RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 1, Introduction and General Description of Plant
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1.1-1 Reformatted Per Amendment 99-01 FSAR NOTE With the exception of the description of 10CFR50.2 design basis parameters, numerical values throughout this FSAR should be considered nominal in nature. They are intended to provide a sense for the value of the parameter and should NOT be viewed
as an actual value observable in the plant.
Plant operation at va lues other than those presented herein is acceptable. The actual values are documented within established Technical Specification, Desi gn Basis Documentation, or other controlled documents.
1.0 INTRODUCTION
AND GENERAL DESCRIPTION OF PLANT NOTE 1.1 Section 1.1 is being retained for historical purposes only.
1.1 INTRODUCTION
This Final Safety Analysis Report is submitted in support of an application by South Carolina Electric and Gas Company (SCE&G) for a class 103 license to operate a
nuclear power station designated as the Virg il C. Summer Nuclear Station. This report presents descriptions and analyses of the final station design.
1.1.1 STATION LOCATION The plant site is located north (2.5 miles) of Parr, South Carolina.
Parr is the site of existing fossil and hydro power stations operated by SCE&G and the decommissioned, experimental Carolinas Virginia Tube Reac tor (CVTR). The nucl ear plant site is adjacent to a manmade reserv oir created by placing a se ries of dams across Frees Creek, a tributary of the Broad River in western Fairfield C ounty, South Carolina. The resulting Monticello Reservoir provides wate r requirements for the nuclear station and a pumped storage facility. The reservoir is loca ted east of the Broad River and west of South Carolina State Highway 215, about 26 miles north of Columbia, South Carolina. The pumped storage facility raises and lowers the reservoir level approximately 4.5 feet when in operation.
1.1.2 NUCLEAR STEAM SUPPLY SYSTEM The station includes a pressurized water reactor nuclear steam supply system (NSSS), designed and furnished by Westinghouse Elec tric Corporation and a turbine generator, designed and furnished by General Electric Company. This equipment is similar in design to several projects licensed or current ly under review by the Nuclear Regulatory Commission (see Section 1.3). The balance of the station is des igned and constructed by SCE&G, with the assistance of it s agent, Gilbert Associates, Inc.
99-01 02-01 1.1-2 Reformatted Per Amendment 99-01 1.1.3 CONTAINMENT Containment is provided by the Reactor Building, a reinforced concrete structure designed by Gilbert Associates, Inc. The Reactor Building is comprised of a flat foundation mat, cylindrical wa ll and shallow dome roof. The foundation mat and cylindrical wall are reinforced with conventional mild steel reinforcing. The cylindrical wall is prestressed in the vertical and horizon tal directions by a post-tensioning system.
The shallow dome roof is prestressed by a three-way post-tensioning system. The inside surface of the Reactor Building is li ned with a carbon steel liner to ensure a high degree of leak tightness under operating and accident conditions.
1.1.4 CORE THERMAL POWER LEVELS The NSSS is designed for a rated power output of 2912 MWt, which is the license application rating, with an equivale nt station net electrical output of approximately 950 MWe. Containment and engineered safety features are designed and evaluated for operation based upon an Engineered Safety Design Rating (ESDR) and Licensed Power Level of 2900 MWt in the core.
Postulated accidents having offsite dose consequences are evaluated at 2900 MWt. A nalyses of all accident consequences, to show conformance to 10CFR50.46, have been performed at 2900 MWt. Conformance to 10CFR50.46 for Large Break LOCA is evaluated at 2900 MWt, the rated core power level following Refueling 9 in the Spring of 1996.
1.1.5 SCHEDULE The project schedule is based upon plant completi on (to fuel loading) in February, 1982 and commencement of commercial operation appr oximately six months after issuance of the operating license.
1.3-1Reformatted PerAmendment 99-01NOTE 1.3Section 1.3 is being retained for historical purposes only.1.3COMPARISON TABLES1.3.1COMPARISON WITH SIMILAR FACILITY DESIGNSTable 1.3-1 provides significant similarities and differences between the Virgil C.Summer Nuclear Station and other designs.1.3.2COMPARISON OF FINAL AND PRELIMINARY INFORMATIONTable 1.3-2 provides significant design changes since submittal of the Preliminary Safety Analysis Report.
99-01 1.3-2Reformatted PerAmendment 99-01TABLE 1.3-1DESIGN COMPARISON WITH SIMILAR FACILITIES (1)ChapterNumberChapter TitleSystem/ComponentVirgil C. Summer NuclearStation FSAR ReferencesSignificant SimilaritiesSignificant Differences 4.0ReactorSection 4.0Similar to North Anna Units 1 and 2 total 1 NSSS heat output of 2785 MWt.Beaver Valley Unit 1 has a total NSSS heat output of 2660 MWt.FuelSection 4.2.1Similar to Beaver Valley Unit 1 and North Anna Units 1 and 2.Small variations in design parameters based on nuclear and thermal-hydraulicdesign.Reactor Vessel InternalsSection 4.2.2Similar to Beaver Valley Unit 1 and North Anna Units 1 and 2.Beaver Valley Unit 1 and North Anna Units 1 and 2 have thermal shields; Virgil C.Summer Nuclear Station has neutron pads.Reactivity Control SystemsSection 4.2.3Similar to Beaver Valley Unit 1 and North Anna Units 1 and 2.
None.Nuclear DesignSection 4.3Similar to Beaver Valley Unit 1 and North Anna Units 1 and 2Small variations in nuclear parameters. Thermal-Hydraulic DesignSection 4.4Similar to Beaver Valley Unit 1 and North Anna Unit 1 and 2Small variations in thermal-hydraulic and heat transfer parameters.(1) Comparisons identified relative to thermal power are compared to the pre-uprate power level of 2775 MWt RTP.
02-01 02-01 1.3-3Reformatted PerAmendment 99-01TABLE 1.3-1 (Continued)DESIGN COMPARISON WITH SIMILAR FACILITIESChapterNumberChapter TitleSystem/ComponentVirgil C. Summer NuclearStation FSAR ReferencesSignificant SimilaritiesSignificant Differences 5.0Reactor Coolant SystemSections 5.1, 5.2Similar to Beaver Valley Unit 1 and North Anna Units 1 and 2.Virgil C. Summer Nuclear Station does not have loop stop valves.Reactor VesselSection 5.4Similar to Beaver Valley Unit 1 and North Anna Units 1 and 2.No significant differences.Reactor Coolant PumpsSection 5.5.1The hydraulics are similar to Beaver Valley Unit 1 and North Anna Units 1 and 2Virgil C. Summer Nuclear Station motors
ar e of modular construction.Steam GeneratorsSection 5.5.2Similar to Wolfe Creek & Millstone Unit 3.Alloy 690 tube material, triangular tubepitch, larger heat transfer area.PipingSection 5.5.3Similar to Surry Units 1 and 2Virgil C. Summer Nuclear Station has no electroslag welding.Residual HeatRemoval SystemSection 5.5.7The piping and fittings are similar to Joseph M. Farley Units 1 and 2No significant differences.PressurizerSection 5.5.10Similar to North Anna Units 1 and 2No significant differences.
6.0Engineered Safety Featur esElectric Hydrogen RecombinerSection 6.2.5Similar to Surry Units 1 and 2No significant differences.Emergency Core Cooling SystemSection 6.3Similar to Joseph M. Farley Units 1 and 2Automatic sump valve opening logic is different. 02-01 1.3-4Reformatted PerAmendment 99-01TABLE 1.3-1 (Continued)DESIGN COMPARISON WITH SIMILAR FACILITIESChapterNumberChapter TitleSystem/ComponentVirgil C. Summer NuclearStation FSAR ReferencesSignificant SimilaritiesSignificant Differences 7.0Instrumentation and Contro lsReactor Trip SystemSection 7.2Similar to Joseph M. Farley, Unit 1No significant differences.Engineered Safety Features ActuationSystemSection 7.3Similar to Joseph M. Farley, Unit 1No significant differences.System Required for Safe ShutdownSection 7.4Similar to Joseph M. Farley, Unit 1No significant differences.Safety-Related DisplayInstrumentationSection 7.5Similar to Joseph M. Farley, Unit 1No significant differences but physical display may differ.Other Safety SystemSection 7.6Similar to Joseph M. Farley, Unit 1No significant differences.Control SystemsSection 7.7Similar to Joseph M. Farley, Unit 1Joseph H. Farley, Unit 1 has a 50% load rejection capability. Virgil C. SummerNuclear Station has an 85% load rejection capability.
02-01 02-01 1.3-5Reformatted PerAmendment 99-01TABLE 1.3-1 (Continued)DESIGN COMPARISON WITH SIMILAR FACILITIESChapterNumberChapter TitleSystem/ComponentVirgil C. Summer NuclearStation FSAR ReferencesSignificant SimilaritiesSignificant Differences 8.0Electric PowerOffsite PowerSection 8.2Three Mile Island Nuclear Station Unit 1 has two offsite sources, 230 kV/6.9 kV.Virgil C. Summer Nuclear Station has one 115 kV/7.2 kV and one 230 kV/7.2 kVoffsite source.Onsite PowerSection 8.3Joseph M. Farley Units 1 and 2 have Colt Industries 12PC2V diesel enginesThree Mile Island Nuclear Station Unit 1 has four 125 volt batteries for supply ofvital d-c power. Virgil C. Summer has two125 volt batteries for supply of vital d-c power.120 Volt Vital a-c Bus SystemThree Mile Island Nuclear Station Unit 1 normal source of power is 480 volt a-cESF bus; alternate source is 125 volt d-csystem.Three Mile Island Nuclear Station Unit 1 has four 15kVA single phase inverters.Virgil C. Summer Nuclear Station has six7.5kVA single phase inverters, arranged infour independent channels.
9.0Auxiliary System Fuel Handling SystemSection 9.1.4Similar to Beaver Valley Unit 1 and North
Ann a Units 1 and 2Each project has difference capacities for spent fuel storage.Chemical and Volume Control SystemSection 9.3.4Similar to Joseph M. Farley Units 1 and 2No significant differences.
02-01 02-01 1.3-6Reformatted PerAmendment 99-01TABLE 1.3-1 (Continued)DESIGN COMPARISON WITH SIMILAR FACILITIESChapterNumberChapter TitleSystem/ComponentVirgil C. Summer NuclearStation FSAR ReferencesSignificant SimilaritiesSignificant Differences 10.0Radioactive Waste Managem entSource TermsSection 11.1Similar to Joseph M. Farley Units 1 and 2Differences in source term analysis.Liquid Waste ProcessingSection 11.2Performance characteristics similar to Joseph M. Farley Units 1 and 2Differences in source term analysis.Gaseous Waste ProcessingSection 11.3Similar to Joseph M. Farley Units 1 and 2Differences in source term analysis.
14.0Initial Tests andInspectionsChapter 14Similar to Beaver Valley Unit 1 and North Anna Units 1 and 2.No significant differences.
15.0Accident AnalysisChapter 15Similar to Beaver Valley Unit 1 and North Anna Units 1 and 2The accident analysis sections have been updated. New sections have been added,e.g., single RCCA withdrawal, accidentaldepressurization of the RCS, code descriptions, etc.
02-01 1.3-7Reformatted PerAmendment 99-01TABLE 1.3-2COMPARISON OF FINAL AND PRELIMINARY DESIGNS ItemFSAR ReferenceChangeService Water Pond Bottom and North Dam Foundation 2.5.6The following changes were made to the servicewater pond as conservative measures to reduce thepotential for seepage losses from the pond during a postulated loss of the Monticello Reservoir:a.Installation of a grout curtain along the c enterline of the North Dam foundation.b.Placement of a clay blanket on the portion of the p ond bottom adjacent to the South Dam.Auxiliary Building Charcoal Exhaust SystemTable 3.2-1 and 9.4.2.2.3The auxiliary building charcoal exhaust system was d eclassified from SC-3 to non-nuclear safety class.Release calculations indicated that this system wasnot necessary to maintain offsite dose rates withinacceptable limits. However, areas exhausted were increased and HEPA filters were added. Change in Seismic Instrumentation 3.7.4Seismic instrumentation program was changed to
m eet the required sensitivity of instrumentation tomeasure the seismic response of nuclear powerplant features important to safety. Increase of Reactor Building Internal Design Pressure3.8.1.3.1.2 and 6.2.1The reactor building internal design pressure was inc reased from 55 psig to 57 psig to maintain therequired margin above the calculated pressure.
99-01 1.3-8Reformatted PerAmendment 99-01TABLE 1.3-2 (Continued)COMPARISON OF FINAL AND PRELIMINARY DESIGNS ItemFSAR ReferenceChangeCompressive Strength of 5000 psi Concrete wasChanged from 28 day Test to 90 day Test3.8.1.6.1.2.1The heat generated during the curing period of mass concrete calculated from results obtained fromactual field tests, utilizing the actual dimensions of the reactor building mat, was higher thanacceptable. The reduction in concrete mix cementcontent and consequent lower temperature in the concrete was acceptable by using the 90 day strength as a design requirement. Control Building Foundation3.8.5.1.4The control building foundation was changed from
c aissons to mat on fill concrete. Fuel 4.2.1The reactor will be fueled with 17 x 17 fuel assemblies in lieu of 15 x 15 fuel assemblies. Reactor Internals 4.2.2The reactor internals have been modified to accept
17 x 17 fuel assemblies. The thermal shield has been replaced by neutron
pads. This change simplifies core support designand reduces flow pressure drop and velocity. Fuel Pellet Enrichments and Density 4.3The fuel pellet enrichments as well as other core
par ameters change between the PSAR and FSAR toreflect the evolution of the design as core performance and safety requirements are met. Fueldensity has been increased from 94, 92, and 91percent of theoretical to 95 percent of theoretical.
1.3-9Reformatted PerAmendment 99-01TABLE 1.3-2 (Continued)COMPARISON OF FINAL AND PRELIMINARY DESIGNS ItemFSAR ReferenceChangeBurnable Poison Loading Pattern 4.3The burnable poison pattern shown in the FSARreflects more detailed calculations than that of thePSAR. Reactor Top Head Penetrations 5.4Reactor vessel top head penetrations and CRDM wer e redesigned to meet requirements for inserviceinspection.Reactor Vessel Nozzle Insulation 5.4Non crushable insulation is used.Reactor Building Ventilation System6.2.2 and 9.4.8HEPA filters and bypass dampers were added to the r eactor building ventilation system to reduce offsitereleases following a LOCA.Post Accident Hydrogen Recombiner 6.2.5Change of post accident hydrogen recombiners from
ex ternally mounted to internally mounted units.Semi-Automatic Emergency Core Cooling System (E CCS) Switchover 6.3System was revised to automatically change residual heat removal (RHR) pump suction from therefueling water storage tank (RWST) to the recirculation sump of the reactor building duringchangeover from the injection phase to therecirculation phase during post accident core cooling. This revision ensures continual water supply to the suction of the RHR pump.Automatic Reactor Building Spray Switchover 6.3.2System was revised to automatically change spray pum p suction from the RWST to the recirculationsump of the Reactor Building on lo-lo level in theRWST. This revision ensures continuous water supply to the suction of the pumps.
99-01 1.3-10Reformatted PerAmendment 99-01TABLE 1.3-2 (Continued)COMPARISON OF FINAL AND PRELIMINARY DESIGNS ItemFSAR ReferenceChangeElimination of Redundancy of Instrument Air Supply forVentilation Control in the Control Room6.4 and 9.4.1Since the control air system is not nuclear safety-related and all dampers/valves fail to the postaccident mode, redundancy is not required.Steam Dump 7.2Power operated steam relief valves were automated to i ncrease dump capacity.Undervoltage Reactor Trip 7.2The undervoltage sensors and underfrequencysensors were moved to the motor side of the reactorcoolant pump breaker. Trip on pump breaker open was removed.Safety Injection 7.2Injection signal on pressurizer low pressure in lieu of c oincident low level and low pressure.Feedwater Control 7.2The programmed setpoint for the feedwater controlvalve uses steam flow/feed flow mismatch and S/Glevel. The bypass valves use S/G level and neutron flux to control valve position. Also, the feedwater pump speed is programmed using line pressuresand steam flow.Main Steam Isolation Signal 7.2Changed the main steam isolation from activating on hig h steam line flow coincident with low steam linepressure or lo-lo T avg to low steam line pressure orhigh steam line flow coincident with lo-lo T avg.Blowdown and Sample Line Isolation 7.2Automatic initiation of emergency feedwater causes
i solation. Manual isolation was removed.
02-01 1.3-11Reformatted PerAmendment 99-01TABLE 1.3-2 (Continued)COMPARISON OF FINAL AND PRELIMINARY DESIGNS ItemFSAR ReferenceChangeChanges to Manual Actuation of the following ThreeSystems: Main Steam Isolation, Containment Isolation Reactor Building Spray and Automatic Initiation of Reactor Building Spray7.2, Figure 7.2-1 and Table 7.3-2 Changes have been made to these systems toprovide the required redundancy for systeminitiation.Post Accident Monitoring 7.5A post accident monitoring system has been added.Analog to Digital 7.7Analog rod position indication has been replaced by
dig ital rod position indication to improve reactorprotection instrumentation.Incorporation of a Circuit Breaker between the Generator
T erminals and the Unit Auxiliary Transformer8.2.1.1The addition of this circuit breaker improves operational flexibility by permitting the use of the unitauxiliary transformer during startup and eliminating the need of station service transfer during turbine
trip.Physical Arrangement and Separation Criteria
Requ irements of Installed Electrical Systems andEquipment8.3.1.1 and 8.3.2.1Electrical equipment such as motor control centers, batter ies, battery chargers, distribution panels,diesel generators, etc., have been located to provideprotection and/or separation from postulatedaccidents (i.e. missiles, seismic event, fire hazards, etc.).Removal of Bus Tie Breaker between Class 1E d-c
B uses 1A and 1B8.3.2.1.2 and 8.3.2.2.1The manually operated bus tie breaker between Class 1E d-c buses 1A and 1B was removed toimprove system reliability and safety.
1.3-12Reformatted PerAmendment 99-01TABLE 1.3-2 (Continued)COMPARISON OF FINAL AND PRELIMINARY DESIGNS ItemFSAR ReferenceChangeDiesel Generator Fuel Oil System8.3.1.1.2.1 and 9.5.4The following changes have been made to the dieselgenerator fuel oil system:a.Incorporation of two independent fuel oil storage s ystems, including independent buried storagetanks, in lieu of a common storage system.b.Incorporation of two a-c motor driven transfer pum ps per system in lieu of one a-c and one d-cmotor driven pump.c.Reduction in operating time for fuel available in da y tanks to a nominal 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to beconsistent with the maximum size of day tank permitted by National Fire Protection Association Standard (NFPA) 37.Spent Fuel Racks 9.1The storage lattice spacing has been reduced to 14
inc hes from 21 inches on center to increase storagefrom approximately 1-1/3 core to 4-1/3 core.Reactor Building Instrument Air System 9.3.1The reactor building instrument air system was
re vised to:a.Delete one air compressor.b.Relocate remaining two compressors outside the r eactor building.
1.3-13Reformatted PerAmendment 99-01TABLE 1.3-2 (Continued)COMPARISON OF FINAL AND PRELIMINARY DESIGNS ItemFSAR ReferenceChangeNuclear Sampling System 9.3.2Recent regulatory guidelines permit declassificationof nuclear sampling system piping outside reactorbuilding without reduction in system reliability or performance.Fuel Handling Building Charcoal Exhaust System9.4.3.2.1Additional areas were added to the exhaust system
to r educe normal plant releases. Redundantplenums and additional HEPA filters were added.Steam Generator Blowdown System10.4.8The steam generator blowdown system was
r edesigned to accommodate the changes in thesecondary cycle chemistry control from phosphate toall volatile treatment (AVT).Redesign of Radiation Monitoring11.4, 12.1.4 and 12.2.4Various changes to the radiation monitoring system to i ncrease surveillance and detection capabilities ofplant effluents.Automatic Emergency Feedwater Switchover from
Cond ensate Tank to Service Water10.4.9System was revised to automatically change emergency feedwater pump suction from thecondensate storage tank to the service water system on lo suction pressure to the EFW pumps. This revision ensures continuous water supply to the suction of the pumps.ATWS Mitigation System Actuation Circuitry (AMSAC) 7.8Added instrumentation and controls necessary to r espond to 10CFR50.62 for mitigation of ATWSevents. 99-01 1.4-1Reformatted PerAmendment 99-01NOTE 1.4Section 1.4 is being retained for historical purposes only.1.4IDENTIFICATION OF AGENTS AND CONTRACTORSSouth Carolina Electric and Gas Company (SCE&G) is the sole applicant for theconstruction permit and operating license for the Virgil C. Summer Nuclear Station.
SCE&G is two-thirds owner of the Virgil C. Summer Nuclear Station. The remaining one-third is owned by the South Carolina Public Service Authority (SCPSA) which together with SCE&G owns the Virgil C. Summer Nuclear Station as tenants in common. It is specifically noted that SCE&G is the agent for the SCPSA with regard to the Virgil C. Summer Nuclear Station, and that SCE&G retains sole responsibility for theoverall technical direction in the licensing, design, construction, operation, management, maintenance and decommissioning. Likewise, it is specifically noted that the ownership change reflected in the amended application does not effect a transfer of control from SCE&G to the SCPSA, nor does the ownership change involve a physical or design change in the facility or site.Gilbert Associates, Inc. (Gilbert), has been retained by SCE&G as architect engineer forthe entire project, including plant layouts and system arrangements, and design ofbalance of plant equipment. A brief summary of Gilbert technical qualifications is included in Section 1.4.2.SCE&G has contracted with the Westinghouse Electric Corporation (Westinghouse) forthe design and manufacture of the complete Nuclear Steam Supply System. Inaddition, Westinghouse supplies competent technical consultation in areas such as initial fuel loading, testing, and initial startup. Westinghouse is involved in the training of SCE&G personnel. A brief summary of Westinghouse technical qualifications is included in Section 1.4.3.Dames and Moore, Inc. (D&M), is retained as environmental consultant and hasassisted SCE&G in the preparation of the environmental portions of this report. Inaddition, D&M has assisted SCE&G in the preparation of the required Environmental Report-Operating License Stage. A brief summary of D&M technical qualifications is included in Section 1.4.4.
99-01 1.4-2Reformatted PerAmendment 99-011.4.1SOUTH CAROLINA ELECTRIC AND GAS COMPANY QUALIFICATIONSAND EXPERIENCESCE&G owns and operates an integrated electric generation, transmission, anddistribution system which serves approximately 484,000 customers in a 15,000 squaremile service area. This area, stretching from the central region to the coastal plains, includes Columbia, the state capital, and Charleston, South Carolina's principal seaport.
SCE&G's transmission system is part of the interconnected grid extending over a large part of the central and eastern portion of the nation.SCE&G, acting as its own general contractor, has constructed generating facilitieswhich provide most of SCE&G's present generating capacity. For the construction ofthe Virgil C. Summer Nuclear Station, SCE&G has engaged Daniel Construction Company (Daniel) of Greenville, S.C. to act as the general contractor. A summary of the qualifications of Daniel is given in Section 1.4.5.SCE&G is a partner with Carolina Power and Light Company, Duke Power Companyand Virginia Electric and Power Company in Carolinas Virginia Nuclear PowerAssociates, Inc. (CVNPA). CVNPA was formed in 1956 to build and operate a 17,000 kWe nuclear steam generating plant at Parr, S.C., for research, operating and engineering experience. This plant, the Carolinas Virginia Tube Reactor (CVTR), was constructed and operated under license granted by the Atomic Energy Commission.
The CVTR was decommissioned in 1967 after completion of a successful operating and research program. SCE&G actively participated in the CVTR project in the planning, management, training, research, technical and operational levels.SCE&G has maintained active participation in the fast breeder programs of bothWestinghouse and Atomics International as well as participating in the Savannah RiverNuclear Study Group which examined the feasibility of developing power from aproduction reactor.Organization charts for SCE&G relevant to this project, are Figures 13.1-1 through13.1-5. The engineering and construction of this project is the responsibility of theVice-President, Group Executive, Engineering and Construction.1.4.2GILBERT ASSOCIATES, INC. QUALIFICATIONS AND EXPERIENCEGilbert Associates, Inc. (Gilbert) engineers and consultants, is the architect engineer forthe Virgil C. Summer Nuclear Station. The company, with its main office in Reading,Pennsylvania, was originally known as W. S. Barstow and Company which was organized in 1906. The corporate name was changed to E. M. Gilbert Engineering Corporation in 1933. In 1942, the corporate structure was revised to provide for complete employee ownership and the name became Gilbert Associates, Inc. In 1973, Gilbert Associates acquired Commonwealth Associates, Inc., a large engineering and consulting firm of similar capabilities.
1.4-3Reformatted PerAmendment 99-01Throughout the past 70 years, Gilbert has progressively grown in size and in scope ofactivity. The collective experience and capabilities of the firm offer complete consulting and engineering services to both investor-owner utilities and general industry in suchdiverse fields as: nuclear and conventional power generation; transmission, substation, and distribution systems; economic engineering and management consulting service; steel making and processing; cement and minerals processing; chemical and general industrial facilities; water desalination plants; institutional and commercial installations;environmental and solid waste treatment; and water production projects. Projects undertaken have ranged from large electric power generating plants and production facilities to small industrial boiler plants and allied service facilities.Since 1942, Gilbert has been responsible for the design of well over 100 thermal generating units, both fossil and nuclear power, representing approximately 30,000,000kW of new generating capacity. The company's experience includes one of the first reheat units, one of the first once through boiler units and one of the first supercritical steam pressure units. Individual unit designs have ranged in ratings up to 1,200,000 kW and stations have varied complexity - nuclear, mine-mouth, closed cycle cooling tower, base-load, peaking and others. At present, Gilbert has over 17,000,000 kW of generation under design, of which 10,600,000 kW is nuclear.Since 1950, Gilbert has played an active and important role in the development ofnuclear energy for private utilities, industry and governmental agencies. Gilbert projectsinclude complete programs of nuclear power development involving analysis of sites, complete evaluations of proposals, contract and fuel program assistance, preparation of license applications, containment vessel design concepts, complete plant design and procurement. More than a score of studies, cost estimates, evaluations, concept developments and preliminary plant designs have been prepared since 1953 for various utility customers and other clients such as the Nuclear Regulatory Commission (formerly the Atomic Energy Commission), Westinghouse Electric Corporation, Atomics International and the Air Force. Development of reactor containment design concepts for plants near large population centers was accomplished in 1963. The first complete station design was the Saxton Experimental Power Reactor (20 MWe) in 1959. Other completed designs include the 490 MWe Robert Emmett Ginna Station for the Rochester Gas and Electric Corporation; the 320 MWe Mihama plant for the Kansai Electric Power Company, Inc.; the 781 MWe Takahama Station for the Kansai Electric Power Company, Inc. as subcontractor to Westinghouse Electric International Company; and the 870 MWe Three Mile Island Nuclear Station, Unit 1 for the Metropolitan Edison Company. Stations for which Gilbert now has overall plant design and engineering responsibility are the 855 MWe Crystal River Plant, Unit 3, for FloridaPower Corporation; the 2 unit, 1200 MWe each, Perry Nuclear Power Plant for the Cleveland Electric Illuminating Company; the 2 unit, 1122 MWe each, Ohi Station (ice condenser), as subcontractor to Westinghouse for the Kansai Electric Power Company, Inc.; the 564 MWe Ko-Ri Station, Unit 1, as subcontractor to Westinghouse for the Korea Electric Power Company; and the 615 MWe Nuclear Power Plant Krsko for Savske Elektrarne, Ljubljana, Slovenia; Elektroprivreda, Zagreb, Croatia, as subcontractor to Westinghouse.
1.4-4Reformatted PerAmendment 99-01The Gilbert organization includes nearly 3700 employees with a complete staff ofengineers, draftsmen and many technical specialists. Included in the total staff are over 2200 engineers, technical specialists and draftsmen. This includes members of management, professional personnel and individuals in other specialized fields.Responsibility for engineering and design of nuclear power plants is centered in theUtilities Division of the Company. Every nuclear project is assigned to a ProjectManager, selected from a staff of engineers having an average of about 15 years withGilbert. Through this divisional control, the production function of the project is carried through to completion.The engineering disciplines, such as nuclear, mechanical, electrical, civil, structural,architectural, environmental engineering, etc., are grouped into departments andprovide technical resources to the project. Each department is managed by a chief engineer who assumes technical and administrative responsibility for the personnelassigned to the various projects. Figures 1.4-1 and 1.4-2 depict the Gilbert organization.Additional support necessary to the project is provided from other service departments,including drafting, estimating, specifications, legal, accounting, purchasing, expediting, etc.1.4.3WESTINGHOUSE ELECTRIC CORPORATION QUALIFICATIONS ANDEXPERIENCEWestinghouse Electric Corporation's (Westinghouse) experience in nuclear plants forthe electric utility industry is demonstrated by the Pressurized Water Reactor (PWR)plants that Westinghouse has designed, developed, and manufactured. Table 1.4-1 lists all Westinghouse PWR plants to date, including those plants currently under construction or on order.Westinghouse has long held a position of leadership in the electrical manufacturingindustry. Traditionally this leadership has been based on technological development ofboth standard and new products, reliability and product quality. Nowhere is this traditional leadership displayed more vividly than in nuclear power. Through early participation in basic research and basic engineering development, Westinghouse has established a broad technological foundation in nuclear power application. This has been followed by a continuing program of sound technological development which enables Westinghouse to offer to the electric utility industry a reliable and safe source ofpower.The experience of Westinghouse in nuclear activity is evident in numerous nuclearpower projects completed, soon to go into operation, or under construction. Thefollowing paragraphs describe Westinghouse designed PWR plants which are presently in operation.
1.4-5Reformatted PerAmendment 99-011.4.3.1Plants in OperationWestinghouse PWR plants in operation are as follows:1.ShippingportShippingport was the world's first large central station nuclear power plant. Thereactor plant was designed by the Bettis Atomic Power Laboratory, which isoperated by Westinghouse under a Nuclear Regulatory Commission (NRC) contract. Shippingport's PWR has produced power for Duquesne Light Company since December 1957.2.Yankee-RoweSingled out by the NRC as a "Nuclear Success Story," Yankee-Rowe went online in November 1960. Owned and operated by the Yankee Atomic Electric Company,Yankee-Rowe has progressed from an initial rating of 120 Mwe to its present 175 MWe rating. Westinghouse supplied the NSSS and the turbine generator.3.Trino Vercellese (Enrico Fermi)The Trino Vercellese nuclear plant was one of the first Westinghouse designed plants to incorporate chemical shim control of reactivity. Chemical shim has sincebecome a standard feature of Westinghouse PWR control. Trino Vercellese achieved initial criticality in June 1964 and began power operation in October 1964. The plant is rated at 261 MWe.4.Chooz (Ardennes)The Chooz plant is unique in that the Westinghouse PWR and its auxiliaries arehoused in man-made caverns. Ardennes, a joint Franco-Belgian undertaking, owned and operated by the Societe d'Energie Nucleaire Franco-Belge des Ardennes (SENA), is located in France near the French-Belgian border. Chooz achieved initial criticality in October 1966 and began power operation in 1967.5.San Onofre No. 1San Onofre No. 1 employs the Westinghouse developed rod cluster control concept which has since become a standard feature on the Westinghouse PWR.Owned by the Southern California Edison Company and the San Diego Gas and Electric Company, the 430 MWe plant is located near San Clemente, California.
Westinghouse supplied the NSSS and the turbine generator. Initial criticality was achieved in June 1967, and power operation began in January 1968.
1.4-6Reformatted PerAmendment 99-016.Haddam Neck (Connecticut Yankee)Owned and operated by the Connecticut Yankee Atomic Power Company, thisplant went critical in July 1967 and attained full power operation in December1967. Like San Onofre No. 1, the plant employs rod cluster control in conjunction with chemical shim control. Westinghouse supplied the NSSS and the turbine generator. The plant has been uprated to 575 MWe.7.Jose Cabrera-ZoritaThe Jose Cabrera-Zorita station is located near Zorita, Spain. The 153 MWe plant employs rod cluster control, chemical shim control and a Zircaloy clad core.Construction began in mid-1965, and power operation began in 1968. Jose Cabrera-Zorita is owned and operated by the Union Electrica, S.A., a Spanish utility.8.Beznau No. 1 and No. 2Beznau No. 1, Switzerland's first commercial nuclear power plant, achieved initial criticality in June 1969 and supplied power to the system in July 1969. The 350MWe plant was designed and constructed by the Westinghouse-Brown Boveri Consortium for the owner/operator utility, Nordostschweizerische Kraftwerke A. G.
The plant started producing power less than four years after award of the plant contract. Beznau No. 2 achieved criticality in October 1971 and began commercial operation in early 1972.9.Robert Emmett GinnaThe Robert Emmett Ginna Plant, owned and operated by the Rochester Gas and Electric Corporation, is located in New York on the south shore of Lake Ontario.Westinghouse supplied the 490 MWe plant on a turnkey basis. Construction began in April 1966 with initial criticality being achieved in November 1969 (just 42 months after start of construction). Power was supplied to the system in December 1969.10.Mihama No. 1 and Takahama No. 1These plants are owned by the Kansai Electric Power Company, Inc. Mihama No.
1 is a two-loop, 320-MWe unit and marks the beginning of a line of WestinghousePWR's supplying the generation needs of the Far East. Westinghouse International Company was the prime contractor for the Mihama project, supplying the NSSS engineering, nuclear fuel, and some major system components.
Mihama No. 1 required only 44 months from the start of site construction to first power production in August 1970. Takahama No. 1 is a three-loop, 781 MWe unit.
Initial criticality was achieved in March 1974.
02-01 1.4-7Reformatted PerAmendment 99-0111.H. B. Robinson No. 2This plant is a three-loop, 700 MWe unit which was built on a turnkey basis for theCarolina Power and Light Company. The plant is located at a site near Hartsville,South Carolina on a man-made cooling lake. The construction permit was granted in April 1967. The plant achieved criticality in September 1970 and first power to the system in October 1970.12.Point Beach No. 1 and No. 2The Point Beach Project consists of two 497 MWe units, which were built on a turnkey basis for the Wisconsin Michigan Power Company and the WisconsinElectric Power Company. The plants are located near Two Creeks, Wisconsin, 90 miles north of Milwaukee on Lake Michigan. This was the first two-unit station to utilize many common facilities and shared auxiliary systems. The construction permit for Point Beach No. 1 was granted in July 1967 with initial criticality and first power to the system in November 1970. Point Beach No. 2 went critical in May 1972 and was available for commercial operation in October 1972.13.Surry No. 1 and No. 2The Surry Power Station, two three-loop 788 MWe units, is owned by the Virginia Electric and Power Company. (The James River Station is about 30 miles fromNorfolk, Virginia). First criticality on Surry No. 1 was achieved July 1972.
Commercial operation began in September 1972. Initial criticality on Surry No. 2 was achieved in March 1973.14.Turkey Point No. 3 and No. 4Florida Power and Light Company is the owner of a four-unit station on Biscayne Bay, Florida. Turkey Point Nos. 3 and 4 of the station are three-loop, 666 MWeplants. Commercial status for Turkey Point No. 3 was achieved in December 1972. Initial criticality for Turkey Point No. 4 was achieved in June 1973.15.Indian Point No. 2 and 3Consolidated Edison Company of New York operates three nuclear units located in Buchanan, New York; two are (units 1 and 2) owned by the company and one (unitNo. 3) is owned by the Power Authority of the State of New York. Units 2 and 3 are Westinghouse PWRs rated at 873 and 965 MWe respectively. Indian Point No. 2 achieved initial criticality in May 1973 and Indian Point No. 3 achieved initial criticality in April 1976.
1.4-8Reformatted PerAmendment 99-0116.Prairie Island No. 1 and No. 2Northern States Power Company is the owner of these two-loop, 530 MWe unitslocated in Welch, Minnesota. Initial criticality was achieved in December 1973 forPrairie Island No. 1 and in December 1974 for Prairie Island No. 2.17.Zion No. 1 and No. 2Commonwealth Edison Company is the owner of these four-loop, 1050 MWe units located on Lake Michigan near Zion, Illinois. Initial criticality was achieved in June1973 for Zion No. 1 and in December 1973 for Zion No. 2.18.KewauneeWisconsin Public Service Corporation, Wisconsin Power and Light Company, and Madison Gas and Electric Company are the owners of this two-loop, 541 MWeplant located in Kewaunee, Wisconsin. Initial criticality was achieved in March 1974.19.Ringhals No. 2Statens Vattenfallsverk (SSPB) is the owner of this three-loop, 822 MWe unit located in Sweden. Initial criticality was achieved in June 1974.20.Donald C. Cook No. 1 and No. 2Indiana and Michigan Electric Company is the owner of these four-loop, 1060 MWe plants located in Bridgman, Michigan. These plants are the first to use theWestinghouse ice condenser containment design. Unit 1 initial criticality was achieved in January 1975, and Unit 2 initial criticality was achieved in 1978.21.TrojanThis four loop, 1130 MWe plant is jointly owned by Portland General Electric Company, Eugene Water and Electric Board, and Pacific Power and LightCompany. In addition to being the first commercial nuclear plant to operate in the Pacific Northwest (located on the Oregon shore of the Columbia River near Rainier, Oregon), Trojan is the first 17 x 17 fuel-rod-per-assembly plant to achieve criticality. Initial criticality was achieved in December 1975.
1.4-9Reformatted PerAmendment 99-0122.Beaver Valley No. 1This three-loop, 852 MWe plant is jointly owned by Duquesne Light Company,Ohio Edison Company, and Pennsylvania Power Company. Beaver Valley No 1 islocated on the Ohio River, 22 miles northwest of Pittsburgh, Pennsylvania.
Commercial operation began in April 1977.23.Salem No. 1Salem No. 1, owned jointly by the Public Service Electric and Gas Company, Philadelphia Electric Company, Atlantic Electric Company, and Delmarva Powerand Light Company, is located on Artificial Island, a man-made peninsula in Salem County, New Jersey. The 1090 MWe, four-loop plant achieved initial criticality in late 1976.1.4.3.2Westinghouse FacilitiesWestinghouse Electric Corporation was contracted to design, fabricate, and deliver the Nuclear Steam Supply System and Nuclear Fuel for the Virgil C. Summer Nuclear Station. Westinghouse provided technical assistance for the installation and startup of their supplied equipment.1.4.4DAMES AND MOORE, INC. QUALIFICATIONS AND EXPERIENCEThe partnership of Dames and Moore, Inc. (D&M) was founded in 1938 in Los Angeles,California. Since then the firm has grown to more than 1400 employees in 25 offices inthe United States and 17 offices in foreign countries.The professional staff of D&M has a diversified background in the fields of land useplanning, socioeconomics, demography, meteorology, air and water quality, biology,ecology, geology, soil and rock mechanics and dynamics, foundation engineering, geophysics, seismology, hydrology, marine engineering and systems management.
1.4-10Reformatted PerAmendment 99-01D&M has served more than 8500 clients in over 100 countries. The firm has performedmore than 30,000 investigations of various types. D&M has been involved in site and environmental studies for close to half of the nuclear power plants under construction or planned in the United States. Within the Southeast, these projects include:Arkansas Power & Light CompanyArkansas Nuclear OneAlabama Power CompanyJoseph M. FarleyCarolina Power & Light CompanyH. B. RobinsonCarolina Power & Light CompanyShearon HarrisDuke Power CompanyOconeeFlorida Power & Light CompanyTurkey PointFlorida Power & Light CompanySouth DadeFlorida Power & Light CompanySt. LucieGeorgia Power CompanyEdwin I. HatchVirginia Electric & Power CompanyNorth AnnaVirginia Electric & Power CompanySurry1.4.5DANIEL CONSTRUCTION COMPANY QUALIFICATIONS ANDEXPERIENCEDaniel Construction Company (Daniel), a division of Daniel International Corporation,Greenville, S.C., has a wide variety of engineering and construction assignments. Theparent company, Daniel International Corporation has in excess of $10,000,000 worth of projects currently in engineering or construction in many parts of the world. A recent survey of the nation's 400 largest contractor rates Daniel 3rd in contract awards, and 32nd in design/construction awards. Daniel has acquired extensive construction andproject management experience in major industrial complexes for the chemical, paper, rubber, textile, aluminum and power generation industries. These construction services involve the ability to meet precise tolerances and specifications on erection, fabrication and equipment installation and have required a thorough knowledge of heavy construction, mechanical, electrical and instrumentation techniques and methods. This experience and the developed capabilities are applicable to the construction of nuclearpower facilities.The Daniel quality assurance program for ASME nuclear code construction has beenevaluated and accepted by an ASME survey team and an interim certificate ofauthorization issued.The certificate of authorization to perform code construction ("N" Stamp) will beobtained following the successful completion of an ASME survey team fieldimplementation and enforcement audit at the Virgil C. Summer Nuclear Station.
02-01 1.4-11Reformatted PerAmendment 99-01Daniel's experience, past and present, include construction of nuclear and fossil fueledpower plants. First project of this nature was construction of the nuclear power CVTR at Parr, S.C. This facility operated several years as a prototype plant. Power plants completed or currently under construction are specified below:1.Alabama Power Co.J. M. Farley Project Unit 1 & 2847 MW (each)2.Carolina Power & Light Co.Shearon Harris Unit 1, 2, 3, & 4900 MW (each)Darlington Elec. Plant (Turbines) 660 MW3.Detroit Edison CompanyFermi II Unit 21150 MW4.Duke Power CompanySpencer Facility
170 M WCliffside Facility 575 MW5.Georgia Power CompanyVogtle Plant (Turbines)300 MW McManus Plant (Turbines) 360 MW Plant Wansley Unit 1 & 2 880 MW (each)6.Kansas City Power & Light CompanyKansas Gas & ElectricLa Cygne Unit 2 600 MW Wolf Creek Unit 1 1150 MW7.Kansas City Power & Light CompanySt. Joseph Light & PowerIatan Station 630 MW 1.4-12Reformatted PerAmendment 99-018.South Carolina Electric & Gas CompanyArthur M. Williams (Turbines)60 MW Arthur M. Williams Unit 1 600 MW Virgil C. Summer Unit 1 918 MW Fairfield Pumped Storage (8 Units) 480 MW9.Union Electric CompanyCallaway Generating Plant Unit 1 & 21150 (each)10.Virginia Electric & Power CompanyBath County Pumped Storage (6 units)2100 MWIn addition to constructing numerous gas turbine power stations, Daniel is currently building a nuclear fuel reprocessing plant for Allied Gulf Nuclear Services.
1.4-13AMENDMENT 97-01AUGUST 1997TABLE 1.4-1WESTINGHOUSE PRESSURIZED WATER REACTOR NUCLEAR POWER PLANTSPlantOwner UtilityLocationScheduledCommercialOperationMweNet Numberof LoopsShippingportDuquesne Light Company; Research
& Development AdministrationPennsylvania1957 90 4Yankee-RoweYankee Atomic Electric CompanyMassachusetts1961 175 4Trino Vercellese (Enrico Fermi)Ente Nazionale per L'Energia Elettrica (ENEL)Italy1965 261 4Chooz (Ardennes)Societe d'Energie Nucleaire Franco-Belge des Ardennes (SENA)France1967 309 4San Onofre No. 1Southern California Edison Co.; San Diego Gas and Electric Co.California1968 430 3Haddam Neck (ConnecticutYankee)Connecticut Yankee Atomic Power CompanyConnecticut1968 575 4Jose' Cabrera-
ZoritaUnion Electrica, S. A.Spain1969 153 1 1.4-14AMENDMENT 97-01AUGUST 1997TABLE 1.4-1 (Continued)WESTINGHOUSE PRESSURIZED WATER REACTOR NUCLEAR POWER PLANTSPlantOwner UtilityLocationScheduledCommercialOperationMweNet Numberof LoopsBeznau No. 1Nordostschweizerische Kraftwerke AG (NOK)Switzerland1969 350 2Robert Emmett GinnaRochester Gas and Electric CorporationNew York1970 490 2Mihama No. 1The Kansai Electric Power Company, Inc.Japan1970 320 2Point Beach No. 1 Wisconsin Electric Power Co.;Wisconsin Michigan Power Co.Wisconsin1970 497 2H. B. Robinson No. 2 Carolina Power and Light Co.South Carolina1971 700 3Beznau No. 2 Nordostschweizerische Kraftwerke AG(NOK)Switzerland1972 350 2Point Beach No. 2 Wisconsin Electric Power Co.;Wisconsin Michigan Power Co.Wisconsin1973 497 2 1.4-15AMENDMENT 97-01AUGUST 1997TABLE 1.4-1 (Continued)WESTINGHOUSE PRESSURIZED WATER REACTOR NUCLEAR POWER PLANTSPlantOwner UtilityLocationScheduledCommercialOperationMweNet Numberof LoopsSurry No. 1Virginia Electric and Power Co.Virginia1972 788 3Turkey Point No. 3Florida Power and Light Co.
Florida1972 666 3Indian Point No. 2 Consolidated Edison Company ofNew York, Inc.New York1973 873 4Prairie Island No. 1 Northern States Power CompanyMinnesota1973 530 2Turkey Point No. 4 Florida Power and Light Co.
Florida1973 666 3Surry No. 2Virginia Electric and Power Co.Virginia1973 788 3Zion No. 1Commonwealth Edison CompanyIllinois19731050 4 1.4-16AMENDMENT 97-01AUGUST 1997TABLE 1.4-1 (Continued)WESTINGHOUSE PRESSURIZED WATER REACTOR NUCLEAR POWER PLANTSPlantOwner UtilityLocationScheduledCommercialOperationMweNet Numberof LoopsKewauneeWisconsin Public Service Corp.;
Wisconsin Power and Light Co.;Madison Gas and Electric Co.Wisconsin1974 541 2Prairie Island No. 2Northern States Power CompanyMinnesota1974 530 2Takahama No. 1The Kansai Electric Power Company, Inc.Japan1974 781 3Zion No. 2Commomwealth Edison CompanyIllinois19741050 4Doel No. 1Indivision DoelBelgium1975 390 2Doel No. 2Indivision DoelBelgium1975 390 2Donald C. Cook No. 1 Indiana and Michigan ElectricCompany (AEP)Michigan19751060 4Ringhals No. 2Statens Vattenfallsverk (SSPB)Sweden1975 822 3 1.4-17AMENDMENT 97-01AUGUST 1997TABLE 1.4-1 (Continued)WESTINGHOUSE PRESSURIZED WATER REACTOR NUCLEAR POWER PLANTSPlantOwner UtilityLocationScheduledCommercialOperationMweNet Numberof LoopsAlmaraz No. 1Union Electrica, S. A.,; Compania Sevillana de Electricidad, S. A.;Hidroelectrica Espanola, S. A.Spain1981 902 3Beaver Valley No.
1Duquesne Light Company; Ohio Edison Company; PennsylvaniaPower CompanyPennsylvania1976 852 3Diablo Canyon No. 1 Pacific Gas and Electric Co.California19801084 4Indian Point No. 3 Consolidated Edison Company ofNew York, Inc.New York1976 965 4Lemoniz No. 1Iberduero, S. A.Spain1982 902 3Salem No. 1Public Service Electric and Gas Company; Philadelphia Electric Co.;Atlantic City Electric Co.; Delmarva Power and Light Co.New Jersey19771090 4 1.4-18AMENDMENT 97-01AUGUST 1997TABLE 1.4-1 (Continued)WESTINGHOUSE PRESSURIZED WATER REACTOR NUCLEAR POWER PLANTSPlantOwner UtilityLocationScheduledCommercialOperationMweNet Numberof LoopsTrojanPortland General Electric Co.; Eugene Water and Electric Board; PacificPower and Light CompanyOregon19761130 4Almaraz No. 2Union Electrica, S. A.; Compania Sevillana de Electricidad, S. A.;Hidroelectrica Espanola, S. A.Spain1984 902 3Asco No. 1Fuerzas Electricas de Cataluna, S. A.
(FESCA)Spain1982 902 3Diablo Canyon No. 2 Pacific Gas and Electric Co.California19811106 4Joseph M. Farley No. 1 Alabama Power CompanyAlabama1977 829 3Ko-Ri No. 1Korea Electric Power Co., Ltd.Korea1978 564 2North Anna No. 1 Virginia Electric and Power Co.Virginia1978 898 3 1.4-19AMENDMENT 97-01AUGUST 1997TABLE 1.4-1 (Continued)WESTINGHOUSE PRESSURIZED WATER REACTOR NUCLEAR POWER PLANTSPlantOwner UtilityLocationScheduledCommercialOperationMweNet Numberof LoopsNorth Anna No. 2 Virginia Electric and Power Co.Virginia1980 898 3Ohi No. 1The Kansai Electric Power Co., Inc.Japan19791122 4Ohi No. 2The Kansai Electric Power Co., Inc.Japan19791122 4Ringhals No. 3Statens Vattenfallsvert (SSPB)Sweden1981 900 3Sequoyah No. 1Tennessee Valley AuthorityTennessee19801148 4Angra dos Reis No. 1 Furnas-Centrais Electricas, S.A.Brazil1981 626 2Asco No. 2Fuerzas Electricas de Cataluna, S. A.
(FESCA); Empresa NacionalHidroelectrica del Ribagorzana, S. A.
(ENHER); Fuerzas Hidroelectricas Spain1983 902 3 1.4-20AMENDMENT 97-01AUGUST 1997TABLE 1.4-1 (Continued)WESTINGHOUSE PRESSURIZED WATER REACTOR NUCLEAR POWER PLANTSPlantOwner UtilityLocationScheduledCommercialOperationMweNet Numberof LoopsDonald C. Cook No. 2 Indiana and Michigan ElectricCompany (AEP)Michigan19781060 4Lemoniz No. 2Iberduero, S. A.Spain1984 902 3Sequoyah No. 2Tennessee Valley AuthorityTennessee19811148 4Watts Bar No. 1Tennessee Valley AuthorityTennessee19811177 4William B.
McGuire No. 1Duke Power CompanyNorth Carolina19811180 4Joseph M. Farley No. 2 Alabama Power CompanyAlabama1980 829 3 KrskoSavske Elektrarne, Ljubljana, Slovenia, Elektroprivreda Zagreb,CroatiaYugoslavia1981 615 2Ringhals No. 4Statens Vattenfallsvert (SSPB)Sweden1982 900 3 1.4-21AMENDMENT 97-01AUGUST 1997TABLE 1.4-1 (Continued)WESTINGHOUSE PRESSURIZED WATER REACTOR NUCLEAR POWER PLANTSPlantOwner UtilityLocationScheduledCommercialOperationMweNet Numberof LoopsSalem No. 2Public Service Electric and Gas Company; Philadelphia Electric Co.;Atlantic City Electric Co.; Delmarva Power and Light Co.New Jersey19801115 4Virgil C. SummerSouth Carolina Electric and Gas Company;South Carolina1981 900 3Watts Bar No. 2Tennessee Valley AuthorityTennessee19821177 4William B.
McGuire No. 2Duke Power CompanyNorth Carolina19821180 4Comanche Peak No. 1 Texas Utilities GeneratingTexas19831150 4Bryon No. 1Commonwealth Edison Co.Illinois19841120 4Seabrook No. 1Public Service Company of New Hampshire; United IlluminatingCompanyNew Hampshire19841200 4 1.4-22AMENDMENT 97-01AUGUST 1997TABLE 1.4-1 (Continued)WESTINGHOUSE PRESSURIZED WATER REACTOR NUCLEAR POWER PLANTSPlantOwner UtilityLocationScheduledCommercialOperationMweNet Numberof LoopsSouth Texas Project Unit No. 1Houston Lighting and Power Co.;
Central Power and Light Co.; CityPublic Service of San Antonio; City of Austin, TexasTexas19831250 4Beaver Valley No. 2Duquesne Light Company; Ohio Edison Company; PennsylvaniaPower Co.; Cleveland Electric Illuminating Company; Toledo Edison CompanyPennsylvania1986 852 3Braidwood No. 1Commonwealth Edison CompanyIllinois19851120 4Callaway No. 1SNUPPS - Union Electric Co.Missouri19831150 4Catawba No. 1Duke Power CompanySouth Carolina19851153 4Ko-Ri No. 2Korea Electric Power Co., Ltd.Korea1983 605 2Braidwood No. 2Commonwealth Edison CompanyIllinois19861120 4 1.4-23AMENDMENT 97-01AUGUST 1997TABLE 1.4-1 (Continued)WESTINGHOUSE PRESSURIZED WATER REACTOR NUCLEAR POWER PLANTSPlantOwner UtilityLocationScheduledCommercialOperationMweNet Numberof LoopsByron No. 2Commonwealth Edison CompanyIllinois19851120 4Catawba No. 2Duke Power CompanySouth Carolina19871153 4Comanche Peak No. 2 Texas Utilities Generating Co.Texas19851150 4Marble Hill No. 1Public Service Company of Indiana, Inc.; Northern Indiana Public ServiceCompanyIndiana19841150 4Millstone No. 3Northeast Nuclear Energy Co.Connecticut19871156 4Seabrook No. 2Public Service Company of New Hampshire; United IlluminatingCompanyNew Hampshire19891200 4South Texas Project Unit No. 2Houston Lighting and Power Co.;
Central Power and Light Co.; CityPublic Service of San Antonio; City of Austin, TexasTexas19861250 4 1.4-24AMENDMENT 97-01AUGUST 1997TABLE 1.4-1 (Continued)WESTINGHOUSE PRESSURIZED WATER REACTOR NUCLEAR POWER PLANTSPlantOwner UtilityLocationScheduledCommercialOperationMweNet Numberof LoopsTaiwan Unit No. 5Taiwan Power CompanyTaiwan1984 950 3Wolf Creek Unit No. 1SNUPPS - Kansas Gas and Electric Company; Kansas City Power andLight Company Kansas19831150 4Alvin W. Vogtle No. 1 Georgia Power CompanyGeorgia19861113 4Taiwan Unit No. 6Taiwan Power CompanyTaiwan1985 950 3Alvin W. Vogtle No. 2 Georgia Power CompanyGeorgia19911113 4Marble Hill No. 2Public Service Company of Indiana, Inc.; Northern Indiana Public ServiceCompanyIndiana19851150 4Shearon Harris No. 1 Carolina Power and Light Co.North Carolina1987 900 3 1.4-25AMENDMENT 97-01AUGUST 1997TABLE 1.4-1 (Continued)WESTINGHOUSE PRESSURIZED WATER REACTOR NUCLEAR POWER PLANTSPlantOwner UtilityLocationScheduledCommercialOperationMweNet Numberof LoopsShearon Harris No. 2 Carolina Power and Light Co.North Carolina1988 900 3Atlantic No.
(O.P.S.)Public Service Electric and Gas Company; Atlantic City Electric Co.;Jersey Central Power and Light CompanyNew Jersey19871150 4Shearon Harris No. 4 Carolina Power and Light Co.North Carolina1988 900 3Sundesert No. 2San Diego Gas and Electric Co.California1988 950 3Sayago No. 1Iberduero, S. A.Spain1980's1000 3Unit No. 4Iberduero, S. A.Spain1980's1000 3Shearon Harris No. 3 Carolina Power and Light Co.North Carolina1990 900 3 1.4-26AMENDMENT 97-01AUGUST 1997TABLE 1.4-1 (Continued)WESTINGHOUSE PRESSURIZED WATER REACTOR NUCLEAR POWER PLANTSPlantOwner UtilityLocationScheduledCommercialOperationMweNet Numberof LoopsUnassigned No. 1 (O.P.S.)Public Service Electric and GasCompany; Atlantic City ElectricCompanyNew Jersey19901150 4Unassigned No. 2 (O.P.S.)Public Service Electric and Gas Company; Atlantic City ElectricCompanyNew Jersey19921150 4
1.5-1Reformatted PerAmendment 99-01NOTE 1.5Section 1.5 is being retained for historical purposes only.1.5REQUIREMENTS FOR FURTHER TECHNICAL INFORMATIONReference [1] presents descriptions of the safety-related research and developmentprograms which are being carried out for, or by, or in conjunction with, Westinghouse Nuclear Energy Systems, and which are applicable to Westinghouse Pressurized Water Reactors.For each program described in this section and still in progress, the safety-relatedprogram is first introduced, followed, where appropriate, by background information.There is, then, a description of the program which relates the program objectives to the problem and presents pertinent recent results. Finally, an alternate position may be given for programs (generally experimental rather than analytical) which have not yet reached a stage where it is reasonably certain that the results confirm the expectation.
The alternate position is one that might be used if the results are unfavorable; it is not necessarily the only course that might be taken.The term "research and development", as used in this section, is the same as that usedby the Commission in Section 50.2 of its regulations, that is:(n)"Research and development" means (1) theoretical analysis, exploration, orexperimentation; or (2) the extension of investigative findings and theories ofa scientific nature into practical application for experimental and demonstration purposes, including the experimental production and testing of models, devices, equipment, materials, and processes.The technical information generated by these research and development programs will be used either to demonstrate the safety of the design and more sharply define marginsof conservatism, or will lead to design improvements.Progress in these development programs will be reported on a timely basis. New safety-related research and development programs, which include existing programswhich become safety-related, will also be described.Included in the overall research and development effort are the programs below whichare applicable to the 17 x 17 fuel assembly. The test programs were completed during1975 in order to support the initial loading of the first 17 x 17 fuel in late 1975.
99-01 1.5-2Reformatted PerAmendment 99-011.5.1VERIFICATION TESTS (17 x 17)The design of the reactor uses a 17 x 17 square array of fuel rods and thimbles in a fuelassembly and is conceptually similar to, but geometrically different from, the 15 x 15array used in previous designs. The 17 x 17 design is considered to be a relatively small extrapolation of the 15 x 15 design. Comprehensive testing has been performed to verify that the extrapolation is sufficiently conservative. Westinghouse maintains that no plant need be designated a prototype andinstrumented to verify the 17 x 17 fuel design. The change in flow induced vibrationresponse of the internals from a 15 x 15 to a 17 x 17 fuel design will be minimal for the following reasons:1.The only structural changes in the internals, other than the fuel assemblies,resulting from the design change from the 15 x 15 to the 17 x 17 fuel assembly arethe guide tube and control rod drive line.2.The guide tube is rigidly attached at the upper core support plate only. The uppercore plate serves only to align the guide tubes. Because of this type of supportarrangement, the guide tube has a minimal contribution to the vibrational responseof the core barrel and other internal components.3.The effective flow area of the 17 x 17 guide tube is essentially the same as that ofthe 15 x 15 guide tube and therefore, there are no significant differences in the flowdistribution in the upper plenum.4.The fuel assembly lateral spring rate was found by applying loads at various gridlocations with both nozzles restrained. This test assures that the effects of the fuelon the vibrational response of the reactor internals will remain essentially unchanged between the 15 x 15 and the 17 x 17 fuel assembly design because of the small variance of mass and spring rate between the tow designs. The preoperational hot functional flow testing presented in Chapter 14 is considered the most conservative test condition since higher flowrates exist.Some of the verification work described herein was conducted using 17 x 17 assemblies of seven grid design whereas, the selected 17 x 17 assembly design has eight grids.Tabulated below are those 17 x 17 tests which utilized a seven grid geometry and the effect of adding an eighth grid.
1.5-3Reformatted PerAmendment 99-01TestParameterEffectFuel AssemblyStructural TestAxial StiffnessNegligible effect at blowdown impact forces[2].Lateral ImpactAdditional grid shares impact load [2].Prototype Assembly Test Pressure DropThe margin between 7 griddesig n P and D-Loop results[3] is adequate tocover the additional Presulting from the additional grid (< 5% increase in P).Lift ForceThe margin between 7 griddesign lift force and D-Loop results[3] is adequate tocover the additional lift force resulting from the additional grid.Rod VibrationDecreased span length results in improvedvibration characteristics and reduced rod wear.Departure from Nucleate Boiling (DNB)DNB CorrelationAddition of a grid increases mixing which increasesDNB margin. The effect of additional grid testing is presented in Reference [4] .Incore Flow MixingThermal Diffusion Coefficient (TDC)TDC increases as grid spacing decreases[5].The above tabulation shows that:1.Additional design changes are not required (e.g., number of new fuel assemblyholddown springs) due to the addition of a grid, and 2.Seven grid test information can be used to assess the adequacy of the eight griddesig n.Additional testing to specifically investigate the eight grid assembly is not required.
1.5-4Reformatted PerAmendment 99-011.5.1.1Rod Cluster Control Spider Tests1.5.1.1.1Test Purpose and ParametersThe 17 x 17 rod cluster control spider is conceptually similar to, but geometricallydifferent from, the 15 x 15 spider. The 17 x 17 spider supports 24 rodlets (the 15 x 15design supports 20) with no vane supporting more than two rodlets (same as the 15 x 15 design).The rod cluster control spider tests verified the structural adequacy of the design. Thespider vane to hub joint was tested for structural adequacy by:1.Vertical static load test to failure, and2.Vertical fatigue test to approximately 3.0 x 10 6 steps.The static load test was performed by applying tensile and compressive loads to thespider. The load was applied parallel to the spider hub and reacted between the spiderhub and fingers. The spider fingers shared the load equally. The number of cycles for the fatigue test was determined from the expected number of steps a control rod drive mechanism would experience during 20 years in a load follow reactor (1.5 x 10 6 steps).The test met the recommended cyclic test requirements of the ASME Code,Section III, Appendix II, paragraph 1520.The spring pack within the spider hub was tested to determine the spring load deflectioncharacteristic as a function of the loading cycles seen by the spring. The test wasterminated after 1000 cycles compared to a 400 cycle (rod drop) design value. The test loads were equal to or greater than that predicted to result in spring material yielding.
These loads were in excess of the design values. The test acceptance criterion was for the spring to retain adequate preload after the repeated cycling.1.5.1.1.2FacilityThe 17 x 17 spider tests were performed at the Westinghouse Engineering MechanicsLaboratory (see Section 1.5.3.2.15).1.5.1.1.3StatusSpider tests have been completed. A vertical static load test approximately seven timesthe design dynamic load did not result in spider vane to hub joint failure. A spider wastested to 2.8 x 10 6 steps without failure. The spider loading was 110 percent of thedesign value for 1.8 x 10 6 cycles and 220 percent of the design loading for 1 x 10 6cycles. Design load is 3600 pounds compression and 1800 pounds tension. The spring test resulted in negligible preload loss.
1.5-5Reformatted PerAmendment 99-011.5.1.2Grid Tests1.5.1.2.1Test Purpose and ParametersThe 17 x 17 grid is conceptually similar to, but geometrically different from the 15 x 15"R" grid. The purpose of the grid tests is to verify the structural adequacy of the griddesign.Load deflection tests have been made on the grid spring and dimple. Grid spring radial(normal) stiffness and the grid dimple radial and tangential stiffness were obtained. Thisinformation was used to verify that the fuel rod clad wear evaluation has been based on conservative values of these parameters. The fuel rod wear evaluation is conservative as shown by the flow test results presented in Reference [3].The grid buckling strength has been determined from tests using both static anddynamic loads. The loads were applied uniformly to the face of the outside strap,transmitted directly through the grid and reacted at the grid face opposite the input. A description of the grid impact test along with a description of the analytical use of the test parameters is given in Reference [2]. These tests are used to verify that grid buckling during a postulated seismic occurrence does not interfere with control rod insertion.1.5.1.2.2FacilityThe grid tests were conducted at the Westinghouse Engineering Mechanics Laboratory(see Section 1.5.3.2.15).1.5.1.2.3StatusThe grid tests have been completed. Test results are in agreement with pretest designvalues. The test results, along with fuel assembly structural test results, were factoredinto the seismic analysis [2].1.5.1.3Fuel Assembly Structural Tests1.5.1.3.1Test Purpose and ParametersThe 17 x 17 fuel assembly tests were performed to determine mechanical strength andproperties. The fuel assembly parameters obtained were as follows:1.Lateral and axial stiffness.2.Impact and internal structural damping coefficients.3.Vibrational characteristics.4.Lateral and axial impact response for postulated accident loads.
1.5-6Reformatted PerAmendment 99-01The parameters obtained from the lateral dynamic tests are used for seismic analysis,while those obtained from the axial tests are incorporated in the loss of coolant (blowdown) accident analysis. The remaining tests are primarily to demonstrate that the assembly has sufficient mechanical strength to preclude damage during shipment, normal handling, and normal operation.The fuel assembly is subjected to both lateral and axial loads to obtain the respectivestatic axial and lateral stiffnesses. The information obtained from these tests is used toestablish parameters primarily for accident analysis since these conditions appear limiting. The axially applied loads, which were well in excess of shipment, normal handling and normal operational design loads, did not result in any fuel assembly permanent deformation or damage.Static loads were incrementally applied to the fuel assembly to determine its lateralstiffness. The tests were accomplished with both nozzles fixed in place and forcesapplied to various grids.The fuel assembly was dynamically tested in a vertical position using core pins tosimulate reactor support conditions. An electrodynamic shaker was attached to thecenter (4th) grid to provide excitation. The fuel assembly mode shapes and corresponding natural frequencies were obtained from displacement transducers. A comparison of analytical and experimental results is given in Reference [2].
Experimental vibrational studies of individual fuel rods were also performed. The rods were tested under simulated fuel assembly support conditions and as assembled in a prototype fuel assembly. The information obtained from these tests included the fundamental frequencies and mode shapes. A general test description and a summary of the results is presented in Reference [3].The fuel assembly axial stiffness was found by incrementally increasing the static load(compressive) and then incrementally decreasing the static load.Lateral impact tests were performed by displacing the center of the assembly with the nozzles fixed in place. The assembly was released and allowed to impact on lateralrestraints at each of the five center grid locations.The axial impact response and damping were found by dropping the fuel assembly fromvarious heights. The axial impact test was performed with the fuel assembly in theupright position.
1.5-7Reformatted PerAmendment 99-01The relevant parameters measured during the lateral and axial impact tests are asfollows:1.Impact duration versus impact load.
2.Impact force versus drop height or initial displacement.
3.Impact damping or restitution as a function of impact force.
A general description of the test procedure, including a description of use of theparameters as related to accident analysis is presented in Reference [2].There is a general axial test buckling criterion which does not allow local buckling of components which could preclude control rod insertion during an accident. The fuelassembly overall buckling and component local buckling is checked during the axial static and dynamic tests. The lateral displacement associated with the fuel assembly overall (beam type) buckling is limited by the reactor internals clearances and therefore buckling does not reduce the fuel assembly ultimate strength. Local component buckling was not experienced during either the static or dynamic tests for loads well in excess of the design values. The general acceptance criteria was not violated.1.5.1.3.2FacilityThese tests were conducted at the Westinghouse Engineering Mechanics Laboratory(see Section 1.5.3.2.15).1.5.1.3.3StatusThe fuel assembly structural tests have been completed. The fuel assembly structuraltest results are factored into the seismic and blowdown analyses [2].1.5.1.4Guide Tube TestsA new guide tube was designed to accommodate the 24 rodlet pattern adopted for 17 x 17 cores. This guide tube is sufficiently strong to provide increased margins of safety over present guide tubes. The main features of the new design are full length enclosures and cylindrical upper enclosures. The 17 x 17 (24 rodlet) pattern reduced the central area available for driveline passage significantly, thus necessitating a generally tighter design of the rod guidance elements.The following guide tube tests are considered as engineering tests.
1.5-8Reformatted PerAmendment 99-01These tests are used as design tools and are not specifically required for demonstrationof plant safety.1.Engineering Prototype Assembly Tests.
2.Guide Tube Drop and Deflection Test.
1.5.1.5Engineering Prototype Assembly Tests1.5.1.5.1Test Purpose and ParametersThe purpose of these tests was to demonstrate that the 17 x 17 fuel assembly anddriveline hardware designs perform as predicted. These tests were run prior to therequired plant functional tests and are used as an engineering information test to obtain experimental data. A single set of driveline hardware, including control rods, was used in the tests. The fuel assemblies and driveline were subjected to flow and system conditions covering those most likely to occur in a plant during normal operation as well as during a pump overspeed transient.These tests are used to verify, from an engineering confidence standpoint, theintegrated fuel assembly and rod cluster control performance in several areas. Dataobtained included pressures and pressure drops throughout the system, hydraulic loadings on the fuel assembly and driveline, control rod drop time and stall velocity, fuel rod vibration and control rod, driveline, guide tube, and guide thimble wear during a lifetime of operation. None of this information is considered to be safety-related.Specifically, two full size 17 x 17 fuel assemblies (one for Phase I and one for Phase IIand III testing), one control rod, drive shaft and control rod drive mechanism wereinstalled and tested in the 24 inch inside diameter x 40 foot high D-Loop at the Westinghouse Test Engineering Laboratory Facility.1.5.1.5.1.1Fuel Assembly Life Test (Phase I)The first fuel assembly was subjected to the maximum expected control rod travelduring one fuel assembly lifetime. The nominal test conditions were a flow velocitybased on the design flowrate, a temperature of 585°F and a pressure of 2000 psig.
These conditions represent an extreme set of conditions.Using an instrumented 17 x 17 prototype fuel assembly, guide tube and rod clustercontrol drive assembly, the tests conducted in the D-Loop obtained information on thefollowing:1.Mechanical integrity and performance 2.Drop time 1.5-9Reformatted PerAmendment 99-013.Fuel rod vibration4.Control rod velocity 5.Hydraulic lift force 6.Guide thimble dashpot pressure Following this, the prototype fuel assembly underwent a complete post test evaluationand the guide tubes and driveline were inspected for any abnormal wear conditions.The purpose of this test was basically to determine the effect of the 17 x 17 fuel assembly and control rod configuration of Items 1 through 6 in Phase I and Items 1 and 2 in Phases II and III. The effect on control rod drop due to a seismic disturbance is evaluated analytically.1.5.1.5.1.2Guide Tube and Rod Cluster Control Life Test (Phase II and Phase III)The second fuel assembly was then installed to continue the test at the same flow andtemperature until 3,000,000 total steps of the driveline were accumulated. For PhasesII and III, testing was run at temperatures between 100°F and 585°F and at flowrates from 50 percent to 140 percent of the design flowrate.The test included a program of control rod drops and mechanism stepping thatapproximates the driveline duty for the design lifetime of an operating plant (Phase II).Specifically, approximately 1,275,000 mechanism steps and approximately 270 control rod drops were accumulated. The components were then inspected. Following inspections, testing was continued until a total of 3,000,000 mechanism steps and approximately 500 control rod drops were accumulated (Phase III testing).These tests were directed toward:
1.Life wear evaluation 2.Drop time 3.Rod stall characteristics These tests are not safety-related tests; they are used as engineering tools. Finalverification is demonstrated during the precritical rod drop tests.At the completion of Phase II tests, the test assembly was inspected to determine guide tube and driveline wear characteristics. This inspection was repeated at the end of thetest (Phase III).
1.5-10Reformatted PerAmendment 99-011.5.1.5.2FacilityThe above testing is conducted in the Westinghouse Test Engineering LaboratoryFacility (see Section 1.5.3).1.5.1.5.3StatusThe D-Loop testing has been completed. The results of the testing are given inReferences [3] and [6].1.5.2LOCA HEAT TRANSFER TESTS (17 x 17)1.5.2.117 x 17 Reflood Heat Transfer TestsExtensive experimental programs have recently been performed with a simulated 17 x17 assembly to determine its behavior under Loss of Coolant Accident (LOCA) conditions. The 17 x 17 tests were conducted in the G-Loop facility at theWestinghouse Forest Hills Laboratory.Results from the 17 x 17 programs were compared with data from the 15 x 15 assemblytest programs and were used to confirm predictions made by correlations and codesbased on the 15 x 15 test results (see Reference [7]).1.5.2.2Facility DescriptionThe 17 x 17 test facility provides experimental measurements on the refloodingbehavior of a 17 x 17 rod array following a LOCA. The test assembly consists of an array of 336 electrically heated rods and 25 guide tube thimbles arranged in a 17 x 17 array. The heater rod diameter, the active heated length, and pitch spacing is identical to that used in the 17 x 17 fuel. There were eight Westinghouse production mixing vane grids in the bundle.1.5.2.3Delayed Departure From Nucleate Boiling Testing1.5.2.3.1IntroductionThe NRC Acceptance Criteria for Emergency Core Cooling Systems (ECCS) forLight-Water Power Reactors was issued in Section 50.46 of 10 CFR 50 on December28, 1973. It defines the basis and conservative assumptions to be used in the evaluation of the performance of the ECCS. Westinghouse believes that some of the conservatism of the criteria is associated with the manner in which transient DNB phenomena are treated in the evaluation models. Transient critical heat flux data presented at the 1972 specialist's meeting of the Committee on Reactor Safety Technology (CREST) indicated that the time to DNB can be delayed under transient conditions. To demonstrate the conservatism of the ECCS evaluation models, Westinghouse has initiated a program to experimentally simulate the blowdown phase of a LOCA. This testing is scheduled for the middle of 1976 as part of an Electric Power 1.5-11Reformatted PerAmendment 99-01Research Institute (EPRI) sponsored Blowdown Heat Transfer Program. This programis scheduled for completion at the end of 1976. Development of a transient DNB correlation for use in the Westinghouse ECCS analysis is planned for completion in 1977.1.5.2.3.2ObjectiveThe objective of the delayed departure from nucleate boiling (DDNB) test is todetermine the time that DNB occurs under LOCA conditions. This information will beused to confirm the existing, or develop a new, Westinghouse transient DNB correlation.
The steady-state DNB data obtained from 15 x 15 and 17 x 17 test programs can be used to assure that the minimal geometric differences between the two fuel arrays can be correctly treated in the transient correlations.1.5.2.3.3ProgramThe program is divided into two phases. Phase I simulates PWR behavior during aLOCA to permit definition of the time delay associated with onset of DNB. Tests in thisphase will cover the large double ended guillotine cold leg break. All tests in Phase I are started after establishment of typical steady-state operating conditions. The fluid transient is then initiated, and the rod power decay is programmed in such a manner as to simulate the actual heat input of fuel rods. The test is terminated when the heater rod temperatures reach a predetermined limit (dependent on power level).Typical parameters which can be studied under Phase I testing are shown inTable 1.5-1.Phase II will provide separate effects data to permit heat transfer correlationdevelopment. The Phase II tests will also start from steady-state conditions, with sufficient power tomaintain nucleate boiling throughout the bundle. Controlled ramps of decreasing testsection pressure or flow will initiate DNB. By applying a series of controlled conditions, investigation of the DNB will be studied over a range of qualities and flows and at pressures relevant to a PWR blowdown.Typical parameters which can be studied under Phase II testing are shown inTable 1.5-2.1.5.2.3.4Test DescriptionThe experimental program is being conducted in the J-Loop at the Westinghouse ForestHills Facility with a full length 5 x 5 rod bundle simulating a section of a 15 x 15assembly to determine DNB occurrence under LOCA conditions.
1.5-12Reformatted PerAmendment 99-01The heater rod bundles to be used in this program will be assembled using internallyheated rods and Westinghouse grids. The heater rods are designed for high reliability, long life, and high power density. The maximum power is 18.8 kW/ft, and the total power is 135 kW for extended periods over the 12 foot heated length of the rod. Heat is generated internally by means of a varying cross-section, rugged, tubular resistor which approximates a U 2 cos U power distribution, skewed to the bottom. Each rod isadequately instrumented with a total of 20 thermocouples (8 resistors and 12 sheath thermocouples).1.5.2.3.5ResultsThe experiments in the DDNB facility will result in cladding temperature and fluidproperties measured as a function of time throughout the blowdown range from 0 to 20seconds.Facility modifications and installation of the initial test bundle have been completed. Aseries of shakedown tests in the J-Loop have been performed. These tests provideddata for instrumentation calibration and check-out, and provided information regarding facility control and performance. Initial program tests were performed during the firsthalf of 1975.1.5.2.4Single Rod Burst TestThe single rod burst test results are used to quantify the maximum assembly flowblockage which is assumed in LOCA analyses.Previously, single rod and multi-rod burst tests have been completed on the 15 x 15 fuelassembly rods under conditions which exist during the LOCA. The conclusion of thesetests was that fuel rods burst in a staggered manner so that maximum average flow area blockage for the assembly is 55 percent during blowdown and 65 percent during reflood, based on the characteristics of the PWR fuel rod and the conservative peak clad temperature predicted during the LOCA transient.The single rod burst test program for the 17 x 17 fuel assembly rods consisted ofbursting specimens at the various internal pressures and heating rates in a steamatmosphere.In addition, tests were run on 15 x 15 fuel assembly rods to assure reproducibility of the1972 single rod burst test results. Results from the program are documented inReferences [8] and [9].
1.5-13Reformatted PerAmendment 99-01The single rod burst tests and evaluation have been completed and are reported inReferences [8] and [9]. Results of the tests showed that the LOCA behavior of the 17 x 17 clad in comparison to that of the 15 x 15 clad exhibited no significant differences in failure ductility. Because of the results and the geometric scaling, the flow blockage
(%) as determined by 15 x 15 multi-rod burst test simulation can be used for the 17 x 17 geometry.1.5.2.5Power-Flow MismatchDesign emphasis has been placed on reliable and effective control and protectionsystems for the reactor core and engineered safety features to ensure adequate margins within incidents can be terminated before the onset of fuel failure. For this reason, investigations into the mechanisms of fuel failure and its propagation, and phenomena of molten fuel-coolant interaction have been limited.Obtaining complete answers to the question related to fuel rod failure will requireextensive testing because of the multiplicity of parameters involved. Such an extensiveinpile test program has been proposed by the Aerojet Nuclear Company under NRC (AEC) sponsorship for the Power Burst Facility (PBF).The proposed PBF program is expected to simulate conditions appropriate to major accidents postulated for reactor systems, i.e., loss of flow, loss of coolant, and reactivityinitiated accidents. Phase I of the proposed PBF program is expected to be completed approximately two years after facility checkout or by mid 1977. Phase I is expected toconcentrate upon establishing the thresholds for and the consequences of fuel failure and utilizes fuel rod clusters of various sizes to determine the extent of failurepropagation. This program would broaden the experimental basis for evaluating reactor safety, but should not be considered essential for the design and safe operation.Westinghouse will closely follow any such experimental program or analytical studiesthat may become available. Until such information is available, clearly demonstratingthat local fuel melting is an acceptable condition, the emphasis in design will continue tobe placed on providing adequate margins to minimize the probability of fuel melting.The margins incorporated in the design within which incidents can be terminated, before the onset of fuel failure, provide a sound basis for safe operation.
1.5-14Reformatted PerAmendment 99-011.5.3WESTINGHOUSE TEST ENGINEERING LABORATORY FACILITY1.5.3.1IntroductionThe Test Engineering Laboratory at Forest Hills, Pennsylvania, has long been the majorWestinghouse center for nuclear research and development. The Test Engineering Laboratory is totally involved with the design and implementation of facilities and programs to prove the reliability of Westinghouse PWR concepts and components.The Test Engineering Laboratory has full in-house capabilities to design and constructPWR loops for both hydraulic and heat transfer testing programs.Historically the Test Engineering Laboratory has been in a state of transition, depending upon the current need for its services. Today's great need is for ECCS data and theverification of many new PWR system components. Past needs and accomplishments have included the development of supercritical heat transfer once through loops; rod cluster control drive mechanisms; fuel assemblies; underwater handling tools; and fuel assembly grid design, among many other earlier projects. Testing has included air filter tests; water chemistry tests; inpile testing for fuel rods; single fuel rod burst tests; hydraulic studies on fuel assemblies; and corrosion testing of Zircaloy and other PWR components and materials, with and without heat transfer.The Test Engineering Laboratory is a very flexible installation, one which will continue toexpand and develop as future needs for its services arise. Its staff, too, variesaccording to requirements.There are currently more than 100 persons involved in laboratory projects, including 12electrical and mechanical engineers, more than 75 highly skilled technicians, and some30 specialists from other divisions of Westinghouse. The Test Engineering Laboratory has the option of obtaining personnel from the entire Corporation, depending upon the need for specific skills, knowledge and experience.Ongoing research performed at the Test Engineering Laboratory continues to demonstrate the reliability of Westinghouse PWR plant components and greatlyfacilitates the development of improved reactor system components. As the test center for Westinghouse Nuclear Energy Systems, the Test Engineering Laboratory is totally committed to the advancement of the nuclear energy industry.
1.5-15Reformatted PerAmendment 99-011.5.3.2Tests, Test Loops and EquipmentThis section contains a brief description of the major tests, test loops and testequipment at the Westinghouse Test Engineering Laboratory Facility.1.5.3.2.1A and B-Loops, Low Flow/High Pressure Hydraulic FacilitiesThese loops are small, high pressure, stainless steel facilities, used for testing smallcomponents and individual parts of larger components under normal working conditions.A canned motor pump circulates water in both the A-Loop and the B-Loop at 150 gpm.
Operating temperatures are obtained from the conversion of the pumping power into heat, as well as from external heaters. Typical tests run in these loops are:1.Full scale gate and check valves; 2.Material corrosion-erosion, with variable water chemistry; and 3.Corrosion product release and transport properties of crud.Characteristics of A and B LoopsMaximum Flowrate150 gpm at 300 feetMaximum Pump Head335 feet at 60 gpmMaximum Allowable Temperature 650FNormal Working Pressure2000 psiNormal Working Temperature 600F1.5.3.2.2D-Loop, Medium Flow/High Pressure Hydraulic Facility The D-Loop is a flexible test facility used for demonstrating the interplay of reactorsubsystems and evaluating component design concepts. It contains a canned motorpump, which produces a 290 foot head at 3000 gpm. All piping (10 inch Schedule 160)in contact with the primary water is stainless steel. Loop pressure is established and maintained by an air driven charging pump operating in conjunction with a gas loaded back pressure valve. Most of the power required to establish and maintain loop temperature is derived from the circulating pump operation, and 75 kW of heat is available from electric strip heaters.
1.5-16Reformatted PerAmendment 99-01The D-Loop services a 24 inch inside diameter x 40 foot long test vessel, whichaccommodates full scale models of large PWR core components for operational studies.Characteristics of D LoopMaximum Flowrate4400 gpmMaximum Allowable Pressure2400 psiMaximum Allowable Temperature 650FNormal Working Pressure2000 psiNormal Working Temperature 600FPump Head at 3000 gpm290 feetMaximum Pump Head340 feet (at 1500 gpm)Main Loop Flow Measurement10 inch venturiAuxiliary Flow Measurement6 inch venturis (2 inch branch lines)1.5.3.2.3E-Loop, Low Flow/Low Pressure Hydraulic FacilityThe E-Loop is a low pressure, six inch, stainless steel loop, with 2 circulating pumps.These pumps may be connected in parallel, giving 2000 gpm at a 130 foot head, or inseries, giving 1000 gpm at a 260 foot head. Flow and vibration studies are conducted with this loop, and, because of its low pressure, plastic models for visual observation or photography may be used. In addition, a four inch Rockwell water meter in a branch line permits the calibration of flow meters up to 800 gpm.Characteristics of E LoopMaximum Flowrate2000 gpm at 130 feet 1000 gpm at 260 feetMaximum Working PressurePump Head1.5.3.2.4G-Loop, Emergency Core Cooling System FacilityThe G-Loop is a high pressure, ECCS test facility designed and fabricated to ASMESection I for 2000 psi and 650°F. It consists of a main test section and vessel, exhaustsystem, piping, separators and muffler, flash chamber steam supply system, and high pressure/low pressure cooling systems.This loop is basically designed to obtain test data for analysis of LOCAs, for breaks upto and including double ended pipe breaks for PWRs. Tests are initiated at simulatedconditions existing 8 seconds after the start of a LOCA. A typical run consists of constant power and pressure, followed by pressure blowdown, power decay and 1.5-17Reformatted PerAmendment 99-01emergency core cooling. The G-Loop is capable of performing the following methods ofemergency core cooling: Current, Upper Head Injection (UHI), UHI w/Current and other Core Spray Systems. It may also be used for constant temperature/pressure small leg break tests (core uncovering tests). These consist of boiling off water at a constant bundle power input until the rods can no longer be cooled.The G-Loop test bundle consists of 480 electrically heated rods, 16 grid supportthimbles, and 33 spray thimbles bounded by an octagonal stainless steel baffle andarranged as per a four-loop 15 x 15 rod bundle configuration. The loop is controlled (fully automated during transients) through a PDP-II-DEC-16K computer with a 600 point Computer Products A-D Converter operating at a sweep rate of 40,000 points per second for data acquisition. Figure 1.5-1 is a schematic of the G-Loop test facility.G-Loop System Components and CharacteristicsRatedTypical OperatingComponentMaterial Pressure (psi)Temperature (F)Pressure (psi)Temperature (F)Test VesselCarbon Steel2000 6501000 545Downcomer Side TankCarbon Steel2000 6501000 545In-Line MixerCarbon Steel2000 6501000 545Mixer AccumulatorStainless Steel2500 6501800 100 Flash Cham berCarbon Steel3000 7002800 660Separators Nos. 1&2 Carbon Steel2000 6501000 545Spray AccumulatorsNos. 1&2Carbon Steel2000 6501800 150Spray Accumulator No. 3 Stainless Steel2500 6501800 150Reflood Tan kStainless SteelAtmosphere 212Atmosphere 150Primary Pipi ngCarbon Steel2000 6501000 545 02-01 1.5-18Reformatted PerAmendment 99-011.5.3.2.5H-Loop, High Flow Hydraulic FacilityThe H-Loop constitutes a versatile hydraulic facility, capable of supplying 14,000 gpm ofwater at a developed head of 600 feet and at temperatures as high as 200°F. Thisfour-loop system can simultaneously handle either full scale prototype test assemblies, or one large scale reactor model. The major purpose of the H-Loop is to permit the use of 1/7 scale reactor models and full scale fuel assemblies for conducting mixing studies, flow distribution studies, and similar low temperature/low pressure hydraulic tests.Characteristics of H LoopMaximum Flowrate14,000 gpmPressure Drop Across Vessel Model 120 psiMinimum Vessel Outlet Pressure10 psigFlow Accuracy 1/2%Water Temperature Range7-200FMaximum Loop-to-Loop Temperature Variation 2FMaximum Loop-to-Loop Flowrate Variation 3%1.5.3.2.6J-Loop, Delayed Departure from Nucleate Boiling Heat Transfer FacilityThe J-Loop is a completely instrumented pressurized water test facility for verifyingDDNB phenomena during a LOCA, and for conducting steady-state heat transferstudies. This test loop is a full size, single-loop simulation of a typical four-loop reactor system; it will accept a full length 5 x 5 bundle of internally heated fuel rods. The J-Loop is designed to operate at 2500 psia at 650°F, and at variable flowrates of up to 450 gpm. During LOCA tests, fluid input to the reactor vessel is closely controlled by 2 servo-controlled mixers, which inject a two-phase water/steam mixture into the test vessel, to simulate flow from the unbroken loops. Figure 1.5-2 is a schematic of the J-Loop test facility.Characteristics of J-LoopTest FluidDemineralized WaterDesign Pressure2500 psiaDesign Temperature 650FMaximum Flowrate (hot)450 gpmPower Input to Test Vessel3,500,000 watts (maximum)Primary Test Heat Exchanger Rating11,400,000 BTU/hour 02-01 1.5-19Reformatted PerAmendment 99-011.5.3.2.7K-Loop, Boron Thermal Regeneration TestThe K-Loop, Boron Thermal Regeneration System (BTRS) test facility is used to studythe performance and to verify the component sizing of both the currently availableTHERM I and the improved THERM II BTRS. The function of this system is to process boron-containing effluents from the Reactor Coolant System (RCS) to yield a high boron concentration fraction, which can be used to borate the RCS. A relatively boron free fraction is also processed, which can be used to dilute the RCS, such as that required in load follow operations.Characteristics of K-LoopTotal Tank Capacity30,000 gallons Chiller Capacity48 ice-tonsMaximum Ion Exchange Resin Test Volume75 ft 3Maximum Test Process Rate Capability10 gpm/ft 2 bed areaMaximum Flow Test Capability200 gpmMinimum Boron Storage Mode Fluid Temperature 50FMaximum Boron Release Mode Fluid Temperature 160F1.5.3.2.8FLECHT-SET, Emergency Core Cooling System FacilityThe FLECHT-SET is a low pressure facility, designed to provide experimental data onthe influence of system effects on ECCS during the reflood phase of a LOCA.The facility consists of a once through system, including an electrically heated testsection (fuel rods and housing), accumulator, steam generator simulators, pressurizer,catch vessels, instrumentation, and the necessary piping to simulate the reactor primary coolant loop. Data acquisition is accomplished through a PDP-II-DEC-16K Computer with a 256 point Computer Products A-D Converter, operating at a sweep rate of 1200 points per second.Characteristics of FLECHT-SET100 Rod Bundle Maximum Power100 kWMaximum Bundle Flooding Rate86 gpmWater Temperature Range100-200°FSystem Pressure0-60 psia 02-01 1.5-20Reformatted PerAmendment 99-011.5.3.2.9Single Rod Loop, Heater Rod Development FacilityThe Single Rod Loop is used to evaluate prototype heater rods and for in-depth study ofexisting rods in pressurized water systems. The test section of the loop is easilyreplaced to facilitate the installation of various length and diameter heater rods. The Single Rod Loop is electrically controlled and operated by one person. Steady-state and blowdown at various conditions can be simulated in the loop. The main test section can be replaced with a quartz tube, and DNB phenomenon can be observed on a single rod with a remotely operated camera.Characteristics of Single Rod LoopMaximum Operating Pressure2250 psiaMaximum Operating Temperature 650°FMaximum Flowrate10 gpmSystem Capacity5 gallonsMaximum Power Available200 kWPiping Size1 and 3 inch1.5.3.2.10Hydraulic Model TestingMiscellaneous hydraulic tests on mock-ups of reactor system parts and components areroutinely performed at the Test Engineering Laboratory. Typical of this type of testingare the two discussed below, which were recently completed:1.Emergency Core Cooling Flow DistributionA 10 x 10 rod bundle was installed in a plastic housing with a water supply at thetop. A grid collection unit at the bottom of the bundle collected the water as itflowed through the model and diverted it to the measuring tubes at the base.
Knowledge of the flow distribution in the bundle was obtained in this manner.2.Sample System Mixing TestThis test used one thermocouple to measure the temperature of water from four locations in a reactor. The purpose of the procedure was to determine whether theindication from the single thermocouple was representative of the average temperature of the four water supplies. A mock-up of the mixing chamber was constructed so that hot or cold water (at closely controlled pressure) could besupplied to any of the four inlets. By running combinations of hot and cold inlets and making simultaneous recordings of the various temperatures, highly useful information was obtained.
1.5-21Reformatted PerAmendment 99-011.5.3.2.11Autoclave TestingThe Test Engineering Laboratory is equipped with autoclaves ranging in size from 1/2gallon to 100 gallons. These devices are in constant use to determine the effect ofvarious water chemistries on core components, as well as to perform corrosion tests.
The units have also been used as boilers to provide steam for miscellaneous development tests, including acoustic leak detection.1.5.3.2.12Mechanical Component and Vibration TestFull scale mechanical and vibration tests are performed at the Test EngineeringLaboratory on plant and reactor components to prove the reliability of equipment design.Vibration testing of reactor components is also performed in the laboratory, using electronically excited shaker heads. Three sizes are available (2 lbs, 50 lbs, and 150 lbs) for regular scale model testing for frequencies from 5 Hz to 50 Hz.1.5.3.2.13Electrical Component AssemblyHighly skilled technicians are available at the Test Engineering Laboratory forconstructing complex control and instrumentation systems. Work is initiated withengineering ideas and sketches, and includes mounting of process controllers, recorders, meters, relay logic, protection circuits, switches and indicators.Point-to-point wiring is used, as required. Final as-built drawings are prepared,inspection and thorough electrical checkout is performed before installation in a facility.1.5.3.2.14Surveillance System DevelopmentSurveillance systems provide on-line monitoring of pressure vessels for flaws.Electronic components are being developed at the Test Engineering Laboratory for anacoustic emission monitoring system for inservice inspection of operating plant vessels and piping. This system is designed to detect and locate initiation and propagation of cracks at various locations, such as welds and stress risers. Vessel flaw growth andrupture data have been obtained through joint programs at the Idaho National Engineering Laboratories and at the Oak Ridge National Laboratories. Pipe rupture data has been obtained from NRC (AEC) sponsored tests, and hydrostatic test data, operational noise and attenuation characteristics have been measured at various Westinghouse operating plants.1.5.3.2.15Engineering Mechanics LaboratoryBench tests are performed in fixtures designed for the particular test using standard testequipment and techniques.
1.5-22Reformatted PerAmendment 99-011.
5.4REFERENCES
1.Eggleston, F. T., "Safety-Related Research and Development for WestinghousePressurized Water Reactors, Program Summaries, Spring 1976," WCAP-8768,June, 1976.2.Chiang, D., Gesinski, L. and Nakazato, S., "Safety Analysis of the 17 x 17 FuelAssembly for Combined Seismic and Loss of Coolant Accident," WCAP-8236(Proprietary), December, 1973 and WCAP-8288 (Non-Proprietary), January, 1974.3.DeMario, E. E. and Nakazato, S., "Hydraulic Flow Test of the 17 x 17 FuelAssembly," WCAP-8278 (Proprietary) and WCAP-8279 (Non-Proprietary),February, 1974.4.Motley, F. E., Wenzel, A. H. and Cadek, F. F., "Critical Heat Flux Testing of the 17x 17 Fuel Assembly Geometry with 22 Inch Grid Spacing," WCAP-8536(Proprietary) and WCAP-8537 (Non-Proprietary), May, 1975.5.Cadek, F. F., Dominicis, D. P. and Motley, F. E., "Effect of Axial Spacing onInterchannel Thermal Mixing with "R" Mixing Vane Grid," WCAP-7941-P-A(Proprietary) and WCAP-7959-A (Non-Proprietary), January, 1975.6.Cooper, F. W., Jr., "17 x 17 Driveline Components Tests - Phase IB, II, III D-LoopDrop and Deflection," WCAP-8446 (Proprietary) and WCAP-8449(Non-Proprietary), December, 1974.7.Burnett, A. J. and Kopelic, S. D., "Westinghouse ECCS Evaluation Model-October1975 Version," WCAP-8622 (Proprietary) and WCAP-8623 (Non-Proprietary),November,1975.8.Kuchirka, P. J., "17 x 17 Design Fuel Rod Behavior During Simulated LOCAConditions," WCAP-8289 (Proprietary) and WCAP-8290 (Non-Proprietary),November, 1974.9."A Temperature Sensitivity Study of Single Rod Burst Tests," WCAP-8289,Addendum 1 (Proprietary) and WCAP-8290, Addendum 1 (Non-Proprietary),December, 1975.
1.5-23AMENDMENT 97-01AUGUST 1997TABLE 1.5-1DELAYED DEPARTURE FROM NUCLEATE BOILINGPHASE I TEST PARAMETERSParametersNominal ValueINITIAL STEADY-STATE CONDITIONS Pressure2250 psiaTest section mass velocity2.5 x 10 6 lb/hr-ft 2Inlet coolant temperature 560FMaximum heat flux531,000 BTU/hr-ft 2TRANSIENT CONDITIONS Simulated breakDouble ended cold leg guillotinebreaks TABLE 1.5-2DELAYED DEPARTURE FROM NUCLEATE BOILINGPHASE II TEST PARAMETERSParametersNominal ValueINITIAL STEADY-STATE CONDITIONS Pressure1750 to 1900 psiaTest section mass velocity2.0 to 3.0 x 10 6 lb/hr-ft 2Core inlet temperature 530 to 560FMaximum heat flux440,000 to 560,000 BTU/hr-ft 2TRANSIENT RAMP CONDITIONS Pressure decrease0 to 350 psi/sec (subcooleddepressurization)Flow decrease0 to 100% secInlet enthalpyConstant
1.8-1Reformatted PerAmendment 99-01NOTE 1.8Section 1.8 is being retained for historical purposes only.1.8TMI ACTION PLAN REQUIREMENTSA cross reference between various TMI Action Plan Requirements (NUREG-0737,November 1980) addressed in the FSAR and the appropriate FSAR sections ispresented by Table 1.8-1.
99-01 1.8-2Reformatted PerAmendment 99-01TABLE 1.8-1 CROSS REFERENCETMI ACTION PLAN REQUIREMENTS TO FSAR SECTIONSACTION PLANREQUIREMENTFSAR SECTION/SCE&G LETTERS I.A.1.113.1.2.1, 13.1.2.2.2, 13.1.3.1.12, 13.2.1, SCE&G letter to NRC dated 1/2/81 I.A.1.2 13.1.2.2, 13.5.1.3.1 I.A.1.3 13.1.2.3, 13.5.1.3, SCE&G Letter to NRC dated 1/22/82 I.A.2.113.2.1, SCE&G letters to NRC dated 10/28/80, 10/31/80, 5/22/84, 7/19/84, 11/28/84 I.A.2.313.2, SCE&G letter to NRC dated 10/28/80 I.A.3.113.2.1, 13.2.2, SCE&G letter to NRC dated 10/28/80 I.B.1.213.1.1.4, SCE&G letter to NRC dated 1/2/81 I.C.16.3.3.3.1, 13.5.2, SCE&G letters to NRC dated 11/14/80, 12/2/80, 3/17/82, and
W estinghouse Owners Group Letter OG-47 dated 12/15/80 I.C.2 13.5.1.3 I.C.313.5.1.3.1, Tech Spec (6.1.2)
I.C.4 13.5.1.3 I.C.5 13.5.1.13 I.C.613.5.1.6, SCE&G letter to NRC dated 12/11/80 I.C.713.5.1.3.2, SCE&G letter to NRC dated 12/2/80 and 12/22/80 I.C.813.5.1.3.3, SCE&G letter to NRC dated 11/14/80 I.D.11.2.3.1, SCE&G letters to NRC dated 11/12/80, 1/15/81, 11/25/81, 2/23/82, 3/26/82, 6/11/82, 4/15/83, 7/21/83, 10/28/83, 4/4/84, 4/15/85 I.D.27.7.3, SCE&G letters to NRC dated 3/31/83, 4/15/83, 12/28/83, 4/18/85, 12/23/85I.G.114.1.4.4, SCE&G letters to NRC dated 10/31/80, 12/2/80, 12/22/80, 3/31/82 and
7/29/82 II.B.15.5.15, SCE&G letters to NRC dated 2/19/81, 12/30/81, 1/18/84 II.B.212.1.2.3, App. 12A, SCE&G letters to NRC dated 8/27/80 and 11/21/80 II.B.39.3.2, App. 12A, SCE&G letter to NRC dated 3/30/82 II.B.4 13.2.1, 13.2.2, SCE&G l e tter to NRC dated 10/28/80 99-01 99-01 99-01 99-01 99-01 1.8-3Reformatted PerAmendment 99-01TABLE 1.8-1 CROSS REFERENCETMI ACTION PLAN REQUIREMENTS TO FSAR SECTIONSACTION PLANREQUIREMENTFSAR SECTION/SCE&G LETTERS II.D.15.5.13.4, SCE&G letters to NRC dated 3/25/81, 7/29/81, 3/26/82, 4/1/82, 6/29/82 and 7/30/82 II.D.31.2.3.1, 1.7, 5.5.10.2.2.4, 5.6, 7.7.4, SCE&G letters to NRC dated 1/13/81, 2/26/82, 3/12/82 and 8/26/82 II.E.1.1SCE&G letters to NRC dated 8/15/80, 11/5/80 and 12/2/80 II.E.1.2 7.3.1.1.1, 7.3.2.2, 7.3.2.3, 7.5.1, 10.4.9.1, 10.4.9.2, 10.4.9.3, 10.4.9.5.3 II.E.3.1 8.3.1.1.1.a, 8.3.1.1.3, SCE&G l e tter to NRC dated 2/23/82 II.E.4.1 6.2.5.2.1 II.E.4.26.2.4.3; 9.4.8.2.2, Items 8.l and 8.m; 9.4.8.2.3, Items 8.h and 8.i II.F.1 6.2.5.1.3, 6.2.5.2.3, 6.2.5.3.3, 6.2.5.4.3, 6.2.5.5.3, 6.2.5.5.4, 7.7.3.1, 11.4, 11.4.2, Fi gure 11.4-2, 12.2.5, Fi gure 12.2-2, 12.3.2.2, SCE&G letters to NRC dated 8/28/80, 12/22/80, 6/30/82, 1/18/84, 3/22/84 II.F.2 1.2.3.1, 5.6, SCE&G l etters to NRC dated 12/4/80, 12/15/80, 12/30/80, 2/19/81, 6/8/81, 7/30/81, 1/18/82, 3/16/82, 4/30/82, 7/20/82, 3/8/83, 3/10/83, 4/22/83, 8/26/83, 2/17/84, 4/13/84, 4/16/84, 4/30/84 and 6/6/84II.G.1 7.4.1.2.1, 8.3.1.1.3 II.K.17.3.1.1, 13.5.1.6, and Technical Specifications II.K.2SCE&G letters to NRC dated 12/11/80, 1/6/81, and 12/31/81 II.K.35.2.2.3; 7.2.1.1.2, Item 6; Figure 7.7-4; Technical Specifications, SCE&G letters to
NRC dated 9/9/80, 1/6/81, 1/12/81, 2/19/81, 3/10/81, 3/23/81, 12/31/81, and
4/22/83, 5.5.1.3.13, 13.5.1.3.4, 13.5.1.14, 15.3.1.2.4, W estinghouse letters to NRC(NS-EPR-2581) dated 3/26/82 and (NS-EPR-2581) dated 6/28/82 III.A.1.1Radiation Emergency Plan, SCE&G letters to NRC dated 5/12/81 and 6/16/81 III.A.1.21.2.3.1, 6.4, 7.7.3, Appendix 12A, SCE&G letters to NRC dated 3/16/82, 7/23/82, 8/6/82, 11/24/82, 12/3/82, 3/31/83, 4/15/83, 5/16/83, 9/14/84 III.A.22.3.3.2, Radiation Emergency Plan, SCE&G letters to NRC dated 5/12/81, 6/16/81, 3/31/83 III.D.1.16.3.2.11.2, Technical Specifications, SCE&G letter to NRC dated 2/23/82 III.D.3.3 6.2.5.1.3, 6.2.5.2.3, 6.2.5.3.3, 6.2.5.4.3, 6.2.5.5.4, 12.1.4.2, 12.3.2.2.4 III.D.3.42.2.1, 2.2.2, 2.2.3, 6.4, 15.4, SCE&G letters to NRC dated 11/15/80, 12/15/80, 1/18/84 99-01 99-01 99-01 99-01 99-01 99-01 99-01