ML19329A656

Consequences of Main Steam & Feedwater Piping Rupture, Current Status of Design Engineering Review.

ML19329A656
Person / Time
Site:	Oconee
Issue date:	12/19/1972
From:	DUKE POWER CO.
To:
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ML19329A654	List: ML19329A653 ML19329A656
References
NUDOCS 8001090641
Download: ML19329A656 (4)

Text

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DUKE POWEi. COMPANY OCONEE NUCLEAR STATION CONSEQUENCES OF MAIN STEAM & FEEDWATER PIPING RUPTURE CURRENT STATUS OF DESIGN ENGINEERING REVIEW .

12-29-72 ,

1) INTRODUCTION in response to the AEC/ DOL's 12-15-72 letter and attached guidelines on the consequences of postulated Main, Steam and Feedwater piping failures in struc-tures other than the Reactor Building, Duke's current position is as outlined herein.

2) PROBASILITy 0F FAILURE .

Although Duke is continuing to review the consequences of postulated pipe I ruptures,' such ruptures are not considered credible for the Oconee Nuclear

! Station based on the following:

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

~" ruptures based on conservative engineering practices.

b. 'The only basis for postulating a line rupture is stress criteria. The

. following describes representative stress conditions for the Main Steam >

and Main Feedwater Systems at Oconee.

Main Steam lines are 100 percent cold pulled so that as the line heats up, all thermal expansion stresses are essentially eliminated throughout the system. For example, at the Reactor Building penetration (terminal end), there is only 1100 psi maximum thermal stress du-ing 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 allowable stress values indicates a tremendous amount of conserva-

• tism. The Main Feedwater System is not cold pulled since it operates at a lower temperature; however, similar to the Main Steam, there is only

- 3645 psi maximum thermal stress during norma ( operation at the Reactor Building penetration (terminal end). Agairt, this is only about 16 percent of the ANSI B3-1.1.0 (1967) Code allowable stress.

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

Main Steam 910 psig 1093 psig 20 Feedwater:- 1070 psig 1383 psig 29 c

9 01 0 90[

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d. The safety related portions of these systems are Duke class F Indicating that the materials of construction were procured fabricated, tested and documented similar to a nuclear system as can be denoted from the following: .

1) Piping Materials----------Traceable

- - II) Velding Filler Metal------Traceable Ill) NDE ----------------------100 percent X-ray Iv) Piping ' Materials QA-------inspection at. fabricators plant and site receiving inspection .

v) Documentation ------------Required . . .,

vi) Support Design QA --------As outlined in FSAR IC.3 4.5

3) POSTULATED RUPTURES REVIEWED TO DATE By definition of the attached guidelines to the 12-15-72 AEC letter, Duke has reviewed double-ended ruptures of the two Main Steam and two Main

' Feedwater lines at the terminal ends of the Reactor Building penetration anchors. Other postulated rupture points are currently being defined; -

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~ however; preliminary studies indicate that additional postulated break poidts will have very little effect on the ability to shut the unit down and. maintain it in the safe shutdown condition.

4) CONSEQUENCES OF POSTULATED RUPTURES

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 dowr.

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-However, pressure effects tain the reactor in a safe shutdown condition.

and . steam concentrations might possibly pode a problem in the penetration T.oom. ,

c. East and West Main Feedwater Lines. These lines enter the Reactor .

Building through the East Penetration 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 br.ckup Auxiliary Service Water System. P. essure ef fects andSince steam concentra-the

-tions could pose a problem in the penetration room.

postulated failure can occur on either the upstream or the downstream side of the Reactor Building isolation check valve, both cases are being analyzed.

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d. Control Room Intcarity. Review of the general arrangement of high energy systems relative to the control room Indicates that pipe whip, Jet Impingement and reaction forces would not affect the integrity of the control room. A structural reinforced concrete wall is located between the control room and the penetr& tion room.-

5) POSSIBLE MODIFICATIONS TO THE STATION At present, possible modifications to the station to reduce the effects of pressure and steam concentrations as described in 4)b. and 4)c. to accept-able limits are being analyzed.

a. East Steam Generator Main Steam Line. As shown on the attached Sketch PO-222, the existing north penetration room wall may be modified to include low pressure blowout panels to relieve pressure and to provide a steam escape route for the postulated failure of this line. The additinn of a new wall to remove the main steam line from the penetra-q' tion room environment may be added along Column Line 65 as shown on I

Sketch PO-222.

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b. East and West 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.

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 oe _

shut down safely and maintained in a safe shutdown condition indefinitely 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.needs for modifications.

Duke,wl,11 contact the AEC on 1-18-73 for another progress report on these matters.

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Firm commitmedts for an application amendment and proposed 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 n n u 4,[ %*tj Ia UNIT #2 & 3 SIMILAR

, . POS MODIFICATIONS TO PENETRATION ROCM p,. j- -

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h 12-29 /2 PO-212

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