ML19329E142

Chapter 9 of AR Nuclear 1 PSAR, Auxiliary & Emergency Sys. Includes Revisions 1-18

ML19329E142
Person / Time
Site:	Arkansas Nuclear
Issue date:	11/24/1967
From:	ARKANSAS POWER & LIGHT CO.
To:
References
NUDOCS 8005300722
Download: ML19329E142 (54)

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TABLE OF CONTENTS Section M 9 AUXILIARY AND EMERGENCY SYSTEMS 9-1 91 MAKEUP AND PURIFICATION SYSTEM 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 9-3 9 1.1.4 Purification Eemineralizer 9-3 9 1.1 5 Makeup Pumps 9-3 9 1.1.6 Seal Return Coolers 9-3 9 1.1 7 Makeup Tank 9-3 9 1.1.8 Filters 9-3 9 1.2 SYSTEM DESCRIPTION AND EVALUATION 9-4 9 1.2.1 Schematic Diagram 9-4 9 1.2.2 Performance Requirements 9-4 9 1.2 3 Mode of operation 9-4 9 1.2.4 Reliability Considerations 9-5 9 1.2 5 Codes and Standards 9-5 9 1.2.6 System Isolation 9-5 9 1.2 7 Leakage Considerations 9-6 9 1.2.8 operating Conditions 9-6 92 CHEMICAL ADDITION AND SAMPLING SYSTEM 9-lo 9 2.1 DESIGN BASES 9-10 9 2.1.1 General System Function 9-lo 9 2.1.2 Boric Acid Mix Tank 9-10

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9 2.1.3 Boric Acid Pumps 9-10 3 9-1 I79 .

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CONTENTS (Cont'd) r]

Section M 9 2.1.4 Potassium Hydroxide Mix Tank 9-10 9 2.2 SYSTEM DESCRIPTION AND EVALUATION 9-10 9 2.2.1 Schematic Diagram and System Description 9-10 9 2.2.2 Performance Requirements 9-11 9 2.2.3 Mode of 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 9-13 9 2.2 9 Operating conditions 9-13 93 COOLING WATER SYSTEMS 9-17 ,

9 3.1 DESIGN BASIS 9-17 932 SYSTEM DESCRIPTION AND EVALUATION 9-17a 6 9 3.2.1- Service Water and Intermediate cooling Systems 9-17a 9 3 2.2 Auxiliary Cooling Water Sys tem 9-21 9323 condenser circulating Water System 9-21 9.4 SPENT FUEL COOLING SYSTEM 9-23 9.4.1 DESIGN BASES 9-23 9 4.2 SYSTEM DESCRIPTION AND EVALUATION 9-23 9.4.2.1 Sche =atic Diagram 9-23 9.4.2.2 Performance Requirements 9-23 9.4.2.3 Mode of Operation 9-23 9.4.2.4 Reliability Considerations 9-24 9.4.2 5 codes and Standards l .. ! 9-24 9-11 7-11-68 3 180 Su,D1ement No. 6 - -

CONTENTS (Cont'd)

Section Page.

9.6.2.6 Leakage Considerations 9-24 9.4.2 7 Failure Considerations 9-24 9.4.2.8 operaticg Conditions 9-25 95 DECAY HEAT REMOVAL SYSTEM 9-26 951 DESIGN BASES 9-26 9 5 1.1 General System Function 9-26 9 5 1.2 Decay Heat Removal Pumps 9-26 9513 Decay Heat Removal Coolers 9-26 952 SYSTEM DESCRIPIION AND EVALUATION 9-26 9 5 2.1 Schematic Diagram 9-26 9 5 2.2 Performance Requirements 9-26 9523 Mode of operation 9-26 9 5 2.4 Reliability Considerations 9-27 9525 Codes and Standards 9-27 9 5 2.6 System Isolation 9-27 9527 Leakage Considerations 9-27 9 5 2.8 Failure Considerations 9-27 96 FUEL HANDLING SYSTEM 9-29 9.6.1 DESIGN BASES 9-29 L

9.6.1.1 General System Function 9-29

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9.6.1.2 New Fuel Storage Area 9-29 9.6.1 3 Spent Fuel Storage Pool 9-29

.9.6.1.4 Fuel Transfer Tubes 9-29 9 6.1 5 Fuel Transfer Canal 9-30 L

9 6.1.6 jgl Miscellaneous Fuel Handling Equipment 9-30 9-111 5-4-70 Supplement No. 17

COnmus (Cont'd)

Section Title Page 9.6.2 SYSTEM DESCRIPTION AND E'IALUATION 9-30 9.6.2.1 Receiving and Storing Fuel 9-30 9.6.2.2 Loading and Removing Fuel 9-30

. 9.6.2 3 Safety Provisicas 9-32 9.6.2.4 operational Limits 9-34 97 PIANT VENTIIATION SYSTEMS 9-35 971 DESIGN BASES 9-35 972 SYSTEM DESCRIPTION AND EVALUATION 9-35 98 FIRE PROTECTION SYSTEM 9-36 :17

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9-iv 5-4-70 Supplement No. 17

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cottrmTS (cont'd)

Section h

9.6.2 SYSTEM DESCRIPTION AND EVALUATION 9-30 9.6.2.1 Receiving and Storing Fuel 9-30 9.6.2.2 Loading and Removing Fuel 9-30 9.6.2 3 .

Safety Provisions 9-32 9.6.2.4 Operational Limits 9-34 97 PIANT VENTIIATION SYSTEMS 9-35 9 7.1 DESIGN BASES 9-35 9 7.2 SYST m DESCRIPTION AND EVALUATION 9-35 9.8 FIRE PROTECTION SYSTEM 9-36 13 l.

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10-31-69 Supplement No. 13 I83 l

LIST OF TABLES .

Table No. Title Page 9-1 Makeup and Purification System Performance Data 9-T 9-2 - Makeup and Purification System Equipment Data 9-8 9-3 Steam Generator Feedwater Quality 9-14 94 Reactor Coolant-Qualtiy 9-14 9-5 Chemical Addition and Sampling System Equipment Data 9-15 9-6 Reactor Building Cooling Units Performance and Equipment Data 9-21 4

9-T Intermediate Cooling and Service Water Systems Design Parameters and Component Data 9-22 9-8 Spent Fuel Cooling System Performance and Equipment Data 9-25 9-9 Decay Heat Removal System Performance Data 9-27 9-10 Decay Heat Removal System Equipment Data 9-28 G

l84 9-v

LIST OF FIGURES (At rear of Section)

Figure No. Title 9-1 Flow Diagram Identifications 9-2 Makeup and Purification System 9-3 Chemical Addition and Sampling System 9-k . Service Water and Auxiliary Cooling Water System 9-5: Intermediate cooling System 9-6 Spent Fuel Cooling System 9-7 Decay Heat Removal System 9-8 Decay Heat Generation versus Time After Shutdown 9-9 Turbine and Auxiliary Building Ventilation System

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9 AUXILIARY AND EMERGENCY SYSTEMS The auxiliary systems required to support the reactor coolant system during normal operation of Russellville Nuclear Unit are described in the following sections and listed below:

a. Makeup and purification system.

b. Chemical addition and sampling system.

c. Cooling water systems,

d. Spent fuel cooling system.

e. Decay heat removal system.

f. Fuel handling system.

g. Plant ventilation systems.

Some of these systems are described in detail in Section 6 since they serve as engineered safeguards. The information in this section deals primarily 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 reactor building and the auxiliary building are equipped with reactor building isolation valves as described in 5.h.

The codes and standards used, as applicable, in the design, fabrication, and testing of components, valves, and piping are as follows: i

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

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

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

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

e. Standards of the American Society of Testing Materials.

f. USASI, B31.1,Section I (Power Piping).

g. USASI, C50.20-195h Test Code for Polyphase Induction Motors and l Generators. I

h. USASI, C50.2-1955 for Alternating Current Motors, Induction Machines, and General and Universal Motors, lbb 9-1

i. Standards of the American Institute of Electrical and Electronics Engineers.

j. Standards of the National Electrical Manufacturers Association.

k. Hydraulic Institute Standards.

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

m. Standards of Tubular Exchanger Manufacturers Association.

n. Air Moving and Conditioning Associat, ion.

o. USASI, B96.1, Aluminum Tanks.

p. Valves and piping, vill be designed and fabricated to meet the re-quirements of USASI B16. ..r MSS SP-66 and USASI B31.1, respectively.

q. The pressure-containing parts of all pumps of stainless steel mate-rial vill be liquid penetrant-examined in accordance with Appendix VIII of Section .VIII of the ASME Code. The pressure-containing velds of all engineered safeguards pumps vill be radiographically examined in accordance with Paragraph UW-51 of Section VIII of the  !

ASME Code.

As an aid to review of the system drawings, a standard set of symbols and ab-breviations has been used and is summarized in Figure 9-1.

91 MAKEUP AND PURIFICATION SYSTEM 9.1.1 DESIGN BASES 9 1.1.1 General System Function i

l The system shown on Figure 9-2 supplies the reactor coolant system with fill l and operational makeup water; circulates seal water for the reactor coolant pumps and control rod drives; receives, purifies, and recirculates reactor

! coolant system letdown to provide water quality and reactor coolant boric acid concentration control; accommodates temporary changes in the required reac-tor coolant inventory and provides makeup to the core flooding tanks.

9 1.1.2 Letdown Coolers The letdown coolers cool the letdown flow from reactor coolant temperature to a temperature suitable for demineralization and injection to the reactor cool-l ant pump seals and control rod drive seals. The maximum letdown flow is re-quired for a startup from a cold condition late in core life wherein the re-l

. actor coolant boron concentration is reduced by an amount corresponding to the change due to moderator temperature reactivity deficit. Heat in the letdown j . coolers is rejected to the intermediate cooling system.

\ST 6-5-68 9_2 Supplement No. 4

9.1.1.3 Letdown control Letdown fica is established by use of a block orifice which is sized fcr 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 the parallel control valve along with the 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 either purification demineralizer to remove reactor coolant impurities other than boron. The purification letdown flow to maintain the reactor coolant water quality is equal to one reactor coolant vol-ume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Each purification demineralizer is sized fbr the maximum let-down flow rate as required for boron concentration control. Fefer to Table 11-3 for the maximim anticipated equilibrium fission product accumulation in the reactor coolant.

9 1.1.5 Makeup Pumps The makeup pumps are designed to return the letdown flow to the reactor cool-ant system and supply the seal water flow to the reactor coolant pumps and the control rod drives. The design flow capacity is equal to the maximum makeup flow plus the seal water flow to the reactor coolant pumps and the control rod drives. The pumpa are sized to meet these requirements with one pump in opera-tion.

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 and the control rod drive seals. Heat from these coolers is rejected to the ser-vice water system.

9.1.1.7 Makeup Tank This tank serves as a surge vessel for the makeup pumps and as a receiver for the letdown flow, chemical addition, and outside makeup; it also accomodates temporary changes in reactor coolant system volume. The volume of the tank is such that the useful tank volume is equal to the maximum expected expansion and contraction of the reactor coolant system during power transients.

9.1.1.8 Filters The filters will prevent the entry of resin fines from the demineralizer and other particulates from the Chemical Addition System, and the plant demineralized water supply into the system and into the seals of the r.eactor coolant pumps and control rod drives. '

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]88 2-8-68 Amendment Noe1

9.l.2 SYSTEM DESCRIPTION AND EVALUATION 9.1.2.1 Schematic Diagram

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The makeup and purification system is shown on Figure 9-2.

9.1.2.2 Performance Requirements Tables 9-1 and 9-2 list the system performance requirements and data for indi-vidual system components.

o 9.1.2.3 Mode of operation 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, to a header which supplies the seals of the control rod drives, and to a makeup line connection to one of the reactor inlet lines.

Makeup-flow to the reactor coolant system is regulated by the makeup control valve, which operates on signals from the liquid level controller of the reac-tor coolant system pressurizer. Control valves in the injection line to the pump seals, and in the header of the control rod drive seals, automatically maintain the desired inlet pressure to the seals. A small part 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 addition, bleed and feed of reactor coolant are required for removal of impurities and boric acid from the reactor coolant. Reactor coolant is removed from one of the reactor inlet lines, cooled during passage through one of the letdown coolers, passed from the reactor building 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 waste disposal system. The level in the makeup tank is maintained with deborated water from storage or with' demineralized water from the plant demineralized water storage tank. The quantity of unborated water received is measured and limited by inline instrumentation and interlocked with shim rod position controls.

The makeup tank also receives chemicals for addition to the reactor coolant. l A hydrogen overpressure maintained in the makeup tank supplies the hydrogen added to the reactor coolant. Other chemicals are injected in solution to the makeup tank.

System control is accomplished remotely from the control room with the excep-

, tion of the seal return coolers. The letdown flow rate is set by remotely opening the stop valve upstream of the block orifice, and/pr positioning the 9-4 2-8-68 <

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!: Amendment No. 1.

letdown control valve to pass the desired flow rate. The spare purification demineralizer can be placed in service.by remote positioning of the demineral-izer isolation valves. Diverting the letdown flow to the waste disposal sys-

~ tem is accomplished by remote positioning of the three-way valve and the valves in the waste disposal system. The control valves in the injection lines to the reactor coolant pump seals and the control rod drive seals are automatically controlled by pressure differential controller connected to the reactor coolant system to maintain the desired inlet pressure to the seals. The pressurizer makeup control valve is automatically controlled by the pressurizer level con-troller. During heatup and cooldown, the reactor coolant system pressure varies from 100 to 2,185 psig, and the discharge pressure of the makeup pumps remains about 2,600 psig. The letdown control valve is designed for letdown flow rate control at reduced reactor coolant system pressure.

The makeup pumps are controlled remotely.

For emergency operation as a high pressure injection supply, the normal let-down coolant flow line and the normal seal injection return line are closed; and flow is diverted to the emergency high pressure injection lines. The pumps and pump motors are designed to operate at the higher flow rates and lower dis-charge pressures associated with the high pressure injection requirements.

Emergency operation is discussed in detail in 6.1.

9.1.2.4 Reliability Considerations The system has three letdown control paths in parallel (block orifice, remotely operated control valve, and manual valve) and two, full-capacity letdown coolers to insure the flow capability needed to adjust boric acid concentra-tion. Two full-capacity seal return coolers are supplied.

Three makeup pumps are supplied; one is capable of supplying the required re-actor coolant pump seal, control rod drive seal, and makeup flow. The letdown coolers transfer heat to the intermediate cooling system, and the seal return coolers transfer heat to the service water system.

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

Components which are designed to the ASME Code are:

Letdown Cooler - ASME Section III-C Seal Return Cooler - ASME Section III-C Purification Demineralizer - ASNE Section III-C Nbkeup Tank - ASME Section III-C i 9.1.2.6 System Isolation The letdown line, the reactor coolant pump seal return line, and the control rod drive seal return line penetrate the reactor building. All three lines ,

contain' electric motor-operated isolation valves inside the reactor building and. pneumatic valves outside which are automatically closed with operation of the engineered safeguards. +

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Amendment No. 1 j

.Four emergency injection lines are used for injecting coolant to the reactor .

vessel after a loss-of-coolant incident. Check valves in the discharge of each makeup pump provide further backup for reactor building isolation if re-quired. - After use of the lines for emergency injection is discontinued, the remotely operated valves in each line outside the reactor building are closed by the control room operator.

9 1.2 7 Leakage considerations Reactor coolant is normally let down to this system. Each purification demin-eralizer vill remove essentially 100 per cent of the ionic and solid contami-nants except for boric acid while gaseous contaminants vill tend to collect in the makeup tank as the letdown flow is sprayed into the gas space of this tank.

The gas void in the makeup tank may be vented to the vaste disposal system by opening a remotely operated valve in the vent line. The equignent in this system is shielded by concrete. Shielding design criteria are discussed fur-ther in Section 11.

9 1.2.8 Operating conditions Tg makeup tank will be maintained with a fluid inventory between 100 and 500 ft . Oxygen accumulation in the tank vill be less than 2 per cent by volume.

One letdown cooler and two makeup pumps vill be functional at all' times.

To prevent an inadvertent excessive dilution of the reactor coolant boric acid concentration, three safety measures are applied to the method of diluting, i.e.,

the bleed and feed method. The first safety measure is a 140 gpn limitation on the maximum rate of adding demineralized water; for feed and bleed, the demin-eralized water makeup control valve to the makeup tank is automatically con-trolled to prevent exceeding a preset flow rate. The second safety measure is a' control rod assembly position interlock which either permits or prohibits dilution depending on the control rod pattern. The third safety measure con-sists of closing the makeup tank makeup valves, and diverting the letdown flow through the three-way valve back to the makeup tank, when the flow has integrated to a preset value back to the makeup tank. Initiation of dilution must be by the operator, and the operator can terminate dilution at any time.

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i 9-6 2-8-68 Amendment No. 1

Table 9-1 Makeup and Purification System Perfomance Data Letdown Flow (cold), gym h5-140 Total Flow to Each Reactor Coolant Pump Seal, gpm h5-50 Seal Inleakage to Reactor Coolant System per Reactor Coolant Pump, gpm 6 Injection Pressure to Reactor Coolant Pump Seals at Startup, psig 850-2,235 Injection Pressure to Reactor Coolant Pump Seals (normal), psig 2,235 Injection Pressure to Reactor Coolant Pump Seals (maximum), psig 2,535 Temperature to Reactor Coolant Pump Seals, normal / maximum,F 125/150 Total Flow to Each Control Rod Drive Seal, gph 30 Seal Inleakage to Reactor Coolant System per Control Rod Drive, gph 5 Injection Pressure to Control Rod Drive Seals at Startup, psig 135-2,235 Injection Pressure to Control Rod Drive

, Seals (normal), psig 2,235 Injection Pressure to Control Rod Drive Seals (maximum), psig 2,535 Temperature to Control Rod Drive Seals, nor=al/ maximum,F 125/150 Purification Letdown Fluid Temperature, normal / maximum,F 125/140 Makeup Tank Normal Operating Pressure, psig 15

. Makeup Tank Volume Between Minimum and Maximum Operating Levels, ft3 400 '

Reactor Coolant Water Quality See Table 9 4 2 97 2-8-68 Amendment No. 1

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I i .g la Table 9-2 Makeup and Purification System Equilment Data

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^ (Capacities are for single components.)

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• 74 Makeup Pump Quantity 3 Type Multistage centrifugal, mechanical seal Capacity, gym See Figure 6-2 Head, ft at sp. gr. = 1 See Figure 6-2 Motor Horsepower, hp 700 Pump Material 3 SS vetted parts Design Pressure, psig 2,850 Design Temperature, F 200 Makeup Filter Quantity 2

. Capacity, gym 150 T Design Pressure, psig 7 150 Design Temperature, F 200 Letdown Cooler Quantity 2 full-capacity Type Shell and tube Heat Transferred Btu /hr 16.1 x 10 IetdownFlow,lb/hr 3 5 x 104 letdown Temperature Change, F 555 to 120 3 Material,shell/ tube CS/SS Design Pressure, psig 2,500 Design Temperature, F 600 Seal Return Cooler Quantity 2 full-capacity Type Heat Transferred, Btu /hr Shell ang tube 2.2 x 10 SealReturnFlow,lb/hr 1.025 x 105 Seal: Return Temperature Change, F IM to 122 Material,shell/ tube CS/SS Design Pressure, psig 100 Design Temperature, F 200

- Cooling Water Flow, lb/hr 1.025 x 105 Makeup Tank Quantity 1 Volume, ft 3 600

. Design Pressure, psig 100 Design Temperature, F 200 Material SS

. 9-8 5-3-68 Supplement No. 3

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Table 9-2. (Cont'd)-

Purifiestion Demineralizer

, Quantity 2 Type. Mixed bed, boric acid saturated

_ Cation: Anion Ratio ~ 2:1 Material. SS Resin Volume, ft3 40 Flow, gpm. 70

. Vessel Design Pressure, psig 150 Vessal Design Temperature, F 200 4

k 194 Q

j 92 CHElMICAL ADDITION AND SAMPIING SYSTEM N

9 2.1 DESIGN BASES /

9 2.1.1 General System Function Chemical addition and sampling operations are required to alter and monitor

.the concentration of various chemicals in the reactor coolant and auxiliary systems. The system shown on Figure 9-3 is designed to add boric acid to the reactor coolant system for reactivity control (see Table 3-5 and Figure 3-1), potassium hydroxide for pH control, and hydrazine for oxygen control.

The system is designed to take reactor coolant samples and steam generator water samples.

9 2.1.1 Boric Acid Mix and Batch Tanks 3 A boric acid mix tank is provided to mix the concentrated boric acid solu-tion, and a batch tank is provided to store the mixed solution. The volume of the batch tank vill provide approximately half the required boric acid solution to increase the boron concentration of the reactor coolant system to that required for cold shutdown with no xenon. Heaters in the tank main-tain the temperature above that required to insure solubility of the boric acid. Transfer lines vill be electrically traced.

92.13 Boric Acid Pumps Two boric acid pumps are provided to facilitate transfer of the concentrated boric acid solution from the boric acid mix tank to the borated water stor-age tank, the makeup tank, or the spent fuel storage pool. The pumps are sized so that when both are operating, one complete charge of concentrated boric acid solution from the boric acid mix tank may be injected into the re-actor coolant system in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

3 9 2.1.4 potassium Hydroxide Mix Tank The tank volume was established to contain a sufficient enount of KOH for

-continual addition to the reactor coolant system so that a concentration of 3-6 ppm can be maintained while letting down at the maximum rate.

9 2.2 SYSTEM DESCRIPIION AND EVAEATION 9 2.2.1 Schematic Diagram and System Description Figure 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 six tank and discharge to either the spent fuel storage pool, borated water storage tank, or upstream of the makeup tank. At the end of core life, both boric acid pumps are re-quired to raise the reactor coolant system boron concentration from the minimum end-of-life concentration to the refueling concentration in approx-imately 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 device and a heating unit.

h 19b9-lo 5-3-68 u

Supplement No. 3

The potassium hydroxide equipment consists of a mix tank, a single positive displacement pump, and connecting pipin6 The pump discharges to the makeup tank.

A hydrazine drum is connected to a positive displacement pump, which dis-charges to a line leading to the makeup tanks. A nitrogen blanket is used to displace the hydrazine as it is removed from the drums.

The liquid sampling portion of the system receives samples of the reactor coolant from the pressurizer, from upstream and downstream of the purifica-tion demineralizers, from the makeup tank, from the core flooding tanks, from the decay heat cooler outlet, and from the secondary side of the steam generators. Water qualities to be maintained are listed in Tables 9-3 and

94. Gaseous samples are taken from the pressurizer vapor space and from the makeup tank. Sample lines from these points are piped to a sampling cu-bicle outsida the reactor building. Samples are collected in containers de-signed for full operating temperature and presure at flow rates of 0.66 and 3 1.67 gpm.

An automatic gas analyzer is used to monitor various tanks and equipment in the vaste disposal system in a continuous sequence for hydrogen-oxygen mix-tures and to alarm at a preset level. Capability is also provided to sample the gas stream to the analyzer.

The pertinent parameters for each major component in the chemical addition and sampling system are shown in Table 9-5 9 2.2.2- Perfomance Requirements This system permits sampling of, and chemical addition to, the reactor cool-ant system and the reactor auxiliary systems during normal operation and has no active emergency function. During a loss-of-coolant accident, this sys-tem is isolated at the reactor building boundary.

9 2.2 3 Mode of Operation The system is capable of drawing reactor coolant samples during reactor oper-ation and during nuclear unit cooldown when the decay heat removal system is in operation. Access to the reactor building is not required.

Sampling of other process coolant, such as process streams or tanks in the vaste disposal system, is accomplished locally. Equipment for sampling non-radioactive fluids is separated frca the equipment provided for reactor cool-ant samples. Leakage and drainage resulting from the sampling operations are collected and drained to tanks located in the vaste disposal system.

During nomal operation, liquid and vapor samples may be taken from the fol-loving points:

Liquid

! a. Steam generator secondary water.

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b. Reactor coolant system pressurizer.

c. Purification demineralizer inlet.

j g(3 9-11 5-3-68 Supplement No. 3

d. Purification demineralizer outlet.

e. Makeup tank. 3

f. Cora flooding tanks.

6 :- Decay heat cooler outlet (Reactor Coolant) 3 Vapor and Gas

a. Pressurizer

b. Makeup tank.

c. Waste gas decay tank 3 In addition, an oxygen and hydrogen analyzer automatically samples the gas spaces in the vaste disposal system tanks and equipment in an automatic se-quence. A gas sample may also be taken with this equipent.

During normal operation, this system also delivers the following chemicals:

a. Boric acid to the spent fuel storage pool, the borated water storage tank, and the makeup tank. 3

b. Potassium hydroxide to the mnkeup tank.

c. Hydrazine to the makeup tank.

9 2.2.4 Reliability Considerations The system is not required to function during an emergency, nor is it re-quired 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 equipent.

9 2.2 5 Codes and Standards The equipment in this system vill be designed to applicable codes and stan-dards tabulated in Section 9 Equipment applicable to the ASIG Codes are:

the Reactor Coolant Pressurizer Sample Cooler which will be designed to ASIE Section III, Class C, and the Steam Generator Sample Cooler which will be designed to ASIE Section VIII.

9 2.2.6 System Isolation Isolation of this system from the reactor building is accomplished by signals from the Safeguards Actuation System as described in Sections 5 4 and 71.

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 sump. Ieakage of radio-197 5-3-68 Supplement No. 3 o

active material from this system outside the reactor building is collected by placing the entire 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 a liquid vaste disposal tank. .

The chemical addition portion of this system delivers additives to the spent fuel storage pool and the makeup and borated water storage tanks. Additives 3 to the spent fuel storage pool are delivered above the water level. Backflow from the tanks to the positive displacement pumps is prevented by check valves and normally closed shutoff valves between them.

9 2.2.8 Failure Considerations To evaluate system safety, the following failures or malfunctions were as-sumed concurrent with a loss-of-coolant accident, and the consequences were analyzed. As a result of this evaluation, it is concluded that proper con-sideration has been given to plant safety in the design of the system.

Corments and Component Failure Consequences Pressurizer Sample Electrically operated Diaphra6:n-operated valve samplin6 valve inside outside the reactor reactor building fails building vill close.

to close on ES signal.

Steam Generator Steam Electric motor-operated Sample line is not con-Sample sampling valve outside nected directly to reac- ,

reactor building fails tor coolany system, and I to close on ES signal. steam generator therefore provides first barrier.

Sample Line from Ei- Line breaks inside re- Remotely-operated valves ther of the Preceding actor building down- outside reactor building Components stream of EbD valves. close on signal from ES system.

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9 2.2 9 operating Conditions 9 2.2 9 1 Boric Acid Concentration The boric acid mix tank is to be maintained at an average temperature of 95 F to maintain a boric acid concentration of 7 per cent.

9 2.2 9 2 Coolant Sample Temperature The high pressure reactor coolant samples leaving the reactor coolant pres-surizer sample cooler should be held to a temperature of 200 F to minimize the generation of radioactive aerosols.

198 9-13 ',

5-3-68 Supplement No. 3

v Table 9-3 Steam Generator Feedwater Quality w

Parameter Value Maximum Total Dissolved Solids, ppm 0.05 Suspended Solids, ppm 0.0 Hardness 0.0 Organic, ppm 0.0 Maximum Dissolved Oxygen, ppm 0.007 Carbon Dioxide 0.0 Maximum Total Silica (as SiO2), Ppm 0.02 Maximum Total Iron (as Fe), ppm 0.01(*)

{ Maximum Total Copper (as Cu), ppm 0.002(*)

pH 9 3 to 9.5 Lead and Heavy Metals None

(*) Included in maximum TDS as a soluble compound.

Table 9 b Reactor Coolant Quality Parameter Value Total Solids, max. (excluding H B03 3 and KOH), 1.0 ppm Boron, ppm See Figure 3-1 KOH, ppm 3-6 pH at 77 F 5.5-6.0 pH at 560 F (calculated) 7-10 02 (max.) ppb 10 C1 (max.) ppm 0.1 H2 , std ec/1- 15 ho Hydrazine (required during .=hutdown), ppm 25 199 9-1u 2-8-68 Amendment No. 1

Table 9-5 Chemical Addition and Sampling System Equipment Data (Capacities are for single components.)

Tanks Boric Acid Mix Tank Quantity 1 Type verti al cynn M eal Volume, ft3 1,000 3 Design Pressure, psig Atmospheric Design Temperature, F 200 Material Al Boric Acid Batch Tank 3 Quantity 1 Type Vertical Cylindrical Volume, ft 3 134 Design Pressure, psig Atmospheric Design Temperature, F 200 Material Al Potassium Hydroxide Mix Tank Quantity 1 Type Vertical Cylindrical Volume, gal 50 Design Pressure, psig Atmospheric Design Temperature, F 200 Material SS Hydrazine Drum Quantity 2 Type Std. Commercial 55 gal Drum Pumps 4 Boric Acid Pump Quantity 2 Type Reciprocating, Variable Stroke Capacity, gpm 0-10 Head, psi 50 Design Pressure, psig 100 Design Temperature, F 200 Pump Material SS 3

Potassium Hydroxide _ Pump Quantity 1 Type Reciprocating, Variable Stroke ' l Capacity, gph 0-10 Head, psi 50 Design Pressure, psig 100

' Design Temperature, F -100 Pump Material SS l

9-15 ,

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l 200- -

5-3-68 Supplement No. 3 1

Table 9-5 (Cont'd)

Hydrazine Pump Quantity 1 Type Reciprocating, Variable Stroke Capacity, gph 0-10 Head, psi 50 Design Pressure, psig 100 Design Temperature, F 100 Pump Material SS Sampling Sampling Containers Quantity k Design Pressure, psig 2,500 Design Temperature, F 670 Reactor Coolant Pressurizer Sample Cooler Quantity 1 Type Shell and Spiral Tube Heat Transferred, Btu /hr 2.1 x 105 Sample Flow Rate, gpm 1 Max. Sample Inlet Temperature, F 650 Sample Outlet Temperature, F 150 Cooling Water Flov, Ib/hr 5 x 103 Coil Side Design Temperature, F 670 Coil Side Design Pressure, psig 2,500 Steam Generator Sample Cooler Quantity 2 Type Shell and Spiral Tube Heat Transferred, Btu /hr 2.0 x 105 Sample Flow Rate, gpm 1 Sample Inlet Temperature, F 525 Sample Outlet Temperature, F 150 Cooling Water Flov, lb/hr 5 x 103 Coil Side Design Temperature, F 600 Coil Side Design Pressure, psig 1,050 201 9-16

93 COOLING WATER SYSTDIS 931 'DESIGH BASIS The cooling water systems are arranged into three separate pumping systems.

a. A service water system and intermediate cooling system cool all nuclear cycle and fuel handling requirements. These

. systems are essentially two independent full capacity sys-tems to insure continuous heat removal from components requiring cooling. During an emergency where power genera-tion is lost, the service water system is powered from emergency power sources while the intermediate cooling system and non vital items are shut-off. In the unlikely 6 event of the ccmplete loss of cooling water from the Dardanelle Reservoir due to the failure of the Dardanelle Dam, water will be supplied by gravity flow from the emergency reservoir through a supply line to the service water compartment in the intake structure. After having been circulated through the service water system, the emergency cooling water is directed back to the emergency reservoir by the service water pumps through a return line.

b. An auxiliary cooling water system furnishes all non-nuclear related cooling water requirements. This system may be permitted to stop functioning in an emergency where power generation is lost.

c. A condenser circulating water system provides cooling water to the main surface condensers under normal operation. on loss of electrical power, this system will not be operated.

The cycle cool-down requirements are adequately handled by other means.

These systems will be sized to insure adequate heat removal based on highest expected temperatures of cooling water, maximum loadings, and leakage ' allowances. - The equipment in these systems will be designed to applicable codes and standards tablulated in Section 9, page 9-1.

All cooling water' systems will be designed to prevent any component fail-ure,from curtailing normal unit' operation. It will be possible to isolate all heat exchangers and pumps.-

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

Isolation valves located external to the reactor building will be in-corporated _in -all cooling service water lines penetrating the reactor building.-

Electrical power requirements for the service water system can be supplied by any of the redundant power sources described in 8.2 3 Neither the

-intemediate cooling system, auxiliary cooling water system, nor the

"~ condenser circulating water system are intended'to be. operated from the diesel generators.

9-17' 7-11-68

All system components will be hydrostatically tested prior to unit startup and will be accessible for periodic inspections during operation.

All electrical components, switchovers, and starting controls will be tested periodically. Design parameters for system components are listed in Tables 9-6 and 9-7 93.2 SYSTEM DESCRIPTION AND EVALUATION 9 3.2.1 Service Water and Intermediate cooling Systems All services cooled by these systems (Figures 94 and 9-5) will be connected with the nuclear plant and, hence, segregated to these systems.

Ihe. service water system will provide cooling water for the following equipment:

9-17a 4

203 s .,

Normal - Decay heat coolers (shutdown) 17 ,s I

- Control room chillers

- hkeup pump L.O. coolers

- Engineered safeguard equipment area coolers

- Intermediate coolers, serving:

Steam generator sample coolers Control rod drives Letdown coolers Reactor coolant pump coolers Waste gas compressor aftercoolers Air ccmpressors Spent fuel coolers Reactor coolant pump seal return coolers Pressurizer sample cooler Reactor coolant pump motor coolers Reactor coolant pump lube oil coolers Main F.W. pump L.O. coolers Iso-phase bus cooler Main chillers Administrative building chiller Emergency.- Reactor building coolers

- Emergency diesel generators

- Emergency feed pump lube oil coolers

- R.B. spray pump L.O. coolers

- Decay heat coolers

- Makeup pump L.O. coolers Redundancy is obtained by dividing the service water systen into two independent I circuits arranged such that the failure of any single vital component will not affect the required performance of the system.

The service water pumps will be located in the intake structure. Raw water at 85 F will be circulated in the coolers located throughout the plant building.

l The intermediate cooling pumps and coolers are located in the auxiliary build-ing. This closed loop system provides an additional barrier between high pres-sure reactor coolant and service water to prevent an inadvertent release of radioactivity. ~

The cooling loops .will be continuously monitored 'for radioactivity, and if any is detected, the leaking cooler will be isolated and the spare cooler will be put in service.

l i

9-18 1 5 4 -70 Supplement No. 17,

The intermediate cooling system pumps and coolers vill be rotated on a scheduled.

basis to monitor their operational capability.

• The four reactor building cooling units each contain separa'te and independent nor- 17 mal and e'mergency cooling coils. The normal cooling coils are served by chilled water to remove the building heat load.- Service uter is circulated through the emergency coils to cool the reactor building atmosphere after a L.O.C. A.

Performance requirements for the service water system during the post-accident '

situations are described in Section 6 of this report. Normal operation require-ments are listed in Tables 9-6 and 9-6.

The operation of the intermediate cooling and service water systems is monitored with the following instrumentation:

a. Temperature indicators in the main inlet and outlet lines of the intermediate coolers. -

b. A pressure indicator on the discharge line of each pump,

c. A flow indicator and alam in the intermediate cooling water line upstream of the intermediate coolers.

d. A level indicator and alarm on the intermediate cooling system surge tank.

e. A temperature indicator located on each outlet line of the letdown coolers and of the mechanixal seal areas of the reactor coolant pumps.

f.

A radiation monitor and alam on each intermediate cooling 17 vater header serving radioactive components upstream of the intermediate coolers.

g.

A radiation monitor and alarm and flow indicator and alam 17 on the outlet line from each set of two reactor building cooling units.

h. A radiation monitor and alam on tre outlet line from each

~

decay heat removal cooler.

1. Temperature indicators and alarms in the control rod drive 17 cooling coil inlet and outlet headers.

Thereliabilityconsiderationsforthesesystemsareprimarilybasedonthe

. considerations for post-accident operations which are discussed in Section 6.

System Isolation 1 Intermediate Cooling System l

The intermediate cooling system inlet lines to the reactor building 17 vil.1 have pneumatically-operated isolation valves'outside of the -

T

' reactor building penetration and check valves inside. The outlet - -) ' ,

lines from the reactor, building vill have' remotely operated stop _ .-

. . J- ,

.. , m-

.m _g.19 -

. . 5-4-70, -

20$ .-

~~

Supp1ement no. 17

. . .z 1 . . .

g E valve 3 on both raideo cf the reactcr building penetration. Exc;pt f;r the normally closed makeup line and equipment vent overflow, and -

drain lines, there are no direct connections between the cooling .

' water and other systems. . The equipment vent and drain lines outside the reactor building have manual valves which am normally closed ]

unless the equipment is being vented or drained for maintenance or n pair operations. The vent lines are also capped as an additional safety feature.

53

. Service Water System -

Each inlet and outlet cooling water line serves two reactor building 17 emergency cooling coils and contains a remotely operated valve out-side the reactor building vall. There are no direct connections to any other system or to the reactor buildig atmosphere. .

Ieakage Considerations Water leakage from piping, valves, equipment, etc., in the system inside the reactor building is not considered to be generally detrimental unless the leakage exceeds the makeup capability. With respect to water leakage from piping, valves, and equipment outside the reactor building, velded construction is used where possible to minimize the possibility of leakage.

Service water could become contaminated with radioactive water due to one of the following:

a. A leak in a letdown. cooler tube or a cooling coil for the mechanical seal on a reactor coolant pump simultaneously with a leak in the intermedia,te cooling heat exchanger. s

b. Ieakage of radioactive products from the reactor building 17 air during actual post-accident operation of the air coolers -

into the cooling water. Tube or coil leaks in the air coolers are detected by radiation monitors located on each of the cool-ing water outlet headers. Two cooling coils comprising 50 per cent of the emergency cooling capacity can be isolated if necessary.

c. Tube or coil leaks in co=ponents being cooled by the inter-mediate cooling system, are detected by a radiation monitor

' located on each of the intermdiate cooler main outlet headers. 17 A leak in a letdown cooler tube is isolated by closing the D D valves on the inlet and outlet lines on the reactor coolant side of the cooler. A defective coil for a mechanical seal is isolated by entering the reactor building and closing the man-ual valves on the inlet and outlet cooling lines.

d. A tube leak in a decay heat removal cooler, either during~ I normal'use for decay heat removal or during post-accident l usage, is detected by a radiation monitor located on the cooler cooling water outlet.line. The cooler may then be isolated by closing the DD valves located on the inlet and outlet cooling water lines of the defective cooler.

j -

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5 4 -70 Supplement No'. 17

Failure Considerations The consequences of equip =ent, piping,' valves, instnicentat, ion, etc. , failures in this system are presented in Section 6. .

9 3 2.2 Auxiliary Cooling Water System This will be a normally operated cooling water system with one of the 17 thzwe service water pumps available to pump 85 F rav vater to heat ex- -

. changers located in the turbine and auxiliary buildings.

All services served by this system vill be expendable in the event of accident and the system can be dropped on unit trip at time of accident.

No nuclear-oriented services vill be serviced by this syste.

9323 Condeneer Circulating water System Rav vater from the Dardanelle Reservoir vill be used as the source of water for the condenser circulating water system.

The intake structure vill be provided with a trash removal system. The circulating water pump intake structure vill have a suction extending below low water level and vill be provided with traveling screens. The intake structure vill have four circulating water pumps.

The condenser circulating water system vill be designed to take advantage of syphon effect to reduce the static head.

Table 9-6 Reactor Building Cooling Units Perfor-ance and Equipment Data (Capacities are for single components. ) ,

Number of Reactor Building Cooling Units 4

~

2.7 ~

Number of Units Normally Operating 3 Normal Air Flow, cfm total .

1.12 x 10 Fan Static Pressure Head, in. WG 7 Emergency Duty EnteringAirConditions,F/RH 286/ Saturated -

~~

HeatRejectedtoServiceWater, Btu /hr 60 x 10 17 Note: Other design parameters are listed in 6.2 for post-accident operations.

9-21 5-4-70 Supplement No. 17; N. ' '

~

._g

Table 9-7 Intermediate Cooling and Service Water Systems Design parameters and Component Data (Capacities are for single components)

Intermediata Cooling Pumps Quantity 3 Half Capacity 17.

Number Nnmally Operating 2 Type Centrifugal Rated Capacity, gpm 2500 Rated Head, Ft 125 Eter Horsepower, hp 125 CasingMaterial/ Trim CastIron/ Bronze Intermediate Coolers Quantity 3 17' Number Normally in Service 2 Type Shell agd Tube .

HeatTransferred, Btu /Hr 30 x 10 '

Shell Side (intermediate cooling system)

NormalInlet/ outlet,F 125/95 17.

Flow Rate, spm 2000 Tube Side (raw river water)

NormalInlet/ Outlet,F 85/106.4 17.

Flow Rate, gpm 2800 MaterialTube/Shell Admiralty /CS Service Water Pumps Quantity 3 Full Capacity 17 Number Normally Operating (1 S.W. & 1 Aux. C.W.) 2 Type Centrifugal ,

Rated Capacity, gpm 8000 Rated Head, Ft 100 2 tor Horsepower, hp 250 CasingMaterial/ Trim Steel / Steel 9-22 l 5-4-70 i Supplement No. 17 l

208

9.4 SPEIfr FUEL COOLING SYSTEM 9.4.1 DESIGN BASES The spent fuel cooling system is shown on Figure 9-6. It is designed to main-tain the spent fuel storage pool at 105 F with a heat load based on removing thedecayheatgenerationfromone1/3 core,whichhasbeenirradiatedfor 930 days and cooled for 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />. In meeting the design bases above, the system has the additional capability to maintain the spent fuel storage pool at 142 F peak temperature while removing the decay heat from the following combination of stored fuel assemblies:

a. 1/3coreirradiatedfor930daysandcooledfor100 days.

b. 1/3coreirradiatedfor720daysandcooledfor150 hours.

1 c.'1/3coreirradiatedfor410daysandcooledfor150 hours.

d. 1/3coreirradiatedfor100daysandcooledfor150 hours.

9.4.2 SYSTEM DESCRIPTION AND EVALUATION l

9 4.2.1 Schematic Diagram The schematic diagram for the spent fuel cooling system is shown in Figure 9-6.

Spent fuel is cooled 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 the spent fuel storage pool water,  !

water contained in the fuel transfer canal during the refueling operation, and the contents of the borated water storage tank (after it has been used in the fuel transfer canal during refueling.)

9.4.2.2 Performance Requirements The first design basis of the system predicates an operating schedule in which the nuclear unit is on an equilibrium refueling period (310 FPD per cycle) with l approximately 1/3 of a core being removed from the unit at the end of each l period. At equilibrium the removed fuel assemblies will have been in the reactor l3 for three cycles, i.e., 930 days at the time of discharge.

l The second design basis for the system considers that it is possible that dur-ing the life of the plant it will be necessary to unload the reactor vessel totallyformaintenanceorinspectionatthetimethatthe1/3coreisalready residing in the spent fuel storage pool.

The basic system performance and equipment data are presented in Table 9-8.

9 4.2.3 Mode of operation During normal conditions 1/3 of a core will be stored in the pool. At this time, operation of two pumps and two , coolers will maintain a peak temperature 3 of 105 F. Operating one pump and one cooler will maintain a peak temperature of 124 F. The pool is initially filled with water from the borated water s storage tank.

9-23 209 '

5-3-68 l Supplement No. 3

For the case where 1-1/3 cores - are stored due to complete ' unloading of the re-actor vessel, two pumps and two coolers vill maintain the spent fuel stcrage pool temperature at 142 F peak temperature. If both a pump and a cooler are out 3 .n) for maintenance when this storage condition exists, the water temperature vill eventually rise to 200 F peak temperature, although considerable time vill 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 212 F for.each of the foregoing quantities of stored fuel is as follows:

One-third of a core (one cooler) 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> b

One and one-third cores (two coolers) 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> The borated water recirculation pump provides circulation of the water contained in the fuel transfer canal during the refueling operation, and of the water contained in the borated water storage tank after the refueling operation through the spent fuel demineralizer and filters, without disturbing the spent fuel cooling operation.

9.h.2.h Reliability Considerations During the time when a 1/3 core is stored in the pool, only one-half of the installed equipment vill be utilized to maintain the pool at 120 F.

9.h.2 5 Codes and Standards The equipment in this system vill be designed to applicable codes and standards tabulated in Section 9 4 Components which are designed to the ASME Code are Spent Fuel Cooler - ASME Section III-C Spent Fuel Coolant Demineralizer - ASME Section III-C 9.h.2.6 Leakage Considerations Whenever a' leaking fuel assembly is transferred from the fuel transfer canal to the spent fuel storage pool, a small quantity of fission products may enter the spent fuel cooling water. A small purification loop is provided for removing these fission products and other contaminants from the water.

The fuel handling and storage area housing the spent fuel storage pool vill be ventilated on a controlled basis, exhausting circulated air to the outside through the plant vent.

Provisions have been made in the design to air-test the valved and flanged ends of the fuel transfer tube for leak-tightness after it has been used.

9.h.2.7 Failure Considerations The most serious failure of this system would be co=plete loss of water in the stor' age pool. To protect against this possibility, the spent fuel storage pool cooling connections enter near or above the water level so that the pool cannot be gravity-drained. For this same reason care is also exercised in the design and installation of the fuel transfer tube.

210 36_3 63 9-2h. . Supplement No. 4

9.h.2.8 oueratine Conditions The pool vill normally be limited to lh2 F peak temperature except in most unusual circumstances as previously described. Boric acid concentration in 3 the pool fluid vill be maintained at 12,000 to 13,000 ppm (2,090 to 2,270 ppm boren).

Table 9-8 Spent Fuel Cooline System Performance and Ecuirment Data (Capacities are for single components.)

System Cooling Capacity, Btu /hr 6

Normal (1/3 core) 8.75 x 10 6 Maximum (1-1/3 cores) 25.85 x 10 System Design Pressure, psig 75 System Design Temperature, F 250 Spent Fuel Cooler Data Quantity 2 Type Tube and Shell Material, shell/ tube CS/SS Duty, Btu /hr 6 Cooling Water Flow, lb/hr 8.75xIg 5 0 x 10 Spent Fuel Pu=p Data Quantity 2 Type Horizontal, Centrifugal Material Stainless Steel Flow, gpm 1,000 Head, ft 100 Motor Horsepower, hp 60 Spent Fuel Coolant Demineralizer FlowR.'te,gpg 160 Bed Volume ft 20 Type Nonregenerative Vessel Material Carbon Steel - Lined Design Pressure, psig 75 Design Temperature, F 200 3

Spent Fuel Storage Pool Water Volume, ft 41,500

[6

. Borated Water Recirculation Pump Quantity 1 Type Horizontal, Centrifugal Material Stainless Steel

( Flov, gpm '

180 Head, ft '

lho Motor Horsepower, hp 15 5-3-68 9-25 7-11-68 . Supplement No. 3 _

I i

95 DECAY HEAT REMOVAL SYSTEM 9.5.1 DESIGN BASES l

9.5.1.1 General System Function The normal function of this system as shown by Figure 9-7 is to remove reac- 6 tor decay heat during the latter stages of cooldova, maintain reactor coolant ttmperature during refueling, and provide the means for filling and draining the fuel transfer canal. The emergency functions of this system are described in 6.1.

9.5.1.2 Decay Heat Removal Pumps The decay heat removal pumps, during shutdown, circulate the reactor coolant from one reactor outle.t line through the decay heat coolers and return it to the reactor injection no::les. The design flow is that required to cool the r: actor coolant system from 250 F to lk0 F Jn lh hours. (The steam generators l3 are used to, immediately after shutdown, reduce the reactor coolant system from operating temperature to 250 F.)

9 5 1.3 Decay Heat Removal Coolers The decay heat removal coolers, during shutdown, reaove the decay heat from the circulated reactor coolant. At 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after shutdown of the reactor (1h hours after reaching 250 F), two coolers and two pumps will reduce the reac-tor coo 3 ant temperature to 140 F.

9.5.2 SYSTEM DESCRIPTION AND EVALUATION 9 5.2.1 Schematic Diagram The decay heat removal system is shown schematically in Figure 9-7 6 9 5.2.2 Performance Recuirements Tables 9-9 and 9-10 at the end of this subsection list system performance data and design data for individual components.

9.5.2.3 Mode of Operation Two pumps and two coolers perform the decay heat cooling function. After the steam generators have reduced the reactor coolant temperature to 250 F, decay htat cooling 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 pump or one cooler is available, the reactor coolant temperature is r:duced at a lower rate.

The equipment utilized for decay heat cooling is also used for low pressure in-jaction into the core during accident conditions.

212 9-26 7-11-68 ,;' 5-3-68 Supplement No. 6 Supplement No. 3

9 5.2.4 Reliability Consideration 4 6 The nuclear unit has two pumps and two coolers.

9525 Codes and Standards Se equipment in this system win be designed to applicable codes and standards tabulated in Section 9 The decay heat removal cooler which is applicable to the ASME Code, will be designed to Section III,' Class C.

9 5 2.6 System Isolation The decay heat removal system is connected to a reactor outlet line on the suc-tion 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 series and on the discharge side through one air-operated gate valve and a check valve in series. All three of these valves are normally closed whenever the reactor is in the operating conditien. 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.

9527 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 will be recirculated through the exterior piping system. To obtain the total radiation dose to the public due to leakage from this system, the potential leaks have been evaluated and discussed in 6.3 and 14.2.

9 5 2.8 Failure Considerations Failure considerations for the accident case are evaluated and tabulated in 6.1.3 Table 9-9 Decay Heat Removal System Ferformance Data Reactor Coolant Temperature at Startup of Decay Heat Removal, F 280 17 Time to Cool Reactor Coolant System From 280 F to 140 F, hr 14 17 Refueli::g Temperature, F 140 Decay Heat Generation Figure 9-8 Fuel Transfer Canal Fill Time, hr 1 Fuel Transfer Canal Drain Time, hr 1 Boron Concentration in the Borated Water Storage Tank, ppm Boron 2,270

,. . i

-' 9-27 l

)

5-4-70 '

hf3 Supplement No. 17 1

Table 9-10 Decay Heat Removal System Eauipment Data ,

(Capacities are given for single components)

Pumps Number 2 Type Single Stage, Centrifugal Capacity,. gym. 3000 Head at Rated Capacity - Ft H O 2 350 Motor Horsepower, hp 400 Material SS (wetted parts)

Design Pressure, psig 300 Design Temperature, F 300 (a)

Coolers Number 2 Type Shell agd Tube HeatTransferred, Btu /Hr 30 x 10 17 Reactor Coolant Flow, gpm 3000 Cooling Water Flow, gpm 3000 Cooling Water Inlet Temperature, F 85 Reactor Water Inlet Temperature, F 140 Material, shell/ tube SS/CS 17 Design Pressure, shell/ tube, psig 450/100 Design Temperature, F 300 Borated Water Storage Tank Number 1 Capacity, Gal. 380,000 Material CS with internal coating 17 l Design Pressure Hydrostatic Head + 10 Ft.

Design Temperature, F. 150 l (a) l Refer to Figure 6-4 for heat transferred as a function of cooler inlet vnter temperature.

l l

l l

9-28 5 h-70 Supplement No. 17 214

~

96 PUEL HANDLING SYSTEM I

/-

'9 6.1 DI: SIGN BASES

. 1 9 6.1.1 .

General System Function '

The fuel handling system is designed to provide a safe, effective means of trans-porting and handling fuel from the time it reaches the plant in an unirradiated condition until it leaves the plant after postirradiation cooling. The system is designed to minimize the possibility of mishandling or maloperations that -

could cause fuel assembly damage and/or potential fission product release.

The. reactor is refueled with equipment designed to handle the spent fuel as-semblies under water from the time they leave the reactor vessel until they are placed in a cask for shipment from the site. Underwater transfer of spent fuel assemblies provides an effective, economic, transparent radiation shield, as well as a reliable cooling medium for removal of decay heat. Borated water insures subcritical conditions during refueling.

9 6.1.2 New Fuel Storage Area The new fuel storage area is a separate and protected area for the dry storage of new fuel assemblies in the fuel storage and handling area. The new fuel storage area is sized to accommodate the maximum number of new fuel assemblies required for refueling of the reactor as dictated by the fuel management pro-gram. 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 suf-ficient to maintain a keff of less than 0 9 when vet.

9 6.1 3 Spent Fuel Storage Pool The spent fuel storage pool is a reinforced concrete pool lined with stainless steel; it is located in the fuel storage and handling area. The pool is sized to accommodate 252 spent fuel assemblies which allows for a full core of irra- 17 diated fuel 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. 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 assemblies requiring removal from the reactors are stored in the spent fuel assemblies.

9 6.1.4' Fuel Transfer Tube A horizontal tube is provided to convey fuel between the reactor building and the fuel storage pool. This tube contains tracks for the fuel transfer car-riage, gate valve on the spent fuel storage pool side, and a means for a flanged cloadre on the reactor building side. The fuel transfer tube pene-trates into the fuel transfer canal at the lover depth, where space is. pro-vided for the rotation of the fuel transfer carriage basket containing a fuel assembly.,

9-29 5-4-70

, Supplement No. 17; v .-

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'9i6.1.5 Puel Transfer canal - - la /~5

^ -

. , ,. JV 4' , tihe fuel transfer canal is a passageway in the reactor building extending from e 3the. reactor vessel to the reactor building vall. It is formed by an upward ex-tension'of the primary shield valls. The enclosure is a reinforced concrete

' , structure lined with stainless steel; it forms a canal above the reactor vessel, which is filled'vith borated water for refueling.

Space is available in the fuel transfer canal for undervater storage of the re-actor vessel internals upper plenum assembly.

The deeper fuel transfer station portion of'the fuel transfer canal can be used for storage of the reactor ve'ssel internals core barrel and ther=al shield as-semblies.

, 9.6.1.'6' Miscelfaneous Puel Handling Equipment This equipment consists of fuel handling bridges, fuel handling tools, new fuel storage racks, spent fuel storage racks, fuel transfer containers, control rod handling tools, viewing equipment, fuel transfer mechanism, and shipping casks.

In addition to the equipment directly associated with the handling of fuel, equipment is provided for handling the reactor closure head and the upper ple-num assembly to expose the core for refueling.

9.6.2 SYSTEM DESCNIPTION AND EVALUATION 9.6.2.1 Receiving and storing Puel +  ;

.i

, 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 ways:

a. After reactor shutdown, new fuel assemblies can be transferred from the new fuel storage area into the reactor building through the equip-ment hatch.

b. After reacter shutdown, nev' fuel assemblies can be transferred from the new fuel storage area into the reactor building by way of'the fuel transfer carriage and the fuel transfer tube.

.9.6.2.2 Loading and Removing Puel

.Following the reactor shutdown and reactor building entry, the refueling proce-dure is begun by removing the reactor closure head and control rod drive as ,

sembly. Head removal.and replacement time is minimized by the use of two stud tensioners. The stud tensioner is a hydraulically operated device that permits preloading and unloading of the reactor closure studs at cold shutdown condi-tions. .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 ,

T and nuts are supported in the closure head bolt holes with specially designed ,

216 .

9-30

spacers. Removal ei the studs with the reactor closure head minimizes handling i

. time and reduces the chance of thread damage.

~Th2 reactor closure head assembly is handled by a lifting fixture supported from the reactor building crane. It is lifted out of the canal onto a head stcrage stand located on the operating floor. The stand is designed to pro-t';ct the gasket surfact of the closure head. The lift i's guided by three clo-s.ura head alignment pins installed in three of the stud holes. These pins also '

provide proper alignment. of the reactor closure head with the reactor vessel i 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 for inspection and cleaning using special stud and nut handling fixtures. A stud and alignment pin storage rack is provided. ,

The annular space b' e tween the reactor vessel flange and the bottom of the fuel transfer canal is sealed off, before the canal is filled, by a seal clamped to the canal shield plate flange and the reactor vessel flange. The fuel transfer canal is then filled with borated water.

The upper plenum assembly is removed from the reactor by the reactor building crane and stored under water on a stand on the fuel transfer canal floor using a lifting device with special adapters.

Ratueling operations are carried out from two fuel handling bridges which' span 3 ths fuel transfer canal. The main bridge is used to shuttle spent fuel assemblies from the' core to the transfer station and new fuel assemblies from the transfer

. stction to the core. During this operation, the auxiliary bridge is occupied with i

ralocating partially spent fuel assemblies in the core as specified by the fuel nanagement program.

Fual assemblies are handled by a pneumatically operated fuel grapple attached to a telescoping and rotating mast which moves laterally on each bridge. Control rod assemblies are handled by a control rod grapple attached to a second mast loccted on the main bridge in the reactor building.

Ttie main (two-mast) bridge moves a spent fuel assembly from the core under water

• to the transfer station where the fuel assembly is lowered into the fuel trans-fcr carriage fuel basket. The control rod grapple attached to the second mast is ussd to transfer a control rod assembly to a new fuel assembly. This new fuel asstably with control rod assembly is carried to the reactor .by the fuel grapple and located in the. core while the spent fuel ~ assembly is being ransferred to the.

cpent fuel' storage pool.  ;

l Spent fuel' assemblies removed from the reactor are transported to the spent fuel

~

l storage' pool from the reactor building via a fuel transfer tube by means of a i fual transfer carriage. The spent fuel assemblies are removed from the fuel trans- I far carriage basket.using's pneumatically operated fuel grapple attached to a l movable mast located on the fuel storage handling bridge. This motor-driven l

. 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.

Tha fuel transfer mechanism is an underwater carriage that runs on tracks

/ ' cxtInuing from the spent fuel storage pool through the transfer. tube '

V .

217 J 9-31 5-3-68 Supplement No. 3'

and into the reactor building. . A rotating fuel basket is mounted on one end of the 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 renoval of the fuel assembly.

Once refueling is completed, the fuel transfer canal water 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.

3 During operation of the reactor, the carriage is stored in the spent fuel stor-age 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 reactor building side of the tube.

A space isolated from the spent fuel storage pool is provided for a spent fuel 17 shipping cask to permit loading spent fuel elements into the shipping cask.

Following a sufficient decay period, the spent fuel assemblies are removed from, storage and loaded into the spent fuel shipping 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 cask washdown and decontamhtion area is located in the building ad,jacent to the 17 spent fuel storage pool; in this area the outside surfaces of the casks can be washed down upon arrival and decontaminated before shipment by using steam, water, or detergent solutions, and manual scrubbing to the extent required. s 9 6.2 3 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 refueling or transfer operations.

All fuel ucscribly stcra~e facilities, new an.1 cract. maintain an eversafe geo-n'tric cracinc of 21 in.between sesemblics. The new and spent fuel storace racks are designed so that it is imposcible 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. spac-ing insures an eversafe geometric array in unborated water. Under these con-ditions, 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. Mechanical damage to the fuel assemblies during transfer operations is possible, although remote. Since the fission product release would occur under water, the amount of activit/ reaching the environ-ment vill present no appreciable hazard. A fuel handling accident analysis is included in Section 14.2.2.1.

4 All spent fuel asccmbly transfer operations are conducted under vater. The , ,

water 1cvel in the fuel transfer canal provides a minimum of 10 ft of water' -

over the active fuel line of the , spent fuel assemblies during movement from  ;'

2lO 9-32 ,

54-7o Supplement No. 17

,Q_, - - -

the core into storage; this limits radiation at the surface of the water to less than 10 mrem /hr. The depth of the water over the fuel assemblies , as well as the thickness of the concrete valls of the transfer canal, is sufficient to limit the maximum continuous radiation levels in the working area to 2.5 mrem /hr.

Water in the reactor vessel is cooled during shutdown and refueling by the de-cry heat removal system described in 9 5 In case of a power failure, this sys-tem vill be operated by the auxiliary power supply. The spent fuel storage pool v;ter is cooled by the spent fuel cooling system as cescribed in 9.k. A power failure during the refueling cycle vill create no immediate hazardous condition owing to the large water volume in both the fuel transfer canal and spent fuel storage pool. With a normal quantity of spent fuel asse=blies in the storage pool and no cooling available, the water temperature in the spent fuel storage pool vould increase as discussed in 9.h.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 valve is centinuously open. This eliminates the necessity for an interlock between the fuel transfer carriage and fuel transfer tube valve operations. The simplified movement of a transfer carriage through the horizontal fuel transfer tube mini-mizes the danger of ja= ming or derailing. To cope with such an eventuality, the open tube design provides access to the entire length of the fuel transfer carriage travel frcm the fuel transfer canal. All operating mechanisms of the system are located in the fuel storage and handling area for ease of maintenance and accessibility for inspection before the start of refueling operations.

During reactor operation, a bolted and gasketed closure plate, located en the reactor building flange of the fuel transfer tube, prevents leakage of water from the spent fuel storage pool into the transfer canal in the event of a leak through the fuel transfer tube valve. Both the spent fuel storage pool and the fu21 transfer canal are completely lined with stainless steel for leak-tight-ntss and ease of decontamination. The fuel transfer tube vill be appropriately attached to these liners to maintain leak integrity. The spent fuel storage pool cannot be accidentally drained since water must be pumped out through a suction pipe. 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. .

l3 All elactrical gear is located above water for greater integrity and ease of maintenance. The hydraulic system that actuates the rotating fuel basket uses storage pool water for operation to eliminate contenination.

The fuel transfer canal and storage pool water will have a baron concentration of 2,270 ppm. Although this concentration is sufficient to maintain 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 replacement. Although not required for safe storage of spent fuel assemblies, the spent fuel storage pool water vill also be borated so that the transfer canal' water vill not be diluted during fuel transfer operations.

Each fuel handling bridge mast travel is designed to limit the maximus lift of 3 a fuel assembly to a safe shielding depth.

2lg 9-33 t-5-3-68 Supplement No. 3

y

)

Relief valves are provided on each stud tensioner to prevent overtensioning of

.the studs due to excessive pressure.

Gross failures of fuel are prevented by safety margins in the design and con-trol of the core. The fuel assembly utilizes a free-standing Zircaloy fuel rod of sufficient length to accommodate the expected fission gas release from the fuel.

Any leaking fuel assemblies will be removed from the core for verification of leakage and placed in a failed fuel container. This operation is done in the fuel transfer canal.

The failed fuel is then transferred in the sealed containers 3 to the spent fuel-storage pool. Offsite shipment, following a suitable decay period, vill require that fuel be transferred t6 a liner compatible with the shipping cask design.

9.6.2.k. operational Limita Certain manipulations of the fuel assemblies and reactor internals during re-fueling may result in short-term exposures with radiation levels greater than 2 5 mrem /hr. The 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 handling of fuel and control rod as-semblies an$ reactor closure head studs only. All lifts for handling the re-actor clor 4 e head and reactor internals will use the reactor building crane.

Travel speeds for the fuel handling bridges , masts , and fuel transfer carriage vill be controlled to insure safe handling conditions.

i s

220 5-3-68 9-3h Supplement No. 3

97 FIANT VENTIIATION SYSTEMS 971 DESIGN BASES 2he plant will be designed to provide maximum safety and convenience for operating personnel with equipment arranged in zones so that potentially contaminated areas are separated from clean areas. The heating, ventih lating, and air-conditioning system for the plant will be designed to provide a suitable environment for equipment and personnel. The path of ventilating air in the Auxiliary Building will be from areas of low activity toward areas of progressively higher activity. Ventilating air will be re-circulated in clean areas only.

972 SYSTEM DESCRIPTION AND EVALUATION The Reactor Building nor=al ventilation system is discussed in 5.5 and shown on Figure 5-6. The remaining ventilation systems for the plant are discussed here and shown en Figu-e 9-9 The equipment used to ventilate cach area is independent frem that used in any other area. The systems handling potentially-contaminated air all discharge to the plant vent.

The Auxiliary Building will be served by separate ventilation systcms for the fuel handling area, the radwaste area, the nonradioactive area, and the control room area. These systems are shown on Figure 9-9 The venti-lation air from the fuel handling area and radwaste area is discharged to the plant vent through multi-filter units. These units will be approxi- 17 mately 21+'-0" long by 19' -2" wide by 17'-8" high. They will include rough- ,

ing filters, high efficiency filters, charcoal filters and roughing filters on the downstream side of the chlrecal filters. The unit will be equipped with a water spray system in the charcoal filter bank to protect against fire. The system serving the non-radioactive areas of the Auxiliary Build-ing will also supply air to the laundry, hot laboratory, showers, toilets, and hot machine shop. The discharge air from these areas will go to the plant vent.

The turbine area vencilation syst(m 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 ventilating equipment will be in accordance with accepted indt.stry standards for power station equipment. Redundant exhaust fans will be provided for the potentially contaminated areas. The control room area system performance will be continually monitored with alarms for high radi-etion, fan failure and excessive pressure drop through filters. The control room operator will have manual control for selecting fan and filter oper-ction in order to' insure satisfactory control room conditions following an accident.

The ventilation systems will be designed in accordance with the applicable codes and standards tabulated in Section 9, page 9-1.

O .9-35

! 5 !+-70 Supplement No. 17

ille ventilating equipment will be accessible for periodic testing and inspection during normal operation. Where redundant equipment is provided, it wi n be operated alternately to provide assurance of operability.

A aigh efficiency particulate filters will be tested for efficiency by a dioctylthalate (DOP) aerosol fog test. At least 90% of the particles generated shall be approximately 0 3 to 0 5 micron. Filter efficiency shall not be less than 99 97% when measured by means of a portable photo-meter upstream and downstream of the filter. 17 '

The charcoal filters will be tested by the use of Freon 112 introduced on the upstream side of the filters. The concentration shall be approximately 500 ppm at the filter rated flow of approximately 340 scfm. Instrumentation will provide for measuring the relative 'Tetream and downstream concentrations of Freon 112. A downstream concentration of 0.25% of the upstream 500 ppm will be cause for rejection of the filters. The maximum time for one test shall not exceed 3 minutes.

98 FIRE PROTECTION SYSTEM With the exception of the intake structure and the emergency generator room, fire protection facilities are not located in areas where engineered safe-guards equipment is located.

The fire pumps and accessories are located in the Class I por, tion of the intake structure. The fire pumps are separated from the Class I service water pumps by a Class I wall. The fire system piping will be physically separated from critical piping to preclude damage to Class I piping. 17 2he deluge valves controlling the sprinkler systems in the emergency gener-ator rooms will be of the type that will open only on a manual signal from the Control Room to preclude accidental flooding of these rooms. Fire or smoke detection will be by ionization type detectors which will transmit the signal to the Control Room.

9-36 5-4-70

• Supplement No. 17 L)

The final design and acceptance test requirements for the auxiliary build- 13 ing charcoal filter trains will be ready for submittal by February 1970.

9.8 FIRE PROTECTION SYSTEM I

An evaluation of the fire protection system will be ready for submittal 13 in December 1969 i

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