ML19317F280
| ML19317F280 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 12/29/1972 |
| From: | Thies A DUKE POWER CO. |
| To: | Anthony Giambusso US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML19317F262 | List: |
| References | |
| NUDOCS 8001090613 | |
| Download: ML19317F280 (5) | |
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December 29, 1972
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Mr. Angelo Giambusso Deputy Director for Reactor Projects Directorate of Licensing U. S. Atomic Energy Cct=ission Washington, D. C.
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Re: Docket Nos. 50-269, -270, and -287
Dear Mr. Giambusso:
Picase refer to your letter of December 15, 1972 requesting additional analysis and other relevant information needed to determine the con-sequences of postulated failure of the main steam and feedwater pipes outside of the Reactor Building for Oconce Nuc1 car Station Units 1, 2, and 3.
Our response to the request for a description of the analysis to be cade, proposed' modifications, and schedule to be followed are given by the attached surr.ary of the current status of Design Engineering review.
V'ery.truly yours, 34fS s
A. C. Thies ACT:vr.
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DUKE POWEl. COMPANY OCONEE NUCLE.AR STATION I
CONSEQUENCES OF MAIN STEAM c FEEDWATER PIPING RUPTURE CURRENT STATUS OF DESIGN ENGINEERING REVIEW
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12-29-22 1)
INTRODUCTION in response to the AEC/ DOL's 12-15-72 letter and attached guidelines on the consequences of postulated Mair. Steam and Fecdwater piping failures in struc-tures other than the Reactor Building, Duke's current position is as outlined herein.
2)
PROBABILITY OF FAILURE Although Duke 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:
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 h
and Main Feedwater Systems at Oconee.
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Main Steam lines are 100 percent cold pulled so that at the line heats up, all thermal expansion stresses are essentially elimiaated throughout the system.
For exampic, at the Reactor Builoing penetration (terminal cnd), 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 allowable stress values indicates a tremendous amcunt 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 normal. operation at the Reactor Build,ing penetration (terminal end). Again, this is only about 16 perc'ent of the ANSI B31';1.0 (1967) Code allowable stress, 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 Pressurc,
Percent Pressure Capability Margin Main Steam:
910 psig 1093 psig 20 Feedwater:
1070 psig 1383 psig 29 s
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The safety related portions of...these systems are Duke Class F Indicatl.ng that the mater!.als of ' construction were procured z
fabricat,cd..,.t.csted and documen,t.cd, similar..to a nuclear system as can be denoted from the following:3-p I) P i p ing Ma t e r ia l s----------Tracea b l e
,,,yli),WeldingFillerMetal------Traceable 111) NDE ----------------------100 percent X-ray
. a Iv). Piping Materials QA-------inspection at fabricators plant and
._s.ite receiving inspection:.
.;c.
.v), Documentation ---------,--Required A
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Support Design QA ------ '-As outlined in FSAR 1C.3.4.5 3)
POSTULATED RUPTURES REVIEWED TO DATE J
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 bcIng defined; however;* preliminary studies indicate that additional postulated break points.wlll have very little effect on the ability to shut the unit down and maintain it in the safe shutdown condition.
{].,4)
CONSEQUENCES OF POSTULA*.~9 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 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, Jct 1mpingement 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 probicm in the penetration room.
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.
Fecdwater for secondary side cooling is assured to the unaffected Steam Generator e
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 Reacto.- Building isolation check valve, both cases are being analyzed.
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Control Room integrity.
Reviewf the general arrangement of h.igh o
' energy systems relative to the'c'ontrol room' indicates"that pipe whip,
.' Jet' impingement and reacti6n'.' forces ~would not affect the integri ty' of
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~ the ' con trol' room'.'
A' structural reinforced concrete wall is located between the control room and t.he penetration room.
5)
POSSIBLE MODIFICATIONS TO THE' STAT'10N' At present', possible modifications t'o' the station to reduce the"effect's of
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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 P0-222, 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 t'his line. The addition of a new wall to remove the main steam line from the penetra-
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tion room environment may be added along Column Line 65 as shown on r
Sketch PO-222.
b.
East and Vest 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 be shut down safely and maintained in a safe shutdown condition indefinitely with possible mincr 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 will contact the AEC on 1-18-73 for another progress report on these matters.
Firm commitments 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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