ML20042E686

LER 90-009-00:on 900316,potential Overpressurization of Auxiliary Feedwater Piping Could Have Occurred During Thermal Expansion of Process Fluid Between Closed Valved. Caused by Design deficiencies.W/900423 Ltr

ML20042E686
Person / Time
Site:	Fort Calhoun
Issue date:	04/23/1990
From:	Gates W, Van Sant B OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-90-009, LER-90-9, LIC-90-0302, LIC-90-302, NUDOCS 9004300272
Download: ML20042E686 (4)

Text

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1 Omaha Public Power District  ;

444 South 16th Street Mall i Omaha, Nebraska 68102-2247  ?

402/636-2000 i April 23, 1990 LIC-90-0302 i i

i U. S. Nuclear Regulatory Commission  !

Attn Document Control Desk Mail Station P1-137  !

Washington, DC 20555 l

Reference:

Docket No. 50-285  ;

1 Gentlemen:

i

Subject:

Licensee Event Report 90-09 for the Fort Calhoun Station  !

t Please find attached Licensee Event Report 90-09 dated April 23, 1990. f This report is being submitted per requirements of 10 CFR  !

50.73(a)(2)(ii)(B). An extension of the submittal date to April 23, 1990  !

was granted by Mr. Les Constable of Region IV. This was based on a f telephone conversation between Mr. Constable and D. J. Matthews of my staff.

v If you should have any questions, please contact me, i i

Sincerely,  ;

8 W. G. Gates i Division Manage i Nuclear Operations i WGG/ tem  !

Attachment I

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c: R. D. Martin, NRC Regional Administrator A. Bournia, NRC Project Manager >

P. H. Harrell, NRC Senior Resident Inspector  ;

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An analysis of the Auxiliary Feedwater System (AFW) )iping between the containment penetration isolation valves has shown t1at, in the event of a Main SteamLineBreak(MSLB)orlossofCoolantAccident(LOCA)insidecontainments the pi)ing would be overpressurized due to thermal expansion of the process fluid )etween the closed valves. The potential pipe failure could result in the inability of the AFW system to provide coolant to the intact steam generator following a MSLB. This is a condition outside the plant design basis.

The corrective actions include implementation of a modification to prevent overpressurization of the AFW line. ASafetyAnalysisforOperability(SAO) will be implemented to justify operation until installation is complete.

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TheFortCalhounStationUnitNo.1UpdatedSafetyAnaly(AFW)Systempipingas designates the design code for the Auxiliary Feedwater UnitedStatesofAmericaStandard(USAS)B31.7. The USAR also defines the AFW  ;

system as Seismic Category I. The design pressure and temperature of the -

piping is 1660 psig at 550 degrees F. There is a separate piping path for AFW .;

! supply to each of two steam generators. Control valves HCV-1107A, HCV-11078, .:

. HCV-1108A, and HCV-1108B are pneumatically operated containment isolation  ;

valves in the AFW lines, and are closed during normal power operation. They areautomaticallyoperatedbytheAuxiliaryFeedwaterActuationSignal(AFAS),

remote-manually operated from the main control room panels or manually .

i operated from the local control panel. The"A"suffixdeslgnatesthatthe valve is inside containment, while the "B" suffix denotes outside containment. t As part of the design basis reconstitution effort currently underway at the Fort Calhoun station, an AFW system design basis document (SOBD-FW-AFW-117) was prepared. Resolution of Open Item No. 119 for this document determined that a 4 potentially reportable concern existed for the AFW piping.

For a postulated Main Steam Line Break (MSLB) inside containment, there would -

be a rapid increase in containment temperature resulting from the break.

Valves HCV-1107A, HCV-11078, HCV-1108A, and HCV-1108B would remain closed until

, the water level in the steam generators reached a low level setpoint. At this point the AFAS would open the valves to the steam generator unaffected by the MSLB. However, the valves would not receive a signal to open in the first 60- '

seconds after a break. Because of the long lengt:1 of uninsulated piping between the containment wall and the inner AFW isolation valves, thermal expansion of the water would cause the pressure in the piping between each ) air  :

of AFW penetration isolation valves to increase ra) idly. The pressure in tie piping would increase to greater than 3000 psi wit 11n the first 60 seconds after a MSLB. This greatly exceeds the 1660 psi design pressure of the piping.

The valves have piston balanced plugs and are designed to remain closed with very high differential pressures across the valves. The differential pressure required to partially lift the plug with pressure under the seat is a) proximately 3900 psi. Therefore, no credit could be taken for relieving of -

tie pressure between the valves due to lifting of the plug in either valve.

I A loss of coolant accident (LOCA) would also result in a similar condition in the piping. However, this postulated accident would result in a lower peak '

containment temperature, and as such would be enveloped by the MSLB scenario.

At 1130 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.29965e-4 months <br /> CST on March 16, 1990, the Auxiliary Feedwater System piping between the containment isolation valves was determined to be outside the plant design basis. Subsequently a "four hour" report was made to the Nuclear Regulatory Commission pursuant to 10CFR 50.72(b)(2)(1). This condition is also reportablepursuantto10CFR50.73(a)(2)(ii).

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h Investigation of the problem revealed that the design deficiency has existed  !

since plant construction. Overpressurization of the piping could have resulted in the failure of the AFW )iping to both steam generators. The consequence of this piping failure would >e the inability to provide coolant to the intact steam generator following a MSLB or LOCA. In this scenario, Emergency Operating Procedure E0P-20, Functional Recovery, would be used for accident mitigation.

The cause of this event is attributed to design and analysis deficiencies by the original plant Architect / Engineer company. The precise root cause can not ,

be determined due to an insufficient amount of informatinn and documentation concerning practices and procedures utilized by this company.

The corrective actions include implementation of a modification to prevent ,

overpressurization of the AFW lines. A Safety Analysis for Operability (SAO) will be implemented to justify operation until the modification is completed.

This event is similar to the events reported in LER 90-03 and LER 90-07 as it describes conditions outside design basis due to design deficiencies by the original A/E. Generic corrective actions noted in those LER's will also apply for this event. LER 89-21 also concerned design deficiencies by contracted companies.

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