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DOCKETED ocEo 4 e:N O JAN 61967> ; i
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ATOMIC UNERGY CO.'15tISSION 4 .a
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., s In the Matter of tlic Application of: DocketNos.(50_2.53 50-265 CO5D10NNE ALTil EDISON COMPANY '
Pr 1 Applicant's Exhibit-No. 1
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Sunmary Description of Quad-Citics Units 1 and 2 and Review of Considerations Important to Safety .
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L. . I lf CONTENTS l~ l ITEM PAGE l
1.0 INTRODUCTION
1.1 General . 1-1 l 1.2 Applicability to Statement of Issues in Notice of licaring . 1-2 2.0 SITU AND ENVIRONS 2-1 3.0 SU,'IMARY DESCRIPTION OF Tile FACILITY 3.1 Introduction 3-1 3.2 Reactor Primary System 3-2
- 3. 3~ Containment Systems 3-3 3.4 Auxiliary and Standby Cooling Systccs 3-4 3.5 Plant Control and Instrumentation 3-5 3.6 Fuel Storage and llandling 3-6 3.7 Electrical System 3-7 3.8 Radioactive Waste Control 3-7 4.0 PRINCIPAL ARCllITECTURAL AND ENGINEERING CRITERIA FOR {
DESIGN 4-1 ! 5.0 INTERACTIONS OF UNITS 1 AND 2 5-1 6.0 NEN FE ATURES ' 6-1 7.0
SUMMARY
OF OFF-SITE EFFECTS 7-1
8.0 CONCLUSION
8-1 LIST OF PLATES Plate 1 Station Perspective - Pinte 2 Index of Differences I Plate 3 Map of Arca Surrounding Station Plate 4 Station Property Plan Plate 5 Artists Conception of (Typical) Reactor and Containteni Plate 6 Reactor Vessel and Core Arrangement
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1.0 INTRODUCTION
1.1 Conoral Commonwealth Edison Company has made application to the Atomic Energy Commission for construction permits and facility licenses for Units 1 and 2 at its. Quad-Ci' ties Station. The Cencral Electric Company s has undertaken to furnish two complete nuclear power units, each to be licensed for operation initially at power icvels of approximately 2300 Mwt. To provide margins which will assure achievement of this objective, cach unit, . including all its various components, is desi;ned for operation at power Icycis of about 2600 Mwt. Eacli unit will utilize a singic cycle, forced circulation, boiling water reactor and other plant features which are substantially similar to Commonwealth's Dresdon Units 2 and 3 now under construction pursuant to CPPR-18 and CPPR-22 ( AEC Dockets 50-237 and 50-249) . Although the Quad-Citics Units 1 and 2 are identical in virtually all > respects to Dresden Units 2 and 3, e.g. design concepts, criteria, capacity and components, the Quad-Citics units incorporate several' changes which reflect improvements in' plant design, improvements in utilization of equipment, or which reficct plant changes arising from site considerations. All such changes or dif ferences are identified and analyzed in the Quad-Citics Plant Design Analysis as amended filed in AEC Dockets 50-254/265 Plate 2, attached hereto, provides an index to sections of the Quad-Cities report in which changes are described or analyzed. None of such changes involve any change in.the Principal Architectural and Engineering Criteria, set forth in Section 4.0 hereof, which governs the do.velopment of the design of Quad-Cities Units 1 and 2, and which are identical to those specified for Dresden 2 and 3
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p . 4 . q < 1-2' Subsequent to the authorization. of constructica of Dresden Unit 3 in CPPR-22 design efforts have procccded relative to the' provisions for cmcrgency core cooling. These provisions and evaluations of their performance as. applicable to Dresden Units 2 and 3.and to Quad-Citics
' Units 1 an'd 2 are.. discussed in.:considorablo detail in Amondment No. 4 to the-Quad-Citics Plant. Design Analysis :These provisions; together with.other incorporated-systems,.provides. assurance that the " General. .i Design Criteria for.Nuc1 car Power Plants" published by the' Atomic Energy Commission.on' November 22, 1965 will be satisfied and that the Quad-Citics Units 1 and 2 can and will be constructed and operated-without' jeopardizing public health and safety.
1.2 _ Applicability to Statement of Issues in Notice of !!carinn This exhibit is responsive principally to the first issue, and incidentally to the fourth issue, stated-in the. Notice of licaring dated December 16, 1966 With respect to subparagraph (a) of the first. issue, the'following sections 2.0, 3.0, 4.0 and 5.0 provido a descriptiol of the proposed cbsign of the facility, set forth the principal architectural and engineering criteria for design and identify the rajor features or components incorporated in the design for protection of the ) health and safety of the public. 1 I Section 6.0 identifics, in response to subparagraph (b) of the first j issue, those components and systems incorporated in the design respectin l' which further technical or design information is to be supplied. In relation to subparagraph (c) of the first issue Applicant does not consider that there arc any safety features or components which require research and development nor that there are any unresolved safet3l i questions associated with any features or conponents of the facility ! i
1 J 1 1-3 l which necessitate any rese. arch and development. Tha further tests and design evaluation which will be perforned to confirm analyses a,nd establish final detailed design of certain ccaponents fall within the scope of the " technical or design information" referred to in subparagraph (b) of the first issue which will be obtained and submitted I for review in the final safety analysis report. 1 h'hile this document in its entirety is responsive to subparagraph (d) of the f.irst issue (other than the clause respecting unresolved safety questions), the Applicant considers that sections 2.0 and 7.0 are , l 1 particularly relevant in this respect.
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2.0 SITE AND UNVIRONS
-~ The site for the Quad-Citics Station is cwned by the Commonwealth i
Edison Company nnd consists of a tract of land of approximately 488 acres. located on the cast bank of the Mississippi River in Rock Island County, Illinois, app.roximately 3 miles north of Cordova, Illinois, at the intersections of sections 7, 8, 17 and 18, 20 North, Range 2 East. l
- i The charactor and contours of the site and the immediate environs are shown in Plates 3 and 4 ,
The conter line of each unit is approximately 1800 feet fro.m the contor line of the Mississippi River to the west', and approximately one. mile from the property line to the east. The northernmost unit, Unit 2, is located approximately 1200 feet from the nearest property line north of the reactor building, and Unit 1 is located approximately
. 2000 foot from the property line to the south. (See Plate 4.)
There are no residences on the site. The residence nearest to the i reactor building is located 1/3 mile north. .\dj ace n t to the site to the north and cast at a distance of approximately 1-1/2 (8000 feet) miles is an industrial park. The remainder of the area around the site is used for agricultural purposes. Studies and review by the Applicant of the site characteristics supportI the conclusion that the Quad-Cities Station is a suitabic site for construction and operation of Units 1 and 2 (Refs. I-2.8; II)* PLANT DESIGN BASES DEPENDENT UPON PLANT SITE AND ENVIRONS CilARACTERISTI(l Information relating to the site and environs for the Quad-Cities Station has been sunnarized in Section II of the Plant Design Analysis report. It is intended to use this information as applicabic to the
*Unicss otherwise identified, references are to sections of the Quad-Citics Plant Design Analysis, j
design of tho station. The several design features which arc dependent.
.or affected by the site characteristics are sur arized below. .
'Off-Cas System
.The Quad-Citics Station will-be designed with a 310 foot stack to be used for the cont ruous disporsal of of f gnses to the atmoshhore.-
Based upon availabic meteorological data, station operations' characteristics and thc stack design assurance is provided that of f-site doses' arising from plant operation will be substantially below 10 CFR 20. (Ref. VII-3.2) Liquid Naste Effluent Liquid Wastes will be discharged to the river through the discharge l canal. The concentration of such wastes at the point of discharge from I the station will luf in compliance with'10 CFR 20. (Refs.,VII-3.1; Amend. 2, Sec. A-1; Amend 3, Sec. 2) ' Nind Loading Design Station structures will be designed to withstand the effects of 100 mph wind with gusts to 110 mph. Features necessary to shutdown either or both units under normal or abnormal conditions and maintain l both units in a safe shutdown condition will be designed to withstand winds of tornado velocity. (Refs. V-3.1; Amend. 2, Sec.11-1; Anend. 2, Sec. B-1; Amend. 3, Soc. 4) - Hydrology . j 1 No special design features are required to accommodate tho l hydrological characteristics of the site, since the station clovation is 8,5 feet above theoretical maximum water 1cvel due to floods. ' Liquid effluents from the site would not' mix with water supplios in the area due to local drainage patterns. (Ref. 11-4.0) Geology The geology of the area indicates that the underlying bedrock is capabic of supporting loads at 1 cast as high as that required for the I b
23 st ation structures. Consequently, no problems or restrictions beyond normal design practice are anticipated. (Ref. II-5.0) Scisnic Design The scismic design for structures and equipment important to safety will be based on dynamic analyses using acceleration or velocity responso spectrum curves which are based on a ground motion'of 0,12 g. As an additional requirement, the design will_be such that a safe shutdown can be made with the containment and heat removal-facilitics intact during a ground motion of 0.24 g. (Refs. II-6.0; Appendix F; V-3.1; /. mend. 2, Sec, C)
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, .: .D3501 EUil'fARYr DESCRIPTION ' Tile- FACI LI TY
- 3;1- Introduction Each of the Quad-Cities Units'is to be a General Elcet'ric boiling water reactor identical in most design features to Dresden Units 2 and.
3, 'tho . construction 'of which woro recontly authorized in the procoodins;; on<tEc applications of the ' commonwealth C Edison Company.in Atomic Encr'gy Commission' Dockets 50-237 and 50-249. Certain common facilitics are provided'for the two units which are summarized in Section 5 horcof Almost all of'the features of thc Quad-Citics Station Units have been demonstrated succes'sfully in the operation of one or nore of the General Electric boiling water reacters which have been constructed, including.the basic fuci design,'zircaloy fuc.1 cladding, hydraulically operated control rods, in-core neutron monitoring instrum ntation,
-pressure suppression containment and radioactive waste control. All new features (see Section 6.0 hereof) which have not yet. been utilized on an operating reactor will have been demonstrated in ope atbn of Dresden Units 2 and 3 and other similar reactors prior to completion of Quad-Citics Units 1 and 2, ,
An extensive technical description and evaluation of Quad-cities Static; Units I and 2 is given in the Plant Design Analysis and amendments of the application, Nos. I through 4 The following is intended to summarize the principal design features which are significant te safety considerations. In order to facilitate review, references to the Plant Design Analysis and amendments have been added where appropriate in case a more "in-depth" discussion is desired or considered to be in order. Table 1 herein provides in capsule form significant plant desig featurcs and some data appropriate to achieve a reactor thermal output of 2255 W. h'hile some of the paranctors listed in the tabic may be refined as design and procurement progresses, the final design will satisfy the principal architectural and engineering criteria for design
l he reactor and contain-j as sc't forth in Section 4.0 hereofo A drawing ci ' ment typical of that contemplated for the Quad-Citics Units is shown in lP_ late 5. 3.2 R'enctor Primary Systen Each reactor is a singic cycle, forced c'irculation, boiling water reactor producing steam for direct use in the steam turbines. Further, details are availabic in the Quad-Citics Plant Design Analysis, Section I-4.'1 (Ref. I-4.1). ] The fuel for cach reactor core co,nsists of slightly enriched uranium 1 ly dioxide pellets contained in sealed Zircaloy-2 tubes. These fuel rods \ are assembled into individual fuel assemblics of 49 fuel rods cach. } { Each comp,lete core loading consis,ts of 724 fuel assemblics. ,( Ref. III-1.C Control of each reactor is achieved by movable control rods and temporary control curtains. Reactor power level control is augmented by contro: ling the recirculation flow rate through the reactor core. Separate standby liquid control systems are provided for each unit as independent redundant!I j control mechanisms to be used in the remote event that the control red l system becomes inoperative. (Re f s . 111 - 1. 0, VI-4.0) . , { i I j E,ch a reactor pressure vessel contains the reacter core and supporting i I structure, the steam separators and dryers , the jet pumps, the contro: ! rod guide tubes, the feedwater, core spray and standby liquid control {; spargers, and other components. The nain connections to each reactor
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vessel include the steam lines, { reactor coolan:) recirculation lines, feedwater lines, control rod drive housing and other connections for core cooling. Plate 6 is a cut-away view typical of the reactor vessel and core arrangement. (Ref. Fig. 54*)
- Figures referred to herein may be found in Volune II of the Quad-Citics Plant Design Analysis.
L ' Containment-Systr -
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-Each-unit.caploy's' identical and independent primary containmen't i systems. .Each prinary containment system, consisting of a.dryucil, a pressure suppression chamber and interconnection; vont pipes, provide th'c principal ' containment barrier surrounding,each reactor vessel, the reactor. coolant recirculation loops, and other service loops connected '
to'the reactor vessel. linch primary con tainment system is f abricated as two large steci pressure vessels designed to accommodate the pressures and temperatures which would result fron or occur subsequent to a failure equivalent to a circumferential rupture of a recirculation line within the primary containment. During plant operation the primary containment ! contains an inert atmosphere. (Ref s . V-1.0 ; Anend. 2, Soc. D) f Each pressure suppression chamber is a steel torus-shaped pressurc
- l vessel approximately half-filled with water, and located below and encircling the'drywell. In the event of a reactor primary system pipe failure within the drywell the released steam would pass directly to the water where it would be condensed. This transfer of energy to the water pool would reduce rapidly (within 30 seconds) the residual pressure in the drywell and substantially reduce the potential for' subsequent leakage from the primary containment. (Ref. V-1.2.2)
Any leakage from.either primary containment system is to the secondary containment system which consists of the reactor building, the standby gas treatnent system, and the 310 foot stack. The latter features of the' secondary containment ~are conmon to both Units 1 and 2 The primary safeguards functions of the secondary containment' are to mininize ground
.lovel release of airborne radioactive materials and to provido for controlled, filtered, and elevated relense of the building atmosphere under accident conditions. (Ref s . V- 2.0 ; Amend. 2, Sec. D-6)
The safeguards features of both the prinary containment and the scactor building are canabic of beinn testinn periodically.
- 3- 4 3.4 Auxiliary and Standbv Cooling Systems In addition to the turbine generator and main condenser system, j 1
multiple independent auxiliary systems are provided for each unit for the purpose of cooling ea.ch reactor and primary containment system under various normal and abnormal conditions. (a) A residual heat removal system is provided which serves four
, functions:
(1) To remove reactor decay heat during shutdown. (Shutdown Cooling System) (Refs. X-3.0; Anend. 2, Sec. 6-3) (2) To inject water into the reactor vessel subsequent to a postulated loss of coolant accident sufficiently rapid as to reflood the core and prevent fuel clad aciting. (Low Pressure Coolant Inj ection System) (Refs. VI-7.0; Amend. 2, Sec. 6-5) (3) To remove heat from the water in the suppression chamber. (Ref X-4. 3.1.1(a)) (4) To spray water into the drywell as an au;mented means of removing energy fron the drywell if required. (Containment _ Spray System) (Refs, X-3,0; Amend, 3, Sec. 12.1; Amend. 4, Sec. IV) (b) A reactor core isolation cooling system is provided for removal of decay heat from the core when the reactor is isolated and other cooling systems are not availabic. (Refs. X-4.0; Amend. 2, Sec. F) (c) A high pressure coolant injection system is provided for removal of decay heat as backup to the reactor core ~ isolation cooling systen, and to provide coolant inventory control and heat dissipation from the core to the suppression chamber under postulate slow depressurization accidents. (Refs. VI-9.0; Amend. 3, Sec. 13-d~ Amend. 4, Sec. III)
55 (d) Two core spray systems are provided each of which as designed te pump water under accident conditions f rc- the pressure suppression chamber pool directly to the reactor core by a spray header or sparger mounted in the reactor vessel ab:ve the core. (Refs. V1-6.0; Anond. 2, Sec. 6-5; Amend. 3, Sec. 13-b; Amend. 4, Sec.,II) (c) A standby coolant supply system is provided by a cross-tic between the service water system and the feedwater system which makes available an inexhaustibic supply of coo:ing water to the reacter core and containment independent of all ether cooling wate,r sources. (Refs. X.-10.0; Amend. 4, Sec. VI) The core cooling provisions itemized above are designed to prevent fuel clad molting for the full range of primary system pipe sizes which may L be postulated to fail. [ 3.5 ' Plant Control and Instrumentation Reactor power for cach unit is controlled by movement of control rods and by regulation of the recirculation f1:w rate. Control rods are used to bring the reactor through the full range of power and to shape the core power distribution. Load folleving and adjustments in reactor power level are accomplished with recirculation flow control. (Refs. 111-5.0; IX-1.0; IX-3.0) ' Rapid shutdown of the reactor is accomplished by insertion of the i hydraulically actuated control rods. (Refs. III-5.2.6; IX-3.2) l The performance of the reactor core and the indication of reactor I power Icyc1 are continuously monitored by the neutron nonitoring system, the sensors of which are located in the reactor core. The system efficiently and accurately provides power Icyc1 monitoring from source range to full power on a gross and local basis. (Ref. IX-4.0)
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i t A bypass system having n capacity of approximately 40%.~of steam flow 3 at'ratedEload is supplied with the turbine to restrict overpressure' l transients resulting from sudden turbine; control valveoor~stop valve closure. Partial load rejection can be accommodated'by the rapid action of the bypass system. (Ref. IX-9.0)
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The reactor protection system overrides the'above control systems to initiate any required safety action. The reactor protection system automatically. initiates-appropriate action whenever the plant conditions conitored by the system approach pre-established limits. The-reactor protection system acts to shutdown the reactor, close isolation valves,
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or initiate the operation of other safeguards systems as required. Components of.the reactor protection system can be removed from service for testing and maintenance w' hout interrupting plant operations and without; negating the ability o; the protection system to perform its protective functions upon receipt of the appropriate signals. (Ref. I X- 7.0) Instrumentation is provided for continuous mo.nitoring of the radio,- activity of specific process systems and to provide alarms or signals for appropriate corrective actions. In addition, radiation surveillance of the site and environs will be maintained. (Ref. I X-6.0) Controls,, instrumentation and alarms necessary for safe operation of cach unit are located in the main control room in the turbine building. i-l 3.6 Fuel Storage and llandling New fuel is stored in a dry vault in the reactor building. Refueling is conducted under water, and the irradiated fuel is stored under water,in the reactor building until shipment from the site. (Refs X-1.0: Amend. 3. Sec. 18)
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'3.7 Electrical System The electrical output of each unit will feed into the 345 KV network through four 345'KV circuits. The primary feature of the cIcctric power' system serving each Quad-Cities unit is th'e diversity of dependable power sources, physically isolated so that any.one ,
instrument of failure affecting one.sourch of supply will not communicate l .- ~to alternate sources, thus assuring a continuous s'ource of auxiliary power to the unit. Auxiliary power for each unit can be supplied fron phree separate and independent sources: the alternate unit, the 345 KV I transmission' system, and the standby diesel generator system. Three' standby diesel generators are provided, any two of which can serve the emergency requirements of one unit and the normal shutdown, requirements-f. of the other. (Re f s .VIII-1.0, 2.0, 3.0) [ 4 station battery for each unit is used for all controls vital to p' ant safety and to power certain functions required for plant shutdown, such as closing of isolation ralves, opening valves to the isolation cooling system, providing lighting, .and required instrumentation. Exampics of such required instrumentation are the control rod position indicators and a neutron channel to monitor the core during shutdown. (Ref. VIII-4.0) . 3.8 Radioactive Waste Control Gascous, liquid and solid waste control facilities are provided to limit the release of radioactive materials from the site with applicabic regulations. (Ref. VII) mW
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TA3LN 1 PRINCIPAL DliSIGN FEATURES OF QUAD-CITIES UNITS 1 AND 2 (Data for Each Unit) Station Not Bicctrical Output , 715 MW per ubit Reactor - ! a Lt , Thornal Output 2255 MW 7 T Core Opnating Pressure 1000 ps Total Core Flow Rato 98x10gg Jb/hr i Steam Flow Rat'e 8.62 x 109 lb/hr Core j Circumscribed Core Diameter 189.7 inches Active Length _ 12 feet Fuel Assembly ,
' Number of ]ucl Assemblics 724 Fuci Rod Agray 7x7 Cladding Material Zircaloy - 2 Fuct Material UO Active Fuel Length 143 inches Cladding Outside Diameter 0.570 inch Cladding Thickness- 0.036 inch Fucl Channci Material Zircaloy - 4 Control System ** Number of Movabic Control Rods 177 Shape of '!cvabic Control Rods Cruciform Pitch of Movable Control Rods 12-0 inches Control Material in Movabic Control Rods Boron Carbide Type of Control Drives Bottom entry, hydraulic actuated Material in Tenporary Control Curtains Boron - S.tainicss steci' Number of Temporary Control Curtains 324 Control 6f Reactor Power Output Movement of control rods and variation of coolant flow rate Core Design Data and Operating Conditions Power Density 36.7 kw/ liter licat Transfer Surface Area 63,527 sq. ft.
Average llcat Flux 116,300 Stu/hr-ft 2 Maximum ifcat Flux 349,000 Btu /hr-ft 2 Mininun Critical llent Flux Ratio at Overpower-Equal to or Greater Than 1.5 Core Average Voids of Coolant within Assembly 37%
, Core Average Exit Quality of Coolant within Assemblies 0.9%
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Design Power Peaking Factors Total Penking Factor 3.0 Additional Allowance for Overpower 1.2 Nue: car Design Data Initial Average Fuci Enrichment 2.0% Nater/UO 2 V lume Ratio - 2.38 ' Ex: css Reactivity of Cican Core (Uncontro11cd) at 65'F 0.22 K Total 'intth of Control -3.26 K Reactivity of Core with All Control Rods In 9.96 K North of Sundby Liquid Control System -0.14 K Reactor Vessel Inside Diameter 20 ft. - 11 in. Overall Length 68 ft. 5/8 in. Design Pressure 1250 psig Coo
- ant Recirculation Loops Location of Recirculation Loops Containment Drywell Xurber of Recirculation. Loops 2 Pipe Size 28 inches Surbor of Jet' Pumps 20 Location of Jet Pumps Inside Reactor Vessel Prirary Containment Type ,
Pressure Suppression Design Pressure of Drywell Vessel 62 psig Design Pressure of Suppression Ch amber Vessel 62 psig Maximum Leakage Rate Less than 0.5% Free Volume por Day Secondary Containment (Common for Units 1 6 2) Type Reinforced concrete and Steel superstructure with netal siding Internal Design Pressure 0.25 psig Maximum Inicakage Rate 130% free volume per day at ,
, 0.25 in water negative pressure Structural Design Scismic Resistance 0.12g Sustained Wind Loading 110 mph Control Room Shielding Dose not to exceed 500 r. rem in 8 hours under accident conditions assuming full core:
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4 . Unit Electrical Systems-Number of' Incoming Power Sources 4 - 345 XV Separate Power Sources Provided 1 Auxiliary transformer 3 Standby diesel generators l shared between Units 1 6 1 Startup transformer 1 Station Battery Reactor Instrumentation System Location of Neutron Monitor System In-core Ranges of Nuclear. Instrumentation Startup Range Source to 0.01% rated power Intermediate Range - 0.0001% to 10% rated power Power Range 1% to 125% rated power Reactor Protection System. Nunber of Channels in Reactor Protection System 2 Number of Channels Required to Scram of Effect Other Protective Functions 2 Nuhber of Sensors por Monitored Variabic in Each Channel 2 Method to Prevent Unwarranted Withdrawal of Control Rods Automatic interlocks l Other Engineered Safeguards - Summary of Systems and Functions 2 Coro Spray Systems To cool the core under assumed loss os coolant accident; capability to
/ reflood the core following
! coolant loss.
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Low Pressure Coolant Injection Systom To provide a means of cool ' ing the core by re' flooding the core subsequent to a postulated loss of coolant accident. Ilich Pressure Coolant Injection System To provide backup cooling i capacity to the reactor corl isolation cooling system an to provide high pressure coolant makeup for postulat. slow depressurization accidents. Reactor Core Isolation Cooling System To avoid overheating of the reactor fuel in the event that reactor feedwater capability is lost and othe normal heat removal systems which require a-c electrica, power for operation are not! available. l l l
Other Ennincerod Saforut as - Sunmarv of Systems a: Functions (Cont'd.) Standby Coolant Supply System To provide an inexhaustib1. supply of water to the cc: and containment independo: of all other core coolin; methods. 2 Containment Spray Cooling Systems To augment removal of encr;
. trem primary centninment
. subsequent to assumed los:
of coolant accident. Rod Velocity Limiter To limit the free fall of r control rod from the core to approximately five fec: per second. Control Rod D' rive Housing Support To prevent a control rod drive mechanism from fall-
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ing away from the rcactor pressure vessel in the unlikely event of a failu: of a drive housing. Main Steam Line Flow Restrictors A constriction in each mair steam line to reduce rate of blow-down in event of postulated severama of the nain steam line. Isolation Valve's To effect reactor contain-ment automatically when j required under postulated f accident conditions. Primary Containment Atmosphere Control System To provide inert atmosphere in the primary containmer.t system to preclude a hydrogen-oxygen. reaction subsequent to a postulated coolant loss accident. Standby Gas Treatment System' To provide a means for re-(Common to Units 1 and 2) moval of particulate and halogens from the reactor building air under postula ted accident conditions prior to discharge of the filtered air through the ventilation stack. Also provides a means for main-taining the reactor build-ing at a negative pressure so that leakage is into th reactor building and thus prevents ground Icyc1 re-lease of building air unde postulated accident con-ditions..
l . Other Engineered Safeguards - Sunmary of Systems a,d Functions (Cont'd.) Standby Liquid Control System . To provide a redundant, . l independent back-up contr: nechanism in the event th: the control system become inoperable. 9 4 e l l
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40 . PRINCIP AL ARCllITECTURAL ' AND ENGINEERING CRITERI A FOR DESIGN The principal architectural and engineering criteria for design for th'e station arc summarized.below. ,
- 4 .' l - Station Design Principal structures and equipment which may serve either to preven accidents or'to mitigate their consequences.will!be designed, fabricated and erected in accordance with applicable codes and to withstand the most severe earthquakes, flooding conditions, windstorms, ice conditions temperature and other. deleterious ' natural phenomena which could occur at the site during the lifetime of this Station.
d.2 Containment , , (a) The primary containment , including the drywell, pressure- l suppression chamber, associated access openings and penetrations, will l be designed, fabricated and erected to accommodate, without failure, , the pressures'and temperatures resulting from or subsequent to the double-ended rupture or equiv51cnt failure of any coolant pipe within the drywell.
. l (b) Provisions will be made both for the removal of heat from .
within the primary containment and for such other measures as may be necessary to maintain the integrity of the containment system as long as necessary following a loss of coolant accident. ! 1 (c) The reactor building, encompassing the primary containment system, will provide primary containment when the pressure suppression i' system is open and secondary containment when the primary containment is closed. ] 1 (d) Provision will be made for prooperational pressuro and leak rate testing of the primary containment system and'for leak testing at
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. 4-2 periodic intervals af ter thc .f acility has conncnced operation. Provision- l will also be nade for Icak testing selected penetrations. Provision will also be made for demonstrating the functiona' integrity.of reactor building containment.
(e) The integrity of the complete containment system and such other engineered safeguards as may be necessary will be designed and maintained so that off-site dosos resulting from postulated accidents will be below the values stated in 10 CFR 100. 4.3 Reactor System (a) A direct-cycic boiling water reactor will be employed to pr'oduce steam at approximately 1000 psig for use in a steam-driven turbine-generato The reference design thermal output of the reactcr is approximately 2255 SM'
(b) The reactor will 'bc fueled with slightly enriched uranium ~
dioxide contained in zircaloy clad fuct rods. (c) At the design overpower condition the mir.imum critical heat flux ratio will be above, and maximum fuel center temperature evaluated at the design overpower condition will be below, that which could lead to fuel rod failures. (d) Fuel rod cladding thickness will be designed to maintain cladding integrity throughout the anticipated fuel life. :ission gas release within the rods and other factors affecting design life will be considered for the maximum expected exposures. (e) The reactor and plant will be designed so that there will be no inherent tendency for undamped oscillations. (f) The reactor will be designed to accommodate tripping of the turbine-generator, loss of power to the reactor recirculation system and other station transients' and marauvers which night be expected without compromising safety and without fuc; damage.
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- 4-3 i
(g) The reactor will include separate systces including over-pressure scram, the reactor core isolation cooling system, safety valves, I and turbine bypass, to prevent serious primary reactor system overpressure (h) Power excursions which could result. fro: any credible reactivity ) J addition accident will not cause damage, either by motion or rupture, { i to the pressure vessel or impair operation of required'saf'egdards. J (i) ilcat removal systems will be provided which are capable of safely accommodating core decay heat under all credible circumstances, including isolation from the main condenser and loss of coolant from the reactor. Each system so provided will have appropriate back'-up'-(.re-dundant) features. Each system can be periodically tested. (j) Reactivity shutdown capability shall be provided to make and hold the core adequately subcritical by control rod action, from any point in the operating cycle and at any temperature down to room temperature, assuming that any one control rod is fully withdrawn and unavailable for use. (k) Rodundant backup reactivity shutdown capability will be provided ind: pendent of normal reactivity control provisiens. This system will have the capability, with adequate margin, to shut down the reactor from any operating condition., 4.4. Control and Instrumentation (a) The station will be provided with a centralized control room having adequate shielding to permit occupancy during all design accident situations. (b) There will be sufficient interlocks or other projections so-that procedural controls are not the only means of preventing serious accidents.
.e
4-4 l , (c) A reliable reactor protection system will be providod' to automatically initiate. appropriate action whenever plant conditions approach pre-established limits. Periodid testing capability will,be provided. Sufficient redundancy will be'provided so that failure or removal from service of.any one component or portion of the system will not preclude sc' ram or actuation of other protective devices when required 4.5 Electrical Power Sufficient normal and emergency auxiliary sources of electrical power will be provid.ed to assure a capability for prompt shutdown and continued maintenance of the station in a safe condition under all credibic circumstances. t 4.6 Radioactive Waste Disposal Cl
~
(a) Gascous, liquid and solid waste disposal facilitics will be designed so that discharge of effluents and off-site shipments shall be in accordance with 10 CFR 20. (b) Process and discharge streams will be appropriately monitored i and'such automatic features incorporated as may be necessary to naintain releases below the permissibic limits of 10 CFR 20. 4.7 Shiciding and Access Control The radiation shielding in the station and the station access contro: patterns will be such that the doses shall be less than those specified in 10 CFR 20, 4.8 Fuel Handling and Storarc Appropriate fuel handling and storage facilities will be provided to preclude accidental criticality and to provide cooling for spent fuel. Note:. The material in this section is .dentical in content to Section I-3,0, Plant Design Analysis. 1
INTERACTIONS OF UNIT. I and 2 5.0 As indicated in Section 3.0, certain portions Of the station are m b utilized in common or are shared wholly or, par:ially by the two units. The criteria followed for the design of this two unit station with regard to the sharing of components, systens, and l'acilities is that each unit will operate independently of the.other. A malfunction of l equipment or an operator error in either of the units will not affect the continued operation of the other unit. Likewise, if by the sharing of a component, system, or facility, the safety of the plant is increase-then such sharing is accomplished. (Refs. I-8.0 ; Amend. 3, Sec. 1)
. . . _ . . . . ~ . ~. . .
There are a number of systems and facilitics which are shared by both units but which are not safety related. Theseincludesuc$facilitics as Administration Offices, Access Control, Machine Shop, Laundry, l Sewage Treatncnt Plant, and Pla.nt Security Fen:ing. Certain service facilitics are also shared including the Fire Protection System, l Circulating Water System, Service Natcr System, Make-up Domineralizer System, Service Air System and llcating System. Failure of such systems would not jeopardize safety of the units. (Ref. I-8.4.10) Following is a summary of other shared systems and fac'ilitics to which special attention has been directed to assure adequate IcVels of safety during normal operation of the units and pos'tu;ated accident conditions. (a) Site and Off-Site Environs 'lonitoring A study of environmental radiation levcis wi'11 be initiated approximately two years before the scheduled operation of the Quad-Citics Station and continued after' operation. The purpose of the program is to provide adequate monitoring for the integrated Quad-Cities Station comple and to assure that the requirencnts of 10 CFR 00 are not for the combined operation of both units. (Ref. II-7.0) ~ 1 1 l
L". -
. 5-2 (b) New Fuel. Storage .
l The storage vault for Unit 1 and 2 has been specifically designed for use as-a common new fuel storage facility. Only twice in the life of the combined plant will the vault be filled with new fuel. Since there is a year's difference in start-up time of the Units 1,and 2, there is no problem in storage. Based on the experience of Dresden Unit 1, full spare. reactor cores are not kept on hand between,refuelings. (Re f . , X-1.1) (c) Spent fuel handling and storage pool The use of the interconnecting transfer canal between Unit 1 and 2 provides for the safe underwater movement of radioactive material and fue.l. Not only does the use of the canal improve the refueling h cycle for each reactor, but it also facilitates any necessary maintenance of each pool. (Re f s . 1- 8. 4. 7 ; X-1. 2) . (d) Reactor Building Closed Cooling Nater System This system serves as an interface between certain other systems which are exposed to the reactor and the river water, and prevents potential leakage of radioactive materials directly to the river.
- Sharing of those systems provides greater flexibility of plant operations which is assessed as contributing to overall plant sa fety (Re fs . I-S .4.4 ;VI-5.!
(c) Auxiliary Power System The two units are only tied togetherelectrically through the Edison transmission system and by electrical inter-ties to the Iowa-Illinois Gas and Electric Company system. The source of at least fifty per cent of the auxiliary power requirements for each unit is through the 345 KV transmission system. The design of the 345 KV switchyard, is such that electrical breakers will sectionalize the switchyard to prevent elect rical problens f ron propagating fron one unit to the other.
5-3 (f) ' Standby Diosc1 Generator System , The standby dicsc1 generator system includes three diesel generators, any two of which are capable of furnishing power requirements for operation of emergency equipment on onc unit and normal shutdown power for the other. (VIII-3.0) , (g) Station Battery - Each unit has a separato stntion battery. However, a 100 per cent
- . I capacity spare charscr is provided which can be connected to the d-c system of either Unit 1 or Unit 2. Electrical breakers are interlocked to prev.cnt connecting the d-c systems of both units together. (VIII-4'.0)
(h) Liquid and Solid Radioactive Waste Control The radwaste facility for U51ts 1 and 2 is designed for use as a common radioactive chemical treatment plant. The purpose,'fo the plant is to process liquid radioactive wastes for reuse, discharge to the river, or concentration for storage and off-site shipment. The source of the liquid wastes is unimportant to accomplishing the design objectives. The facility is designed to maintain the quantities of radioactive waste released to the environment with the limits of . 10 CFR 20. (I- 8. 2 ; VIII-3.1,3. 3) (i) Gaseous Radioactive Waste The gaseous waste effluents from Units 1 and 2 will use a common 310-foot stack. The off-gas system of each unit, which constitutes the principal source of the radioactive gaseous effluent, is continuously monitored and recorded in the control room and can be shut off if . required. The total stack effluent discharged from Units 1 and 2 is
'also monitored and recorded in each control room. (Ref s . I-8.1; VIII-3.2)
(j) Inerting Gas Sypply System - The inerting gas supply system for Units I and 2 are shared, although each unit has its own gas makeup system. (VI-8.0; Figure SS)
.. . _ ____ y ___ _ __ -
J+ *
~
.,.. , l5-4 (k): Common Turbino Building The use of a common turbine building with multi-unit pla.nts.
improves the operability and maintenance 6f each individual unit c o n'-
~
.tained within the building. The turb'ine building ventilation air supply-and exhaust systems are ' operated as a combined systen. (.Refs.I-8.4.5;; Fig.32' ,
(1) . Reactor. Containment - Units 1 and 2 have separate and independent primary containments. and pressure suppression systems. The secondary containment-for each unit below the operating floor IcVel is constructed to serve its own unit. The' operating floor will be open and common to both units. Both units sh'are the same standby gas treatment, ventilation and heating-systems, each ' system having capac,iti As to accommodate the -combined . 7 s'ocondary. cont ainment volume. . .. ....... (Re fs . I-8. 4. 6 ; V-2.0. ;_Fig. ... ,- 28; Amend. 3,Sec.9: (m) Control Rooms Evaluations have been made of the inter-plant effects of postulated accidents. It has been asc,ertained that an accident in'either
. of the units, up to and including maximum postulated accidents, will_not ,,
prevent access to the control room of either of the units or prevent
~
safe operation of shut'down of the other unit. (Ref. I-8,5) a e 4 4 i _ - _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ -.___1_ - - - - - - - - - - -
'6.0 NEW FE ATURES
- The' successful operation of a number of boiling water reactors has L
demonstrated that existing technology respecting such reactors is sufficient to give reasonabic assurance that reactors designed and constructed as" proposed for Quad-Cities Units 1 and 2 can be safely l operated. The Quad-Citics Units 1 and 2 design incorporates sovera'l 1 new features itemized and referenced below, cach of which contributes to overall plant safety. Each of these features is also incorporated in the design of Dresden Units 2 and 3, now under. construction and has been reviewed by the staff of the AEC and the Advisory Committee for i Reactor Safeguards. Technical information pertaining to these design features, as appropriate, will be provided to the nEC Staff. Since Dresden Units 2 and 3 are scheduled to commence regular operation prior , to completion of Quad-Cities Units 1 and 2, the operability of each of the new features will have been demonstrated in full scale reactor plant operation prior to operation of the Quag-Cities Units. No11owing is a listing of the new features which are described in the 3 __ j reference section of the Plant Design Analysis and amen'dments thereto, j I Reference Section in j Item Plant Design Analysis j Control Rod Worth Minimizer IX-8.0 , Control Rod Velocity Limiter VI-2.0 Control Rod ilousing Support VI-3.0 Main Stcan Line Flow Restrictors V-5.0 Core Spray Systems VI-6.0 and Amendments No. 3 and 4 Jet Pumps IV-2.3.2, App. D, and Anendment No.l Containment Spray Cooling Systems V-1.2.7 and Amendments No. 3 and 4 Containment Atmosphere Control
- System VI-8.0 In-Core Neutron Monitoring System X-4.0 liigh Prossuro Coolant Injoction System VI-9.0 and Amendments No. 3 and 4 Low Pressure Coolant Injection System VI-7,0 and Amendaents No. 3 an'd 4 Standby Coolant Supply System VI-10.0 and Amendment No. 4 See also Section I-5.0, Plant Design Analysis i
- u. .
' ~
, ,' 7-i
-l
-7.0. SU,\lMARY '0F - OFF-SITE ~ E FFECTS Analyscs have been made .to evaluate the of f-site effects from both I
normal plant operation and' postulated accident conditions. l During normal full-power operation of the plant it is-anticipated that.
.the maximum-annual average exposure to persons off-site'will'not exceed approximately 5 mrems. This may be compared.with the limitations of
~10 CFR 20 of 500 mrens por year. 1 The general classos of postulated reactivity,. transients and . loss of coolant-accidents have been analyzed under' conditions when the primary
~
containment would proyide an offective barrier'to the escape of fission products (i.e.,. rod drop and recirculation line rupture or-loss of coolant accidents) and when the primary containment would'not'be e f fe ctive (i_. e_. , fuel loading accident and' steam line break outside ,
.the primary containment). '
The following summarizes the maximum calculated off-site doses resulting l from such accidents involving either one of the two units. Tital Accident Exposure.- Rem l
' ~
Accident . Whole Body Thyroid Rod Drop (Ref. XIV-3.1) 8.3 x 10-2 1.9 x 10'3 - Fuc1 Loading (Refs. XIV-3.2; Amend. 2, ' l Sec 11-3 ; Amend. 3, Sec.18) 5.5 x.10-3 ' 3.2 x 10-3 l Steam Lino Break (Ref. XIV-3.3) 5.6 x 10-6 3.6 x 10-4 ! Loss of Coolant (Refs. XIV-3.4; Amend 4, 1 Secs. II-4, III-4, IV-4) 8.6 x 10-4 6.0 x 10-4 i It has been concluded from these analyses that the off-site exposure levels are only a small fraction of the maximum guideline values of 10'CFR 100, and thus the plant provides substantial. and adequate protecti against hazards,to the public. i e u
1, k- .. ; 8-1 i .
8.0 CONCLUSION
On'th'c basis.of-the foregoing the App'licant' respectfully submits 1that: 1 . , . j (a) . A description of the proposed design of the Quad-Cities Units c 1and'2,includingtheprincipalarchitecturalandengineeringlcriheria ~ therefor, has'been furnished; (b) Applicant has identified the major new features incorporated ' in the design. of the Quad-Cities Units 1 and 2 which have not been ,
- demonstrated in actua.1 operation .of other nucl.ca.r power. plants to.date;
~
. and- the technical or design information which will be supplied in the final safety analysis report. ,
(c') There are. no unresolved sa'fety. questions other than those related to the demonstration of the effectiveness an'd-reliability of such new features by actual operation; f (d) The- effectiven,ess and reliability of such new features 'will be demonstrated by the operation of other similar boiling. water reactor plants now under construction and by pre-operational.and initial startup _ tests of the Units 1 and' 2 prior to the last date stated in the application for completion of construction of each unit; and , (c) Taking into consideration the characteristics of the site and environs and the proposed design of the Quad-Citics Units 1 and 2 such - 1 facility can be constructed and operated within the limita,tions established by 10 CFR 20, within the site' criteria set forth in 10 CFR 1: and without undue risk to the health and safety of the public. 9 4 4 9
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y V . . PLATE 2 s INDEX TO SECTIONS OF Tllu QUAD-CITIES TLANT DESIGN ANALYSIS REPORT IN WilICll DESIGN AND PLANT CilARACTERISTICS DIFFER FRO \1 Tii0SE REVIEh'ED IN AEC DC:KET 50-249 i Section Reference in Quad-Cities Unit Item 1 and 2 Plant :csign Analysis Report Site Characteristics Section II Electrical Transmission Inter-ties Section VIII-1.0 Number of Core Spray Pumps Section VI-6.0 and Figure 72 Sharing of Auxiliary Cooling Systems
- Shutdown Cooling System Section X-3,0 and Amendment 4
- Low Pressure Coolant Injection System Section VI-7.0 and Amendment 4 l
- Containment Spray System Section V-1.2.7 and Amendment 4 Reactor Core Isolation Cooling System Section X-4.0 cad Amendments 2, 3, and 4 Interconnection Between Fuci .
Storage Pools Section X-1.2 i Use of Powdex Unit in Reactbr Cleanup Domineralizer System Section X-2.0 , Use of Powdex Units in Condensate Demineralizers Section XI-2.8 Effect of Powdox Units on , Radioactive Waste System Section VII-4.1 , 1
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