ML19309C546

Chapter 9 to TMI-1 PSAR, Auxiliary & Emergency Sys. Includes Revisions 1-11

ML19309C546
Person / Time
Site:	Three Mile Island
Issue date:	05/01/1967
From:	JERSEY CENTRAL POWER & LIGHT CO., METROPOLITAN EDISON CO.
To:
References
NUDOCS 8004080741
Download: ML19309C546 (13)

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~im =r_:. uer. %w.Ii_.._e Section Page 9 AL~ClIAP.? AIID DERGE: ICY SYSTDIS 9-1 9.1 MAKFJP AND PURIFICATICU SYSTEM 9-2 9.1.1 DESIGN BASES 9-2 9.1.1.1 General Sysee: Functicn 9-2

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9.1.1.2 Letdevn Ccoler

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9-2 9.1.1.3 Letdevn Centrcl Valve 9-2a 9.1.1.L Purificacien De=ineralizer 9-2a 9.1.1.5 Makeup Fu= s 9-3 9.1.1.6 Seal Return Coclers 9-3 9.1.1.7 Makeup Tank 9-3 C l.l.8 Filters 9-3 9.1.2 SYSTEM DESCRIPTICN AND EVALUATICN 9-3 w

j 9.1.2.1 Sche =atic Diagra= 9-3 9.1.2.2 Perfer=ance Requirements 9-3 9.1.2.3 Mede cf Operatica 9-3 9.1.2.L Reliability Ccnsideratiens 9-5 9.1.2.5 Cedes and Standards 9-5 9.1.2.6 Systa= Isolation 9-5 9.1.2.7 Leakage Consideratiens 9-5 9.1.2.8 Operating Ccnditiens 9-6 i

9.2 CHDIICAL ADDITICN AND SAMPLING SYSTDI 9-9 9.2.1 DESIGN 3ASES 9-9 9.2.1.1 General Sysee= Function 9-9 9.2.1.2 Scric Acid Mix Tank 9-9 3 9.2.1.3 Scric Acid Fu= s 9-9 J

9.2.1.h Caustic Mi:: Tank 9-9 800 9-1 (Fevised 7-21-67) 40 @+7f '-

CONTENTS (Cont'd)

O Section M

9 2.1 5 Causcie Pmp 9-9 9 2.1.6 Potassium Hydroxide Mix Tank 9-9 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 et Operation 9-11 9 2.2.4 Reliability Considerations 9-12 9 2.2 5 Codes and Stanaamis 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 INTERMEDIATE COOLING SYSTEM 9-18 931 DESIGN BASIS 9-18 932 SYSTEM DESCRIPTION AND EVALUATION 9-18 9 3 2.1 Schematic Diagram 9-18 9 3 2.2 Performance Requirements 9-18 9323 Mode of Operation 9-18 9 3 2.4 Reliability Considerations 9-19 9325 Codes and Standards 9-19 9 3 2.6 System Isolation 9-19 l 9327 Leakage Considerations 9-19 9 3 2.8 Failure Cansiderations 9-19 94 SPENT WEL C00 LIM SYSTEM 9-22 9.k.1 DESIGN BASES 9-22 0001 155 9-11 l

i CONTENTS (Cont!d)

O Section .

Page 9 4.2 SYSTEM DESCRIPTION AND EVALUATION 9-22 9 4.2.1 Schematic Diagram 9-22

9.k.2.2 Perfor

nance Requirements 9-22 9.4.2 3 Mode of operatien 9-22 l 9.k.2.k Reliability Considerations 9-23 9 4.2 5 Ccdes and Standa ds 9-23 9 4.2.6 Leakage Considerations 9-23 9 4.2 7 Failure Considerations 9-23 9 4.2.8 operating Conditions 9-2h .

I 95 DECAY HEAT REMOVAL SYSTEM 9-25 951 DESIGN BASES 9-25 S'-

O '" "*" '""= " '-"

l 9 5 1.2 Decay Heat Re::: oval Pu::rps 9-25 9513 Decay Heat Re::cval Cooler 9-25 4

952 SYSTEM DESCRIPTION AND EVALUATION 9-25 9 5 2.1 Schematic Diagram 9-25 9 5 2.2 Performance Requirements 9-25 9523 Mode of operation 9-25 l 9 5 2.4 Reliability Considerations 9-26 9525 Codes and Standards 9-26 9 5 2.6 System Isolation 9-26 9527 Leakage Considerations 9-26 9 5 2.8 Failure Censiderations 9-26

! 96 C00L2iG WATER SYSTEMS 9-29 96.1 DESIGN BASES 9-29 0001 156 9-111

CONTENTS (Cent'd)

O Section Pm ,

9.6.2 SYSTEM DESCRIPTION AND EVALUATION 9-30 9.6.2.1 Condenser Circulating Wy.:ar System 9-30 9.6.2.2 Secondary Services Cooling Water System 9-30s 9.6.2.3 Nuclear Servicer Cooling Water System 9-31 9.6.2.h Reactor Building Emergency Cooling System 9 - 32 9.6.2 5 Decay Heat Services Cooling System 9-32 9.7 FUEL EANDLING SYSTE4 9-35 9.7.1 DESIGN BASES 9-35 9.7.1.1 General Sysum Function 9-35 9.7.1.2 New Fuel Storage Area 9-35 9.7.1.3 Spent Fuel Storage Pool 9-35 9.7.1.h Fuel Transfer Tubes 9-35 9.7.1.5 Fuel Transfer Canal 9-36 9.7.1.6 Miscellaneous Fuel Handling Equipment 9-36 9.7.2 SYSTEM DESCRIPTION AND F/ALUATION 9-36 9.7.2.1 Receiving and Storing Fuel 9-36 9.7.2.2 Leading and Removing Fuel 9-36 9.7.2.3 Safety Provisions 9-38 9.7.2.h Onerational Limits 9 h0 9.8 STATION VENTILATION SYSTEMS 9 h1 9.8.1 DESIGN BASES 9 h1 9.8.2 SYSTEM DESCRIPTION AND F/ALUATION 9 h1 0 -

9-iv (Revised 12-22-67)

LIST OF TABLES O Table No. Title Page 9-1 Makeup and Purification System Perfor=ance Data 9-7 9-2 Makeup and Purification System Equipment Datt 9-8 9-3 Steam cenerator Feedwater Quality 9-14 9-k Reactor Coolant Quality 9-14 9-5 chemical Addition and Sampling System Equipment Data 9-15 9-6 Intermediate cooling System Performance Data 9-20 9-7 Intermediate cooling System Equipment Data 9-21 9-8 Spent Fuel Cooling System Performance and Equipment Data 9-24 9-9 Decay Heat Removal System Perfor=ance Data 9-27 9-10 Decay Heat Removal System Equipment Data 9-28 o-11 Cooling Water Systems Performance and O- Equipment Data 9-33 l

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O 0001 158 9-v 1

LIST OF FIGURES (At rear of Section)

Figure No. Title

9-1 Flev Diagram Identifications 9-2 Makeup and Purification System 9-3 Chemical Addition and Sampling System 9h Intermediate cooling System 9-5 Spent Fuel Cooling System ,

9-6 Decay Heat Removal System 5

9-6a Cecay Heat-Services Cooling Water System 9-7 Decay Heat Generation Versus Time After Shutdown 9-8 Circulating and River Water 3

9-9 secondary Services cooling Water Systems 9-10 Nuclear Services Cooling Water Systems 9-11 Fuel Handling Systems 9-12 Turbine and Auxiliar/ Suilding Vrntilation System 9-13 Administration Building Ventilati:n System O

9-vi (Revised 12-22-67) 000! !59

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l 9 AUXILIARY AND EMERGENCY SYSTEMy The auxiliary syste=s required to cupport the reactor coolant system dur- )

ir4 nor=al operatica of the Three Mile Island Nuclear Station are de-scribed in the folleving sections and listed below:

a. Makeup and purificatien system.

b. Chemical additien and sa=plin6 systen.

c. Intermediate ecolin6 system.

d. Spent fuel cooling system.

e. Decay heat re= oval system.

f. Cooling vater syste=s.

g. Fuel handling system.

h. Station ventilation systems.

Detailed descriptions of seme of these syste=s have been presented in Sec-tion 6 since they serve as engineered safeguards. The information in this section deals primarily with the functions served durin6 nor=al operation.

Most of the ecmponents within these systems are located within the aux 1.

O 1 =7 ==1141=e- rae 7 te== wits co==ect1=e P1P1=6 *etwee= the re etor building and the auxiliary building are equipped with reactor buildin6 isolation valves as described in 5 2.

The folleving ecdes and standards are used as applicable in the design, fabrication, and testing of eccpenents, valves, and piping:

a. ASME Boiler and Pressure Vessel Cede, Sectica II, Material Specifications.

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

c. ASME Boiler and Pressure vessel Ccde,Section VIII, Unfired Pressure Vessels and ASME Nuclear Case Interpretations,

d. ASME Boiler and Pressure Vessel Cede,Section IX, Welding Qualifications.

e. Standards of the American Sociecy for Testing Materials.

f. American Standard Code for Pressure Pipin6, ASA B31.1,Section I (Power Piping).

S g. American Standard C50.20-1954 Test Code fer Polyphase Induction Motors and Generators. -

t.,' ' ' ,

9-1 0001 160

h. American Standard C50.2-1955 for Alternating Current Motors, Induction Machines, and General an:1 t'niversal Motors.

1. 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 COnditionin6 Cuide; A=erican Society of Heating, Refr16eratin$, and Air Condi:1 nin6 Engineers.

=. Standards of Tubular Exchanger Manufacturers Association.

n. Air Moving and Conditionin6 Association.

o. A=erican Standard 395.1. Alu=inum Tanks .

p. Those ce=ponents not covered by the ASME Code, such as valves and piping, vill be desi ned 6 and fabricated to =eet the requirements of ASA 316.5 or MSS SP-66 and ASA 331.1, respectively.

i l q. Pressure-containing castings will be radiogrepted to meet Severity Level 2 of ASTM E-71. The pressure-containing parts of all pu=ps vill be liquid penetrant-tested in accordance with Appendix VIII of Section VIII of the ASMS Code.

As an aid to review of the system drawings, a standard set of sy=bols and ab- O breviations has been used and is su==arized in Figure 9-1.

91 MAKEUP AND PURIFICATION SYSTEM 9 1.1 DESIGN BASEE 9 1.1.1 ceneral Syste= Function The syste= shcun en Figure 9-2 supplies the reactor ecolant syste= with fill and operational =akeup water; circulates seal water for the reactor coolant l pu=ps and centrol red drive seals; receives, purifies, and recirculates reac-l ter coohnt syste= letdevn to provide water quality and reactor coolant boric

) acid concentration control; and acccc=cdates te=porary changes in the required reacter coolant inventory.

9 1.1.2 Letdevn Cooler The letdevn cooler ecols the letdown flow fr = reactor (colant te=perature +4 a temperature suitable for demineralization and injection to the reactor coolant pu=p seals and centrol rod drive seals. The =aximum letdown flov is required for a startup frc= a cold condit1cn late in core life wherein the reactor cool-ant boren 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 acount corresponding to s the change due to =cderator temperature reactivity deficit. Heat in the let-devn coolers is rejected to the inter =ediate cooling system.

0001 161 gi ..';.

9-2 (Revised 7-21-67)

9 1.1 3 Lstdown Centrol valve Each letdown control valve is sized for the =aximum letdown rate.

9 1.1.4 Purification Demineraliser

• The letdown flow is passed through the purification de=ineralizer to remove re-actor coolant impurities other than boron. The purification l

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f.,b2 9-2a (Revised 7-21-67) e e - ,

letdevn flow to =aintain the reactor coolant water quality is equal to one re-acter c0clant volu=e per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The purification de=ineralizer is si:ed for t "N the =axi=u letdown flow rate as per=itted by the letdown control valve. Re-fer to Table 11-3 for the =aximu= anticipated equi'*S '"- **ssion product ac-cumulation in the reactor :colant.

9.1.1.5 Makeup Pu=ps The =akeup pu=ps are designed to return the letdevn flow to the reactor coolant syste= and supply the seal vater flew to the reactor ecolant pumps and the cen-trol red drive seals. The design flev capacity is equal to the maxi =u= =akeup flow plus the seal water flew to the reactor coolant pumps and the control red drive seals. The pu=ps are sized to meet these requirements with one pu=p in operation.

9 1.1.6 Seal Return Coolers The seal return coolers are sized to re=ove the heat added by the =akeup pu=p and the heat picked up in passage throu6h the reactor coolant pump seals and the control rod drive seals. Heat from these coolers is rejected to the nuc-lear services cooling wate'r syste=.

9 1.1 7 Makeup Tank This tank serves as a surge vessel for the =akeup pu=ps and as a receiver for the letdown flow, chemical addition, and outside =akeup; it also acecm=odates te=porary changes in reactor coolant system volume. The volu=e of the tank is such that the useful tank volu=e is equal to the =ani=un expected expansion and f-~ contraction of the reactor coolant system during power transients.

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9 1.1.S Filters The filters vill prevent the entry of resin fines fro = the demineralizer and other particulates from the Waste Disposal System, Chemical Addition System, and the Station demineralized water supply into the system and into the seals of the reactor ecolant pumps and control rod drives.

9 1.2 SYSTDI DI:SCRIPTICH A:st E:VAMATION

).l.2.1 Sche =atic Diagram The makeup and purification system is shown on Figure 9-2.

9 1.2.2 Perfor=ance Require =ents Tables 9-1 and 9-2 list the system perfor=ance require =ents and data for indi-vidual system eccponents.

9 1.2 3 shde of operation During nor=al operation of the reactor coolant system, one =akeup pu=p centin-uously supply high pressure water from the sakeup tank to the seals of each of-tne reactor coolant pumps, to a header which supplies the seals of the control red drives, and to a =akeup lina ecnnection to one of the reactor inlet lines.

b.

),

y-3 (Revised . c-e~c/

- 0001 163 l

Makeup flcw to the reactor coolant syste= is regulated by the =akeup centrol vaAve, which operates on signals frem the liquid level centroller of the reac-tor coolant system pressuri:er. A control valve in the injection line to the pu=p seals, and in the header of the centrol rod drive seals, aute=atically g maintains the desired inlet pressure to the seals. A s=all part of the water supplied to the seals leaks into the reactor coolsnt syste=. The re=ainder re-tums to the =e.keup tank after passing through one of the two seal return coolers.

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9-3a (Revised 7-21-o7) 0001 164

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Seal vater inleakage to the reactor coolant system requires a continuous letdown of reactor coolant to maintain the desired coolant inventory. In i

additica, bleed and feed of reactor coolant are required for removal of impurities and beric acid fres the reactor coolant. F.eactor ecolant is j

re=cved frc= cne of the reactor inlet lines , cooled during passage through

' one of the letdown coolers, passed from the reactor building through a I

reactor building isolation valve, reduced in pressure during flow thrcugh

' cne of the three letdcyn control valves, and then passed through one puri-fication deminerali:er to a three-way valve, which directs the coolant either to the makeup tank er to the vaste disposal system. i I

Normally, the three-way valve is positioned to direct the letdown flew to -

, the =akeup tank. If the boric acid concentration in the reactor coolant is to be reduced, the three-way valve is positioned to divert the letdown flew to the vaste disposal system. 3eric acid removal is acccmplished in

, the varte dispcsal system either by directing the letdown flow through a

deborating de=ineralizer with the effluent returned directly to the =ake-4 i

up tank or by directing the leticwn flow to a reacter ecolant bleed holdup tank and maincaining the level in the makeup tank with demineralized water i pumped from tht Station desinerali:ed water storage tank. The quantity of unborated water received is measured and limited by inline instrumentation  !

and interlocked with shim red position centrols.

The makeup tank also receives chemicals for addition to the reacter cool-

ant. A hydregen overpressure =aintained in the makeup tank supplies the ,

hydrogen added to the reactor coolant. Other chemicals are injected in '

! solution to the makeup tank. l 1

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System centrol is accceplished remotely from the control room with the  ;

excepcion of the seal return coolers. The letdown flev rate is set by i remotely positiening the letdown centrol valve to pass the desired flev rate. The spare 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 re=cte posi-tiening of the three-way valve and the valves in the waste disposal sys-tem. The control valve in the injection line to the reacter coolant puzp seals and the control rod drive seals is automatically controlled by the pressure differential centroller connected to the reacter ecolant system l to =aintain the desired inlet pressure to the seals. The pressurizer makeup centrol valve is autcmatically centro 11ed by the pressuri:er level controller. Curing heatup and cooldown, the reactor ecclant system pres-l sure varies from 100 to 2,135 psig, and the discharge pressure of the makeup pumps remains about 2,600 psig. Cne of the three letdown control valves is designed for full letdown flow rate centrol at reduced reactor 1

coolant system pressure.

The =akeup pumps are centro 11ed remotely, a

For emergency operation as a high pressure injection supply, the normal  ;

letdevn ecolant flev line and the ner=al seal injection return line are  !

closed; and flew is d'verted to the emergency high pressure injection l lines. The pu=ps and pump motors are designed to operate at the higher ficv rates and lower discharge pressures associated with the high pres- l sure injection requirements. I=ergency operation is discussed in detail '

in o..a.

4 9L (?.evised T-21-c7)

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9.1.2.h Emlirbility Considtratiens The syste= has three full-capacity letdevn centrol valves and two full-capacity letdown ecclers to insure the flev espability needed to adjust Cm boric acid concentratien. Tvc full-capacity seal return ecclers are sup-plied.

Ihree =akeup pu=ps are supplied; one is capable of supplying the required reactor ecolant pu=p seal, centrol red drive seal, and makeup flow. The letdevn ecclers transfer heat to the inter =ediate cooling syste=, and the seal return coolers transfer heat to the nuclear services ecoling water syste=.

9.1.2.5 Cedes and Standards The equip =ent in this syste= will be designed to applicable codes and.stan-dards tabulated in Secticn 9 Cc=ponents which are designed to the ASMZ Code are:

Letdevn Cooler - ASME Section III-C Seal Return Cooler - ASME Section III-C Purification De=ineralizer - ASME Section III-C Makeup Tank - ASME Section III-C 9.1.2.6 syste= Is9 1 atien The letdown line and the reactor coolant pump seal return line penetrate the reactor building. Both lines centain electric =cter-operated isola-(~') tien valves inside the reactor building and pneumatic valves outside which

\s are autc=atically closed with operation of the engineered safeguards.

Four e=ergency injection lines are used for injecting coolant to the reac-ter vessel after a less-of-coolant incident. Check valves in the dis-charge of each =akeup pu=p provide further backup fer reactor building isolation if required. After use of the lines for e=ergency injection is discentinued, the pneu=atic valves in each line outside the reacter build-ing are closed re=otely by the centrol rec = operators.

9.1.2.7 Leakage Ccnsiderstiens Reactor coolant is nor-ally let devn to this syste=. The purificatica de-

=inerali:er will re=cve essentially 100 per cent of the icnic and solid conta=inants except for beric acid, while gasecus conta=inants vill tend to collect in the makeup tank as the letdevn flev is sprayed into the gas space of this tank.

The gas void in the nakeup tank =ay be vented to the vaste disposal syste=

by opening a re=otely operated valve in the vent line. The equipment in this syste= is shielded by concrete. Shielding design criteria are dis-cussed further in Section 11.

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9-5 (Revised 1-3-cS) 04101 166

9 1.2.3 cterating cenditiens The =akeup tank vill be =aintained with a fluid inventory between 100 ft 3 and 500 ft3 Oxygen accu =ulation in the tank will be less than 2 per g

cent by volu=e. Cne letdevn eccler and one =akeu; ;u=; v:.11 be functien- 1 al at all ti=es.

Te ;revent an inadvertent excessive dilutien of the reacter ecciant beric acid concentration, three safety =easures are applied te each cf the tvc metheds of diluting, i.e., the bleed and feed =ethod and the deberating de=ineralizer =ethed. The first safety =easure is a 70 g;= limitatien en ,

the =axi=u= rate of adding de=ineralized water; for feed and bleed the de=1:eralized water =akeup centrol valve to the =akeup tank is aute=at-ically centrolled te ;revent exceeding a preset flev rate; and for de-berating through the de=ineralizer, the three-way valve ;csitten is aute-

=atically changed te step dilution if the de=inerali:er flev rate exceeds the preset flev rate. The secend safety =easure is a centrol red asse:-

bly ;csitten intericek which either per=1ts or prohibits dilutten depend-ing en the centrol red pattern. Because of this intericek, the de=iner-alized water =a.keup valve,and the deberating de=inerali:er inlet icola-tien valve can be c;ened culy vhen the centr:1 red asse=blies are with-drawn te a preset ;csition. The de=inerali:ed water =akeup valve is aise autc=atically closed, and the three-way valve ;csitica is aute=atically changed when the reds have been inserted te a preset ;csition. The third 1 safety =easure censists of closing all the valves as described above when the flev has integrated te a preset value. Initiatien of dilution =ust be by the c;erator, and the c;erater can terminate dilutien at any time.

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O lv i .e 0001 157 9-6 (Revised 1-8-65)

Table 9-1 Makeup snd purift:stien System perfer ance Data

(';SLI'"ID )

Letdevn Flev Maximum (ccid), sp= 70 Tstal Seal Fl:V to Iach Reacter Ccolant Pu=p Seal, gp 45-50 Seal Inleakage to Reactor Coclant Syste= ;er Reactor Coolant Pa=p, 6;m 2

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Injectica Pressure to Reactor Ccolant Pmp Secis et Startup, psig 135-2,255 Injection Pressure to Reactor Ccolant Pa=p Seals (nc=al), psig 2,235 Injection Pressure to Reactor Coolant Pa=p Seals (=axi=u=), psid 2,535 re=perature to Reacter Ccolant Pa=p Seals, acr=c1/=axi=u=, F 125/150 Total Flev to Each Control Red Drive Seal, g;h 25 Seal Inleakage to Reacter Ccolant Syste=

per Control Red Drive, sph 5 Injection Pressure to Control Red Drive Secis c Startup, pois 135-2,255 Injection Pressure to Control Red Drive Seals (nc=al), psig 2,235 Injeatica Pressure to Centrcl Red Drive Seals (=cximu=), psig 2,535 Tc=perature to Control Rcd Drive Seals, nomal/=cxi=u=, F .

125/150 Parificction Letdevn Fluid Te=perature, no=al/=cx1=u=, F 125/140 Makeup Tank Nc=al Cperating Pressure, psig 15 Makeup Tcnk Volu=e Between Mini =u= and K:xi=u= Cperating Levels, r:3 hoo Reac cr Occiant ' dater Quality See Table 9-4

, , + ' .-

". ' ;i' (f .~ 4i 9-7 (Revised 7-21-67)

.000,1 168 I

Table 9-2 Makeup and Parification System Ecuip=ent Data (capacities are on a unit basis)

Makeup Pa=p g

Quantity 3 Type Multistage centritagal, l

=echanical seal Capacity, g;= See Figure 6-2 Head, ft H2 O at sp. gr. = 1 See Figure 6-2 Motor Horsepower, hp 600 Pa p Material SS vetted parts l Design P; essure, psig 2,650 Design Te=parature, F 200 Letdown Ccoler quantity 2 full capacity Type Shell and, spiral tube Heat Transferred, Stu/hr 16.1 x 190 Letdown Flev, lb/hr l 3 5 x 10 Letdevn Te=perature Change, F 555 to 125 Material,shell/ tube CS/GS Design Pressure, psig 2,500 Design Te=perature, F 600 Seal Return Cocler Quantity 2 full capacity Type Shell and tube Heat Transferred, Stu/hr Seal Return Flov, lb/hr 2 x 10D 1.25 x 105 h

Seal Return Te=perature Change, F lhC to 125 Material,shell/ tube Design Pressure, psig CS/SS 100 l:

Design Te=perature, F 200 Ccoling Water Flev, Ib/hr 1.25 x 105 Makeup Tardt Quantity 1 Volu=e, ft3 6c0 Design Pressure, psig 100 Design Te=perature, F 200 Material CS with SS clad 1

Parification De=inerali::er '

Quantity 2 Type Mixed bed, beric acid saturated Cation:Anien Ratio P:1 Material QS 1 Resin Vclu=e, ft3 to '

Flev, sp= 70 1

I i Vessel Design Pressure, psic 100 1 Vessel Design Te=perature, F K?

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92 CEIMIcAL ADDITION AND SAMFLING SYS m O 92.2 DESIGN EASES 9 2.1.1 General Syste= Function i Chemical addition and sampling operations are required to alter and =eni-tor 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 syste= for reactivity control (see Table 3-5 and Figure 3-1), potassium hyd:cxide for pH control, and hydro-gen or hydrazine for oxygen control. The syste= is also designed to add chemicals to the turbine unit feedvater acd condensate systems. The sys-tem is designed to take reactor coolant samples and steam generator water sa=ples.

9 2.1.2 Boric Acid Mix Tank A single boric acid :51x tank is provided as a source of concentrated boric acid solutien. The volu=e of the tank will provide sufficient boric acid solution to increase the boren concentration of the reactor coolant sys-tem to that required for cold shutdown. Hee.ters in the tank =aintain the te=pernture above that required to insure solubility of the boric acid.

Transfer lines vill be electrically traced.

9 2.1 3 Boric Acid Pu=ps Two boric acid pumps are provided to facilitate transfer of the concen-trated boric acid solution from the boric acid mix tank to the borated water storage tank, the makeup tank, or the spent fuel storage pool.

The pumps are sized so that when both are operatics, one ecmplete charge of concentrated boric acid solution frem the boric acid mix tank cay be injected into the reactor coolant system in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

9 2.1.4 caustic Mix Tank The volume of the caustic mix tank was established so that a Na0H solu-tion of sufficient concentration eculd be =aintained at rocm te=perature to neutralize a vaste neutralization tank full of reactor ecolant contain-ing a boric acid concentratica equivalent to that used during refueling.

9 2.1 5 Caustic Pump The caustic pu=p capacity is set so that at the =ax1=u= capacity, the foregoing neutralization operation can be perfor=ed in 15 minutes.

9 2.1.6 Potassium Hydroxide Mix Tank The tank volu=e was established to contair a sufficient a= cunt of KOH for continual addition to the reactor coolant system so that a concentratica of 3-6 ppm can be maintained while letting dec2 at the max 1=um rate. '

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9 2.2 sysTIM DEscaIPrICN /JID EVALUATION 9 2.2.1 Sche =atic Diazram and System Description Figure 9-3 is a sche =atic diagram illustrating the features of the system.

The syste= is operated from local controls. Two beric acid pu=ps, con-nected in parallel, take suction from the boric acid mix tank and dis-charge to either the spent fuel stcrage pool, borated water storage tank, or the =akeup tank. At the end of core life, both beric acid pu=ps are required to raise the reactor ecolant system boren concentration from the mini =um end-of-life concentration to the refueling concentration in ap-prox 1=ately 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 =echanical mixing device and a heating unit.

The potassium hydroxide equip =ent consists of a mix tank, a single pcci-tive displacement pu=p, and connecting piping. The puq discharges to the =akeup tank.

Hydrazine drums are connected to cce of two positive displace =ent pu=ps, l vhich discharge to a line leading to the =akeup tank and to the feedvater system or condensate system. A nitrogen blanket is used to displace the hydrazine as it is removed from the dru=s.

A hydrogen supply =anifold with controls and a distribution line is used I

to supply the desired overpressure in the =akeup tank during operation.

A nitrogen supply =anifold with controls and distribution lines is used to supply a gas blanket or a gas purge for the makeup tank, sodiu= thio-g sulfate mix tank, hydrazine drums, liquid vaste disposal tanks, and gas vaste disposal tanks.

l A caustic mix tank and pump are used to supply chemicals for pH adjust-i

=ent in the liquid vaste disposal system.

l Chemical additions t,? the turbine unit feedvater system or condensate system consist of hydrazine as an cxygen scavenger and a==enia for pH control. The a==cnia is supplied frem a solution tank with a pump to the feedvater line or conden. sate line. The ner~' feedvater quality is shova in Table 9-3 The liquid sampling portica of the; system receives sa=ples of the reactor coolant frem upstream and devastream of the purification demineralizers, frem upstream of the letdevn coolers, frem the =akeup tank, and frc= the secondary side of the steam generators. Water qualities to be =aintained are listed in Tables 9-3 and 9 h. casecus sa=ples are taken f cm the pressurizer vapor space and frem the =akeup tank. Sample lines frem these points are piped to a sa=pling cubicle cutside the reactor building. Sa=

• ples are collected in containers designed for full cperating temperature and pressure at flev rates of 1 and 2 gym.

An e.utematic gas analyzer is used to =cnitor varicus tanks and equip =ent in the vaste disposal system in a centinuous sequence for hydregen-cx,rgen

=ixtures and to alarm at a preset level.

.c ;

n 0001 171

\

9-10

The pertinent para =eters for each =ajor component in the chemical addition and sa=pling syste= are shown in Table 9-5 9 2.2.2 Ferfor=ance Require =ents This system permits sampling of, and chemical addition to, the reactor coolant system, reactor auxiliary syste=s, and turbine unit systems dur-ing normal operation and has no active e=ergency function. During a loss-

-f-coolant accident, this system is isolated at the reactor building cundary.

9'.2.2 3 Mode of operation The system is capable of drawing reactor ecolant 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.

Samplin6 of other process coolant, such as process streams or tanks in the vaste disposal spstem, is acco=plished locally. Equipment for sam-plin6 turbine unit and nonradioactive fluids is separated frem the equip-ment provided for reactor coolant samples. Leakage and drainaSe result-ing from the sampling operations are. collected and drained to tanks lo-cated in the vaste disposal system.

During nor=al operation, liquid and vapor samples =ay be taken from the following points:

Liquid

a. Steam generator secondary rater.

b. Reactor coolant system.

c. Purification demineraliser inlet.

d. Purification demineralizer outlet.

e. Deborating demineralizer outlet.

f. Makeup tank.

Vapor and Gas

a. Pressurizer.

b. Makeup tank.

In addition, an oxygen and hydrogen analyser aute=atically samples the gas spaces in the vaste disposal system tanks and equipment in.an auto-

=atic sequence. The makeup tank Gas space can also be analyzed with this

,- unit.

k"' Durin6 nor=al operation, this syste= also delivers the following chemicals:

g

< ~ . ' .?*}

ff2 9-11

a. Scric acid to the spent fuel storage pool, the borated water storage tank, and the =akeup tank.

b. Caustic to a vaste neutralication tank.

c. Potassiu= hydroxide to the =akeup tank.

d. Hydracine to the =akeup tank and to the feedvater or conden-sate system.

e. A=monia to the feedvater or condensate system.

f. Hydrogen to the makeup tank.

g. Nitregen as required for the core flooding tanks, makeup tanks, hydracine dru=s, sodium thiosulfate storage tank, and tanks and equipment in the vaste disposal system.

9.2.2.h Reliability Considerattens The system is not required to function during an emergency, nor is it required to take action to prevent an e=ergency condition. It is there-fore designed to perform in accordance with standard practice of the chenical process industry with duplicate equip =ent such as pu=ps and high pressure gas regulating valves as required.

9 2.2.5 Cedes and Standards The equipment in this system vill be designed to applicable codes and standards tabulated in Section 9 Equipment applicable to the ASME Codes are: the Reactor Coolant Sa=ple Cooler which vill be designed to ASME Section III, Class C, and the Steam Generator Sa=ple Cooler which vill be designed to ASME Sectica VIII.

9.2.2.6 System Isclatten Isolation of this system frcm the reactor building is accomplished by signals frcm the Safeguards Actuation System as described in Sections 5.2 and T.

9.2.2.7 Leakage Considerattens Leakage of radioactive reactor coolant frcm this system within the re-actor building vill be collected in the reactor building sump. Leakage of radicactive material frem this system outside the reactor building is collected by placing the entire sampling station under a hcod pre-vided with an offgas vent to vaste gas processing. Liquid leakage from the valves in the hoed 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 =akeup tank. Additivec to the spent fuel storage pool are delivered above the water level. Backflow frc=

the =akeup tank to the positive displace =ent pu=ps is prevented by a check valve and a remotely operated valve between them. 3ackflow from g

the makeup tank through the hydrogen additica line is prevented by a

[.,checkvalveandaremote=anualhydrogenadditionvalve.

9-12 (Revised T-21-67) 0001 173

I 9 2.2.8 Failure Considerations To evaluate system safety, the following failures or malfunctions were j assumed concurrent with a loss-of-coolant accident, and the consequences were analped. As a result of this evaluation, it is concluded that proper consideration has been given to Station safety in the design of the system.

Comments and Co=ponent Failure Consequences Pressurizer Sample Electrically operated Diaphragm-operated sampling valve inside valve outside the reactor building fails reactor buildin6 Vill to close en E3 signal. close.

Peactor Letdown Sample Electrically operated Same as above.

sampling 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 ES signal. tem, and steam gen-erator therefore pro-vides first barrier.

Sample Line From 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 Concentratiot.

The boric acid mix tank is to be maintained at an average te=perature of 95 F to =aintain a boric acid concentration of 7 per r ent.

9 2.2 9 2 Coolant Sample Temperature The hi 6h pressure reactor coolant samples leaving the reactor coolant sam-pie cooler should be held to a temperature of 200 F to minimize the genera-tion of radioactive aerosols.

n U

e . .c.  : 0001 174

Table 9-3 O

Steam Generator Feedvater Quality Parameter Value Max 1=um Total Dissolved Solids, pp O.05 Suspended Solids, pp 0.0 Hardness, ppm 0.0 organic, ppm 0.0 Maxi =um Dissolved oxygen, pp O.007 Carbon Dioxide 0.0 Maxd=um Total Silica (as SiO2), ppm 0.02 Maximum Total Iron (as Fe), pp 0.01 Maximum Total Copper (as Cu), pp 0.01 pH 9 3 to 9 5 O

Table 9 h Reactor Coolant Quality Parameter Value Total Solids, =ax. (excludi:6 H E033 "ad KOH)>

pp 1.0 Boron, ppm See Fisure 3-1 KOH, pp 3-6 pH at 77 F 5 5-6.0 pH at 560 F (calculated) 7-10 02 (=ax.), ppb 10 C1 (=ax.), ppm 0.1 H2 , std ec/l 15 ho p a:ine (required during shutdown), ppm 25 h

. 4 ' -'

175 0001 I

9-14

Table 9-5 Che=1 cal Addition and Sa=pling Syste= Eculp=ent Data Tanks Beric Acid Mix Tank Quantity 1 Type Vertical Cylindrical Volu=e, ft3 2,050 Design Pressure, psig At=cspheric Design Te=perature, F 200 Material Al Potassiu= Hydroxide Mix Tank quantity 1

?/pe . Vertical Cylindrical Volu=e, gal 50 Design Pressure, psig At=cspheric Design Te=perature, F 200 Material SS A==cniu= Hydroxide Mix Tank quantity 1

?/pe Vertical Cylindrical Volu=e, gal 50 resign Pressure, psig At=capheric Design Te=perature, F 200 k Material SS Caustic Mix Tank Quantity 1 j

?/pe Vertical Cylindrical Volu=e, gal 500 Design Pressure, psig At=cspheric I Design Te=perature, F 200 l Material CS

)

Hydrazine Dru=s Quantity 2

?/pe Std. Cc=mercial 35 gal Drums Pu=ps Scric Acid Pu=p quantitf 2

?/pe Reciprocating, Variable Stroke Capacity, gym 0-10 Head, psi 50 l Design Pressure, psig 100 -

Design Te=perature, F 200 Pu=p Material SS 0001 176;-

.;r  ;.,..

3-15

Table 9-5 (Cont'd) g Potassium Hydroxide Pu=p Quantity 1

?/pe Reciprocating, Variable Stroke Capacity, sph 0-10 Head, psi 50 Design Pressure, p;1g 100 Design Temperature, F 100 Pump Material SS Ac=cnium Hydroxide Pu7 Quantity 1 T/pe Reciprocating, Variable Stroke Capacity, sph 0 h0 Head, psi 50 Design Pressure, psig 100 Design Te=perature, F 100 Pump Material SS Hydrazine Pu=ps Quantity 2

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

?/pe Reciprocating, Variable Stroke Capacity, gpr. 0-600 Head, pai 10 Design Presture, psis 25 Design Temperature, F 100 Pu=p Materiil CS Samplig Sa=pli g Centa:cers Quantity 10 Desig1 Pressure, psig 2,500 Desigu Te=perature, F 670 Reactor Ccolant Sa=ple Ccoler Quantity 1

?/pe Shell and Spiral Tube l Heat Transferred, Stu/hr 2 9 x 104 l Sample Flov 3 ate, g;= 2 l Max. Sa=ple Inlet Te.*pcrature, F 650

..g g

• ', ,g,Sa=ple Outlet Te=perature, -

'F 150 0001 177 -

9-16

l Table 9-5 (Cont'd) I O CoolingWaterFlev,lb/hr 5 x 103 i

Coil Side Design Te=perature, F 670 1 Coil Side Design Pressure, psig 2,500 )

l Steam Generator Sample Cooler quantity 1 ,

Type Shell and Spiral Tube l Heat Transferred, Stu/hr 2.8 x 104 l Sample Flev Rate, spm 2 Sample Inlet Temperature, F 525 Sample Outlet Temperature, F 150 Coolin6 WaterFlow,lb/hr 5 x 103 Coil Side Design Te=perature, F 600 Coil Side Design Pressure, psig 1,050 j

O L

I s

O W' !.;. ,.o .

. 0001 q7g 9-17

l l

93 InTEms:DIATE CCOLING SYSTEM g

931 DESIGN BASES The system is designed to provide coolin6 vater for various components in the reactor bu11 din 6 as follows: letdown coolers, reactor coolant pump =echanical seal areas, reactor coolant drain tank ecoling coils, and pr1=ary shield cooling coils, i.e., gamma heat remceal. The total design coolin6 requirement for these sources is based en the maximum heat values frem these sources. Se system also provides an additional barrier between high pressure reactor coolant and nuclear services cool-ing water to prevent an inadvertent release of radioactivity.

932 SYSTEM DESCRIPTION AND EVALUATION 9 3 2.1 Schematic Diagram Figure 9 4 is a schematic diagram illustrating the features of the sys-tem. De system is a~ closed loop containin6 two inte=ediate coolers, two inter =ediate coolic6 pumps, an inte=ediate cooling surge tank, and associated piping, valves, and instrumentation which serve to cool the four types of ecmponents in the reactor building. The intem ediate cool-in6 vater in turn is cooled by river water. The pumps, coolers, and i surge tank are located in the auxiliary building.

9 3 2.2 Perfomance Requirements Tables 9-6 and 9-7 list the system performance requirements and data for O individual system ccmponents.

9323 Mode of operation One pump and one couler are normally operated to provide cooling water for r,he ccmponents in the reactor building. Cooli:6 vater flow through the components being cocled is initially balanced with throttle valves.

The pumps and coolers are rotated on a scheduled basis to monitor their operational cape'-111ty.

The water level in the surge tank is maintained between prescribed oper-ating limits by adding demineralized vnter or drainin6 to the auxiliary building sump if either action is required. The intermediate cooling vater is chemically treated to inhibit corrosion. The chemicals are added at the surge tank.

The operation of the system is monitored vith the following instrumenta-tion:

a. Temperature detectors in the main inlet and outlet lines for

! the intermediate coolers.

b. Pressure detector on the line between the pumps and the coolers.

c. Flav detector in the line between the pumps and coolers.

M i ;" 0001 1-79 9-16

d. Level detector on the surge tank.

O e. Temperature detectors located on the outlet lines for the let-down coolers and =echanical seal areas for the reactor coolant pumps.

f. Radiation monitor on the main suction line for the pumps.

9 3 2.4 Reliability considerations The system is not required to funccion during an e=ergency; however, it is designed to handle any abnomalities that =ay occur during operation of the system.

9325 codes and Standards The equipment in this system vill be designed to the applicable codes and standards tabulated in Section 9 9 3 2.6 System Isolation Isolation of the lines passing through the reactor building is accom-plished with two check vr.lves connected in series in the inlet line and with an electric =otor-operated valve and a diaphragm-operated valve in series in the return line. 'Ihe two remotely operated valves are actuated by signals from tce Safeguards Actuation System as described in 71.2.

The electric motor-operated valve and one check valve are located inside J the reactor building, while the other check valve and the diaphra6m-operated valve are located outside the reactor building.

9327 Leakage considerations Operational leakage or maintenance draina6e of water from this system vill be collected in either the reactor buildin6 su=p or the auxiliary building sump. Should a letdown cooler tube leak occur, the leak is re-motely isolated with the electric motor-operated valves on the reactor coolant side. Should a leak occur at the reactor coolant pump mechani-cal seal area, the leak is isolated with an elee w ic =otor-operated valve on the coolin6 vater outlet line and a stop-check valve on the cooling water inlet line. The tet::perature detectors and the radiatice detector located downstream of the letdown coolers and the =echanical seal areas, as ven as the level instrumentation on the surge tank, are used to de-termine such leaks.

9 3 2.8 Failure Considerations To evaluate system safety, the failures or malfune:1ons listed belov vere evaluated concurrently with a loss-of-coolant accident, and the con-sequences were analyced. As a result of this evaluation, it is concluded that proper consideration has been given to Station safety in the design of this rystem.

. i; s i* k. i! '

000!180 9-19

Comments and Component Pa11ure Consequences Cooling Water Supply Check valve inside or The other check valve in Line outside of the reactor the line vill close.

building sticks open.

! Cooling Water Return Electric motor-operat- The other remotely oper-Line 'ed valve or diaphra6m- ated valve closes on ES operated valve fails signal. Additionally, to close on ES signal. the diaphrasm-operated valve vill be an air-to-

! open actuated valve.

l l

Table 9-6 Internediate Cooling System Perfor=ance Data (capacities are on a per unit basis)

Number of Pts::ps 2 Number of Pumps Normally Operating 1 Pump Flev Requircments, gym 650 Number of Coolers 2

(

l Number of Coolers Normally Operating 1 Cooler Heat Removal Require =ents, Btu /hr 17 25 x 10 6 I

O

, ". -(- 0001 181 3ei .

9-20

O Table 9-7 Intermediate Cooling System Equi; ment Data (capacities are en a per unit basis)

Pumps Number 2 Type Centrifugal Rated Capacity, g;m 650 Rated Head, ft H 20 150 Motor HorsepcVer, hp 40 Casing Material Bronze Design Temperature, F 225 Design Pressure, psig 100 Coolers Number 2 Type O' Heat transferred, Btu /hr Shell and Nbe 16.9h x 10$ l Tube Side (River Water)

Inlet Temperature, F 85 Outlet Temperature, F 111 Flev Rate, gym 1,300 Design Temperature, F 225 Design Pressure, psig 100 Shell Side (Intermediate Cooling Water)

Inlet Temperature, F 152 Outlet Temperature, F 100 Flev Rate, g;m 650 Design Temperature, F 225 Design Pressure, psig 100 Tube Material Admiralty Metal Shell Material Carbon Steel Surge Tank Number 1 Capacity, ft 3 So Design Temperature, F 225 Design Pressure, psig 0 Material Carbon Steel O

~0001 lgg

[j'_ ' " ' ! !I ' ' 9-21 (Revf sed T-21-67)

l 94 SPENT FUEL CCoLING SYSTM 9 4.1 DESIGN BASES O

The spent fuel coolinS system is shown on Figure 9-5 It is designed to maintain the spent fuel storage pool at 120 F vith a heat load based on re=oving the decay heat generation frem a 1/3 core that has been irradi-ated for 930 days and cooled for 160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br />. ne system vill do this with 50 per cent of the installed equip =ent. In ceeting the design bases above, the system has the additional capability to =aintain the spent fuel storage pool at 126 F vh11e re=oving the decay heat from the follow-ing combination of stored fuel assemblies:

a. 1/3 core irradiated for 930 days and coolect for 100 days.

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

c. 1/3 core irradiated for klo days and cooled for 160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br />.

d. 1/3 core irradiated for loo days and cooled for 160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br />.

9 4.2 SYSTEM DESCRIPTION AND EVALUATION 9 4.2.1 Sche =atic Diagra=

The sche =atic diagram for the spent fuel cooling system is shown in Fig-ure 9-5 Spent fuel is cooled by pumping spent fuel storage pool water through coolers and back to the spent fuel storage pool. In addition to g

this primary function, the system also provides for purification of both the spent fuel storage pool water and the contents of the borated water storage tank (after it has been used in the fuel transfer canal during refueling).

9 4.2.2 Perfomance Requirements The first design basis of the system predicates an operating schedule in which the nuclear unit is on an equilibrium refueling period (310 FFD per cycle) with approxi=ately 1/3 of a core being removed from the unit at the end of each period. The removed fuel asse=blies vill have been in the reacter for three cycles, i.e., 930 days at the ti=e of discharge.

De second design basic for the syste= considers that it is possible that during the life of the St-tion it vill be necessary to unload the reactor vessel totally for =ainteuance or inspection at the ti=e that the 1/3 core is already residing in the spent fuel storage pool.

D e basic system perfo.~ance and equip =ent data are presented in Table 9-8. .

9 4.2 3 Mode of operation During no=al conditions 1/3 of a core vill be stored in the scol. At this time one of the pu=ps and one of the coolers vill handle the load y.s e.

! 0001 183 9-22

and maintain 120 F. The pool is initially filled with water frc= the g borated water storage tank.

V For the case where 1-1/3 cores are stored (due to cesplete unicading of the reacter vessel), two pu=ps and two coolers vill =aintain the spent fuel storage pool temperature at 126 F. If both a pu=p and a cooler are out for =aintenanc'e when this storage condition exists, the water tem-perature vill eventually rise to ikT F, although considerable time vill be required to heat the large spent fuel storage pool to this tempera-ture. If all cooling is lost, the time rate of temperature rise for the spent fuel storage pool for each of the foregoing quantities of stored fuel is as follevs:

One-third of a core 2.1 F/hr One and one-third cores 5.7 F/hr 9.L.2.h Reliability Censideratiens During the time when a 1/3 core is stored in the pool, caly 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 Components which are designed to the ASE Code are

\ Spent Fuel Cooler - ASE Section III-C Spent Fuel Ccolant Demineralizer - ASE Section III-C 9.k.2.6 Leakage Consideratiens k*henever a leaking fuel assembly is transferred from the fuel transfer canal to the spent fuel storage pool, a s=all quantity of fission prod-ucts may enter the spent fuel ecoling water. A s=all purification lecp is provided for removing these fission products and other contaminants frem the water.

The fuel handling and storage area housing the spent fuel storage ecol vill ce ventilated on a controlled basis, exhausting circulated ali to the outside through the Station vent.

Frovisiens have been made in the design to air-test the valved and flanged l ends of each fuel transfer tube for leak-tightness after it has been used. l A valve and blind flang are used to isolate each fuel transfer tube. I l

9.L.2.7 Failure Considerations l l

The most serious failure of this system vculd be ce=plete less of water I in the stcrage pool. To protect against this possibility, the spent fuel l storage pool ecoling connections enter near or above the water level so that the pcol cannet be gravity-drained. Fcr this sa=e reason care is s also exercised in the design and installation of the fuel transfer tube.

]

pet -

m 0001 184 ,

9-23 (Revised 7-21-67) I i

9.h.2.8 Operatir4 Conditions The pool vill nor= ally be limited to 120 F except in =ost unusual circum-stances as previously described. Boric acid concentration in the pool fluid vill be =aintained at 12,000 to 13,000 ppm.

Table 9-8 Spent Fuel Cooling System Perfor=ance and Equip =ent Data (capacities are on a per unit basis)

System Cooling Capacity, Btu /hr Normal (1/3 core) x 10 6 6 Miximum (1-1/3 cores) 8 7)5 x 10 25.c System Design Pressure, psig 75 System Design Te=perature, F 250 Spent Fuel Cooler Data Nu=ber 2

?/pe Tube and Shell Material (Shell/ Tube) CS/SS Duty, Btu /hrl") 8.75 x 196 Coolirq '4ater Flow, lb/hr 0.5 x 100 Spent Fuel Pu=p Data number 2 g

?/pe Horizontal, Centrifugal Material Stainless Steel Flev, spm 1,000 Head, ft 100 Motor Horsepover, hp 80 Spent Fuel Coolant Demineralizer Flow Rate, gpm 160 Bed Volu=e, ft3 20 Type Non egenerative Vessel Material Carbon Steel - Lined Design Pressure, psig 75 Design Temperature, F 200 Spent Fuel. Storage Pool '4ater 68,250 Vol'me, it 3 (a) Assumes pool water to cooler at 120 F and cooling vater to cooler at 95 F.

O k' -

9-24 (F.evised 7-21-o7) 0001 185

95 DECAY HEAT REMOVAL SYSTEM 951 DESIGN 3ASES

\,

951.1 General Syste= Function

• The ner=al function of this syste= as shown by Figure 9-6 is to re=ove reactor decay heat during the latter stages of cooldown, =aintain reac-tor coolant te=perature during refueling, and provide the =ean.s for fill-106 and draining the fuel transfer canal. The e=ergency functions of this syste= are described in 6.1.

9 5 1.2 Decay Heat Removal Pu=ps The decay heat re= oval pu=ps, durin6 shutdown, circulate the reactor coolant from one reactor cutlet line through the decay heat ecclers and return it to the reactor injection no::les. The design flow is that re-quired to ecol the reactor coolant syste= frc= 250 ? to lLO F in 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />.

(te stea= generators are used to reduce the reactor coolant syste= from operati 6 te=perature to 250 F.)

9 5.1 3 Decay Heat Re=cval Cooler The decay heat re= oval coolers, during shutdown, re=cve the decay heat frc= 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 (14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> after reaching 250 F), two coolers and two pu=ps vill reduce the reactor coolant te=perature to 140 F.

g 952 SYSTEM DESCRIPTION AND EVALUATION 9 5 2.1 Sche =atic Diagram The decay heat re= oval syste= is shown sche =atically in Figure 9-6.

9 5 2.2 Perfor=ance Recuire=ents Tables 9-9 c.nd 9-10 at the end of this subsection list syste= perfor=ance data and desi6n data for individual ec=ponents.

9523 Mede of Operatien Two pu=ps and two ecclers perfc:: the decay heat cooling function. After the steam generators have reduced the reactor ecolant te=perature to 250 F, decay heat coolin6 is initiated by aligning two pu=ps to take suction frem the reactor cutlet line and discharge throu6h the coolers into the reactor vessel. If only one pu=p or one cooler is available, the reactor coolant te=perature is reduced at a lover rate.

The equ:.p=ent utilized for decay heat ecolin6 is also used for low pres-sure injecticn into the core during accident conditions.

O w %n

,_,3 (sevu e1 1_3_ m 0001l86

9.5.2.k Reliability considerations The nuclear unit is provided with two pumps and two coolers. O 9525 Codes and standards The equipment in this system will be designed to applicable codes and standards tabulated in Section 9 The decay heat removal cooler which is applicable to the AEME Code, will be designed to Section III, Clas: C.

95.2.6 system Isolation The decay heat re= oval syctem is connected to a reactor outlet line on the cuc-tion side and to the reactor vecsel on the discharge side. On the cuetion 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 cloced whenever the reactor is in tl.c operating condition. In the event of a loss-of-coolant accident, the valve on the discharge side opens, but the valvec cn the suction side remain closed throughout the accident.

952.7 teakage Considerations During reactor operation all equipment of the decay heat removal system is idle, lnd all isolation valves are closed. During the accident condition, fission products vill be recirculated through the exterior piping system. To obtain the total radiation dose to the public due to leakage from thic system, the potential leaks have been evaluated and discussed in 6 3 and 14.2.

).5 2.8 Failure Considerations

?a11ure considerations for the accident case are evaluated and tabulated in s.13 O

l Til

'ill!! i 0001 187 4C{

9-26 (Revised 1-8-68)

,ja

O Table 9-9 Decay Heat Removal System Perfo:=ance Data Reactor Coolant Temperature at Startup of Decay Heat Removal, F 250 Time to Cool Reactor Coolant System From 250 F to l k F, hr 14 Refueling Temperature, F 140 O nee F ae t oe e= tic = F16ure 9-7 Fuel Transfer Canal Fill Ti=o, hr 1 Fuel Transfer Canal Drain Time, hr 1

't Boron Concentration in the Borated ' dater Stora6e Tank, ppm boron 2,270 0

o a. , e . .;,,,. 9-27 0f)0 I i88

O Table 9-10 Decay Heat Re= oval Systc= Iculp=ent Data -

(capacities are on a per unit basis)

Pmps Nu=ber 2 Type Single stage, centrifugal Capacity, gp= 3,000 Head at Rated Capacity, ft E 02 350 Motor Horsepower, hp 400 Material SS (vetted parts)

Design Pressure, psig 300 Design Te=perature, F 300 Coolers (*)

Nu=ber 2 Type Shell and tube Heat Transferred, Stu/hr 32 5 x lo6 Reactor Coolant Flov, gp=

Cooling Water Flov, gp=

3,000 3,000 $

Cooling Water Inlet Te=perature, F 95 Material,shell/ tube CS/SS Design Pressure, shell/ tube, psig 150/300 Design Te=perature, F 300 Borated Water Storage Tank Nu=ber 1 Capacity, gal 350,000 Material Al Design Pressure Hydrostatic head Design Te=perature, F 150

(" Refer to Figure 6-4 for heat transferred as a function of cooler inlet vater ?,c...perature.

1 1

1 1

i 1

1 0'

k ." ! ' 1M 9-28 (Revised 1-S-66) 0001 l

9.6 COOLING WATER SYSTDiS O 9.6.1 DESIGN BASES The cooling water syste=s are arranged into five separate pumping systems:

a. A decay heat services cooling water system which is comprised of two 100 percent capacity separate systems from the decay heat coolers back to the altimate heat sinz (Susquehanna River). Two 100 percent capacity decay heat river water pu=ps will cool two 100 percent capacity decay heat service water ecolers. Two 100 percent capacity decay heat closed cycle cooling water pumps vill circulate closed cycle cooling water *.hru two 100 percent capacity decay heat coolers and those pu=ps and =otors associated with decay heat re=cval system that require cooling.

b. A reactor building emergency cooling vnter system ec= prised of two separate 100 percent capacity syste=s. Following a LOCA, two 100 percent capacity reactor bui', ding emergency cooling river water pumps vill deliver water frem the river directly to the emergency cooling coils at a pressure in excess of the contain-

=ent design pressure, which is maintained by a pressure control valve en the discharge. Means will be provided to detect a 1crge leak during a LOCA in any of the three emergency ecoling units by differential pressure alarming of individual coil supply and return lines with subsequent isolation. The emergency fs

\

coils nay be tested via a connection from the nuclear services cooling water system, and small leaks can be detected by a roto-meter. After initial testing of the complete system the emergency ,

cooling coils, separate frem the normal cooling coils, vill be I flushed and filled with nuclear ' service water. The river water pumps vill be tested, after the initial tcst, by discharging to the screen house river basin. I c.

A nuclear services closed cooling water system cc= prised of three 50 percent capacity nuclear services river water pumps, four 33-1/3 percent capacity nuclear services ecclers, and three 50 percent capacity nuclear services closed cycle pumps. This system vill satisfy the cooling requirements of all nuclear orientated services other than decay heat and reactor building emergency cooling (s. & b. atove). In the event of a LOCA, cooling vater is diverted frca no.2-essential services to those required for post accident ecoling. This system, while having redundancy in itself, can be supple =ented by the secondary services river vater pumps , by valving, as an additional back-up.

d. A secondary services cooling water system covers all non-nuclear-related cooling vater requirements. This system =ay be permitted to stop functioning in an emergency where power generation is lost. A backup arrangement from the filtered vater fire service tank per=its one instrument air compressor

{ and the emergency feed water pumps to operate in an emergency.

I

['"  :;t:

90 9-29 (Revised 12-22-67)

e. A condenser circulating water system provides cooling water to the =ain surface condenser and feedvater pump turbine con- lll densers under normal operatica. On loss of electrical power, this system vill not be operated. The cycle cooldown require-

=ents are adequately handled by other =eans.

These systems vill be sized to insure adequate heat re= oval based on 85 F river water, =aximum loadings, and leakage allowances. The equip-ment in these systems vill be designed to applicable codes and standards tabulated in Section 9, page 9-1.

All cooling vater systems will be designed to prevent a component failure from curtailing normal Station operation. It vill be possible to isolate all heat exchangers and pumps. Each pressure-reducing valve vill be pro-vided with a bypass.

All systems vill be monitored and operated frem the control room.

solation valves located external to the reactor building vill be incor-porated in all cooling service water lines penetrating the reactor building.

Electrical power requirements can be supplied from any of the redundant sources described in 8.3.3 for: (a) decay heat service water system, (b) reactor building emergency cooling water gystem, and (c) nuclear services cooling vater system. The condenser circulating water system is not intended to he cperated from the diesel generator or the engineered safeguards transformer power sources. If a nuclear services river water pump is not available, a secondary services river water pu=p can be manually started by the operator and fed from the emergency safeguards source.

All system cocponents vill be hydrostatically tested prior to Station startup and vill be accessible for periodic inspections during operation.

All electrica.'. components, switchovers, and starting controls vill be tested periodically.

l 9.6.2 SYSTm DESCRIPTION AND EVALUATION l 9.6.2.1 Condenser Circulating and River Water System l

l The condenser circulating vater will be cooled in two hyperbolic natural draft ecoling towers located to the east of the station proper. Figure 9-8 shows sche =atically the arrange =ent of the system with respect to the Susquehanna River, condenser, cooling tovers , and service water heat exchangers.

Make-up for tower evaporation, vind loss, and blevdown vill be obtained from the secondary servie river water pumping system. After passing through the secondary servi 's coolers, river water vill be mixed with the circulating vater in a cooling taver basin.

O

,"o coi ooo1 191

• • ' 9-30 (Revised 12-22-67)

Blowdown frcm the cooling towers will be discharged to the Susquehanna River and used to dilute treated nuclear vastes.

The circulating water pump building vill be located between the station and the cooling towers. It vill contain six circulating water pumps.

The intake structure vill be provided with trash rakes , traveling screens, and a de-icing water line. Under nor=al operation in sub-freezing weather, condenser circulating vater discharge vill be the source. On loss of circulating water pump power, the nuclear services cooler river water discharge vill be utilized. Make-up to the circulating water system to offset de-icing water loss vill be provided from the discharge of the nuclear service river water pumps by running additional pump (s).

A cross-tie !s provided between nuclear and secondary service river vater syste=s in :ne service water screen and pump house to permit the secondary services river water pumps to pump into the nuclear services river water and pump house and in the nuclear services heat exchanger vault to provide a redundant river vater line between the two.

The nuclear services river water normally discharges to the river and is thus normally available for additional dilution of treated nuclear vastes; when utilized for de-icing during " blackout", it can be provided on an intermittent basis for vaste dilution with alternate operation between the two.

() The nuclear services decay heat services and intermediate service coolers are located in an underground vault near the auxiliary building.

An emergency river water booster pump is provided in the vault to pump from the nuclear services cooler river water discharge to the condenser hotvell in the unlikely event that condensate inventory has been depleted. Fump control and isolating valves are locked closed in the control room.

9.6.2.2 secondary services cooline Water system This vill be a closed ccoling water system (Figure 9-9) with three pumps available to pump 85 F (max) river water to four heat exchangers located in the turbine building where 95 F (nax) closed circuit cooling water is maintained through shell side of the coolers with three closed pumps available.

t l Sufficient redundancy vill be =aintained so that loss of any one pump and/or l cooler vill not affect nornal operation.

The secondary services cooler pu=ps vill be located in a service water intake structure which is so arranged that river water vill be the source.

On the closed cycle portion of the system an elevated surge tank of 10,000 gal capacity vill provide storage of cendensate with makeup frem the con-p densate system being automatically added. Abnormal tank levels vill be V arm:nciated in the control room.

30I i97 9-30a (Revised 12-22-o7)

i l

l l

All heat loads served by this system vill be expendable in the event of O accident, except those for the instrument air compressor and the emergency feedvater pump which are provided with backup source of ecoling vater at 5

that time frem the Station filtered water, fire service head tank. This pumping system can be dropped on unit trip at time of accident. No nuclear-criented services vill be serviced by this system.

9.6.2.3 Nuclear Services cooling Water system All services cooled by this closed system (Figure 9-10) vill be of a nuclear nature and, hence, segregated to this one system. The system vill be divided into two basic circuits as follevs:

Intermediate Coolers - serving: - letdevn coolers

- reactor ecolant pump seal area

- concrete shield ecoling

- reactor coolant drain tank Nuclear Services Cooler - serving:

Normal - spent fuel coolers

- seal return coolers

- evaporator condensers

- sample coolers

- reactor coolant pump motor coolers and reactor coolant pump oil coolers O - spent fuel pumps motors make-up pumps moters Emergency - control room air conditioning

- spent fuel coolers

- control building air conditioning

- spray pumps-motors

- spent fuel pumps-motors make-up pumps-motors In an emergency, the intermediate ecclers vill be shut off and the flov diverted from the normal services to the emergency services on the nu-clear services cooler cycle.

Redundancy is obtained as follows:

3 - Nuclear Services Cooler Pumps 2 run for normal ecoling.

1 Req'd for energency ecoling.

h - Nuclear Services Coolers 3 operated for nor=al ecoling. .

2 Req'd for emergency cooling.

1

\_/ .

Ot'1 0001 193' 9-31 (Revised 12-22-67)

l 3 - ruelear Services closed cooling Water Pu=ps 3

llh l 2 run for nor=al cooling.

i 1 Req'd for emergency cooling.

2 - Inter =ediate Coolers 1 operated for normal cooling.

2 - intermediate Closed Cooling Water Pumps

1 run for normal cooling i

The nuclear services river water pumps are located in the service water screen and pump house and are arranged to take suction in the same manner as the secondary services cooler pumps. River water at 85 F (max) vill be circulated in the coolers located in the nuclear services cooler vault with 95 F (max) closed cooling vater circulated on the shell side by the closed cooling water pumps. Radioactive fluid leakage vill be prevented from returning to the river with this closed system unless a tube leak

, occurs sbanitaneously in a nuclear services cooler and in a cooler served l by the closed system. This is a remote possibility. Thus this design l provides a double contingency against leakage. This is true also for the intermediate coolers and the coolers which they serve. The closed loops vill be centinuously monitored for radioactivity, and if any is detected, the system vill be tested and the leaking cooler isolated. Where spare coolers exist, the spare vill be put into service. g On the closed cycle portion of the nuclear services coolers system an elevated surge tank of 10,000 gal capacity vill provide storage of condensate with make-up from the condensate system being automatically added. Abnormal tank levels will be =enitored in the control room. Overflow vill be piped to the vaste disposal system.

9.6.2.h Reactor Building E=ergency Cooling System This vill be an open system whereby only in the event of an LOCA vill 85 F river water be pumped directly into the emergency coils. Two 100 percent capacity river pumps, located in the service water intake house pump via redundant lines to the coolers ' manifold outside containment.

In the event of coil leakage, inflow of river water into containment vill occur and will be detected by the detection method described in 9.6.lb.

9.6.2.5 Decay Heat Services Cooling System This closed cooling system as coupled with the decay hea+ coolers and pumps is designed to provide two entirely separate 100 percent capacity tystems l

from the core all the way to the ulti= ate heat sink.

l Two 100 percent capacity river vater pumps are lor.ated in the service g

vater intake structure.

h' "

0001 194 9-32 (Revised 12-22-67) l l

a-...____._--_-__ . _ .

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i i

i No 100 percent capacity coolers are located in the vault along side the .

) auxiliary building.

4 Two 100 percent capacity closed cycle pumps ar- located in the auxiliary

buildi 6 I

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O 0001 195 9-32s (Fe'tised 12-22-67) i

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Tablo 9-11 Cooling Water Systems Performance and Equipment Data (Capacities are on a per unit basis. )

Condenser Circulating Water Pu=ps Number 6 Flow, gpm 68,000 Design Pressure, psig 75 Design Temperature, F 85 Secondary Services Cooler Pu=ps Number 3 Flow, gpm 6,000 Design Pressure, psig 75 Design Temperature, F 85 Secondary Services Coolers Number h Type Shell and Tube River Cooling Water Flow (tubeside), gym h,000 River Cooling Water Temperature, F 85 Closed Cycle Cooling Water Outlet Temperature, F 95 Closed Cycle Cooling Water Flow (shell side), gpm h,000 Tube Material Admiralty Shell Material Carbon Steel Design Pressure, shell/ tube, psig 100/150 Design Temperature, shell/ tube, F 150/150 Secondary Services Closed Cooling Water Pumps Number 3 Flow, gym 6,000 Design Pressure, psig 60 Design Temperature, F 95 l

Nuclear Services Cooler Pumps Number 3 Flow, spm 7,000 Design Pressure, psig 75 Design Temperature, F 85 Nuclear Services Coolers Nunber 4 Type Shell and Tube River Cooling Water Flow (tubeside ), gpm h,500 River Cooling Water Temperature, F 85 Closed Cycle Cooling Water Outlet Temperature, F 95 Closed Cycle Cooling Water Flow (shell side), gpm h,500 Tube Material Admiralty Shell Material Carben Steel ,

4 Design Fressure, shell/tute, psig 100/100 l Design Tempera:ure, shell/ tube, y 200/150 f CL.,

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9-33 (Revised 7-21-67) l GNr(% 1

Table 9-11 (Cont'd)

Nuclear Services Closed Cooling '4ater P=ps Nu=ber Flow, gpm 7' Design Pressure, psig 60 Design Te perature, F 95 l1 O-i 00011910 ,

s .

g..

9-3h (Revised 7-21-67)

97 FUEL HANDLING SYS m O 971 DESIGN BASES 9 7 1.1 General System Function

, The Pael handlin6 system (Figure 9-11) is designed to provide a safe, i

effective = cans of +fansportin6 and handling fuel frem the ti=e it reach-es the Station in an unirradiated condition until it leaves the Station after postirradiatien ecoling. The system is designed to minimize the possibility of mishandlin6 or maloperations that could cause fuel assem- '

bly da=a6e and/or potential fission product release.

The reactor is refueled with equipent designed to handle the spent fuel assemblies under water frem the time they leave the reactor vessel until they are placed in a cask for shipment frem the site. Undervater trans-fer of spent fuel assemblies provides an effective, econcmic, transparent radiation shield, as well as a reliable cooling =ediu= for removal of de-cay heat. Berated water insures suberitical conditions during refuelin6 9 7 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 acecmmodate the =ax1=um number of new fuel assemblies required for refuel-ing of the reactor as dictated by the fuel manage =ent prc6 ram. Se new fuel asse=blies are stored in racks in parallel revs having a center-to-O- center distance of 21 in. in both directions. This spacin6 is sufficient to maintain a keff of less than 0 9 when vet.

9713 Spent Fuel Storage Pool The spent fuel storage pool is a reinforced concrete pool lined with stain-less steel; it is located in the fuel stora6e building. The pool is sized to acccmmedate a full core of irradiated Pael asse=blies in addition to the concurrent stora6e of the largest quantity of alient Pael assemblies frcm the reactor as established by the fuel manage =ent program. The spent fuel assemblies are stored in racks in parallel revs having a cen-ter-to-center distance of 21 in. in both directions. Centrol red asse=-

blies requirin6 removal frem the reactor are stored in the spent Pael as-semblies.

9 7 1.4 Fuel Transfer rabes

?.ro hori:cutal tubes are provided to convey fuel between the reactor build-ing and the Pael storage building. These tubes contain tracks for the fuel transfer carriades, gate valves on the spent fuel storage pcol side, and a =eans for flanged closure on the reactor bu11 din 6 side. The Pael '

transfer tubes penetrate into the fuel transfer canal at the icver depth, where space is provided for the rotation of the fuel transfer carria6e basket containing a fuel assembly.

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.e  : g.

9-35

Fuel Transfer Canal 9713 g

The fuel transfer ennal is a passa6evny in the reactor building extending frc= the reactor vessel to the reactor buildi:6 vall. It is fo=ed by an upward extension of the pri=ary shield valls. The enclosure is a rein-forced concrete structure lined with scainless steel; it fo=s a canal above the reactor vessel, which is filled with borated water for refueling.

Spee is available in the fuel trancter canal for undervater stora6e of the reactor vessel internals upper plenum asse=bly.

The deeper fuel transfer station portion of the fuel transfer canal con-tains the new fuel hardlir4 racks. This portion can also be used for storage of the reactor vessel internals core barrel and the=al shield asse=blies by te=porarily re=ovin6 the new fuel hnnMing racks.

9 7 1.6 Miscellaneous Fuel Eardling Ecui;=ent "his equip =ent consisi a of fuel hardli=g bridges, fuel hardling tools, new fuel stora6e racks, spent fuel stora6e racks, new fuel handling racks, fuel transfer containers, control rod handling tools, viewin6 equip =ent, fuel transfer =echanis=s, and shippin6 casha. In addition to the equip-

=ent directly associated with the handlin6 of fuel, equip =ent is provided for hardlin6 the reactor closure head and the upper plenu= asse=bly to ex-pose the core for refueling.

972 srsTEM DEscaIPIION AND EVAWATION 972.1 Receivire and Storing Fuel New fuel asse=blies are received in shipping containers and stored dry in racks havin6 a center-to-center distance of at least 21 in. They are sub-sequently =oved into the reactor buildin6 in or.e of the following ways,

a. After reactor shutdown, new fuel asse=blies can be transferred l from the new fuel stora6e area into the reautor building throu6h the equip =ent hatch and stored directly in the new fuel handling racks in the transfer canal.

b. After reactor shutdown, new fuel asse=blies can be transferred frc= 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 trans-fer tubes.

l 9 7 2.2 Leading and Re=cving Fuel Followin6 the reactor shutdevn and reactor building entry, the refuelin6 procedure is begun by re=oving the reactor closure head and control rod drives asse=bly. Head re= oval and replace =ent ti=e is =inimized by the use of two stud tensioners. The stud tensioner is a hydraulically operat-k p j ad device that pe=its prelcading and unicading of the reactor closure studs'at cold shutdown conditions. The studs are tensicned to their cper-stional lead in tvo steps in a predetemined sequence. Required stud elon6ationaftertensioningisverifiedby=icrc=eter=easure=ents.QGQ} )hh

}--SN.. ,

9-30 L

Folleving re= oval of the studs frc= the reacter vessel tapped holes, the P studs and nuts are supported in the closure head bolt holes with specially desig::ed spacers. Re=cval of the studs with the reactor closure head =in-i=1:es handling ti=e and reduces the chance of thread da= age.

The reactor closure head asse=bly is handled by a lifting fixture support-ed frc= the reactor building crane. It is lifted out of the canal onto a head storage stand located on the operating ficer. The stand is designed to protect the gasket surface of the closure head. The lift is guided by three closure head alignment pins installed in tnree of the stud holes.

These pins also provide proper alig=ent 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 stcrage location for inspection and cicaning using special stud and nut handling fixtures. A stud and alig=ent pin stcrage rack is provided.

The annular space between the reactor vessel flange and the bottc= 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 reacter vessel flange. ,

The fuel transfer canal is then filled with borated water. '

The upper plenum assembly is re=oved frc= the reactor by the reactor building crane and stored under water en a stand en the fuel transfer canal ficor usi g a lifting device with special adapters.

l Refueling operations, aided if necessary by an underwater TV viewing sys- l te=, are carried cut frc= two fuel handling bridges which span the fuel l transfer canal. One bridge is used to shuttle spent fuel asse=blies frc=

the core to the transfer station and new fuel asse=blies frc= the new fuel handling racks to the core. During this cperation, the second bridge is occupied with relocating partially spent fael asse=blies in the core as specified by the fuel =anage=ent progra=.

Fuel asse=blies are ha-M ed by a pneu=atically operated fuel handling tool attached to a telescoping and rotating =ast which = oves laterally on each bridge. Centrol red asse=blies are handled by a centrol red handling tool attached to a second =ast located on ene of the bridges in the reac-ter : tilding.

The two-cast bridge = oves a spent fael asse=bly frc= the core under water to the transfer station where the fuel asse=bly is lowered into the fael transfer carriage fuel basket. The control red handling tool attached to the second =ast is used to transfer a centrol red asse=bly to a new fuel l asse=bly in the adjacent new fuel handling racks. This new fael asse=-

bly with centrol red asse=bly is carried to the reactor by the fael han-dling teci and located in the core while the spent fael asse=bly is being transferred to the spent fael stcrage pecl.

Spen: fael asse=blies re=cved frc= the reacter are tra.sperted to the spent fael storage pec' "- " e reacter building via a fael transfer tube by p =eans cf a cable-cperated fuel transfer car-iage. he spent fuel asse=-

U blies a-- -*-avad '- -'e fuel transfer car-iage basket using a pneu=ati-cally cperated fael handling teci attached to a =cvable =ast iccated en a I t

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c. . ,. - 000I 200 4

fuel handling bridge. This =otor-driven bridge spans the spent fuel stor-age pool and per=1ts the rePaeling crew to store er re=ove new t,.el asse=-

blies in any one of the =any vertical storage rack positions.

h R e fuel transfer =echanis=s are undervater cable-driven carriages that run en tracks extendir4 frc= the spent fuel stora6e pool through the transfer tubes and into the reactor building. Each of the two fuel trans-fer =echanis=s is independently operated so that either can be used for the fuel asse=bly transfer operaticns. A rotating fuel basket is =cunted on ene end of each fuel transfer carriage to receive fuel asse=blies in a vertical position. The hydraulically operated fael basket on the end of the carria6e being used for refueling is rotated to a horizontal position for passa6e throu6h the transfer tube, and then rotated back to a verti-cal position in the spent fuel storage poc1 for vertical removal of the fuel asse=bly.

Once refueling is cc=pleted, the fuel transfer canal vater is drained by suction throu6h a pipe located in the deep transfer station area. The canal vater is pu= ped to the berated water stor36e tank to be available for the next refuelir4 cr for emergency coolin6 folleving a loss-of-coolant accident.

During operation of the reactor, the carria8es are stored in the spent fuel storage pool, thus pe=1tting gate valves on the spent fuel stora6e 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 stora6e. Folleving a sufficient decay period, the apent fael asse=blies are re=oved frc= storage and leaded into the spent fuel shipping cask under water for re= oval frc= the site. Casks up to 100 tons in weight can be healed by the fuel stora6e building err.ne.

A decontamination area is located in the building adjacunt to the spent Pael storage pool; in this area the outside surfaces of the casks can be decentn=1nated before ship =ent by usin6 steam, water, or detergent solu-tions, and -a"nal scrubbing to the extent required.

9723 Safety Provisions Safety provisions are designed into the fuel h uA'ing syste= to prevent the development of hanardous conditions in the event of ec=penent =alfanc-tions, accidental da= age, or cperational and administrative failures dur-ing refueli::6 or transfer operations.

All fuel assembly storage facilities, new and spent, =aintain an eversafe 6ec=etric spacing of 21 in. between asse=blies. The new arai spent fuel storage racks are designed so that it is i=possible to insert fuel asse=-

blies in other than the prescribed locations, thereby insuring the neces-sary spacing between asse=blies. Although new fuel asse=blies are stored l dry, the 21 x 21 in. spacin6 insures an eversafe gec=etric array in un-berated water. Under these conditions, a criticality accident during re-fueling or storage is not censidered credible.

h,

,,,,. 0001 201 h-r .. "A

All fuel handling and transfer containers are also designed to =aintain an eversafe sec=etric array. MechMeal da= age to the fuel assemblies during transfer operations is possible, althcugh re=cte. Since the ff.s-sien product release vculd cecur under water, the a= cunt of activity reaching the enviren=ent vill present no appreciable hazari. A fuel han-d11n6 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=um of 10 ft of water over the active fuel line of the spent fuel assemblies during =cve-

=ent frem the core into storage; this limits radiatica at the surface of thewatertolessthan10 mrem /hr. The spent fuel storage racks are lo-cated to provide a minimum of 13 ft of water shielding o<er stored assem-blies to limit radiatien at the surface of the water to no =cre than 2 5

= rem /hr during the storage period. Se depth of the water over the fuel asse=blies, as well as the thickness,of the concrete valls of the trans-fer canal, is sufficient to limit the =axi=u= centinuous radir. tion levels in the verking area to 2 5 mrem /hr.

Water in the reactor vessel is eccled during shutdevn and refueling by the decay heat removal system described in 9 5 In case of a pcVer fail-ure, this system vill be operated by the auxiliary power supply. The i

spent fuel storage pool water is coole1 by the spent fuel cooling syste=

as described in 9 4. A pcVer failure duri::g tne refuelin6 cycle vill create no immediate hazardcus condition cving to the large water volume in both the fuel transfer canal and spent fuel stors6e pool. With a nor-

=al quantity of spent fuel asse=blies in the storage pec1 and no coolin6 O available, the water temperature in the spent fuel storage pcol vould in-crease as discussed in 9 4.2 3 .

Duri:6 the refueling period the water level in both the fuel transfer canal and the spent fuel storage pool is the sa=e, and the fuel transfer tube valves are continucusly open. This eliminates the necessity for interlocks between the fuel transfer carriages and Pael transfer tube valve operations. The si=plified movement of a transfer carriage through the hori ental fuel transfer tubes minimizes the danger of Jammin6 or de-railin6 To ecpe with such an eventuality, the open tube design provides access to the entire length of the fuel transfer carriage travel frem the fuel transfer canal. All operating =echan's=s i of the system are located

1. the fuel storage buildin6 for ease of maintenance and accessibility for inspection before the start of refueling eieratiens.

During reacter operation, bolted and gasketed closure plates, located on the reacter building flanges of the fuel transfer tubes, prevent leakage of water frem the spent fuel storage pocl into the transfer canal in the event cf a leak through the fuel transfer tube valves. 3cth the spent fue.1 s:crage pcol and the fuel transfer canals are ccmpletely lined with str'.nless steel for leak-tightness and ease of decentamination. The funi transfer tubes vill be appropriately attached to these liners to =aintad leak integrity. The spent fuel storage pec1 cannot be accidentally draine:

since vater =ust be pumped cut thrcugh a suction pipe. The fuel transfer

~

i mechanis=s are designed to per=1 initiation of the carriage travel and l the carriage fuel basket rotaticc frem the building in which the carriage j fuel basket is being leaded er unicaded.

l g . ,, . V ,

"' x- 9-:9 0001 202

All electrical gear is located above vcter for greater integrity and ease of maintenance. "'he hydraulic systems that actuate the rotating fuel baskets use stora6e pool water for operation to eliminate contmination.

g The fuel transfer canal and stora6e pool water win have a boron concen-tration of 2,270 ppm. Although this concentration is sufficient to =ain-tain core shutdown if all of the control rod assemblies were re=oved from the core, only a few control rods vill be removed at any one time during the fuel shufflin6 and replacement. Although not required for safe stor-a6e of spent fuel assemblies, the spent fuel storage pool vater 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 overtension-ing of the studs due to excessive pressure.

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

Any leaking fuel assemblies vill be removed from the core for verification of leaka6e and placed in a failed fuel coni,ainer. This operation is done in the fuel transfer canal and completely neals off the leaking fuel as-sembly before a fuel transfer mechanism transfers it out of the fuel transfer canal into the spent fuel stora6e 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 7 2A operational Limits Certain =anipulations of the fuel assemblies and reactor internals during refueling may/hr. result than 2 5 mrem in short-term The exposure time exposures with rtuliation vill be limited so that levels greater the inte-

, grated doses to operatin6 personnel do not exceed the limits of 10 CFR 20.

l The fuel handling bridges are limited to handling of fuel and centrol rod assemblies and reactor closure head studs only. All lifts for handling the reactor closure head and reactor internals vill use the reactor building crane.

Travel speeds for the fuel band 7 4ng bridges, =asts, and fuel transfer carriages vill be controlled to insure safe handling conditions.

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98 STATICN vrILATICN SYSTES

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9 8.1 DESIGN EASES The Station vill be designed to provide =aximum safety and convenience for the operating personnel, with equip =ent arra Ged in :enes so that potentially conta=inated areas are separated from clean areas. The heati 6, ventilating, and air conditioning syste=s for the Station vill be designed to provide a suitable environment for equip =ent and per-sonnel. The path of ventilating air in the auxiliary building vill be frem areas of lov activity toward areas of pro 6ressively higher activity. .

Ventilatin6 air will be recirculated in clean areas only.

9 8.2 SYSTat DESCRIP: ion AND EVALUATICN The reactor building ventilation syste= is discussed in 5 3 and shown en Figure 5-5 The re=aicing ventilation systems for the Station are lc discussed here and shown on Figures 9-12 and 9-13 The equip =ent used to ventilate each building is independent from that used in any other building. The syste=s handling potentially contaminated air all dis-charge to the Station vent.

The centrol building vill be equipped with recundant fans, filters, l

and =echanical refr16eration equip =ent, plus the necessary da=pers and controls for automatic switching (with a =anual override) to full recir- t culation for postaccident ventilation.

\ The fuel handling and auxiliary buildings are each ventilated by a separ-ate supply system and a common exhaust system. Each system incorporates filters, and heating coils as required. Air is exhausted to the Station vent throuch EEPA and charcoal filters.

The turbine building ventilation syste: vill be pov,ered by roof exhaust fans discharging directly to atmosphere. These vill be operated to induce outside air to enter through =otor-operated louvers when cooling is required. Heating vill be provided by =eans of ther=ostatically -

controlled unit heaters The servica building ventilation syste=, shown en Figure 9-13, vill con-sist of air handling equip =ent containing dust filters, cooling coils, heating coils, and fans. The syste= will supply a =ixture of recirculated and outside air properly te=pered to =eet the require =ents of the space served. Exhaust air vill be discharged directly to the at=ccphere.

The ventilating equip =ent vill be in accordance with accepted industry standards for power station equip =ent- Redundant exhaust f ans vill be provided for the potentially contaminated areas, and a completely redun-dant ventilation syste vill be provided for the control bullaing. l-The centr:1 building syste perfor:1nce vill be continually menitored with alar s f:r high radiation, fin failure, and excessive pressure drop l:

turough filters. The control rocm operator vill have =anual override of

(') the automatic control for selecting backup fan and filter operation in order y7 to ,ingure satisfactory control rocm ccnditions following an accident.

s .s-9 kl (Revised 1.d_co) 04

All control rece ventilating syste= fans and filters vill be re=ote from -

the control rocc and vill not be exposed to fire ha:ards, g

The ventilating syste=s vill be designed in accordance with the applicable ecdes and standards tabulated in Section 9, page 9-1.

The ventilating equipcent vill be accessible for periodic testing and inspection during nor=al operation. Where redundant equipment is pro-vided, it will be operated alternately to provide assurance of operability.

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