05000412/LER-1993-001, :on 930126,identified That Stress on AFW Piping to C SG Exceeded Design Stress Due to Water Hammer.Station Will Perform ISI Insps of AFW Piping & Replace Mechanical Snubbers W/New Hydraulic Snubbers

:on 930126,identified That Stress on AFW Piping to C SG Exceeded Design Stress Due to Water Hammer.Station Will Perform ISI Insps of AFW Piping & Replace Mechanical Snubbers W/New Hydraulic Snubbers

ML20034F226
Person / Time
Site:	Beaver Valley
Issue date:	02/25/1993
From:	Freeland L DUQUESNE LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-93-001, LER-93-1, ND3MNO:3417, NUDOCS 9303020427
Download: ML20034F226 (6)

LER-1993-001, on 930126,identified That Stress on AFW Piping to C SG Exceeded Design Stress Due to Water Hammer.Station Will Perform ISI Insps of AFW Piping & Replace Mechanical Snubbers W/New Hydraulic Snubbers

Event date:
Report date:
4121993001R00 - NRC Website

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'%f Telephone (412) 393-6000 Nuclear Group P.O Box 4 Shippingport FA 15077MM i

February 251993

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ND3MNO:3417 1

i Beaver Valley Power Station, Unit No. 2 l

Docket No. 50-412, Licensee No. NPD-73 LER 93-001-00 l

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United States Nuclear Regulatory Commission l

Document Control Desk Washington, DC 20555 i

Gentlemen:

In accordance with Appendix A, Beaver Valley Technical Specifications, the following Licensee Event Report is submitted:

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i LER 93-001-00,10 CFR 50.73.a.2.ii.B, " Design Stress for the Auxiliary i

Feedwater System Exceeded Due to Water Hammer."

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j Freeland l

General Manager i

Nuclear Operations

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Mr. T. T. Martin, Regional Administrator United States Nuclear Regulatory Commission Region 1 475 Allendale Road i

King of Prussia, PA 19406 Mr. G. E. Edison, BVPS Licensing Project Manager United States nuclear Regulatory Commission Washington, DC 20555 Larry Rossbach, Nuclear Regulatory Commission, BVPS Senior Resident Inspector l

J. A. Holtz, Ohio Edison 76 S. Main Street Akron, OH 44308 l

Larry Beck Centerior Energy 6200 Oak Tree Blvd.

Independence, OH 44101-4661 INPO Records Center Suite 1500 1100 Circle 75 Parkway Atlanta, GA 30339 t

G. E. Muckle, Factory Mutual Engineering 680 Anderson Drive #BLD10 Pittsburgh, PA 15220-2773 Mr. Richard Janati Department of Environmental Resources P.O. Box 2063 16th Floor, Fulton Building Harrisburg, PA 17120 Director, Safety Evaluation & Control Virginia Electric & Power Co.

P.O. Box 26666 One James River Plaza t

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Design Stress Fcr The Auxiliary Feedwater System Exceeded Due To Water Harnmer EVENT DATE (5)

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On 1/26/93 at 1720

hours, the results of extensive engineering analysis of previous plant events identified that the stress on the Auxiliary Feedwater (AFW) piping to the "C" Steam Generator, for combined water hammer and seismic events, exceeded the design stress allowables.

It has been hypothesized that the "C" AFW line has been subjected to water hammer due to steam pocket formation (voiding) and steam bubble collapse.

There have been two events at Beaver Valley Power Station Unit 2 since it went operational in 1987 that have been attributed to water hammer.

The location of the inside containment AFW check valves (inside CV's) and piping layout have the potential to

cause

voiding (see Figure 1).

The inside CV's are located near the main feedwater lines causincJ the piping upstream to be at elevated temperatures.

The inside CV's are also at a

higher elevation

.a n the AFW supply tank minimum water level.

If the inside CV's leak rate is less than the outside CV's, the AFW piping will depressurize causing voiding upstream of the inside CV's due to elevated temperatures.

AFW automatic actuation with unthrottled flow to a

voided AFW line is the main concern for water hammer.

There were minimal safety implications as a result of this event.

The mechanical snubbers were replaced with new hydraulic snubbers with a higher load rating.

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Description of Event

On 1/26/93 at 1720

hours, the results of extensive engineering analysis of previous plant events identified that the stress on the Auxiliary Feedwater (AFW) piping to the "C"

Steam Generator, for combined water hammer and seismic events, exceeded the design stress allowables.

It has been hypothesized that the "C" AFW line has been subjected to water hammer due to steam pocket formation (voiding) and

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steam bubble collapse.

There have been two events at Beaver Valley Power Station Unit 2

since it went operational in 1987 that could be attributed to water hammer.

In event No 1, identified in April 1989, mechanical snubbers 2FWE-PSSP-009 and 2FWE-PSSP-010 on the "C" AFW line were found mechanically frozen durincJ an ISI inspection.

The cause of i

event no.

I was unknown at that time and the snubbers were replaced with identical new snubbers.

Event No.

2, identified in April 1992, t

j mechanical snubber 2FWE-PSSP-009 was found mechanically frozen and l

snubber 2FWE-PSSP-010 Pas found damaged during an ISI inspection on the l

l "C"

AFW line.

The cause for the snubber failure / damage was apparently i

due to excessive compressive or impact loading as a result of an unanticipated water hammer.

The mechanical snubbers were replaced with-new hydraulic snubbers with a higher load rating.

j i

l The location of the inside containment AFW check valves (inside CV's) l and piping layout have the potential to cause steam voiding (see Figure 1).

The inside CV's are located near the main feedwater lines causing l

the piping upstream to be at elevated temperatures (measured 268.3 degree F on top of pipe, 200 degrees F on bottom of pipe, approximately 2

feet upstream of the "C" AFW check valve).

The inside CV's are at a higher elevation than the AFW supply tank minimum water level.

If the I

inside CV's leak rate is less than the outside CV's, the AFW piping will depressurize causing voiding upstream of the inside CV's due to elevated temperatures.

The "C" AFW line has been depressurized.

Cause of Event

e The results of extensive engineering analysis of previous plant events identified that the stress on the Auxiliary Feedwater (AFW) piping to the "C"

Steam Generator, for combined water hammer and seismic events, exceeded the design stress allowables.

The AFW piping and pipe l

supports are classified as QA Category I, seismic, safety class II.

The piping is designed to the requirements of the ASME Boiler and j

Pressure Vessel

Code, Section

III, 1971 Edition including addenda through Wint^r 1972.

The supports are designed to the guidelines established 7

the 6th Edition of the AISC manual of Steel Construction.

The AFW piping and supports have been designed for

thermal, pressure,

deadload, and seismic conditions.

Anticipated or l

" designed" water hammer transients were not previously analytically i

considered.

It was previously assumed that the piping layout and j

components had been designed to preclude the development of postulated I

water hammer events.

In retrospect, there is the potential for l

unanticipated water hammer as a

result of unthrottled AFW flow from j

automatic AFW pump starts to a potentially voided "C" AFW line.

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Corrective Actions

i The following corrective actions have been or will be taken as a result of this event:

1.

Frequent monitoring of the AFW piping for evidence of elevated containment penetration temperatures, system pressures (temporary pressure gages t

installed), noise / water hammer.

2.

A Basis for Continued Operation (BCO) has been prepared to document operation in this condition.

This is based upon analysis to quantify loads resulting from a postulated water hammer event.

3.

The Station is pursuing a long term solution.

4.

The Station will perform ISI inspections of the

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AFW piping and piping supports in containment following an unplanned AFW pump start that results in unthrottled AFW flow.

5.

The mechanical snubbers were replaced with new hydraulic snubbers with a higher load rating.

The new hydraulic snubbers are adequately sized to withstand a water hammer event.

Reportability

i This report is being submitted in accordance with 10CFR50.73.a.2.ii.B because this event resulted in operation outside the Unit 2 design basis.

Safety Implications d

The safety implications are minimal.

Engineering analysis has shown that in cases where the pipe stress levels under the-revised load combinations (to include water hammer) occurring simultaneously with a

seismic event exceeded original design code normal / upset allowables, an assessment of system operability was made to the design code faulted allowables.

The pipe stress

levels, for combined water hammer and seismic

events, are now within the design code faulted allowables.

In developing the BCO, it was determinen that the use of alternate code faulted allowables (i.e. Appendix F or later revisions of the ASME code) was not required due to the rigorous analysis performed to more adequately quantify the stress.

Previous Similar Events

There were no previously reported similar events.

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PDETRATIONS: X-79 X-88 AUX. FEED PUMPS X-83 2FWE-P22 2FWE-P23A 2FWE-P238 NRC Form 3ASA (649)