ML19316A641

Responds to AEC Requesting Addl Analysis & Info Re Consequences of Postulated Failure of Main Steam & Feedwater Pipes Outside Reactor Bldg.Summary of Current Status of Design Engineering Review Encl

ML19316A641
Person / Time
Site:	Oconee
Issue date:	12/29/1972
From:	Thies A DUKE POWER CO.
To:	Anthony Giambusso US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 8001130177
Download: ML19316A641 (5)

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-I December 29, 1972

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Mr. Angelo Giambusso Deputy Director for Reactor Projects

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Directorate of Licensing U. S. Atomic Energy Commission Washington, D. C.

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i Re: Docket Nos. 50-269, -270, and -287 I'

Dear Mr. Giambusso:

Please refer to your letter of December 15, 1972 requesting additional O

analysis and other relevant information needed to determine the con-sequences of postulated failure of the main steam and feedwater pipes j

outside of the Reactor Building for Oconee Nuclear Station Units 1, 2, and 3.

i Our response to the request for a description of the analysis to be made, proposed' modifications, ana schedule to be followed are given by the attached summary of the current status of Design Engineering review.

Very.truly yours,

.?4'S A. C. Thi;es ACT:vr.

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,.DUKL POWEi. COMPANY OCONEE NUCLEAR STATION j

CONSEQUENCES OF MAIN STEAM & FEEDWATER PIPING RUPTURE

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CURRENT STATUS OF DESIGN ENGINEERING REVIEW

, j,, '. - "'" 1, 12-29-22 1)

INTRODUCTION In response to the AEC/ DOL's 12-15-72 letter and attached guidelir.cs on the consequences of postulated Mair Steam and Fcedwater piping failures in struc-

. tures other than the Reactor Building, Duke's current position is as outlined herein.

2)

PROBABILITY OF FAILURE Although,0 uke is continuing to review the consequences of postulated pipe ruptures, such ruptures are not considered credible for the Oconee Nuclear Station based on the following:

Oconce's Main Steam and Feedwater Systems are designed to preclude pipe a.

ruptures based on conservative engineering practices, b.

The only basis for postulating a line rupture is stres' criteria. The following describes representative stress conditions for the Main Steam

-(h and Main Feedwater Systems at Oconce.

Main Steam lines are 100 percent cold pulled so that as the line heats up, all thermal expansion stresses are essentially climinated throughout the_ system.

For example, at the Reactor Building penetration (terminal end), there is only 1100 psi maximum thermal stress during normal.

operation; this is only about 4 percent of the ANSI B31.1.0 (1967) Code allowable stress. This fact coupled with the safety factor built into Code allowabic stress values indicates a tremendous amount of conserva-tism. The Main Feedwater System is r.ot cold pulled since it operates at a lower temperature; however, similar to the Main Steam, there is only 3645 psi maximum thermal stress during normal. operation at the Reactor Build.ing penetration (terminal end). Again, this is only about 16 perc'ent of the ANSI B31'.l.0 (1967) Code allowable stress.

Overpressure capability of the piping based on wall thicknesses actually c.

used is as follows.

It should be noted that these figures are extremely conservative as they are based on ANSI B31.1.0 (1967) Code equations.

Normal Operating Actual Code Pressure,

Percent Pressure Capability Margin Main Steam:

910 psig 1093 psig 20 Feedwater:

1070 psig 1383 psig 29 h

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d-The safety related portions of,.,these systems are Duke class F indicat,I.ng that,the materi.als of ' construction.wcre procured

<fabricat,cd..atested and documen.t.ed,similar to a nucicar system as can be denoted from the following:3. y 1)

Piping Materials----------Traceable

. 11)., Velding Fi ll er Metal------Traceabl e iii) NDE

---

100 percent X-ray s

.Iv). Piping Materials' QA-------Inspection at fabricators plant and

_s.ite receiving inspection:-

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. v)

Docum'entation ---------.--Required

. vi)

Support Design QA ------ '-As outlined in FSAR 1C.3.4.5 3)

POSTULATED RUPTURES REVIEWED TO DATE By. definition of the attached guidelines to the 12-15-72'AEC letter, Duke i

has reviewed double-ended ruptures of the two Main Steam and two Main Teedwater lines at the terminal ends of the Reactor Building penetration anchors.

Other postulated rupture points are currently being defined; however,' preliminary studies indicate that additional postulated break

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points will have very little effect on the ability to shut the unit down t

and ma'intain it in the safe shutdown condition.

{7 )

CONSEQUENCES OF POSTULATED RUPTURES -

4 a.

Vest Steam Generator Main Steam Line. This line runs external to the

. Reactor and Auxiliary Buildings until it enters the Turbine Building and does not pass near any essential equipment necessary to shut down safely and maintain the reactor in a safe shutdown condition.

b.

East Steam Generator Main Steam Line.

This line leaves the Reactor Building wall and passes through one corner of the East Penetration Room as shown on attached Sketch PO-222.

Postulated pipe whip, jet Impingement or reaction forces resulting from failure of this line would not damage any equipment necessary to shut down safely and main-tain the reactor in a safe shutdown condition.' However, pressure effects-and steam concentrations might possibly pose a problem in the penetration room.

c.

East'and West Main Feedwater Lines. These lines enter the Reactor Building through.the East Penet ation Room as shown on attached Sketch PO-222.

Postulated failure of either of these lines could damage several auxiliary systems and related electrical components due to pipe whip, Jet impingement and reaction forces.

Feedwater for secondary side cooling is assured to the unaffected Steam Generator by either the steam-driven Emergency Feedwater System or the backup Auxiliary Service Water System.

Pressure effects and steam concentra-tions could pose a problem in the penetration room.

Since the

. postulated failure can occur on either the upstream or the downstream I

side of the Reactor Building isolation check valve, both cases are i

bcIng analyzed.

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-d. 'Cuntrol Room Integrity.

Revie Vof.the. general arrangement of high

'Mergy systems relative to the' c'oritrol' room' indicates' that pipe whip,

. Jet' impingement and 'reacti0n forces ~would not affect ~ the~1ntegrity of

" thel 'cdntrol' room.'

A' structural reinforced concrete wall is located between the control room and t.he penetration room.

5)

POSSIBLt MODIFICATIONS TO THE~ STAT'l0N'

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At present~, possible modifications t'o' the station to reduce' the~effect's of pressure and steam concentrations as described in 4)b. and 4)c. to accept-able limits are being analyzed.

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,U a.

East Steam Generator Main Steam Line. As shown on the attached Sketch I

P0-222, the existing north penetration room wall may be modified to include low pressure blowout panels to relieve pressure and to provide l

a steam escape route for. the postulated failure of t'his line. The-addition of a new wall to remove the main steam line from the penetra-tion room environment may be added along Column Line 65 as shown on l

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S.k.e..tch PC-222.

b.

East and k'est Main Feedwater Lines.

One way low pressure blowout panels may be added to the new wall along Column Line 65 to relieve pressure and provide a steam escape route for.the postulated failure of these lines.

p 6)

CONCLUSIONS AND SCHEDULE

-Based on preliminary studies and review of the postulated Main Steam and Main Feedwater line ruptures, Duke has confidence that the unit can be shut down safely ar.d maintained in a safe shutdown condition indefinitely 1

with possible minor changes in design.

Standby core cooling is assured

,during the safe shutdown condition.

As discussed with Mr Al Schwencer and Mr Irv Peltier on-December 29, 1972, Duke is performing a detailed review necessary to confirm the above preliminary information and establish possible nee,ds for modifications.

Duke wil.1 contact the AEC on 1-18-73 for another progress report on these Firm commitments for an application amendment and proposed matters.

station modifications'will.be made as appropriate. Unless detailed studies indicate otherwise, changes are expected to be the same for_ all

.three units.

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OCONEE NUCLEAR STATION - UNIT #1 3

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