specifications and drawings did not indicate significant differences in design and installation practice that was confirmed during the equipment walkdown. r Page 3-13

i Each walkdown team consisted of at least two seismic capability engineers and at least one j ~ systems engineer or operator from Waterford 3 was also available for the duration of each : j walkdown to swing from team to team when systems / operations input was required. j The seismic capability walkdowns gathered the necessary data to support the SMA screening analysis. The seismic capability walkdowns also served to collect the necessary data for the l component evaluations in the SMA program to implement the requirements of NP 6041-SL. j i There were three walkdowns performed. An outage walkdown was performed during the Spring i of 1994, and included equipment located in containment, as well as the containment structures review. The structures and distribution system reviews were performed during the course of the outage walkdown. There were two on-line walkdowns performed, one during the Fall of 1993 for the B train equipment and one during the winter of 1994 for the A train equipment. l A typical day during the walkdown consisted of: l l 1. review'mg issues identified on previous days for determination as to whether the item is screened or an outlier 2. planning the day's walkdown effod 3. performing the walkdown i 4. seismic contractors briefing Waterford 3 on the day's progress 3.1.1.9.2 Walkdown Results l t Three Seismic Screening-and-Verification Walkdowns were performed at the Waterford 3 facility l for the IPEEE SMA. The purpose of the walkdowns was to assess the relative seismic capacity based on earthquake and testing experience of a large number of selected safety-related plant _ i structures and components. The screening was performed to the licensing basis SSE at Waterford -

3. The results for the walkdown are in separate sections below. The Seismic Verification Data Sheets for all the walkdowns are in Reference 3-7.

j r t ( 3.1.1.9.3 Walkdown Resultsfor the Train A and Train B On-line Walkdown } i The " Train A" on-line walkdown occurred the week ofJanuary 25,1994. The train was walked i down by one team of Waterford 3 engineers. The " Train B" on-line walkdown was performed the l week of November 7,1993. The train was walked down by two teams consisting of Waterford 3 l and Stevenson & Associates engineers.. Section 3.1.1.7 contains the walkdown teams for each train. Train A consisted of 273 items and Train B consisted of 388 items. Train B consisted of more items since it contained the majority of the containment isolation valves. The seismic walkdowns found Waterford 3 to be seismically rugged and identified no outliers affecting plant l operability. However, there was one item that could not be screened. Table 3.1 lists the i equipment that was not screened. i e f Page 3-14 l

1 l I Table 3.1 - Equipment Not Screened During Walkdown and the Reason Why Class Equipment ID No. Issue to Resolve 3 4KVESWGR3B XPANEL Station air pipe can impact switchgear during an earthquake. 3.1.1.9.4 Walkdown Resultsfor the Outage Walkdown The outage walkdown occurred the week of March 14,1994. The walk down was completed by l one team as stated in Section 3.1.1.7. There were 149 items on the initial list for the walkdown. During the outage walkdown the control room ceiling was evaluated for possible interaction with control room equipment and operators. The control room ceiling is built from light-weight panels l' x l' square that interlock with one another. The ceiling support is a unistrut frame cantilevered from the concrete floor above. The unistrut frame only supports the ceiling and recessed lighting. Other distribution systems such as HVAC duct and cable trays are supported by independent structural members (structural angles, channel, etc.) as they are in the plant at large. To guard against falling of the lights and panels in the event of an earthquake, each panel and light have independent tie wires that go around a horizontal unistrut in the frame. Panels or lights that may come loose will hang from these wires. The lights above the main control board have a latch at one side for changing the lights. In the worst case the latches may come loose and the reflectors may hang from the hinges on the other side. In any case they will not fall on the operators. Waterford 3 has a qualification calculation for the control room ceiling to meet the SSE seismic criteria for the plant. Light fixtures throughout the plant were well attached without open hooks. Platform grating were all clipped down. Masonry walls by equipment reviewed were reinforced and had been previously evaluated in the Waterford 3 IE Bulletin 80-11 program. All equipment reviewed was located in seismic Category I structures, therefore, it was unnecessary to review non-seismic Category I stmetures. There were no outliers affecting plant operability found during the walkdown However, there were several equipment items that could not be screened. Table 3.2 lists the equipment that was not screened. Page 3-15

Table 3.2 - Equipment Not Screened during the Outage Walkdown and the Reason Why Class Equipment ID No. Issue to Resolve 20 IC ECP41 XPANEL Unsecured personal storage lockers and file IC ECP42 XPANEL cabinets are in close proximity of these panels. IC ECP25 XPANEL These lockers and cabinets should be moved or IC ECP26 XPANEL secured so they cannot interact with the panels. IC ECP43 XPANEL l IC ECP44 XPANEL IC ECP45 XPANEL PMCICDSMC XPANEL IC ECP27 XPANEL IC ECP28 XPANEL IC ECP29 XPANEL IC ECP30 XPANEL IC ECP31 XPANEL IC ECP08 XPANEL IC ECP49 XPANEL IC ECP48 XPANEL IC ECP18 XPANEL 3.1.1.9.5 Containment andStnictures Walkdowns The Waterford Unit 3 structures were reviewed on March 21 and 23 as part of the IPEEE program in response to the requirements contained in NUREG-1407. The particular structures evaluated included: 1) the reactor containment shell 2) the shield building 3) the containment internal structure 4) the auxiliary building i a) emergency diesel generator rooms b) control room 5) containment penetrations including hatches and locks 6) containment isolation valves 7) containment ventilation and backup air systems The calculations concerning these structures have been evaluated for: 1) Seismic inertia forces and moment on structures 2) Seismic induced displacements The structures and systems listed in (1) to (7) above were also walked down to evaluate: Page 3-16 i

a) Any potential spatial interactions between structures or systems not considered in design b) Installations which could effect the seismic adequacy given the seismic design level of 0.lg SSE-ZPGA defined for the plant c) Seismic evaluations of containment penetrations, ventilation and backup air and supponing systems have been performed by a containment walkdown. The seismic walkdowns and structures review found Waterford 3 to be seismically rugged and identified no outliers afTecting plant operability. A walkdown of containment penetrations indicates that penetrations were, in general, supported off the inside of the containment steel shell with flexible cennections to the interior of the containment internal structure. It was also noted that large ventilation ducts were supponed on the internal structure at the operating deck level and the containment shell above the operating deck. However in each such instance there was a flexible bellows connecting the sections of the duct supponed on the two separate structures. The deflection of the containment shell and other building structures are contained in Table 3.8-23 of the FSAR for two horizontal directions of SSE input earthquake. The maximum SSE relative displacements between the containment shell and the containment internal structure according to this table may be approximately 1.8 inches at the Elev. +60.3. These deflections are a combination of the in-phase and out-of-phase deflections. Almost all of this deflection is due to SSE translation and rocking. More recent evaluations of relative displacements between the auxiliary building and the shield building and between the shield building and the steel containment shell have separated the in-phase and out-of-phase deflections. The maximum relative of differential motion between the steel containment and containment internal structure will be no greater than 0.0595 inches at Elev. +60.3. The maximum SSE relative displacements at the operating deck level (Elev. +46.0) and at Elev. +21.0 are 0.048 inches and 0.030 inches, respectively. A walkdown of containment concluded that this gap existed between the containment shell and the internal structure. A walkdown of containment also concluded that any component that was supported by the containment shell and the internal structure had sufficient flexibility. The area between the shield building and the auxiliary building was also walked down in the areas of electrical and mechanical penetrations and it was noted that all penetrations were flexibly connected between the shield wall and the first support on the auxiliary building. While performing the walkdowns for Train A and B, the SRT also looked for other potential interactions and none were noted. Page 3-17

4 3.1.2 System Analysis 3.1.2.1 Equipment Necessary to Achieve and Maintain Hot Shutdown and Development of SSEL 3.1. 2.1.1 Summary This section describes the process for creating the Safe Shutdown Equipment List (SSEL). The process depended on procedure OP-902-008, " Safety Function Recovery Procedure," wiring diagrams, flow diagrams, and the station component database (SIMS). The Full SSEL has the equipment needed to combat the IPEEE-seismic scenario: a SSE level earthquake that causes a LOOP and a 1" SBLOCA The IPEEE-seismic analysis is based on techniques and an accident which are both described in EPRI-6041. It requires plants to prevent core melt during the IPEEE-seismic scenario for 72 hours. EPRI-6041 hypothesizes that core melt can be prevented by successfully maintaining four safety functions: Reactivity Control Reactor Coolant Pressure Control Reactor Coolant Inventory Control Decay Heat Removal Waterford procedure OP-902-008 describes how the operators maintain seven safety functions, including the four called out by EPRI-6041. The additional safety functions are: Containment Isolation Containment Temperature & Pressure Control Combustible Gas Equipment on the Full SSEL is associated with steps in OP-902-008 that maintain the four EPRI-6041 safety functions. That " front-line" equipment as well as support equipment (found on CWDsi) and passive equipment (found on flow diagrams) appear on the Full SSEL. Because the goal of the IPEEE-seismic at Waterford is to walkdown the equipment on the Condensed SSEL, equipment location is provided on the Full SSEL. Equipment in the plant was located primarily with SIMS. 3.1.2.1.2 Introduction This section describes how equipment came to be listed on the " Full SSEL." There are two SSELs, a " Full SSEL" and a " Condensed SSEL." The " Condensed SSEL" is created by applying I CWDs are the plant control wiring diagrams. Page 3-18

the EPRI-6041 " rule of the box" criteria to the " Full SSEL", e.g., only a motor control center 1 cabinet needs to be examined during the IPEEE-Seismic walkdown, not all its cubicles. The EPRI-6041 "mle of the box" is straight forward and assumes that the " Full SSEL" is correct. Therefore, this report will focus on the method employed for adding equipment to the " Full SSEL." The IPEEE requires a walkdown of equipment needed to combat the following scenario for 72 hours. The IPEEE assumes the plant suffers a SSE level earthquake that causes both a loss-of-off-site-power (LOOP) and a 1" diameter small break loss of coolant accident (SBLOCA). The Individual Plant Examination ofInternal Events (IPE) had a scope that included many types of accidents, but only 24 hour coping. Thus, a new basis for equipment selection (something besides the IPE) is needed for the IPEEE. There is one operating procedure that covers all safety functions needed to combat a LOOP & SBLOCA. It is OP-902-008, " Safety Function Recovery Procedure." OP-902-008 addresses seven safety functions that form a superset of the four safety functions called out in EPRI-6041, i.e., Reactivity Control, Reactor Coolant Pressure Control, Reactor Coolant Inventory Control, Decay Heat Removal. The EPRI-6041 functions address preventing core damage rather than mitigating consequences of a core melt. Safety-related equipment seeks to reduce the probability of core melt and the consequences of core damage. The IPEEE presumes that preventing core damage eliminates the need to mitigate consequences. Therefore, the only equipment needed on the Full SSEL is that equipment that will prevent core damage. SAFETY FUNCTION in OP-902-008 Corresponding EPRI-6041 function.s Reactivity Control Reactivity Control Vital Auxiliaries Support Systems RCS Inventory & Pressure Control Reactor Coolant Pressure Control Reactor Coolant Inventory Control RCS & Core Heat Removal Decay Heat Removal Containment Isolation Support Systems Containment Temperature & Support Systems Pressure Control Combustible Gas NA to SMA (re page 3-20) 3.1.2.1.3 How The FullSSEL Accounts For The EPld Safety Functions 3.1.2.1.3.1 General Notes The IPEEE asks that licensees list a primary and alternate means of achieving each EPRI-6041 safety function. For the IPEEE, it is assumed that all equipment connected to the diesel generators is available to the operators. Page 3-19

Due to the format of OP-902-008, the Full SSEL contains equipment from at least two trains of one safety system for each pertinent step. Some steps are impertinent because of the 72 hour LOOP assumption, i.e., some equipment is not normally connected to the diesel generators. Many times the pertinent OP-902-008 steps gives the operator several systems to choose from. When only one system can fulfill the OP-902-008 step, two independent trains comprise the primary and alternate means of achieving each EPRI-6041 safety function. When more than one system can fulfill the safety function, the number ofindependent trains increase. Thus, the Full SSEL incorporates more than just a primary and secondary means of achieving each EPRI-6041 safety function. A certified Shift Technical Advisor (STA), who is also an SRO, helped determine that steps and that equipment implement the EPRI-6041 safety functions. EPRI-6041 safety functions prevent core damage. This allowed exclusion of Waterford equipment geared solely to mitigating the consequences of either core damage (e.g., combustible gas control), or station blackout (e.g., controls needed to shed load off the DC busses). OP-902-008 guides the operators into using equipment that won't be available given a LOOP. Equipment not normally powered via the emergency on-site power system do not appear on the Full SSEL, e.g., condensate. However, the IPEEE scenario is LOOP only, so both Emergency Diesel Generators (EDG) operate in the IPEEE scenario. Each equipment train (i.e., Train A and Train B) totally depends on its respective EDG for electric power over the course of 72 hours. Some systems have an AB train. Train AB represents a third alternative in some systems, e.g., Component Cooling Water (CCW), High Pressure Safety Injection (HPSI). Because it is a third choice, there is no need to consider Safety Train AB in the Full SSEL. There is one exception. Those components designated as AB that separate train A from AB and train B from AB are included on the Full SSEL. Because the operators will initiate shutdown cooling during the 72 hours, shutdown cooling equipment is on the Full SSEL. The IPE scope did not include shutdown cooling equipment because the IPE success criteria depended upon Reactor Cooling System (RCS) cooling via the steam generators and emergency feedwater. Therefore, simply listing equipment included in the IPE would not meet the intent of the IPEEE. OP-902-008 asks operators to check parameters shown by instruments and the [ Class IE, seismic Category 1] Qualified Safety Parameter Display System (SPDS) in the control room. Instruments not specifically required by OP-902-008 are excluded, e.g., Heating, Venting and Air Conditioning (HVAC) temperatures. Note, the IPE scope did not include any measuring instmments. Once again, simply listing equipment included in the IPE would not meet the intent of the IPEEE. The method described below for creating the Full SSEL does meet the intent of the IPEEE. Systems that control containment temperature and pressure are on the Full SSEL. The closest design basis accident at Waterford to a 1" SBLOCA has a 0.01 square foot break. Based on that analysis, the followirg conclusions underpin listing containment spray (CS), and containment fan Page 3-20

r i i coolers (CFCs) on the Full SSEL. Given that bo;h containment spray pumps operate when commanded by the Engineering Safety Features Actuation System, the RWSP empties out and recirculatior u the Safety Injection (SI) sump begins before RCS conditions permit Shutdown Cooling (SDC) operation. Therefore, the means to condense steam in the containment fall into the scope of the Full SSEL. Given a 1" SBLOCA, no amount of hydrogen can be produced in a non-core damage situation to threaten containment integrity. Therefore, combustible gas does not perform " core melt prevention" and it is not in the scope of the Full SSEL. The IPEEE-seismic scenario includes LOOP, not SBO. Therefore, the OP-902-008 steps that conserve DC power were ignored in compiling the Full SSEL. For example, the controls, etc., needed to turn-off seal oil pumps for the main feedwater turbines were not included on the Full SSEL. 3.1.2.1.3.2 The EPRI Safety Functions The paragraphs below address each EPRI-6041 safety function in turn. At the system level, they describe why a system appears on the Full SSEL. After these functions, there is a heading for SUPPORT SYSTEMS. The description there accounts for the large amount of equipment shown on the attached Full SSEL. Reactivity Control OP-902-008 gives the operators three options for keeping the core subcritical. CEA insertion Boron injection with charging pumps Boron injection with safety injection system pumps CEA (control element assembly) insertion happens either automatically, or manually. The Full SSEL only includes the cabinets that cause automatic CEA insertion. Cabinets related to CEA motion for start-up and maneuvering were found unnecessary for responding to the IPEEE-seismic scenario. High boron concentrations in the RCS can keep the nuclear chain reaction from returning to critical conditions. All of the various means of baron injection are included in the Full SSEL. Thus, gravity feed valves, Boric Acid Makeup (BAM) pumps, the BAM tanks, the Volume Control Tank (VCT), and two of the three charging pumps are part of the Full SSEL. The AB charging pump is not included for reasons given above. The SI pumps fulfill this function as well as the inventory control function so they are also included in the Full SSEL. The hot leg injection valves, etc., are a means to keep boron from precipitating out of solution. However, for a 1" SBLOCA, hot leg injection should not be necessary for reactivity control. Hot leg injection valves that are also containment isolation valves are on the Full SSEL. Page 3-21

Esactor Coolant Pressure Control Reactor coolant pressure control is accomplished with either ADVs or MSSVs, and EFW feeding the SGs that relieve steam via the ADVs and SDC. MSSV Main Steam Safety Valves EFW Emergency Feedwater ADV Atmospheric Dump Valves SDC Shutdown Cooling All of this generally described pressure control equipment and its supporting equipment are part of the Full SSEL. Reactor Coolant Inventory Control A 1" SBLOCA is a hole that can drain the RCS faster than three charging pumps can fill the RCS. Thus, the IPEEE-seismic scenario requires HPSI to provide RCS inventory control. HPSI needs a borated source of water before the SI sump fills so the RWSP is part of the Full SSEL too. Because leakage through the 1" SBLOCA continues even after SDC begins, HPSI must be able to recirculate SI sump water back into the RCS. Therefore, all equipment needed to establish SI sump recirculation is also part of the Full SSEL. During the IPEEE-seismic scenario, all HPSI pumps are presumed operable, thus, only the Train A and Train B pumps were considered for the Condensed SSEL. Equipment solely for HPSI pump AB was excluded from the Condensed SSEL. D_egy Heat FAmoval During a 1" SBLOCA, the emergency operating philosophy is to initiate SDC as soon as possible. The large amount of energy trapped upstream of the MSIVs needs to be diverted quickly. The operators can then gradually cooldown the RCS until SDC entry conditions exist. The IPEEE scenario causes MSIS to close both the MFWIV and MSIV.2 LOOP will render BOP systems, e.g., Main Feedwater System (MFWS) & Steam Bypass Control System (SBCS), unavailable. Thus, heat from the RCS going into the SGs will be relieved near-term by the MSSVs and long-term by the ADVs. EFW will replenish the boiled off SG water (by drawing on the CSP and the wet tower basins). The passive MSSVs relieve steam in the very beginning of the IPEEE-seismic scenario. The ADVs will control SG pressure during cooldown to SDC, not the SBCS (steam bypass control system). The MFWIVs, MSIVs, MSSVs, and ADVs as well as the electric motor driven trains of EFW appear on the Full SSEL. Because both motor driven 2 MSIS Main Steam Isolation Signal t. MFWlV Main Feedwater Isolation Valve MSIV Main Steara Isolation Valve Page 3-22

EFW pumps are available, no equipment rotely related to the turbine driven EFW pump appears on the Full SSEL. Given LOOP, RCS circulation from the core to the SGs must be natural, i.e., only due to pressure and density differences at different points in the RCS. Natural circulation is enhanced by proper pressurizer heater control, but it is not required according to the STA staff. Although heaters were considered for the Full SSEL, the final conclusion that they were not absolutely necessary meant leaving pressurizer heaters ofTthe Full SSEL. Once the RCS reaches SDC entry conditions, a combination of the LPSI system and CCW system will form the front line decay heat removal system. Operators realign LPSI so that a closed loop is created. During a 1" SBLOCA, CCW can remove decay heat (deposited in the shutdown cooling heat exchangers) with just the dry cooling towers. Thus, wet cooling tower equipment is not included on the Full SSEL. 3.1.2.1.4 Compiling The FidlSSEL OP-902-008 was evaluated to determine equipment needed to implement each step. The object was to uniquely list all of the equipment available after a LOOP that OP-902-008 calls out. Regardless of how many steps use the same piece of equipment, a piece of equipment (i.e., a component ID called UNID) appears on the SSEL only once. On the other hand, it was common to list more than one component ID for a single OP-902-008 step. Therefore, the cross-reference shown in the list between component ID and step is only a representative OP-902-008 step. Because OP-902-008 covers all EPRI-6041 safety functions, this was an effective rneans of identifying all active equipment needed to implement the safety functions. 3.1.2.1.4.1 Support Systems Next, three systems were unconditionally included. They were emergency diesel generators, RAB IiVAC, and component cooling water. Equipment associated with these systems appears on the Full SSEL. Measuring instruments in HVAC systems were explicitly excluded. Operators would be aware (from their plant tours) when 11VAC was not working adequately. Since rain-fall is not part of the IPEEE-seismic scenario, the sump pumps for the dry cooling tower areas were excluded from the Condensed SSEL. There chief function is to prevent flooding the two MCC 315s. MCC 315 provides power for the dry tower fan motors. Consideration was given to adding balance of plant equipment to the Full SSEL, e.g., instrument air and non-safety power for reactor coolant pumps. Based on conversations with Design Engineering and Plant Operations, it was concluded that the plant could reach SDC conditions and remain in SDC conditions up to 72 hours after the IPEEE initiator without using any BOP system. Page 3-23 j

1 Air operated valves, (e.g., the ADVs) can be controlled manually during the cooldown. Specifically, the ADVs can be controlled manually. Note, ADVs typically do not " hunt" during a cooldown. ADV position needs to be adjusted infrequently to match the decay heat rate. 1 Natural circulation of the primary coolant will allow operators to approach SDC conditions in i roughly 10 hours. The time to reach SDC could be shortened by restarting reactor coolant pumps i (RCPs). However, the RCP restart steps in OP-902-008 are for couwnience only. Restarting - l RCPs is not necessary to reach SDC conditions. Finally, the equipment called out on Attachment 12 and Attachment 13 to OP-902-008 was included. These attachments list the equipment that responds to SIAS and CIAS automatic actuations. The importance of having SIAS start ECCS pumps to make-up water into the RCS is j obvious. CIAS is important because radioactive water will be sprayed and recirculated in the containment. The water is radioactive even without a core melt. Because EPRI-604I was unclear on this point, valves responding to CIAS were included as a support system for the safety functions. 3.1.2.1.4.2 Other Supporting Equipment The next step was to identify CWDs that showed equipment called out by OP-902-008, or that showed any of the unconditionally listed systems, e.g., RAB HVAC. CWDs show the equipment needed to support the initial set of equipment created from OP-902-008. With this equipment-to-CWD cross reference, it was possible to use the CCL (cable conduit list database) to list the interconnected equipment shown on the CWDs (e.g., panels, instrument contactors). This list completed the effort to itemize support systems. i t I i Page 3-24

l Process for creatina the Waterford 3 SSEL GOAL SCENARIO OP-902-008 List Step ust allequipment " or Attachment needed to restore Earthquake that uses equipment safety functions. LOOP still available given scenario a How to List the UNID that the 1* SBLOCA restore Peratorlooks at or CWD xref from SIMS touches to safety (CWD shows the equip. 'P _ functions._ that supports the UNID.) n hm n CWD Sheet vs. CCL (a database of equipment shown on a CWD) CCL TAG CCL Tag o Ebasco Tag CCL TAG CCL TAG y y . identify steps in OP-902-008 UNIO

• Determine which UNID is needed to implement the step UN/D y
• Find the CWD for the UNID number
• Get a list of equipment on the CWD UNIO

. Translate the CCL equipment number into a UNIO y y y . Physically locate tne UNID in the plant with SIMS Location Location Location Y SAFE SHUTDOWN EQUIPMENT LIST (SSEL) 3.1.2.1.4.3 Passive Equipment At this point the Full SSEL is complete except for the " passive" equipment the operator needs to ensure safety function success. The list of passive equipment comes from tracing the systems called out by OP-902-008 as well as the unconditional systems, e.g., RAB HVAC. The equipment highlighted on the flow diagrams was listed and compared to the nearly complete Full SSEL. The equipment that did not already exist on the Full SSEL was added, e.g., tanks. Important equipment that may be damaged following a tank collapse is on the Full SSEL. There was no specific effort to try and identify equipment in the same room with a tank. Equipment does not appear on the Full SSPL solely because it was next to a tank. 3.1.2.1.4.4 Location Cross-Reference Via SIMS and searches through the electrical design documents, a location cross-reference for this equipment was included in the Full SSEL. Although the equipment list obtained from CCL proved difficult to completely translate into UNID numbers, all of the information needed for the Condensed SSEL was found. l Page 3-25

l 3.1.2.1.4.5 Screening Applied To Equipment Added Only As A Result Of The CWD Search Adding each and every component to the Full SSEL found on the cross-referenced CWDs would have expanded the scope of the Full SSEL. Many of the CWD circuits are for indication only and j do not perform fundamental safety functions, e.g., reactivity control. Components such as transmitters, limit switches, cabinets, sampling equipment, etc. that only monitor processes were excluded from the Condensed SSEL. Instrument cabinets only associated with non-safety function systems (e.g., blowdown system) were excluded from the Condensed SSEL (e g., IC ICDC103 houses only blowdown system instruments). At times, CWDs show multiple solenoid operated valves. Only those valves directly related to one of the safety functions was retained on the Condensed SSEL (e.g., CAP-101 was screened out while creating the Condensed SSEL). The CWD search resulted in adding PDP-383A to the Full SSEL. Because that PDP only supplies equipment in the Water Treatment Building, PDP 383 A was excluded from the Condensed SSEL RWSP to CVC is normally closed. For the IPEEE-seismic scenario, it needs to stay closed. There is no automatic or remote-manual action required to reach SDC and stay there for 72 hours. Therefore, equipment related to connecting CVC to the RWSP was excluded from the Condensed SSEL. Both MCC 314s were added to the Full SSEL. During the screening described in this section, Waterford 3 determined that MCC 314 power supplies were not relevant to achieving shutdown cooling conditions. MCC 314 supplies loads in the Fuel Handling Building. Therefore, neither MCC 314 appears on the Condensed SSEL. 3.1.2.1.4.6 Conclusion The Full SSEL needed to include at least two independent means for maintaining the four EPRI-6041 safety functions. The Full SSEL provides at least two trains (or paths) for each pertinent step in OP-902-008. Pertinent steps implemented one of the feur core damage prevention EPRI-6041 safety functions. The Full SSEL needed to include the suppon systems needed by the front line safety systems identified above. Using CWDs and flow diagrams, the process used to compile the Full SSEL assured that all important suppon equipment is part of the Full SSEL. The "mle of the box" grouped equipment under a common banner, e.g., a motor control center. Thus, no accuracy was lost after applying the " rule of the box" (in terms of maintaining the four safety functions). EPRI-6041 lists 16 fields related to each piece of equipment on the SSEL. It was considered a guide. The attached Full SSEL does not include the following information: description, desired state, power required, and supporting system. The additional work needed to include this information merely duplicates information in SIMS. The Full SSEL includes the critical information needed to create the Condensed SSEL. Page 3-26

l 3.1.2.1.5 Creating The CondensedSSEL After the SSEL was developed by Waterford 3 Design Engineering and reviewed, Waterford 3 condensed the list using " rule of the box" considerations. Equipment categories were numbered as shown in Table 3.3. For equipment included in Classes 1 through 20, all the components mounted on or in this equipment were considered to be part of that equipment and do not have to be evaluated separately. For example, the diesel generator (Equipment Class #17) includes not only the engine block and generator, but also all other items of equipment mounted on the diesel generator or on its skid. Components needed by the diesel generator but not included in the " box" (i.e., not mounted on the diesel generator or on its skid) were identified and evaluated separately. Table 3.3 EQUIPMENT CLASSES 0 OTHER 11 CHILLERS 1 MOTOR CONTROL CENTERS 12 AIR COMPRESSORS 2 LOW VOLTAGE SWITCHGEAR 13 MOTOR-GENERATORS 3 MEDIUM VOLTAGE SWITCHGEAR 14 DISTRIBUTION PANELS 4 TRANSFORMERS 15 BATTERIES ON RACKS 5 HORIZONTAL PUMPS 16 BATTERY CHARGERS & INVERTERS 6 VERTICAL PUMPS 17 ENGINE-GENERATORS 7 FLUID-OPERATED VALVES 18 INSTRUMENTS ON RACKS 8 MOTOR-OPERATED AND 19 TEMPERATURE SENSORS SOLENOID OPERATED VALVES 9 FANS 20 INSTRUMENTATION AND CONTROL PANELS AND CABINETS 10 AIR HANDLERS 21 TANKS AND HEAT EXCHANGERS Rule of the box considerations were incorporated by Waterford 3 engineers during a preliminary " walk by" and researching plant drawings and manuals for equipment on the SSEL shortly after SSEL development. 3.1.3 Analysis of Structure Response 3.1.3.1 Seismic Analysis of Seismic Category I Structures The seismic analyses of all seismic Category I structures were performed using either the normal mode time history technique or the response spectrum technique. In the case of seismic Category I structures, the seismic response was determined by the response spectra developed for the OBE (0.05g) and the SSE (0.10g). l i Page 3-27

t As seismic Category I stmetures et Waterford 3 are on a common foundation mat, therefore, the mathematical modeling involved constmetion of a single composite model for each directional seismic analysis. The model comprises five individual cantilevers, representing the Reactor Building, the containment vessel, the reactor internal structure, the Reactor Auxiliary Building and the Fuel Handling Building. The Component Cooling Water System is not separately identified and is included in the Reactor Auxiliary Building and Fuel Handling cantilevers. The five cantilevers are founded on the same base, that is in turn supported by foundation springs that modeled the Soil Structure Interaction between the buildings and soil. For each cantilever, the distributed masses of the structure are lumped at certain select points and connected by weightless elastic bars representing the stiffness of the stmeture between the lumped masses. In determining the stiffnesses, the deformation due to bending, shear and joint rotation are considered throughout. Every mass point of the two dimensional horizontal model can have two degrees of freedom, namely, translation and rotation. For the vertical model only one translational degree of freedom is considered. Torsional modes of vibration were analyzed by three-dimensional lumped-mass system using the MRI/Stardyne computer program. Each mass point of the system was given two onhogonal horizontal degrees of freedom and a third rotational degree of freedom in the same plane. The mass points were then idealized as a rigid diaphragm with three degrees of freedom, two translational and one rotational. In this analysis, torsional effect results from the translational seismic inputs because of the eccentricity between the mass center and the shear center of each floor (mass polar moment ofinenia). Torsional soil stmeture interaction was considered by including a torsional spring at the base. Once the dynamic models were developed, analysis of the structures determined the natural periods for vibration of each structure. In these analyses, periods and mode shapes were determined for each mode. These data define participation factors for each structure. These participation factors together with the spectral acceleration masses and relative displacement define resultant seismic forces in each mode. These forces were applied to Seismic Category I Stmetares to determine resultant shears and moments for design purposes. 3.1.3.2 Seismic Design of Mechanical and Electrical Equioment i The design basis for safety-related equipment furnished for installation at Waterford 3 is equivalent to current NRC licensing requirements. Electrical equipment was seismically qualified per the requirement ofIEEE 344-75. Generally for equipment, the vertical ground acceleration specified was equal to the horizontal and is combined with the horizontal acceleration using the SRSS method. The mechanical and electrical equipment were purchased under specifications that include a description of the seismic design criteria for the plant. Page 3-28

3.1.3.3 Seismic Design of Tanks Tanks at Waterford 3 were designed to current licensing criteria. These criteria included the amplified frequency response of the impulsive fluid mass. Therefore, Waterford 3 has implicitly - satisfied the concerns raised for flat bottom tanks in Unresolved Safety Issue A-40, (see Reference 3.2) per NUREG-1233 (September 1989). The percentage of critical damping for welded steel plate assemblies in the Waterford FSAR applicable to tanks, is given as a percentage for both the OBE and SSE. Large tanks included in the IPEEE SMA effort include the Boron Management Holdup Tank A, B, C and D and the Diesel Oil Storage Tank A and B. The walkdown confirmed that there was no bolt degradation for the tanks, and the tanks meet the plant design basis with additional margin. 3.1.3.4 Seismic Design of Distribution Systems 3.1.3.4.1 Piping All seismic Category I Piping 1/2 inch or larger, was seismically analyzed as follows: a) All the Code Class 1 piping systems are analyzed by the Modal Response Spectra Method. b) All the Code Class 2 and 3 piping systems except as described in (c) below using either: Equivalent Static Load Method or the Modal Response Spectra Method. c) All Code Class 3 chilled water piping is analyzed by Chart Method. Some additional lines with a design temperature less than 275 F were also analyzed by Chart Method. The adequacy of the seismic design of the Reactor Coolant System components other than the main loop was determined by the Modal Response Spectra Analysis. The mathematical models employed in the analysis is in sufficient detail to reflect the dynamic response of all significant modes. All modes with natural frequencies in the range of 33 Hz and below are considered significant. In the analysis of complex systems where closely spaced modal frequencies are encountered, the responses of the closely spaced modes are combined by summing the absolute values method and, in turn, combined with the responses of the remaining significant modes by the square root of the sum of the squares method. Modal frequencies are considered closely spaced when their difference is less than 10 percent of the lower frequency. Dynamic loads of a piping system are calculated using the acceleration values of the floor response spectra with an appropriate damping factor. These loads are then used in an elastic analysis to calculate stresses. Page 3-29

For all ASME Code Class I piping, the loadings and, in turn, the primary stresses produced by inertial effects are determined by applying the modal responses, mode-by-mode, to the piping system with the supports / restraints maintained " fixed" The loadings and, in turn, the secondary stresses produced by the relative displacements of the piping supports / restraints are determined by imposing the relative displacements on the piping system. The displacements are imposed in a manner to produce maximum primary plus secondary stresses in the piping when the total inertial effects are added to the effects resulting from the imposed relative displacements. There is no Reactor Coolant System piping routed between buildings. For all ASME Code Class 2 and 3 piping differential displacement between buildings is taken into account in the seismic analysis but displacements at different support points within a structure are not considered, because they are negligible. This is based on a review of all ASME Code Class I calculations that indicated that the maximum relative displacement between the two extremes of any calculation. For subsystems that would normally be analyzed by the Modal Response Spectra Method, if the first mode period of the piping is 70 percent or less of the first mode period of the structure (i.e., peak of the floor response spectra) a modal period of the structure was not performed. Equivalent Static Load Method, ESLM, is used as specified in Standard Review Plan Section 3.7.2. In all cases the stiffness matrix method of natural mode analysis is employed to determine first natural period. The preset value for the maximum allowable period is 0.20 seconds that is not greater than 70 percent of the first mode period of the structure. The ESLM is made directly, using an accelerating value of 1.5 times the maximum value of the floor response spectra in the period range equal to 0.20 seconds or less. The acceleration value that is multiplied by 1.5 is taken from Floor Response Spectra at a period of 0.20 seconds. To justify the ESLM analysis procedure for piping, three sample problems were prepared using both ESLM analysis and modal response spectra methods. The static analyses used 1.0 g horizontal and 0.666 g vertical accelerations. The dynamic analysis utilized seven modes for sample one and five modes for sample two and three. For all modes the horizontal acceleration was taken as 1.0 g and vertical acceleration 0.666 g. The periods for the analyzed modes of all systems were between 0.20 seconds and 0.08 seconds. In all cases the maximum computed stress was higher for the ESLM analysis than for the dynamic analysis, with the maximum stress occurring at the same point for both methods in all three problems. Since the ESLM yields higher stresses than the dynamic when t t. :he same acceleration, the use of 1.5 times the peak of the floor response spectra for piping periods less than 70 percent of the first period of the structure is conservative by at least a 1.5 factor for the three systems analyzed by both methods. Page 3-30

r The chart method of analysis consists oflocating restraints such that the period of the first mode of vibration will not exceed the preset value of 70 percent of the first mode period of the supporting structure. This method involves the use of appropriate and comprehensive charts and tabulations that include correction factors for the effects of concentrated loads, branch connections and other effects. The piping system is studied for loading effects in each of the three coordinate directions to assure that it is adequately restrained in all directions. An additional analysis is performed to evaluate the thermal effects of the restraints of the system. This is done by means of charts that define the minimum distance required for placing restraints adjacent to any expanding leg to stay within allowable stress limits. The computer program (PIPESTRESS 2010) was used for the modal analysis and simplified dynamic analysis using the same stiffness matrix method. The program automatically determines forces, moments and deflections in the three coordinate directions and the stresses applied to both bending moment and torsional moment. The adequacy of seismic loadings used for the design of the Reactor Coolant System Piping were confirmed by the methods of dynamic analysis employing time-history and response spectmm techniques. To account for possible dynamic interaction effects between the components of the system, a composite coupled model was employed in the dynamic analysis of the reactor, the two steam generators, the four reactor coolant pumps and the interconnecting reactor coolant piping. The analysis of these dynamically coupled multi-supported components utilized different time dependent input excitations applied simultaneously to each support. The representation of detail of the reactor vessel assembly used in this coupled model included sufficient detail of the reactor internals to account for possible dynamic interaction from the Reactor Coolant System to the internals. 3.1.3.4.2 Cable TrayandConduit All cable tray and conduit supports were qualified by analysis, using the response spectrum method. Maximum cable tray spans and physical properties were selected. Two types of supporting systems have been used.

a. Rigid supports (defined as having a fundamental frequency > 33 Hz)

b. Non-Rigid supports (defined as having a fundamental frequency < 33 Hz)

For rigid supports, g values were selected for a static analysis of the support based m thg 1 response of the system. For non-rigid supports, a system frequency was selected such that 1.5 times the corresponding response g was within the system capacity. The support frequency was then calculated and a dynamic analysis was performed using a simplified three dimensional model having the desired support and cable tray system frequency. The g value obtained from averaging the responses was then used in a static analysis of the supports. Page 3-31

i ' Supports in the Reactor Auxiliary Buildmg were ongmally all designed rigid.' Supports in the Reactor Containment Building were originally all designed non-rigid. For the Fuel Handling Building, both types of suppons were used. 3.1.3.4.3 HVAC HVAC suppons were qualified by analysis similar to the cable tray supports. Maximum allowable loads on HVAC supports were 50% of the duct weight plus 25% of the support weight. 3.1.3.5 Seismic SpatialInteraction Issues Seismic interaction with non-seismic equipment was addressed in the Wateriord 3 FSAR. Seismic interaction with block walls was also evaluated by Waterford 3 in response to IE Bulletin 80-11. Seismic spatial interactions were addressed in the IPEEE SMA performed as discussed in Section 3.1.4.2.5 of this report. Seismic spatialinteractions_were also the subject of USI A-17. This evaluation satisfies the seismic interaction issues addressed in USI A-17. 3.1.3.6 Structural Damping The damping factors used in the analyses of the various structures, equipment and distribution systems at Waterford 3 are as follows: i Table 3.4 - Damping Values for Waterford 3 PERCENT DAMPING PERCENT DAMPING FOR OBE FOR SSE Soil 7.5 7.5 Reinforced Concrete Frames, and 2.0 5.0 Buildings Concrete Equipment Supports 2.0 5.0 Bolted Steel Framed Structures 2.5 2.5 Welded Steel Framed Structures 2.0 2.0 Welded Steel Plate Assemblies 1.0 1.0 Steel Piping Systems 0.5 1.0 Steel Piping Systems < 12 in. 0.5 1.0 i l Page 3-32 l

3.1.4 Evaluation of Seismic Capacities of Components and Plant 3.1.4.1 Overall Approach P The project used a SMA as the means to investigate the seismic external event of the IPEEE. The final SMA documented in Reference 3-7 represents the plant configuration as it exists at the beginning of the project, except that changes made in connection with the project have been included in this report. Voluntary actions that Waterford 3 implemented befcre assembling the licensing submittal for the IPEEE are also documented. The success path equipment was identified based on operational and systems considerations for Waterford 3. Operational and systems considerations have been implicitly considered by utilizing the Waterford procedure OP-902-008. This is the symptom based emergency operating procedure used by Waterford plant operators to maintain safety functions and shut down the plant during an accident. This includes prioritized success paths and exceeds the requirements of NP-6041-SL. The project approach utilized a screening walkdown that emphasized the importance of experienced Seismic Review Team Engineers familiar with equipment and structural performance in real earthquakes. Each SRT was composed of one Stevenson & Associates engineer, one Waterford 3 seismic engineer, and one Waterford 3 systems engineer or operator that also served as a plant guide. A 100 percent walk by of all equipment on the Condensed SSEL was performed. The interface between the systems engineer and seismic engineer is important since these parts of the process depend on each other for input and guidance. For example, the Waterford 3 system engineer and/or operator gave guidance from the systems part of the analysis in terms of the components and systems that must function for the equipment to perform its safe shutdown function. Before the screening walkdown portions of the Waterford plant seismic criteria, drawings, and documents were reviewed. The intent of this effort was to define a technical baseline from which the SMA was performed. This review included the design documentation for the structures, equipment and components. The Waterford 3 plant is a reduced scope IPEEE facility. Therefore, the Review Level Earthquake (RLE) for the Seismic Margins Assessment (SMA) is the plant's licensing basis Safe Shutdown Earthquake (SSE). The vast majority of the equipment included in the SSEL have existing seismic qualification data to the SSE level at Waterford 3. This data became the baseline for the walkdown. The walkdown included a detailed review of a sample of equipment in a given category (i.e., Motor Control Centers) to have reasonable assurance that the equipment has been installed to the existing criteria. The remaining equipment included in the walkdown was inspected primarily for seismic interaction as discussed in Section 3.1.4.2.5 of this report. The walkdown and review effort assured that the equipment seismic capacity was not reduced through modification or design change, and has not been reduced through programmatic measures. Page 3-33

Relay failure and chatter effects have been explicitly eliminated from the IPEEE seismic effort for reduced scope plants. The only exception to this is if a seismic intera:: tion is identified during the walkdown that is potentially damaging if the equipment contains protective or essential relays. Soil failure evaluations have also been explicitly eliminated from the IPEEE seismic effort for reduced scope plants. During the screening walkoown, modi 9ed Seismic Verification Data Sheets were employed. The more detailed SMA caveats and anchorage review check lists were brought with the SRT in the field during the walkdown for reference purposes. Equipment not screened were evaluated further to insure that it met the Waterford 3 seismic licensing basis. The seismic input and allowable stress criteria used to evaluate unscreened equipment were per the Waterford 3 original design basis. Meeting these criteria closed out the potential issue from further evaluation. The approach taken for the entire peer review process was to have peer review as a concurrent activity during the entire project, rather than a subsequent activity after the project is almost complete. The peer reviewers began their reviews with the project plan. Selected reviews were also conducted at key milestones in the overall effort. 3.1.4.2 Screening Criteria The requirements of the IPEEE SMA for a reauced scope plant such as Waterford 3 is that the plant meet its original seismic design basis requirements. The design basis requirements include: the equipment seismic capacity is greater than demand, the constmetion adequacy of the equipment, and anchorage adequacy. The Waterford 3 IPEEE SMA also addressed seismic spatial interaction. The specific criteria for satisfying these requirements are discussed in this Section. 3.1.4.2.1 Seismic Capacity Vs. Demand Seismic capacity vs. demand for the equipment was addressed for the plant in the FSAR. The design basis for safety-related equipment meet current NRC licensing requirements. The question of seismic capacity vs. demand for equipment was therefor ejudged to be acceptable on a generic basis and this requirement was not included as an item on the equipment lists or check lists. 3.1.4.2.2 Equipment Construction Adequacy EPRI NP-6041-SL contains criteria for equipment construction adequacy for various equipment categories in the form of caveats. A representative sample of equipment was checked for similarity to the equipment that had been subjected to strong motion earthquakes or seismic Page 3-34

( analysis or testing and also met the intent of the specific caveats for that class of equipment. Cavus define vulnerabilities observed in strong motion earthquakes or seismic tests. If equipment-specific seismic qualification data were used, then any specific restrictions or caveats for that qualification data apply instead. The guidance for the sample selection is contained in Section 3.1.1.9.1. The SRT member engineers have significant experience and training in the area of seismic "^ adequacy of equipment and equipment performance during earthquakes. They are very familiar with the issues in regard to equipment adequacy from their experience. The SRT engineers inspected each equipment item, and any detail they felt was seismically vulnerable was addressed. 3.1.4.2.3 Anchorage Criteria An assessment of the anchorage adequacy was performed on a representative sample of equipment included on the safe shutdown list. This included an assessment of the seismic demand on the equipment anchorage (forces and stresses on the anchorage), the seismic capacity of the anchorage components (attachment of the equipment to the anchorage, the anchorage itself, and the development of the anchorage to the foundation), and whether the capacity of the weak link of the anchorage system exceeded the demand. The assessment also included whether the anchorage system had adequate stiffness. The guidance for the sample selection is contained in Section 3.1.1.9.1. To perform the anchorage assessment the guidance contained in Section 7 (Equipment Anchorage) of the SSRAP report (see Reference 3.4) and the URS anchorage report (see Reference 3.5) was used. Other sources for performing the evaluation were knowledge of the design basis ground response spectra, elevation of the equipment, and the available floor response spectra. The main guidance documents referred to above were supplemented when appropriate with Reference 3-6. SRTjudgment based on their experience was also used as discussed below. The discussion contained in Reference 3-5 was primarily used to estimate the seismic demand on the anchorage of mechanical and electrical equipment. The damping used met or was more conservative than the Reference 5 recommendations. The SRT used the floor response spectra when available for estimating the demand. An example of using more conservative assumptions than the criteria was the use of 2% damped spectra for the Waterford 3 evaluations. This is because 5% damped floor spectra were not available. Equipment weight estimates used engineering experience or the guidance given in Reference 3-5. Estimates of equipment fundamental frequency were also made during the course of estimating the seismic demand for some items. This estimate was very rough (for example this item has a frequency above 5 Hz) and could be made using the experience of the SRT members of tests and analyses and afler looking at the construction of the equipment. The primary source for estimating the seismic capacity of the anchorage is Reference 3-5. The details with regard to anchorage allowable loads and cracked concrete considerations are in this reference. Page 3-35 l

The load path from the center of gravity, c.g.,'of the equipment was analyzed (by judgment and/or ' calculation) to the anchorage and ultimately to the supporting structure. Anchorage components j above the foundation were seen and verified. The embedded or buried components were verified using plant specific documentation (e.g., drawings).- An ' anchorage calculation was not performed for the vast majority of equipment items. When the anchorage was obviously rugged and installed to the design basis, SRTjudgment assessed anchorage adequacy. This judgment was performed in the context of the above criteria. l -3.1.4.2.4 Distribution System Adequacy 3.1.4.2.4.1 Piping The walkdown effort concentrated on identifying certain construction details that have led to past . earthquake damage in industrial facilities. The limited plant walkdown of piping identified any of the following potential piping failure modes. Valve failure caused by impact resulting from large displacements of flexible pipes. Pipe failure caused by large displacement ofinadequately anchored equipment (e.g., tank) l Failure of small, stiff pipes attached to large, flexible pipes. 7 Failure of piping between buildings as a result oflarge relative displacement caused by rocking or sliding of the buildings. l 1 Failure of brittle connections (e.g., threaded pipe), eroded or corroded piping, and ] brittle cast iron piping. A limited plant walkdown was conducted by the seismic review team using the sampling guidance in Section 3.1.1.9.1. An inspection of typical safety-related piping runs at Waterford 3 that were l accessible provided a reasonable sample to verify that no problems exist. Piping sections between buildings were carefully investigated. 3.1.4.2.4.2 Cable Tray and Conduit { Extensive tests performed on cable trays have shown that high capacities exist. Cable trays in . nuclear power plants in the past have not always conformed to design drawings and calculations. Therefore, a sampling walkdown of cable tray raceways assessed the as-built seismic capacity of the Waterford 3 cable trays and conduits. The principal failure modes of concern include failure of taut ' cables due to large relative displacement (e.g., relative motions between buildings), and failure of connections (unistmt clips or in threaded rods). During the plant walkdown, example cable trays were verified to have adequate support. Page 3-36

3.1.4.2.4.3 HVAC The only HVAC-related problems were with loss of anchorage of fans and blowers and possibly fan blade misalignment resulting in rubbing and banging after the earthquake. Fans associated with safe shutdown are on the SSEL and were part of the walkdown effort. This report concludes that the dominant failure modes for HVAC systems are anchor bolt and support failures. Ducting failure is not a major concern in SMAs for reduced scope plants because ofits high capacity. A briefinspection using the sampling methodology described in Section 3.1.1.9.1 verified that the equipment has proper anchorage. Shock-mounted HVAC equipment was to be evaluated, however this condition was not encountered at Waterford 3. For ducting that spans between buildings, potential failure because oflarge relative displacements were evaluated.

3. L 4.2.5 Seismic SpatialInteraction The Safe Shutdown equipment was evaluated for the effect of possible seismic spatial interactions with nearby equipment, systems, and structures. The review assures that these possible interactions do not cause the equipment to fail to perform its intended safe shutdown ft'nction.

There was a 100% walkdown of safe shutdown equipment for seismic interactions. To verify the seismic adequacy of an item of mechanical or electrical equipment using the EPRI NP-6041-SL methodology, there was a confirmation that there are no adverse seismic spatial interactions with nearby equipment, systems, and structures that could cause the equipment to fail to perform its required safety function. The interactions of concern are (1) proximity effects, (2) stmetural failure and falling, (3) flexibility of attached lines and cables, and (4) seismic induced fire and flooding. It is the intent of the SMA seismic mteraction evaluation to identify and evaluate real (i.e., credible and significant) interaction hazards. The interaction evaluations focused on areas of concern based on past earthquake experience. Equipment not specifically designed for seismic loads was not arbitrarily assumed to fail under earthquake loads. The SRTs differentiated between likely and unlikely interactions, using their judgment and knowledge of past earthquake experience. Although relay reviews were out of scope for the Waterford 3 SMA, attention was given to the seismic interaction of electrical cabinets containing relays. If the relays in the electrical cabinets were found to be essential, i.e., the relays should not chatter during an earthquake, then impact on the cabinet was considered as potentially unacceptable seismic interaction. The existing seismic interaction criteria for the plant was reviewed. Attention was focused on the following architectural features during this review and subsequent walkdowns: Page 3-37

Control Room Ceiling - T-bar suspended tiles, recessed fixtures, and sheet rock are used in some plant areas (such as the control room). Seismic capabilities of these ceilings may be low. The SRT Engineers checked for details known to lead to failure such as open hooks, no lateral wire bracing, etc. However, falling of relatively small light fiber board tiles does not - constitute an unacceptable interaction. Ligh_t Fixtures - Normal and emergency light fixtures are used throughout the plant. Fixture designs and anchorage details vary widely. Light fixtures may possess a wide range of seismic capabilities. Pendant-hung fluorescent fixtures and tubes pose the highest risk of failure and damage to sensitive equipment. The SRT Engineers checked for positive anchorage, such as closed hooks and properly twisted wires. t Typically, this problem is not a lack of strength; it is usually because of poor connections. Emergency lighting units and batteries can fall and damage safe shutdown equipment due to impact or spillage of l acid. Platform Gratings - Unrestrained platform gratings and similar personnel access provisions may pose hazards to impact-sensitive safe shutdown equipment or components mounted on them. Some reasonable positive attachment is necessary, if the item can fall and thereby interact with equipment being evaluated. Unreinforced Masonry Walls - Unreinforced, masonry block walls were evaluated for possible failure and potential seismic interaction with safe shutdown equipment unless the wall has been seismically qualified as part of the IE Bulletin 80-11 program. The SRT Engineers reviewed the documentation for IE Bulletin 80-11 masonry walls to determine which walls have and which walls have not been seismically qualified during that program. This determination was made by Waterford 3 Engineering. Non-Seismic Category I Structures - If any safe shutdown equipment is in non-Seismic Class I stmetures, then potential structural vulnerabilities were identified. Page 3-38

Distribution systems were also checked for possible interaction problems caused by large relative [ motion between the systems and building structures, as discussed in section 3.1.4.2.4. 3.1.4.2.6 Tanks andHeat Exchangers Vertical tanks were part of the IPEEE SMA at Waterford 3 because of the occurrence of tank failures in past earthquakes. Much research has been performed to provide guidance and evaluation methods for vertical tanks. The procedure used for resolution ofIPEEE SMA uses current methods to provide justification of the structural integrity of vertical tanks for the Safe Shutdown Earthquake. 3 There are several key technical issues addressed in this evaluation regarding the seismic demand on the tank, the seismic capacity of the tank, and the tank critical components. The following discussion provided the background for the SMA evaluation oflarge flat bottom tanks at. Waterford 3. The response of a vertical tank to a seismic event is a combination of sloshing fluid and the impulsive mode from fluid-structure interaction. The sloshing of the fluid at the top surface contributes to over topping of the tank as well as producing loads on the tank roof. Adequate freeboard should be provided to prevent the roof failure. The sloshing effect occurs at very low frequencies and damping values. The impulsive mode includes the tank shell responding to seismic events at frequencies associated with the shell modes of vibration. This response includes the tank and its contents moving together. Though the tank shell may be rigid under empty conditions, a tank shell and its contained fluid act together become flexible. This tank flexibility usually produces a response in the amplified region of the response spectra. This mode

contributes significantly to total seismic response.

The seismic demand on the tank is in terms of the base shear and overturning moment produced at. I the tank base. Recent technical research has produced several simplified procedures that develop the response of fluid-filled tanks. The evaluation procedure for vertical cylindrical tanks includes l the sloshing and impulsive modes of the tank. The USI A-40 issues regarding tanks were based on the assumption that several tanks in existing nuclear facilities were designed considering that the impulsive mode of the tank shell and fluid i were rigid. In some instances the original design of these tanks make this erroneous assumption. l The evaluation for the IPEEE SMA will resolve all USI A-40 issues regarding these tanks. The tanks at Waterford 3 were designed considering the impulsive mode as rigid, because of the i modern vintage of the facility, therefore this issue is not a concern at Waterford 3. The existing documentation will be reviewed and it was determined that the following issues described below were adequately addressed. 1. The overturning moment capacity is based on the compressive strength of the shell to resist buckling and the tension capacity of the anchor bolts and associated connections to the tank was greater than demand. 2. The shear capacity based on the sliding friction between the tank base plate and supporting foundation was greater than demand. t Page 3-39

3. The available freeboard was adequate to prevent damage to the tank roof from sloshing. The compressive strength of the tank shell is the major resistance to the overturning moment. There are two buckling modes common to tanks; " elephant-foot" buckling and diamond buckling will not occur at the SSE level at Waterford 3. The tensile capacity is limited by (1) tensile l capacity of the bolts, (2) the embedment of the bolts into the concrete foundation, and (3) the ability of the tank chair to transfer the tensile load to the tank shell. - The overall shear capacity of i the tank is based on the friction developed between the base plate and foundation and is greater than the shear demand. The anchor bolts need not be included in determining the shear resistance since the friction load path is much stiffer than the bolt load path. The failure of a tank roof to sloshing was prevented by insuring appropriate freeboard. Attached piping was neglected in computing tank responses. However, flexibility of attached piping was checked during the walkdown and it wasjudged that it can accommodate slight uplift expected in the tank base. Piping attached to upper parts of the tanks have the necessary flexibility to accommodate larger deflections, horizontal and vertical, expected in the tank. There were no major seismic issues regarding horizontal heat exchangers identified during the walkdowns. 3.1.5 Analysis of Containment Performance NUREG-1407 includes a requirement for considering early failure of containment function in the performance of the SMA. + The scope of this task consisted of a brief review of the primary containment, its internals and penetrations, and the auxiliary building, to insure that they meet the licensing basis SSE criteria. The primary purpose of the evaluation was to identify seismic vulnerabilities invoMng containment, containment functions and systems that are different from those in the IPE internal events evaluation. This included reviewing penetrations, hatches and locks, supporting actuation and control systems, and containment ventilation systems. These potential vulnerabilities are discussed in EPRI NP-6041-SL, The Containment Isolation valves were included in the equipment walkdowns. 3.2 USI A-45, GI-131, OTHER SEISMIC SAFETY ISSUES 3.2.1 USI A-45 (Decay Heat Removal) Because the seismic events evaluation showed that there are no seismic vulnerabilities, we conclude that there are no significant or unique seismic vulnerabilities in the decay heat removal function. Therefore, USI A-45 should be considered resolved with respect to seismic events. Page 3-40

3.2.2 GI-131 (Potential Seismic Interaction Involving the Movable In-Core Flux Mapping i System Used in Westinghouse Plants) i This issue is not apphcable, since Waterford-3 is a Combustion Engineering plant. 3.2.3 USI A-46 (Verification of Seismic Adequacy of Equipment in Operating Plants) Waterford-3 is not a USI A-46 plant. The issue of spatial interaction, however, has been addressed as part of the reduced scope seismic margins method. l The Waterford 3 IPEEE has not been used to evaluate any other USIs or GIs. REFERENCES 3-1. EPRI NP-6041-SL, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," Rev.1, Jack R. Benjamin and Associates, Inc. et. al., August 1991. 3-2. NUREG/CR-3480, "Value/ Impact Assessment for Seismic Design Criteria USI A-40," Lawrence Livermore National Laboratory, August 1984. 3-3. I&E Bulletin 80-11, " Masonry Wall Design," U.S. Nuclear Regulatory Commission, Washington, D.C.,1980. 3-4. SSRAP Report, "Use of Seismic Experience Data to Show Ruggedness of Equipment in Nuclear Power Plants," Senior Seismic Review and Advisory Panel, Sandia National laboratories Report DE 92-019328, Rev. O, June 1992. 3-5. EPRI NP-5228, " Seismic Verification of Nuclear Plant Anchorage, Volume 1: Development of Anchorage Guidelines; Vol. 2: Anchorage Inspection Workbook, URS Corpljohn A. Blume & Assoc., Prepared for Electric Power Research Institute, Palo Alto, CA, May 1987. t 3-6. American Concrete Institute (ACI), ACI 349-85 Appendix B - Steel Embedments, September 1,1985. 3-7. IPEEE Reduced Scope Seismic Margins Assessment (SMA) Report for Waterford 3, Prepared by Stevenson and Associates, December 1994. f Page 3-41

5. HIGH WINDS. FLOODS, AND OTHERS. The focus of this part of the IPEEE submittal is: "does the plant meet acceptance criteria listed in the 1975 SRP in terms of high winds, on-site storage of hazardous materials, and off-site developments?" The original licensing action for Waterford 3 considered all manner of winds, i floods, and industrial accidents. During the original licensing action, the NRC used the 1975 SRP as a basis for finding Waterford 3 acceptable in terms of external hazards with a few exceptions that the NRC reviewed and accepted. The exceptions are technical rather than substantive, e.g., the SRP recommended technique for calculating tornado loading on the shield building was not appropriate given the shallow dome roof of the shield building. The Waterford 3 staff reviewed the assumptions in FSAR Chapter 2 with respect to external event initiating frequency. Whenever that frequency is explicit in the FSAR, it is compared to the IPE Level 1 initiating event frequencies. If the external event frequency were small compared to the related Level 1 initiating event frequency, then the external event has an insignificant affect on our estimate of both the core damage frequency and distribution of containment release categories. When the external event initiating frequency is indeterminate, the corresponding part of Section 5 describes the plant design features and related basis with respect to external events. Note, the external events are typically a subset of the Level 1 initiating events. For example, storm damage that causes a loss of off-site power is part of the Level 1 assumption behind the transient initiator T5. As a further review of plant specific hazard data and licensing bases, the Waterford 3 staff also qualitatively reviewed the external events postulated in FSAR Chapter 2 against spectacular events in southeast Louisiana. We sought assurance that the postulated events bounded the spectacular events since initial plant startup. i The final part of the review involved re-visiting the 10 CFR 50.59s written since initial plant start-up regarding changes that exposed the plant to new external hazards, e.g., a new hydrogen gas pipeline across the site. The probability and consequences of the new configuration is qualitatively related to either FSAR Chapter 2 assumptions or IPE assumptions. Insignificant changes in this group are those that change neither the estimate of the core damage frequency, nor the distribution of containment release categories. In summary, Waterford 3 found no high winds, floods, or off-site industrial facility accidents that significantly alters the Waterford 3 estimate of either the core damage frequency, or the distribution of containment release categories. Waterford 3 further concludes that the plant is in conformance with the 1975 SRP that pertains to high winds, on-site storage of hazardous materials, and off-site developments. Page 5-1

5.1. HIGli WINDS. 5.1.1. Plant-Specific Hazard Data and Licensing Basis. FSAR Section 2.3.1.2.2 describes the hurricanes considered during Waterford 3 licensing. The maximum sustained winds were measured at 98 mph during Hurricane Betsy (1965). The FSAR recounts that 26 hurricanes have passed New Orleans between 1886 and 1974, including three hurricane eyes. That translates to a 0.30 per year frequency for nearby hurricane passage. The plant structures defined as seismic Category I structures are designed for a maximum sustained wind of 200 mph at 30 feet above plant grade. Those same seismic Category I structures were designed to resist a tornado of 300 mph tangential wind velocity and a 60 mph translational wind velocity. 5.1.2. Identified Significant Changes Since OL Issuance. Hurricane Jaun (1985) and Andrew (1992) would have rated inclusion in FSAR Table 2.3-137 had they occurred before initial plant startup. However, neither storm resulted in winds greater than Hurricane Betsy (1965). Although these storms were spectacular in their own rights, the consequences of a hunicane are still bounded by the current FSAR. These two occurrences since initial plant startup result in a frequency ofless than 0.23 per year; less than initially postulated by the FSAR Chapter 2 analysis.1 There has been one tornado on-site (1994), but it did not enter the protected area immediately surTounding the nuclear plant island structure. That one tornado in the last twelve months matches the frequency estimated in FSAR Section 2.3.1.2.4. Waterford 3 reviewed tornado protection features as documented in FSAR Table 3.5-3 for leads in a search for changes that might significantly affect core damage frequency with respect to high winds. The search yielded no significant changes to any of the features described on Table 3.5-3. In addition, the plant has been declared in conformance with Regulatory Guide 1.76, Design Basis Tornado for Nuclear Power Plants (re FSAR Section 3.3.2.1 and FSAR 1.8). 5.1.3. Plant Robustness in Relation to 1975 SRP Criteria. This part of the review focuses on the protection offered by the plant design against tornadoes. j The FSAR identifies all safety-related structures, systems, and components that need protection from externally generated missiles. All safety-related systems and components as well as stored nuclear fuel are within tornado missile-protected structures or they have tornado missile barriers. An exception is a portion of the emergency feedwater system pipe and portions of the wet cooling j 1 FSAR Table 2.3 137 ( five hurricanes in 10 years) and FSAR 2.3.1.2.2 (55 tropical or better storms in 107 years) Page 5-2

towers (i.e., the ultimate heat sink). The SRP refers to several standards and techniques as one acceptable basis for plant design features that protect against tornadoes. But, the NRC found the overall design of safety-related stmetures at Waterford 3 acceptable while at the same time considering those same standards and techniques. i Category I structures exposed to tornado forces and needed for the safe shutdown of the plant l were designed to resist a tornado of 300 mph tangential wind velocity and a 60 mph translational wind velocity. The simultaneous atmospheric pressure drop was assumed to be 3 psi in 3 seconds. The tornado tangential velocity and translational velocity are summed algebraically, and applied on the entire building stmeture. Shape factors and drag coefficients are based on the procedures outlined in ASCE Paper No. 3269. The NRC reviewed the procedures that were used to determine the loadings on Seismic Category I stmetures, and the NRC made a determination that the procedures were acceptable. Since that time, Waterford 3 has maintained this as the technique for determining the adequacy of safety-related structures. To assure integrity of safety-related structures in the face of tornadoes, Waterford 3 has committed to Regulatory Guides 1.13, " Spent Fuel Storage Facility Design Basis," 1.27, " Ultimate Heat Sink for Nuclear Power Plants," 1.115, " Protection Against Low Trajectory Turbine Missiles," and 1.117, " Tornado Design Classification." The commitment to these regulatoiy guides has not changed since initial plant startup. Thus, the plant is everything the SRP exoects in terms of tornado protection. The SRP refers to several standards and techniques as one acceptable basis for plant design features that protect against hurricanes. The NRC found the design of safety-related structures at Waterford 3 acceptable while considering those standards and techniques. Regarding hurricanes, Waterford 3 design is such that each safety-related structure can withstand a maximum wind of 200 mph at 30 feet above plant grade. The design wind specified has a velocity of 200 mph based on a recurrence of 100 years. The corresponding dynamic wind pressure and corresponding load on the structures was calculated in a manner the NRC found acceptable. Thus, Waterford 3 has maintained these design requirements since initial plant startup. l s i Page 5-3 +

i 1 4 5.2. FLOODS. 5.2.1. Plant-Specific Hazard Data and Licensing Basis. Safety-related equipment is housed within the Nuclear Plant Island Structure (NPIS). The NPIS is a reinforced concrete box stmeture with solid exterior walls and is flood protected up to elevation +29.25 feet MSL. Flooding as a result of the probable maximum hurricane (PMH) and the instantaneous break of the levee were analyzed, and the maximum water levels at the NPIS were determined to be 25.4 fl MSL and 27.6 fl MSL respectively. The maximum combined static and dynamic loads would increase linearly from zero at elevations 25.4 ft and 27.6 ft to 493 psf and 630 psf at elevation 17.5 MSL respectively. These conditions result in water levels and loads being below the design criterion of flood protection to elevation 29.25 fl MSL. In addition to the 29.25 ft flood protection feature, Technical Specification 3.7.5 and the corresponding operating procedures require securing flood tight doors when the Mississippi River exceeds 27.0 fl mean sea level USGS datum. Roof design has been reviewed according to the criteria for load combinations listed in FSAR Table 3.8-39, Formula 5. Both the six hour probable mazimum precipitation (PMP) giving the maximum intensity, and the 48 hour PMP giving the maximum accumulation have been considered. All roofs of safety related structures can safely store the maximum possible ponding resulting from the PMP. FSAR Figure 2.4-8 shows the locations and sizes of all roof drains and scuppers, and the heights of parapets. 5.2.2. Identified Significant Changes Since OL Issuance. LDCR No. 92-0010 was issued to correct the design flood level in the FSAR. The FSAR had stated that the design flood level was elevation 30 ft However, the top of the floodwall was l surveyed at elevation 29.27 ft Since under the most severe flooding condition, the highest level the water will reach is 27.6 ft., the safety related equipment is protected from flooding. The exterior walls of the NPIS were designed based on calculation 6W12-RAB-002(Q). The actual design stresses are far below the allowable design stresses. Therefore, the extra buoyancy and earth pressure on the walls due to 9 inches of settlement of the NPIS will not affect the design of the walls. The original analysis conducted to evaluate effects of Standard Project Storm (SPS) on cooling tower areas is still considered valid. The new PMP analysis (re Generic Letter 89-22) did not update either the data or the guidelines employed in the SPS study. The Generic Letter was intended for future plants and did not require a change to plants' design basis. Page 5-4

No events since plant startup lead anyone to believe that the levees near Waterford 3 are any more vulnerable than during initial licensing. Levee slumps have occurred downstream of Baton Rouge since initial plant startup. But, the Army Corps of Engineers lined the levees on the Waterford side of the Mississippi with concrete in 1993. That feature gives the levees greater stability. Besides that, the Waterford 3 design features to protect against an external flood greatly over-estimates the likely water level on-site after a levee break. Thus, the external flooding hazard associated with Mississippi River levees is at worst unchanged and probably less since initial plant startup. The discussion in Section 5.1 above implies that the frequency of a hurricane storm surge is also essentially the same as assumed during initiallicensing. As discussed above, the plant is still designed to be flood resistant to 29.25 MSL. 5.2.3. Plant Robustness in Relation to 1975 SRP Criteria. Three possible types of flooding were analyzed: (1) probable maximum hurricane (PMH) surges; (2) levee failures during Mississippi River floods; and (3) local intense precipitation. External water levels around the turbine building six or more inches deep will leave electrical equipment in the Turbine Building Switchgear Room disabled. Floods that reach 25' MSL will put the instrument air, auxiliary feedwater (not emergency feedwater inside the NPIS, see below), condensate and main feedwater system out-of-service should the Turbine Building switchgear survive that long. The Level I has already included secondary plant initiators as well as an internal flood analysis. The frequencies associated with the Level 1 and its flood analysis overshadow the frequencies associated with severe external events like levee failures. Thus the focus of the following analysis is on the design features of the Nuclear Plant Island Structure (NPIS) done in terms of design basis external events and the related Standard Review Plan sections. By itself, the NPIS contains all the equipment necessary to bring the plant to a safe shutdown condition and keep it there. 5.2.3.1 Probable Maximum Hurricane Surges Two credible approaches of a hurricane surge to the site were considered: (1) up the Mississippi River from Head of Passes; and (2) from the open Gulf across the low-lying wetlands to the south. The upriver path is the most critical. The analysis assumed that the PMH would occur coincident with a hypothetical severe flood on the Mississippi River. The flood used was one which was developed by the Corps of Engineers and the National Weather Service, producing a 3 discharge of 1,250,000 ft /sec south of Red River Landing. The maximum water level run-up at the NPIS, assuming an instantaneous levee failure, for this event was 25.4 ft MSL. i Page 5-5 1

I i I 5.2.3.2 Probable Maximum Flood-Induced Levee Failure Waterford 3 estimated potential flooding from rainfall over the Mississippi River basin upstream - f the site. The probable maximum flood (PMF), which is the hypothetical flood that is the most j o severe precipitation-induced flood reasonably possible, was estimated to produce flow of 5 million 3 ft /sec at Red River Landing. This estimate was made by using 165% of the Corps of Engineers 1 3 project design flood (PDF) estimate of 3,030,000 ft /sec at the same location. A flood of this - i magnitude would overtop the Mississippi River levees upstream of Waterford 3 and because of-the resultant spillage, produce water levels equal to or less than those associated with the levee. 1 PDF which produces a water level within the levees of 27 ft MSL. This level is lower and less critical than that estimated for a hurricane surge. Waterford 3 also performed an analysis of an instantaneous failure of the levee with a river stage ' of 30 ft MSL. This analysis resulted in an estimated flood level higher than that calculated for the-i hurricane surge. The calculated maximum water level mn-up at the NPIS was 27.6 ft MSL. l The NRC reviewed the analysis and concluded that the estimated run-up level is conservative. The NPIS has been designed for a maximum flood level of 29.25 ft MSL. Since this is 2.4 ft higher than the maximum water level conservatively calculated assuming the most critical flood conditions, the NRC concluded that the flood analysis for the NPIS meets the criteria suggested in 1 Regulatory Guide 1.59, " Design Basis Floods for Nuclear Power Plants," and Regulatory Guide. l 1.102, " Flood Protection for Nuclear Power Plants." l 5.2.3.3 Local Intense Precipitation Waterford 3 is located so that, with the exception of the cooling tower basin areas, nmoff from ' local intense precipitation will not affect its safety, External walls are flood proofed to elevation 29.25 ft MSL. This elevation is a minimum of 12.5 ft above plant grade and is far above any ponding which could be expected due to severe rainfall up to and including the probable maximum precipitation (PMP). The PMP for various durations is as follows: i Duration Amount (hr) m ' in i 6 0.78 30.7. 12 0.88 34.6 l 24 1.00 39.4 48 1.10 43.5 Waterford 3 has wet and dry cooling towers which are open at the top. There are two open l cooling tower areas A and B. Local intense precipitation which falls directly over these open i areas plus runoff from adjacent roofs will accumulate and pond on the floors of the dry cooling i l. 1 Page 5-6 [ t

~ ~, ~. =_ - ] tower areas. A combination of floor drains and a network of drainage piping will convey this water to two sumps where a set of duplex pumps in each sump will remove water from the '~ cooling tower areas. In Amendment 21_to the FSAR, Waterford 3 performed a revised analysis ofpotential flooding in _j . cooling tower areas A and B. In this analysis roof drains were assumed to be 33 percent blocked. l ' The analysis also assumed that one of the sump pumps in each cooling tower area would be - . inoperable during a probable maximum precipitation (PMP) event. This revised analysis resulted i in lower ponding levels in the cooling tower areas. These levels, however, were not low enough to prevent flooding of the motor control centers which are located on the floor of the dry cooling towers. To further reduce ponding levels in the cooling tower areas, Waterford 3 proposed to_- l allow water to flow into and pond in the Fuel Handling Building via eight 4-in diameter openings : between the cooling tower areas and the Fuel Handling Building. Waterford 3 estimated that by -r allowing water to pond in the Fuel Handling Building, a maximum of 1.6 ft ofwater will pond in l the cooling tower areas and in the Fuel Handling Building. The maximum height to which water can pond in the cooling tower areas before flooding of essential portions of the station service transformers occurs is 3.0 ft, and for the motor control centers it is 1.71 ft. The NRC reviewed the material presented by Waterford 3 and performed an independent analysis. The NRC concluded that, with the eight 4-in. diameter opening installed as indicated, water l depths in the cooling tower areas will remain below 1.6 ft following a PMP event and will thus not affect the safe operation of Waterford 3. [ Waterford 3 also considered an event which included an OBE, which fails the cooling tower area i sump pumps, in combination with a standard project storm of 96 hour duration. This event was i determined to produce a total rainfall of about 23 inches and would result in a ponding depth of about 1.9 ft in the cooling tower areas assuming that all four sump pumps are inoperable. This water level is higher than the maximum allowable ponding depth of 1.71 ft. Waterford 3 committed to provide a portable pump with a pumping capacity of 100 gpm and sufficient head to pump over the cooling tower wall. Waterford 3 also committed to include the pump in the i surveillance testing program which would include a demonstration at least once per refueling that the pump will circulate water. Additionally, Waterford 3 committed to store the pump on pallets away from any non-seismic Category I equipment, and, as part of the station's emergency j procedures, to incorporate a provision for setting up the portable pump within 6 hours of a seismic event if the installed pumps fail. The NRC, based on this resolution, concluded that with l respect to potential flooding of the cooling tower areas, Waterford 3 meets the requirements of 10 CFR 50, Appendix A, GDC-2, and the criteria of Regulatory Guides 1.59 and 1.102. l NRC Generic Letter 89-22 communicated the revised Probable Maximum Participation (PMP) data developed by the National Oceanic Atmospheric Administration (NOAA) of the National-- ( i t i Page 5-7 . =

y 1 2 Weather Service. Waterford 3 performed a study to evaluate the potential for increased roof loads, cooling tower areas ponding, site drainage, and the plant area flooding. The analysis concluded the following: The revised 6-hour and longer duration,10 square mile PMP depths used for flooding analysis, are slightly higher than the values previously used and specified for the Waterford 3 Design Basis. The new 6,12, and 24 hours PMP depths are 32.0,38.7 and 47.1 inches compared to previously used depths of 30.7,34.6 and 39.4 inches respectively. However, the revised PMP depths, when modified specifically for the Waterford 3 catchment area, reduce to 22.44,28.51 and 35.24 inches for 6,12, and 24 hour durations respectively. These PMP depths are lower than the original design basis. Therefore, the revised criteria has no adverse impact on Waterford 3 site flooding. The short-duration, point PMP intensities are higher than the intensity originally used for design of the plant. The site drainage design is based on a rainfall intensity of 81/4 inches per hour, 50 year recurrence interval, whereas the roof runoff and the cooling tower area ponding calculations are based on 6-hour rainfall distributed hoarly ranging from a low of 3.08 inch in the first hour to 11.67 inches in the 4th hour. The revised intense PMP depths for the most critical 5 min.,15 min.,30 min., and I hour durations are 6.21,9.70,14.16 and 19.4 inches respectively. These PMP depths are based on the most severe storm ever recorded with and upward moisture maximization adjustment, which makes the PMP worse than even the 100 year recurrence storm. The local site flooding due to the new intense PMP will not have any adverse effect on the nuclear plant island stmetures because the exterior walls of the plant are flood protected up to El. +29.25 ft (MSL),12.5 ft. to 15.5 ft above grade, which is far above any ponding that could be expected due to a severe intense rainfall up to and including the revised PMP and assuming blocked culverts. The ponding on the RAB and the Fuel Handling Building roofs will increase above the e values originally expected due to the revised PMP data. Analysis for the RAB and the Fuel Handling Building roof slabs indicates that the original design is adequate to safety withstand the additional roof ponding loads. Safety-related MCCs and transformers are located in the cooling tower areas. The critical depth ofwater in this area before flooding of essential portions of this equipment is 1.71 l ft This depth will be exceeded for the revised PMP intensity, which is based on adjusted worst ever recorded storm, if the Fuel Handling Building sump pump and one of the two cooling tower area sump pumps are assumed operable. However, for the estimated 50 year recurrence rainfall, the maximum ponding depth is 1.70 ft. which is less than the i 2 PEIR 20066 I i Page 5-8

l critical depth. If both cooling tower area sump pumps are considered operWe in addition to the Fuel Handling Building sump pump, then the maximum ponding degQ for the worst ever recorded storm PMP intensity is 1.65 ft. which is less than the critical depth. Considering the conservatism in the revised PMP criteria and availability of an additional portable pump, it was concluded that additional measures are not required due to the revised PMP criteria. The safety-reN llack-up Fuel Pool Heat Exchanger is located in the Fuel Handling Building. 'T witom floor of that building is at the same level as the bottom of the cooling to- >. Water roughly two or more inches deep on the floor of the cooling towers will !!ow laterally through small pipes in to the FHB sub-basement. Thus, the maximum ponding depth is the same in the cooling tower area and in the sub-basement of the FHB. All equipment and instrumentation in the FHB are located well above the maximum ponding depth of 1.70 fl. and the critical depth of 1.71 fl. j 5.3. TRANSPORTATION AND NEARBY FACILITY ACCIDENTS. 5.3.1. Plant-Specific Hazard Data and Licensing Basis. FSAR Sections 2.2.1,2.2.2,2.2.3, and 2.3.4 describe the location and distances ofindustrial and transportation facilities. The nature and extent of activities conducted at nearby facilities, including the products and materials likely to be processed, stored, used, or transported have been identified and evaluated. Sufficient statistical data has been documented to establish a basis for evaluating the potential hazards to the plant. 5 The potential hazards are of two types; (1) toxic gas hazards, and (2) fire and explosion hazards. These hazards can result from various manufacturing industries, pipelines, roads and railroads, and ship traffic in the Mississippi River. Technical Specification 6.9.1.9 and 6.9.1.10 help us assure that there is an appropriate design basis for the toxic gas hazard protection features of the plant and its procedures. The FSAR includes analyses of over-pressures caused by explosion hazards. Those numbers become part of the design input for safety-related structures. i t Page 5-9

i i 5.3.2. Identified Significant Changes Since OL Issuance. In April 1992, construction started on the Evangeline pipeline.3 This pipeline supplies natural gas to the Waterford 1, Waterford 2 and Little Gypsy LP&L fossil fired power plants. A 24-inch pipeline enters the Waterford 3 property to the south-east of the Waterford 3 buildings. The 24-inch pipeline connects to two 20-inch pipelines at a " pig trap station"just south of the Union & Pacific Railroad tracks. One 20-inch pipeline then runs parallel to and slightly west of the existing 26-inch Bridge!!ne pipeline to supply Little Gypsy. The other 20-inch pipeline then runs parallel to and just south of the railroad tracks and then parallel to the existing 16-inch LP&L pipeline to supply Waterford I and Waterford 2. The maximum natural gas flow rate in the 24-inch pipeline is 250E6 scf/ day. The maximum flow rate in each 20-inch pipeline is 150E06 scf/ day. With the exceptions of the " pig trap station" and the connection to the Waterford header, the Evangeline pipeline is buried for its entire length. The hazard that the new pipeline poses to Waterford 3 is that of a potential pipeline break followed by an explosion. The explosion shock wave is considered whenjudging the adequacy of Waterford 3 structures. A break in a natural gas pipeline does not present a control room habitability concern, per FSAR Section 2.2.3.3.1, because the natural gas is likely to explode or burn before reaching the main control room. FS AR Section 2.2.3.1.3.1 presents an analysis of the effects of a natural gas pipeline break and explosion. The new Evangeline pipeline has the potential to cause this type of accident. It should be noted, that this is an external event accident that has absolutely no effect on Waterford 3 safety-related structas. Calculation EC-M92-010 shows that the explosion hazards because of the new pipeline are bounded by the present FS AR analyses. The maximum over-pressure from an explosion on the new pipeline is 1.6-psi. The allowable over-pressure is 3.0-psi in the Waterford 3 design basis. Concluding Remarks Technical Specifications 6.9.1.9 and 6.9.1.10 require Waterford 3 to perfonn a survey and analysis of toxic chemicals and explosive hazards from chemical plants within the vicinity of Waterford 3. The first survey was performed in May 1988, reference 4, and the second survey was performed in July 1992, reference 3. These surveys are intended to ensure that the toxic chemicals and explosive hazards analyses of record remain bounding. Any new toxic chemicals or explosive hazards would need to be evaluated and resolved with the NRC. 3 LDCR 92-0441 Page 5-10

5.3.3. Plant Robustness in Relation to 1975 SRP Criteria. 13.3.1 Fire Postulated accidents involving fires can be evaluated in two principal ways with respect to the nuclear plant. First, there may be thermal effects due to a fire at the scene of an accident. Locations of all fixed storage facilities, pipelines, and transportation routes were examined and a determination was made that the distance to the plant in each case is sufficient such that the thermal effects with respect to the plant are acceptable. Second, postulated leakage or spillage of materials such as propane or other flammable hydrocarbons could lead to the formation and drift of clouds toward the plant. Analyses were performed in this regard and a fMng made that thermal effects due to delayed ignition and deflagration of flammable vy. sud are insignificant with respect to the plant. This conclusion is based primarily on the observation that the thermal fluxes and time durations of the vapor cloud fireballs are insufficient to produce appreciable heating of giant structures and equipment. For example, the maximum thermal fluxes are about 287 kW/m-in conjunction with a postalated LPG truck accident on LA 18, near the plant. However, the estimated time duration for the fireball is less than 10 seconds, which is more than two orders of magnitude less than what is needed to produce significant heating of plant stmetures. All other pcstulated delayed ignition of flammable vapor clouds involves similar short duration times, and distances from the plant which are significantly greater than 300-foot minimum distance estimated for the propane cloud relative to the plant in the event of a LPG truck accident on LA 18. In view of the above considerations, the NRC concluded that the thermal effects due to postulated fires in the vicinity of the nuclear power plant do not pose a significant threat to the safe operation of the plant. 5.3.3.2 Explosion Waterford 3 analyzed four events for explosive hazards. (a) Explosion of 300,000 barrel tanker containing gasoline passing on Mississippi River 120011 north of safety related structures. (b) Explosion of LPG truck carrying 10,500 gallons of LPG on route 18 passing a critical distance of 462 feet north, east, or west of nuclear plant island structures. (c) Explosion of Bridge line's 26 inch natural gas pipeline which is approximately 3168 fl from the plant. (d) Explosion of Union Carbide Propylene tank containing approximately Page 5-11 j

)j ~ i 6 4 5.78 x 10 lbs of propylene and located approximately 6300 ft from the plant. The over-pressure from these events were (a) 1.3 psi, (b) 3.0 psi, (c) 1.0 psi, and (d) less than 1 psig. Acceptable over-pressure in later hazard analyses has thus become less than or equal to 3.0 - psi. j 13.3.3 Toxic Gas Hazards- 'l 5.3.3.3.1 i Evaluation ofStationary Chlorine Sources .l i Potential hazards posed by stationary sources of chlorine were evaluated by comparing such sources to the allowable quantities listed in Table 1 ofRegulatory Guide 1.95. Waterford 3 was assumed to have a Type II control room. Waterford 3 control room has local detectors, a normal l air exchange rate of 0.06 vol/hr, and a measured leak rate ofless that 0.06 vol/hr. j The stationary source of chlorine posing the greatest potential hazard is a tank on the site of the -l Occidental Chemical Co., which contains 500 tons and is located 1490 meters from the Waterford i 3 control room. At this distance, the maximum allowable quantity calculated by log-log interpolation, in accordance with the guidance of RG 1.95, is 662 tons. the Waterford 3 control room therefore satisfies the guidance of RG 1.95. 5.3.3.3.2 Evaluation of Other Stationary Sources 1 The analysis ofpostulated accidents involving stationary sources of chemicals other than chlorine j were performed in accordance with the general guidance ofRegulatory Guides 1.78 and 1.95 and j utilized the detailed release and atmospheric transport model described in NUREG-0570, June 1979. The atmospheric transport and dispersion of the initial puff was calculated according to the general model presented in Regulatory Guides 1.78 and 1.95. The concentration of a toxic chemical inside the control room was based on a control room air exchange rate of 0.6 per hour. 1 The toxic chemical concentrations calculated inside the main control room were assessed against their "Immediately Dangerous to Life or Health" (IDLH) concentrations. ) -1 The analysis modeled the detection of ammonia by the ammonia detectors and of most other chemicals by the Broad Range Toxic Gas Detectors (BRTGDs). Credit was taken for odor i detection. Operators were assumed to don breathing apparatus two minutes after the alarm or i odor detection, whichever occurs first. Sets of meteorological conditions were constructed which included all combinations of stability class and wind speed. Since the quantity ofliquid that is vaporized increases with temperature, summer temperatures were assumed for the sake of conservatism. Stability classes E-G were 4 FSAR 2.2.3.1.3.3 Page 5-12

I assumed to occur primarily at n;ght. For these cases, the average ambient night-time temperature for the summer months, June through August, was calculated by taking the average of the mean temperatures and the mean minimum temperatures for each of these months. Since stability j classes A - D may occur in the daytime, average daytime temperatures were calculated for those cases by substituting mean maximum for mean minimum temperatures. Daytime ground temperatures were assumed to be 10 C higher than the air temperatures. Night-time ground temperatures were assumed to be the same as the ambient air temperatures. The frequency of occurrence of a given set of meteorological parameters for a given wind direction was calculated as follows. Each value of the annual averagejoint frequency of wind t speed and stability class was divided by the fraction of time the wind was in that sector (i.e., the - joint frequencies for the given compass direction were normalized to 1). Thus, each new value represented the probability of thejoint occurrence of that particular wind speed and stability class combination, assuming the wind is in the given sector. Accidents under each set of meteorological conditions for the given wind direction were then modeled. The control room was assumed to be habitable if the concentration inside did not exceed the IDLH level by the time the operators were assumed to have donned breathing apparatus (two minutes after the alarm or after odor detection, whichever occurs first.) If the control room was habitable under meteorological conditions occurring not less than 95% of the time for the given compass direction, the given source does not pose a hazard, according to the guidance of RG 1.78. 5.3.3.3.4 Analyses of Transient ChemicalSources Transient chemicals transported by truck, barge or rail in the WSES-3 vicinity were first analyzed in the same manner as the stationary sources. The release was postulated to occur at the point on the road, river channel, or rail line closest to the plant. For those postulated accidents for which the habitability criteria discussed above were not met, a probabilistic safety analysis was performed as follows. The ponion of the given transportation route within a five-mile radius of the control room was divided into a number segments. An accident involving the total loss of lading of a single container was postulated to occur at the center of each segment. The probability that such an accident could cause the concentration in the control room to exceed the IDLH level within two minutes of detection was calculated, using the data on thejoint frequency of occurrence of stability class, wind speed and direction. An overall annual probability of such an event was then calculated, using data on the frequency of shipment of that chemical in the particular transport mode. Over 130 sources, either stationary sources or transient sources treated as stationary, were analyzed using the methods described above. None of the stationary sources were found to pose a hazard under the 95% percentile meteorological conditions. I Page 5-13 i

- ~. i ~, .i i 5.4. OTHERS. l Information on the population centers near Waterford _3 appears in Emergency Plannmg documents in accordance with 10 CFR 50.47 and Part 50 Appendix E. No new large populat:on centers have formed near the Waterford 3 site since initial plant startup. i l Air traffic as represented in the FSAR and the corresponding NRC SER continues to bound the actual air traffic hazard to the plant.5 Waterford 3 knows of no other plant-unique external event that poses any significant threat of severe accident within the context of the screening approach for "High Winds, Floods,' and Others." I I I i i l l 1 i I I 5 W3F1-93-0055 of 14May93 -{ Page 5-14 ~...

6. LICENSEE PARTICIPATION AND INTERNAL REVIEW TEAM GL 88-20, Supplement 4, requested significant participation by utility personnelin the performance of the IPEEE. This would allow the maximum benefit to be realized and facilitate integration of the knowledge gained into operating procedures and training. NRC also recommended that an independent review be conducted to assure the accuracy and validity of the results. This section describes the Waterford 3 IPEEE organization, the extent of utility personnel involvement, and the independent reviews that were condected. i 6.1 IPEEE ORGANIZATION The NRC encouraged utility staff participation in IPEEE preparation. The IPEEE team consisted of Waterford 3 Design Engineering personnel from the Safety and Engineering Analysis (PSA) and Civil Engineering groups. With the exception of the seismic evaluation, all of the IPEEE was performed by Waterford 3 staff. The seismic evaluation utilized the expertise of a recognized seismic consultant (Stevenson & Associates). Utility personnel prepared the seismic safe shutdown equipment list, participated in the walkdowns, and provided detailed review and comment to the seismic consultant. This ensures that knowledge and skills gained during the evaluation would be retained in-house so insights and lessons learned could be incorporated into plant procedures and programs more expeditiously. The personnel involved in the IPEEE were: Maria Rosa Gutierrez Seismic, Condensed SSEL, High Winds, Floods, and Others Richard Finch Others Howard Brodt IPEEE Report Assembly Steven Farkas High Winds, Floods, and Others, Full SSEL Greg Ferguson Seismic Walkdown r John Burke Seismic Walkdown Siddarth Munshi Seismic Walkdown Robert Murillo Licensing George Thomas (S&AI) Seismic Stephen Anagnostis(S&A) Seismic Dr. John Stevenson (S&A) Seismic IStevenson & Associates Page 6-1

i ) 6.2 COMPOSITION OF INDEPENDENT REVIEW TEAM The Waterford 3 IPEEE was submitted to an independent peer review. The seismic peer review + was performed by consultant personnel, contracted by Stevenson & Associates (the seismic consultant), who were not involved in the IPEEE. The evaluation of other events was independently reviewed by Waterford 3 personnel. The members of the peer review team were: Howard Brodt High Winds, Floods, and Others Harry Johnson Seismic Robert Budnitz Seismic f These reviews ensured the accuracy of the IPEEE process and its results. In addition, a second level of review was performed by plant personnel not directly involved with the IPEEE. This consisted ofindividuals from Operations, Engineering, and Licensing management who reviewed the IPEEE submittal. 6.3 AREAS OF REVIEW AND MAJOR COMMENTS The peer review for the seismic events portion of the IPEEE covered all seismic evaluation portions of the project and included a review of the Project Plan of Work, the Safe Shutdown Equipment List (SSEL) development process, all systems aspects of the project, and the draft report, and included a visit to the plant site for a sample walkdown and a review of the documentation. The peer review concluded that there were no significant deficiencies in the seismic IPEEE portion of this report, the SSEL, the walkdown process and its documentation. The peer review of the High Winds, Floods, and Others portion of the IPEEE confirmed that the screening method was properly followed and the conclusions werejustified. The review concluded that the evaluation was appropriately performed. 5.4 RESOLUTION OF COMMENTS All comments were resolved to the satisfaction of the independent reviewers. No comments were made which required significant changes in the analysis. Page 6-2

M r 7. PLANT IMPROVEMENTS AND UNIOUE SAFETY FEATURES One of the purposes of the IPEEE is to identify plant specific severe accident vulnerabilities. The l results of the vulnerability review are described in this section. The criteria used in the Waterford 3 IPEEE to define a vulnerability are a combination of quantitative and qualitative criteria. The quantitative criteria come from the NUMARC Severe Accident Issue Closure Guidelines (see Reference 7-1). The criteria are listed below.

1) A mean core damage frequency greater than or equal to lx10-4 per year for any external event scenario

2) A single failure which has an unusual and significant effect on the core damage frequency

3) A common cause failure of two or more components which has an unusual and significant effect on the core damage frequency

4) A support system failure that causes multiple front line system failures and thereby has an unusual and significant impact on core damage frequency Severe accident vulnerabilities and potential plant improvements are discussed below for each type of external event.

1 7.1 SEISMIC EVENTS No seismic vulnerabilities were identified. The walkdowns resulted in no outliers that are operability issues at the plant. However, there were three unresolved issues at the completion of the walkdowns. These issues are not significant to seismic risk and are being made to conform with standard practice in seismic design. 7.1.2 Resolution of Outlier Concerns CR-94-1019 was issued to document all loose items in the Control Room. This condition report addresses the loose items in the control room and justifies why it is not an operability concern. However, the corrective action will be to remove or restrain the lockers and file cabinets in the control room, remove book shelves in the vicinity of safety-related cabinets, and relocate or restrain other loose items in the vicinity of safety-related cabinets. Waterford 3 will complete a modification package by February 15,1995, for any equipment that will be restrained by means bolting or equivalent. Page 7-1

o L CR-94-1111 was issued to document that the station air pipe which is adjacent to 4KVESWGR3B XPANEL does not meet the clearance requirements stated on drawing B288 Sht.10-2A. A rod hanger which supports the station air pipe is within 1/16 of an inch from the panel. However, the station air pipe will not have any impact on the operability of the relays in the panel. The relays that would be affected are not essential relays. They are_overcurrent induction disk relays and do not contain an instantaneous unit. No corrective action is necessary because there is no adverse impact to the equipment in the panel. The response to CR-94-1111 will be completed by March 30,1995, to fonnally evaluate and document the reasons why the existing clearance is acceptable. 7.1.2 Design Enhancement Opportunities Waterford will revise procedure FP-001-17, Transient Combustibles and Designated Storage Areas, to include guidance for temporary storage of temporary equipment inside the Seismic Category I buildings to prevent hazardous seismic interactions. The guidance will provide assurance that the safety function of components, equipment, and systems will not be affected by temporary storage ofloose items. This is one of the corrective actions for CR-94-1019. This task will be completed by April 1,1995. Until this procedure is approved, the Design Engineering Civil Department will be informed of any new items that will be stored near safety-related equipment so that they will ensure that a seismic interaction concern is not created. 7.2 IIIGH WINDS, FLOODS, AND TRANSPORTATION AND NEARBY FACILITY ACCIDENTS The IPEEE found no high winds, floods, or off-site industrial facility accidents that significantly alter the Waterford 3 estimate of either the core damage frequency, or the distribution of containment release categories. Waterford 3 concludes that the plant is in conformance with the 1975 SRP that pertains to high winds, on-site storage of hazardous materials, and off-site developments. No changes to the plant are required. REFERENCES l 7-1. NUMARC 91-04, " Severe Accident Issue Closure Guidelines", January 1992 Page 7-2

8.

SUMMARY

AND CONCLUSIONS ONCLUDING PROPOSED RESOLUTION OF USIS AND GIS) 8.1 SEISMIC EVENTS Waterford 3 developed and implemented a project to satisfy requirements of the IPEEE seismic evaluation. The project implemented a Generic Letter 88-20, Supplement 4 allowed reduced scope seismic margins analysis (SMA per EPRI 6041). This project consisted of developing a Project Plan and a Walkdown Plan in that it concentrated on potential seismic vulnerabilities for equipment, large tanks, distribution systems, and structures. The implementation was appropriate and cost effective for addressing IPEEE seismic concerns at our " reduced scope" site. The basic requirement for walkdowns is that the equipment, tanks, distribution systems, and structures can all withstand the design basis SSE at the plant and still provide their safe shutdown functions. The SMA uses primarily EPRI report NP-6041-SL as guidance which is not overly prescriptive but relies on thejudgment of an experienced team to meet the basic requirement. I A Safe Shutdown Equipment List (SSEL), using safety and non-safety-related components, was selected for achieving and maintaining plant shutdown in accordance with plant operating procedures. The SSEL also included items that are potential seismic-induced fire and seismic-induced flood sources within the plant. There were three walkdowns performed; the Train "B" on-line walkdown during November of 1993, the Train "A" on-line walkdown during 1994 and the outage walkdown during March 1994. Documentation for the walkdowns was gathered and was available during the walkdowns. Equipment specific documentation was placed in individual file folders for a sample of equipment on the SSEL. Generic documentation (floor response spectra, etc.) was available for review during the walkdowns. Seismic Verification Data Sheets that included each equipment item of the equipment list were developed. These sheets contain walkdown observations as well as screening results. The walkdowns resulted in no outliers that are operability issues at the plant. However, there were some unresolved issues at the completion of the walkdowns, which will be resolved per discussion in Section 7.1. Page 8-1

8.2 HIGH WINDS, FLOODS, AND TRANSPORTATION AND NEARBY FACILITY ACCIDENTS The IPEEE found no high winds, floods, or off-site industrial facility accidents that significantly alters the Waterford 3 estimate of either the core damage frequency, er the distribution of containment release categories. Waterford 3 concludes that the plant is in conformance with the 1975 SRP that pertains to high winds, on-site storage of hazardous materials, and off-site developments. 8.3 PROPOSED RESOLUTION OF USIS AND GIS 8.3.1 USI A-45 (Decay IIcat Removal) The stated purpose of Unresolved Safety Issue (USI) A-45 (NUR EG-1289, see Reference 8-1)is to " evaluate the adequacy ofcurrent designs to ensure that LWRs do notpose unacceptable risk as a restdt ofDHR [ decay heat removal] systemfailures. The primary objectives qf the USI A-45 program are to evaluate the safety adequacy ofDHR systems in existing LWRpowerplants and to assess the value and impact (or benefit-cost) ofalternative measuresfor improving the overall reliability of the DHRfunction." t Since the seismic events evaluation showed that there are no seismic vulnerabilities, we conclude that there are no significant or unique seismic vulnerabilities in the decay heat removal function. The high winds, floods, and nearby facility accidents evaluation determined that these external events pose no significant risk of core damage (see Section 8.2). Therefore, USl A-45 should be considered resolved for Waterford-3 with respect to seismic, high wind, flood, and nearby facility accident events. 8.3.2 GI-131 (Potential Seismic Interaction Involving the Movable In-Core Flux Mapping System Used in Westinghouse Plants) This issue is not applicable, since Waterford-3 is a Combustion Engineering plant. 8.3.3 USI A-46 (Verification of Seismic Adequacy of Equipment in Operating Plants) 7 Waterford-3 is not a USI A-46 plant. The issue of spatial interaction, however, has been addressed as part of the reduced scope seismic margins method. 1 i i Page 8-2

- - ~. - - - - i-I I I i i - The Waterford 3 IPEEE has not been used to evaluate any other USIs or GIs. I - REFERENCES 8-1. NUREG-1289,

• Regulatory and Backfit Analysis: Unresolved Safety Issue A-45, Shutdown Decay Heat Removal Requirements."

? I 4 e -5 i k i i i l F Page 8-3 4 +}}