ML20085G372

Third Interim Deficiency Rept Re Design of Interior Concrete Structure of Reactor Bldgs.Initially Reported on 731001. Evaluation of Reactor Bldg Equipment & Piping Design Adequacy Expected by 740501 & Seismic Analysis by 740215

ML20085G372
Person / Time
Site:	Browns Ferry, Sequoyah, 05000000
Issue date:	02/13/1974
From:	Gilleland J TENNESSEE VALLEY AUTHORITY
To:	Knuth D US ATOMIC ENERGY COMMISSION (AEC)
Shared Package
ML20085G374	List: ML20085G372 ML20085G384
References
10CFR-050.55E, NUDOCS 8308290021
Download: ML20085G372 (4)

Text

r PC O s M -J TENNESSEE VALLEY AUTHOAiTY FJr\,! \ pr

. CHATTANOOGA, TENNESSEE 37401 b' '

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February 13,197 +1 :QL amug,ymifAd) 05r?rYEN$$$

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Mr. Donald F. Knuth, Director Directorate of Regulatory Operations U.S. Atomic Energy Comission Washington, DC 20545

Dear Mr. Knuth:

TVA made an initial report to the AEC-DRO Region II office by telephone on October 1, 1973, of a potential design deficiency of Sequoyah Nuclear Plant units 1 and 2 Interior Concrete Structure. Interim reports were submitted on October 29, 1973, and December 14, 1973.

The enclosed report is submitted as a third interim report on the potential design deficiency.

Very truly yours, y,'$ /

v J. E. Gilleland Assistant to the Manager of Power Enclosure '

CC (Enclosure);

Mr. Norman C. Moseley, Director Directorate of Regulatory Operations U.S. Atomic Energy Commission Region II - Suite 818 230 Peachtree Str set , 2M.

Atlanta, Georf f' M303 p e; e; .

8308290021 740213 PDRADOCK05000g S

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IUTEICIAL DE3IC:i ECTICI2::CY  !

IETERIM R2 TORT I!O. 3 An initial report of a potentic1 design deficiency in the interior concrete structure of the reactor buildings for Units 1 cnd 2 of our Sequoyah Huclear Plant was made by telecon to AEC-ERO Region II Office on October 1,1974, in co=pliance Trith 10CFR50.55(e). Interim reports were submitted to AEC-DRo Headquarters on October 29, 1973, and December 14, 1973. The second interim report extended the problem to include all Category I structures.

Reactor Buildings - Units 1 and 2 Interior Concretc Structure The effects of the shift in the location of peak response of the floor responso spectra are not quite co=plete. Interim report number 2 indicated ,

the evcluation of the design adequacy of the reactor building equipment and pipisg therein would be co=pleted by Februsry 1,1974. We now expect this to be completed by I'ay 1,1974.

A sumary of the major items still outstanding, with evaluation to date is as follows: -

1. Ice Condenser There is a slight increase above the previously developed Sequoyah seismic design loads. This increase occurs in the tangentici distri-bution of the seisnic i= pact loads at the center portion of the 48 foot ice colu ns between the lattice frames and the ice baskets.

The increase in the seismic i= pact tancential load distribution is ,

approximately seven percent (70) which is we.11 within the present design capability.

D O g 1

2. Auxiliary Equipment Reanclysis of the ernipulator crane verified the structural adequacy of the manipulator crane durin5 the SS2. The fundamental frequenc'/ of the manipulator crane was below the peak of the original spectra and there is no appreciable effect for the revised spectra where the pech was shifted to a higher frequency. Equipment rodification is not required.

3. Nuclear Steam Supply System (!!SSS) and I!SSS/ Supports

a. The reactor vessel internals have been examined and the resulting ctress values are lover than for the standard design plant. No design change is required.

b. SCRAM time still bein5 evaluated; however, it is not expected to -

be affceted.

<c. The CRDM seismic support frame has been evaluated end a desi6n change is not required.

d. NSSS piping seis=ic loads are still being evaluated but are apper-ently within ellouables.

c. IISSS/ support seismic loads have increased in isolated instances.

The design adequacy of the systems and supports are still being evaluated.

Work still in progress includes the normal verification of the seismic desiCn adequacy of other =ajor reactor equipment. and their internels such as the steam generators, reactor coolent pgs,

.and pressurisers. They vill be evaluated for the revised spectra when scheduled and no design uroblems are enticinated.

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4. Auxilit.ry Piping Auxiliary piping systems analyzed to date have been reenclyzed for the revised spectra. Piping and fluid system ec.-ponent stresses are within allovables for these systems. Sone support locations require change.

Auxilicry piping systens yet to be done vill be cnalyzed using the revised spectra.

Other Seisnic Category I Structures Auxilicry control Building The dynamic seismic analysis end response spectra for attached equipment for the auxiliary-control building, based upon the revised modulus is expected to be co=pleted by Februcry 15, 1974. Seis=ic Category equipment -

and pipin6 supported by the structure vill then be examined.

It appears the additional ice condenser equip =ent buildings are the only other Category I structures to be affected by the revised modulus. The analyses for these tuo structures will begin upon completion of the cuxiliar/-control building.

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