ML20235D908

Forwards Request for Addl Info Re Design Repts 7 & 8 Containing Structural Design Info & Analysis of Consequences of Small Steam Line Break within Drywell,Respectively

ML20235D908
Person / Time
Site:	Brunswick, 05000000
Issue date:	04/07/1971
From:	Morris P US ATOMIC ENERGY COMMISSION (AEC)
To:	Jackie Jones CAROLINA POWER & LIGHT CO.
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 8709250403
Download: ML20235D908 (4)

Text

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I Dit tbution:

' Dockee (2)d:::-- ACRS (16)

AEC PDR (2) WNyer (2)

DR Reading Newmark & i DRL Reading Apg BWR-1 File 7 197l CKBeck Docket Nos. 50-324' MMMann and 50-325 SHanauer FSchroeder Asst. Directors, DRL EGCase, DRS j

Mr. J. A. Jones RRMaccary, DRS I Senior Vice President SERS Group (2)

BGrimes, DRL Operating & Engineering Group Br. Chiefs, DRL/DRS Carolina Power & Light Company SMKari 336 Fayetteville Street Raleigh, North Carolina 27602 WRButler DKartalia, oGC

Dear Mr. Jones:

Compliance (2)

We have reviewed Design Report Hos. 7 and 8 on the Brunswick Steam Electric 1970 andPlant February which26, you 1971. filed with your ~1etters dated December 22 ,

As you know Design Report No. 7 contains structural design information for the Brunswick containments and Design Report No. 8 provides your analysis of the consequences of a small steamline break within the drywell.

We have found. a need for additional information, as detailed in the enclosed " Request for Additional Information," to complete our review of the matters addressed in these reports.

If you have any questions on the above, please contact me.

Sincerely.

Originel SWd by Fras: S.'. .? r

/

[ p Peter Division of Reactor Licensing A. Forris, Director

Enclosure:

Request for Additional Information ec: G. F. Trowbridge, Esquire Shaw, Pittman, Potts, Trowbridge 6 Madden 910 17th Street, N.W.

Washington. 7. C. 20006 OFFICE ) .D.E1 [.SM.d7.1..

/ s IN. ..pR ,,, .Q WRButler:e oy m

SURNAME > ..DERMth... ...... .m_. .

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4./5/71 oArE >

Form A EC.318 (Rev, B-63 4/h/.71 4 /,[. / 71 4/7./7,1 u.s. Govtna um remTwo orrict 2196,- o-364-s os 0709250403 870921 Z8 1 PDR

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UNITED STATES

- ATOMIC ENERGY COMMISSION WASHINGTON. D.C. 20$45 April 7, 1971 Docket Nos. 50-324 and 50-325 Mr. J . - A. Jones Senior Vice President Operating & Engineering Group Carolina Power & Light Company 336'Fayetteville Street Raleigh, North Carolina 27602 .

Dear Mr. Jones:

We have reviewed Design Report Nos. '7' and 8 on 'the Brunswick Steam Electric Plant which you filed with your letters dated December 22, 1970 and February 26, 1971. As you know Design Report No. 7 contains structural design information .for the Brunswick containments and Design Report No. 8 provides your analysis of the consequences of a small steamline break within the drywell.

We have found a need for additional information, as detailed in the enclosed " Request for Additional Information," to complete our review of the matters addressed in these reports.

If you have any questions on the above, please contact me.

Sincerely, Peter A. Morris, Director Division of Reactor Licensing

Enclosure:

l Request for Additional Information cc: G. F. Trowbridge, Esquire Shaw, Pittman, Potts, Trowbridge & Madden 910 17th Street, N.W.

Washington, D. C. 20006

REQUEST FOR ADDITIONAL INFORMATION 1.0 DESIGN REPORT NO. 7 1.1 The statement that the suppression chamber liner is in complete tension under accident loading requires further substantiation in view of the temperature gradients assumed in Figure 45. A gradient profile different from that indicated in Figure 45 which would show a significant drop at the liner / concrete interface due to the thermal effects of the interface boundary and the masses of water and concrete is likely.

Consider such a gradient to exist during the containment pressurization accident and, assuming that a long-term period is required before stabilization of the gradient can take place, discuss whether the suppression chamber liner is then in a state of tension, or whether compressive buckling may take place. (ref: p.1-3, Fig. 45).

1.2 Describe the checking procedures which will be used as a means of design control (as part of the AEC Quality Assurance Criteria for Nuclear Power Plants - 10 CFR Part 50 - Appendix B) to determine the validity of the computer analyses, with regard to correct input data, procedures and results, as applicable to the design of the containment structure.

1.3 Furnish the design stress limits used in the analyses to check the adequacy of design as referred to on pages 1-3 and 6-1.

1.4 Since the analysis presented is limited to axisymmetrical loading, describe the manner in which non-symmetric loads, such as earthquake, were taken into account in the design of the containment structure.

1.5 The combined load equations on page 4-7 do not include local load effects, such as jet load or pipe thrust and the statement on page 6-1 does not provide an adequate explanation of how these loads are accounted for.

Furnish the loading assumptions made, and the design methods and criteria used for these loads.

1.6 It appears that the cracking pattern at the top and bottom of the drywell which is described on pages 4-20, 21 and in Figure 24 is assumed to be l oriented at 45 . Justify the use of this angle of cracking, since it may not be conservative with respect to the amount of shear reinforcing provided.

l 4

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2.0 DESIGN REPORT NO. 8 2.1 State why the vent pipe penetration and the personnel lock penetration will be the only sleeves designed to meet the governing load equation of 1.1D+P20+T340 as noted on page 6-2, in view of the fact that all containment penetrations may be subjected to the loadings of this equation.

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