ML18018A679

Forwards Response to Draft Auxiliary Sys Branch SER Open Item 225 Re NUREG-0737,Item II.E.1.1 Concerning Auxiliary Feedwater Sys Requirements

ML18018A679
Person / Time
Site:	Harris
Issue date:	08/12/1983
From:	Mcduffie M CAROLINA POWER & LIGHT CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.E.1.1, TASK-TM LAP-83-382, NUDOCS 8308190062
Download: ML18018A679 (25)

Text

REGULATOR+NFORNATION DISTRIBUTION S'EN

. (RIBS)

ACCESSION NBR: 830819006?,

DOC ~ DATE s 83/08/1'2 NOTARIZED; NO FACIL:50-400 Shear on Herr)s Nuclear Power Plant< Unit 1~ Carolina

.50-401 Shearon Harris Nuclear Power PlantF Unit 2~ Carolina AUTH ~ NAME AUTHOR AFFILIATION MCDUFFIE<M ~ AD Carolina

<<Power 8 Light Co, RECIP,NAME RECIPIENT AFFILIATION DENTONzH ~ RE Office of Nuclear Reactor Regulationr Director

SUBJECT:

Forwards response to dr af t Auxi1 iary Sys Br anch SER Open Item 225 re NUREG 0737'tem II.ED 1,1 cancerning auxiliary feedwater sys requirements'I'STRIBUTION CODE s 8001S COPIES RECEIVED: L<<TR

-ENCL "SIZE:,

TITLE: Licensing Submittal:,PSAR/FSAR Amdts L Related Correspondence NOTES'OCKET 05000400 05000401 RECIPIENT IO CODE/NAME NRR/OL/ADL NRR L83 LA INTERNAL ELD/HDS1 IE/DEPER/EPB 36 IE/DEQA/QAB

'21 NRR/DE/CEB 11 NRR/DE/EQB 13 NRR/OE/MEB 18 NRR/DE/SAB 24 NRR/DHFS/HFEB40 NRR/DHFS/PSRB NRR/DS I/AEB 26 NRR/DS I/CPB 10 NRR/DS I/I'CSB 16 NRR/DSI/PSB 19 NRR/DSI/RSB 23 RGN2 EXTERNAL; ACRS 41 DMB/DSS (AMDTS)

LPOR 03 NSIC

.05

~COPIES LTTR ENCL 1

0

.0 1

0 3

<<3 1

1 1

1 2

2 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1

>3 6

6 1,1 1

1 1

RECIPIENT ID CODE/NAME NRR L83 BC KADAMBIg P 01 IE FILE IE/DEPER/IRB 35 NRR/DE/AEAB NRR/DE/EHEB NRR/DE/GB 28 NRR/DE/MTEB 17 NRR/DE/SGEB 25 NRR/DHFS/LQB 32 NRR/DL/SSPB NRR/DSI/AS 8 NRR/DS I/CSB 09 NRR/DS I/MET8 12 AB 22 EG FILE 04 8

BNL(AMDTS ONLY)

FEMA"REP DIV 39 NRC PDR 02 NTIS COPIES LTTR ENCL 1

0 1

1 1

1

,1 1

~

1 0

1 1

?

2 1

1 1

1 1

1 0

1 1

1 1

1 1

1 1

1" 1

0 1

1 1

1 1

1 1

)TOTAL NUMBER OF COPIES REQUIRED; LTTR

'53 ENCL 46

~

tI W

.D

~

~

M W

W W

W W

CM0, Carolina Power & Light Company pgG 1.8 1983 SERIAL:

LAP-83-382 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NOS.

1 AND 2 DOCKET NOS, 50-400 AND 50-401 DRAFT SAFETY EVALUATION REPORT RESPONSES AUXILIARYSYSTEMS BRANCH

Dear Mr. Denton:

Carolina Power

& Light Company (CP&L) and forty copies of the response to one Shearon Draft Safety Evaluation Report Open Item.

This System Branch, and is CP&L Open Item Number 225 hereby transmits one original Harris Nuclear Power Plant response is for the Auxiliary (Part 4).

Yours very truly, SAL/ccc (7671NLU)

Attachment

~

~

M. A. McDuffie Senior Vice President Engineering

& Construction Get Mr. E.

A. Licitra (NRC)

Mr. G. F. Maxwell (NRC-SHNPP)

Mr. J.

P. O'Reilly (NRC-RII)

Mr. Travis Payne (KUDZU)

Mr. Daniel F.

Read (CHANGE/ELP)

Chapel Hill Public Library Wake County Public Library Mr. Wells Eddleman Dr. Phyllis Lotchin Mr. John D. Runkle Dr. Richard D. Wilson Mr.

G.

O. Bright (ASLB)

Dr. J.

H. Carpenter (ASLB)

Mr. J.

L. Kelley (ASLB)

Mr. Norm Wagner (NRC-ASB) mt@

I 8308i90062 8308i2 PDR ADOCK 05000400 E

PDR 411 Fayetteville Street o P. O. Box 1551

~ Raleigh, N. C. 27602

o A

II 4 ~ II

'4 ~ 'I ~ ~

~

1 ~ C

, l 1T E

Shearon Harris Nuclear Power Plant Draft SER Open Item No. 225(4)

Res onse to NRC Question 410.22 Provide a response to the NRC's liarch 10, 1980 letter to near-terra operating license applicants concerning the AFWS design (NUREG-0737, Item II.E.1.1).

The response should include the following:

(4)

A discussion of the design basis for the AFWS flow requirements and verification that the AFWS will meet these requirements (see Enclosure 2

of the March 10, 1980 letter).

~Res ense The Carolina Power

& Light Company response is attached as a study addressing Enclosure 2 of the NRC letter of March 10, 1980.

Table 2-1, the Summary of Assumptions Used in the Design Verification, is currently being reviewed by Westinghouse and will be supplemented when the required additional information becomes available.

CAROLINA POWER

& LIGHT COMPANY SHEARON HARRIS NUCLEAR POWER PLANT UNITS 1

AND 2

RESPONSE

TO ENCLOSURE 2 OF NRC LETTER OF MARCH 10, 1980 "BASIS FOR AUXILIARYFEEDWATER SYSTEM FLOW REQUIREMENTS" y,~

<<,+ p,> qw~np* sq

~

yypegs}

g w gpyvsqwqeygyp z ', Aw ~syvrytq <<p <<'ty w <a -e-,

', ~<,

ww' w w

~ w g

ew g p, ~ ~ ~ +,o

~ y >g.,

~

Question 1

A.

Identify the plant transient and accident conditions considered in establishing ASS flow requirements including the following events:

1.

2.

3.

4.

5.

6.

7.

8.

9.

10.

Loss of main feed (LMFN)

LMFH with loss of off-site AC power LMFW with loss of on-site and off-site AC power Plant cooldown Turbine trip with and without bypass Main steam isolation valve closure Main feedline break Main steamline break Small break LOCA Other transient or accident conditions not listed above.

B.

Describe the plant protection acceptance criteria and corresponding technical bases used for each initiating event identified above.

The acceptance criteria should address plant limits, such as:

1.

Maximum RCS pressure (PORV or safety valve actuation) 2.

Fuel temperature or damage limits (DNB, PCT, maximum fuel central temperature) 3.

RCS cooli.ng rate limit to avoid excessive coolant shrinkage 4.

Minimum steam generator level to assure sufficient steam generator heat transfer surface to remove decay heat and/or cool down the primary system.

Res onse to 1A The Auxiliary Feedwater System serves as a backup system for supplying feedwater to the secondary side of the steam generators at times when the feedwater system is not available, thereby maintaining the heat sink capabilities of the steam generator.

As an Engineered Safeguards

System, the Auxiliary Feedwater System is directly relied upon to prevent core damage and system overpressurization in the event of transients, such as loss of normal feedwater or a secondary system pipe rupture and to provide a means for plant cooldown following any plant transient.

Following a reactor trip, decay heat is dissipated by evaporating water in the steam generators and venting the generated steam either to the condensers through the steam dump or to the atmosphere through the steam generator safety valves or the power-operated relief valves.

Steam generator water inventory must be maintained at a level sufficient to ensure adequate heat transfer and continuation of the decay heat removal process.

The water level is maintained under these circumstances by the Auxiliary Feedwater

System, which delivers an emergency water supply to the steam generators.

The Auxiliary Feedwater System must be capable of functioning for extended periods allowing time either to restore normal feedwater flow or to proceed with an orderly cooldown of the plant to the reactor coolant temperature where the Residual Heat Removal System can assume the burden of decay heat removal.

The Auxiliary Feedwater System flow and the emergency water supply capacity must be sufficient

to remove core decay heat, reactor coolant pump heat, and sensible heat during the plant cooldown.

The Auxiliary Feedwater System can also be used to maintain the steam generator water levels above the tubes following a LOCA.

In the latter function, the water head in the steam generators serves as a barrier to prevent leakage of fission products from the Reactor Coolant System into the secondary plant.

DESIGN CONDITIONS The reactor plant conditions, which impose safety-related performance requirements on the design of the Auxiliary Feedwater

System, are as follows for the Shearon Harris Nuclear Power Plant Units l and 2:

Loss of main feedwater trans1ent with off-site power available without off-site power available Secondary system pipe ruptures feedline rupture steamline rupture Loss of coolant accident (LOCA)

Loss of all AC power Cooldown Loss of Main Feedwater Transients The design loss of main feedwater transients are those caused by:

Interruptions of the Main Feedwater System flow due to a malfunction in the feedwater or condensate system Loss of off-site power or blackout with the consequential shutdown of the system pumps, auxiliar1es, and controls Loss of main feedwater transients are characterized by a rapid reduction in steam generator water levels which results in a reactor trip, a turbine trip, and auxiliary feedwater actuation by the protection system logic.

Following reactor trip from h1gh power, the power quickly falls to decay heat levels.

The water levels continue to decrease progressively uncovering the steam generator tubes as decay heat is transferred and discharged in the form of steam either through the steam dump valves to the condenser or through the steam generator safety or power-operated relief valves to the atmosphere.

The reactor coolant temperature increases as the residual heat in excess of that dissipated through the steam generators is absorbed.

With increased temperature, the volume of reactor coolant expands and begins filling the pressurizer.

Without the addit1on of sufficient emergency feedwater, further expansion will result in water being discharged through the pressurizer safety and relief valves.

If the temperature rise and the resulting volumetric

expansion of the primary coolant are permitted to continue, then l) pressurizer safety valve capacities may be exceeded causing overpressur1-zation of the Reactor Coolant System and/or 2) the cont1nuing loss of fluid from the primary coolant system may result in bulk boiling in the Reactor Coolant System and eventually in core uncovering, loss of natural circulation, and core damage.

If such a situation were ever to occur, the Emergency Core Cooling System would be ineffectual because the primary coolant system pressure exceeds the shut-off head of the safety injection system

pumps, the nitrogen overpressure in the accumulator

tanks, and the design pressure of the Res1dual Heat Removal Loop.

Hence, the timely introduction of sufficient emergency feedwater is necessary to arrest the decrease in the steam generator water levels to reverse the rise in reactor coolant temperature, to prevent the pressurizer from filling to a water solid condition, and eventually to establish stable hot standby conditions.

Subsequently, a decision may be made to proceed with plant cooldown if the problem cannot be satisfactorily corrected.

The blackout transient differs from a simple loss of main feedwater in that emergency power sources must be relied upon to operate vital equipment.

The loss of power to the electric-driven condenser circulating water pumps results in a loss of condenser vacuum and condenser dump valves.

Hence, steam formed by decay heat is relieved through the steam generator safety valves or the power-operated relief valves.

The calculated transients are similar for both the loss of main feedwater and the blackout except that reactor coolant pump heat 1nput is not a

considerat1on in the blackout transient following loss of power to the reactor coolant pump bus.

Seconda S stem Pi e

Ru tures The feedwater line rupture accident not only results in the loss of feedwater flow to the steam generators, but also results in the complete blowdown of one steam generator within a short time if the rupture should occur downstream of the last nonreturn valve 1n the main feedwater piping to an indiv1dual steam generator.

Another significant result of a feedline rupture may be the spilling of auxiliary feedwater from the faulted steam generator.

This could result 1n the pumping of a disproportionately large fraction of the total emergency feedwater flow to the faulted steam generator and out the break because the system preferentially pumps water to the lowest pressure steam generator rather than to the effective steam generators which are at relatively high pressure.

The system design must allow for terminating, limiting, or minimizing that fraction of emergency feedwater flow which 1s delivered to a faulted loop in order to ensure that sufficient flow will be delivered to the remaining effective steam generator(s).

The concerns are similar for the main feedwater line rupture as those explained for the loss of main feedwater transients.

f~h'~f le'\\0gg\\hl%0-

~ PAvi k,f%fan,"

~

Main steamline rupture accident conditions are characterized initially by plant cooldown, and for breaks inside containment, by increasing containment pressure and temperature.

Auxiliary feedwater is not needed during the early phase of the transient, but flow to the faulted loop will contribute to the release of mass and energy to containment.

Thus, steamline rupture conditions establish the upper limit on auxiliary feedwater flow delivered to a faulted loop. Eventually,

however, the Reactor Coolant System w111 heat up again and emergency feedwater flow will be required to be delivered to the unfaulted loops but at somewhat lower rates than for the loss of feedwater transients described previously.

Provisions must be made in the design of the Auxiliary Feedwater System to limit, control, or terminate the auxiliary feedwater flow to the faulted loop as necessary in order to prevent containment overpressuriza-tion following a steamline break inside containment and to ensure the minimum flow to the remaining unfaulted loops.

Loss-of-Coolant Accident (LOCA)

The loss of coolant accidents do not impose on the emergency feedwater system any flow requirements in addition to those required by the other accidents addressed in this response.

The following description of the small LOCA is provided here for the sake of completeness to explain the role of the Auxiliary Feedwater System in this transient.

Small LOCAs are characterized by relatively slow rates of decrease in reactor coolant system pressure and liquid volume.

The pr1nc1pal contribution from the Auxiliary Feedwater System following such small LOCAs is basically the same as the system's function during hot shutdown or following a spurious safety injection signal which trips the reactor.

Maintaining a water level inventory in the secondary side of the steam generators prov1des a heat sink for removing decay heat and establishes the capability for providing a buoyancy head for natural circulation.

The Auxiliary Feedwater System may be utilized to assist in a system cooldown and depressurization following a small LOCA while bringing the reactor to a cold shutdown condition.-

Cooldown The cooldown function performed by the Emergency Feedwater System is a partial one since the reactor coolant system is reduced from normal zero load temperatures to a hot leg temperature of approximately 350'F.

The latter is the maximum temperature recommended for placing the Residual Heat Removal System (RHRS) into service.

The RHR system completes the cooldown to cold shutdown conditions.

Cooldown may be required following expected transients following an accident such as a main feedline break, or during a normal cooldown prior to refueling or performing reactor plant maintenance.

If the reactor is tripped following extended operation at rated power level, the EFS is capable of delivering suff1cient emergency feedwater to remove decay heat and reactor coolant pump (RCP) heat following reactor trip while maintaining the steam generator (SG) water level.

Following transients or accidents, the recommended cooldown rate is consistent with expected needs and at the same time does not impose additional requirements on the capacities of the emergency feedwater pumps considering a single failure.

In any event, the process consists of being able to dissipate plant sens1ble heat in addition to the decay heat produced by the reactor core.

Loss of all AC Power The loss of all AC power is postulated as resulting from accident conditions wherein not only on-si,te and off-site AC power is lost but also AC emergency power is lost as an assumed common mode failure.

Battery power for operation of protection circuits is assumed available.

The impact on the Auxiliary Feedwater System is the necessity for providing both an auxiliary feedwater pump power and control source which are not dependent on AC power and which are capable of maintaining the plant at hot shutdown until AC power 1s restored.

Res onse to lB Table 1B-1 summarizes the criteria, which are the general design bases for each event discussed in the response to Question 1A above.

Specific assumptions used in the analyses to verify that the design bases are met are discussed in response to Question 2.

The primary function of the Auxiliary Feedwater System is to prov1de sufficient heat removal capability for heatup accidents following reactor tr1p to remove the decay heat generated by the core and prevent system overpressurization.

Other plant protection systems are des1gned to meet short-term or pretrip fuel failure criteria.

The effects of excessive coolant shrinkage are bounded by the analysis of the rupture of a main steam pipe transient.

The maximum flow requirements determined by other bases are incorporated into this analysis resulting in no additional flow requirements.

n

'4 hr TABLE 1B-1 CRITERIA FOR AUXILIARYFEEDWATER SYSTEM DESIGN BASIS CONDITIONS CONDITION OR TRANSIENT CLASSIFICATION CRITERIA*

ADDITIONALDESIGN CRITERIA Loss of main feedwater Condition II Peak RCS pressure not to exceed design pressure.

No adverse effects on the core

~

No water released through the pressurizer power-operated relief valves or safety valves.

Station Blackout Condition II Same as LMFW.

Same as LMFW.

Steamline Rupture Condition IV DNB design basis is met assuming most reactive RCCA stuck in fully withdrawn position.

10CFR100 dose limits not exceeded.

Feedline Rupture Condition IV RCS design pressure not exceeded.

10CFR100 dose limits not exceeded.

Core does not uncover.

Loss of all AC power N/A Note 1

Same as blackout assuming turbine driven pump operation.

Loss of Coolant Condition III Condition IV 10CFR100 dose limits 10CFR50 PCT limits (Same as for Condition III)

Cool down N/A 100'F/Hr.

556'F to 350'F

• Ref:

ANSI N18.2 (This information provided for those transients performed in the FSAR).

Although this transient establishes the basis for auxiliary feedwater pump powered by a diverse power source, thi,s is not evaluated relative to typical criteria since multiple failures must be assumed to postulate this transient.

~uestion 2

Describe the analyses and assumptions and corresponding technical justification used with plant condition cons1dered in 1A above 1ncluding:

A.

Maximum reactor power (including instrument error allowance) at the time of the initiating transient or accident.

B.

Time delay from initiating event to reactor trip.

C.

Plant parameter(s) which 1nitiates AFWS flow and time delay between 1nitiating event and introduct1on of AFWS flow into steam generator(s).

D.

Minimum steam generator water level when initiating event occurs

~

E.

Initial steam generator water inventory and depletion rate before and after AFWS flow commences identify reactor decay heat rate used.

F.

Maximum pressure at which steam is released from steam generator(s) and against which the AFW pump must develop sufficient head.

G.

Minimum number of steam generators that must receive AFW flow; e.g.,

one out of two?

Two out of fours H.

RC flow condition continued operation of RC pumps or natural circulation.

I.

Maximum AFW inlet temperature.

J.

Following a postulated steam or feedline break, time delay assumed to isolate break and direct AFW flow to intact steam generator(s).

AFW pump flow capacity allowance to accommodate the time delay and maintain minimum steam generator water level.

Also identify credit taken for primary system heat removal due to blowdown.

K.

Volume and maximum temperature of water in main feed lines between steam generator(s) and AFWS connection to main feed line.

L.

Operating condition of steam generator normal blowdown following initiating event.

M.

Primary and secondary system water and metal sensible heat used for cooldown and AFW flow sizing.

N.

Time at hot standby and time to cooldown RCS to RHR system cut in temperature to size AFW water source inventory.

The limiting transients which define the ASS performance requirements for Shearon Harris are as follows:

- Loss of main feedwater (station blackout)

Rupture of main feedwater pipe Rupture of a main steam pipe inside containment Loss of Main Feedwater (Blackout)

A loss of feedwater assuming a loss of power to the reactor coolant pumps was performed in FSAR Section 15.2.6 for the purpose of showing that for a station blackout transient, two auxiliary feedwater pumps delivering flow to three steam generators does not result in filling the pressurizer.

Furthermore, the peak RCS pressure remains below the criterion for Condition II transients and no fuel failure occurs (refer to Table 1B-1).

Table 2-1 summarizes the assumptions used in this analysis.

The transient analysis begins at the time of teactor trip.

This can be done because the trip occurs on a steam generator level signal;

hence, the core

power, temperatures, and steam generator level at time of reactor trip do not depend on the event sequence prior to trip.

Although the time from the loss of feedwater until the reactor trip occurs cannot be determined from this analysis, this delay is expected to be 20 to 30 seconds.

The analysis assumes that the plant is initially operating at 102% of the Engineered Safeguards Design (ESD) rating shown on the table, a very conservative assumption in defining decay heat and stored energy in the RCS.

The reactor is assumed to be tripped on low-low steam generator level allowing for level uncertainty.

Ru ture of Main Feedwater Pi e

The double-ended rupture of a main feedwater pipe downstream of the main feedwater line check valve is analyzed in FSAR Section 15.2.8.

Table 2-1 summarizes the assumptions used in this analysis.

Reactor trip is assumed to be actuated by low-low level in the affected steam generator.

This conservative assumption maximizes the stored heat prior to reactor trip and minimizes the ability of the steam generator to remove heat from the RCS following reactor trip due to a conservatively small total steam generator inventory.

As in the loss of normal feedwater analysis, the initial power rating was assumed to be 102% of the ESD rating.

The Shearon Harris auxiliary feedwater design is assumed to supply a total of 380 gpm to the two intact steam generators, including allowance for feeding the affected steam generator.

The criteria listed in Table 1B-1 are met.

E

Ru ture of a Main Steam Pi e Inside Containment Because the steamline break transient is a cooldown, the EFS is not needed to remove heat in the short term.

Furthermore, addition of excessive emergency feedwater to the faulted steam generator will affect the peak containment pressure following a steamline break inside contain-ment.

This transient is performed at four power levels for several break sizes.

Fmergency feedwater is assumed to be initiated at the time of the break independent of system actuation signals.

The maximum flow is used for this analysis considering pump runout.

Table 2-1 summarizes the assumptions used in this analysis.

At 10 minutes after the break, it is assumed that the operator has isolated the EFS from the faulted steam generator which subsequently blows down to ambient pressure.

The criteria stated in Table 18-1 are met.

This transient establishes the maximum allowable emergency feedwater flow rate to a single-faulted steam generator assuming all pumps operating; establishes the basis for runout protection, if needed; and establishes layout requirements so that the flow requirements may be met considering the worst single failure.

Plant Cooldown Maximum and minimum flow requirements from the previously discussed transients meet the flow requirements of plant cooldown.

This operation, however, defines the basis for tankage size based on the required cool-down duration, maximum decay heat input, and maximum stored heat in the system.

As previously discussed in Response 1A, the emergency feedwater system partially cools the system to the point where the RHRS may complete the cooldown; i.e., 350'F in the RCS.

Table 2-1 shows the assumptions used to determine the cooldown heat capacity of the emergency feedwater system.

The cooldown is assumed to commence at 102% of engineered safeguards design power; and maximum trip delays and decay heat source terms are assumed when the reactor is tripped.

Primary metal, primary water, secondary system metal, and secondary system water are all included in the stored heat to be removed by the EFS.

See Table 2-2 for the items constituting the sensible heat stored in the NSSS.

This operation is analyzed to establish minimum tank size requirements for emergency feedwater fluid source which is normally aligned.

2

<\\2l TABLE 2-1

SUMMARY

OF ASSUMPTIONS USED IN AFWS DESIGN VERIFICATION ANALYSIS TRANSIENT LOSS OF FEEDWATER (STATION BLACKOUT)

COOLDOWN MAIN FEEDLINE BRFAK MAIN STEAMLINE BREAK (CONTAINMENT) e5 C

a.

Max Reactor Power b.

Time Delay from Event c.

AFWS Actuation Signal/

time delay for flow d.

SG Water Level at Time 102% of ESD rating (102%

of 2910 Mwt) 2 seconds Low-low SG level/

1 minute Low-low SG level 2910 Mwt 2 seconds N/A N/A 102%. of ESD rating (102% of 2910 Mwt) 2 seconds Low-Low SG level/

1 minute Low-Low SG level 0, 30, 70, 102%

(% of Mwt)

Variable Assumed immediately N/A e.

Initial SG Inventory Ra te of change before and after AWFS actuation Decay heat

~

'+

P.

2 22

~.2 f.

AFW Pump Design Pressure g.

Minimum 8 of SGs which must receive RFW h.

RCS Pump Status i.

Max AFW temperature j.

Operator Action k.

AFW Purge Plow/Volume l.

Normal Blowdown m.

Sensible Heat 1265 psia 2of 3 Tripped 8 reactor trip 120'F None None None assumed See cooldown 1265 psia N/A Tripped 120'F N/A None None assumed Table 2-2 1265 psia 2of3 All operating 120'F 30 minutes None None assumed See cooldown N/A N/A All operating Equal to main feed temperature 10 minutes None None assumed N/A n.

Time at standby/time to cool to RHR 2 hr./4 hr.

2 hr./4 hr.

2 hr./4 hr.

N/A o.

AFW flow rate 400 gpm Variable 430 gpm 380 gpm 5

11

TABLE 2-2 SIMfARY OF SENSIBLE HEAT SOURCES Primary water sources (initially at engineered safeguards design power temperature and inventory)

RCS fluid Pressurizer fluid (liquid and vapor)

Primary metal sources (initially at engineered safeguards design power temperature)

Reactor coolant piping,

pumps, and reactor vessel Pressurizer Steam generator tube metal and tube sheet Steam generator metal below tube sheet Reactor vessel internals Secondary water sources (initially at engineered safeguards design power temperature and inventory)

Steam generator fluid (liquid and vapor)

Secondary metal sources (initially at engineered safeguards'design power temperature)

All steam generator metal above tube sheet, excluding tubes 12

• , ~

I

~

D Question 3

Verify that the AFW pumps in your plant will supply the necessary flow to the steam generators as determined by Items 1 and 2 above considering a single failure.

Identify the margin in sizing the pump flow to allow for pump recirculation flow, seal

leakage, and pump wear.

Res onse to 3 The AFW pumps in the Shearon Harris AFWS design will supply the required flow to the steam generators considering single failure.

The two 100%

capacity motor-driven pumps are sized for 450 gpm each and are capable of delivering 490 gpm.

The turbine-driven pump (200% capacity) is sized to deliver 900 gpm.

Thus, for Condi.tion IV events, the AFS has the capability of supplying 200% of the required flow even with a failure of the largest pump.

13

~ ~

~ A I