ML19319C256

Chapter 9 to Davis-Besse PSAR, Auxiliary & Emergency Sys. Includes Revisions 1-8

ML19319C256
Person / Time
Site:	Davis Besse
Issue date:	08/01/1969
From:	TOLEDO EDISON CO.
To:
References
NUDOCS 8002110720
Download: ML19319C256 (54)

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D-B TABLE OF CONTENTS Section Pm 9 AUXILIARY AND EMERGENCY SYSTEMS 9-1 9.1 MAKEUP AND PURIFICATION SYS'IEM 9-2 9.1.1 DESIGN BASES 9-2 9.1.1.1 General System Function 9-2 9.1.1.2 Letdown Coolers 9-2 9.1.1.3 Letdown control Valves 9-2 9.1.1.h Purification Demineralizer 9-2

,.1.1.5 Makeup Pumus 9-2 9.1.1.6 Seal Return Coolers 9-2 9.1.1.7 Makeup Tank 9-3 9 1.1.8 Filters 9-3 l 9 1.2 SYSTEM DESCRIPTION AND EVALUATION 9-3 9 1.2.1 Schematic Diagram 9-3 9.I.2.2 Performance Requirements 9 9.1.2.3 Mode of Operation 9-3 9.1.2.h Reliability Considerations 94 9 1.2.5 Codes and Standards 9-5 9 1.2.6 System Isolation 9-5 9 1.2.7 Leakage Considerations 9-5 9 1.2.8 Operational Limits 9-5 92 CHEMICAL ADDITION SYSTEM f-9 9 2.1 DESIGN BASES 9-9 9.2.1.1 General System Function 9-9 9.2.1.2 Boric Acid Mix Tank ,9-9

, 9.2.1.3 Boric Acid Addition Tank 9-9 9-1 9

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Section M 9 2.1.4 Boric Acid Pumps 9-9 9 2.1 5 Lithium Hydroxide Mix Tank 9-9 9.2.2 SYSTEM DESCRIPTION AND EVALUATION 9-9 9 2.2.1 Schematic Diagram and System Description 9-9 9 2.2.2 Performance Requirements 9-10 9.2.2.3 Mode of Operation 9-10 9 2.2.h Reliability considerations 9-10 9 2.2.5 codes and Standards 9-10 9 2.2.6 System Isolation 9-10 9.2.2 7 Leakage considerations 9-10 9.2.2.8 Operational Limits 9-10 9.3 COOLING WATER SYSTEMS 9-13 9.3.1 DESIGN BASES 9-13 _)

9.3.2 SYSTEM DESCRIPTION AND EVALUATION 9-13 9.3.2.1 condenser circulating Water System 9-13 3l9.3.2.2 Service Water System 9-13 9 3.2.3 component cooling Water System 9-lh 9 3.2.h Turbine Plant Recirculated Cooling Water 9-15 System SPENT FUEL FOOL COOLING SYSTEM 9-16 3 ( 9.4 9.h.1 DESIGN BASES 9-16 9.h.2 SYSTEM DESCRIPTION AND EVALUATION 9-16 .

9.h.3 SYSTEM RELIABILITY 9-16 9.h.h CODES AND STANDARDS 9-17 9.'h.5 TEST AND INSPECTION 9-17 95 DECAY HEAT REMOVAL SYSTEM _ 9-18 .

I 9-18 9.5.1 DESIGN BASES gg

's 9-11 ,

Amendment No. 3 E '

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D-B Section Page

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9.5.1.1 General System Function 9-18 9 5 1.2 recay Heat Removal Pumps 9-18 9.5.1.3 Decay Heat Removal Coolers 9-18 9.5.2 SYSTEM DESCRIPTION AND EVALUATION 9-18 9.5.2.1 Schematic Diagram 9-18 9.5.2.2 Performance Requirements 9-18 9.5.2.3 Mode of Oreration 9-18 9 5.2.h Reliability Considerations 9-19 ,

9 5.2.5 Codes and Standards 9-19 9 5.2.6 System Isolation 9-19 9.5.2.7 Leakage Considerations 9-19 9 5.2.8 Failure Considerations 9-19 9.6 FUEL HANDLING SYSTEM 9-21 9.6.1 DESIGN BASES , 9-21 9.6.2 SYSTEM DESCRIPTION AND OPERATION 9-21~

9.6.2.1 Receiving and Storing Fuel 9-21 9.6.2.2 Loading and Removing Fuel 9-21 l

9.6.2.3 Storage of Spent Fuel 9-23 l l

9.6.2.h Safety Provisions 9-2h 9.6.2.5 Operational Limits 9-25 9.6.2.6 Miscellaneous Fuel Handling Eauitment 9-25 9.7 SAMDLING SYSTEM 9-26 9.7.1 DESIGN BASES 9-26 9

7.2 DESCRIPTION

AND OPERATION 9-26 9.8 STATION VENTILATION SYSTEMS 9-27 s 9.8.1 DESIGN BASES 9-27 0238 9-111

D-B Section Page

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9.8.2 SYSTE4 DESCRIPTION hND EVALUATION 9-27 9.8.2.1 Containment Vessel Ventilation 9-27 7 19.8.2.2 Emergency ventilation 9-27 9.6.2.3 Auxiliary Building Ventilation 9-2I 9.8.2.h Turbine Building Ventilation 9-2b 9.8.2.5 Station Heating 9-28 9.8.2.6 System Relictility 9-20 99 INSTRLHENT_AND, SERVICE AIR SYSTEM 9-20 9.9.1 DESIGN BASES 9-29 9.

9.2 DESCRIPTION

AND GPERATION 9-29 9.9.3 TESTS AND INSPECTIONS 9-29 9.10 AUXILIARY FEEDWATER SYS'IU4 9-30 9.10.1 DESIGN LASES 9-30 )

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10.2 DESCRIPTION

AND OPERATION 9-30 9.11 FIRE PROTECTION SYSTEM 9-32 9.11.1 DESIGN BASES 9-32 9.11.2 SYSTEM DESCRIPTION AND GPERATION 9-32 0240 x)

Amendment No. 7 ,

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D-B LIST OF TABLES

-- s . Table Pagg, 9-1 Makeup and Purification System Performance Data 9-6 9-2 Makeup and Purification System Equipment Data 9-7 9-3 Chemical Addition System Equipment Data 9-11 9k Decay Heat Removal System Performance Data 9-20 9-5 Decay Heat Removal System Equipment Data 9-20

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LIST OF FIGURES (At Rear of Section)

Figure No.

Title 9-1 Drawing Symbols 9

9-2 Makeup and Furification System 9-3 chemical Addition System 9-4 condenser circulating Water System 9-5 Service cooling Water System 9-6 cceponent coo.'ing Water System 9-7 Spent Fuel Fool Cooling System 9-8 Decay Heat Removal System 9-9 Fuel Handling System

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9-10 Sempling System 9-11 Station Ventilation System; 9-12 Flow Diagram, Instrumentation and Service Air Systems 9-13 Flow Diagram, Auxiliary Feedwater System 9-lh Decay Heat Generation Versus Time After Shutdown 9

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,m 9 AUXILIARY AND EMENGENCY SYSTEMS s_.

The auxiliary systems required to support the reactor coolant system during normal operation of the Davis-Besse Station are described in the following sections.

Some of these systems are described in detail in Section 6 since they serve as engi-neered safety features. The information in this section deals primiarily with the functions served during normal operation.

Most of the components within these systems are located within the auxiliary building. Those systems with connecting piping between the containment and the auxiliary building are equipped with containment isolation valves as described in 5.1.5 The codes and standards used, as appropriate, in the design, fabrication, and testing of ecmponents, valves, and piping are as follows:

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

b. ASME Boiler and Pressure Vessel Code,Section III, 'Tuclear Vessels.

c. ASME Boiler and Pressure Vessel Code, Sectien VIII, Pressure Vessels and ASME Nuclear Case Interpretations.

,i . ASME Boiler and Pressure Vessel Code,Section IX, Welding Qualifica-

, tions.

e. Standards of the American Society for Testing and Materials. (ASTM).

f. ANSI, B31.1.0 - 1967, Power Piping 5

g. ANSI, B31.7 - 1969, Nuclear Power Piping'

h. Standards of the American Institute of Electrical and Electronics Engineers. (IEEE).

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Standards of the National Electrical Manufacturers Association. (NERA)

J. Hydraulic Institute Standards.

k. Standards of Tubular Exchanger Manufacturers Association.

1. Air Moving and Conditioning Association Codes. (AMCA)

m. ANSI, B96.1, Aluminum Tanks.

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n. American Gear Manufacturers Association Standards. (AGMA)

o. American Water Works Association Standards. (AWWA).

p. Draft ASME Code for Pumps and Valves for Nuclear Power November 1968.

q. National Fire Protection Association (NFPA) Standards.

0?43 9-1 Amendment No. 5

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r. ANSI Standard C 50.20 - Test Code for Polyphase Induction Motors

- and Generators.

s. ANSI Standard C 50.2 for Alternating Current. Motors, Induction Machines in General and Universal Motors.

t. Heating, Ventilating, and Air Conditioning Guide; American Society of Heating, Refrigerating and Air Conditioning Engineers,

u. The pressure-containing parts of all engineered safety features systems pu=ps of stainless steel caterial vill be liquid penetrant-examined in accordance with Appendix VIII of Section VIII of the ASME Code. The pressure-containing velds of all engineered safety feature systems pumps will be radiographically examined in accordance with Paragraph UW-51 of Section VIII of the ASME Code,

v. Valves and piping vill be designed and fabricated to meet the requirements of ANSI B16.5 or MSS SP-66, and ANSI B31.1.0.

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To aid in reviewing the system drawings, a standard set of symbols and abbreviations has been used and is seme ized in Figure 9-1.

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9-la~ Amendement No. 5

D-B 9.1 MAKEUP AND PURIFICATION SYSTEM 9.1.1 DESIGN BASES

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9.1.1.1 General System Function 8 l The system shown in Figure 9-2 supplies the reactor coolant system with fill and operational makeup water; circulates seal water for the reactor coolant pumps; receives, purifies, and recirculates reactor coolant system letdown to provide water quality and reactor coolant boric acid concentration control; and 3 accommodates temporary changes in the required reactor coolant inventory. The makeup and purification system performs no emergency function.

9.1.1. 2 Letdown Coolers The letdown coolers cool the letdown flow from reactor coolant te=perature to a temperature suitable for demineralization and injection to the reactor cool-ant pump seals. The maximum letdown flow is required for startup at 100 F/hr late h core life wherein the reactor coolant boron concentration is reduced by an amount corresponding to the change due to moderator temperature reactiv-ity deficit. Heat in the letdown coolers is rejected to the component cooling system.

9.1.1.3 Letdown Control Letdown flow is established by use of a block orifice which is sized for the normal purification rate. However, during a startup or shutdown phase when the reactor coolant system is at low pressure, the desired letdown flow is maintained by the supplemental use of a parallel control valve along with the i orifice. Both flow paths are also used when high letdown flow is required, e.g., reactor coolant boron concentration adjustment.

9.1.1. 4 Purification Demineralizer The letdown flow is passed through the purification demineralizer to remove rcactor coolant impurities other than boron. The purification letdown flow to maintain the reactor coolant water quality is equal to one reactor coolant vol-une per 2h hours. Each purification demineralizer is sized for the letdown flow rate as required for boron concentration control. Refer to Table 11-3 for the maximum anticipated equilibrium fission product accumulation in the r: actor coolant.

9.1.1. 5 Makeup Pumos 8lThemakeuppumpsaredesignedtoreturntheletdownflowtothereactorc_oolant system and supply the seal water flow to the reactor coolant pumps. The normal flow capacity is equal to the normal makeup 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 Seal Return Coolers The seal return coolers are sized to remove the heat added by the makeup pump and the heat picked up in passage through.the reactor coolant pump seals. Heat '

from these coolers is rejected to *tlie.'tomponent cooling system. /

624s Amendment No.8 9-2

D-B 9.1.1.7 Makeup Tank 311s tank rc:ves as a surge vessel for the makeup system and as a receiver for the letdown flow, chemical addition, and outside makeup; it also accommodates temporary changes in reactor coolant system volume. The volume of the tank is such that the useful tank volume, in conjunction with the pressurizer, will accommodate the expected e cpansion and contraction of the reactor coolant system during normal power transients. Normal power transients are the transients associated with reactor operation between 15% and 100% power.

9.1.1.8 Filters The filters will prevent the entry of resin fines from the demineralizer and other particulates from the radioactive waste treatment system, chemical addi-tion system, and the station demineralized water suppRy into the makeup tank and seals of the reactor coolant pumps. .

9.1. 2 SYSTEM DESCRIPTION AND EVAWATION 9.1. 2.1 Schematic Diagram The makeup and purification system is shown in Figure 9-2.

9.1. 2. 2 Performance Recuirements Table 9-1 lists the system performance requirements.

9.1. 2. 3 Mode of ooeration During normal operation of the reactor coolant system, one makeup pump contin-uously supplies high-pressure water from the makeup tank to the seals of each of the reactor coolant pumps, and to a makeup line which is connected to the reactor inlet line by a high pressure injection line. This line is the cnly interconnection between the makeup system and the high pressure injection system.

3 {8 Makeup flow to the reactor roolant system is regulated by the makeup control valve, which operates on si,nals l from the liquid level controller of the reac-tor coolant system pressurizer. Control valves in the injection lines to the punp seals automatically maintain the desired flow rate to the seals. A part 3 of the water supplied to the seals leaks into the reactor coolant system.

The remainder returns to the makeup tank after passing through one of the two seal return coolers.

Seal water inleakage to the reactor coolant system requires a continuous let-down of reactor coolant to maintain the desired coolant inventory. In addi-tion, treatment of reactor coolant is required for removal of impurities and boric acid from the reactor coolant and to accommodate volume changes in the c reactor coolant system during changes in power level. Reactor coolant is re-moved from one of the reactor inlet lines, cooled during passage through one of the letdown coolers, passed from the containment vessel through a reactor building isolation valve, reduced in pressure during flow through the letdown flow control station, and then passed through one purification demineralizer to a three-way valve which. directs the coolant either to the makeup tank or '

to the radioactive waste disposal system.

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Uormally, the three-way valve is positioned to direct the letdown flow to the makeup tank. If the boric acid concentration in the reactor coolant is to be reduced, the three-way valve is positioned to divert the letdown flow to the radioactive waste disposal system. Boric acid removal is accomplished in the radioactive waste disposal system either by directing the letdown flow through a deborating demineraliser with the effluent returned directly to the makeup 3 tank, or by directing the letdown flow to a clean vaste receiver tank.

The level in the makeup tank is maintained with deborated water from storage or with demineralized water frcm the plant demineralized water storage tank.

The quantity of unborated water received is msasured and limited by inline in-strumentation and interlocked with shim rod position controls.

The makeup tank also receives chemicals for addition to the reactor coolant.

A hydrogen overpressure maintained in the makeup tank supplies the hydrogen added to the reactor coolant.

System control is acco=plished remotely from the control room with the excep-tion of the seal return coolers. The letdown flow rate is established by the block orifice during normal operation, but may be increased by opening the let-down control valva. The spare purification demineralizer can be placed in ser-vice by remote positioning of the demineralizer isolation valves. Diverting the letdown flow to the radioactive waste treatment system is accomplished by remote positioning of the three-way valve and the vaves in the radioactive waste disposal system. The control valves in the injection lines to the reactor 3 coolant pu=p seals are auto =atically controlled by a flow rate controller installed ,

in each seal injection line to maintain the desired flow rate t6 the seals. The pressurizer makeup control valve is automatically controlled by the pressurizer level controller. The letdown control valve is designed for letdown flow rate control at reduced reactor coolant system pressure. The makeup pumps are con-trolled remotely.

. 3 (Deleted) 9.1. 2. h Reliability Considerations This system provider essential functions for the normal operation of the unit.

3 l Redundant components and flow paths have beta provided to improve system reliability.

In addition to the letdown orifice the sy; tem has two full-capacity control valves in parallel with the orifice. One of these control valves is manually operated and one is remotely operated.

Two purification demineralizers are provided. One is normally in use while the 1 other is a spare.

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Amendment No. 3 ,

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D-B There are two makeup m s, each capable of supplying the required reactor coolant pump seal and makeup flow. One is normally in operation while N another is in standby status to be used as needed. 3 There are two letdown coolers and two seal return coolers. One of each is in operation normally, while the other is a spare. Similarly, there are two make-up filters provided.

9.1. 2. 5 codes and standards The equipment in this system will be designed to applicable codes or standards as noted in Table 9-2 9.1. 2. 6 System Isolation The letdown line and the reactor coolant me-; seal supply and return lines penetrate the containment vessel. These lines contain power-operated isola-tion valves which are automatically closed by the containment emergency isolation signal.

(Deleted) 3 9.1. 2. 7 Leakage Consideration ,

Design and installation of the components and piping to the Makeup and Purifica- l 8 tion System considers the radioactive service of this system. Except where flanged connections have been installed for ease of maintenance, the system is l8 an all-welded system. Valves have double packing with provisions for stem leak-off connections.

9.1. 2. 8 operational Limits Alarms or interlocks are provided to limit variables or conditions of. operation that might affect system or station safety. These variables or conditions of operation which are limited are as follows:

Makeup Tank Level The low-level alarm point signals an alarm and also signals the three-way valve to switch from bleed operation to makeup operation. A 1pss of level in this tank could cause a loss of prime to the makeup pumps, thus causing loss of make-up flow which could have a serious effect on the reactor coolant system if it occurred during cooldown or when makeup for contraction of the reactor coolant system is required.

_ Letdown Line Temoerature A high-temperature alarm signals a high temperature in the letdown line down-stream of the letdown coolers and also signals the letdown isolation l3

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valve to close. Hot reactor coolant, if allowed to pass through the purifica- '

tion demineralizer, could damage the demineralizer resins.

Dilution Control To prevent an inadvertent excessive dilution of the reactor coolant boric acid concentration, two safety measures are provided to limit the normal process of deboration or dilution of the reactor coolant. The normal process of debora-tion of the reactor coolant has previously been defined in Section 9.1.2.3 as the feed and bleed method or the deborating demineralizer method. These safety measures are as follows:

a. The process of normal deboration of the reactor coolant by cither method cannot start unless specific control rod groups are withdrawn to a certain point which allows for deboration. This control rod group position interlock either permits or prohibits continuous dilu-tion depending upon the control rod group position. Because of this interlock, the denineralized water makeup valve and the three-way valve can be operated simultaneously only when the control rod group is withdrawn to a preset position. The demineralized water makeup valve is automatically closed, and the three-way valve position is automatically changed when the rods have inserted to a preset posi-tion.

b. The second safety measure consists of closing the demineralized water makeup valve and switching the three-way valve position when the /

makeup flow has integrated to a preset valve.

Initiation of dilution must be by the operator, and the operator can terminate dilution at any time.

Table 9-1 Makeup and Purification System Performance Data Letdown Flow (Cold), gpm 45-140 Total Flow to Each Reactor Coolant Pump Seal, gpm 10-12 Seal Inleakage to Reactor Coolant System per Reactor 8-10 Coolant Pump, gpm Injection Pressure to Reacter Coolant Pump Seals, 2,190 psig Temperature to Reactor Coolant Pump Seals, F 120 Purification Letdown Fluid Terrh ur;, F 120 Makeup Tank Normal Operating Pressure, psig 15-35 Makeup Tank Volume Egtveen Minimum and Maximum 250 Operating Levels, ft Amendment No. 3 9-6 )

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D-B Table 9-2 Makeup and Purification System Em11pment Data

. .P.

C;' Makeup Pump Quantity 2 Type Centrifugal, Mechanical Seal Rated Capacity, gpm 80 {6 Rated Head, ft at sp gr = 1 5800 Motcr Horsepower 400 Pump Material SS Wetted Parts Design Pressure, psig 3050 3 Design Temperatu"e, F 200 Makeup Filter Quantity 2 Capacity, gpm 80 Material SS Design Pressure, psig 150 -

Design Temperature, F 200 Code ASME Section III-C -

Letdown Cooler Quantity 2 Full Capacity Type Shell and Tube Heat Transferred, Btt/hr 16.2 x lg6 i Letdown Flow, lb/hr 3.5 x 10 Letdown Temperature Change, F- 557 to 120 Material,Shell/ Tube CS/SS Design Pressure, psig 2,500 Design Temperature, F 600 Code ASME Section III-C and VIII Seal Return Cooler Quantity 2 Full Capacity Type ShellangTube Heat Transferred, Stu/hr 7.2 x 10 Seal Return Flow, li/h 3000 Seal Peturn Te=perature Change, F lkh to 120 3 1 Material,Shell/ Tube CS/SS Design Pressure, psig 150 Cooling Water Flow, lb/hr 3000 13 Code ASME Section III-C and VIII Makeup Tank ,

1 Quantity Volume, ft 3 600 Design Pressure, psig 100 resign Temperature, F 200 Material SS j Code ASME Section III-C 9-7 Anendment No. 6-0250 ,

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D-B Table 9-2 (Cont'd) ~' ,

Purification Demineralizer Quantity 2 Type Mixed Bed, Boric Acid Saturated Material SS 1 l'ResinVolume,ft 3 So Flow, gpm-- 70 Vessel. Design Pressure, psig 150 Vessel Design Temperature, F 200 Code' ASME Section III-C

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D-B 9.2 CHEMICAL ADDITION SYSTEM

. 9.2.1 DESIGN BASES 9.2.1.1 General System Function Chemical addition operations are required to alter the concentration of varicus chemicals in the reactor coolant and auxiliary systems. The system shown in l8 Figure 9-3 is designed to add boric acid to the reactor coolant system for re-activity control (see Table 3-6 and Figure 3-1), lithium hydroxide for pH con-trol, and hydrazine for oxygen control.

9 2.1.2 Boric Acid Mix Tank A single boric acid mix tank is provided as a source of concentrated boric acid solution. Heaters in the tank maintain the temperature above that required to insure solubility of the boric acid. Transfer lines vill be electrically traced with dual heating circuits. This tank supplies the boric acid addition tank by gravity flow. ,

9.2.1.3 Boric Acid Addition Tanks Two boric acid addition tauks are provided for storage of boric acid solution in sufficent quantity. One tank will provide sufficient storage to accomplish a cold 3 shutdown. These tanks are used to store, for eventual use, the boric acid from the boric acid mix tank and also the processed boric acid from the radvaste system.

9.2.1.h Boric Acid Pumes Two boric acid pumps are provided to facilitate transfer of the concentrated boric acid solution frcm the beric acid addition tank to the borated water storage tank, the makeup tank, or the spent fuel storage pool. The pumps are sized so that when both are operating, one ecmplete charge of concentrated boric acid solution from the boric acid addition tank may be injected into the reactor coolant system in 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

l6 9.2.1.5 Lithium Hydroxide Mix Tank The tank volume contains a sufficient amount of LiOH for addition to the reactor coolant system so that a concentration of 0 5-2.0 ppm of Li7 can be maintained. 6 (See Table h-2) 9.2.2 SYSTEM DESCRIPTION AND EVALUATION 9.2.2.1 Schematic Diagram and System Description Figure 9-3 is a schematic diagram illustrating the features of the system. The system (except boric acid pumps) is operated from local controls. Two beric <

acid pumps, connected in parallel, take suction from the boric acid addition tank and discharge to either the spent fuel storage pool, borated water storage tank, or upstream of the makeup tank. At the end cf co.re life, both boric acid pumps are required to raise the reactor coolant system boren concentra-tion from the minimum end-of-life concentration to the refu.eling concentration in approximately 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The boric acid mix tank has a mechanical mixing de-vice and a heating unit. -

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ozsz kN4 g_9 Amendment No. 8

D-B The lithium hydre ude equipment consists of a mix tank, a single positive dis-placement pump, and connecting piping. The pump discharges upstream of the f makeup tank.

A hydrazine drum is connected to a positive displacement pump, which discharges to a line leading to the makeup tank. A nitrogen blanket is used to displace the hydrazine as it is removed from the drum.

9.2.2.2 Performance Pec_uirements This system permits chemical addition to the reactor coolant system and the reactor tuxiliary systems. Water qualities to be maintained are listed in Tables L-2 and h-6.

9.2.2.3 Mode of operation During normal operation, this system delivers the following chemicals:

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

b. Lithium hydroxide to the makeup tank.

c. Hydrazine to the makeup tank.

9.2.2.4 Reliability Censiderations i

The system is not required to f anction during an emergency, nor is it required 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.

9.2.2.5 codes and standards Each co=ponent of this system vill be designed to the code or standard, as ap-plicalle, as noted in Table 9-3.

9.2.2.6 System Isclatien H containment isolation is required of this syster since its boundaries do not 2

l penetrate the containment.

9.2.2.7 Leakage considerations This system delivers additives to the spent fuel storage pool and the makeup, and borated wate" stcrage tynks. Backflow from the tanks'to the positive dis-placement pu=ps is prevented by check valves and normally closed shutoff valvas between them. Additives to the spent fuel storage pool are delivered above the water level, thereby preventing backflow.

9.2.2.8 operational Limits

' s The boric acid mix and addition tanks are to be maintained at an average tem- )

perature of 95 F to maintain a boric acid in solution at a concentration of 0253 Amendment No. 2 9-10

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D-B 7 per cent. The a=ount of solution maintained in the boric acid addition tank

__, is to be sufficient to borate the reactor coolant system for cold shutdown near the end of core life.

Table 9-3 Chemical Addition System Ecuip=ent Data Tanks Beric Acid Mix Tank Quantity 1 Type Vertical Cylindrical Volume, ft 3 130 Design Pressure, psig At=ospheric Design Temperature, F 200 Material SS Code ANSI 396.1 3-Boric Acid Addition Tank quantity 2 Type Hori:orr,al Cylindrical 3 Volu=e, ft3 1,000 Design Pressure, psig *t=ospheric Design Temperature, F iOO Material SS Lithium Hydroxide Mix Tank Quantity 1 Type Vertical Cylindrical Volume, gal 50 Design Pressure, psig Atmospheric Design Temperature, F 150 Material SS Hydrazine Drums Quantity 1 Type Std Cc==ercial 55 gal Drums O

Pu=rs Boric Acid Pu=p Quantity 2 Type Diaphrag=, Variable Stroke Capacity, gpm 0 - 25 ]

Head, ft 1h0 6

.. Design Pressure, psig 100 Design Temperature, F 200 Material SS 0254 9-11 A=endment No. 6 l

D-B Table 9-3 (Cont'd) ]

Lithium Hydroxide Pump Quantity 1 Type Diaphrag=, Variable Stroke Capacity, gph 0 - 10

, 1 l Head, psi 100 Design Pressure, psig 150 Design Te perature, F 200 Material SS Hydrazine Pump Quantity 1 Type Diaphrag=, Variable Stroke Capacity, gph 0 - 10 1 Head, psi 100 l

Design Pressure, psig 150 Design Temperature, F 200 Material SS 0255 J'

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93 COOLING WATER SYSTD4S 9 3.1 DESIGN BASES The cooling water systems remove heat from the station equipment to permit a sustained operation and safe shutdown of the station. The condenser is cooled by the condenser circulating water system described in Section 9 3.2.1.

A station service water system furnishes cooling water to the component cool-ing water system used to cool nuclear plant and safety features equipment, and g to the turbine recirculated cooling water system (described in Section 9 3.2.4) which cools turbine building cooling and sealing water.

9 3.2 SYSTD4 DESCRIPTION AND EVALUATION 9 3.2.1 condenser circulating Water System I

The condenser circulating water system is sized to handle the maximum (VWO Load) condenser heat loads and consists of a closed system utilizing a hyperbolic natural draft cooling tower and the associated circulating vater pumps, piping and valves. Fill and makeup water is taken from Lake Erie through the intake water system and intake structure. Four circulating water pumps with suction lines from the base of the cooling tower pump through the condensers and back to the cooling tower. The circulating water. system is shown in Figure 9 h.

8 Cooling tower makeup (to account for evaporative losses, entrainment losses, and blowdown) is provided by two cooling tower makeup pumI.s located in the intake structure. To help maintain the desired purity of the circulating vater, provisions are included for controlled blevdown of the tower to the station dischargc eater system.

Circulating water piping vill be partly concrete and underground, and partly steel piping. The condensers as a minimum vill be in general accord to Heat Exchange Institute Standards for condensers.

9 3.2.2 Service Water System l2 The service water system as shown in Figure 9-5 takes water fr m the pump suction pit after the travelling screens. This system suppliec cooling water to the component cooling water system, normal and emergency containment cool-era, the turbine plant cooling water systems, and serves as a backup suction source to the auxiliary feed pumps. 3 8

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0256 Amendment No. 8 9-13

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D' Three service water pumps are located in the pump house to serve this system. ,

8l3 ITwopu=psareusedduringnormaloperation. A single pump will take care of l the essential cooling requirements following an accident.

Following a loss-of-coolant accident, loss of electric power and use of emergency diesels, the supply to the turbine plant recirculated cooling water system is automatically valved off, with all the flow then being routed

- to the engineered safety features, auxiliaries.

2 l In the event of the loss of outside power, the two emergency diesel generator s each automatically energizes one service water pump.

The intake structure, the service water pumps, and the emergency containment coolers are designed to seismic Class I standards.

Adequate redundancy in equipment and piping is provided to guard against any single equipnent or pipe failure.

2 l The station service water piping vill be designed to ANSI E317 Class III.

9 3.2.3 component cooling water system A schematic description of the component cooling water system is shown on Figure 9-6. This system is a closed loop system which provides cooling y water to the nuclear and engineered safety features systems and also acts as an intermediate barrier between the radioactive system and the service cooling water 2 system. The system consists of three circulating pu=ps, three heat exchangers, a surge tank, associated valves , piping, instrumentation, and controls.

One ccaponent cooling water pump and one heat exchanger are required during normal plant operation as well as following a loss-of-coolant accident.

2 Two component cooling pumps and two heat exchangers are utilized to remove the decay and sensible heat during normal reactor shutdown and refueling.

All the components in this system are missile protected and designed for seismic Class I standards.

2 l In the event of the loss of outside power the two emergency diesel generators vill each energize one pump.

The surge tank allows for expansion of the system and provides sufficient NPSH for the ccanponent cooling water pumps. Make up is provided frce the deminer-alized water source to maintain surge tank level. Provisions will be made for 3 the addition of a chemical for corrosion and pH control.

The operation of this system is monitored with the following instruments:

8l a. Temperature indicators on the inlet'and outlet lines '

for the component coolitig heat exchangers and alarms ,

in the outlet line.

8 b. Pressure indicators on the lines between the pumps and .

the heat exchangers.

.]

8l c. Level alarms on the surge tank.

• Q'2,$

Amendment No. 8 9-lh

D--B

(.

d. Temperature indicators located on the outlet lines l8 of the components being cooled.

e. Radiation monitors on the return line to the component cooling water pumps.

The valves used in this system are of standard construction and although the component cooling water system is not normally radioactive, velded construc-tion is used,where possible, to minimize the possibility of leakage. All piping in this system is carbon steel.

The piping is so mi w ed that following a less-of-coolant accident the supply to the non-critical comI.onents can be valved off.

Complete redundance is provided in the piping to the critical equipnents. Two full capacity headers are provided for the component cooling water system. .

The component cooling water piping vill be designed to ANSI B31 7 Class III. 3 It vill be of carbon steel material.

9 3.2.h Turbine Plant Recirculated Cooling Water System This system vill furnish purified and treated cooling water to turbine and pump oil coolers, various pump seals, generator hydrogen equipment auxiliaries including generator hydrogen coolers and stator liquid cooler, isolated phase bus, air compressor jackets and coolers, and turbine plant sample coolers.

The system vill contain a low level supply and drain tank, two pumps, two heat exchangers, and a high level supply tank and associated piping. The high level supply tank will supply vcter to the turbine plant ecmponents needing sealing or cooling. The pumps will maintain the controlled level in the high level supply tank. The heat exchangers will be cooled by station service water.

g The Engi?.eered Safety Features equipment vill not be dependent on this turbine plant cooling water system.

The piping vill be designed to ANSI B31.1.0 power piping code.

3 0258 9-15 Amendment No. 8 I

D-B 9.h SPENT FUEL POOL COOLING SYSTEM

);

9.h.1 DESIGN BASES The spent fuel pool cooling system is designed to .aintain the borated spent fuel pool water at 100 F, with heat load based on removing decay heat from 1/3 core which is assu=ed to have undergone infinite irradiation, and to have r

been cooled for 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />.

The spent fuel pool is, however, designed for a total storage capacity of 11/3 core, plus 19 spare locations. The cooling capacity for this additional core vill be provided, through temporary connections, by the decay heat removal system should it be required.

9.k.2 SYSTEM DESCRIPTION & EVALUATION 1 lThespentfuelpoolcoolingsystemisshowninFigure9-7 It consists of two half capacity recirculation pumps and two half capacity heat exchangers ,

associated valves, piping and instruments. A bypass system consisting of a filter and a decinerali::er is also proviaed for purification of both spent 1 l fuel pool water and the contents of the borated water storage tank (after it has been used in the refueling pool).

During normal operation, when 1/3 core is freshly stored in the pool, both pumps and both heat exchangers are in continuous operation; as the decay heat emitted by the spent fuel decreases or during the colder weather when the service cooling water temperature may be lov, one pump and one heat exchanger 3 can be shutdown. The spent fuel water temperature is normally maintained at /

100 F and will be limited to 150 F maximum.

During cold shutdown and refueling conditions, the reactor refueling cavity is filled with water fro = the berated water storage tank. The reactor refueling 8 1 cavity water can be cooled by the decay heat coolers and purified as needed by the spent fuel pool cleanup system or the make-up and purification system.

The decay heat Tiumps empty the refueling cavity water into the borated water

, storage tank after refueling is co=pleted. The fuel tilting mechanism pits can be filled and e=ptied through the decay heat cooling system piping interconnee-1 tions and decay heat pumps.

Surface skimmers are provided in the spent fuel pool to facilithte removal of accumulated material from the surface of the pool.

9.k.3 SYSTEM RELIABILITY The two motor driven pumps and two heat exchangers vill provide sufficient heat removal capability in the system to keep the spent fuel pool water temperature at an acceptable level.

The most serious failure of the system would be the co=plete loss of the spent fuel pool water. To protect against this possibility, all connections to the pool enter near to or above the water level, so the pool could not be gravity drained through leaking valves .or piping. . .

s Amendment,No. 8 9-16 0259 u

D-B m

In the event that a leaking fuel assembly is transferred from the refueling l -

pool to the spent fuel pool, a small quantity of fission products may enter the spent fuel pool water. The bypass purification system is provided for removing these fission products and other contaminants from the water.

9.h.k CODES AND STANDARDS The equipment in this system will be designed to applicable codes and standards tabulated on page 9-1.

9.4.5 TESTING AND INSPECTION Each component is inspected and cleaned prior to installation into the system.

Demineralized water vill be used to flush the system.

Instn1ments will be calibrated during testing. Alarm functions will be checked .

for operability and limits during preoperational testing.

The system will be operated and tested initially with regard to flow paths, flow capacity and mechanical operability. At least one pump of each type vill be tested to demonstrate head and capacity.

Data vill be taken periodically during normal station operation to confirm heat

~

transfer capabilities and puriff cation efficiency.

i.

, 8

( 0260

., un

-!,; w -

9-17

D-B 95 DECAY HEAT RDOVAL SYSTD4 9 5.1 DESIGN BASES 9 5.1.1 General System Function 1 l The normal function of this syste= as shown by Figure 9-8 is to remove reactor decay heat during the latter stages of cooldown and maintain reactor coolant temperature during refueling. The emergency functions of this syste= are de-scribed in 6.1.

1 l 1s given in Figure 9-lb.A curve of decay heat generation versus time after shutdown 9

9.5.1.2 Decay Heat Pumns The decay heat pumps, during shutdown, circulate the reactor coolant from one reactor outlet line through the decay heat coolers and return it to the reactor injection nozzles. The design flow is that required to cool the reactor coolant syste= from 280 F to lho F in lh hours. (The steam generators are used i==ediately after shutdown to reduce the reactor coolant system from operating te=perature to 280 F in a 6-hour period.)

9 5.1.3 Decay Heat Coolers The decay heat coolers, during shutdown, remove the decay heat from the circulated reactor coolant. Two coolers and two pumps will reduce the reactor coolant tem-perature to lho F in lh hours after the reactor coolant temperature reaches 280 F.

)

9.5.2 SYSTD4 DESCRIPTION AND EVALUATION 9 5.2.1 Schematie Diaera=

1 ,

The decay heat removal syste= is shown schematically in Figure 9-8.

9.5.2.2 rerformance Recuirements Tables 9-h and 9-5 at the end of this subsection list system performance data and design data for individual components.

9 5.2.3 Mode of oeeration Two pumps and two coolers perfor= the decay heat cooling function. After the steamcooling heat generators have reduced the reactor coolant temperature to 280 F, decay is initiated.

Normally two pumps vill take suction from the reac-tor outlet line and discharge through the coolers into the reactor vessel. If only one'pu=p or one cooler is available, the reactor coolant te=perature is reduced at a lower rate. The equipment utiliz,ed for decay heat ecoling is also used for low-pressure injection into the core during accident conditions.

During refueling the decay heat from the reactor core is rejected to the decay heat coolers in the same =anner as it is during cooldovn to 1h0 F. At the be-1lginningoftherefuelingperiodbothcoolersandbothpumpsarerequiredto maintain lho F in the core and refueling canal. Later, as core decay heat decreases, one cooler.and pump can maintain the required lho F.

9-18 Og

D-B

(} The refueling canal is initially fil] ad by switching the suction of the 1

\

decay heat pumps from the reactor outlet to the borated water storage tank.

As the borated water passes through the reactor vessel into the refueling canal it absorbs the decay heat from the core. '4 hen the refueling canal is filled, y suction to the pumps is switched back to the reactor outlet pipe. The refuel-ing canal is drained after refueling by switching the discharge of one of the pumps from the reactor injection no::le to the borated water storage tank.

The other pump will continue the recirculation mode of decay heat removal.

9 5.2.4 Reliability Considerations Two pumps and two coolers are provided.

9 5.2.5 codes and standards The equipment in this system vill be designed to the applicable code or stan-dard ss noted in Table 9-5 .

9.5.2.6 system Isolation The decay heat re= oval system is connec;ed to a reactor outlet line on the suc-tion side and to the reactor vessel, through the core flooding lines, on the discharge side. On the suction side the connection is through two remotely l 1 operated gate valves in series, and on the discharge side through one remotely operated gate valve and two check valves in series. All of these valves are nor= ally closed whenever 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.'( Leakage considerations During reactor operation all equipment of the decay heat removal system is idle, and all isolation valves are closed. During the accident condition, fission products vill be recirculated through the exterior piping system. To determine the total radiation dose released from this system, the potential leaks have been evaluated and discussed in lb.2.2.k.2 and section 6.3.5 9.5.2.8 Failure considerations a

re considerations for the accident case are evaluated and tabulated in 0262 t

9-19 L

D-B Table 9-h Decay Heat Removal System Performance Data Reactor Coolant Temperature at Startup

'of Decay Heat Removal, F 280 ,

Time to Cool Reactor Coolant Systen Frc= 280 F to lho F, hr 1L Maximum Refueling Temperature, F 1h0 Boron Concentration in the Borated Water Storage Tank, pp= Boron 1,800 Table 9-5 Decay Heat Removal System Eauipment Data Pumps Number 2 Type Single Stage, Centrifugal Capacity, gp: 3,000 Head at Rated Capacity, ft 350 Motor Horsepower, hp LOO Material SS (Wetted Parts) 1 Design Pressure, psig 450 Design Te=perature, F 350 Coolers "

Number 2 IYpe Heat Transferred, Btu /hr ShellagdTube 30 x 10 Reactor Coolant Flow, gp= 3,000 2 l Component Cooling Water Flow, gpm 6,000 Reactor Water Inlet Temperature, F 1h0 Material, Shell/ Tube CS/SS 1

Design Pressure, Shell/ Tube, psig 100/h50 Design Temperature, F 350 Code ASME Section III and VIII Borated Water Storage Tank Number 1 Capacity, gal 360,000 Material CS/ Coated Inside Design Pressure Atmospheric Design Temperature,_F 125 Code.,

, AWWA D-100

" Refer to Figure 6-4 for heat transferred as a function of N cooler inlep; water temperature.

s_,

3.

0263 '

Amendment No. 2 9-20

D-B p 9.6 FUEL HANDLING SYSGf i

9.6.1 DESIGN 3ASES The fuel handling system shown in Figure 9-9 is designed to provide a safe, effective means of transporting and handling fuel from the time it reaches the l1 station in an unirradiated condition until it leaves the station after post-irradiation cooling. The system is designed and constructed to minimize the possibility of mishandling or maloperations that could cause fuel assembly damage and/or potential fission product release. This is accomplished through the use of interlocks, travel and load limiting devices and other protective measures.

The reactor is refueled with equipment designed to handle the spent fue'l assemblies under water from the time they leave the reactor vessel until they are placed in a cask for shipment from the site. Undervater transfer of spent ftiel assemblies provides an effective, transparent radiation shield, as well as a reliable cooling medium for removal of decay heat. Borated water insures  :

suberitical conditions during refueling.

9.6.2 SYSTEM DESCRIPTION AND OPERATION 9.6.2.1 Receiving and Storing Fuel New fuel assamblies are received in shipping containers and stored in the new fuel storage area, which is a separate and protected area for the dry storage of new fuel assemblies. The new fuel storage area is sized to accom=odate the maximum number of new fuel assemblies required (80 assemblies). 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 maintoin a ke rf of less than 0 90 even if flooded with non-borated water.

9.6.2.2 Leading and Re=oving Fuel New fuel assemblies are transferred from a new fuel storage area into the spent fuel pool. They are then transferred into the containment vessel by the fuel transfer carriages operating through the fuel transfer tubes. Transfer of new fuel occurs concurrently with removal of spent fuel after the reactor is shut down and the refueling canal is filled with borated water. 1 Two hori:: ental tubes are provided to convey fuel between the containment vessel and the fuel storage pool. Each tube contains tracks for the fuel transfer carriage, a gate valve on the spent fuel storage pool side, and a flanged closure on the containment vessel side. Each fuel transfer tube tene-trates into the refueling canal, inside the containment vessel where 1 space is provided for the rotation of the fuel transfer carriage basket containing a fuel assembly. The other end terminates in the spent fuel storage <

pool, where space is provided for rotation of the fuel transfer carriage basket.

The refueling canal is a passageway in the containment vessel extending 1 from the reactor vessel to the fuel transfer tube. This reinforce nerete I

enclosure, lined with stainless steel, forms a canal above the re vessel, ,

which is filled with borated water for refueling. The refueling al is 1

. ,.(p 0264 9-21

[

D-B also used for storage of the reactor vessel upper plenum and core barrel assemblies.

i Following the reactor shutdown, containment vessel entry, and missile shield removal, the refueling procedure is begun by removing the reactor closure head and control rod drives and their service structure. Head removal and replace-ment time is minimized by the use of two stud tensioners. The stud tensioner is a hydraul;cally operated device that permits preloading and unloading of the reactor closure studs at cold shutdown conditions. The studs are tensioned to their ?rerational load in two steps in a predetermined sequence. Required stud elongatson 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.

The. reactor closure head assembly is handled by a handling fixture supported frem the containment vessel crane. It is lifted out of the canal onto a head -

storage stand located on the operating floor. The stand is designed to pro-tect the gasket surface of the closure head. The lift is guided by two closure head alignment studs installed in two of the stud holes. These studs also provide proper alignment of the reactor closure head with the reactor vessel and internals when the closure head is replaced after refueling. The studs and nuts can be removed from the reactor closure head at the storage location 2 l for inspection and cleaning using special stud and nut handling fixtures. A stud storage rack is provided.

The annular space between the reactor vessel flange and the bottom of the 1

refueling canal is sealed off before the canal is filled by a seal cla= ped to the canal shield plate flange and the reactor vessel flange. The refueling canal is then filled with borated water.

The plenum assembly is removed from the reactor by the containment vessel crar.e using an internals handling adapter and stored under water on a stand on the 1 refueling canal floor.

Refueling operations are carried out from two fuel handling bridges which span 1 the refueling canal. The nain bridge is used to shuttle spent fuel asses-lbliesfromthecoretothetransferstationandnewfuelassembli transfer station to the core. During this operation, the auxiliary bridge is occupied with relocating partially spent fuel assemblies in the core as spec-ified by the fuel management program.

Fuel asse=blies are handled by a hydraulically operated fuel grapple attached to a telescoping and rotating mast mounted on a trolley which moves laterally on each bridge. Control and orifice rod assemblies are handled by a grapple attached to a second mast located on the main bridge in the reactor building.

The main (two-mast) bridge moves a spent fuel assembly from the core under .'

water to the transfer station where the fuel asse=bly is lowered into the. fuel.

transfer carriage fuel basket. The grapple attached to the second mast is'used to transfer a rod assembly to a new fuel assembly. This new fuel assembly with rod assembly is -carried to the reactor by the fuel handling mechanism and

{}

~'

Amendment No. 2 9-22 -

0?6d

i D-B p . located in the core while the cpent fuel assembly is being transferred to the spent fuel storage pool.

Spent fuel assembs removed from the reactor are transported to the spent fuel storage pool uom the containment vessel via the fuel transfer tubes by means of the fuel transfer carr%ges. The spent fuel assemblies are removed from the *uel transfer carriage basket using a hydraulically operated fuel grapple attached to a telescoping mast mounted on a trolley which moves later-ally on the-fuel storage handling bridge. This motor-driven bridge spans the spent fuel storage pool and permits the refueling crew to store fuel assemblies in any one of the many vertical storage rack positions.

The fuel transfer mechanisms are underwater air motor driven carriages that run en tracks extending from the spent fuel storage pool through the transfer tubes and into the containment vessel. A rotating fuel basket is mounted on one end of each fuel transfer carriage to receive fuel assemblies in a vertical '

position. The hydraulically operated fuel basket on the end of the carriage is rotated to a horizontal 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.

During operation of the reactor, the carriages are stored in the spent fuel storage pool, thus permitting the gate valve on the spent fuel storage pool side of the transfer tube to be closed and a blind flange to be installed on the containment vessel side of the tube.

Once refueling is completed, the refueling canal water la drained and 1 pumped to the borated water storage tank as described in 9.h.

9.6.2.3 Storage of Scent Fuel After removal from the reactor and transfer to the fuel storage building, spent fuel is stored ut the spent fuel storage pool.

~The spent fuel storage pool is a reinforced concrete pool, lined with stainless steel, located in the fuel storage building. The pool is sized to accommodate s full core of irradiated fuel assemblies (177 assemblies) in addition to the concurrent storage of the largest quantity of spent fuel assemblies from the reactor as established by the fuel management program (59 assemblies + 19 spare locations). The spent fuel assemblies are stored in racks having spacing and/or poison sufficient to maintain a k gp of less than 0 90 if immersed in fresh (unborated) water. Control red assemblies requiring removal from tne reactor are stored in the spent fuel assemblies. '

The spent fuel storage pool has space for a spent fuel shipping cash, as well as for required fuel storage. Following a sufficient decay period, the spent fuel assemblies are removed frem storage and loaded into the spent fuel ship- j ping cask under water for removal from the site. Casks up to 100 tons in weight can be handled by the fuel storage building crane.

A decontamination area is located in the building ad,jacent to the spent fuel storage pool; in this area the outside surfaces of the casks can. be decontami-nated before shipment.

0266 W

9-23 st. .-

s. ,

D-B 9.6.2.4 Safety Provisions  ;

Safety provisions are designed into the fuel handling system to prevent the development of hazardous conditions in the event of component malfunctions, accidental damage, or operational and administrative failures during refuel-ing er transfer operations. A mechanical lock prevents disengagement of the

• uel assembly grapple latches as long as fuel assembly weight is suspended Jrom the grapple mechanism. Bridge and trolley controls are interlocked to prevent movement until the fuel assembly has been completely withdrawn into the protective mast tube.

The new and* spent fuel assembly storage facilities are designed for noncrit-icality by use of adequate spacing and/or poison. The new and spent fuel storage racks are designed to prevent insertion of a fuel assembly in other than the prescribed locations, thereby insuring a safe geometric array. A safe condition is insured even if new fuel is immersed in unborated water.

Under these conditions, a criticality accident during refueling or storage is not credible.

All spent fuel assembly transfer operations are conducted under water. The 1 l vater level in the refueling canal provides a minimum of 9-1/2 ft of water over the top of the active fuel in the spent fuel assemblies during movement from the core into storage. The depth of the water over the fuel assemblies, 1 l as well as the thickness of the concrete valls of the refueling canal, is suf-ficient to limit the maximum continuous radiation levels in the workiy e. ea to values consistent with the radiation zoning described in Section 11.

1 The spent fuel storage pool water is cooled by the spent fuel cooling system as described in 9.h. A power failure during the refueling cycle vill create no immediate hazardous condition owing to the large water volume in both the 1 refueling canal and spent fuel storage pool.

1 l During the refueling period the water level in both the refueline canal and the spent fuel storage pool is the same, and the fuel transfer tube valve is continuously open. This eliminates the necessity for an interlock between the fuel t.ansfer carriage and fuel transfer tube valve operations except to verify full open valve position.

The simplifiec movement cf a transfer carriage through the horizontal fuel transfer tube midmizes the danger of ja==ing or derailing. All operating mechanisms of the system are located in the fuel building for ease of mainte-nance and accessibility for inspection before the start of refueling operations.

During reactor operation, a bolted and gasketed closure plate located on the containment vessel flange of the fuel transfer tube, together with the fuel transfer tube valve located on the fuel storage building end of the tube, 1 provide containment vessel isolation as described in Section 5.3 Both the spent fuel storage pool and the refueling canal are completely lined with stainless steel for leak-tightness and ease of decontamination. The fuel transfer tubes are appropriately attached to these liners to maintain leak integrity. The spent fuel storage pool cannot be accidentally drained.

s" QM7 #

l 9-2h i

D-B f' The fuel transfer mechanism is designed to permit initiation of the carriage fuel basket rotation from the building in which the carriage fuel basket is being loaded or unloaded.

All electrical gear is located above water for greater reliability and ease of maintenance. The hydraulic system that actuates the rotating fuel basket uses storage pool water for operation to minimize contamination.

The refueling canal and storage pool water will have a boron concentration l1 of 1,800 ppm. Although this concentratica is sufficient to maist'ain core shut-down if all of the control rod assemblies were removed from the core, only a few control rods vill be removed at any one time during the fuel shuffling and l1 replacement. Although not required for' safe storage of spent fuel assemblies, the spent fuel storage pool water vill also be borated so that the refueling l1 canal water vill not be diluted during fuel transfer operation.

Each fuel handling bridge mast travel is designed to limit the maximum 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.

Leaking fuel assemblies removed from the core and verified for leakage are transferred to the spent fuel pool via the fuel transfer tube. The failed fuel is then placed in a failed fuel container in the spent fuel storage rack. Off-site shipment, following a suitable decay period, will require that fuel be

('

transferred to a shipping container compatible with the shipping cask design to comply with 10 CFR 71.

9.6.2.5 overational Limits Personnel exposure time vill be limited so that the integrated doses to operating personnel do not exceed the limits of 10 CFR 20.

The fuel handling bridges are limited to the handling of fuel rod assemblies.

All lifts handling the reactor closure head and reactor internals vill be made u:,ing the containment vessel crane.

Travel speeds for the fuel handling bridges, trolleys, grapple hoists, masts, and fuel transfer carriages vill be controlled to insure safe handling conditions.

9.6.2.6 Miscellaneous Fuel Handling Ecuitment This equipment consists of fuel handling bridges, fuel handling tools , new fuel storage racks, spent fuel storage racks, new fuel transfer containers, control i "

rod handling tools, shipping casks, and fuel transfer mechanisms. In addition to the equipment directly associated with' the handling of fuel, equipment is provided for handling the reactor closure head and the upper plenum. assembly to exposed;t,te ore for refueling.

r -

(_ 0268 9-25

D-B 97 SAMPLING SYSTEM

~')

9.T.1 DESIGN BASES Th; sampling system provides samples for laboratory analyses which serve to guide the operation of the Reactor Coolant System, the Makeup and Purifica-tion System, the Chemical Addition System and the Process Steam System. These samples flow to a central location in the auxiliary building; access to the containment vessel for this purpose is not required during power operation.

Typical of the analyses performed on such samples are reactor coolant boric acid concentration, fission, product activity level, dissolved gas content, corrosion product concentration and main steam gross activity. Analytical results are used for regulating boron concentration adjustments, evaluating the integrity of fuel rods and the performance of the demineralizers, and regulating chemical addition to the reactor coolant.

972 DESCRIPTION AND OPERATION 1l Figure 9-10 is a schematic diagram for the station illustrating the features of the system. This system permits sampling of the reactor coolant system and the reactor auxiliary systems during normal plant power and shutdown operation. Fol-loving a loss-of-coolant accident, samples of the recirculating coolant are taken from the discharge of the decay heat removal pumps. The sampling system is oper-ated manually, and on an inter =ittent basis, under conditions ranging from full power to cold shutdown operation. During a loss-of-coolant accident, this system is isolated at the containment vessel boundary.

)

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

a. Liquid

1. Reactor Coolant System Pressurizer

2. Purification Demineralizer Inlet

3. Purification Demineralizer (Ntlet

h. Makeup Tank 5 Decay Heat Pump Discharge

6. Core Flooding Tanks T. Boric Acid Solution Makeup

8. Radwaste System (Local Grab Samples) 9 Steam Generator Secondary Side

10. Borated Water Storage Tank

(,3 O$Y9

^

9-26

1 l

D-B f b. Vapor and Gas

1. Pressurizer

2. Makeup Tank

3. Quench Tank

h. Radvaste System (Local Grab Samples) 5 Main Steam Samples are collected in containers designed for full operating temperatures and pressure and at flow velocities which insure transport of suspended parti-cles where appropriate. Sample lines are purged to insure that representative samples are obtained.

~

Gaseous leakage is collected by placing the sampling station under a hood pro-vided with an off-gas vent to the radvaste area ventilation system. Liquid leakage from the valves in the hood and sampling effluents are drained to the vaste disposal system.

9.8 STATION VENTILATION SYSTEMS 9.8.1 DESIGN BASES

(

The heating, ventilating, and air-conditioning systems for the station are de-signed to provide a suitable environment for equipment and personnel sith equipment arranged in zones so that potentially contaminated areas are sep-arated from clean areas. The path of ventilating air in the auxiliary build-ing is from areas of lov activity toward areas of progressively higher acti-vity. Conditioned air is recirculated in clean areas only.

9.8.2 SYSTEM DESCRIPTION AND EVALUATION A flow diagram of the containment vessel cooling water system is shown on l Figure 6-6 and the containment vessel, shield building, and penetration l7 room ventilation and filtration system on Figure 6-7 The remaining ven-tilating systems for the station are shown on Figure 9-11. 1 9.8.2.1 containment vessel ventilation The containment vessel normal ventilation system is. discussed in Section 5.h.

9.8.2.2 Emergency ventilation.

The Emergency Ventilation System, which ventilates the shield building and l7 penetration rocms, is discussed in Section 6.3 l1 9.8.2.3 Auxiliary Building Ventilation

(- The auxiliary building is served by separate ventilation szste for the fuel V handling area, the radvaste area, the non-radioactive areaf ... l7 the control room area.

. ~.~.~ ;-. -,

0270 9-2T Amendment do.7

D-B i The ventilation systems for the fuel handling area and the radvaste areas are designed on a "once through" be. sis to control and direct all potentially con-taminated air to the veau stack via_ roughing and high efficiency particulate filters. Exhaust air from these areas 1s nonitored before it iY~ discharged from the station through the vent stack. Redundant exhaust fans are provided for these areas.

In the event of an accident within the fuel-handling area, a negative pressure may be created immediately. By means of ductverk and dampers, the fuel-hand-ling area is connected to the Shield Building and Penetration Roo= Ventilation 6 System (section 5 5 and 6.3), and thus exhaust air is passed thru the HEPA and charcoal filters before being discharged through the stack. The system is shown in Figures 6-7 and 9-11.

The clean non-radioactive areas are ventilated by means of heating and venti-lating units employing outside and recirculated air to conserve heat during vinter operation.

The control room area system employs redundant fans, filters and mechanical refrigeration equipment, plus the necessary dampers and controls for switch-ing to full recirculation for post-incident ventilation. The control roo=

area system performance vill be continually monitored with alarus for high radiation and equipment malfunction. The control room operator vill have remote manual control for selecting damper position, and fan and filter opera-tion in order to ensure satisfactory control room conditions following an in-cident. All control area ventilation equipment vill be remote from the con- )

trol area and vill not be exposed to fire hazards. The control room vill be maintained at a pressure somewhat higher than the cable spreading room and the rest of the plant.

9.8.2.h Turbine Buildine ventilation Turbine building ventilation is provided by means of several fresh air supply systems which conduct cooling air to motors and equipment. The air is recir-culated in cold weather to conserve heat. Air is exhausted from the turbine building through roof exhaust fans. Unit heaters will provide space heating during station shutdown.

Ventilation for the emergency diesel generator rooms, the intake structure and the service water building are all served by individual fresh air systems in the same manner as the turbine building systems.

9.8.2 5 Station Heating The station uses low-pressure extraction steam for station heating. Station heating is provided by the auxiliary boiler during station shutdown.

9.8.2.6 System Reliability Ventilation systems and equipment are designed in accordance with the recom-mended practices of the American Society of Heating, Refrigeration and Air- -

s Conditioning Engineers Guide, the Air Moving and Conditioning Association ,/

and National Fire Protection Association. Redundant exhaust fans are provided for the potentially contaminated areas.

. .. i Amendment No. 6 9-28 02r;tt

i D-B 9.9 INSTRUMENT AND SERVICE AIR SYSTEM 991 DESIGN BASES

.he Instrument and Service Air System is designed to provide a reliable con-tinous supply of dry, oil-free compressed air for pneumatie instrument oper-ation and for control of pneumatic valves. The system also supplies air to service outlets throughout the station for operation of pneumatic tools or other requirements. The compressors supply air at a pressure of 100 psig with pressure reduced as necessary for the various service requirements.

9

9.2 DESCRIPTION

AND OPERATION Figure 9-i2 is a schematic diagram of the system. During normal operation, one of the two full capacity air compressors may operate to supply statiort. .

instrument and service air requirements. The remaining air empressor is placed in automatic standby frem the control room and vill start upon de- .

crease of supply air header pressure. Operation of the standby air compressor is annunciated in the control room. To ensure instrument air supply, the service air header is automatically valved off when the compressed air sys-tem pressure drops to a preset value.

A smaller non-lubricated instrument air empressor vill also be provided to supply essential instrument air during a loss of electric power. This com-pressor vill operate from either emergency diesel generator as required.

Protection against loss of instrument air is provided by redundancy in active emponeats emprising the instrument air system. In addition, in the event of a loss of all instrument air supply, all pneumatically operated valves are arranged to assume their respective safe positions.

To maintain acceptable purity and lov dev point, a dual tower dessicant type air dryer and two full capacity filters are provided from which the instrument air supply is split into headers; branch lines are taken off to supply all areas of the station.

The power source for the station compressor motors is the normal a-c distribution systen. Backup from the emergency diesel generators is provided by manuni transfer. .

9 9.3 TESTS AND INSPECTIONS Each cmpressor is inspected periodically to ensure equipment operability.

During nomal operation, the two station compressors are operated alternately.

The standby instrument compressor vill be checked and operated periodically.

D 02?2 - '-

9-29

D-B 9.10 AUXILIARY FEEDWATER SYSTEM 1

9 10.1 DESIGN BASES The auxiliary feedvater system is designed to provide feedvater to the steen generators when the turbine driven main feedvater pu=ps are not available or following a loss of nor=al, standby, and emergency electric power. All com-ponents and piping in the system are designed to seismic Class I requirements ,

and are tornado protected.

9 10.2 DESCRIPIION AND OPERATION The schematic diagram for the auxilian feedvater system is shown in Figure 9-13. The auxiliary feedvater system may be utilized during normal unit startup and shutdown. On startup, the auxiliary feedvater pu=ps may be ased to fill and maintain level in the steam generators until the main feedvater pumps are capable of supplying this load using steam from the auxiliary boiler.

On station shutdown, the auxiliary feedvater pu=ps can be used to remove the 2 decay heat until the decay heat removal system can be placed in service.

Following a complete loss of all station power, or following a postulated steam line break, the auxiliary feedvater syste= supplies water direct to the stea=

generators through the auxiliary feedvater ring header to remove reactor decay heat. Reactor decay heat removal after coastdown of the reactor coolant pu=ps is provided by the natural circulating characteristics of the reactor coolant system. Use of the auxiliary feedvater system for cooldown is discontinued when the reactor coolant system temperature decreases to about 280 F; further )

cooldown is accomplished by the decay heat removal syster.

8 lTheauxiliaryfeedvatersystemconsistsoftwosteamturbine-driven feedvater pumps, the condensate storage tank, the deaerator, suction and dis-ch'arge water piping, steam piping valves and associated instrumentation and controls. The pamps take suction from the condensate storage tank and the 8 3 deaerator or from the condensate feed pumps during startup. A backup is also provided from the fire protection system and from the service water system.

The turbine driver receives steam from the steam generators (auxiliary .

boiler during startup) and exhausts to the. atmosphere. The condensate storage vill be sized such that a total condensate inventory is available to the pumps sufficient to remove decay heat for approximately eight hours plus a 2 subsequent cooldown to 280 F. Lov tank level is annunciated in the control room.

The capacity of the pu=ps is deter =ined by the decay heat removal requirements 40 seconds after reactor trip at full power (assu=ing infinite irradiation at 2,772 MWt). Each pu=p vill have 1000 GPM capacity. One pu=p meets the capac-ity criteria.

System reliability is achieved by the following features:

1. Two turbine-driven pu=ps are provided

2. Steam is supplied by separate steam lines from separate s steam generators 3 ,

j 0273 Amendment No. 8 9 -30

p.

D-B

.r '- 3. In the event of loss of water supply from the demerator or condensate storage tank, a manuel backup is provided from the service water system, and from the fire protection 3 system.

h. Feedwater to the steam generator is supplied through lines separate from the main feedwater lines and through separate steam generator nozzles.

All active components of the system are accessible for inspection during station operation. The auxiliary feedwater pumps will be tested periodically during station operation discharging via the recirculated piping back to the condensate storage tank or to the condenser.

O h

f

? . oan 9-31 Amendment No. 3 1 I - -I j

1 D-B 9 11 FIRE PROTECTION SYSTEM 9.11.1 DESIGN BASES The fire protection system furnishes water to all points throughout the station area, where water for fire fighting may be required.

The fire protection system is designed to provide the following:

a. A reliable supply of fresh water for fire fighting.

b. A reliable system for delivery of water to potential fire locations.

c. Autc=atic fire detection in those areas where the danger of fire is marked.

d. Fire extinguishment by fixed equipment activated autcicatically or manually for those areas where the danger of fire is marked.

e. Manually operated fire extinguishing equipment for use by Station personnel at selected points throughout the property.

8\ Portable CO2 and chemical fire extinguishers are provided at key locations.

The design of the fire protection syste= will be in general accordance with the requirements of the National Fire Protection Association, the American Insurance Association, the Nuclear Energy Property Insurance Association and the applicable codes and regulations of the State of Ohio.

Additionally, the fire protection system provides back-up service for the following: ,

, a. Auxiliary Feedvater Pump Suction 1 b. Spent Fuel Pool

c. Diesel Generator Cooling Water (Ref. 8-10) 8 (DELETED)

.9.11.2 SYSTEM DESCRIPTION AND OPERATION Fire protection is provided by means of fixed fog deluge systems, sprinklers, hose lines, portable extinguishers, instrumentation, and controls.

Water for the fire protection system is provided by two full capacity fire pumps. One pump is electrically driven, the other is diesel-engine driven.

Both pumps are arranged to start manually, or automatically on low fire system header. pressure, with.the diesel engine-driven pump being started at a lower pressure switch setting than the motor-driven pu=p. The diesel pump may be started from the main control board. A jockey pump, complete Amendment No. 8 9-32

l i

D-B with local controls, is provided to maintain the system full and pressurized.

[ System pressure is monitored in the control roca with low system pressure and i fire pump start annunciated.

l The motor driven fire pump and jockey pump take suction from a fire water g j storage tank. The diesel fire pump is located in the intake structure with

suction taken from Lake Erie. The pumps discharge into the main fire protec-tion water header which completely encircles the station building. Branch l lines are taken .off the fire main header to supply each fire hydrant, each j deluge system, each sprinkler system, and each wet standpipe system. Fixed

fog deluge systems, automatically actuated, protect the main generators, and 8

! the main, startup and station auxiliary transfomers. Wet pipe fusible heat

! sprinkler systems provide fire protection to the oil storage rooms for the

! main turbines (containing the oil storage tanks, oil reservoirs and oil filters),

electrical penetration rocas, the diesel generator rooms, hydrogen seal oil g.

room and the turbine building basement areas where oil could spill.

Readily accessible reel mounted hose line stations with tog type nozzles are i located throughout the plant, so that all areas in the turbine and aul .11ary building are withi'n 50 feet (maximum) of a 35 psig (minimum) hose reel.

Portable fire extinguishers are provided at convenient and accessible loca-tions. The extinguishing media are pressurized water, CO2 or dry chemicals as required by the regulatory agencies for the service requirements of the area.

\

With the exception of underground piping, all equipment and components are l readily accessible for periodic inspection. Tests are periodically conducted i in accordance with the requirements of the regulatory and insurance agencies.

1 4

5 cy

. s

~0276 Amendment No. 8 9 i. .

T I

I 1

9 4

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4 I f9

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1 0277 I

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ABBRlVlATIONS FLOW DEVICES ES EMERGEhCY IN,ECI.ON ACTUAT.ON 5.GNAL DR DDAIN )

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F FL0s An ANAll2ER L LEVEL C CONTROLLER OR P PRES $11RE 1 IN0lCATOR Ra GADIAll0N R RECORDER T T E MP ER AT URE S SWITCH PIPE SYMBOL M

f HAND POSITION Z

I INTEGRATOR TEST POINT

( )

l- SCH hf MAIL dP OlFFERENTIAL E ELEMENT ~

PRESSURE I TRANSMITTER SYSTEM ABBREVIA110NS RC REACTOR C00LANI WU M AlfilP AND PURIFIC AT10N DH DEC AY HF AT REMOVAL h CC COMPONENT COOLING CA CHEMICAL ADDITION WO R A010ACilVE W ASTE TRE ATMENT (WASTE DISPOS AL) N SCW SERVICE COOLING WATER F05 FEE 0BATER DE DEMINERAltIED WATER SF SPENT FUEL P0OL COOLING CS CONTAINMENT SPRAY MS MAIN STEAM LINE DESIGNATIONS HEAT TRACEO PRIMARY FLOW SECOND ARY FLOW f J' - / -' PNEUMATIC

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CAPlLLARY TUBING E LE C T R I C A L

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(O't" DAVIS-BESSE NUCLEAR POWER STATION

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AMENDMENT NO. 6 .

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AMENDMENT NO. 6

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, DAVIS-BESSE NUCLEAR POWER STATION CONDENSER CIRCULATING WATER SYSTEM

-, FIGURE 9-4 AMEN 0 MENT NO. 8 0285 }

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DAVIS-BESSE NUCLEAR POWER STATION SERVICE COOLING WATER SYSTEM FIGURE 9-5 AMENDMENT NO. 8 OM e

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WA57E GAS COMPRESSORS, SAMPLE ConLERS, CONTROL ROD ORNE COOLERS AND DAVIS-BESSE NUCLEAR POWER STATION NQUENCH TANK COOLER */ COMPONENT COOLING WATER SYSTEM y FIGURE 9-6 AMENDMENT ha B 0283

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-- SPENT FUEL POOL COOLING SYSTEM FIGURE 9-7 b AMENDMENT NO. 7 0291

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AMENDMENT NO. 8 0237 I

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DAVIS-BESSE NUCLEAR POWER STATION STATION VENTILATION SYSTEllS I FIGURE 9-11 k

AMENDMENT NO. 8 0233

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M M JL

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