ML19340F036

Forwards Request for Addl Info Re First Set of Second Round Questions on OL Application

ML19340F036
Person / Time
Site:	Byron, Braidwood
Issue date:	01/02/1981
From:	Tedesco R Office of Nuclear Reactor Regulation
To:	Abel J COMMONWEALTH EDISON CO.
References
NUDOCS 8101160543
Download: ML19340F036 (18)

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f, Y.KECUt Io UNITED STATES y

g NUCLEAR REGULATORY COMMISSION y ') ),.,

g WASHINGTON, D. C. 20555

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JAN 2 1981 Gi Docket Nos: STN 50-454/455 f

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O Mr. J. S. Abel 12h; Wij s

Director of Nuclear Licensing y

a gy Commonwealth Edison Company g

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P.O. Box 767 0

E Chicago, Illinois 60690 m

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Subject:

Second Round Questions on the Byron and Braidwood OL Application During our continuing review of your application for operating licenses for the Byron Station, Units 1 and 2, and the Braidwood Station, Units 1 and 2, we have identified a need for additional information which we require to complete our review. The specific requests contained in the enclosure to this letter are the first set of our round two questions and cover some of the areas of review performed by (1) the Auxiliary Systems Branch, and (2) the Structural Engineering Branch. Some items in the enclosure are statements of staff positions developed after reviewing responses to our first round questions.

Please contact us if you desire any discussion or clarification of the enclosed requests.

Sincerely, i

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l Robert L. Tedesco, Assistant Director for Licensing Division of Licensing I

Enciosure:

As stated l

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Mr. J. S Abel Director of Nuclear Licensing Commonwealth Edison Company

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Post Office Box 767 Chicago, Illinois 60690 ces:

Mr. William Kortier Mr. Edward R. Crass Atomic Power Distribution Nuclear Safeguards and Licensing Division Westinghouse Electric Corporation Sargent & Lundy Engineers P. O. Box 355 55 East Monroe Street Pittsburgh, Pennsylvania 15230 Chicago, Illinois 60603 Paul M. Murphy, Esq.

Nuclear Regulatory Commissico, Region III Isham, Lincoln & Beale Office of Inspection and Enforcemant One First National Plaza 799 Roosevelt Road 42nd Floor Glen Ellyn, Illinois 60137 Chicago, Illinois 60603 Myron Cherry, Esq.

Mrs. Phillip B. Johnson Cherry, Flynn and Kanter 1907 Stratford Lane 1 IBM Plaza, Suite 4501 Rockford, Illinois 61107 Chicago, Illinoi: 60611 Ms. Julianne Mahler Marshall E. Miller, Esq., Chairman Center for Governmental Studies Atomic Safety and Licensing Northerd Illinois University Board Panel CeKalb, Illinois 60115 U. S. Nuclear Regulatory Commission Washington, D. C.

20555 C. Allen Bock, Esq.

P. O. Box 342

. Dr. A. Olxon Callihan Urbanan, Illinois 61820-

' Union Carbide Corporation P. O. Box Y Thomas J. Gordon, Esq.

Oak Ridge, Tennessee 37830 Waaler, Evans & Gordon 2503 S. Neil Dr. Richard F. Cole Champaign, Illinois 61820 Atomic Safety and Licensing Board Panel Ms. Bridget Little Rorem U. S. Nuclear Regulatory Commission Appleseed Coordinator Washington, D. C.

20555 117 North Linden Street Essex, Illinois 60935 Kenneth F. Levin, Esq.

Beatty, Levin, Holland, Basofin & Sarsany 11 South LaSalle Street Suite 2200 Chicago, Illinois 60603 e

ENCLOSURE 010-19 010.0 Auxiliary Systems Branch 010.37 You have not provided an adequate response to Q010.25. Indicate how the (3.*.1)

(3.5.2)

Syron statien essential service water system can be furnished adequate (3.2.1)

(9.2.5) makeuo water for long term plant cooling in the event of loss of function (3yron only) of the essential service water makeup pumps at the river screen house due to a probable maximum flood or tornado generated nissiles. Indicate how onsite wells can perform this function when the accident is coupled with loss of offsite power and assuming a single failure.

It is our position that adequate essential servica water system makeup be assured in the event of a probable maxirc. flood or tornado m'ssiles assuming a loss of offsite power and a single failure in accordance with the recommendations of Regu-latory Guide 1.27.

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010.38 Your response to QO10.15 does not analy:e or evaluate the protactive (3.5.1.1) features provided safety-related equipment assuming internal missiles are generated outside of containment by failures of equipment such is valves, instrument wells, pump impellers, drive couplings and fan blades. You state that protection is. achieved by remote location or physical separa-I tion. provide an analysis and an evaluation of how those protective measures are achieved for a typical safety-related system. The auxiliary fee.'..ater system is considered a suitable example. The analysis should cover the entire system including the diesel and motor driven pumps, routing in the auxiliary building and pipe tunnel, junction with its respective tempering feedwater line, and tennination at the primary containment. Equip-ment and pipe routing drawings should illustrate the protection afforded by

l 010.20 spacing and separation

..;m adjacent high or moderate energy systems and potential missile source. listed above. The evaluation of this typical system should verify the. no damage to safety-related equipment will result which would prev,. r, use of the equipment necessary to reach a safe shutdcwn.

010.39 Your response to QO10.2

.id Q010.16 nas not considered the effect of (3.5.2) multiple missiles genera y:d by one torn.tdo on the various safety-related components located outdgs and on air intakes, exhausts and other building openings.

It is our pos-r.icn that redundancy alone is insufficient assurance against the loe s of safety-related functions in the event of missile impacts in a tor, ado and that specific design capability must be provided each component.

Provide a description of the methods used to protect these structures, system and components from damage by multiple missiles generated by a.~rnador Include the following:

Byron Station Only Describe the protection...ovided to the essential service water cooling towers to prevent damage.ar loss of the fans or motor drives from the impact of multiple verti:al tornado missiles falling into all the cell openings.

Byron /Braidwood Stations Describe the protec e.;n provided to the exposed exhaust stacks of the a.

station emergency dig.el engines to prevent unacceptable damage or stack blockage from a sin;.: or multiple missiles impacting both stacks for one unit.

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010.21 b.

Describe the protection provided to prevent obstruction of flew of ventilating and ccmbustion air to both emergency diesel engines of one unit from the impact of multiple missiles.

c.

Describe the capability of the fuel handling building railroad freight door to withstand the forces of tornado wind and missile impact and the degree of protection or hazard presented by the wash dcwn area structure. Consider the probabilities and potential adverse affects of lightweight objects of large area being impelled through an open, damaged or missing freight door into the spent fuel pool. Cescribe the administra-tive or other controls to assure closure of the freight door during normal plant operatien.

010.40 Provide a response to question QO10.17 and include the following in your (3.6.1) response. Provide.the results of_ analyses.of the. effects.on safety-related _.

systems of failures-in any high or moderate energy piping system in accor-dance with the J. F. O' Leary letter of July 12, 1973, as defined in Branch Technical Position ASB 3-1. Appendix C.

Provide a table which identifies the method of protection provided all safety-related systsa s listed in FSAR Table 3.6.1 from failures of any high.or moderate energy systems. listed in FSAR Table 3.5-2.

Include ff gures depicting the locations of failures relative to the systems of FSAR Table 3.6-1 giving dimensions, locations and protective method for each postulated break or crack in a high or moderate energy system. Include the assumptions uied in your analysis such as flowrates through postulated cracks, pump room areas, sump capacities, and floor drainage system capacities.

010.22 010.41 Your response to question 010.4 is not complete. Discuss the worst case (3.6.1) accidental environmental conditions of temperature, pressure, humidity, potential flooding consequences, and the duration of these effects which would result from an assumed crack, equivalent to the flow area of a single ended pipe rupture in the high energy lines located in the compcrtment between the containment and the safety valve house.

It is our position that a break in these lines not impair the safe shutdcwn capability of the facili ty. Any equipment, which can be affected by the resulting environ-ment including valve operators needed for ss.fe shutdown, shall be qualified to withstand the worst case effects.

010.42 You have postulated a failure of the spent fuel pool cooling system in (g.1.3) section 9.1.3.1 of the FSAR and state that "it is anticipated that evapora-tive heat loss to the environment would limit the pool water to a 180*F maximum if all heat sinks -and a nonadiabatic process is considered." Describe -

the steady state distribution of spent fuel decay heat losses to their ultimate heat sinks in this nonadiabatic process. Provide this information for both the design base case and the maximum stored fuel case. Describe the effects of the evaporative or steam losses from the pool surface in the fuel handling building environment, ventilation system and the handling of condensate from the hot pool within the fuel handling building.

In addition, describe the routing and ultimate residence of the makeu'p water used to prevent boiling after it leaves the spent fuel pool during the postulated loss of normal cooling.

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010.23 During a seismic event, the non-Category I primary water makeup system and the fire protection centrifugal pumps could be lost coupled with a single failure in the common portion of the spent fuel pool cold water return line (OFC03314, FSAR Fig. 9.1-8). Describe the method of cooling after the capacity of the refueling water storage tanks is exhausted, and the non-Category I backup makeup water sources are not available.

010.43 Your response to QO10.7 is not complete. You have not provided a suffi-(9.1.4) cient description of the precise methods, crane interlocks, administrative controls, structures, etc. to restrict the fuel handling building crane hook travel over the spent fuel pool. It is our position that administrative controls alone are an inadequate,means to restrict movement to a parti-cular position. Provide a description of the design used to prevent move-ment of the spent fuel cask laterally over the spent fuel pool while the fuel handling building crane bridge ' positioned longitudinally to handle the spent fuel cask within the spe, fuel cask storage area. Also provide this same information for movement of the fuel handling building crane hook when transferring new fuel to the new fuel elevator.

010.44 Your response to Q010.9 is not complete. You have indicated that tests of (9.2.2) l the reactor coolart pumps performed oy Westinghouse indicate that the pumps cari function satisfacto % for 10 minutes without component cooling j

water supply. Low comorr*p. tuling water flow alams and high component cooling water temperatuc aL e d from the reactor coo: 2nt pump oil coolers are provided in the control room to adicate a loss of component cooling water supply. Operator action can be taken within the T

080..'4 It is our

'l0 minutes available to secure the reactor coolant pumps.

position that the alarm indica: ion of loss of c mocnent cooling water flow to the reactor coolant pu-ps :e safety grade and,mee the requirements for Class IE instrumentation. "cai fy your response accordingly.

010.45 Provide a piping and instrumen:ation di29r2m (?MD) which shows that the

( r,. 2. 4 )

potable and sanitary water s:.s: ems.do not interface with any system that might discharge radioactive.a:erials anc thereby contaminate the potable and sanitary water systems.

Explain how the essential sanice water makeup pumos and the travelling 010.46 (9.2.5)

(Byron only) trash screens shown in FSAR :igare 1.2-15 can accomnocace a failure of the Oregon Dam downstream of the p en Statica concurrent with a low river discharge condition of the Eca River of 564 fee: %.

FSAR Figure 1.2-16 shows a basemat elevation of 654 fce: 91Wut m:2sses or sumps ar the pumps or

creens. This figure is no: 22'si W S* di~h Ze description in FSAR Section 9.2.5.3 where such provisions "'e *Wi5d-Your response to QO10.30 has.. proMed 2n adequate analysis to demonstrate 010.47 (9.3.3)

O safety-related components or that drainage of leakage wa e.may n :MMg resal ting from postulated pipe breaks cr systems is adequate for. ors case cracks in high or moderate e-t 'O Pg.,.

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.%se safety-related components

. 2. s " age by natural routes such as

' C'd or systems. The analysis

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stainvells or equipment hat;'fe er ay *ne non-seismic Category I drainage wr3e o orevent the loss of function system under f ailed conditic-< c M ' #'5

• 5 an example, show that a crack of safety-related conocnen:s :-

'r "s e *Ne essential service water pump in one essential service 3:r-room will not flood out :"e :Y' '"1,r: ant Dumo before operator action can

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be taken to isolate the leak assuming a failed non-safety grade sump alar.:.

system. Worst case locations should be assumed for this example and for other safety-related systems listed in FSAR Table 3.6-1.

It is our position that unless drainage capability by natural or by failed non-seismic Category I drainage systems can be demonstrated, you should provide the following for all areas housing redundant safety-related equipment.

1.

Leak detection sumps shall' be equipped with redundant safety grade alarms which annunciate in the control room. Verify that if operator action is required on receipt of the alarm that flooding of redundant safety grade equipment will not occur within 30 minutes; OR 2.

provide separate watertight rooms and ind$ pendent drainage paths with leak detection sumps for each redundant safety-related component.

010.48 Provide an analysis of the minimum temperature conditions which will be (9.4.5)

(Byron only) reached in the Byron river screen house following prolonged loss of the building unit heaters or loss of offsite power during extreme cold weather. Define f

the minimum operating temperature conditions at the essential service water makeup pump diesel drive units, the diesel oil supply system, and the essential service water lines as a function of time from heating system failure and of ambient temperature. State the reliability of starting the diesel drive units and of provisions to prevent freezing in stagnant water lines during the minimum temperatura period.

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r 010.26 010.49 In Amendment 21, you revised your response to Q010.10 to delete your commit-(10.3.1) ment to verify the operability of the air-operated atmospheric relief valves with no offsite power during low-power testing of the plant. It is our position that you recommit to perfom this verification, or verify that the air-operated atmospheric relief valves can be opened remotely frca the control roca assuming loss of offsite power. Any backup air source for this purpose should be seismic Category I.

010.50 Your response to Q010.33 concerning the effects of flooding resulting from (10.4.5) a failure of the circulating water system transport barrier is incomplete.

You have not provided an adequate response to items (4) & (5) of QO10.33.

Our cencern is for the consequences of a major circulating water system leak in the turbine building caused by failures of such non-seismic Category I components as the main water headers or expansion joints to the condenser coupled with failures of their corresponding butterfly isolat. ion -valves.~ The potential ~ exists to~

flood the turbine building basement to the water level elevation of the cooling tower basin (Byron) or the cooling pond (Braidwood) by simple gravity draining from these large reservoirs.

Describe the designs and locations with the aid of drawings, if necessary, of the watertight barriers provided to prevent floodwater leakage from the turbine building to the auxiliary building or any other safety-related enclosure.

Include a discussion of the consideration given to passageways, pipe chases and/or ca'bleways joining the flooded space to spaces containing safety-related system components. As an example, discuss the means of preventing floodwater from er.tering the main steam tunnel and eventually reaching the auxiliary building at its termination with the main steam tunnel near the safety valve room.

Include in the discussion water exiting the turbine w

s 010.27 building at or above grade level and entering other safety-related enclosures througn watertight barriers removed for maintenance.

010.51 In 0010.13 we indicated that we were evaluating the preheat model steam (10.4. 7) generator (such as those utilfred at the Byron /Braidwood Station) for hydraulic instabilities (water hammer phenomencn potential) and may impose further requirements. Based on these studies we have established the need for a verification test to demonstrate that no damaging water hammer will occur in the steam generator and/or feedwater system. It is our position that you commit to perfom a test using the standard plant operating proce-dures to verify that unacceptable water hammer will not occur. We require that you provide us with a copy of the test procedure prior to perfoming the test.

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010.52.

Your response to. Questions 010.14.and.D10.34.concerning.. cur request for_a..

(10.4.9) reliability analysis for the auxiliary feedwater direct diesel driven pump is not acceptable. As indicated in NUREG-75/023 Supplement 1, dated l

August,1975, we require that you provide.us with evidence of the reliability of this pump to assure that its reliability is at least consistent with the reliability of the emergency diesel generators.

It is our position that the direct diesel drive system for the auxiliary feedwater pump meet those aspects of Regulatory Guide 1.9 " Selection, Design & Qualification of Diesel-Generator Units Used As Standby (Onsite)

Electric Pcwer Systems at Nuclear Power Plants." as are applicable to a diesel-pump unit. We recognize that Regulatory Guide 1.9 and its referenced IEEE Standards are designed for diesel-generator units but that many of its requirements can be adapted to a non-electrical output device. Clearly such

010.28 requirements as starting, load acceptance, vibration, overspeed, automatic control, and site testing are applicable to a diesel-pump unit as well as a diesel-generator unit.

Provide a comparison analysis of the reliability of similar features between the emergency diesel-generator and the auxiliary feedwater diesel driven pump. Include comparative reliabilities of the folicwing subsystems:

starting, combustion air, exhaust, flywheel, fuel oil, lubricating oil, cooling, governor, control, protection, surveillance, and cubicle environ-ment.

The comparative analysis shall be based on the applicant's or other's experience with similar equipment or subsystems. Where similarities between proposed existing equipment and subsystems are poor, the applicant shall justify his reliability assessment based on the specific differences between the subsystems. Test data comparisons of existing duplicate or nearest similar diesel drive arr aaments should be included.

010.53 Provide-a response to our March 10,'1980 letter concerning your auxiliary (10.4.9) feedwater system (AFS) design. This response should. include the following:

1.

A detailed point-by-point review of your AFS design against Standard Review Plan Section 10.4.9 and Branch Technical Position ASB 10-1.

2.

A reliability evaltation similar to that performed for operating plants (refer to Enclosure 1 of the March 10, 1980 letter) and discussed in ttuREG-0611.

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010.29 3.

A point-by-point review of your AFS design, technical specifications and operating procedures against the generic short term and long term requirements discussed in the March 10, 1980 letter.

4.

An evaluation of the design basis for the AFS flow requirements and verification that your AFS will meet these requirements (refer to of the March 10, 1960-ietter).

We note that your present AFS design provides two safety grade auxiliary feedwater pumps. We wish to point out to you that previous reliability studies for two pump auxiliary feedwater systems have indicated that j

installation of a third automatically started pump powered from a redundant emergency bus significantly improves AFS reliability.

It is our position that you achieve a system reliability comparable to other recently approved operating Westinghouse plants-with three safety grade auxiliary feedwater pumps.

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130-5 130.0 Str_uctural Engineering Branch 130.06 Question 130.06 (3.7.2.1)We have reviewed your response to Question 130.6 and we conclude that it is not adequate and not acceptable for the following reasons:

1) Selection of SSE and OBE Design Earthouakes A considerable portion of your response is based on the conservatism you feel is in toe SSE and OBE design earthquakes. You also presented arguments for reducing the design earthquake to those originally proposed in the PSAR (zero period acceleration of 0.06g for OBE and 0.12 for SSE).

These values have been subsequently increased to 0.09g and 0.29 respectively and rationale for the Regulatory staff position was stated in the Question 2.5.63.

Furthermore,- on the basis of further investigation the staff came to the conclusion that the deconvolution procedures are not acceptable and that the Regulatory Guide 1.60 Design Response Spectra snould be applied at the foundation level.

2) Effect of Foundation Size on Design Spectra The response suggests that the design spectra can be reduced based on previous studies. performed by Dr. Newmark for the Diablo Canyon Site.

These studies justify reduced effective spectra as a result of considering the effect of foundation size on design spectrum. You pointed out in the response that the reduced effective spectra were developed for the specific site of the Diablo Canyon Plant and the basic reason for its acceptance

130.6 was the postylated near-field earthquake.

Since the Byron /Braidwood sites are located in an entirely different tectonic province the argument whicn was used in case of Diablo Canyon application cannot be applied to the subject sites.

3) Conservatism in Analysis The staff does agree that the three components of earthquake motion are probably not the same acceleration. The magnitude of the actual acceler-ation of each component should be found by means of a 3-dimensional analysis.

It is the position of the sta'ff that the response spectrum for vertical motion can be taken as 2/3 of the response spectrum for horizontal motion for the Western United States only.

For other locations, the vertical response spectrum should be the same as that given in Regulatory Guide 1.60.

(See Enclosure)

As far as the damping values are concerned, the referenced report, NUREG-CR 0098 was developed for a' specific purpose of evaltiating seismic risk of nuclear plants which are already operating.

The damping values contained in that report cannet be applied in licensing of new plants.

The response claims that the elastic analysis which 1r used in design of new plants may be unreasonably conservative.

In view of the fact that there is a lot of safety-related equipment which might produce catastrophic consequences in case of excessive deformation of supporting members, this position of the Regulatory staff is not unreasonable. You neglected to mention in your response that the referenced criteria for the Diablo Canyon plant stipulate that the ductility of 1.3 for concrete and 3 for steel are

130.7 for turbine building and intake structure.

These structures are non-Category I per se and the only reason that they have been reviewed by the staff was that in certain locations they are housing some safety-related equipment. Thus the criteria which are applicable to those two structures cannot be automatically applied to all Category I structures.

4) Evaluation of Structures using 0.09g OBE and 0.20g SSE Regulatory Guide 1.60 Spectrum The evaluation of structures using the Regulatory criteria provided in the response have been reviewed.

It is recognized that there is a general increase in the stress level of many structural members.

We find, however, that without re-analysis of the affected structures and determination of the shear forces and moments imposed by the new loads the evaluation cannot be considered to be conclusive. You are, therefore, requested to compare the structural responses of Category I structures and the design parameters (bending moments, shears and axial loads) actually used in design of Byron /Braidwood plant with those which would have been obtained if the criteria stated in Question 130.06 were used.

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130.8 i

130.9 For the river screen house at the Byron station there is a marked (3.7.2.4) increase in the response spectra for most of the frequencies of interest in structural design. The technical position of the Regulatory staff is that the results of the two methods, i.e., the half space and the finite element method should be enveloped in order to be used

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in the design. This position is stated in the Enclosure and designated as Method 3(a). As an alternate solution, the staff would find acceptable the two other options which are designated in the Enclosure i

as Methods 3(b) and 3(c). You are requested to perforn a seismic 4

analysis using one of the above noted three options, quantitatively I

assess its impact on structural desigg of the river screen house at

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the Byron plant and submit the results for our review.

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130.9 1

SUMMARY

dFSEBINTERIMLICEflSING POSIiI0ti5 AND SiATUS OF SRP REVISION PARCH 1979 SRP SECTION INTERIN LICENSING POSITION IN i

ADDITION TO OR DIFF. FROM THOSE LISTED Ifl CORRESPONDING SRP SECTIONS 3.7.1 Seismic Input 1.

Use of site dependent input design spectra is acceptable if the input l

spectra are reviewed and accepted by GSB (Ref. SRP. Section 2.5) i 2.

For western United States (West of Rock-les), the response spectrum for vertical l

motion can be taken as 2/3 the response spectrum for horizontal motion over the entire range of frequencies.

3.

Methods for implementing the soil-struc-ture interaction analysis should include i

both the half space lumped spring and mass representation and-the-finite- -

L' went approaches. Category I struc-tt..es, systems and. components-should be.

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designed to responses obtained by any j

l one of the following methods:

' a) Envelope of'results of the two l

methods, b) Results of one method with conserv-ative design consideration of impact from use of the other method, i

el Combination of-(a) and (b) with provision of adequate conservat, ism in design.

4.

Consideration of the effects due to accidental torsional forces in design (as a minimum, the 5% times base dimen-ston off-setting criteria should apply),

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