Text

Chapter 9 of Oconee 1,2 & 3 PSAR, Auxiliary & Emergency Systems. Includes Revisions 1-6
	ML19322B513
Person / Time
Site:	Oconee
Issue date:	12/01/1966
From:	; DUKE POWER CO.
To:	;
References
	NUDOCS 7912040271
	Download: ML19322B513 (59)
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TABLE OF COICEUTS 1

Section Page 9

AUXILIARY AND DERGENCY SYSTEMS 9-1 91 HIGH PRESSURE INJECTION AND PURIFICATION SYSTEM 9-2 9 1.1 DESIGN BASES 9-2 9 1.1.1 General System Function 9-2 l

9 1.1.2 Letdown Cooler 9-2 9 1.1 3 Letdown Control valve 9-2 9 1.1.4 Purification Demineralizer 9-2 i

9 1.1 5 High Pressure Injection Pumps 9-3 9 1.1.6 Reactor Coolant Pump Seal Return Coolers 9-3 9 1.1 7 Letdown Storage Tank 9-3 9 1.2 SYSTEM DESCRIPTION AND EVALUATION 9-3 9 1.2.1 Sche =atic Diagram 9-3 j

9 1.2.2 Perfomance Requirements 9-3 9 1.2 3 Mode of Operation 9-3 9 1.2.4 Reliability Considerations 9-5 j

9425 Codes and Standards 95 i

9 1.2.6 System Isolation 9-5 9 1.2 7 Leakage Considerations 9-5 9 1.2.8 Operation Limits 9-6 92 CHEMICAL ADDITION AND SAMPLING SYSTEM 9-9 9 2.1 PESIGN BASES 9-9 i

l 9 2.1.1 General System Function 9-9 i

9 2.1.2 Boric Acid Mix Tank 9-9 4

4 p

9 2.1 3 Boric Acid Pumps 9-9 g

4 9 2.1.4 Caustic Mix Tank 99 l

188 9-1 l

COITfEIITS (Cont'd)

Section M

9 2.1 5 Caustic Pu=p 9-9 9 2.1.6 Potassium Hydroxide Tank 9-9 9 2.2 SYSTDI DESCRIPIION AIID EVALUATION 9-10 9 2.2.1 Sche =atie Diagram and System Description 9-10 9 2.2.2 Perfor=ance Requirements 9-11 9 2.2 3 Mode or Operation 9-11 9 2.2.4 Reliability Considerations 9-12 9 2.2 5 Codes and Standards 9-12 9 2.2.6 System Isolation 9-12 9 2.2 7 Leakage Considerations 9-12 9 2.2.8 Failure Considerations 9-12 9 2.2 9 Operating L1=its 9-13 93 HI" rCZ 2*J~I.:Z:0 JJ:" CcMPOIarr COOLING SYSTEM 9-17 931 DESIGN BASES 9-17 932 SYSTEM DESCRIPTION AIID EVALUATION 9-17 9 3 2.1 Schematic Diagra=

9-17 9 3 2.2 Perfomance Requirements 9-18 e

9323 Mode or Operation 18 9 3 2.4 Reliability Considerations 9-19 9325 Codes and Standards 9-19 532.6 system Isolation 9-19 9327 Leaksge Considerations 9-19 9 3 2.8 Failure Considerations 9-20 O

94 SPE?iT FUEL COOLING SYSTEM 9-21

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9.4.1 DESIGN BASES 9-21 gg 9-11

O COUTEUTS (Cont'd)

Section P3 9 4.2 SYSTEM DESCRIPTION AND EVALL%2 ION 9-21 9 4.2.1 Sche =atic Dia.3 ram 9-21 9 4.2.2 Performance Requirements 9-22 9 4.2 3 Mode of operation 9-22 9 4.2.4 Reliability Considerations 9-22 9 4.2 5 Codes and Standards 9-22 9.4.2.6 Leakage Considerations 9-22 9 4.2 7 Failure Considerations 9-23 9 4.2.8 Operation Limits 9-23 95 LOW PRESSURE IUJECTION AND DECAY HEAT REMOVAL SYSTEM 9-25 g

951 DESIGN BASES 9-25 9 5 1.1 General System Function 9-25 I

9 5 1.2 Lcv Pressure Injection and Decay Heat l

Removal Pumps 9-25 9513 Lov Pressure In.jection and Decay Heat Removal Cooler 9-25 952 SYSTE4 DESCRIPTION AND EVALUAIION 9-25 9 5 2.1 Sche =atic Dia,3 ram 9-25 9 5 2.2 Performance Requirements 9-25 9523 Mode of operation 9-25 9 5 2.4 Reliability Considerations 9-26 9525 Codes and Standards 9-26 9 5 2.6 System Isolation 9-26 9527 Leakage Considerations 9-26 l

9 5 2.8 Failure Considerations 9-26 lN 9-111 l

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O COITIEI;TS (cont'd)

Section M

96 cc0 LING WATER SYSTEMS 9-29 9 6.1 DESIGN BASES 9-29 9.6.2 SYSTEM DESCRIPTION AND EVALUATION 9-29 9 6.2.1 condenser circulating Water System 9-29 9.6.2.2 High Pressure Service Water System 9-30 9.6.2 3 Low Pressure Service Water System 9-30 9 6.2.4 Recirculated cooling Water System 9-30 97 FUEL HANDLING SYSTEM 9-32 971 DESIGN BASES 9-32 9 7 1.1 General System Function 9-32 9 7 1.2 New Fuel Storage Area 9-32 9713 Spent Fuel Storage Pool 9-32 9 7 1.4 Fuel Transfer Tubes 9-32 9715 Fuel Transfer Canal 9-33 9 7 1.6 Miscellaneous Fuel Fnnaling Equipment 9-33 972 SYSTE4 DESCRIPTION AND EVALUATION 9-33 9 7 2.1 Receiving and Storing Fuel 9-33 9 7 2.2 Icading and Removing Fuel 9-34 9723 Safety Provisions 9-36 972.4 Operationci Limits 9-37 9.8 STATION "EITTILATION SYSTDIS 9-38 9 8.1 DESIGN BASES 9-38 9 8.2 SYSTDI DESCRIPTION AND EVALUATION 9-38

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i LIST OF TABLES i

Table No.

Title Page 9-1 High Pressure Injection and Purification System 9-7 9-2 Component Data 9-8 9-3 Steam Generator Feedvater quality 9-14 9-4 Reactor Coolant quality 9-14 9-5 Chemical Addition and Sampling System Equipment Data 9-15 9-6 Reactor Building and Component Cooling System Design Parameters for Ucrmal Operation on a Per Unit Basis 9-20 9-7 Reactor Building and Component Cooling System Component Data for Normally Operating Equipment 9-20 9-8 Spent Fuel Cooling System and Component Data 9-24 9-9 Low Pressure Injection and Decay Heat Removal System 9-27 9-10 Low Pressure Injection and Decay Heat Removal System Co=ponent Design Data 9-28 9-11 Cooling Water Systems Component Design Para =eters 9-31 v

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LIST OF FIGURES (At Rear of Section)

Figure No.

Title 9-1 Flow Diagram Identifications t

9-2 High Pressure Injection and Purification System 9-3 Chemical Addition and Sampling System 9-4 normal operating Fode for Reactor Building and component cooling System 9-5 Spent Fuel Cooling System 9-6 low Pressure Injection and Decay Heat Removal System 9-7 Decay Heat Generation Versus Time After Shutdown 9-8 condenser circulating Water System 9-9 High Pressure and Low Pressure Service Water Systems 9-10 Recirculated Cooling Water System s

9-11 Fuel Handling System 9-12 Turbine and Auxiliary Building Ventilation Systems 9-13 Administration Building Ventilation System O'~'

194 9-vi 4

9 AUXILIAM AND DETMCY SYSTHE The auxiliary systems required to support the reactor coolant syste=s during nor=al operation of the Oconee Nuclear Station are described in the following sections. They include the following:

a.

High pressure injection and purification system, b.

Chemical addition and sampling system.

c.

':::'" hillin; ;;i component cooling system.

d.

Spent fuel cooling syste:::.

e.

Iow pressure injection and decay heat removal system.

f.

Cooling water systems.

g.

Fuel handling system.

h.

Station ventilation systems.

The design of the auxiliary and emergency systems has given consideration to the dual-unit concept, i.e., a two-reactor station. This section describes the equipment for each unit and states where equipment is shared by both units.

Shared equipment vill be installed with Unit 1.

Detailed descriptions of some of these systems have been presented in Section 6 since they serve as engineered safeguards. The information in this section deals primarily with the functions served during normal operation.

The majority of the components within these systems is located within the auxiliary building. Those systems with connectin; piping between the reactor buildings and the auxiliary building are equippei. tith reactor building isola-tion valves as described in 5 2.

The following codes and standards are used as applicable in the design, fabri-cation, and testing of components, valves, and piping:

ASME Boiler and Pressure Vessel Code,Section II, Material a.

Specifications, b.

ASME Boiler and Pressure 7eesel Code,Section III, miclear Vessels.

ASME Boiler and Pressure Vessel Code,Section VIII, Unfired c.

Pressure Vessels and ASME Nuclear Case Interpretations, d.

ASME Boiler and Pressure Vessel Code,Section IX, Welding Qualifications.

Standards of the A=erican Society for Testing Materials.

e.

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f.. American Standard Code for Pressure Piping, ASA B31.1,Section I (Power Piping).

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9-195 9

g.

A=erican Standard C50.20-1954 Test Code for Polyphase Induction Motors and Generators.

h.

A=erican Standard C50.2-1935 for Alternating Current Motors, Induc-tion Machines, and General and Universal Motors.

1.

Standards of themb ri nn Institute of Electrical and Electronics Engineers.

J.

Standards of the Iational Electrical Manufacturers Association.

k.

Hydraulic Institute Standards.

1.

Heating, Ventilating, and Air Conditioning Guide; A=erican Society of Heating, Refrigerating, and Air Conditioning Engineers.

m.

Southern Standard Building Code.

n.

Standards of Tubular Exchanger Manufacturers Association o.

Air Moving and Conditioning Association.

To assist in review of the system drawings, a stnMnd set of symbols and abbreviations has been used and is presented in su==ary in Figure 9-1.

91 HIGH PRESSURE INJECTION AND PURIFICATION SYSTEM l

9 1.1 DESIGN BASES 9 1.1.1 General System Function The system shown on Figure 9-2 supplies the reactor coolant system with fill and operational makeup water; circulates seal water for the reactor coolant jumps; receives, purifies, and recirculates reactor coolant system letdown to provide water quality and reactor coolant boric acid concentration control; and acco==odates temporary changes in the required reactor coolant inventory.

9 1.1.2 Iatdown Cooler The letdown cooler cools the letdown flow from reactor coolant temperature to a temperature suitable for demineralization and injection to the reactor cool-ant pump seals. The maximum letdown flow is required for a startup from a cold condition late in core life wherein the reactor coolant boron concentra-tion is reduced in 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months by an amount corresponding to the change due to moderator temperature reactivity deficit. Heat in the letdown coolers is re-jected to the n:cM tdllin6 n-component cooling system.

l 9 1.1 3 Letdown Control Valve 1

Each letdown control valve is sized for the maximum letdown rate.

9 1.1.4 Purification Demineralizer The letdown flow is passed through the purification demineralizer to re=ove reactor coolant impurities other than boron. The purification letdown flow to =aintain the reactor coolant water quality is equal to one reactor coolant volume per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The purification decineralizer is sized for the maximum letdown ficw rate as permitted by the letdown control valve. Refer to Table 11-3 for the max:.=um anticipated equilibriu= fission product accu =ulation in the reactor cool' ant.

9-2

O 9 1.1 5 High Pressure Injection Pumps The high pressure injection pumps are designed to return the letdown flow to the reactor coolant system and supply the seal water flow to the reactor cool-ant pumps. The design flow capacity is equal to the maximum letdown flow plus the seal water flow to the reactor coolant pumps. The pumps are sized to meet these requirements with one pump in operation.

~

9 1.1.6 Reactor Coolant Pump Seal Return Coolers The seal return coolers are sized to remove the heat added by the high pres-sure injection pumps and the heat picked up in passage through the reactor coolant pump seals. Heat from these coolers is rejected to the recirculated coolin6 water system.

9 1.1 7 Ietdown Storage Tank This tank serves as a surge vessel for the high pressure injection pumps, and as a receiver for the letdown flow, chemical addition, and outside makeup; and also accommodates temporary changes in reactor coolant system volume. The volume of the tank is such that the useful tank volume is equal to the maxi-mum expected expansion and contraction of the reactor coolant system during power transients.

9 1.2 SYSTEM DESCRIPTION AND EVAWATION

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v 9 1.2.1 Schematic Diagram The high pressure injection and purification system is shown schematically on Figure 9-2.

91.2.2 Performance Requirements Tables 9-1 and 9-2 list the system performance requirements and data for in-dividual system components.

91.23 Mode of operation Durin6 nomal reactor coolant system operation, one high pressure injection pump continuously supplies high pressure water from the letdown storage tank to the seals of each of the reactor coolant pumps and to a makeup line con-nection to one of the reactor inlet lines. Makeup flow to the reactor cool-ac.., systez is regulated by a flow control valve, which operates on signals from the liquid level controller of the reactor coolant system pressurizer.

A control valve in the injection line to the pump seals autccatically main-tains the desired inlet pressure to the seals. A s=all part of the water supplied to the puzp seals leaks into the reactor coolant system. The remain-der returns to the letdown storage tank after passing throu6h one of the two reactor coolant pump seal return coolers.

Seal water inleakage to the 2 setor coolant system requires a continuous let-down of reactor coolant to :m stain the desired coolant inventory. In addi-tion, bleed and feed of react r coolant are :tequired for removal of impurities

'v' and boric acid from the reactor coolant. Reactor coolant is removed from one i

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9-3

of the reactor inlet lines, cooled during passage through cne of the letdevn ecclers, passed frc= the reactor building through a reactor building isolaticn valve, reduced in pressure during flow through cne of the three letdevn ficv control valves, and then passed through one purification demineralizer to a three-way valve which directs the coolant either to the letdevn stcrage tank er to the vaste disposal system.

Normally, the three-way valve is positioned to direct the letdevn flcw to the letdcvn storage tank.

If-the boric acid concentratien in the reactor coolant is to be reduced, the three-way valve is positioned to divert the letdown ficv to the vaste disposal system. Boric acid re= oval is accc=plished in the vaste disposal system either by directing the letdown flew through a deborating de-

=ineralizer with the effluent returned directly to the letdevn storage tank or by directing the letdevn flow to a reactor coolant bleed holdup tank and main-taining the level in the letdevn storage tank with de=ineralized water pu= ped frc= the station de=ineralized rater storage tank. The quantity of unborated water received is measured and limited by inline instrumentation and interlocked with shi= rod positien controls.

The letdown storage tank also receives chemicals for addition to the reactor coolant. A hydrogen overpressure maintained in the letdevn storage tank sup-plies the hydrogen added to the reactor coolent. Other chemicals are injected in solutien to the letdevn storage tank.

Syste= control is accc=plished remotely frc= the control roc = with the excep-tien of the reactor coolant pump seal return ecoling. The letdevn flev rate is set by remotely positiening the letdevn flev control valve to pass the de-sired flow rate. The spare purification de=ineralizer can be placed in ser-vice by remote positiening of the demineralizer isolation valves. Diverting the letdown flow to the vaste disposal syste= is acce=plished by remote posi-tioning of the three-vey valve and the valves in the vaste disposal system.

The control valve in the injection line to the reactor ecolant pump seals is autcmatically centrolled by the pressure differential cor. troller connected to the reactor coolant syste= to maintain the desired inlet pressure to the seals.

The pressurizer makeup centrol valve is autc=atically centrolled by the pres-surizer level centroller. During heatup and ecoldown, the reactor ecclant syste= pressure varies frc= 100 to 2,185 psig, and the discharge pressure of the high pressure injection pu=ps re=ains about 2,600 psig. Additional centrol valves are cperated remotely in parallel with the nomal pressurizer makeup een-trol valve and the normal seal injection centrol valve to centrol the seal end makeup flow during this period. One of the three letdown centrcl valves is designed for full letdevn flev rate centrol at reduced reactor coolant system pressure.

The high pressure injecticn pumps are centrolled re=ctely.

For e=ergency operation as a high pressure injection system, the normal let-down coolant flow line and the normal pump seal injection line are closed; and flew is diverted to the emergency high pressure injection lines. The pu=ps and pu=p motors are designed to cperate at the higher flev rates and lever discharge pressures associated with the high pressure injecticn require-ments. E=ergency cperation of this syste= is discussed in detail in 6.1.

9h (Revised 4-1-67)

'd 9.1.2.h Reliability Censiderations The system has two, full capacity, letdown centrol valves, and two, full capa-city, letdown coolers for each unit to insure the flow capability needed to adjust beric acid concentration.

A spare purification demineralizer is shared between the two units. Inter-locks prevent opening the stop valves to cne unit if the stop valves to the other unit are not closed.

Two, full capacity, reactor coolant pu=p seal return coolers are supplied for each unit.

1 Each unit has two high pressure injecticn pumps, each capable of supplying the required reactor coolant pu=p seal and makeup flow. The third injection pump can be used for pump seal and makeup flow, but it is primarily provided for emergency use only.

The letdcun coolers transfer heat to the component cooling system, and the reactor coolant pu=p seal return coolers transfer heat to the recirculated i

cooling water system.

9.1.2.5 codes and standards The equipment in this system vill be designed to applicable codes and stan-dards tabulated in Section 9.

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9.1.2.6 system Isolation The letdown line and the reactor coolant pu=p seal return line penetrate the reactor building. Both lines centain electric motor-operated isolaticn valves inside the reacter building and pneumatic valves outside which are automati-cally closed by the engineered safeguards circuitry. The injection line to the reactor coolant pump seals, and the ner=al makeup line to the reactor coolant system, are inflow lines penetrating the reactor building. Each 'line centains in reacto uilding and a pneumatic valve outside which a peck va{yg%gide g; t%.?g;; inn..I : f.gn.~

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Check valves in the discharge of each high pressure injection pu=p provide further backup for reactor building isolatien. The two emergency injection lines are used for injecting coolant to the reactor vessel after a loss-of-coolant incident.

Furt' er, after use of the lines for emergency injection is disecntinued, the pnescatic valves in each line outside the reactor building vculd be c1csed remotely by the centrol room cperators.

9.1.2.7 Leakage censiderations Reactor coolant is normally letdevn to this system. The purification demin-eralizer vill remove ionic and solid contaminants while gaseous centaminants will tend to collect in the letdevn storage tank as the letdown flow is sprgyed into the gas space ' of this tank. The removal efficiencies of the purification demineralizer for impurities other than boric acid and potassium

(T hydroxide are described in 'll.l.1.3.

(Refer to Table 11-3 for the maximum

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anticipated equilibrium fission product accumulation in the reactor coolant.)

9-5 (Revised 4-1-67)'

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O The gas void in the letdown stora6e tank may be vented to the vaste disposal system by opening a remotely operated valve in the vent line. The equipment in this system is shielded by concrete. Further discussion of shieldire de-sign criteria is presented in Section 11.

9 1.2.8 creration Limits The letdown storage tank vill be maintained with a fluid inventory between 100 ft3 and 500 ft3. Oxygen accumulation in the tank will be less than 2 per cent by volu=e.

One letdown cooler and one high pressure injection pump will be functional at all times.

To prevent e.n inadvertent excessive dilution of the reactor coolant boric acid concentration, three safety measures are applied to each of the two methods of

dilutinc, i.e., the bleed and feed method and the deborating de=ineralizer (Figure 11-1) =ethod.

The first safety ceasure is a 70 gpa limitation on the maximum rate of adding demineralized water; for feed and bleed the demineral-ized water takeup control valve to the letdown storage tank is automatically controlled so as not to exceed a preset flow rate; and for deborating through the demineralizer the three-vay valve position is automatically changed to stop dilution if the demineralizer flow rate exceeds the preset flow rate.

The second safety measure is a control rod cluster assembly position interlock which either pe=its or prohibits dilution depending upon the control red pat-tern.

Because of this interlock, the demineralized water =akeup valve and the deboratin; demineralizer inlet isolation valve can be opened only when the cluster control assenblies are withdrawn to a preset position. The de=ineral-ized water makeup valve is also automatically closed, and the three-way valve position is automatically changed when the rods have inserted to a preset po-sition. The third safety ceasure consists of closing all the valves as de-scribed above when the flow has integrated to a preset value. Initiation of dilution must be by the operator, and the operator can teminate dilution at tny time.

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200 9-o-1

OV Table 9-1 Eich Pressure In.jection and Purification System (System Perfomance Data)

Letdown Flow for Purification, gpm, cold 45 Letdown Flow Maximum, spm, cold 70 Total Seal Flow to Each Reactor Coolant

c

$I 7" I#

c Pump, Spm Seal Inleakage to Reactor Coolant System per Reactor Coolant Pump, spa

$2 Injection Pressure to Reactor Coolant Pump Seals, psig, at startup 135-2,235 Injection Pressure to Reactor Coolant Pump Seals, psig, nor=al 2,235 Injection Pressure to Reactor Coolant Pump Seals, psig, maximum 2,535 Temperature to Seals, normal / maximum, F 125/150

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Purification Letdown Fluid Temperature, normal / maximum, F

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/1# /F Letdown Storage Tank Iiormal Operating Pressure, psig 15 Letdown Storage Tank Volume Between Minimum and Maximum Operating Levels, ft3 E00 Reactor Coolant Water Quality See Table 9-4 TABLE 9-1 9-7 201

1 D

Tahle 9-2 Component Data (Component quantities for two units)

High Pressure Injection Pump Quantity 6

Type Multistage centrifugal, mechanical seal Capacity, gpm See Figure 6-3 Head, ft H O at sp. gr. = 1 2

Motor Horsepower, hp 600 Pump Material SS wetted parts Design Pressure, psig 2,850 Design Temperature, F 200 Letdown Cooler Quantity 4 Full capacity Type Shell and spiral tube 6

Heat Transferred, Btu /hr

/4,/ -16. ' x 10 letdownFlow,lb/hr 3 5 x 104 Ietdown Temperature Change, F 555 to -itet /20 Material,shell/ tube CS/SS f-Design Pressure, psig 2,500 f

Design Temperature, F 600 Reactor Coolant Pump Seal Return Cooler Quantity k Full capacity Type Shell13nd tube Heat Transferred, Btu /hr 2 x 100

)

SealReturnFlov,lb/hr 1.25 x 105 Seal Return Temperature Change, F 140 to 225 Material, shell/ tube A1/A1 Design Pressure, psig 100 Design Temperature, F 200 Recirculated Cooling 'Jater Flow (each),lb/hr 1.25 x 105 Letdown Storage Tank Quantity 2

Volume, ft3 600 Design Pressure, psig 100 Design Temperature, F 200 Material CS with SS clad Furification Demineralizer Quantity 3

Type Mixed bed, boric acid saturated Cation /AnionRatio 2:1 Material SS I.

Resin Volume, ft3

-he-SO Flow, gpm 70 i

Vessel Design Pressure, psig 100 TABLE 9-2 Vessel Design. Temperature, F 200 9:8

O 92 CHE!CCAL ADDITICII A:TD SA!?LI'M SYSTE!

N 9 2.1 DEIGIT BASES 9 2.1.1 cenep l System Function 2emicaA additian and sampling operations are required to citer and monitor the concentration of various chemicals in the reactor coolant and auxiliary sys-tems. The system shown on Figure 9-3 is desicned to add boric acid to the reaecor coolant system for reactivity control (see Table 3-5 and Figure 3-1),

potassium hydroxide for pH control, ami hydrogen or.' hydrazine for oxygen con-trol. The tystem is also designed to add chemicals tc the steam generator feedvater and e ndensate systc=s. The system is designed to take up to 16 re-actor coolant samples and eight stesa generator water samples (from two units) in an eight hour sh'.ft.

This system serves both reactors.

9 2.1.2 Boric Acid Mix Tank A single horic acid =ix tank is provided as a source of concentrated boric acid solution. The volume of the tank has been chosen to provide sufficient boric acid solution to increase the reactor coolant system bcron concentration to that required for cold shutdown. Heaters are p ~'rided in the tank to maintain the temperature above that required to insure uu; ility of the boric acid. Trans-fer lines vill be electrically traced.

9 2.1 3 Boric Acid Pumps Two boric acid pumps are provided to facilitate transfer of the concentrated boric acid solution from the boric acid tank to the borated water storage tank, letdown storage tanks, or the spent fuel storage pool.

The pu=ps are sized so that when both are operating, one complete charge of con-centrated boric acid solution from the boric acid mix tank may be injected into the reactor coolant system in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

9 2.1.4 Caustic Mix Tank The volume of the caustic mix tank was established so that a solution of IIaOH of sufficient concentration could be =aintained at room temperature to neutral-ice a vaste neutralization tank full of reactor coolant containing a boric acid concentration equivalent to.that used during refueling.

9 2.1 5 Caustic pu=p The caustic pump capacity is set so that at the maximum capacity, the above neu-tralization operation can be performed in 15 minutes.

9 2.1.6 Potassium Hydroxide Tank The tank volu=e was established to contain a sufficient amount of KOH for con-tinual addition to the reactor coolant system so that a cencentrction of 3-o ppm can be =aintained while letting down at the =aximum rate.

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9 2.2 SYSTEM DESC7PIION AND EVALUATION 9 2.2.1 Sche =atic Dinaram and System Description Fisure 9-3 is a schematic diagram illustrating the features of the system. The system is operated from local controls. Two boric acid pumps, connected in parallel, take suction from the boric acid mix tank and discharge to either the spent fuel stora6e Pool, borated water storage tanks, or the letdown storage tanks. At the end of core life, bot. boric acid pumps are required to raise the reactor coolant system boron concentration from the minimum end-of-life concen-tration to the refueling concentration in approximately 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The boric acid mix tank has a mechanical mixing device and a heating unit. The equipment is provided with the instrumentation and controls so that the required boren concentrations in the reactor coolant and auxiliary systems can be maintained.

The potassium hydroxide equipment consists of a mix tank, a single positive dis-placement pump, and connecting piping. This system discharges to the letdown storage tanks.

Hydrazine drums are connected to one of two positive displacement pt=ps, which discharge to a line leading to the letdown storage tanks and to the feedvater systems. Nitrogen overpressure is used to displace the hydrazine as it is re-moved from the drums.

A hydrogen supply manifold with controls and a distribution line is used to

/O s9pply the desired overpressure in each of the letdown storage tanks during V

operation.

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A caustic solution tank and pu=p are used to supply chemicals for pH adjustment in the liquid vaste disposal system.

Chemical additions to the steam generator feedvater system consist of hydrazine as an oxygen scavenger and ea:monia for pH control. The am=onia is supplied from a solution tank with a pump to the feedvater lines. The nor=al feedvater quality is shown in Table 9-3 The liquid sampling portion of the system receives samples of the reattor cool-ant from upstream and downstream of the reactor coolant purification equipment, from upstream of the letdown coolers, from the letdown storage tanks, and from the s'econdary side of the steam Generators. Water qualities to be maintained are listed in Tables 9-3 and 9-4.

Gaseous samples are taken from the pressur-iser vapor space and from the letdown storage tank. Sample lines from these points are piped to L. sampling cubicle outside the reactor building. Samples are collected in containers designed for full operating temperature and pres-sure at flow rates of 1 and 2 gpm.

An automatic gas analyzer is used to monitor various points as mentioned in 9 2.2 3 in a continuous sequence for hydrogen-oxygen mixtures and to alarm at a preset level.

A nitrogen supply manifold with control and distribution lines provides an inert gas to purge accu =ulated gas mixtures from the, storage tanks.

V The pertinent parameters for each =ajor component in the chemical addition and Jampling system are shown in Table 9-5 9-lo 204

9 2.2.2 Performance Require =ents This system permits sampling of, and chemical addition to, the reactor coclant, feedvater, and steam syste=s, and other reactor auxiliary syste=s during normal operation and has no active emergency function. During a loss-of-coolant acci-dent, this system is isolated at the reactor building boundary.

9 2.2 3 Mode of operation The system is capable of drawin6 reactor coolant samples during reactor opera-tion, and durin6 ur.it cooldown when the low pressure injection and decay heat removal system is in operation. Access is not required to the reactor building.

Sampling of other process coolant, such as streams cr tanks in the vaste dispos-al system, is a:complished locally. Equipment for sampling secondary and non-radioactive fluids is separated from the equip =ent provided for reactor coolant samples. Leakt.ge and drainage resulting from the sampling operations are collect-ed and drained to tanks located in the vaste disposal system.

During normal opeation, liquid and vapor samples may be taken from the follow-ing points:

Liquid a.

Steam generator secondary water.

b.

Reactor coolant system.

c.

Purification demineralizer inlets, d.

Purification demineralizer outlets, e.

Deborating demineralizer outlets.

f.

Letdown storage tanks.

Vapor and Gas a.

Pressurizers, b.

Letdown storage tanks.

In addition, an oxygen and hydrogen analyser automatically samples the gas spaces of the following vessels in a predetermined sequence.

a.

Letdown storage tanks.

b.

Reactor coolant bleed holdup tanks.

c.

Miscellaneous vaste holdup tank, d.

Waste neutralization tank.

e.

Waste gas decay tanks.

f.

Evaporator condenser.

During nor=al operation, this system also delivers the following chemicals:

Boric acid to the spent fuel storage pool, the borated water storage a.

tank, and the letdown storage tanks.

b.

Caustic to the vaste neutralization tank and to the deborating demin-eralizers.

9-11 205

s c.

Potassium hydroxide to the letdown storage tank.

d.

Hydrazine to the letdown storage tanks and to the feedvater.

e.

Ammonia to the feedvater.

f.

Hydrogen to the letdown storage tanks.

g.

Nitrogen as required for the liquid vaste storage tanks and the letdown storage tanks.

9 2.2.4 Reliability considerations The system is not required to function during an emergency, nor is it required I

to take action to prevent an emergency condition.

It is therefore designed to perform in accordance with standard practice of the chemical process industry with duplicate equipment such as pumps and high pres-4 sure gas regulating valves as required.

9 2.2 5 codes and Standares

[

The equipment in this system vill be designed to applicable codes and standards tabulated in Section 9 9 2.2.6 System Isolation Isolation of this system from the reactor building is accomplished by signals from the Engineered Safeguards Protective System as further described in Sections 5 2 and 7 9 2.2 7 Leakage considerations Leakage of radioactive reactor coolant from this system within the reactor building vill be collected in the reactor building su=p.

Leakage of radioactive materi-1 from this system outside the reactor building is collected by placing the en ire sampling station under a hood provided with an offgas vent to vaste gas processing. Liquid leakage from the valves in the hood is drained to the misce]'aneous vaste storage tank.

The chcalcal addition portion of this system delivers additives to the spent fuel stora6e pool and the letdown storage tank. Additives to the spent fuel storage pool are delivered above the water level. Backflow from the letdown i

storage tank to the positive displacement pumps is prevented by two check valves in series with a throttling valve between them. Backflow from the letdown stor-s6e tank through the hydrogen addition line is prevented by a check valve and a remote manual hydrogen addition valve.

9 2.2.8 Failure Considerations To evaluate system safety, the following failures or malfunctions were assumed concurrent with a loss-of-coolant accident, and the consequences analyzed. 'As i

a result of this evaluation it is concluded that proper consideration has been given to station safety in the. design of the system.

206 9-12

00==ents and Component Failure Consecuences Pressurizer Sa=ple Electrically operated Diaphrap-operated sampling valve inside valve outside the reactor buildin; fails reactor building to close on ES signal.

vill close.

F.eactor Letdown Sa=ple Electrically operated Same as above, sampling valve inside reactor building fails to close on ES signal.

S t e 9.") Generator Steam Di.a;hra6m operated Sa=ple line is not Sample sa=pling valve outside connected directly to reactor buildin; fails reactor coolant, and to close on ES signal.

steam generator there-fare provides first barrier.

Sample Line From Any One Line breaks inside Diaphrap -op? rated of the Three Preceding reactor building valves outsidt reac-Cc ponents downstream of EMO tor building close on valves.

signal from ES system.

9 2.2 9 operating Limits 9 2.2 9 1 Boric Acid Concentration The boric acid mix tank is to be =aintained at an average te=perature of 95 F to =aintain a boric acid concentration of 7 per cent.

9 2.2 9 2 Coolant Sample Te=perature The high pressure reactor coolant samples leaving the reactor coolant sa ple cooler should be held to a temperature of 100 F to cinimize the generation of radioactive aerosols.

O_

207

~~'

9-13

i Table 9-3 i

Steam Generator Feedvater Cuality Parameter Value Maxi =um Total Dissolved Solids, ppm 0.05 Suspended Solids, ppm 0.0 Hardness, ppm 0.0 Organic, ppm 0.0 Faximum Dissolved Oxygen, ppm 0.007 Carbon Dioxide 0.0 Maximum Total Silica (as SiO ), ppm 0.02 2

Maxi =um Total Iron (as Fe), ppm 0.01 Maximum Total Copper (as Cu), ppm 0.01 pH 8.5 to 9 2 O

Table 9-4 Reactor Coolant Cuality Parameter Value Total Solids (excluding H)BO) and KOH) 1.0 ppm max.

H)BO)

See Figure 3-1 KOH 3-6 ppm pH at 77 F 5 5-6.0 pH, at 560 F (calculated) 7-10 0

10 ppb max.

2 Cl 0.1 ppm max.

H 15-40stace/l 2

Hydrazine(requiredduringshutdown) 25 ppm TABLE 9-3, 9.h m

9-14 208 1

r--

g we wrw

Table 9-5 i

Chemical Addition and Sampling System Eculpment Data Tanks Boric Acid Mix Tank Quantity 1

Type Vertical Cylindrical Volure, ft) 2,050 Design Pressure, psig 10 Design Temperature, F 200 Material Al Potassium Hydroxide Mix Tank Quantity 1

Type Vertical Cylindrical i

Volume, gal 50 Design Pressure, psig 10 Design Temperature, F 100 Material SS Ammonia Mix Tank Quantity 1

Type Vertical Cylindrical O

Volume, gal 50 Design Pressure, psig 10 i

s Design Temperature, F 100 Material CS Caustic Mix Tank Quantity 1

Type Vertical Cylindrical Volume, gal 50 Design Pressure, psig 10 Design Temperature, F 100 Material CS Hydrazine Drums Quantity 2

Type Std. Co::nercial 55 gal Drums Pump Boric Acid Pump Quantity 2

Type Reciprocating, Variable Stroke Capacity, gpm 0-10 Head, psi 25 Design Pressure, psig 100 Design Temperature, F 200 O

Pump Material SS TABLE 9-5 209 9-15

Table 9 5 (Cont'd)

Potassium Hydroxide Pump Quantity 1

Type Reciprocating, Variable Stroke Capacity, gph 0-10 Head, psi 50 Design Pressure, psig 100 Design Temperature, F 100 Pu=p Material SS An:=onia Pump C,uantity 1

Type Reciprocating, Variable Stroke Capacity, gph 0-40 Head, psi 50 Design Pressure, psig 100 Design Temperature, F 100 Pump Material CS Hydrazine Pump No. 1 C.uantity 1

Type Reciprocating, Variable Stroke Capacity, gph 0-10 Head, psi 25 Design Pressure, psig 100 Design Temperature, F 100 Pump Material SS Hydrazine Pump No. 2 Quantity 1

Type Reciprocating, Variable Stroke Capacity, gph 0-10 Head, psi 50 Design Pressure, psig 100 Design Temperature, F 100 Pu=p Material SS Caustic Pump quantity 1

Type Reciprocating, Variable Stroke Capacity, gph 0-40 Head, psi 10 Design Pressure, psig 25 Design Temperature, F 100 l

Pump Material CS Sampling Sampling Containers Quantity 10 Design Pressure, psig C,500 Design Temperature, F 670 TABLE 9-5 (cont'd)

-l 210 e

9-16 l

~

nU Table 9-5 (Cent'd)

Reactor Coolant Sa=ple Cooler Quantity 1

Type Shell and Spiral Tube Heat Transferred, Btu /hr 2.9 x lo" Sample Flow Rate, gp=

2 Max. Sample Inlet Te=perature, F 650 Sample Outlet Temperature, F 150 3

RCW Flow, lb/hr 5 x 10 Coil Side Design Temperature, F 670 Coil Side Design Pressure, psig 2,500 I

Steam Generator Sample Ccoler Quantity 1

Type Shell and Spiral Tube Heat Transferred, Btu /hr 2.8 x 10k Sample Flow Rate, gp=

2 Sample Inlet Temperature. F 525 Sample Outlet Te=perature, F 150 3

RCW Flov, lb/hr 5 x 10 i

Coil Side Design Te=perature, F 600 Coil Side Design Pressure, psig 1,050 d

9.3 CrMPONENT COOLING SYSTEM 9.3.1 DESIGN BASES The <;yste= is designed to previde cooling water for various components in the reactor buildings as follows:

letdown ecclers, reactor coolant pu=p cooling coils, quench tank cooling coils, and concrete shield cooling coils, i.e.,

ga==a heat re= oval. The total design cooling requirement for these sources is based on the maximum heat values from these sources. The syste= also provides an additional barrier between high pressure reactor coolant and service water to prevent an inadvertent release of radioactivity.

9.3.2 SYSTEM DESCRIPTION AND EVALUATION 9.3.2.1 Schematic Diagra=

Figure 9 h shows a schematic diagra= of the system. For each unit, the system contains two component cooling pumps, one ec=ponent ::coler, a surge tank, piping to the cooling coils for the four types of components, and associated valves and instrumentation. Tne cc=ponents being cooled are located inside the reactor building while the co=ponent. coolers, the component cooling pumps and the surge tank are located in the Auxiliary Building. A spare cc=ponent-ecoler is shared between the two units. This is the only shared ec=penent in this system.

O)

(

9-17 (Revised.4-1-67) 211

9.3.2.2 Perfor=ance Recuire=ents The performance requirements are listed in Tables 9-6 and 9-7.

9.3.2.3 Mode of Operation One pu=p and one ec=ponent cooler are normally operated to provide cooling water for the four types of ec=ponents in the reactor building. The pumps are rotated on a scheduled basis to =cnitor their operational capability, and cnly the co=ponent cooler associated with a reactor unit is normally used unless this cooler is out of service for maintenance or repair.

The component cooling water is chemically treated to inhibit corrosion.

Tne surge tank takes the expansion of the loop water and provides sufficient NFSH for the ec=penent cooling pu=ps.

Makeup can be added to the loop thrt, ugh a line connected to the loop at a point devnstream of the ec=ponent coolec. The source of makeup water is the station de=inerali::ed water syste=.

The operation of the syste= is =cnitored with the following instru=entation:

a.

A te=perature detector in the inlet line for the ec=ponent coolers, and temperature alar =s in the outlet line.

b.

A pressure detector en the line between the pumps and the co=ponent

coolers, c.

A flev indicator in the line between the pu=ps and coolers.

d.

A level detector en the surge tank.

e.

Te=perature detectors located on the outlet lines for letdown cool-ers and cooling coils of the reactor coolant pu=ps.

f.

A radiation monitor en the main cutlet line for the component ecclers.

2k

~~~

9-18 (Revised 4-1-67)

L

(~

U) 9.3.2.4 Reliability Considerations The system performs no emergency functions, but is designed to handle all abnormal system conditions which may occur during operation.

9.3.2.5 Codes and Standards The equipment in this system will be designed to applicable codes and standards tabulated in Section 9.

9.3.2.6 System Isolation Since the system is not used as an engineered safeguards system, Reactor Building isolation valves are automatically closed in case of an emergency.

The line going into the Reactor Building is isolated by two check valves -

the outside and one on the inside of the Reactor Building. The line one on coming out the Reactor Building is isolated by an electric motor operated valve on the inside and by a diaphragm valve on the outside of the Reactor Building.

9.3.2.7 Leakage Considerations Water leakage from piping, valves, squipment, etc., in the system inside tha Reactor Building is not considered to be generally detrimental unless the Og leakage exceeds the makeup capability. With respect to water leakage from piping, valves, and equipment outside the Reactor Building, welded con-struction is used where possible to minimize the possibility of leakage.

The component cooling water could become contaminated with radioactive water due to a leak in a letdown cooler tube or a cooling coil of a reactor coolant pump.

Tube or coil leaks in components being cooled are detected by a radiation monitor located on the main inlet cooling line. A defective coil of a coolant pump is remotely isolated with an EHO valve on the outlet cooling line and a stop-check valve on the inlet line. A letdown cooler leak is remotely isolated with EHO valves on the reactor coolant side of the cooler.

The cooling water side is completely isolated by closing a remotely operat-ed EMO valve on the inlet of the cooler and by entering the Reactor Build-ing and closing the manual valves on the outlet cooling lines. Leakage from the coil in the quench tank or the concrete shield is not likely as these involve low pressures, and the coil for the shielding is embedded in the concrete.

(N s-)

o

^

213 9-19 (Revised 4-1-67)

' ~

9.3.2.8 Failure Censideratiens 1.

Cooling line Check Valve Other Check \\alve Inlet sticks open vill be closed 2.

Cooling line EMO Valve or Other valve v: 11 Outlet diaphry= valve close en engi: leered f ai l a close safeguard ngaal.

Diaphrag: valve is en Air-to-Opern valve Tcble 9-6 Compenent Cooling System Desien Parereters for Normal Operation on a Per Unit Basis Number of Cc=ponent Cooling Pumps 2

Nu=ber of Pump-Normally Operating 1

Flew Require =er.cs, gp= per pump 650 Number of Cc=penent Coolers 1 + 1 Shared Spare Number of Coolers Normally Operating 1

Heat Removal Requirements, Btu /hr per ecoler 17.25 x 10 Table 9-7 Cc=penent Cooling Syste=

Cc=ponent Data en a Per Unit Basis Cc=ponent Cooling Pumps Quantity 2

Type Centrifugal Rated Capacity, gp=

650 Rated Head, ft H O lh5 2

Normal Operating Capacity, gp=

650 Motor Horsepower, hp L0 Casing Material Bron::e resign Pressure, Lsig 100 Design Te=perature, F 225 Cc=penent Coolers Quantity 1 + 1 Shared Spare Type Shell and Tube Heat Transferred, Btu /hr 17.25 x 10 Shell Side -

Cc=penent Cooling Water Inlet Temp., F 153 Cc=penent Ccoling Water Outlet Temp., F 100 k

i 9-20 (Revised 4-1-67)

v)

Table 9-7 (Cont'd)

Reactor Building Component Coolers (Cont'd)

Component Cooling Water Flow Rate, spm

-1, C C - to" Design Temperature, F

-30G-22#

Design Pressure, psig

-Gee- /00 Tube Side -

LPSW Inlet Temperature, F 75 LPSW Outlet Temperature, F

-Ee- /*

LPSW Flow Bate, gpm

- 5,000- 6 3##

i External Design Pressure, psig

-fee /##

External Design Temperature, F

-30026 Internal Design Pressure, psig 150 Internal Design Temperature, F 150 Tube Material Ad=iralty Metal Shell Material Carbon Steel 94 SPENT FUEL COOLING SYSTD.

9 4.1 DESIGN EASES The spent fuel cooling system shown on Figure 9-5 serves both units. It is de-

[d--)

signed to maintain the spent fuel storage pool at 120 F, with a heat load based on removing the decay heat generation from two 1/3 cores, one of which has been irradiated for 930 days and cooled for 160 hours0.00185 days 0.0444 hours 2.645503e-4 weeks 6.088e-5 months , and one which has been irra-diated for 930 days and cooled for 430 hours0.00498 days 0.119 hours 7.109788e-4 weeks 1.63615e-4 months .

In meeting the above design basis, the system has the capability to maintain the spent fuel storage pool at 150 F vhile removing the decay heat from the following combination of stored fuel assemblies:

2/3 core irradiated for 930 days and cooled for 100 days.

a.

b.

1/3 core irradiated for 720 days and cooled for 160 hours0.00185 days 0.0444 hours 2.645503e-4 weeks 6.088e-5 months .

1/3 core irradiated for 410 days and cooled for 160 hours0.00185 days 0.0444 hours 2.645503e-4 weeks 6.088e-5 months .

c.

d.

1/3 core irradiated for 100 days and coolel for 160 hours0.00185 days 0.0444 hours 2.645503e-4 weeks 6.088e-5 months .

9 4.2 SYSTEM DESCRIPTION AND EVALUATION 9 4.2.1 Schematic Diagram 4

The schematic diagram for the spent fuel cooling system is shown in Figure 9-5 Spent fuel cooling is accomplished by pumping spent fuel storage pool water through coolers and back to the spent fuel storage pool.

In addition to this primary function, the system also provides for purification of both the spent fuel storage pool water and the contents of the borated water storage tank p

(after it has been used in the fuel transfer canal during refueling).

t i

O 9-21 l

l

9 4.2.2 Ferformance Require =ents The first design basis of the system predicates an operating schedule in which each unit is on an equilibrium refueling period (310 FFD per cycle) with ap-proxicately 1/3 of a core being removed from each unit at the end of each pe-riod. The removed fuel assemblies will have been in the reactor for three

cycles, i.e., 930 days at the ti=e of discharge.

The second design basis for the system considers that it is possible that dur-ing the life of the station it will be necessary to totally unload one reactor vessel for maintenance or inspection at the ti=e that the 2/3 cores are already residing in the spent fuel storage pool.

The basic system and co=ponent data are presented in Table 9-8.

9.h 2 3 Mode of Operation During normal conditions either 1/3 or 2/3 of a core will be stored in the pool.

When 1/3 of a core is present, one of the pumps and one of the coolers will handle the load and =aintain 120 F.

When 2/3 of a core is stored, both pumps and both heat exchangers will be operated to maintain the 120 F temperature.

The pool is initially filled with water from the borated water storage tank.

For the case where 1-2/3 cores are stored (due to complete unloading of one re-actor vessel), two pumps and two spent fuel coolers will caintain the spent fuel storage pool temperature at 150 F.

If both a pu=p and a cooler are out for =aintenance when this storage condition exists, the water temperature will eventually rise to 205 F, although considerable time will be required to heat the large spent fuel storage pool to this temperature. If all cooling is lost, the time required for the spent fuel storage pool to reach 205 F for each of the above quantities of stored fuel is as follows:

One-third of a core kh hours Two-thirds of a core 27 hours3.125e-4 days 0.0075 hours 4.464286e-5 weeks 1.02735e-5 months One and two-thirds cores 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months 9 4.2.4 neliability considerations During the time when 2/3 core is stored in the pool, two pumps and two coolers will be utilized to caintain the pool at 120 F.

If both a heat exchan6er and a pump are lost at this time, the spent fuel storage pool will be maintained at 150 F.

9 4.2 5 Codes and Standards The equip =ent in this system vill be designed to applicable codes and standards tabulated in Section 9 9 4.2.6 Ieakage Considerations Whenever a leaking fuel assembly is transferred from the fuel transfer canal to the spent fuel storage pool, a s=all quantity of fission products may enter the spent fuel cooling water. A small purification loop is provided for re=oving these fission products and other contnMnants from the water. t 216 9-22

m 4

The fusi handling and storage area housing the spent fuel storage pool will be f

ventilnted on a controlled basis, e:taustin6 circulated air to the outside throu6h the station vent.

Provisions rave been made in the des 1 n to air-test the valved and flanged ends 6

of the fuel transfer tubes for leak-ti6htness after it has been used. For each reactor unit, a valve and blind flan 6e are used to isolate cach fuel transfer i

tube.

4 9 4.2 7 Failure Considerations The most serious failure of this system would be complete loss of water in the stora6e pool. To protect a6ainst this possibility, the spent fuel stora6e Pool coolin6 connections enter near or above the water level so the.t the pool can-not be Gravity-drained. For this same reason care is also e::ercised in the design and installation of the fuel transfer tube.

9.4.2.8 Operation Limits The yaol will normally be limited to 150 F except in most unusual circu= stances as previously described. Boric acid concentration in the pool fluid will be i

=sintained at 12,000 to 13,000 ppm.

t 1

l

\\

l i

i i

V[

+

9-23

i i

e t

i Table 9-8 Spent Fuel Cooling.iystem and Component Data 4

SystemCoolingCapacity,)Bfu/hr i

6 Normal (two 1/3 cores 14 3 x 10-i Faximum (1-2/3 cores) 27 5 x 10' System Des 1 n Pressure, psig 50 6

l System Design Temperature, F 250 I

j Spent Fuel Cooler Data Quantity 2

Type Tube and Shell Material Aluminun Duty, Btu /hr/ cooler (3) 7 15 x 10 i

RCWFlow,lb/hr/ cooler 5 x 105 3 pent Fuel Pump rata Quantity 2

Type Horizontal, Centrifugal l

Mat;erial Stainless Steel Flow,gpm/pu=p 1,000 I

Head, ft H O 100 2

Motor Horsepower, hp 40 i

1 l

Spent Fuel Storage Pool Volume, ft 75,000 3

i (a) Assumes pool water to cooler at 120 F and cooling water to

.f cooler at 90 F.

b

)

i i

J TABLE 9-8 m

9-2h 218

/

95 Im PRESSURE INJECTION AND DECAY HEAT Em0 VAL SYSTEM 931 DESIGN EASES 9 5 1.1 General System Function The normal function of this system as shown by Figure 9-6 is to remove reactor decay heat during the latter stages of cooldown, maintain reactor coolant te=-

perature during refueling, and provide the means for filling and drainin6 the fuel transfer canal. The emergency functions of this system are described in 6.1.

9 5 1.2 MW Pressure Injection and Decay Heat Removal Pumps pupi;m 4 hvTsM/

.w"TP &

The low pressure injection and decay heat removal pumps, in n:rri :;r-tirn, circulate the reactor coolant from one reactor inlet line throu6h the decay heat cooler and return it to the reactor injection nozzles. The design flow is that required to cool the reactor coolant system from 250 F to 140 F in 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months .

(The steam generators are used to reduce the reactor coolant system from operating temperature to 250 F.)

9513 low Pressure Injection and Decay Heat Removal Cooler Dunsso A R*"r!Mr k"W" The low pressure injection and decay heat removal coolers, t r

' ^p - ti:n, remove the decay heat from the circulated reactor coolant. Each cooler is de-p signed to remove the decay heat being generated 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months after shutdown of the Q

m : tor (14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months after reaching 250 F).

952 SYSTEM DESCRIFIION AND EVAWATION 952.1 Schematic Diagram The low pressure injection and decay heat re= oval system is shown schematically in Figure 9-6.

9 5.2.2 Performance Requirements Tables 9-9 and 9-10 at the end of this subsection list system performance data and design data for individual components.

9523 Mode of Operation Two pumps and one cooler perform the decay heat cooling functions for each re-actor unit. After the steam generators have reduced the reactor coolant tem-perature to 250 F, decay heat cooling is initiated by-i;ul; tin;

p-re w-

-.. 1_r: M th: ;hr.r;i j7, -- ' aligning pumps to take suction from the re-actor M line and discharge through one cooler into t' e reactor vessel. If only one pu=p and one cooler are available, reduction el reactor coolant tem-perature is accomplished at a lower rate.

The equipment utilized for decay heat cooling is also used for low pressure injection during accident conditions.

'~'

/N(b 2l9 9-25

O 9.5.2.4 Reliability Considerations gg

@+' onPA c tf1 g*,

. b 1 Y. - Y h, Eh$ $__'YA..b.._Y.l h., Y___. $.... ?? $..t. N. 5.2 Y... $

YtE*"** - -

ou m m u mo um.._

y

^-^ g-acy M ckup.

9.5.2.5 Codes and Standards The equipment in this system will be designed to applicable codes and standards tabulated in Section 9.

9.5.2.6 f.ystem Isolation The decay heat removal system is connected to the reactor line on the suction side and to the reactor vessel on the discharge side. On the suction side the connection is through two electric motor-operated gate valves in 7w,, C"/*" "#'"#3 series and on the discharge side through one air-operated gate valve and e-

/w UAW 4.c c'a 12PJe 1" eariee. All eino+e of these valves are normally closed when-ever the reactor is in the operating condition.

In the event of a loss-of-coolant accident, the valve on the discharge side opens but the valves on the suction side remain closed throughout the accident.

9.5.2.7 Leakage Considerations During reactor operation all equipment of the low pressure injection and decay heat removal system is idle, and all isolation valves are closed.

During the accident condition fission products will be recirculated through the exterior piping system. To obtain the total radiation dose to the public due to leak-age from this system, the potential leaks have been evaluated and discussed in 6.4 and 14.2.

9.5.2.8 Failure Considerations Failure considerations for the accident case have been evaluated and tabulated in 6.1.2.9 and 6.2.2.9.

i 220 9-26 1

O i

}

4 i

Table 9-9 Iow Pressure Injection and Decay Heat Removal System i

(SystemFerformanceData)

Reactor Coolant Temperature at Startup of Decay Heat Removal, F 250 Time to Cool Reactor Coolant System from 250 F to 140 F, hr 14 Refueling Temperature, F 140 s

Decay Heat Generation Figure 9-7 Fuel Transfer Canal Fill Time, hr 1

Fuel Transfer Canal Drain Time, hr 1

H B0 j

3 3 concentration in Borated Water Storage Tanks, ppm 13,000

)

[

e 4

4 t

t' TABLE 9-9 9-27

~

{

4 s

Table 9-10 Iow Pressure Injection and Decay Heat Re= oval System Component Design Data (Component Quantities for Two Units) d Pumps Quantity 6

Type Single stage, centrifugal Capacity (each),gpm 3,000 Head at rated capacity, ft H O 350 2

Motor horsepower, hp 400 Material SS (wetted parts)

Design pressure, psig 300 Design temperature, F 300 Coolers (a)

Quantity 4

Type Sht.31 and tube

'.feattransferred, Btu /hr 65 x 100 3eactor coolant flow, gpm 6,000 L.P. service water flow (each), gym 6,000 L.P. service water inlet temperature, F 75 Material,shell/ tube A1,'Al Designpressure,shell/ tube 300/150 Desi n temperature, F 300 6

Borated Water Storage Tanks Quantity 2

Capacity (each), gal

^2,T^

Ji# ###

s Material Al e

Design pressurc Hydrostatic head Desi n temperature, F 150 6

(a) Refer to Figure 6 5 for heat transferred as a function of cooler inlet water temperature.

TABLE 9-10 t

J 9-28 222

/

96 COOLING WATER SYSTEMS 9.6.1 DESIGN EASES The coolin6 water systems for this station will have sufficient redundancy to insure continuous heat removal from components requiring cooling.

The systems will be sized to insure adequate heat removal based on highest expected temperatures of coolin6 water, maximum loadings, and leaka$e allow-ances. The equipment in these systems will be designed to applicable codes and standards tabulated in Section 9 The entire coolin6 sater systems will be desi ned to prevent a component fail-6 ure from curtailing normal station operation. It will be possible to isolate all heat exchangers and pumps and bypass all pressure-reducin6 valves.

All systems will be monitored and operated from the control room.

Isolation valves will be incorporated in all service water lines penetratin6 the Reactor Buildings.

Electrical power requirements for all coolin6 water systems can be supplied by any of the redundant power sources described in 8.2 3 All system components will be hydrostatically tested prior to Station startup O

and will be accessible for periodic inspf.:tions during operation. All elec-b trical components, switchovers, and start ng controls will be tested period-ically. Design parameters for system com,onents are listed in Table 9-11.

9 6.2 SYSTEM DESCRIPTION AND EVATATION j

9.6.2.1 condenser Circulating water system The Little River Arm of Iake Keowee will be the source of water for the con-denser circulating water (CC4) systems. F16ure 9-8 shows the arrangement of this system with respect to the two branches of lake Keowee. The intake struc-ture will have a suction extending below maximum drawdown of the lake with screens that can be manually re=oved for periodic cleaning. Each unit will have four circulating water pumps serving two supply lines to the condenser i

intake conduit. Only one pump is necessary to insure an adequate supply of I

water to meet the cooling water requirements of a unit under emergency con-ditions.

The CCW systems will be designed to take advantage of the siphon effect so that the pumps will only be required to overcome pipe and condenser friction l

losses. The siphon will be initiated at startup by a mechanical vacuum pump and sustained, if necessary, by steam operated air removal equipment during operation.

The discharge from the CCW systems will go to the Keo ee River Arm of Lake Keowee as shown on Figure 9-8.

Should all pumps be lost, an ecergency dis-charge line to the Keovee Hydro tailrace will be opened. The CCW syste=s U

'223 i,

9-29 e ?,,

would continue to operate as unassisted ciphon systems supplying sufficient water to the condensers for all emergency cooling requirements.

9 6.2.2 High Pressure Service Water System There will be one high pressure service water (HPSW) system for the station with three motor-driven pumps taking suction from either one or both conden-ser intake conduits. One pump per unit will be satisfactory for fire protec-tion service with the third pump serving as a spare. Figure 9-9 shows the ar-range =ent of this system as well as the low pressure service water (LPSW) sys tem.

A 1CO,000 gallons elevated storage tank will serve as backup for the HPSW sys-tem.

9 6.2 3 Iow Pressure Service Water System There will be one low precsure service water (LPSW) system for the station with three motor-driven pumps taking suction from either one or both conden-ser intake conduits. One pump per unit will be required for nor=al service with the third pump serving as a spare. Any two of the three pumps will pro-vide for both the emergency condition of one unit and the shutdown or normal requirements of the other unit. Figure 9-9 shows the arrangement of this sys-tem as well as the HPSW system.

For emergency operation, the LPSW system supplies cooling water to tne low grnes6sce pressure injection coolers,"h; opt-j :^^1c r -and the Reactor Building. _.

c,s., da 37 --

__1.

with automatic opening of the disenarge valves on each cooler.

veer $

The low pressure service water discharge from the low pressure injection cool-ers,e_h; :prn; ^'c1 r and the Reactor Euilding component coolers will te con-tinuously monitored for radioactivity.

If radioactivity is detected, the cooler ser vice water flow to the CCW discharge from the defective cooler will be terminated by placing spare cooler in service.

Should tne pressure on the LPSW system drop for any reason, backup is achieved by eith er one of two connections to the HPSW system through pressure-reducing valves.

9 6.2.1 Recirculated Cooling water System There will be one recirculated cooling water (RCW) system for the station to serve the Turbine Building and Auxiliary Building components. This closed loop system consists of a surge tank, two motor-driven pumps and two heat ex-changers cooled by condenser circulating water. Figure 9-10 shows the ar-rangement of this system.

For operation, one of the RCW pumps and one of the heat exchangers will be re-quired. The other components will serve as spares.

Makeup to the RCW system is from the condensate system of either unit.

The 20,000 gallon surge tank provides storage of RCW until makeup can be manually added. High and low surge tank water levels will be annunciated in the con-trol room.

9-304

O Table 9-11 Cooling Water Systems Component Design Parameters Parameter Value Condenser Circulating Water Pumps 4 per unit Flow (per pump), gpm 164,000 Design temperature, F 75 Desi n pressure, psig 50 5 High Pressure Service Water Pumps 3 for both units Flow (perpump), gym 5,100 Design temperature, F 75 Desi6n pressure, psig 150 Iow Pressure Service Water Pumps 3 for both units Flow (perpump),gpm 20,000 Design temperature, F 75 ] Design pressure, psig 150 l Recircuhted Cocling Water Pumps 2 for both units Flow (perpump),gpm 4,000 Design temperature, F 250 Design pressure, psig 150 Recirculated Cooling Water Heat Exchangers 2 for both units Type Shell and tuce Recirculating cooling water flow, each (shellside),gpm 4,000 Recirculating coolin'g water outlet temperature, F 90 Condenser circulatirg water inlet temperature, F 75 Desi6npressure,ehell/ tube,psig 100/100 Designtemperature,shell/ tube,F 250/150 Tube material Admiralty metal Shell material Carbon steel i TABLE 9-11 ) 9-31 ~~ 225

pb 97 FUEL HANDLING SYSTEM 971 Dm IGN EASES 9 7 1.1 General System Function The fuel handling system shown on Figure 9-11 is designed to provide a safe, effective means of transporting and handling fuel from the time it reaches the station in an unirradiated condition until it leaves the station after postirradiation cooling. The system is designed to minimize the possibility of mishandlin3 or maloperations that could cause fuel assembly dama6e and/or potential fission product release. Separate fuel handling equipment is provided for each reactor, but a common fuel handling and stora6e area will serve both reactors. The reactors are refueled with equipment desi ned to handle the spent fuel 6 assemblies underwater from the time they leave the reactor vessels until they are placed in a cask for shipment from the site. Underwater transfer of spent fuel assemblies provides an effective, economic, and transparent radiation shield, as well as a reliable cooling medium for removal of decay heat. Borated water insures suberitical conditions during refueling. 9 7 1.2 New Fuel Storage Area The new fuel storage area is a separate and protected area for the dry stora6e Q of new fuel asse=blies. The new fuel storage area is sized to accommodate the maximum number of new fuel assemblies required for refueling of both reactors as dictated by the fuel management program. The new fuel assemblies are stored in racks in parallel rows having a center to center distance of 21 in. in both directions. This spacing is sufficient to maintain a keff of less than o.9 v' en wet. 9713 Spent Fuel Storage Pool The spent fuel storage pool is a reinforced concrete pool lined with stainless steel located in the auxiliary building. The pool is sized to accommodate a full core of irradiated fuel assemblics in addition to the concurrent storage of the largest quantity of spent fuel 1sse=blies from both reactors as estab-lished by the fuel =anagement pro 6 ram. The spent fuel assemblies are stored in racks in parallel rows having a center to center distance of 21 in. in both directions. Control rod cluster assemblies requiring removal from the reactors are stored in the spent fuel assemblies. 9 7 1.4 Fuel Transfer Tubes Two horizontal tubes are provided to convey fuel between the respective reactor buildings and the auxiliary building. These tubes contain tracks for the fuel transfer carriages, gate valves on the spent fuel storage pool side, and a means for flanged closure on the reactor buildin6 side. The fuel transfer tubes pene-trate into the fuel transfer canals at their lover depth, where space is pro-vided for the rotation of the fue3 transfer carriage baskets containing a fuel ( assembly. \\ N. i 4, 226 9-32

9715 Fuel Transfer Canal The fuel transfer canal is a passageway in the reactor building extending from the reactor vessel to the reactor building vall. It is formed by an upward extension of the primary shield valls. The enclosure is a reinforced concrete structure lined with stainless steel which for=s a canal above the reactor ves-sel which is filled with borated water for refueling. Space is available in the fuel transfer canal for underwater storage of the re-actor vessel internals upper plenum chamber assembly. The deeper fuel transfer station portion of the fuel transfer canal contains the new fuel handling racks. This portion of the fuel transfer canal can also be used for storage of the reactor vessel internals core barrel and thermal shield assemblies by temporarily removing the new fuel handling racks. 9 7 1.6 Miscellaneous Fuel Handling Equipment This equipment consists of fuel handling bridges, fuel assembly handling tools, new fuel storage racks, spent fuel storage racks, new fuel handling racks, fuel assembly transfer containers, control rod assembly handling tools, viewing equip- =ent, fuel transfer mechanisms, and shipping casks. In addition to the equip-ment directly associated with the handling of fuel, equip =ent is provided for handling the reactor vessel closure head and the upper plenum chamber assembly to expose the core for refueling. 972 SYSTEM DESCRIPIION AND E'/ALUATION 9 7 2.1 Receiving and Storing Fuel New fuel assemblies are received in shipping containers and stored dry in racks having a center to center distance of at least 21 in. They are subsequently moved into the reactor building in one of the following vays. a. During reactor operation, new fuel transfer containers can be used to transfer new fuel assemblies from the storage area to the reactor building. The new fuel assemblies, in their containers, are moved into the reactor building in a horizontal position through the per-sonnel access hatch air locks. The new fuel assemblies are removed from the transport vehicle and stored vertically in racks which pro-vide an eversafe geometric pattern of 21 in. in both directions. Following reactor shutdown for refueling and prior to the start of actual refueling operation, the new fuel assemblies are transferred by crane to the new fuel handling racks located in the deep area of the transfer canal adjacent to the transfer tube. b. After reactor shutdown, new fuel assemblies can be transferred from the new fuel storage area into the reactor building through the equip- =ent hatch and stored directly in the new fuel handling racks in the transfer canal. After reactor shutdown, new fuel assemblies can be transferred from th c. new fuel storage area to the new fuel handling racks in the transf l 9-33 i

4 canal by way of the spent fuel storage pool with the use of the fuel 4 transfer carriages and the fuel transfer tubes. 9 7 2.2 loading and Removing Fuel Following the reactor shutdown and reactor building entry, the refueling pro-cedure is begun by removal of the reactor closure head and control rod drive assemblies. Head removal and replacement time is minimized by the use of two stud tensioners. The stud tensioner is a hydraulically operated device which pemits preloading and unloading of the reactor vessel closure studs at cold shutdown conditions. The studs are tensioned to their operational load in two steps in a predetermined sequence. Required stud elongation after tensioning is verified by micrometer measurements. Following removal of the studs from the reactor vessel tapped holes, the studs and nuts are supported in the closure head bolt holes with specially designed spacers. Removal of the studs with the reactor closure head minimizes handling time and reduces the chance of thread damage. cQ )$M,wre n S The reactor closure head is initially raised by use of t er 11!t N 7 [Ms o hold } the flange surfaces parallel. Following the initial lift, the reactor closure 2 head assembly is lifted out of the canal onto a head storage stand on the oper-ating floor by a head handling sling attached to the polar crane. The stand is i designed to protect the gasket surface of the closure head. The lift is guided by two alignment pins installed in two of the stud holes. These pins also pro-vide proper align =ent of the reactor closure head with the reactor vessel and in-i \\ ternals when the closure head is replaced after refueling. The studs and nuts can be removed from the reactor closure head at the storage location for inspection and cleaning using special stud and nut handling fixtures. A stud and alignment pin storage rack is provided. The annular space between the reactor vessel flange and the bottom of the fuel transfer canal is sealed off prior to filling of the canal by a seal clamped to the canal shield plate flange and the reactor vessel flange. The fuel j transfer canal is then filled with borated water. l The upper plenum chamber assembly is removed from the reactor by the polar crane and stored underwater on a stand on the fuel transfer canal floor using a lift-ing device with special adapters. Refueling operations aided, if necessary, by an underwater TV viewing system are carried e from two fuel handling bridges which span the fuel transfer canal in each reactor building. One bridge is used to shuttle spent fuel as-semblies from the con to the transfer station and new fuel assemblies from the new fuel handling racks to the core. During this operation, the second bridge is occupied with relocating partially spent fuel assemblies in the core as specified by the fuel management program. A manually operated fuel handling tool is suspended from a hoist on each bridge for handling the fuel assemblies. A manually operated control rod' assembly han-dling tool is suspended from a second hoist on one of the bridges for handling the control. rod cluster assemblies. Long handled hook tools are'available to aid in guiding the handling tools, m 228 ia 9-34 .--y +%

A two-hoist bridge moves a spent fuel asse=bly from the core underwater to the transfer station where the fuel assembly is lovered into the fuel transfer car-riche fuel basket. The control cluster asse=bly handling tool attached to the second hoist is used to transfer a control cluster assembly to a new fuel Assem-bRr in the adjacent new fuel handling racks. This new fuel assembly with con-t,rol cluster assembly is carried to the reactor by the fuel assembly handling tool and hoist, and located in the core while the spent fuel asse=bly is being transferred to the spent fuel storage pool. Spent fuel assemblies removed from the reactors are transported to the spent fuel storage pool from the reactor buildings via fuel transfer tubes by means of a cable-operated fuel transfer carriage. The spent fuel asse=blies are re-coved from the fuel transfer carriage fuel basket using a manually operated fuel asse=bly handling tool suspended from a =onorail electric hoist located on a movable fuel handling bridge. This motor-driven bridge spans the spent fuel storage pool and permits the refueling crew to store or re=ove new fuel assem-blies in any one of the many vertical storage rack positions. Long handled hook tools are available to the refueling crew to help guide the handling tool. The fuel transfer mechanisms are underwater cable-driven carriages that run on tracks extending from the spent fuel storage pool through the transfer tubes and into the reactor buildin6s. Each of the two independently operated fuel trans-fer mechanisms is designed to operate in two directions so that either of the two reactor buildings can be serviced by one or two mechanisms as required. Ro-tating fuel baskets are maintained on each end of both fuel transfer carriages to receive fuel assemblies in a vertical position. The hydraulically operated fuel basket on the end of the carriage being used for refueling is rotated to a horisontal position for passage through the transfer tube, and then rotated back to a vertical position in the spent fuel storage pool for vertical removal of the fuel assembly. Once refueling is completed, the fuel transfer canal vater is drained by suction through a pipe located in the deep transfer station area. The canal water is pumped to the borated water storage tank to be available for the next refueling or for emergency cooling following a loss-of-coolant accident. During operation of the reactors, the carriages are stored in the spent fuel storage pool, thus permitting gate valves on the spent fuel storage pool siac of each transfer tube to be closed and blind flanges to be installed on the reactor building side of the tube. Spa:.e is provided in the spent fuel storage pool to receive a spent fuel shipping cast as well as provide for required fuel storage. Following a sufficient decay period, the spent fuel assemblies are removed from storage and loaded into the spent fuel shipping cask underwater for removal from the site. Casks up to 100 tons in veight are handled by the fuel building crane. A decontamination area is located in the building adjacent to the spent fuel storage pool where the outside surfaces of the casks can be decontaminated prior to chipment by using steam, vater, or detergent solutions, and =anual scrubbing to t:w am.nt required. ~ ~ 229 9-35

(D V 9723 Safety Provisions o Safety provisions are designed into the fuel handling system to prevent the development of hazardous conditions in the event of component malfunctions, accidental da=sge, or operational and administrative failures during refueling or transfer operations. All fuel assembly storage facilities, new and spent, maintain an eversafe geo-metric spacing of 21 in. between assemblies. The new and spent fuel storage racks are designed so that it is impossible to insert fuel assemblies in other than the prescribed locations, thereby insuring the necessary spacing between assemblies. Although new fuel assemblies are stored dry, the 21 x 21 in. spacing is to insure an eversafe geometric array in unborated water. Under these conditions, a criticality accident during refueling or storage is not considered credible. All fuel handling and transfer containers are also designed to maintain an eversafe geometric array. 1/echanical damage to the fuel assemblies during trcusfer operations is possible although remote. Since the fission product relesse would occur underwater, the amount of activity reaching the environ-ment will present no appreciable hazard. A fuel handling accident analysis is included in Section 14. All spent fuel assembly transfer operations are conducted underwater. The water level in the fuel transfer canal provides a minimum of 10 ft of water T over the active fuel line of the spent fuel assemblies during movement from d the core into ctorage to limit radiation at the curface of the water to less than 10 mrem /hr. The spent fuel storage racks are located to provide a min-imum of 13 ft of water shielding over stored assemblies to limit radiation at the surface of the water to no more than 2 5 mrem /hr during the storage period. The depth of the water over the fuel assemblies, and the thickness of the con-crete walls of the transfer canal, are sufficient to limit the maximum contin-uous radiation levels in the working area to 2 5 mrem /hr. Water in the reactor vessel is cooled during shutdown and refueling by the low pressure injection and decay heat removal system as described in 9 5 In case of a power failure, this system will be operated by the auxiliary power sup-ply. The spent fuel storage pool water is cooled by the spent fuel cooling system as described in 9 4. A power failure during the, refueling cycle will create no immediate hazardous condition due to the large water volume in both the transfer canal and spent fuel storage pool. With a normal quantity of spent fuel asceablies in the storap pool and no cooling available, the water temperature in the spent fuel stcrage pool would increase as discussed in 9 4.2 3 During the refueling period the water level in both the fuel transfer canal and the spent fuel storage pool is the same, and the fuel transfer tube valves are continuously open. This eliminates the necessity for interlocks between the fuel transfer carriages and transfer tube valve operations. The simpli-fled movement of a transfer carriage through the horizontal fuel transfer tubes minimizes the danger of Ja= ming or derailing. The open tube design pro-m vides access to the entire length of the fuel transfer carriage travel from [V) the fuel transfer canal to cope with such an eventuality. All operating mech-anisms of the system are located in the fuel handling and storage area for -l L.. 9-36

ease of maintenance and accessibility for inspection prior to start of refuel-s ing operations. During reactor operations bolted and gasketed closure plates, located on the reactor building flanges of the fuel transfer tubes, guarantee that spent fuel storage pool water will not leak into the transfer canal in the event of a leak through the transfer tube valves. Both the spent fuel storage pool and the fuel transfer canals are completely lined with stainless steel for leak-tightness and for ease of decontamination. The fuel transfer tubes will be appropriately attached to these liners to maintain leak integrity. The spent fuel storage pool cannot be accidentally drained since water must be pu= ped out through a suction pipe. The fuel transfer mechanisms are designed to per-mit initiation of the carriage travel and the carriage fuel basket rotation from the building in which the carriage fuel basket is being loaded or un-loaded. All electrical gear is located above water for greater integrity and ease of =aintenance. The hydraulic systems which actuate the rotating fuel baskets use storage pool water for operation to eliminate contn*ation. The fuel transfer canal and storage pool water will have a boron concentration of 2,270 ppm. Although this concentration is sufficient to caintain core shutdown if all of the control cluster assemblies were removed from the core, only a few control rods will be removed at any one tice during the fuel shuf-fling and replacement. Although not rep tred for safe storage of spent fuel assemblies, the spent fuel storage pool water will also be borated so that the transfer canal water will not be diluted during fuel transfer operations. The fuel handling bridge hoist travel, tool, and sling length are designed to limit the -nWmm lift of a fuel assembly to a safe shielding depth. Relief valves are provided on each stud tensioner to prevent overtensioning of the studs due to excessive pressure. Gross failures of fuel are p;evented by safety margins in the design and con-trol of the core. The proposed fuel assembly utilizes a free-standing Zire-aloy fuel rod of sufficient length to accom=odate the expected fission gas release from the fuel. Any leaking fuel assemblies, if they occur, are removed from the core for veri-fication of leakage and placed in a failed fuel assembly container. This operation is accomplished in the fuel transfer canal and completely seals off the leaking fuel assembly before it is transferred out of the fuel transfer canal into the spent fuel storage pool by the fuel transfer mechanism. The design of the failed fuel assembly containers will comply with 10 CFR 71 so that defective fuel assemblies can be safely stored and shipped while sea. Lect in the failed fuel container. 9 7 2.4 operational Limits Certain manipulations of the fuel assemblies and reactor internal assemblies during refueling ray result in short term expcsures with radiation levels greater than 2 5 mrem /hr. The exposure time will be limited so that the in-tegrated doses to operating personnel do not exceed the limits of 10 CFR 20. 9-37 s

O The fuel handling bridges are limited to handling of fuel and control rod cluste-a.ssemblies and reactor closure head studs only. All lifts for handling of reactor closure heads and reactor internal assemblies will be made using the reactor building cranes. Travel speeds for the fuel handling bridges, hoists, and fuel transfer car-riages will be controlled to insure safe handling conditions. 98 STATION VENTILATION SYSTEMS 9 8.1 DESIGN BASES The station will be designed to provide maximum safety and convenience for operating personnel with equipment arranged in zones so that potentially con-taminated areas are separated from clean areas. The heating, ventilating, and air conditioning systems for the station will be designed to provide a suit-able environment for equipment and personnel. The path of ventilating air in

he Auxiliary Building will be from areas of low activity toward areas of pro-gressively higher activity. Ventilating air will be recirculated in clean areas only.

9 8.2 SYSTEM DESCRIPTION AND EVALUATION The Beactor Building nor=al ventilation system is discussed in 5 3 and shown on Figure 5-6. The penetration room ventilation system is discussed in 6.5 ( and shown on Figure 6-15 The remaining ventilation systems for the station y are discussed here and shown on Figures 9-12 and 9-13 The equipment used'to ventilate each area is independent from that used in any o%er area. The sys-tems handling potentially contaminated air all discharge to the station vent. The Auxiliary Building will be served by separate ventilation systems for the fuel handling area, the radwaste area, the nonradioactive area, and the con-trol room area. These systems are shown on Figure 9-12. The system serving the nonradioactive areas of the Auxiliary Building will also supply air to the laundry, hot laboratory, showers, toilets, and hot machine shop. The dis-charge air from these areas will go to the station vent. The control room area system will be equipped with redundant fans, filters, and mechanical re-frigeration equip =ent, plus the necessary dampers and controls for =anually switching t'o full recirculation for postaccident ventilation. The turbine area ventilation system will recirculate air with provisions for makeup as required from a fresh air louver. Exhaust air will be discharged directly to the atmosphere through roof ventilators. The Administration Building ventilation system shown on Figure 9-13 will aon-sist of a =ultizone-type air handling unit with a chilled water cooling coil and a heating coil. The system will be arranged to receive makeup from a fresh air louver. Exhaust air will be discharged directly to ' the atmosphere through an exhaust fan. The ventilating equipment will be in accordance with accepted industry stan-dards for power station equip =ent. Redundant exhaust fans will be provided [v) for the potentially contaminated areas, and a completely redundant ventil.a+1on 9-38 N2

system will be provided for the control room area. The control room area sys-tem perfor=a::ce will be continually =onitored with alarms for high radiation, fan failure and excessive pressure drop through filters. The control room operator will have manual control for selecting backup fan and filter opera-tion in order to insure satisfactory control room conditions following an ac-cident. All control area ventilating system fans and filters will be remote from the control area and will not be exposed to fire hazards. The ventilation systems will be designed in accordance vi.h the applicable codes and standard.a tabulated in Section 9 The ventilating equipment will be accessible for periodic testing and inspec-tion during normal operation. Where redundant equipment is provided, it will be ~ operated alternately to provide assurance of operability. O O__ i' 9-39 2

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me X .e saw (O*9PCWfM V COOLeMJ s Aa,70 70 sm waa n n. n! r ~- J-9r-, ,m a~,' no. 2 Q,, :: -6 --e> Sa" r_r2 sm_w_ _ Co** m ENT CN PWP4P4 Do-N-~ x ,A er --rw M3?ts sewt t AnManetMear MsCArt D* sMr! / AM A8092 JACfPr 904 Tm40 Cth=9*'tw2Nr COOstM M Mne9fMCLATL/Af 426tND JN PseuM2 9 / "Habeo...uw, c= =.13. COMPONENT COOLING SYSTEM A00t0 Sumu tw \\ omst rema OCONEE NUCLEAR STATION FIGURE 94 240 i '1

c 4 INSIDE 0g73499 UNIT 1 RB REA(JOR -l BLDG ~~~ I LEAKAGE I RB d5 M f FUEL TRANSFER f UNAL I g UNIT 1 V I l AL SPENT FUEd h j [g i STORAGE poc I fe l r i i (o at co)- I O'A' 4 TO DECAY HEAT RB POMP SUCTION LINE X FROM us gg 63 RATED WATER STORA(.h TANKS

I I

/ X gg RB NOTE: FOR LEGEND NOMEN SEE FIGURE 9 3 2kl

r \\ l 4 OUT6lDS INSIDE UNIT 2 Rb REA I _ - ~ _ ~ ~~ g, 5D LEAKAGE l L3 I i FUEL TRANSFER l g i CAN AL I Qi l V l UNIT Z lp mi : i m l I ,1 l TO BORATED WATER STORAGE TANKS 4 4 I I g3 LTO DECAY HEAT PUMP 50CTioN LINE pcw 43e SF LOOLERb Rcw - 6 AL40) f-RS (lik T LO REV; 4-4-67 PERCUTED FUEL TRANSFER CANAL DRAIN LINE .\\ .\\ "O

t AT URE SPENT FUEL COOLING SYSTEM S F PORIFILATION EQUIPMENT M

8 m f0WER OCONEE NUCLEAR STATION FIGURE 9-5 242 T / 4 i

l I L IMS)DE OUTS)DE RB ) i RB ~ M @ 2 r.- TEST 57 M uuEt jg FROM CORE O FROM CORE j FLOOOtNG TAMKl 1 FLOODiMG TAMR 1 P e l ~ 'l l dk ? I N.r-N-- --t-9. ' A e REACTOR VELSV L pg UM)T 1 DECAY MEA'T RB uut FROM ToCA - ->4 - - - - REACTOR CootAMT SYSTEM SYSTEM pop,SAMpuuG FRoM FUEL TRAMSFER CAMAL UM\\T 1 1 I pg e M k d ! n 19 a_ rm REACTOR gg BLOG g EMERGEMCY' &f2 7' SUMP M O T E *. U Ml'

f:

iS D

== Fon SEE i 243 i

i ( BORATED 1 WATE R l ADDmN l - FROM SF PuR\\F) CAT)OM { l EQU)PMEMT l I DW q y I I JL l l To I i AE PUMPS f~ ~ ~ ' ' b"' i 1 BOR AT ED V LO l 3 k/D J L WATE R l STORAGE STORAGE TAK1W TAMW HEATER uu nT \\ Q H\\ / N V LO MAL % bwo To SF POR)FsCAviou l EQut PME.uT PSW LPSw p + TO HP COOLERS LP POMPS U M)T \\ 1 ! h T-- rb L_ _ _. g's ES ] -ES-l Xl ~ To RBS POMPS ~ 4 LOW PRESSURE INJECTION AND l 4 h e' DECAY HEAT REMOVAL SYSTEM 2 ARRAMGEMEMT REV: 4-1-67 hptm7g op gugT REDRAWN-SHARED EQUlPMENT (g g g ELIMitJATED F at OCONEE NUCLEAR STATION NOMtuC.tATORE, LEGEMD FIGURE 9-6 Flo,C-1 244 i 4 l 5

O m g 'N \\ ..nerAcron now \\ I N 1 N M \\, ___ < _1_n___n_ -cwn,en,,-, t t t t t Q acxaus oecenear (surwrt ex^riav) yy y \\ ~ x ,=m,,ru,nn ancour,ou '-.----. x x-s f M ~ m D4 g 1 N O %m .e v N s d1 {R N I s_->s R l' i o 8?Acrom Acw - a \\n AfW/M *D a ~ / x %m I --a l ____ E 01a l 10 to 1, ( _ .. - 243 m I c.

1 I L N _R g__-@. - "h t 3 k I __Q 6 4 I N t 2 Q _I ~~ I! I t l I I I i i l ~- N w ~ y N w ~~._ x -m-s m N N'N ~ ~ w N

' % m 4

0 0 Y '3 /O /0 C D me, s:conos i DECAY HEAT GENERATION VERSUS TIME AFTER SHUTDOWN mirente OCONEE NUCLEAR STATION FIGURE 9-7 246 g. ?I 4

e N a I i j LAKE KE0 WEE CONNECTED camat TO L.'tE RE0*LE w v - s [=t cer t c.w] AN C i "2 \\ T w._._ w....-, g= .v s c \\ c.~.OM sedT.A[ BOTT <etno cb. 5,.?W1me IN.ME SY8vctu L % \\T 1 r UNIT 3 (MT 2 vedf 7 l - D .Nu.i ~~e .c..c C C#dL** CONDirt condAp? i _ <. n_,., i _.., ] =u ]- .- _}_,~_._r, _- ~, ~. -== =v 'N' e u=u u= ~ 1 -u,N.os.=* rt,. wtin t anc.d ety c m t- '\\ UNIT 3 ve+T 2 vvt i 1=f,"fe'- es<,,..a o.3c

4 esc m cor co w p,p g l'

-.ro.cc q i I l b SCurM ATIC 4 AN t 247 l 4, -.- ~. - _ - _....., _ _, _ _ _, _, _, "-''-*'---*,p--

'l 2 \\ J T*9 OF DAM EL M ~6 t wr entzuc e p.vang 9Ju* Mr.;s On ] WTAst Msap WTC# EWL88GE*d[W DMMAAGE WWE Fin g pr2 F t mW-S* T 8*d A E C A8sS FUL L PCNC f t.800& L l } I, l l -ior.r.Em, Eut nGFart,' e W l ll l ' GCL ** 8E

a MM pesnaf eyw

w. 4.

A uAs rea s? 'ss. g> sM (f MAK ORA *?tMN EL TYW'Ef f

g 7Au EE .A,E w g, -( L e gt i sNTAuf PUMP t onwat OK H AuGE h ~ 70 LMI 40*LC - FWacANtv DT>eARGE COPOUIT 10 Y AL R ACI g-E ME RCE NCY 06CHAA[I WALVC MAR FLOOD Lt VFL EL 697-O* Nos* MAL 8hvf A LFVFL ft f.6Cft Shvf e BE0 EL 653*t SCHE MATIC ELE VATION No sc AL E ree+op

6. s t 4T M* L HA M LE E'a^Ai!

CONDENSER CIRCULATING WATER SYSTEM an sent reme OCONEE NUCLEAR STATION FIGURE 9-8 248 .I

e I s 4 1

~n c.
c.

.: N. 1 .~ i 61 f-- q r 1 n'; ,,. a O f (. N \\ l ) i / w ~., - v n,.., 5 i jej m.. 7,.. pe [ C, _l.,g -w. - $__.l i-V 6 l l l V l9 ,. ~ m. L A_ L } awl _j L l lVI l .e _ _I e j"9= _,,i C u 'f,"I.".. %..j.'_ l 'l V I l6 ! ~$ j -- _. A.. i. l 1-4-- .v i, _._ a.. I 'J E*% L",IC1','.," = oa. - o..e Aa*[R 9t h42!s i f 8rs.* (( > (Y-M -d =0 i A ~^} w~t n was t 4-W e. j g,*< 4

T,*, \\

2' ~ ~ ~ r / YM 9"' f

mres ( e sw

' -Qg 9 c+J'. 't" eaa e r w'r (-Vr -#- e +w m s'a'c. arta s. MM M & Jes S k ^ t, de G i g VW '" ="= s ~~~5 5 S [S5? ^ $Y? ' k ^ ^ ~ - .e4etgy4 -a [ =' m......._,........

.~

wm m. ma.v.t u-k k p,. g e,s.. & #4 C av / 44'f C(At AS \\ (*> n1,' h kh m V>g > w o fo un t. < < e punear4 _ .S Alb 49

- C C,

~ l~ l l I~ O I i i i i i i ? m 2 A _v _ ._r.._, __v._, a... _ _ 1. q _.n.,_ _._1._. l r l q p _. .e .e .e .e e ..e .c m a vee nat7'o* was eauto# uAt ne*ca um[ aso'on ( _j j dest PLA88 PLnam Amap m ese ptase w pues> punar N, \\ 9 = *elaxxawr, .s..+ a ct an.<. o.t m.<. = uo m eaa.<. ik { Con te ahtat t (m te (Org f a jar af t Cros.Em anc.tf J-g- g" r, sac.af f T 7-r- y-gy Iu;j ag A "4.pe m gamow se' tutA t.83 f f f t . m, I h ab p v-v-i e wt. CC e htLW Ct. Desa.no a aus GLOf. i 249 I~

4 F ) 1 [ -:L.. t }C .n'.!a .? % -r ..~ ,m -] g wu w

~: -

A l.T...,,_',,- ,--. 7l,., _. 4.I v l '_/. -] + ! re ;i_ - c...s .-- _i ; f N._. / ( v L- _, a i <x,..s m. -, i ~4 .-.~ i.:. ' ! K_/ I w{ i 4. m --./e+I_ I ,vj. L, o,. _ m, -3 L _.t __.i ,.,,,u. 4 A* l &i tass 1 e _i....,

L

.s.'e. / mon L -) v. - - - w h s' p p.; (a*-'s;- 1 x v .y _____w__ ., wme - .m<<,, ~., ,,.;,7. _ ma ~/- -..I ,~..- I ' N[N ex ~s- - b .~1 =.~ 1 Y=h I,_ E - - - - 1.- m ~_k' b as t(%s t 1 I. 1.. c- . =. /. 3 <.. um,. p ' fe w esdj ' g1

p'.

%h Lv t u o.a.* su e :im a m.- I t ? ? ? 4 -;s-- ,1;-- .-.y ,3, 35,. .,5-.l qgp ..,5,, f~~ .c .c c .c .i .c ~ o. l~ P & L P SERVICE WATER SYSTEMS m -,c n m e,- m

m..

,.~. w-.,...- ,.c- .a ~ .-.,...i,..~,,...--. u. ~ o u. 4 4 4 4 e 4 4 4 t t t t j $=:'-~~'" suit rom OCONEE NUCLEAR STATION ..<,,~..a., ,a ..%*.T, e.,.~..~.c=.'.--, FIGURE 9-9 e. LOfm.efe 18e T'S 250 a

k, l m o s.E== FN 1l*l l l IN o m m P -as ~,, - Ivl Ivl l l 1l Il M\\ l 6 o o o n ,q pr Pq jl ,,,t-;gg .....e...- ....m w OL COOLER 5 ot CJ0 Lips g mg, i i l I I 1 ip 1 1 I czwetsson RCw f R$ u. l COWmE SSOR 2 4> _Z5, _a r . -==ai-7E:; ~YEh. i=-F~--~'3--' _,. _ _ : - r=-= i FEl {9l' u_; -f El (3[_%$ kaiws avuoany i m I. 1 5tf as ii see as# WD se iP y sig.as C=r.on "g E mmon.To.n gt,,, 4 2 ..t omm A d I k h t 251 j

i \\ J i m o e .P 9 ... P,o M 1 F"~ c [I ~95 IOl L i o ...pyg'Tl;.g I-~~ .. D =.t ~6d'df x f k = u .c. wu w. o Omass oc ++ %.* > 1," "M ~ ,~ 1 , -_ a_ - _== ===========- -== =NW Ps _"g,3,=d ...n.., . _.,. _ = . = -

g. -,

RECIRCUIATED COOLING WATER SYSTEM ^_ ufjff.

yf1, Ml'&~n'u?

.s A WM T ** I OCONEE NUCLEAR STATION FIGURE 9-10 2_92_ e A

e <t PERSOM4EL MATCM UPPER PLENUM j N / CHAMBER ASSY FU[L ASSY HANDLNG TOOL \\ / j '\\ 7' SPENT FUEL STORAGE POOL / \\ [ CONTROL ROD ASSY t M*w, \\, .. / HANDLING TOOL / e , tw_ :_ _ :_,, 9 '. ~ s' y y, Si\\\\ k-'[ q --i \\~ Y { g. s .h s j .3 p1 ~~ ~ >= T*{I

i L-sQ. g, lI - '

4 ~ a 7 h. h l '\\ f gN e [ -. f i '/ b 1 s l \\ -NEW FUEL HANDLING RACKS -j \\ l - CORE BARREL AND FUEL LOADING AREA ' T)ERMAL $HIELD FLEL TRANSFER CANAL-., ASSEMBLY STORAGE CASK DECONTAMINATION AREA AREA EMERGENCY HATCH- - POLAR CRANE f' SPENT FUEL STORAGE RACK [ FUEL BUILDING CRANE I ff CARRIAGE ACTUATNG SYSTEM J n-t j UPPER PLENUM FUEL ASST HANDLING TOOL s CHAneER ASSEMBLY q f N / [ -SPENT FUEL ASSY STORAGE STANO \\\\ \\ / j ; STUO AND NUT 'N p i l a / 'f, i $

AGE adc" Ns

.g g/j l r "If$uREHEAo\\N m[ g j' j/ y\\.x\\ };cali - -- +~ C - $~A-m ___.m $1Ase / 4i 1 ! EQUIP HATCH \\ j l su yj - \\m i

i i

N N)r u 'm 7. m_ sn-r i "Trn rU 1(il-E in x x ( s \\ '_ sj -FUEL TRANS CARRIAGE y y ruEL TRANS TUBE =vE { K' FUEL TRANS TUBE CARRIAGE FUEL BASKET j \\ SPENT FUEL SHIPPING CASK- { bFUEL TRANS STATION FUEL TRANS TUBE COVER PLATE s 253 5 s

1 t 1 / ~~~ . p / i a g,Il {Q, N l ~ %. i _. ) Ll i I ( i (~' + A ~ 1 1 sEF-qJ k J t .'D tr

1 j

~ 1r_ _3(

w m

FUEL HANDLING SYSTEM bal i a-l O' OCONEE NUCLEA FIGURE 9-11 s ,-......,.-e.-,. r, r,- ,.,,v-vm c .n. + v.

a / / { E H C fN P g aas EL 833+0 EL. 822 + 0 rt eog + 3 FUEL HANDLING ARE A f 3+J e n H S DE A@ p l TA C EL 785 + 9 ST AT ION ' EL 775 + O h A & l R ADWASTE ARE A HI ( h b M C OUTSIDE AIR If p p INTAKE l GEb ~ = g p-- r 4 EL 822 + 0 3 ^ EL 809 + 3 L_ 2 i' 9& i, sl ,< m.e NON - RADIOACTIVE ARE A y fg=l______________i p NOTE [FGEND b SE FIG RE 9 -1 REV 5 25* LOUVER COOLING COIL TO 5HOW C uONITOR L H HE AT COIL C L I W5

t \\ wza P AC r= gg IC ',N -f P )U? SCI A:R H C l P L C C si qu w-- d h/ E ~ CONTROL ROOM P A C EL 822 + 0 + l US C ABLE ROOM EL 809 + 3

~

U CHAR E,_ l ,4______,,_,__M_. EQUIPMENT ROOM El 796 + 6 CONT ROL ARE A [ OOF EXH AUST ERS \\> / ( f& Y-r H C s P a 1 _ _ _ _ _ _. _ _ _ _ _. _ _ _ _ _ _ +..._ ) P -e-b OUTSIDE AIR ~~~ j w g /g g,- = ^ 4-INTAKE n C P e i [, t p, b TURBINE ARE A i d 'g [ [ M9?? f I I uia TURBINE & AUXILIARY BUILDING VENTIIATION SYSTEMS 7 OI ATIO N HC.NC E S OCONEE NUCLEAR STATION FIGURE 9-12 e f-w

O po TYPICAL FOR 6 ZONES T F 7 E2l '/ / P --+ - "?T!'E"3 O d I 4 e l l k V k If IF lt 4 ylL ING I I i 'r l T T T T ogC,ngg,E __jj ) l l l l l PO I I I I I _] _ _ _ _ _ _ _ _ _ _ _ = LEGEND NOTE E IfE C NOM ENCL ATU RE S IG 9. /, LOUVER HEAT Colt b d e COOLING Colt J 1 ADMINISTRATION BUILDING VENTILATION SYSTEM s outrow t OCONEE NUCLEAR STATION b FIGURE 9-13 257 l}}

ML19322B513

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