ML19319D683

Chapter 6 to Crystal River 3 & 4 PSAR, Engineered Safeguards. Includes Revisions 1-10

ML19319D683
Person / Time
Site:	Crystal River, 05000303
Issue date:	08/10/1967
From:	FLORIDA POWER CORP.
To:
References
NUDOCS 8003240662
Download: ML19319D683 (33)

Text

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TABLE OF CONTENTS ps Section 6 ENGINEERED SAFEGUARDS 6-1 6.1 EMERGENCY I?iJECTION 6-1 6.1.1 DESIGN BASES 6-1 6.1.2 DESC3IPTION 6-2 6.1 3 DESIGN EVALUATION 6-3 6.1 3 1 Failure Analvsis 6-5 6.1 3.2 Emergency In.jection Response 6-5 6.133 Special Features 6-6 6.1.3.h Check 'isive Lesksee - Core Flooding System 6-6 3 6.1.h TESTS AND INSPECTIONS 6-6a 6.2 REACTOR BUILDING ATMOSPHERE COOLING AND WASHING 6-13 p 6.2.1 DESIGN BASES 6-13 V

6.

2.2 DESCRIPTION

6-13

- 6.2 3 DESIGN EVALUATION 6-14 6.2 3 1 Failure Analysis 6-15 6.2 3 2 Reactor Building Cooling Response 6-19 6.2 3 3 Special Features 6-19 6.2.4 TESTS AND INSPECTIONS 6-19 6.3 ENGINEERED SAFEGUARDS LEAKAGE AND

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RADIATION CONSIDERATIONS 6-20 l 6.3 1 II."fRODUCTION 6-20 l

6.3 2

SUMMARY

OF POSTACCIDENT RECIRCULATION AND LEAKAGE CONSIDERAiIONS 6-20 6.3 3 LEAKAGE ASSUMPTIONS 6-21 l

6.3.h DESIGN BASIS LEAKAGE 6-22 635 0083 o LEAKAGE ANALYSIS CONCLUSIONS 6-22

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6-1 (Revised 3-1-68) l

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[) LIST OF TABLES Table No. Page i Title 6-1 Core Flooding System Ferfor=ance and Equipment Data 6-4 ,

6-2 Single Failure Analysis-Emergency Injection 6-7 6-3 Emergency Injection Equipment Performance Tbsting 6 -12 64 Reactor Building Cooling Unit Performance and Equipment Data 6-13 6-5 Reactor Building Spray System Performance and Equipment Data 6-14 6-6 Single Failure Analysis-Reactor Building Atmosphere Cooling and Washing 6-16 6-7 Leakage Quantities to Auxiliary Building Atmosphere 6-22

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6-11

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LIST OF FIGURES (At rear of Section)

Figure No. Title 6-l' Emergency Injection Safeguards 6-2 Makeup Pump Characteristics 6-3 Decay Heat Removal Pump Characteristics 6*h Decay Heat Removal Cooler Characteristics -

6-5 Reactor Building Atmosphere Cooling and Washing Safeguards 6-6 Reactor Building Emergency Cooler Characteristics 6-7 Reactor Building Spray Pump Characteristics 0

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6-111

6 ENGINEERED SAFE ~,UARDS Engineered safeguards for each nuclear unit are provided to fulfill four func-tions in the unlikely event of a serious less-of-coolant accident:

a. Protect the fuel cladding.

b. Insure reactor building integrity.

c. Reduce the driving force for building leakage,

d. Re=ove fissica products frca the reacter building atmosphere.

Emergency injection of coolant to the reactor coolant sys a satisfies the first function above, while building atmosphere ecoling and washing satisfy the latter three functions. Each of these cperations is performed by two or more systems which, in addition, employ multiple components to insure operability. All equip;-

ment requiring electrical power for operat!on is supplied by the emergency elec-trical power cources as described in 8.2 3 The engineered safeguards include a core flooding system, high pressure injection equipment, the decay heat removal system, the reactor building cooling system, and the reactor building spray system. Figures 6-1 and 6-5 show the operation of these systems in the engineered safeguarde mode, together with associated in-strurnentation and piping.

Applicable codes and standards for design, fabrication, and testing of compo-nents used as safeguards are listed in the introduction to Section 9, and seis-mic requirements are given in Section 2. The safety analysis presented in Sec-tion 14 demonstrates the perfomance of installed equipment in relation to func-tional objectives with assu=ed failures. l 1

The engineered safeguards functions noted above are acccmplished with the post-accident use of equipment serving nor=al functions. The design approach is based on the belief that regular use of equipment provides the best possible means for monitoring equipment availability and conditions. Because some of the equipment used serves a nor=al function, the need for periodic testing is minimized. In cases where the equipment is used for emergencies only, the sys-tems have been designed to pemit meaningful periodic tests. Additional de-scriptive infomation and design details on equipment used for normal operation are presented in Section 9 This Section 6 will present design bases for safe-guarda protection, equipment operational descriptions, design evaluation of equipment, failure analysis, and a preliminary operational testing program for systems used as engineered safeguards.

6.1 EMERGENCY INJECTION 6.1.1 DESIGN BASES The principal design basis for the emergency injection is as follows:

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U 6-1

Emerc~nev core injection is provided to ornvent clad m-1 tine for 2 the entire spectrum of reactor coolant system failures rancing from the smallest leak to the complete severance of the larcest reactor coolant pipe.

High pressure injection is provided to prevent uncovering of the core for small coolant piping leaks at high pressure and to delay uncovering of the core for intermediate-sized leaks. The core flooding system and the decay heat removal system (which provides low pressure injection) are provided to recover the core at intermediate-to-low pressures to maintain core integrity during leaks rang-ing from intermediate to the largest size. This equipment has been conserva-tively sized to limit the temperature transient to a clad temperature of 2,300 F or less.

6.

1.2 DESCRIPTION

Figure 6-1 is a schematic flow diagram for the emergency injection and associ-ated instrumentation.

Emergency injection fluid, pumped to the reactor coolant system during safe-guards operations, is supplied in each case from the borated water storage tank. This tank contains the volume of borated water necessary to fill the fuel transfer canal during refuelirg operations and is connected to the injec-tion pump suctions by two lines. Additional coolant for the emergency injec-tion supply is contained in core flooding tanks which inject without fluid pumping, as described later in this section.

Emergency injection into the reactor coolant system vill be initiated in the event of an abnormally low reactor coolant system pressure of 1,800 psi dur-ing power operation. The low pressure signals vill automatically increase high pressure injection flow to the reactor coolant system with changes in the operating mode of the makeup and purification system as described in Sec-tion 9:

a. The standby makeup pumps will start and come on the line.

b. The stop check valves in each injection supply line to the makeup and decay heat pumps will open.

c. The injection valve in each of four injection lines vill open.

Emergency high pressure injection vill continue until the reactor coolant sys-tem pressure has dropped to the point where core flocding tanks begin emer-gency injection. The flow characteristic curves for each makeup pump are given in Figure 6-2.

The ccre flooding system has two flooding tanks, each directly connected to a reactor vessel no :le ty a line containing two check valves and one stop (icclatien) valve; the system provides fer autcmatic flooding injection with initiation of flev when the reacter coolant system pressure reaches approxi-mately 603 psi. lhis injection provisicn does not require any electrica.

power, automatic switchi*g, or operatcr action to ensure a supp? cf emergency coolant to the reactor vessel. Operator action is required only during reac-tor cooldown, at which time the stcp valves in the core flooding lines are 6-2 (Revised 2-7-68) 0087 L . _ _ - __. _ _ _ _ _ _ _ _ .

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closed to contain the centents of the core flooding tanks. The combined cool- 2 ant content of the two flooding tanks is sufficient to recover the core hot

' spot' assuming that no liquid is contained in the reactor vessel, while the

gas overpressure and flooding line sizes are sufficient to ensure core reflood-ing within approximately 25 see after the largest pipe rupture has occurred.

The decay heat removal system (described in Section 9) is normally maintained

, , on standby during power operation and provides supplemental core flooding flow through the two core flooding lines after the reactor coolant cyctem pressure reaches 135 psi. Emergency operation of this system vill be initiated by a 4

reacter coolant system pressure of 200 psi during any accident. The flow char- l 1

' acteristics of each decay heat pump for injection are shcvn in Figure 6-3;  ;

these pumps are designed to deliver 3,000 gpm flow into the reactor vessel at '

t a vessel pressure of 100 psi.

Lov pressure injection, with supply from the borated water storage tank, using i

the decay heat pumps vill continue until a low level signal is received from

! the tank (39 min at a combined 1cv pressure injecticn and reactor building spray flow of 9,000 gpm). At this time, the operator vill open the valves controlling suction from the reactor building sump, and recirculation of cool-ant from the sump to the reactor vessel vill begin. The decay heat coolers will cool the recirculated flow, thus removing heat from the reactor building ,

t fluid and preventing further building accumulation of decay heat generated by i

the core.

l The decay heat removal pumps are located at an elevation below the reactor j

building sump with dual suction lines routed outside the reactor building to 3

the pumps. In the event that one suction line is unavailable for recircula-tion, the lines have been sized so that one line vill be capable of handling I

'() the total potential recirculation flow of one 3,000-gpm decay heat removal pump and one 1,500-gpm reactor building spray pump. The NPSH available has been conservatively calculated to be greater than the NPSH requirement of the decay heat removal pumps and the reactor building spray pumps.

The heat transfer capability of each decay heat cooler as a function of recir-culated water temperature is illustrated in Figure 6-h. The heat transfer capability at the saturation temperature corresponding to reactor building pressure is in excess of the heat generation rate of the core following stor-l age tank injection.

i Design data for core flooding system components are given in Table 6-1. De-sign data for other emergency injection ccmponents are given in Section 9 j except for those shown in Figures 6-2, 6-3 and 6-h.

6.1.3 DESIGN EVALUNIION In establishing the required components for the emergency injection the fol-loving factors were considered:

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a. The probability of a major reactor coolant system failure is very lov.

'b. The fraction of a given ecmponent lifetime for whien the component is unavailable because of maintenance is estimated to be a small 6-3 (Revised 2-7-68) 0%B

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part of lifetime. On this basis, it is estimated that the proba- 2 bility of a major reactor coolant system accident occurring while a protective component is out for maintenance is two orders of magni-tude below the lov basic accident probability.

c. The equipment downtime for maintenance in a well-operated station often can be scheduled during reactor shutdown periods. When main-tenance of an engineered safeguard component is required during oper-ation, the periodic test frequency of the remaining equipment can be increased to ensure availability.

d. Where the systems are designed to operate normally or where meaning-ful periodic tests can be performed, there is also a low probability that the required emergency action would not be performed when needed. That is, equipment reliability is improved by using the equipment for other than emergency functions.

(DELETED)

e. Three makeup pumps are installed: One is normally operating, one can be down for maintenance, and one is required for engineered safe-guards.

Table 6-1 Core Flooding System Performance and Equipment Data (for one nuclear unit)

Core Flooding Tanks (*)

Number 2 Design Pressure, psig 700 Normal Pressure, psig 600 Design Te=perature, F 300 Operation Temperature, F 110 Total Volume, ft3 1,410 Nomal Water Volume, ft3 gao Material of Construction Carbon Steel-lined Check Valves Nu=ber per Flooding Line 2 Size, in. 14 Material SS Design Pressure, psig 2,500 Design Te=perature, F 650 Isolation Valves Nu=ber per Flooding Line 1 Size, in. 14 Material S3 Design Prosure, psig 2,500 Design Te=perature, F 650 QQ89

(*) Designed to ASMS Section III, Class C.

6-4 (Revised 2-7-68)

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Table 6-1 (Cont'd)

Piping

() Number of Flooding Lines 2 Size, in. 14 Material SS Design Pressure, psig 2,500 Design Temperature, F 650 6.1 3 1 Failure Analysis The single failure analysis presented in Table 6-2 is based on the assu=ption that a major loss-of-coolant accident had occured. It was then assumed that an additional malfunction or failure occurred either in the process of actuating the emergency injection systems or as a secondary accident effect. All credible failures were analyzed. For example, the analysis includes malfunctions or failures such as electrical circuit or motor failures, stuck check valves, etc.

It was considered incredible that valves would change to the opposite position by accident if they were in the required position when the accident occurred.

In general, failures of the type assumed in this analysis should be unlikely be-cause a program of periodic testing and service rotation of standby equipment will be incorporated in the Plant operating procedures.

The single failure analysis (Table 6-2) and the dynamic postaccident performance analysis (Secticn 14) of the engineered safeguards considered capacity reduction as a result of equipment being out for maintenance or as a result of a failure to start or operate properly. This amounts to adding another factor of conser-vatism to the analyses because good operating practice requires repairing equip-ment as quickly as possible. Plant maintenance activities will be scheduled so that the required capacity of the engineered safeguards systems will always be available in the event of an accident.

The adequacy of equipment sizes is demonstrated by the postaccident performance analysis described in Section 14, which also discusses the consequences of achieving less than the maximum injection flows. There is sufficient redundancy in the emergency injection systems to preclude the possibility of any single credible failure leading to core melting.

6.1 3 2 Emergency Injection Response The emergency high-pressure injection valves are designed to open within 10 sec.

One makeup pump is normally in operation,.and the pipe lines are filled with 2 coolant. The four high-pressure injection lines contain thermal sleeves at their connections into the reactor coolant piping to prevent overstressing of the pipe juncture owing to the 90 F water being injected into these high tem-perature lines. The equipment normally operating is handling 125 F vater, and hence vill experience no thermal shock when 90 F water is introduced.

Injection response of the core flooding system is dependent upon the rate of re-duction of reactor coolant system pressure. For a maximum hypothetical rupture, the core flooding system is capable of reflooding the core to the hot spot within a safe period after a rupture nas occurred.

Emergency low pressure injection by the decay heat removal system will be deli-  !

vered within 25 rec after the reactor coolant system reaches the actuating pres-sure of 200 psig. This anticipated delay time consists of these intervals:

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a. Total instrumentation lag -- = 1 see

b. Emergency power source start -- < 15 see

c. Pu=p motor startup (from the time the pu=p motor line circuit breaker closes until the pu=p attains full speed) -- = 10 see

d. Injection valve opening time -- < 10 see

e. Borated veter storage tank outlet valves -- < 10 see Total (only b and c are additive) = 25 see 6.1 3 3 Special Features

'Ihe core flooding nozzles (Figure 3-61) vill be specially designed to insure that they will safely take the differential temperatures imposed by the acci-dent condition. Special attention also will be given to the ability of the injection lines to absorb the expansion resulting from the recirculating water temperature.

For most of their routing, the emergency injection Jines will be outside the reactor and steam generator shielding, and hence protected from missiles orig-inating within these areas. The portions of the injection lines located be-tween the primary reactor shield and the reactor vessel vall are not subject to missile damage because there are no credible sources of missiles in that area. To afford further missile protection, a high-pressure injection line connects to each reactor coolant inlet pipe, and the twc core flooding noz-zles are loccted en opposite sides of the reactor vessel.

All vater used for emergency injection fluid vill be maintained at a minimum concentration of 2,270 ppm of boron. The temperature, pressure, and level of these tanks will be displayed in the control room, and alams will sound when any condition is outside the nomal limits. The water vill be periodically sampled and analyzed to insure proper boron concentration.

6.1.3.h Check Valve Leakare - Core Fleedine System The action that would be taken in the case of check valve leakare would be a 3 function cf the m2gnitude of

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Limited Oneck valve leskve vill rave nc adverse eff"et on reacter cperaticn.

The valver vi . te r;aci:iea ': ne " *:e tivi.tnace requi rements c f 57-fF-C1.

For these neces, this 1~. cant: :: a nsxirr ;ermiscitle leakage cf ILO cc/hr p e r '. n h e . S; ta. ze u seri.-s are ;rcviaed in each core flecding line; hence, leekugc sn:ui; L< .tcw t.a /sia leikage T0rC05 2

• r. e 5 *. Oh * . /h;ves 2an %Sve *hree effec *C:

(a) it Can cauSe a tenperat ure in:ruze in tne line sn 2 ::. re f1 coding tank, (b ) it can cause a level and resan snt ; rec ure :n r+eze ir. the tsnk, and (c) it can cause dilu-tien cf the u rned water in th- acre riccaine tank.

O 6-6 (Ftevised 3-1-68) )

Leakage at the rate mentioned ab,ove causes insignificant changes in any of these 3 parameters . A leakage of ILO cc/hr causes level increase in the tank cf less f^')

N-than 1 in./co. The associated temperature and pressure increase is correspond-ingly lov.

If it were assumed that the leakage rate is 100 times greater than specified, then chere would still be no signigicant effect en reactor operation since the level change would be approximately 2 in./ day. A 2-in. level change vill re-sult in a pressure increase of approximately 10 psi. 'dith redundant temperature, pressure, and level indicators and alarms available to monitor the core flooding tank ccnditions, the most significant effect en reactor operations is expected to be a more frequent sampling of tank boric' acid concentration.

To insure that no temperature increase vill occur in the tank, even at higher leakage rates, the portion of the line between the two check valves and the line to the tanks vill be left uninsulated to promote convective losses to the building atmosphere.

In summary, reactor operation may continue with no adverse effects coincident with check valve leakage. Maximum permissible limits on core flooding tank parameters (level, temperature, and boron concentration) vil; be established to insure ccepliance with the core protection criteria and final safety analyses.

6.1.h TESTS AND INSPECTICUS All active components, as listed in Table 6-3, of the emergency injection sys-tems vill be tested periodically to demonstrate system readiness. In addition,

~N normally operating components vill be inspected for leaks from pump seals, valve packing, flanged joints, and safety valves, (d

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6-6a (Revised 3-1-68)

-O .O ,O Table 6-2 Single Failure Analysis-Emergency Injection Component Malfunction Comments and Consequences

. A. High Pressure Injection

1. Power-operated valve at Valve remains open. When the tank is empty, tank pres-makeup tank outlet. sure would be.less than the high-pressure injection pump suction pressure (with borated water storage tank on the line), thus preventing l the release of . hydrogen from the tank to the pump suction line.

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2. Pnetaatic operated suction Fails to open. Similar valve in other makeup string 2 valve for makeup pumps from will deliver required flow.

borated water storage tank.

os la 3. Makeup pump. Out for maintenance. Two pumps will still be available. 2 Only one pump is required for engi-Ed neered safeguards.

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$ h. Makeup pump. Fails (stops). Other makeup pump dell.ders required 2 Ok flow.

ru L 5 Makeup pump isolation valve. Left inadvertently closed. See Item A-4 above. Valves will os . normally be left open since the en check valve in each pump discharge will prevent backflow. Operating procedures will call for pump iso-lation valves to be closed only for maintenance.

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Table 6-2 (Cont'd)

Component Malfunction Comments and Consequences

6. Makeup pump discharge check Sticks closed. This is considered incredible valve. since the pump discharge pressure of 2,700 psi at no flow would tend to open even a very tightly stuck check disc.

7 Pressuriser level con trol Fails to close. fio consequences.

valve. l2

8. Deal injection control valve. Fails to close. Injection flow through this line would be small compared to the flow through the two injection lines due to the high flow resis-tance of the reactor coolant pump os seals.

a cn 9 Power operated valve in

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high-pressure injection Fails to open. Flow from one pump will go through 2 gi

< 1Inc. the alternate lcne. Other pump will g.

m operate as normal.

it t (r e e ctor u Id- Sticks closed. See comment on Item A-6 above.

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11. Injection line inside reactor Rupture.

building.

Flow rate indicators in the four injection lines would indicate the l2 gross difference in flow rates.

Check valve in the injection line would prevent additional loss of coolant from the reactor. The line is protected from missiles by reac-tor coolant system shielding.

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0 O O Table 6-2 (Cont'd)

Component Malfunction Comments and Consequences

12. Power operated valve from Inadvertently left open. No significant consequences. A 2 decay heat coolers to makeup small percentage of LP injecticn

. pumps, flow will be bypassed to ifP suction.

13. Power operated valve from de- Fails to open. Alternate line will deliver required cay heat coolers to makeup flow pumps.

14. Double manual valves' con- Inadvertently left open. Not. credible that both valves will necting pump lines. inadvertently be left open.

B. Core Flooding System

1. , Flooding line check valve. Sticks closed. This is considered incredible based 7s y3 on the valve size and opening pres-sure applied.

-Ed y C. Decay Heat Removal System El 1. Check valve'at reactor Sticks closed. This is considert' incredible since m vessel. these valves will be used periodi-a cally during decay heat removal,

ss and the opening force vill be ap-53 proximately 5,000 pounds.

2. Power-operated injection Fails to open. Second injection line will deliver valve. required flow. l2

3. Safety valve. Stuck open. Loss of injection flow is small since valve is small.

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Table 6-2 (Cont'd)

Camponent Malfunction Comments and Consequences 1

h. Decay heat cooler. Isolation valve left closed. Other near exchanger will take re-quired injection flow and remove re-quired heat. Valves will be closed only for maintenance of heat ex-changer.

5. Decay heat cooler. Massive rupture. Not credible. During normal decay heat removal operation, heat ex-changer will be exposed to higher pressure and approximately the same temperature as the postaccident tem-perature and pressure.

f> . Decay heat cooler. Out for icintenance. Remaining heat exchanger will take g, required injection flow.

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7 Decay heat pump isolation Left closed. Remaining pump will deliver required 2

~~ valve. injection flow.

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d. 8. Decay heat pump discharge St.cks closed. See comment on Item C-1 above.

$ check valve.

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9' 9. Decay heat pump. Falls to start. Remaining pump will deliver required 2 Y injection flow.

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10. Stop-check valve at borated Sticks closed. Alternate line will permit required 2 water storage tank outlet. flow.

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Power operated valve permit-ting suction from reactor Fails to open. Two lines and valves are provided, but need not be actuated until 39 l2

'i} building sump. min. after start of accident which C7%

provides time for manual operation.

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Table 6-2 (Cont'd)

Component Malfunction Comments and Consequences

12. Reactor building sump out- Becomes clogged. Cloggind of a single line does not let pipe

• impair functicu because of the dual sump line arrangement, the size of the lines, and the sump design. 'Ihe two recirculation lines take suction from the different portions of the sump. A grating will be provided over the sump, and additional heavy duty strainers will be provided.

13. Dual manual valves connect- Inadvertently 1ert open. Not credible that valves will be in- 2 ing outlet or decay heat advertently left open.

coolers.

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O Table 6-3 E=ergency Injection Equip =ent Perfor=ance Testing Makeup Pu=ps One pu=n is operating continuously. The 2 other two will be periodically tested.

High Pressure Injection Tne re=otely operated stop valve in each line Line Valves vill be opened partially one at a time. Tne flow devices will indicate flow through the lines.

Makeup Pu=p Suction Valves The makeup tank pressure vill be reg- 2 ulated to eaunlize the pressure exerted by the corated water storage tank. The valves will then be opened individually and closed.

Decay Heat Pu=ps In addition to use for shutdown cooling, these pu=ps will be tested singly by opening the borated water storage tank outlet valves and the bypasses in the borated water storage tank fill line. This vill allow water to be pu= ped from the borated water storage tank through each of the injection lines and back to the tank.

Borated Water Storage Tank The operational readiness of these valves will Outlet Valves be established in co=pleting the pu=p opera-tional test discussed above. During this test, each of the valves will be tested separately for flow.

Irv P. essure Injection With pumps shut down and borated water storage Valves tank outlet valves closed, these valves vill be opened and reclosed by operator action.

Valve fc c Suetlua 1' rom With pumps shut down and borated water storage Su=p tank outlet val s close: these valves vill be opened and re lot

• by o1.rator action.

Valves n Core Flooding Valves can '.e opere ed d' -ini; each shutdown to Injection Lines deter =ine perfor=e ;: Isolation valves will be closed to conta.n water in core flooding tanks during shutdown.

0098 O 6-12 (Revised 2-7-68)

6.2 REACTOR EUILDI'M ATMOSPFERE COOLINO AND WASHI'!O 6.2.1 DISION BASES

. O' E=ergency building ctrosphere cooling and vashing is provided to limit post-accident building pressures to design values and reduce the postaccident level of fission products in the building atmosphere.

Fcactor building air recirculation and cooling units, backed up by recetor tuilding sprays, are used for emergency atmosphere cooling. Chemical additives contained in the building sprays are used to reduce postaccident fission pro-duct concentrations in the building atmosphere.

, 6.2.2 DESCRIP1' ION

The schematic flow diagram of the emergency reactor building atmosphere cool-

ing and washing and associated instrunentation is given in Figure 6-5 Emergency and normal cooling are performed with the same basic units. Each unit contains an e=ergency cooling coil, a normal cooling coil, and a two-speed fan. For emergency cooling, all units vill operate under postaccident conditions with the heat being rejected to the nuclear servic'es cooling water system. Each of these units can remove 80 x 106 Btu /hr under peak reactor building te=perature conditions. Figure 6-6 shows the heat exchenge charac-teristics versus building ambient conditions for these units. The design data for the cooling units are shown in Table 6-4.

i Table 6-4 Reactor Building Cooling Unit Performance and Equipment Data (capacities are for single cc=penents) 5 Duty Duergency Normal No. Installed 3 3 No. Required 3 2 Type Coil FinnedTubg Finned Tubg Peak Heat Lead, Btu /hr 80 x 10 2.15 x 100 Fan Capacity, cfm 54,000 l1 108,000

, Reactor Building Atmosphere Inlet Conditions Temperature, F 281 110 Steam Partial Pressure, psia 50 --

, Air Partial Pressure, psia 20 --

! Total Pressure, psig 55 Atmospht te

! Cooling Water Flov, gpm 1,780 250 i Cooling Water Inlet Temperature, F 95 85 Coolint; Water Outlet Temperature, F 185 95 t ;; '

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' Simultaneously with.the air recirculation cooling, reactor building sprays are uupplied with water by two pu=ps which take suction on the borated 6-13 (Revised 4-8-68)

water storage tank until this coolant source is exhausted. The sodium thiosul-fate chemical additive required for the reactor building sprays is supplied from a stora6e tank connected by dual lines containing check valves to the suc-g tion of the spray and decay heat re= oval pu=ps. Sufficient sodium thiosulfate is injocted into the borated water to create a 1 vt % concentration in the re-actor cuilding water inventory.

After the supply from the borated water storage tank is exhausted, the spray pumrs take suction from the reactor building su=p recirculation line. This cor.tinued spraying serves to reduce the reactor building atmosphere to the tem-parature of the reactor building su=p.

Design data for the reactor building spray system components are given in Table 6-5, and the flow characteristics of the reactor building spray pumps are given in Figure 6-7 Design data for components of the reactor building cooling and decay heat removal systems used in this phase of engineered safeguards opera-tion are given in Section 9 and supplemented by Figures 6-3, 6-4, 6-6, and 6-7 of this Section.

Table 6-5 Reactor Building Spray System Performance and Equipment Data (capacities are for single cc=ponents) l5 Reactor Building Spray Pumps Number 2 Flow, gpm Developed Head at Rated Flow, ft 1,500 430 l1

.h Motor Horsepower, hp 250 Material SS Design Pressure, psi 300 Design Te=perature, F 300 Sodium Thiosulfate Tank Number 1 Volume, ft3 1,500 Material SS l1 Design Pressure, psi 50 Design Te=perature, F 150 Sodium Tniosulfate Concentration, wt % 30 Spray Header Number 2 Spray No::les per Header 375 6.2.3 DESIGN EVALUATION The function of cooling the reactor building atmosphere is fulfilled by either of the two methods described above, and redundancy of equipment within both methods will provide for protection of building integrity. The reactor build-ing sprays through duplication, basic washing concept, and chemical additive will serve to reduce fission product levels in the building atmosphere.

O 6-14 (Revised L-S-6S)

For the first 30 ho sin following the maximum blowdown loss-of-coolant acci-(]

k dent, i.e., during the time that the reactor building spray pu=ps take their suction from the borated water storage tank, this system provides more than 100 per cert of the heat removal capacity of the reactor building cooling sys-ter..

The reactor building spray system design is based on the spray water being raised to the tenperature of the reactor building in falling through the steam-air r.ixture within the building. Detailed evaluation of system performance is presented in Section 14. Each of the following equipment arrangements will provile sufficient heat removal capability to maintain the postaccident reac-tor building pressure below the design value:

a. Reactor building spray system.

b. All emergency units in the reactor building cooling system.

c. Two emergency cooling units and the reactor building spray system at one-half capacity.

The reactor building spray system shares the suction line from the borated water storage tank and the tank itself with the high and low pressure injection safeguards.

6.2 3 1 Failure Analysis A single failure analysis has been made on all active components of the systems used to show that the failure of any single active component wi.1 not prevent fulfilling of the design functions. This analysis is shown in Table 6-6. As-sumptions inherent in this analysis are the same as those presented in 6.1 3 in regard to valve functioning, failure types, etc. Results of full and par-tial performance of these safeguards are presented in Section 14 under analy-sis of postaccident conditions.

O v 0101 6-15

Table 6-6

, Single Philure Analysis-Reactor Building Atmoaphere Cooling and Washing Con.ponent Fhlfunction CoIrJnents and Consequences

1. Reactor bu1] ding spray nozzles. Clogged. Large number of nozzles (375 on each of two headers) renders elogging of cignificant number of nozzles as in-credible.

2. Reactor building spray header. Rupture. This is considered heredible due to low operating prereure differential.

3 Check valve in spray header Sticks closed. This is cor.sidered incredible due to line. large opening force available at pump shu'.off head.

4. Air-operated valve in spray Fails to open. Second header delivers 50 per cent P header line. flow.

5 Spray pump isolation valve. Irft closed. Flow and coolinc capacity reduced to 50 per cent of design. In combination with emergency coolers,150 per cent of total design requirement is still provided.

6. Reactor building spray pump. Fails to start. Flow and cooling capacity reduced to 50 per cent of design. In combination with emergency coolers, 150 per cent of total design requirement ic still provided.

O

@ flormal anel emergency cooling Stops. Emergency cooling by the other operat-C unit fan. ing units with supplemental cooling N by the sprays.

G 9- 9

-. . - - . =_ . _ . . __ - -

O O O Table 6-6 (Cont'd)

Component Malfunction Comments and Consequences

8. Homal and emergency cooling Rupture of emergency cooling 'Ihe tubes are designed for 200 psi and unit. coil. 300 F which exceeds maximum operating conditions. Tubes are protected against credible missiles. Hence, rupture is not considered credible.

9 normal and emergency cooling Rupture of casind and/or Consideration will be given during de-unit, ducts. tailed design to the dynamic forces resulting from the pressure buildup during a postaccident situation. 'Ihe units will also be inspectable and protected against credible missiles.

Cooling with these units will be sup-plemented by the sprays.

os A 10. Normal and emergency cooling Rupture of system piping. Rupture is not considered credible units. since all piping is Schedule 40, per-mitting an allowable working pressure of at least 500 psi at 650 F for all sizes. Piping is inspectable and pro-tected from missiles. Maximum actual internal pressure will be less than 200 psi at temperatures below 300 F.

11. Air-operatel valve at inlet Sticks closed. Flow will be periodically established penetration. through the line to check the opera- ,

tional capability of system. Such tests will show if valve is malfunc-tioning.

O

_ 12. Air-operated valve at outlet Fails to open. Comments for Item 11 apply. '

o penetration.

W

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'Ibble 6-6 (Cont'd)

Component Malfunction Coranents c_nd Consequences 13 Power operated valve at co- Fails to open. Alternate check valve will permit flov dium thiosulfate storage tank required for sprays.

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5.2 3 2 Reactor Building Ccoline Respense

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O Air recirculation established during nornal operatien thrcush two of the three building ventilution units continues under acciden , conditiens. In addition, the third unit vill te placed in operatien for the accident ccnditien.

Alternate ecoling coils in these ventilation units, supplied with e .ercency ecoling water, are placed into service after react 0r buildinc pressure increases to . psi. Ccoling centinues utilicing these coils ur.til the building pressure ren:hes near-atncsphrie, 3rd the deca'/ heat removal systen is placed into emergency service, recirculating and ecoling fluid frcs the reacter building su.p.

Se reacter 'cu11 ding spray systen vill likeuise be activated by a single para-meter signal. ~'vo of three signals signifying high reacter building pressure vill start both of the reactor building spray pumps, open the reactor building spray inlet valves, and open the suction valves frc the berated water storage ta:n. The system ccaponents may also be actuated by cperator action from the control roem for perfernance testin6 The total time to delivery of reactor building sprays is approximately 1 min 1 after building pressure reaches 10 psi.

6.233 dpecial Features The casing design for the ventilation units vill be of a conventional nature

• unless additional analysis shows the possibility of pressure wave collapse.

p; In that event, quick, inward-cpening hinged decrs, or other protective features, V vill be inccrporated into the design to maintain postaccident operability. The ventilation units are lccated outside the concrete shield for the reactor ves-sel, steam generators, and reactor coolant punps at an elevation above the water level in the bottom of the reactor buildin6 at postaccident ccnditions.

In this location, the systems in the reactor building are protected frcm cred-ible missiles and from flooding during postaccident operations. Also, this location provides shielding so that the design radiation dose level is 25 mren/hr and allows for maintenance and repair, and inspections to be perfomed during power operation.

The spray headers of the reactor building spray system are located outside and above the reactor and stea Generator concreta shield. During operation, a shield also provides missile protection for the area immediately above the re-actor vessel. The spray headers are therefore protected from missiles origi-nating within the shield. The spray pumps are located cutside the reacter building and are thus available for operative checks during Station operation.

6.2.h T.:STS AND INSFECTICIS Active components of the ventilation units vill nomally be in service. Valv-ing on the emergency coils can be periodically cycled, thus placing the coils into service periodically during operation.

0105 C  %

6-19 (Revised 1-15-68)

The active cc=ponents of the reactor building spray syste= will be tested on a regular schedule as follows:

Reactor Building These pu=ps will be tested singly by opening the Spray Pu=ps valves in the test line and the berated water storage tank outlet valves. Each pu=p in turn vill be start-ed by Station operator action and checked for flow establish =ent to each of the spray headers. Flov vill also be tested through each of the borated water storage tank outlet valves by operating these valves.

33 rated Water These valves will be tested in perfor=ing the pu=p Stcrage Tank test listed above.

Outlet Valves Reactor Building With the pu=ps shut down and the borated water Spray Injection storsge tank outlet valves closed, these valves will Valves each be opened and closed by operator action.

Reactor Building Under the conditiens specified for the previous test, Spray Noccles and with the reactor building spray valves closed, 1 low pressure air vill be blown through the test con-nections.

63 ENGINEERED SAFEOUARDS LEAKA0E AND RADIATION CONSIDERATIONS 6.3 1 INTRODUCTION The use cf ner= ally operatinc equipment for engineered safeguards func- O tions and the location of sc=e of this equip =ent outside the reactor building require that consideration be given to direct radiation levels after fission productc have accu =ulated in these syste=s with leakage frc= these syste=s. Although the engineered safeguards equip =ent is de-signed for re=ote cperation following an accident, long-ter= postaccident operation could necessitate =anual operation of certain valves.

The shielding fer cc=ponents of the Engineered SafeCuards is designed to provide protectica for personnel to perfor= all operations necessary for

=itigation of the accident within the li=its of 10 CFR 100 in the event of an MHA.

632 SLM'.ARY OF KSTAOCIDENT RECIRCULATION AND LEAKAGE CCNSIDERATIONS Following a loss-of-ccclant accident and exhaustion of the borated water stcrage tank, reactcr building su:.:p recirculation to the reactor vessel and the reactor building sprays is initiated.

While the reactor auxiliary syste=s involved in the recirculation co= plex are closed to the auxiliary building at::osphere, leakage is possible through co=penent flan 6es, seulc, instru=entaticn, and valves.

o-20 (hevised 1-15-8)

The leakage sources consideret are:

a. Valves.

(1) Disc leakage when valve is on recirculation ecmplex boundary.

(2) Stem leskage.

., (3) Bonnet flange leakage.

b. Flanges.

c. Pt.:mp stuffing boxes.

While leakage rates have been assuced for these sources, maintenance and periodic testing of these systems vill preclude all but a amall percentage of the assumed amounts. With the exception of the boundary valve discs, all of the potential leakage paths may be examined during periodic tests or normal operation. The boundary valve disc leakage is retained in the other closed systems and therefore vill not be released to the auxiliary building.

While valve stem leakage has been assumed for all valves, the manual valves in the recirculation complex are backseating.

633 LEAKAGE ASSUMPfIONS Source Quantities

a. Valves - Process gl) Disc leakage 10 cc/hr/in of nominal disc diameter.

(2) Stem leakage 1 drop / min (3) Bonnet flange 10 drops / min

b. Valves - Instrumentation Bonnet flange and stem 1 drop / min

c. Flanges 10 drops / min

d. Pump Stuffing Boxes 50 drops / min

! For the analysis, it was assumed that the vatar leaving the reactor build- 1 I

ing was less than 200. F when recirculation occurs.

O OI07 i

6-21 (Revised 1-15-68) ,

s _ _ _ __ .-_ __.

1 ILELE*E;) h 6 3.4 DESIGN BASIS LEAKAGE The design basis leakage quantities derived fro = these assu=ptions for postaccident su=p recirculation are tabulated in Table 6-7 635 LEAKAGE ANALYSIS CONCLUSIONS It =ay be concluded fro = this analysis (in conjunction with the discussion and analysis in 14.2.2.4.4) that leaka6e from Engineered Safeguards equip-

=ent outside the reactor building does not pose a public safety proble=.

Table 6-7 Leakage Quantities to Auxiliary Bai.lding At=osphere 1

No. of Per Source, Total, Leakage Source Sources drops /=in ec/hr

a. Pu=p Seals 2

h Decay heat pu=ps 2 50 300 Spray pu=ps 2 50 300

b. Flanges (a) 114 10 3,320

c. Process Valves 35 1 105 IDELETED)

d. Instru=entation valves 25 1 75

e. Valve Seats at Boundaries 11 (b) 750 2

Total 1,850

(")Asru=es process and boundary valves, and process cc=ponents are flarged.

(b)Assu=es 10 cc/hr/in. cf nc=inal dise dia=eter.

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