ML20235C118

Advises That Review of SAR Completed.No Further Questions Re Status of Plant Foundations Raised.Questions Re Seismology & Reactor Primary Vessel Supports Included for Clarification

ML20235C118
Person / Time
Site:	Quad Cities, 05000000
Issue date:	01/13/1969
From:	Hall W NATHAN M. NEWMARK CONSULTING ENGINEERING SERVICES
To:	Morris P US ATOMIC ENERGY COMMISSION (AEC)
Shared Package
ML20235B311	List: IA-87-111, Submits Rept on Plant,As Discussed in ACRS 150th Meeting on 721012-14.Listed Items Can Be Resolved During Const & Facility Can Be Constructed W/Reasonable Assurance for Operation W/O Undue Risk to Public Safety IR 05000254/1972009 IR 05000358/1972003 IR 05000358/1973001 IR 05000358/1973002 IR 05000358/1973003 IR 05000358/1973005 IR 05000358/1973006 IR 05000358/1973007 IR 05000358/1974001 IR 05000358/1974002 IR 05000358/1974003 IR 05000358/1975004 ML20235B308 ML20235B364 ML20235B401 ML20235B405 ML20235B424 ML20235B456 ML20235B461 ML20235B465 ML20235B477 ML20235B488 ML20235B507 ML20235B552 ML20235B582 ML20235B585 ML20235B588 ML20235B605 ML20235B618 ML20235B623 ML20235B649 ML20235B656 ML20235B662 ML20235B670 ML20235B679 ML20235B683 ML20235B687 ML20235B702 ML20235B708 ML20235B713 ML20235B714 ML20235B718 ML20235B720 ML20235B730 ML20235B734 ML20235B748 ML20235B754 ML20235B762 ML20235B764 ... further results
References
FOIA-87-111 NUDOCS 8709240355
Download: ML20235C118 (3)

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NATHAN M. NEWMARK m-. CONSULTING ENGINEERING SERVICES 1114 CIVIL ENGINEERING BUILDING URBANA. ILLINOIS 618ol 13 January 1969 l

1 Dr. Pe t e r A. Mor r is , D i rec tor Divis ion of Reactor Licens ing U. S. Atomic Energy Commiss ion Washington, D.C. 20545 Re: Contract No. AT(69-5)-2667 Quad Cities Station Units 1 and 2 AEC Dockets 50-254 and 50-265 I l

Dear Dr. Morr is :

Dr. N. M. Newma rk , D r . W. H. Walker, and I have reviewed the Safety l Analys is Report for Quad Cities Station Units I and 2. In accordance with our discuss ions with members of your s taf f, we are not rais ing any f urther l

ques t ions here concerning the s ta tus of the foundat ions of the noted plant  !

s ince this has been a subject of discussion in recent months and we are awaiting further inf ormat ion f rom the applicant on this mat ter.

We are 1is t Ing below two ques t ions which probably cons is t more of clarification of points in the Safety Analysis Report, and these are followed by a request for additional information on matters which are not clearly enough delineated in the report for us to evaluate and in some cases for i t ems which we can f ind essent ially no ment ion in the Saf et y Analys is Report.

Our ques t ions and corments f ollow.

1. On page 2.6-1 under the sec t ion ent i tled "Se lsnology" the re a re

. statements to the effect that " Critical components of the plant will be des igned f or the ground acceleration of 0.12 .9 Safe shutdown will be assured for a ground acceleration of 0.249 For noncr i t ical components , a s tat ic des ign coef ficient of 0.10g with one-third increase in ellowable structural 8709240355 870921 PDR FOIA /

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stresses was used." .By way of clarification, we wish to know whether the term " critical components"' ref ers to the Class I critical structures and cr it ical equipment as lis ted on pages 12.1-1 through 12.1-3 In the same manner, we wish clarification that the noncritical components are those noted as Class !! structures and equipment listed on the pages cited.

2. On page 12.1-4 of the Safety Analys is Report under the heading

" Reactor Pr imary Vesse l Suppor ts ," I t is noted that for the load combination involving the Orerating. Bas is Earthquake, express ion (b) the allowable concrete 's trets was 85% of the ultimate stress for concrete. Does this mean 0.85fl7 Under the loading combinat ion involving the Des ign Bas is Earthquake, expression (c) there is an Indication that stresses do not exceed the yleid point of the netcrial or the ultimate strength of the concrete. Does the latterstatenentrefertofl7

'f the answers to these quest ions are af firnet tve then we should like identification of the situations and details where th these stress limits were employed in proportioning the structure.

Some summary detall is presented in Sect ion 5 and section 12 of the Safety Analysis Report concerning the dynamic analyses of the stack, reactor building, dry well, suppress ion chamber, etc. However, with the exception of i

the presentation for the stack, the remainder of the presentation is not suf ficiently detailed to permit us to draw any conclus ions readily as to the adequacy of the procedures that were employed and the resulting design. Ve l

can find little or no information in the report concerning s imilar studies for the piping and Ins trumentat ion. Accordingly, we should like to request a that you attenpt to obtain so f ar as poss ILle supplementary reports and naterial f rom the applicant which would permit us to s tudy in more detall the

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. methods of analys is employed and the result ing des ign conclus ions. If It is a natter of selecting only a few of these items for further detailed information we would suggest as a first step attempting to obtain Information concern!'ng the analys is- of the dry well, and sone t ypica'l ma in piping runs.

We should also be Interested in any Information or criteria as relsted to the selsmic res is tance of the Class I Ins trumentat ion placed in this plant.

If the infornat ion cited above can be made available to us in the near future, it will aid us in our examination of the plant later this spring when we Intend to nake a site visit. We shall be pleased to discuss the natters noted above . In more detall with your staf f as may be appropriate.

Sincerely yours ,

s W. J. Hall b , k fl(,Qf f)l b) w .

cc: N. H. Newma r k W. H. Walker J. D. Haltiwanger i

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