ML19319D686
| ML19319D686 | |
| Person / Time | |
|---|---|
| Site: | Crystal River, 05000303  |
| Issue date: | 08/10/1967 |
| From: | FLORIDA POWER CORP. |
| To: | |
| References | |
| NUDOCS 8003240667 | |
| Download: ML19319D686 (63) | |
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[Y TABLE OF CONTENTS Section Page 9 EUARY AND EMERGENCY SYSTEMS 9-1 9.1 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 Letdevn Coolers 9-2 9 1.1.3 Letdown Centrol Valves 9-2 9.1.1.4 Purification Demineralizer 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 O 92 srsrta orsCa er on inn zvituir on 9-3 9.1.2.1 Schematic Diagram 9-3 9.1.2.2 Performance Requirements 9-3 9.1.2.3 Mode of Operation 9-h 9.1.2.h Reliability Considerations 9-5 9.l.2.5 Codes and Standards 9-5 9.1.2.6 System Isolatien 9-5 9.1.2.7 Leakage Considerations 9-6 9.1.2.8 operating Conditions 9-6 9.2 CHEMICAL ADDITION AND SAMPLING SYSTEM 9-9
.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 Pures 9-9 0205 9-1 L = _ l
((,>) . CONTENTS (Cont'd)
Section Page 9.2.1.h Caustic Mix Tank . 9-9. ,
9.2.1.5 Caustic Pump 9-9 9 2.1.6 Potassium Hydroxide Mix Tank 9-9 9.2.2 SYSTEM DESCRITTION AND EVALUATICH 9-10 9.2.2.1 Schematic Diagram and System rescriptien 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
() 2.2.2.8 Failure Considerations 9-13 9 2.2.9 Operating Conditions 9-13 9.3 COOLING WATER SYSTEMS 9-18 9 3.1 DESIGN BASES 9-18 9.3.2 SYSTEM DESCRIPTION AND EVALUATION 9-19 9.3.2.1 Nuclear Services and Intermediate Cooling Water Systems 9-19 9.3.2.2 Secondary Services cooling Water System 9-20 9.3.2.3 Cendenser circulating Water System 9-20 9.h SPENT FUEL COOLING SYSTEM 9-2h s
9.h.1 DESIGN HASES 9-2h 9.h.2 SYSTEM DESCRIPTION'AND EVALUATION 9-2h 9.h.2.1 Schematic Diagram 9-2h
, 9.h.2.2 Performance Requirements ^ -2 '-
{~NsJ a9.h.2.3 Mode of Operation 0206 9-2h 9-11
CONTENTS (Cont'd) l)'
q
. Section Page 9.h.2.h Reliability Considerations .9-25 , _.
9.h.2.5 Codes and Standards 9-25 9.h.2.6 Leakage considerations 9-25 9.h.2.7 Failure Considerations 9-25 9.h.2.8 Operating conditions 9-26 9.5 DECAY HEAT REMOVAL oISTEM 9-27 9.5.1 DESIGN BASES 9-27 9.5.1.1 General System Function 9-27 9.5.1.2 Decay Heat Removal Pumps 9-27 9.5.1.3 Decay Heat Removal Coolers 9-27 9.5.2 SYSTEM DESCRIPTION AND EVALUATION 9-27 9.5.2.1 Schematic Diagram 9-27
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9.5.2.2 Performance Requirements 9-27 9.5.2.3 Mode of Operation 9-27 9.5.2.4 Reliability Considerations 9-28 9.5.2.5 Codes and Standards 9-28 9.5.2.6 System Isolation 9-28 9.5.2.7- _ Leakage Considerations 9-28 9.5.2.8 Failure Consi te.-ations 9-28 9.6 FUEL HANDLING SYSTEM 9-31 9.6.1 DESIGN BASES 9-31 9.6.1.1 General System Function 9-31 9.6.1.2 New Fuel Storage Area 9-31 9.6.1 3 Spent Fuel Storage Pool .9-31 9.6.1.h Fuel Transfer Tubes 9-31
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0207 9-111
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CONTENTS (Cont'd)
Section Page s
9.6.1.5 . Fuel Transfer Canal 9-32 1
9.6.1.6 Miscellaneous Fuel Handling Equipment 9-32 9.6.2 SYSTD4 DESCRIPTION AND EVALUATION 9-32 9.6.2.1 Receiving and Storing Fuel 9-32' 9.6.2.2 Leading and Removing Fuel 9-32 9.6.2.3 Safety Provisiens 9-3h 9.6.2.h operational Limits 9-36
, 9.7 PLANT VENTILATION SYSTEMS 9-37 1
4 9.7.1 DESIGN BASES 9-37 9.7.2 SYSTE4 DESCRIPTION AND EVALUATION 9-37 O
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- - - , . - . . - - .~. - - . - . . . - . - .
d LIST OF TABLES
~f Table No. Title Page 9-1 Makeup and Purification Sys, tem Performance Data .9-T 9-2 Makeup and Purification System Equipment Data 9-8 9-3 Steam Generator Feedvater Quality 9-lh 9h Reactor Coolant Quality 9-lh-9-5 ~ Chemical Addition and Sampling System
. Equipment Data 9-15 I
9-6 Cooling Water Systems Performance and
- Equipment Data. 9-22 l 9-T Intermediate cooling System Performance j and Equipment Data 9-23 9-8 Spent Fuel Cooling System Performance and Equipment. Data 9-26 a
9-9 Decay Heat Removal System Performance Data 9-29
] 9-10 Decay Heat Removal System Equipment Data 9-30 4
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0209 g
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LIST OF FIGURES
'O (At rear of Section) j .. Figure No - -
' Title -
9-1 Flow Diagram Identifications 9-2 Makeup and Purification System 9-3 chemical Addition and Sampling System
.-4 Nuclear Services Cooling Water System 9-ka nuclear Services cooling water System 9-4b nuclear Services Sea Water System 9-5 Intennediate cooling System 9-6 Secondary Services cooling Water System 9-7 condenser circulating water System 9-8 Spent Fuel Cooling System O 9-9 o e r a
- a ev 1 sF te-9-10 Decay Heat Generation versus Time after Shutdown 9-11 Fuel Handling System 9-12 Turbine and Auxiliary Building Ventilation System 9-13 Nuclear Plant Office Building Ventilation System O -
0210 9-vi (Revised k-8-68)
(D 9 AUXILIARY AUD EMERGENCY SYSTEMS U
The auxiliary systems required to support each reactor coolant system during normal operation of Crystal River Plant Units 3 and h are described in the i following sections and listed below: -
- a. Makeup and purification system.
- b. Chemical addition and sampling system.
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- 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
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building. Those systems with connecting piping between the reactor buildings and the auxiliary building are equipped with reactor building isolation valves as described in 5.2.
The codes and standards used, as applicable, in the design, fabrication, and testing of components, valves , and piping are as follovs:
- a. ASME Boiler and Pressure Ves:cl Code, See6n II, Material Specifica-tions.
- b. ASME Boiler and Pressure Vessel Code,Section III, Nuclear Vessels.
- c. ASME Boiler and Pressure Vessel Code, Section 'JIII, Unfired Pressure Vessels and ASME Nuclear Case Laterpretations.
- d. ASME Boiler and Pressure Vessel Code,Section IX, Welding Qualifica-tions.
- e. Standards of the American Society for Testing Materials.
- f. USASI, B31.1,Section I (Power Piping).
- g. USASI, C50.20-195h Test Code for Polyphase Induction Motors and Generators.
- h. USASI, C50.2-1955 for Alternating Current Motors, Induction Machines,
-- and General and Universal Motors.
V 9-1
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- i. Standards of the Institute of Electrical and Electronics Engineers.
J. Standards of the National Electrical Manufacturers Association.
- k. Hydrauli: Institute Standards.
- 1. Heating, Ventilating, and A'ir Conditioning Guide; American Society of lleating, Refrigerating, and Air Conditioning Engineers.
- m. Standards of Tubular Exchanger Manufacturers Association.
- n. Air Moving and Conditioning Association.
- o. USASI, B96.1, Aluminum Tanks.
- p. Those components not covered by the ASME Code, such as valves and piping, will be designed and fabricated to meet the requirements of USASI B16.5 or MSS SP-66 and USASI B31.1, respectively. '
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- q. Safeguards systems pressure-containing castings will be radiographed to meet Severity Level 2 of ASTM E-71. The pressure-containing parts of all pumps will be liquid penetrant-tested in accordance with Appendix VIII 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.
h 9.1 MAKEUP AND PURIFICATION SYSTEM 9.'.I DESIGN BASES 9.1.1.1 Generra System Function i
The system shown on Figure 9-2 supplies the reactor coolant system with fill and operational makeup water; circulates seal water for the reactor coolant 7
pumps; receives, purifies, and recirculates reactor coolant system letdown to' provide water quality and reactor coolant boric acid concentration control; and accommodates toporary changes in che required reactor coolanc inventory.
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-ant pump seals. The maximum letdown flow is required for a startup from a cold I conditian late in core life wherein the reactor coolant boron concentration is reduced in 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> by an amount corresponding to the change due to moderator temperature reactivity deficit. Heat in the letdown coolers is rejected to the intermediate cooling system.
9.1.1.3 Letdwn Control Valves Each letdoun control valve is si::ed for the maximum letdown rate.
9-2 (Revised 7-15-69) 0212
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9.1.1.4 Purification Demineralizer
(-} The letdown flow is passed through the purification demineralizer to remove re-
\_,/ actor coolant impurities other than boron. The purification letdown flow to l- maintain the reactor coolant water quality is equal to one reactor coolant vol-tme per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The purification demineralizer is sized for the maximum let-down flow rate as. permitted by the letdown control valve. Refer to Table 11-3 ,
for the maximum anticipated equilibrium fission product accumulation in~the- -i-reactor coolant.
- 9.1.1.5 Makeup Pumps The makeup pumps are designed to return the letdown flow to the reactor coolant system and supply the seal water flow to the reactor coolant pumps. The design lT flow capacity is equal to the maximum makeup flow plus the seal water flow to 2 7
the reactor coolant pumps. The pumps are sized to meet these requirements with r 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 2 and the heat. picked up in passage through the reactor coolant pump seals. 7 Heat f rom these coolers is rejected to the nuclear services cooling water system.
9 .1.1. 7 Makeup Tank This tank serves as a surge vessel for the makeup penps and as a receiver for
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the letdown flow, chemical addition, and outside maksup; it also accommodates tempcrary changes in reactor coolant system volume. The volume of the tank is such that the useful tank volume is equal to the ma.timum expected expansion and
- contraction of the reactor coolant system during power transients.
i 9.1.1.8 Filters
! The filters will prevent the entry of resin. fines from the demineralizer and i other particulates from the Waste Disposal System, Chemical Addition System, and the Plant demineralized water supply into the system and into the seals of the reactor coolant pumps. lT 9.1.2 SYSTEM DESCRIPTION AND EVALUATION 9.1.2.1 Schematic Diagram The makeup and purification system is shown on Figure 9-2.
1
- 9.1.2.2 Performance Requirements Tables 9-1 and 9-2 list the system performance requirements and data for indi-l ~ vidual system components.
l' 9-3 (Revised 7-15-69) i 02 B
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9.1.2.3 Mode of Operation g
During normal cperation of the reactor coolant system, one makeup pump contin- p uously supplies high pressure water from the makeup tank to the seals of each of the reactor coolant pumps, and to a makeup line connection to one of the 7 reactor inlet lines.
Mrkeup flow to the reactor coolant system is regulated by the makeup control valve, which operates on signals from the liquid level controller of the reactor coolant cystem pressurizer. A control valve in the injection line to the pump seals, I automatically maintains the desired inlet pressure to the seals. A small part of the water supplied to the seals leaks into the reactor coolant syatem. The remainder returns to the makeup tank after passing through one of the two seal return coolers.
S:al 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 pascage through one of the letdown coolers, passed from the reactor building through a reactor building isolation valve, reduced in pressure during flow through one of the three letdown control valves, and then passed through one purification demineralizer to a three-way valve which directs tb coolant either to the makeup tank or to the waste disposal system.
Normally, the three-way valve is positioned to direct the letdown flow to the rakeup tank. If the boric acid concentration in the reactor coolant is to be g r:duced, the three-way valve is positioned to divert the letdown flow to the W waste disposal system. Boric acid removal is accomplisned in the waste dispo-sal system either by directing the letdown flow through a deborating demineral-izer with the effluent returned directly to the makeup tank or by directing the letdown flow to a reactor coolant bleed holdup tank and maintaining the level in the makeup tank with demineralized water pumped from the Plant demin-eralized water storage tank. The quantity of unborated water received is mea-sured and limited by inline instrumentation and interlocked with shim rod posi-tion controls.
The makeup rank 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. 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 po-sitioning the letdown control valve to pass the desired flow rate. The spare a purification demineralizer can be placed in service by remote positioning of the demineralizer isolation valves. Diverting the letdown flow to the waste disposal system is accomplished by remote positioning of the three-way valve cnd the valves in the waste disposal system. The control valve in the injec-tion line to the reactor coolant pump seals is automatically controlled by the T pressure differential controller connected to the reactor coolant system to 9-4 (Eevised 7-15-69) 0214
. maintain the' desired inlet presst e to the seals. The pressurizer makeup
. control valve is automatically controlled by the pressurizer level controller. 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. One of the three letdown control valves is designed for full letdown flow rate control at reduced reactor coolant system pressure.
The makeup pumps are contro11ed' remotely.
i 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 discharge 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, full-capacity letdown control valves and two, dual-unit, full-capacity letdown coolers for the nuclear unit to insure the flow capabil- 5 ity needed to adjust boric acid concentration. Two full-capacity seal return coolers are supplied for the unit.
A spare purification demineralizer is provided.
Three makeup pumps are supplied; one is capable of supplying the required reac- 2 tor coolant pump seal, and makeup flow. The letdown coolers transfer heat to the intermediate cooling water system and the seal return coolers transfer heat lT '
to the nuclear services cooling water system.
9.1.2.5 Codes and Standards The eouipment 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 - ASME Section III-C Makeup Tank - ASME Section III-C 9.1.2.6 System Isolation The letdown.line and the reactor coolant pump seal return line penetrate the reactor' building. Both lines contain electric motor-operated isolation valves inside the reactor building and pneumatic valves outside which are automati-cally closed with operation of the engineered safeguards.
Four emergency injtetion lines are used for injecting coolant to the reactor 2 vessel after a loss-of-coolant incident. Check valves in the discharge of each O
J 9-5 (Revised 7-15-69) 0215
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makeup pu p provide further backup for reactor building isolation if required.
After use of the lines for emergency injection is discontinued, the pneu=atic g valves in each line outside the reactor building are closed remotely by the control room operators.
9 1.2.7 Leakage considerations Reactor coolant is normally let down to this system. The purification demin-eralizer vill remove essentially 100 percent of the ionic and solid contami-nants except for boric acid, while gaseous contaminants will 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 equipment in this sys-tem is shielded by concrete. Shielding design criteria are discussed further in Section ll.
9... 8 operating conditions The makeup tank will be maintained with a fluid inventory between 100 ft3 and 500 ft3 Oxygen accumulatien in the tank vill be less than 2 percent by volume. One dual letdown cooler and two makeup pu=ps will be functional at all times.
To prevent an inadvertent excessive dilution of the reactor coolant boric acid concentration, three safety measures are applied to each of the two methods of diluting, i.e. , the bleed and feed method and the deborating demineralizer g method. The first safety measure is a 70 gpm limitation on the maximum rate of adding demineralized water; for feed and bleed the demineralized water makeup control valve to the makeup tank is automatically controlled to prevent exceeding a preset flow rate; and for deborating through the demineralizer, the three-way valve position is automatically changed to stop dilution if the demineralizer flow rate exceeds the preset flow rate. The second safety mea-sure is a control rod assembly position interlock which either permits or pro-hibits dilution depending on the control rod pattern. Because of this inter-lock, the demineralized water makeup valve and the deborating demineralizer inlet isolation valve can be opened only when the control rod assemblies are withdrawn to a preset position. The demineralized water makeup valve is also automatically closed, and the three-way valve position is automatically changed when the rods have been inserted to a preset pcsition. The third safety mea-sure consists of closing all the valves as described above when the flow has integrated to a preset value. Initiation of dilution must be by the operator, and the operator can terminate dilution at any time.
0216 9 9-6 l
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Table 9-1 1
(. / Makeup and Purification System Performance Data ,
Letdown Flow Maximum (cold), gpm 140 l1 Total Flow to Each Rea'etor Coolant Pump
~
2 Seal, gpm 45-50
! Seal Inleakage to Reactor Coolant System per Reactor Coolant Pump, gpm 4-6 l1 Injection Pressure to Reactor Coolant Pump Seals at Startup, psig 850-2,235 l1 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, F 125 7
Deleted b) l1 Purification Letdown Fluid Temperature,
! F 120 Makeup Tank Normal Operating pressure, psig 15 MakeupTankVolumeBetweenMjnimumand Maximum Operating Levels, ft 400 Reactor Coolant Water Quality See Table 9-4 9-7 (Revised 7-15-69) n tJ Oj?17 L
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Table 9-2 Makeup and Purification System Equipment Data (Capacities are for single components.) l5 Makeup Pump Quantity 3 l5 Type Multistage centrifugel, mechanical seal Capacity, gpm See Figure 6-2 Head, ft H 3 0 at sp. gr. = 1 See Figure 6-2 Motor Horsepower, hp 600 Pump Material SS wetted parts 2 Design Pressure, psig 2,850 Design Temperature, F 200 Letdown Cooler Quantity 2 dual-unit, full-capacity l5 Type Shell and Spiral tube 6
Heat Transferred, Btu /hr 16.1xlg l1 Letdown Flow, Ib/hr 3.5 x 10 Letdown Temperature Change, F 555 to 120 Material, shell/ tube CS/SS Design Pressure, psig 2,500 Design Temperature, F 600 Seal Return Cooler Quantity 2 full capacity l5 Type Shellangtube Heat Transferred, Btu /hr 2.2 x 10 Seal Return Flow, lb/hr 1.025 x 10 5 1 Seal Return Temperature Change, F 144 to 122 Material, shell/ tube CS/SS Design Pressure, psig 150 Design Temperature. F 250 l7 5
Cooling Water Flow, lb/hr 1.025 x 10 Makeup Tank Quantity 3
1 l5 Volume, ft 600 Design Pressure, psig 100 Design Temperature, F 200 Material SS l1 Purification Demineralizer Quantity 2 5 Type Mixed bed, boric acid saturated Cation:Anton Ratio 2:1 Material SS Resin Volume, ft 40 Flow, gpm 70 Vessel Design Pressure, psig 150 1 Vessel Design Temperature, F 200 9-8 (Eevised 7-15-69)g
92 CHDiICAL ADDITION AND SAFSLING SYSTEM p) ( v 9 2.1 Lesion SAsEs 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 sys-te=s. The system shown on Figure 9-3 is designed to add boric acid to the re-actor coolant system for reactivity control (see Table 3-5 and Figure 3-1),
potassium hydroxide for pH control, and hydrogen c" hydrazine for oxygen con-trol. The system is designed to take reactor coolant samples and steam genera-tor water sa=ples. 9 2.1.2 Boric Acid Mix Tank A single boric acid mix tank is provided as a source of concentrated boric acid solution. The volume of the tank will provide sufficient boric acid solution to increase the boron concentration of the reactor coolant system to that re-quired for cold shutdown. Heaters in the tank maintain the temperature above that required to insure solubility of the boric acid. Transfer lines will be electrically traced. 9.2.1.3 Boric Acid Pumps Two boric acid pumps are provided to facilitate transfer of the concentrated 1 boric acid solution from the boric acid mix tank to the borated water storage tanks, the makeup tanks, 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 reactor coolant system in 12 hours.
A third beric acid pu=p (high head, low capacity) is provided to effect make-up of the concentrated boric acid solution to the core flooding tanks. 9.2.1.h caustic Mix Tank The volume of the caustic mix tank was established so that a NaOH solution of sufficient concentration could be =aintained at room temperature to neutralize a vaste neutralization tank full of reactor coolant containing a boric acid con-centration equivalent to that used during refueling, or to regenerate a deborat-ing decineralizer. 9 2.1 5 Caustic Pump The caustic pump capacity is set so that at the maximum capacity, the foregoing neutralization operation can be performed in 15 minutes. 9 2.1.6 Potassium Hydroxide Mix Tank The tank volu=e was established to contain a sufficient amounc of KOH for con-tinual addition to the reactor coolant system so that a concentration of 3-6 ppm can be maintained while letting down at the =axi=um rate. f t.)3 02:19 l 9-9 (Revised 1-15-68) l
9.2.2 SYSTEM DESCRIPTION AND EVALUATION 9.2.2.1 Schematic Diarram and Syste= Descrittien O Figure 9-3 is a schematie diagram illustrating the features of the system. The system is operated from local controls, Two boric acid pumps, cornected in parallel, take suction from the boric acid mix tank and discharge to either the spent fuel storage pool, borated water storage tank, or the makeup tank. At the end of core life, both boric acid pumps are required to raise the reac-tor coolant system boron concentration from the minimum end-of-life concentra-tion to the refueling concentration in approximately 12 hours. The boric acid mix tank has a mechanical mixing device and a heating unit. The potassium hydroxide equipment consists of a mix tank, a single positive displacement pump, and connecting piping. The pump discharges to the makeup tank. A hydrazine drum is connected to a positive displacement pump, which discharges to a line leading to the makeup tanks. A nitrogen blanket is used to displace the hydrazine as it is removed from the dru=s. A hydrogen supply manifold with controls and a distribution line is used to supply the desired overpressure in the makeup tanks during operation. A nitrogen supply manifold with controls and distribution lines is used to sup-ply a gas blanket or a gas purge for the makeup tanks, sodium thiosulfate mix tank, core flooding tanks, hydrazine drum, liquid vaste disposal tanks, and gas vaste disposal tanks. A caustic mix tank and pump are used to supply chemicals for pH adjustment in the liquid vaste disposal system, and for regenerating the deborating deminera-liners. The liquid sampling portion of the system receives samples of the resctor cool-ant from upstream and downstream of the purification demineralizers, from up-stream of the letdown coolers, from the makeup tanks, and from the secondary side of the steam generators. Water qualities to be maintained are listed in Tables 9-3 and 9 L. Gaseous samples are taken from the pressurizer vapor spaces and from the makeup tanks. ' mple lines from these points are piped to a sam-pling cubicle outside the v .. tor building. Samples are collected in containers designed for full operating temperature and pressure at flow rates of 1 and 2 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 mixtures and to alarm at a preset level. 0220 9 9-10
The pertinent para =eters for each major component in the chemical addition (7 and sanpling system are shown in Table 9-5 V 9 2.2.2 Performance Requirements This system permits sa;apling of, and chemical addition to, ' the reactor - coolant system and the reactor auxiliary systems, during nor=al operation and tas no active emergency function. During a loss-of-coolant accident, this system ic isolated at the reactor building boundary. 9 2.2 3 Mode of operation The system is capable of drawing reactor coolant samples during reactor operation 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 waste disposal system, is accomplished locally. Equipment for sampling non-radioactive fluido is separated from 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 waste disposal system. During nor=al operation, liquid and vapor samples may be taken from the following points: Liquid
- a. Steam generator secondary water,
- b. Reactor coolant system.
- c. Purification demineralizer inlet.
- d. Purification demineralizer outlet.
- e. Deborating demineralizer outlet.
- f. Makeup tank.
- g. Core flooding tanks. 1 Vapor and Gas
- a. Pressurizer.
- b. Makeup tank.
i In addition, an oxygen and hydrogen analyzer automatically samples the gas spaces in the vaste disposal system tanks and equipment in an automatic sequence. The makeup tank gas space can also be analyzed with this unit. l' A (j During normal-operatior., this system also delivers the following chemicals: 9-11 (Revised 1-15-68) 0_221
- c. Boric acid to the spent fuel storage pool, the borated water storage tank, core flooding tanks, and the makeup tank.
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- b. Caustic to a vaste neutralization tank and to the deborating demin-eralizer for resin regeneration. h
- c. Pctassium hydroxide to the makeup tank.
- d. Hydrazine to the makeup tank.
- e. Hydrogen to the makeup tank.
- f. Ilitrogen as required for the core flooding tanks, makeup tank, hydra-zine drums, sodium thiosulfate storage tank, nuclear services cooling water system surge tank, and tanks and equipment in the vaste dispoi.al system.
9.2.2.L Reliability Considerations Tne system is not required to function during an emergency, nor is it required to take action to prevent ar emergency condition. It is therefore designed to perform in accordance with standard practice of the chemical process industry with duplicate equipment such as pumps and high pressure gas regulating valves es required. 9.2.2.5 Codes and Standards The equipment in this system will be designed to applicable codes and standards tabulated in Section 9. Equipment applicable to the ASME Codes are: the Re-cetor Coolant Sample Cooler which will be designed to ASME Section III, Class C, and. the Steam Generator Sample Cooler which will be designed to ASME 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 5.2 and 7. 9.2.2.7 Leakage Considerations Leakage of radioactive reactor coolant from this system within the reactor buildir.g vill be collected in the reactor building sump. Leakage of radioac-tive 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 portien of this system delivers additives to the spent fuel storage pool and the makeup tanks. Additives to the spent fuel storage pool are delivered above the water level. Backflow from the makeup tank to the positive displacement pumps is prevented by a check valve and a remotely operated valve between them. Backflow fro = a makeup tank through the hydrogen addition line is prevented by a check valve and a remote manual hydrogen addi-tion valve. O 9-12 (Revised 1-15 68) 0222 l
j- . 9 2.2.9 Failure Considerations V To evaluate system safety, the following failures or =alfunctions were assumed concurrent with a loss-of-coolant accident, and the consequences were analyzed. As a result of this evaluation, it is concluded that proper consideration has been given to Plant safety 'in the design of the system. Co=ments and Component Failure Consequences Pressurizer Sample Electrically operated Diaphragm-operated sampling valve inside valve outside the reactor building fails reactor building vill to close on ES signal. close. Reactor Letdown Sample Electrically operated Same as above, sa=pling valve inside reactor building fails to close on ES signal. Steam Generator Steam Diaphragm-operated Sample line is not Sample sampling valve outside connected directly to reactor building fails reactor coolant sys-to close on "' signal. tem, and steam gen-rS erator therefore pro-(_) vides first barrier. Sample Line Frcm Any One Line breaks inside Diaphragm-operated of the Three Preceding reactor building valves outside reac-Components downstream of EMO tor building close on valves. signal from ES system. 9 2.2 9 operating Conditions 9 2.2 9 1 Boric Acid Concentration The boric acid mix tank is to be =aintained at an average temperature of 95 F to maintain a boric acid concentration of 7 per cent. 9 2.2 9 2 Coolant Sample Te=perature
- The high pressure reactor coolant samples leaving the reactor coolant sa=-
ple cooler should be held to a- te=perature of 200 F to minimize the genera-tion of radioactive aerosols.
)
x./ 9-13 e
Table 9-3 E'.eam Generator Feedwater Quality Parameter Value Maxi =um Total Dissolved Solids, ppm 0.05 Suspended Solids, ppm 0.0 Hardness 0.0 Organic, pp= 0.0 Maximum Dissolved Oxygen, pps 0.007 Carbon Dioxide 0.0 Maxi =um Total Silica (as SiO2), ppm 0.02 Maximum Total Iron (as Fe), ppm 0.01(* Maxiaum Total Copper (as Cu), ppm 0.002(*} l1 pH 9.3 to 9.5 Lead and Heavy Metals None l1 (*) Included in maximum TDS as a soluble compound. Table 9-4 Reactor Coolant Quality h Parameter Value Total Solids, max. (excluding H)B05 and KOH), ppm 1.0 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 ), pp= 0.1 H2 , std cc/l 15-40 Hydrazine (required during shutdown), ppm 25 0 9-14 (Revised 1-15-68) l
Table 9-5 Chemical Addition and Sampling System Equipcent Data
' x/
f'% s (Capacities are for single components.) -l 5 Tanks - Boric Acid Mix Tank Quantity 1 Type Vertical Cylindrical Volume, ft3 2,050 ' Design Pressure, psig Atmospheric Design Te=perature, F 200 ~ Material Al Potassium Hydroxide Mix Tank Quantity 1 Type Vertical Cylindr'. cal Volume, gal 50 Design Pressure, psig Atmospheric Design Temperature, F 200 Material SS Caustic Mix Tank Quantity 1 Type- Vertical Cylindrical Volume, gal 500 (~'\ Design Pressure, psig At=ospheric Design Temperature, F 200 Material CS Hydrazine Drum ' Quantity 1 Type Std. Commercial 55 gal Drum Pumps Boric Acid Pump Quantity 2 Type Reciprocating, Variable Stroke Capacity, gpm 0-10 Head, psi 50 I Design Pressure, psig 100 Design Temperature, F 200 Pu=p Material SS Ecric Acid Pump 1 quantity 1 Type Beciprocating, Variable Stroke ! Capacity, gpm 1 ! Head, psi- 630 Design Pressure, psig 700 Design Temperature, F 300 i
,e s Pump Material SS
\_
9-15 (Revised h-8-68) 0225
Table 9-5 (Cont'd) Potassium Hydruxide Pump h; Quantity 1 J Type Reciprocating, Variable Stroke Capacity, gph 0-10 Head, psi 50 Design Pressure, psig 100 Design Te=perature, F 100 Pu=p Material SS Hydrazine Pump ! Quantity 1 l Type Reciprocating, Variable Stroke Capacity, gph 0-10 Head, psi 50 Design Pressure, psig 100 Design Temperature, F 100 l Pump Material SS Caustic Pump Quantity 1 Type Reciprocating, Variable Stroke , Capacity, gph 0-600 l Head, psi 10 Design Pressure, psig 25 Design Temperature, F 100 Pump Material CS Sa=pling Sa=pling Containers Quantity Design Pressure, psig 6 2,500 l5 Design Temperature, F 670 Reactor Coolant Sample Cooler Quantity 1 l5 Type Shell and Spiral Tube HeatTransferred, Btu /hr 2.1 x 105 Sample Flow Rate, gp 1 Max. Sa=ple Inlet Te=perature, F 650 Sa:Ple Outlet Temperature, F 1% 0226 0 9-16 (Revised h-8-68) L_
Table 9-5 (Cont'd) 3
's Cooling Water Flow, lb/hr 5 x 10
(# Coil Side Design Temperature, F 6TO Coil Side Design Pressure, psig 2,500
. Steam Generator Sample. Cooler Quantity 1 ' l 5~,
Type Shell and Spiral Tube Heat Transferred, Etu/hr 2.0 x 105 Sample Flow Eate, gym 1 Sample Inlet Temperature, F 525 Sample Outlet Temperature, F 150 3 Cooling Water Flow, lb/hr 5 x 10 Coil Side Design Temperature, F 600 Coil Side Design Fressure, psig 1,050
-3 (u s's 9 /~S l u 0227 9-1T (Revised h-8-68) 0227
9.3 COOLING WATER SYSTEMS 9.3.1 DESIGN BASIS O The cooling water systems are arranged in three separate pumping systems, 2 each having sufficient redundancy to operate with loss of one pump and/or one cooler (rav sea water to closed cycle water).
- a. A nuclear services cooling water system and intermediate cooling vater system cool all nuclear cycle and fuel handling requirements.
These systems have sufficient redundancy from the heat source to the heat sink to irsure continuous heat removal from components requir-ing cooling during normal and emergency conditions.
- b. A secondary services cooling water system cools all nonnuclear re-lated cooling water requiremente. 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 condenser which serves both the main turbine unit and l5 the feedvater pump turbines. On loss of off-site electrical power, this system will not be operated. The cycle cool-down requirements are adequately handled by other means.
These systems vill be sized to insure adequate heat removal based on highest expected temperatures of cooling water, maximum loadings, and leakage allow-ances. The equipment in these systems will be designed to applicable codes and standards tabulated in Section 9, page 9-1. g All cooling water systems will be designed to prevent a component failure from curtailing normal unit operation. It will be possible to isolate all heat ex-changers and pumps. Each pressure-reducing valve vill be prcvided with a by-pass. All systems will be monitored and operated from the control room. Isolation valves located external to the Reactor Building vill be incorporated in all cooling service water lines penetrating the Reactor Building. Activity monitors will be installed in cooling water return headers. 2 Electrical power requirements for the nuclear services cooling water system can be supplied by any of the redundant power sources described in 8.2.3. Neither the secondary services cooling water system nor the condenser circulating water system are intended to be crerated frcs the diesel generatcr er the engineered safeguards transformer power sources. All system conpenents will be hydrostatically tested prior to unit startup and vill be accessible for periodic inspecticns 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 9-le (Revised h-6-6S) 0228
9.3.2 SYSTEM DESCRIPTION AND EVALUATION 9.3.2.1 Nuclear Services and Inter =ediate Cooling water Systems 2 All service's cooled by these closed systems (Figures 9 h, 9-ha, and 9-5) vill
, be of a nuclear nature and,-hence, segregated to these systems. Each system
~ ~
vill provide cooling water for the following equipment: Nuclear Services Coolers - serving: Normal - makeup pump motor cooler and pump bearing cooler
- standby makeup pump motor coolers and pump bearing coolers
- spent fuel coolers
- reactor coolant pump seal return coolers
- evaporator condensers
- sample coolers
- reactor coolant pump motor coolers
- reactor coolant pump oil coolers
- fan motor coolers
- instrument air compressor and instru-ment air compressor after-cooler
- emergency feedwater pump cooler
- control building air conditioning Emergency - reactor building emergency coolers
- spent fuel coolers
(~ ) - control building air conditioning
- fan motor coolers
- instrument air compressor and after-cooler
- makeup pump motor cooler and pump bearing cooler
- emergency feedvater pump cooler Decay Heat Service Coolers - serving:
Normal - decay heat cooler (shutdown)
- decay heat pump motor coolers and bearing coolers
- decay heat closed cycle cooling water pump motor coolers and bearing coclers Emergency - decay heat coolers
- decay heat pump motor coolers and bearing coolers
- decay heat closed cycle cooling water pump motor coolers and bearing coolers
- makeup pump (high pressure injection) motor coolers and bearing c- 61ers
- spray pump motor coolers ana 'aring coolers Intermediate Coolers - serving: - letdown coolers
- reactor coolant pump seal area l
{~/} s_
- concrete shield cooling (if required)
- reactor coolant drain tank
, 9-19 (Revised 2-7-68)
In an emergency, the intermediate coolers will be shut off and the flow diverted from the normal services to the emergency services on the nuclear services cooler cycle. Redundancy is obtained as follows: 2 3 - Nuclear Services Sea Water Pumps 1 run for normal cooling 2 run for emergency cooling h - Nuclear Services Coolers 1 operated for normal cooling 3 operated for emergency cooling 3 - Nuclear Services Closed Cycle Cooling Water Pu=ps 1 run for normal cooling 2 run for emergency cooling 2 - Decay Heat Service Cooler Sea Water Pumps 1 operated for normal shutdown 1 operated for emergency cooling 2 - Decay Heat Service Coolers 1 operated for normal shutdown 1 operated for emergency cooling 2 - Decay Heat Closed Cycle Cooling Water Pumps 1 operated for normal shutdown 1 operated for emergency cooling 2 - Intermediate Coolers 1 operat2d for norral cooling 2 - Intermediate. Closed Cycle Cooling Water Pumps 1 run for normal cooling O 9-19a (Revised 2-7-68) 0230
2 The nuclear services sea water pumps and the decay heat service cooler sea water pumps will be located in the auxiliary building. Rav sea water at max. 85 F vill be circulated in the nuclear service coolers and in the decay heat service coolers with 95 F max. closed cooling water circulated on the shell side by closed cycle cooling water pu=ps. This closed cooling water system represents a double barrier which will not allow radioactivity to enter the Gulf in case a primary cooler leaks. The closed loops vill be continuously monitored for radioactivity, and if any is detected, the systen vill be tested and the leaking cooler isolated. Radioactivity monitors vill also be installed in the sea water effluent headers. On the closed cycle portion of the above cooling systems, elevated surge tanks of sufficient capacity vill provide storage of condensate with make-up from the condensate system being automatically added. Abnormal tank levels vill be monitored in the control room. On receiving a radioactive monitor alarm, overflow will be diverted to vaste treatment. 9.3.2.2 Secondary Services cooling Water System This vill be a closed cooling water system (Figure 9-6) with three pumps available to pu=p 85 F raw sea water to three heat exchangers located in the turbine building where 95 F closed circuit cooling water is pumped through 2 the shell side of the coolers with three closed cycle pumps available. Sufficient redundancy vill be maintained so that loss of any one pu=p and/or cooler vill not affect normal operation. On the closed cycle portion of the system an elevated surge tank of 10,000 gal capacity will provide storage of condensate with makeup from the condensate system being automatically added. Abnormal tank levels vill be annunciated in the control room. All services served by this system vill be expendable in the event of an accident. This system can be dropped on unit trip at time of accident. No nuclear-oriented services vill be serviced by this system. 9.3.2.3 condenser circulating Water System Water from the Gulf of Mexico vill be used as the source of water for the con-denser circulating water (CCW) system. Figure 9-7 shows schematically the arrangement of the system. The intake structures vill be provided with a trash removal system. The cir-culating water pu=p intaxe structures vill have a suction extending below lov vater level and will be provided with traveling screens. The intake structures vill each have four circulating water pumps. O 9-20 (Revised 2-7-68) 0231 o
4 I h.'
' The :CCW system will be designed to take advantage of the syphon effect _ to reduce the static head. The system will be primed at startup by mechanical 2
- i. - O'-- '
vacuum pumps, these_ vacuum pumps will also serve for air re= oval in all the condenser-water boxes during normal operation.
~
- s. . .- ,
c a 4 e d k O 1 ? 1 1 i i 1-j. I 2 ( s 1 O 9-21 (Revised 2-7-68) 0237
- u... .. . .. .: . . - = . . _ , -
Table 9-6 Cooling Water Systems Performance and Eculpment Data (Capacities are for single components) g
)5 Nuclear Services Sea Water Pu=ps 2 Quantity 3 Flow, gpm 5 2-lk,000; l-10,000 Normal Design Pressure, psig 50 Design Temperature, F 85 Decay Heat Services Cooler Sea Water Pumps 2 Quantity 2 Flow,gpm 9,000 l5 Design Pressure, psig 50 Design Temperature , F 85 Nuclear Services Quantity k Type Shell and Tube Raw Sea Cooling Water Flov (tubeside), gpm h,660 [5 Rav Sea Cooling Water Temperature, F 85 Closed Cycle Cooling Water Outlet Temperature, F 95 Closed Cycle Cooling Water Flov (shellside), gym 3,000 l5 Design Pressure, shell/ tube, psig 100/100 Design Temperature, shell/ tube, F 200/150 Tube Material 90-10 Cu-Ni Shell Material Carbon Steel Decay Heat Service Coolers Quantity 2 2
Type Shell and 'fube Raw Sea Cooling Water Flov (tubeside), gpm 9,000 Rav Sea Cooling Water Temperature, F 85 l5 Closed Cycle Cooling Water Outlet Temperature, F 95 Closed Cycle Cooling Water Flov (shellside), gpm 3,000 Design Pressure, shell/ tube, psig 100/100 Design Temperature, shell/ tube, F 200/150 Tube Material 90/10 Cu-Ni Shell Material Carbon Steel Nuclear Services Closed Cycle Cooling Water Pumps Quantity 3 Flow, gpm 2-9,100; l-6,k00 5 Normal Design Pressure, psig 60 Design Temperature, F 95 Decay Heat Closed Cycle Cooling Water Fu=ps 2 Quantity 2 Flov, gpm 3,000 Design Pressure, psig 60 Design Temperature, F 95 9 9-22 (Revised h-8-68) 0233
~
Surge Tanks 2 Quantity 3, Total Capacity, gal
- () Design Te=perature, F one 10,000, two 5,000 120 Design Pressure, psig 75 Material Carbon Steel Secondary Servi--i Jualer Sea Water 5 Supplied by Main Condenser Circulating Pumps Secondary Services Coolers t
quantity 3 Type Shell and Tube Raw Sea Cooling Water Flow (tubeside), gym 9,000 Raw Sea Cooling Water Temperature, F 85 Closed Cycle Cooling Water Outlet Temperature, F 95 Closed Cycle Cooling Water Flow (shellside), gpm 4,500 Design Pressure, shell/ tube, psig 100/150 Design Temperature, shell/ tube, F 150/150 Tube Material 90-10 Cu-Ni Shell Material Carbon Steel 4 i I l O . 0234 ; l l 9-22a (Revised k-8-68)
, _ . - - . - - . - , - - . _ _. = - _. . . , - - - ,_
Table 9-6 (Cont'd) b) V 2 Secondary Servicec Closed Cycle Ccoling Water Pumps Quantity 3 Flov, cum h,500 Design Pressure, psig 60 - Design Temrerature, ? 95 Condenser Circulating Water Pumps Quantity h Flav. crm 240,000 Design Pressure, esig 50 Design Temperature, F 80 Table 9-7 Intermediate Cooling System Performance and Equipment Data (Capacities are for single components. ) Pumps Quantity 2 Full Capacity Capacity, cpm 650 Design Fressure, psig 100 Lesien Temuerature, F 225 q Coolers (g Quantity 2 Full Capacity Type Shell and Heat Transferred, Btu /hr 17.25 x 10gube Tube Side (rav sea water) Inlet Temperature, F 85 Outlet Temuerature, F 113 Flov Pate, epm 1,300 Design Pressure, psig 100 Design Temperature, F 225 Shell Side (intermediate cooling water) Inlet Temperature, F 148 Outlet Temperature, F 95 Flow Rate, spm 650 Design Pressure, psig 100 Design Temperature, F 225 Tube Material 90-10 Cu-Ni Shell Material Carbon Steel Surge Tank Quantity 1 Catacity. gal 375 Design Prescure, psic 100 Design Temperature, F 225 Material Carbon Steel Cs, 023*f l 9-23 (Revised 2-7-68) l
9.4 SPENT FUEL COOLING SYSTEM 9.4.1 DESIGN BASES The spent fuel cooling system is shown on Figure 9-8. It is designed to main- 5 tain the spent fuel storage pool at approximately 120 F vith a heat load based on removing the decay heat generation from 1/3 core, which has been irradiated for 930 days and cooled for 150 hours. In meeting the design bases above, the system has the additional capability to maintain the spent fuel storage pool at 143 F while removing the decay heat from the follovir combination of stored fuel assemblies:
- a. 1/3 core irradiated for 930 days and cooled for 100 days.
- b. 1/3 core irradiated for 720 days and cooled for 150 hours.
- c. 1/3 core irradiated for 410 days and cooled for 150 hours.
- d. 1/3 core irradiated for 100 days and cooled for 150 hours.
9.4.2 SYSTEM DESCRIPTION AND EVALUATION 9.4.2.1 Schematic Diagram The sche.satic dia6 ram for the spent fuel cooling system is shown in Figure 9-8. 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 both the spent fuel storage pool water contained in the fuel transfer canal during the refueling operation, and 3 the contents of the borated water storage tank (after it has been used in the fuel transfer canal during refueling). 9.h.2.2 Performance Beauirements The first design basis of the system predicates en operating schedule which is 5 on an equilibrium refueling period (310 FPD per cycle) with approximately 1/3 of a core being removed from the unit at the end of each period. The removed fuel assemblies will have been in the reactor for three cycles, i.e., 930 days at the time of discharge. The second design basis for the system considers that it is possible that during the life of the Plant it will be necessary to unload the reactor vessel totally 5 for maintenance or inspection at the time that the 1/3 core is already residing in the spent fuel storage pool. The basic system performance and equipment data are presented in Table 9-8. 9.h.2.3 Mode of Operation During normal conditions 1/3 of a ccre vill be stored in the pool and one of 5 the pumps and one of the coolers will handle the load and maintain 127 F. Op-eration of both pumps and coolers will maintain a pool temperature of 111 F. g{t Tg pool is initially filled with water from the borated water storage tant. sw T7.ay 9-2L (Revised h-8-68) JMt(P 0236
r"N' Fer'the case where 1-1/3 cores are stored due to complete unloading of the re- 5 (_/ acto: vessel, two pumps and two coolers will maintain the spent fuel storage pool temperature at ih3 F. -If both a pu=p and a cooler are out for maintenance when this storage condition exists, the water temperature will eventually rise to .190 F, although considerable time will be required to heat..the large spent fuel storage pool to this temperature. If all cooling is lost, the time re-quired for the spent fuel storage pool to reach 212 ? for each of the fore-going quantities of stored fuel is as follows: One-third of a. core 54 hours , 5 One and one-third cores 15 hours 9.4.2.h Feliability Considerations During the time when 1/3 core is stored in the pool, only one half of the in- 5 stalled equipment will be utilized to maintain the pool at 127 F. 9.h.2.5 Codes and Standards The equipment in this system will be designed to applicable codes and standards tabulated in Section 9 Ccmponents which are designed to the ASME Code are Spent Fuel Cooler - ASME Section III-C
/O
(-) Spent Fuel Coolant Demineralizer - ASME Section III-C 9.h.2.6 Leakage Considerations
'4henever 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 spant fuel cooling water. A small purification loop is provided for removing these fission products and otner contaminants from the water.
The fuel handling and storage area housing the spent fuel storage pool will 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 each fuel transfer tube for leak-tightness after it has been used. A valve and blind flange are used to isolate each fuel transfer tube. 9.h.2.7 Failure Considerations The most serious failure of this system would be complete loss of water in the storage 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.
.O t/
9-25 (Revised 4-8-68)
}}f L .
9.4.2.8 operating Conditions The pool vill normally be limited to lh3 F except in most unusual circumstances l5g as previously described. Boric acid concentration in the pool fluid will be w maintained at 12,000 to 13,000 pp= (2,090 to 2,270 ppm boron). Table 9-8 Spent Fuel Cooling System Performance and Eauipment Data (Equipment capacities are for single components.) l5 System Cooling Capacity, Btu /hr Normal (1/3 core) 8.75 x 106 5 Maximum (1-1/3 cores) 25.85 x 106 System Design Pressure, psig 75 System Design Temperature, F 250 Spent Fuel Cooler Data Quantity 2 Type Tube and Shell Material CS/SS Duty, Btu /hr k.375 x 106g7 = 111 F 5 1n 8.75 x 106@Th = 127 F 25.85 x 106@T h = 190 F Cooling Water Flow, lb/hr 7 5 x 105 Spent Fuel Pump Data Quantity 2 Flow, gpm 1,500 He4d, ft H 2O 100 Spent Fuel Coolant Demineralizer Flow Rate, spm 160 Bed Volume, ft3 20 Type Nonregenerative Vessel Material Carbon Steel - Lined Design Pressure, psig 75 Design Temperature, F 200 Spent Fuel Storage Pool Water Volume, ft3 87,500 l5 9-26 (Revised h-8-68) 0238 0
95 DECAY HEAT RD' OVAL SYSTE! n 951 DESIGN BASES V 9 5 1.1 General System Function
,. The normal function of this system as shown by Figure 9-9 is to remove reae, , ,.
tor decay heat during the latter stages of cooldown, maintain pactor coolant temperature 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. '. 2 Decay Heat Removal Pumus The decay heat removal pu=ps, during shutdown, circulate the reactor coolant from one reactor outlet line through the decay heat coolers and return it to l2 the reactor injection no::les. The design flov is that required to cool the reactor coolant system from 250 F to 140 F in 14 hours. (The steam generators are used to reduce the reactor coolant system from operating temperature to 250 F.) 9513 Decay Heat Removal Coolers The decay heat removal coolers, during shutdown, re=ove the decay heat from the circulated reactor coolant. At 20 hours after shutdown of the reactor (14 hours after reaching 250 F), two coolers and two pumps vill reduce the reac-tor coolant temperature to 140 F.
.p 952 SYSTEM DESCRIPIION AND EVALUATION v
9 5 2.1 Schematic Diagram The decay heat removal system is shown schematically in Figure 9-9 9 5 2.2 Performance Requirements Tables 9-9 and 9-10 at the end of this subsection list system performance data and design data for individual components. 9523 Mode of Operation Two pumps and two coolers perform the decay heat cooling function. After the steam generators have reduced the reactor coolant temperature to 250 F, decay heat 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 l only one pump or one cooler is available, the reactor coolant temperature is reduced at a lower rate. l ' The equipment utilized for decay heat cooling is also used for low pressure in-jection into the core during accident conditions. g 0239 V ' 9-27 (Revised 2-7-68)
9 5.2.h Reliability Consideration The nuclear unit has two pumps and two coolers. 9525 Codes and standards The equipment in this system will be designeu to applicable codes and standards tabulated in Section 9 The decay heat removal cooler which is applicable to the ASY2 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 tide through one air-operated gate valve and a check valve in series. All three of these va' es are normally closed whenever the reactor is in the operating condition. i the event of a loss-of-coolant accident, the valve on the discharge side opens, but the valves on the suct: )n 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.
952.8 Failure Considera'. ions Failure considerations for the accident case are evaluated and tabulated in 6.1 3 0240 9-26 (sevised 2-7-65) h
I Table 9-9 O Decay Heat Renoval System Perfomance Data Reactor Coolant Temperature at Startup
.of Decay Heat Removal, F 250 . .
Time to Cool Reactor Coolant System From 250 F to 140 F, hr 14 Refueling Temperature, F 140 Decay Heat Generation Figure 9-10 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 O O 9-29 0241 l
O Table 9-10 Decay Heat Removal System Equipmen+. Data (Quantities are given for one nuclear unit. Capacities are given for single components.) Pumps 2 Number 2 Capacity, gpm 3,000 Design Pressure, psis 300 Design Temperature, F 300 Coolers (a) Number 2 Type Shell and Tube Heat Transferred, Btu /hr 32 5 x 10 Reactor Coolant Flow, gpm 3,000 Cooling Water Flow, gpm 3,000 Cooling Water Inlet Temperature, F 95 Material,shell/ tube CS/SS 1 Design Pressure, shell/ tube, psi 6 150/300 Desi5n Te=perature, F 300 Borated Water Storage Tank Number 1 Capacity, gal 350,000 Material Al Design Pressure Hydrostatic Head Design Temperature, F 150 ("} Refer to Figure 6-h for heat transferred as a function of cooler inlet vater te=perature. 9-30 (Revised 2-7-68) -
,_ 9.6 FUEL HANDLING SYSTEM t \~h' 9.6.1 DESIGN BASES 9.6.1.1 General System Fun:tien The fuel handling system (Figure 9-11) is designed to provide a safe, effective means of transpcrting and handling fuel frcs the Lime it reaches the Plant in an unirradiated condition until it leaves the Plant after pestirradiation cool-ing. The system is designed to minimize the possibility c,f mishandling or mal-operations that coul' cause fuel assembly damage and/or potential fission prod-uct release. 5 The reactor is refueled with equipment designed to handle the spent fuel as-se=blies under vater fram the time they leave the reacter 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 mediun for removal of decay heat. Borated water insures suberitical conditiens 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. 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 program. The new fuel assemblies are stored {5 in racks in paralled revs having a center-to-center distance of 21 in. in both - (- ) directions. This spacing is sufficient to maintain a keft of less than 0.9 when vet. 9.6.1.3 Spent Fuel Storage Pool The spent fuel storage pool is a reinfoeced concrete pool lined with stainless steel; it is located in the fuel storage building. The pool is sized to ac-commodate 250 spent fuel assemblies which allows for a full core of irradiated l5 fuel aseemblies in addition to the concurrent storage of the largest quantity of spent fuel assemblies from the reactor as established by the fuel manage- l5 ment program. The spent fuel assemblies are stored in racks in paralled rows having a center-to-center distance of 21 in. in both directions. Control rod assemblies requiring removal from the reactor are stored in the spent fuel as-semblies. l5 9.6.1.4 Fuel Transfer Tubes Two horizontal tubes are provided to convey fuel between the reactor building and the fuel storage building. These tubes contain tracks for the fuel transfer carriages, gate valves en the spent fuel storage pool side, and a means for flanged closure on the reactor building side. The fuel transfer tubes penetrate into the fuel transfer canals at their lover depths , where space is provided for the rotation of the fuel transfer carriage basket con-taining a fuel assembly. O .. 0243 1 9-31 (Revised h-8-68) ' i l 1
9.6.1.5 N el Transfer Canal The fuel transfer canal is a passageway in the reactor building extending from the reactor vessel to the reactor building wall. It is formed by an upward ex-tension of fhe primary shield walls. The enclosure is a reinforced concrete structure lined with stainless steel; it forms a canal above the reactor vessel, which is filled with borated water for refueling. Space is available in the fuel transfer canal for underwater storage of the reactor vessel internals upper plenum asse=bly. The deeper fuel transfer station portion of the fuel transfer canal contains the new fuel handling racks. This portion can also be used for storage of the reactor vessel internals core barrel and thermal shield assemblies by tempo-rarily removing the new fuel handling racks. 9.6.1.6 Miscellaneous Fuel Handling Equipment This equipment consists of fuel handling bridges, fuel handling tools, new fuel storage racks, spent fuel storage racks, new fuel handling racks, fuel transfer containers, control rod handling tools, viewing equipment, fuel transfer mech-anisms, and shipping casks. In addition to the equipment directly associated with the handling of fuel, equipment is provided for handling the reactor clo-sure head and the upper plenum assembly to expose the core for refueling. 9.6.2 SYSTEM DESCRIPTION AND EVALUATION 9.6.2.1 Receiving and Storing Fuel New fuel assemblies are received in shipping containers and stored dry in racks having a center-to-center distance of at least 21 in. They are subsequently moved into the reactor building in one of the following 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 and storel directly in the new fuel handling racks in the transfer canal.
- b. After reactor shutdown, new fuel assemblies can be transferred from the new fuel storage area to the new fuel handling racks in the transfer canal by way of the spent fuel storage pool with the use of the fuel transfer carriages and the fuel transfer tubes.
9.6.2.2 Loading and Removing Fue' Following the reactor shutdown and reactor building entry, the refueling proce-dure is begun by removing tne reactor elesure head and control rod drives as-sembly. head removal c.nl replacement time is minimized by the use of two stud tensioners. The stud ticsioner is a hydraulically operated device that permits preloading and unloading cf the reactor closure studs at cold shutdown conditions. The studs are tensioned to their cperational load in two steps in a predeter-mined sequence. Required stud elongation after tensioning is verified by mi-crometer messurements.
'O
*~
_ 0244
- Following re= oval of the studs fro = the reactor vessel tapped holes, the
( ) studa and nuts are supported in the closure head bolt holes with specially designed spacers. Rc= oval of the studs with the reactor closure head =in-i=1zes handling ti=e and reduces the chance of thread da= age. The reactor closure head assembly is handled by a lifting' fixture support-ed fro = the reactor building crane. It is lifted out of the canal onto a head storage stand located on the operating floor. The stand is designed to protect the gasket surface of the closure head. The lift is guided by three closure head align =ent pins installed in three of the stud holes. These pins 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 re=oved frc= the reactor closure head at the storage location for inspection and cleaning using special stud and nut handling fixtures. A stud and align =ent pin storage rack is provided. The annular space between the reactor vessel flange and the botto= of the fuel transfer canal is sealed off, before the canal is filled, by a seal cla= ped to the canal shield plate flange and the reactor vessel flarge. The fuel transfer canal is then filled with borated water. The upper plenu= assembly is re=oved frc= 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.
,'~, ' Refueling operations are carried out fro = two fuel handling bridges which i span the fuel transfer canal. One bridge is used to shuttle spert fuel as-semblies fro = the core to the transfer station and new fuel assemblies fro =
the new fuel handling racks to the core. During this operation, the second bridge is occupied with relocating partially spent fuel assemblies in the core as specified by the fuel management progra=. Fuel asse=blies are handled by a pneu=atically operated fuel handling tool attached to a telescoping and rotating =ast which coves laterally on each bridge. Control rod asse=blies are handled by a control rod handling tool attached to a second cast located on one of the bridges in the reac-tot building. The two-=ast bridge = oves a spent fuel asse=bly frc= the core under water to the transfer station where the fuel asse=bly is lowered into the fuel transfer carriage fuel basket. The control rod handling tool attached to the second =ast is used to transfer a control rod asse=bly to a new fuel assembly in the adjacent new fuel handling racks. This new fuel asse=- bly with control rod asse=bly is carried to the reactor by the fuel han-dling tool and located in the core while the spent fuel assembly is being transferred to the spent fuel storage pool. Spent fuel asse=blies removed frc= the reactor are transported to the spent fuel storage pool frc= the reactor building via a fuel transfer tube by
=eans of a cable-operated fuel transfer carriage. The spent fuel asse=-
blies are re=oved frc= the fuel transfer carriage basket using a pneu=ati-
.]
cally operated fuel handling tool attached to a =ovabla =ast located on a 9-33
~ '
fuel handling bridge. This motor-driven bridge spans the spent fuel stor-age pool and pemits the refueling crew to store er remove new fuel asses-blies in any one of the many vertical storage rack positions. The fuel transfer mechanis=s are undervater cable-driven carriages that run on tracks extending fro = the spent fuel storage pool through the transfer tubes and into the reactor building. Each of the two fuel trans-fer mechanisms is independently operated so that either can be used for the fuel assembly transfer operations. 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 being used for refueling is rotated to a horizontal position for passage through the transfer tube, and then rotated back to a verti-cal position in the spent fuel storage pool for vertical removal of the fuel acsembly. Once refueling is completed, the fuel transfer canal vater is drained by suction through a pipe located in the deep transfer station area. The canal water is pumped to the borated water storage tank to be available for the next refueling or for e=ergency cooling following a loss-of-coolant accident. During operation of the rcntor, the carri4es are stored in the spent fuel storage pool, thus permitting gate valves on the spent fuel storage pool side of each transfer tube to be closed and blind flanges to be in-stalled on the reactor building side of the tube. The spent fuel storage pool has space for a spent fuel shipping cask, as well as for required fuel storage. Following a sufficient decay period, the spent fuel asse=blies are removed from storage and loaded into the spent fuel shipping cask under water for renoval 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 adjacent to the spent . fuel storage pool; in this area the outside surfaces of the casks can be decontaminated before ship =ent by using steam, water, or detergent solu-tions, and manual scrubbing to the extent required. 9 6.2 3 safety Provisions Safety provisions are designed into the fuel handling syste= to prevent the develop =ent of hazardous conditions in the event cf component malfunc-tions, accidental damage, or operational and administrative failures dur-ing refueling or transfer operations. All fuel asse=bly storage facilities, new and spent, =aintain an eversafe geometric spaci.g of 21 in, between asser.:blies. The new and spent fuel storace racks are designed so that it is impossible to insert fuel asser-blies in other than the prescribed locations, thereby insuring the neces-sary spacing between assemblies. Although new fuel assemblies are stored dry, the 21 x 21 in. spacing insures an eversafe gec=etric array in un-borated water. Under these conditions, a criticality accident during re-fueling or storage is not considered credible. O
,_s
~.[ ,f ,4 '.) ', 9-34 fW 0246
p All fuel handling and transfer containers are al.co designed to =aintain
\v an eversafe geo=etric array. Mechanical da= age to the fuel asse=blies during transfer operations is possible, although re=ote. Since the fis-sien product release vould occur under vater, the enount of activity reaching the environ =ent.Will present no appreciable hazard. A fuel han-dling accident analysis is included in Section 14.
All spent fuel assembly transfer operations are conducted under water. The water level in the fuel transfer canal provides a =ini=u= of 10 ft of water over the active fuel line of the spent fuel asse=blies during =ove-nent frcm the core into storage; this limits radiation at the surface of the water to less than 10 crem/hr. The spent fuel storage racks are lo-cated to provide a minimum of 13 ft of water shielding over stored assem-blies to limit radiation at the surface of the water to no = ore than 2 5
= rem /hrduringthestorageperiod. The depth of the water over the fuel assemblies, as well as the thickness of the concrete valls of the trans-fer canal, is sufficient .to 11=it 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 decay heat re= oval system described in 9 5 In case of a power fail-ure, this system vill be operated by the auxiliary power supply. The spent fuel storage pool water is cooled by the spent fuel cooling system as described in 9.4. A power failure during the refueling cycle vill create no i= mediate hazardous condition owing to the large water volume in both the fuel transfer canal and spent fuel storage pool. With a nor-hV =al quantity of spent fuel assemblies in the storage pool and no cooling available, the water temperature in the spent fuel storage pool would in-crease as discussed in 9 4.2 3 During the refueling period the water level in both the fuel transfer canal and the spent fuel storage pool is the same, and the fuel transfer tube valves are continuously open. This eliminates the necessity for interlocks between the fuel transfer carriages and fuel transfer tube valve operations. The simplified movement of a transfer carriage through the horizontal fuel transfer tubes minimizes the danger of Jamming or de-1 railing. To cope with such an eventuality, the open tube design provides access to the entire length of the fuel transfer carriage travel from the fuel transfer canal. All operating mechanisms of the system are located in the fuel storage building for ease of maintenance and accessibility for incpection before the start of refueling operations.
- During reactor operation, bolted and gasketed closure plates, located on j
the reactor building flanges of the fuel transfer tubes, prevent leakage of water from the spent fuel storage pool into the transfer canal in the event of a leak through the fuel transfer tube valves. Both the spent fuel storage pool and the fuel transfer canals are co=pletely lined with stainless steel for leax-tightness and ease of decontamination. The fuel transfer tubes vill be appropriately attached to these liners to maintain leak integrity. .The spent fuel storage pool cannot be accidentally drained l siace water must be pu= ped out through a suction pipe. The fuel transfer mechanis=s are designed to per=it initiation of the carriage travel and
= p., the carriage fuel basket rotation frcm the building in which the carriage
(,1 fuel basket is being loaded or unloaded. 9-35 m7
A31 electrical gear is located above water for greater integrity and ease of maintenance. The hydraulic systems that actuate the rotating fuel baskets use g storage pool vater for operation to eliminate contamination. The fuel transfer canal and storage pool water will have a boron concentration of 2,270 ppm. Although this concentration is sufficient to maintain core shut-down if all of the control rod asse=blies verc removed from the core, only a few control rods will 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. The 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. 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 accormodate the expected fission gas release from the fuel. Any leaking fuel asse=blies 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 and co=pletely seals off the leaking fuel assembly before a fuel transfer mechanism transfers it out of the fuel transfer canal into the llg spent fuel storage pool. The design of the failed fuel containers will comply with 10 CFR 71 so that defective fuel assemblies can be safely stored and shipped while sealed in the failed fuel container. 9.6.2.4 operational Limits Certain manipulations of the fuel asse=blies and reactor internals during re-fueling may result in short-term exposures with radiation levels greater than 2 5 mres/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 li=ited to handling of fuel and control rod as-se=blies and reactor closure head studs only. All lifts for handling the re-actor closure head and reactor internals vill use the reactor building crane. Travel speeds for the fuel handling bridges, masts, and fuel transfer carriages vill be controlled to insure safe handling conditions. O 0248 9-36
9.7 PLANT VENTILATION SYSTEMS 9.7.1 DESIGN BASES {) The Plant vill be designed to provide maxtsun safety and convenience for the operating personnel, with equipment arranged in zones so that potentially' con- . taminated areas are separated from clean areas. The heating, ventilating,.and air conditoning systems for the Plant will be designed to provide a suitable environment for equipment and personnel. The path of ventilating air in the Auxiliary Building vill be from areas of lov activity toward areas of progres-sively higher activity. Ventilating air will be recirculated in clean areas only. 9.7.2 SYSTEM DESCRIPTION AND EVALUATION The Reactor Building ventilation system is discussed in 5.3 and shown on Fig-ure 5-6. The remaining ventilation systems for the Plant are discussed here and shown on Figures 9-12 and 9-13. The equipment used to ventilate each build-ing is independent from that used in any other building. The systems handling potentially centaminated air all discharge to the Plant vent. The Control Building vill be equipped with redundant fans, filters, and mechan-ical refrigeration equipment, plus the necessary dampers and controls for auto- 1 matic switching (with a manual override) to full recirculation for post-accident ventilation. The Fuel Handling and Auxiliary Buildings are each ventilated by a separate cupply system and a common exhaust system. Each system incorporates filters, (~
\
and heating coils as required. Air is exhausted to the Plant vent through HEPA and charcoal filters. The Turbine Building ventilation system vill consist of gravity roof ventilators discharging directly to atmosphere. These vill be operated to induce outside air to enter through windows and louvers. Heating vill be provided by means of thermostatically-controlled unit heaters. The Nuclear Plant Office Building ventilation system, shown on Figure 9-13, will consist of air handling equipment containing dust filters, cooling coils, heating coils, and fans. The system will supply a mixture of recircualted and outside air properly tempered to meet the requirements of the space served. l The ventilating equipment vill be in accordance with accepted industry standards l for power station equipment. Redundant exhaust fans vill be provided for the potentially contaminated areas, and a completely redundant ventilation system vill be provided for the control room building. The Control Building syttem performance vill be continually monitored with alarms for high radiation, fan failure, and excessive pressure drop through filters. The control room operator vill have manual override of the auto-matic control for selecting backup fan and filter operation in order to i insure satisfactory control room conditions following an accident. l s l
- v) 9-37 (Revised 1-15-68)
All control room ventilating system fans and filters vill be remote from the control room and vill not be exposed to fire hazards. The ventilating systems vill be designed in accordance with the applicable codes and standards tabulated in Section 9, page 9.1. The ventilating equipment vill be accessible for periodic testing and inspection during normal operation. Where redundant equipment is provided, it will be operated alternately to provide assurance of operability. O 0250 O 9-37a (Revised 1-15-68)
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