ML19317E240

Forwards Seismic Review of FSAR Vols 1 & 2 Per Request

ML19317E240
Person / Time
Site:	Oconee
Issue date:	09/02/1969
From:	Sharpe R JOHN A. BLUME & ASSOCIATES, ENGINEERS
To:	Morris P US ATOMIC ENERGY COMMISSION (AEC)
References
CON-AT(49-5)-3011 NUDOCS 7912170491
Download: ML19317E240 (13)

Text

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. 2. b . ' f September 2, 1969 Dr. Peter tiorris, Director '

Division of Reactor L: censing . .

U.S. Atomic Energy Commission , $'p , ( '-Q Vashington, D.C. 20343 /'

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Subject:

Oconce fluclear Str.sion Units 1, 2 and 3 .i D,$,. ,' ',,' ' '/

Duke Power Company ,

Final Safety Analysis Report ' D' , ' ,gk Y Docke t ilos . 50-269, 270, 287

Dear Dr. tiorris:

In accordance with your request we have performed a general review of the FSAR, Volumes I and 2 for the Oconce Plant and also made a visit to the plant site on August 13 and 14.

We are enclosing herewith three copies and tripoftoa the list of questions and comments resulting from this review site, i t is possible that additional questions will be generated af ter re-ceipt of the information requested in the attached.

Very truly yours, JOHit A.BLUME & ASSOCIATES, Et1Gif1EERS

//7/7_) 03 f Roland L. Sharpe Executive Vice President RLS:Js Enclosure t

3 SlilS?!!C RiiviliW OCGNiil:

NilC1.EAll STATION UNITS 1, 2 AND 3 DUKE l'OKElt CO31PANY

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(Docket Nos. 50-269, 270, 287}

The attached is a list of comments resulting f the reference documents (see attached s li t) f rom a preliminary review of view was primarily directed toward seismic considor the Oconce The re- Plant.

' review of the structural characteristi erations, but a general

'the questions and comments have been arrcs and design concepts was perform anged in the following categories:

I.

Introduction and Summary -

II.

Reactor Coolant System 111.

Reactor Duilding, I V.

Interior Structure V.

VI. Class I Piping Systems - Reactor Building Class I Piping Systems - Other Buildings VII. Class I Equipment Vill.

Auxiliary Building IX.

Turbine Building i,3

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I. INTI:0 DUCT ION ANI) SU'Dtil:Y lt is stated that Appendix IC, " Systems Design Critoria for Natural Phe-nomena" will be submitted later. We will complete our review of Section 1 of the FSAR when this appendix is received.

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l JOHN A. E l_ U !.UC & % 5CO:ta F r D . r,c E.- G

11.

REACIOR COOLANT SYST121 1

Picase describe in detail the uset) to determine that the nuclearanalysis and/or testing es procedu vessel, steam generators, reactor o co lsteam supply system (

meets Scismic Class I criteria ant pumps, piping, etc.)

following: .

a.

Include in this discussion e th A detailed description and sketch of the system, including a discu of the mathemat'ical model(

dom and methods of lumping mass ssion of the degrees of free-crties, etc.

es, determining section prop-b.

A discu<sion of the analytical '

methods of computing periods procedures used, including the

, moda factors, and the procedures f shapes, modal participation displacements, shears, and momor computing designons, accelerati

c. ents.

A discussion of the possibilit coupling between the nuc1 car stey and significance of dynamic structure (internal structuret wi hiam system and the supporting ing).  !

d.

n the containment build-A listing of the damping values

c. used. i The results of the analyses \

used, the natural periods, mode sh, including the input and placements, shears, moments, and strapes, accelcrations, di esses.

3-u JOHN ^ - Su>"

111 11/4101: bu t ! )CG 1.

In order to properly evaluate the adeqtacy of the design of the contai nunt st ructure, it is necessary to know which loading con-ditions are actually critical to the de;ign. To facilit ate this evaluation, please provide a sunnary of the stresses frou the var-ions loading conditions (dead load, thernal, pressure, wind, carth-quale, etc.), the total combined stress, and the allowable stress for each critical point on the structure and each load conbina-tion case. Explain how the results of the seismic analyses were incorporated in the final design. -

2. On page 5-12 it is statul that the finite elenent nesh for the base slab was extended down into the foundation naterial to take into consideration the clastic nature of the foundation material and its effect upon the behavior of the base slab. This exten-sion below the base slab is apparently not shown on Figure 5-4,

" Reactor Building Finite Elenent Mesh." Please provide a draw-ing of the nesh used to account for the effects of the foundation waterial.

3. It is understood that the tendon access gallery is structurally separated in the vertical direction from the base slab. Picaso describe how the prestress gallery was considered in the design of the base slab.

4. On page 5-14, it is stated that the liner was treated as an in-tcgral part of the structure. Does this nean that it was included in the finite element resh of the containment structure? If so, please provide a detailed sketch of the mesh.

5.

For which loading cases do the isostress plots shown on Figures

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5-6 and 5-7 apply?

6. In regard to the seismic analyses:

a. What were the pcriods of vibration as computed for the con-tainment r,odel?

b. Please justify the assunption of fixity of the base slab as shown in the noici on Figure 5-10.

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c. Once the inertia forces were obtained as explained on page 5-18, how ucre the inteinal forces in the containment walls and base slab computed?

7. bliat provisions were nade to transfer seismic and wind shcar for-ces across construction joints?

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8. blien the, structural tests are complete, plea'sc provide a summary of the predicted stresses, strains, and deficctions versus the actual recorded values for cach increment of pressure testing for each instrument. Provide an evaluation of the results of these tests as related to the adequacy and conservativeness of the design and analysis assumptions.

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.;O M N A C LUM E G ASS OCt AL .3. C* '0"s ': d R C

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IV.

INTElt10R ST:tuC_TURE 1.

Please describe the procedures utilized to ascertain the scismic adequacy of the interior structure within the containment structure.

Include in this discussion a sketch of the mathematical model used o

and the results of the analysis in terms of periods, accelerations, shears, taoments, displacements, etc. ,,

2.

The finito clement mesh shown for the containment building appar-ently does not include the interior structure. What influence does the interior structure have on the stresses in the base slab .

computed by the finite element analysis? Ilow was the base slab designed to resist the scismic shear and overturning moment from the interior structure?

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! .JO"N A. C LU V E C: AC COCW,5 0, CNGNEER G

V. Cl. ASS I PIPING SYSTDIS - El: ACTOR liUILDl';G

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1.

Please describe in detail the analysis procedures used to deternine that,the Class I piping within the reactor building meets seismic Class I criteria. Include in this discussion the following:

,- a.

% e methods utilized to determine the input (response spectra) for the piping analyses. Also, please provide graphs of these spectr$at the various elevations in the structure. Include a comparison of the postulated spectra for the site and the spec-tra determined from time-history used in the analysis of the reactor building.

b. Typical mathematical models for several piping systems for the Oconce plant. Include specifically the model of the main steam line (and steam generator).

c.

A discussion of the analytical procedures used, including the methods of computing the stiffness and mass matrices, periods, mode shapes, and participation factors, and the procedures for computing design accelerations, displacements, shears, moments, and stresses.

d.

The detailed results including items listed under c, above, for several systems as specified in b, above.

c. Provide for all systems a summary of the stresses from the various loading conditions (dead load, thermal, earthquake, etc.), the total combined stress, and the allowabic stress for each critical point on the system and for each load com-bination case.

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VI. Cl. ASS ,I PIPING SYSTliMS - OTilER BUILDINGS

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1. We understand that the Class I pipes being designed by Duke Power Company are designed for a uniform static coefficient equal to the peak spectral acceleration from the spectrum for the appropriate support point in the auxiliary building. For these systems, please f

provide the following: ,

a. Summarize the procedures currently being used,

b. Demonstrate that the method used is conservative, that is, it results in seismic stresses equal to or greater than those that would be obtained by dynamic analyses.

c. Provide for all systems a sunmary of the stresses from the various loading conditions (dcad load, thermal, carthquake, etc.), the total combined stress, and the allowabic stress for cach critical point on the system and for each load combination case.

2. We niso understand that spectra from the highest point in the Auxiliary Building at which the piping systems are anchored are used for piping in both the Auxiliary Building and the Turbine Building. tiill not the spectra for the two buildings be dif-ferent and exhibit different amplifications at different frequen-cies? lias rocking of the Turbine Support Structure been considered?

Please demonstrate that if the spectra from the Auxiliary Building are utilized for pipes in the Turbine Building, the resulting scismic stresses will be conservative.

t JCHN A. CLUf.iC & ACSOCMTEC, CSM. M u

Vi1. Cl, ASS I If{ll!P.1111NT

1. Please describe the procedures used to assure that the Class I equipment (tanks, pumps, etc.) meet the scismic design criteria.

Provide a summary of all such pieces of equipment and the types y

and results of the scismic analyses performed, i

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f JOHN A. BLUME & ASSOCIATES. ENGIN'_~TRS

VIII. _AUXII.I ARY BilII DING 1.

Please provide a sketch and detailed description of theematical math model used for the dynamic analysis of the auxiliary building a the results of this analysis.

e Include in these results the re-sponse spectra developed for use in the analyses of piping system .

l JCHN A. GLUME a ASSOCIATES, ENGINEERS

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SEISMIC REVlliti REFERENCE IOCUMENTS Oconec Nuclear Station Units 1, 2 and 3 Duke Power Company Docket Nos. 50-269, 270, 287 ,

Construction Permit Preliminary Safety Analysis Report Vol. 1 Preliminary Safety Analysis Report Vol. 2 Report No. I to the ACRS dated May 24, 1967 Report No. 2 to the ACRS dated June 16, 1967 Addendum to Report No. 2 to the ACRS dated July 6, 1967 Op rating License Final Safety Analysis Report Vol. 1 Final Safety Analysis Report Vol. 2

" Reactor Internals Stress and Deflection due to Loss-of-Coolant Accident and Maxinum !!ypothetical Earthquake," Babcock G h'ilcox Topical Report bah'-10008, Part 1, June 1969.

JOHN A. OLUME & ASSOCI ATES. ENGINEERS

.