ML19326A253

Forwards Request for Addl Info from Util Based on FSAR Review

ML19326A253
Person / Time
Site:	Davis Besse
Issue date:	07/16/1973
From:	Tedesco R US ATOMIC ENERGY COMMISSION (AEC)
To:	Deyoung R US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 8002030114
Download: ML19326A253 (8)

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Docket No. 50- 346 Richard C. DeYoung, Assistant Director for Pressurized Water Reactors, L.

REQUEST POR ADDITIONAL INFORMATION FOR DAVIS-BESSE Plant Name: Davis-Besse Licensing Stage: OL USSS Suppliar: Babcock and Wilcox Architect Engineer: Bechtel Containment: Dual Docket No. : 50-346 Responsible Ernnch & Project '4anager: FWR #4, I. Peltier Requested Completion Date: July 6, 1973 Applicant's Response Date: October 12, 1973 Description of Response: Additional Inforeation Review Status: Awaiting Information The enclosed request for additional information for the Davis-Besse Nuclear Power Station operating license review has been prepared by the Containment Systems Branch af ter having reviewed the applicable sections of the FSAR.

The following comments are based on our review:

.l.

The applicant has not performed a complete pipe break spectrum analysis which smuld identify the break -ize and locatica that results in the highest calculated containment pressure.

2.

The applicant has not discussed the conservatisca in the analysis of the core flooding rate or presented curves of core flooding rate as a function of tina.

3.

The applicant has not described t!m core reflood model.

4.

The applicant has not provided suffiefent information to permit us i

to perform confirmatory containment response analyses.

5.

The applicant proposes to repressurize the containment following a loss-of-coolant accident as a means for diluting possible hydrogen evolution from metal-water reactors or radiolizers. Since the

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Request for Additional Informatioa cc: w/o enci A. Cf= h =a W. W ald w/enci J. M. Hendrie S. H. Hannuer D. Vassallo I. Peltier J. Glynn J. Shapaker i

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I EEQUEST FOR ADDITIONAL INFORMATIGE DAVIS-BESSE N T LEAR POWER STATION DOCKET No. 50-346 I

6.1 The FSAE indicates that cold leg, pump suction and pump discharge breaks have not base analyzed, and it is not apparent that the i

3 ft2 hot les break results in the highest calculated contain= ant pressure; therefore, provide the results of contain==gt pressure response analyses for a spectrum of break areas for a cold leg (pg suction) pipe and a cold les (pump discharge) pipe to j

identify the break size and location that results in the highest concaisment pressure. Include the following information for j

each case analysed: break area, break location, pest sontain-3 ment pressure, tims of peak pressure, and energy released to the f,

contaiammet up to the time of peak pressure. For the loss-of-coolant accident at each of the assumed break locations, i.e.,

i the hot lag and cold leg, pump suction and pump discharge pipes, that results in the highest calculated containment pressure, pro-vida a table of mess release rate (1b /hr) and enthalpy (Bru/lb )

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as a function of tzme (br) throughout the blowdown and cora reflood phases of the accidents, i

6.2 Provide an analysis of the containment pressure response for a spectrise of =tcan generator, steam line and feedvater pipe

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ruptures. Specify the postulated break sizes and locations and initial plant conditions. Provide justification for the ascu:ned

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initial plant conditions. Describe the analytical model used in i

the analysis. Discuss the conservatism in the analysis with regard to mariniting the energy release to the containment. Provide a cable of mass release rate (Ibalar) and enthalpy (Btu /lb ) cs a a

function of time (hr) for the secondary system pipe rupture that results in tha highest contain= ant pressure.

L 6.3 The FLASH computer code is used to predict the mass and energy release to the coar=4== ant during blowdown. Discuss the assump-tions made to obtain conservatively high energy release rates I

from the core for cou nin===t evaluation studies. Discuss the r

f criterian used to establish the tims to DNB considering cl.at a j

• amamevative approach would be to delay DMB until the core was a

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6.4 During bisodena, amersy may be t====Farred from the' steam smeersters to the primary coelmat by eseducatan thromsk the tube walls. Discuse tbs heat transfer correlations used for both the primary and secondary sides of the steem i

gaaerator during blowdown. Give the additinami emergy that could be released to the containment if DNB was delayed on the primary side of the steam generator tubes.

6.5 Frevide e description of the core reflood model. Discuss i

the eammmatiam in the model with respect to emas i{

the amargy release to the comenia===t.

Include the following in year dime===4a= of the core reflood =ad=1:

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(1) Discuse the assumptions made regarding the water remaining l

La the reactor vessel at the and of blowdown. We believe i

a conservative approach for containnsat analyses would be j

to assume that the water remaining in the reactor vessel is saturated and at the bottom of the core.

(2) Discuss the asamupticas made regarding the core flooding

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rate. We believe a conservative approach for contain:sont analyses would be to assuras full H:CS operation.

(3) Discuss the assump'tions made regarding the core quench l

beight and carryout fraction. We believe a conservativa

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approesh for sostainamat analyses would be to assume a carryout fraction of 0.3 and that the core would be quenched.st the 10-foot level.

I (4) Provide a tabulation of the system resistances used in the reflood analysis. If these resistances were detarained far normal system operating conditions, describe the method i

used to extrapolate them to reflood randttious.

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6.6 Meer the core has been recovered with water following a pump section break, boiling will occur to cool the core, and a

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two-phase mixture of steam and water will be generated. Provide sa amelvste showing the bei$t that the two-posse adxture will rise above the cose. If any unter le calemlated to encor the I

steen gamerators, peeride the energy release rate to the con-kl

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6.7 With respect to the heat alaks listed in Table 6-1 of the ygA1,

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identify the heat sinks that are exposed to the aamem1====e atmospheus en both sides, and specify.whether the esposed

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sides. Also, provida the exposed surface area of the mis-e c=11===a== sheathed concreta (itam 7 in Table 6-1).

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6.8 Discuss the method (s) sad the accuracy of the anthod(s) used 3

to determine the fras containm=nt volume. Provide a sensiti-I vity study of the effect of the uncertainty in calculating the full volume on the containment vessel pressure response under loss-of-coelmat meaidaat conditions. Discuss hour the coa-i eah===t full volumn will ba verified.

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6.9 Fot the==haa-paetment analysas, provide assurance that there i

are ao flair restrictions within a subcorgartment that could s

cause pressure differences. Discuss the difference between

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the orifice flair area and the miscellaneous flow area that i

are given for each subcompartment, and how the areas are treated by the computer code C0FRA.

I 6.10 The arrnage===t drawings of the plant indicate that the come=4====t emergency sump is not at the lowest elavetion in the plant, and that a significant amount of water could be reemin=d belour the elevation at which water would begin to overfloor into the emergemey sump. The reactor vessel cavity, normal sump, ref=aling canal, incore instrumentation tunnai, i

pipe tuamm1, and value pit are some of the areas that lie belour the emergency sump. Also Figure t'-17 indicatas that the refueling canal draias to the reactor vessel cavity which draina to the reactor vessel cavity which drains to the normal l

sump, and the emergency sump also drains to the normal sump.

Specify the water level in the cont =h===t following a LOCA j

assuming the containment voltsae below the elevation of tha energancy susep is uniformly filled with water. Discuss the adequacy of available NPSR to the containment spray ptarps in the sentext of Safety Guide 1 " Net Positive Suction Head i

for Emergency Core Cooling and Containment Heat Removal Peeps".

6.11 The intaka senses inat=11ad over the com ni== mat emergency sump e

does met appear to be strecturally t y*=.

For example,

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enly a single, completely esposed wire uneh screen is provided,

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and if the aereen was damaged debris could estar both recircu-

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laties lines. Provide the following information:

i (1) a menu da*=4taa drawing of the intake screen wuich shows hour the screen is attached to the *==*=i-t vessel wall amid floor, i

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i screen will not negate the effectiveness of the entire screen, and l

(3) assurance that the screen cannot be readily damaged ny k'

a missile or large debris that could be carried in the I

water following a LOCA.

t 6.12 Specify the manufacturer of the contain==nt air cooler mits.

Describe the qualification test program that was conducted to detaruine the performance capability of an air cooler Provide a curve of air cooler performance showing mait.-

energy removal rate as a function of containment atmosphere j

temperature.

Since lake water will be circulated through the i

air coolers and since the air coolers will be used under both normal and accident conditions, discuss how fouling of i

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the secondary side of the cooling coils was factored into the analysis of the heat removal capability of an air cooler.

i Specify the service water (lake water) temperature used in I

I the analysis, and provide a table of the =wi-*m ar.d minimm, I

and monthly average temperature of the lake water at the service water system intake.

I 6.13 Identify the ductvork of the contaf rment air cooling syatan that mest remain intact following a loss-of-coolant accident i

to assure that the functional capability of the system is

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not impaired.

Discuss the design provisions to assure that j

the air cooler unit housings and systen ductwork can with-i stand the differential pressures resulting from a loss-of-l coolant accident.

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6.14 Describe how the fusible dropout register (s) associated with the contain nent air cooling system (as shown on Figure 9-12A) l will function.

6.15.

Provide the following information in Table 6-8, Contalument i

Yessel Isolation Yalve Arrangements:

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is shovm.

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clarification.

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6.16 The coarmi===ne vessel penetrations that are exceptions to i

General Design Criterion 56 are listed on page 6-46 of the FSAR. With respect to items 6 and 7; i.e., the isolation valve arrsagements for the containment vossal hydrogen dilution and pur;;e system, rad the containment vessel air l

sample inist and outlet linas, the rationale.~or exempting them from the requirements of GDC 56 was not prasented.

Therefore, discuss why these penetrations are being con-i-

sidered to be exempted frca the requirements of GDC 56.

Ynclude the containment vessel spray lines in the discussion.

6.17 Table 6-4 in the FSAR indicates that the core flooding c:mk sample' and vent 14w are each provided with a single isolation valve outside cone =ine=at.

The cora flooding ennh are not

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considered closed systems inside contain:nent and, therefore, i

General Design Critarion 57 does not apply to these lines.

E Discuss any other basis that you may have which would demon-strate that the valve arrangement meets the intent of the CDC.

i 6.13 Tabla 7-5, SPAS Actuation Stsnaary, indicates that the contain-i meet valves are grouped into three systems. Provide a tabu-latiam of the isolation valves in each oystem and specify the trip setpoints.

6.19 Describe the q==14fication test program that was condmeted to j

assure the operability of comemi-e isolation valves, valve

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drives, pesiefan *M ia-tors, seesing elements, cables, etc.

felicering a IACA or steam line break accident. Identify the i

equipment that was test d. Graphiem117 shoir the envira====tal test conditions as a f==* tion et time.

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6.20 Identify all limas penetratlag the contain===t that do not terminate within areas served by the emergency ventilation i

system. Provide an estimate of the total amount of contain-l meat leakass which esa bypass the areas served by the ty emergency ventilation system.

6.21 Provida the following information with respect to the plant combustible gas control systems, i.e., the hydrogen dilution 1

system, the hydrogen purge system, and the containment air

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recirculation systan:

j (1) Provide sa analysis of the differential pressures that may t

occur following a LOCA for the fan housings and doctwork j

of the containment air recirculation system.

1i (2) Da page 3-3 of the FSAR, the hydrogen purge - dilution i.

system is identified as being seismic Category I.

How-i ever, the purge line is not seismic Category I (as indi-i cated on Figure 9-12A), and is subject to a singla active failure. Since the proposed method of hydrogen control for the plant involves repressurizing the containment, the purge lina should be designed to engineered safety featura standards to assure that continuous hydrogen control i

capability will exist. Therefore, provide a hydrogen purge system that meets the design criteria for an engineered safety feature.

(3) Specify the

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allowable pressure that the contain-ment will be repressurized to using the hydrogen dilution system before hydrogen purge system operation becomes necessary.

(4) Specify the power source for each isolation valve in the hydrogen dilution system (hT 5064, HV 5065, W 5090, and hT 5091) to assure that the hydrogen dilution system is f

not subject to a single active failure, i

s 6.22 Provida a F and I drawing of the containment gas monitoring

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systen. Discess the accuracy of the hydrogen analyser.

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JUN 2 4 1970 the " failure of an interlock" was assumed to occur as the operator or the control system was imposing a demand signal which would normally be limited by the subject interlock. Satisfactory clarification has been obtained from the applicant. This concern has been satisfied. Reference IIC 4.

Onsite Power The response is acceptable on the basis that the applicant under-stands that the continuous rating is the 8,000-hour rating. This statement was not included in the response. Reference IIIB 5.

Environmental Testing (Valves in Containment)

Response is acceptable. Reference IVC Os

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7, I fs sd l D. Tondi ESB-50 Electrical Systems Branch DRS:ESB:DT Division of Reactor Standards cc:

E. G. Case R. Boyd R. Tedesco R. Powell s