ML20005D710

Discusses Planned Installation of Third Auxiliary Feedwater (AFW) Pump to Resolve Generic Issue 124, AFW Sys Reliability. Util Intends to Modify AFW in Other Areas to Enhance Overall Sys Reliability.W/One Oversize Figure

ML20005D710
Person / Time
Site:	Fort Calhoun
Issue date:	12/07/1989
From:	Morris K OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
REF-GTECI-124, REF-GTECI-NI, TASK-124, TASK-OR LIC-89-955, NUDOCS 8912140307
Download: ML20005D710 (4)

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Omaha Public Power District 1623 Harney Omaha, Nebraska 68102 2247 402/536-4000 t

December 7, 1989 LIC-89-955 U. S. Nuclear Regulatory Commission

-Attn: Document Control Desk Mail Station F1-137 Washington DC 20555

References:

1.

Docket No. 50-285 2.

Letter OPPD (W. C. Jones) to NRC (T. E. Murley) dated February 17, 1988 (LIC-88-078) 3.

Letter NRC (P.- D. Milano) to OPPD (R. L. Andrews) dated May 9, 1988 4.

Letter OPPD (W. C. Jones) to NRC (R. D. Martin) dated December 9, 1988 (LIC-88-1094)

Gentlemen:

SUBJECT:

Resolution of Generic Issue No.124 Auxiliary Feedwater System Reliability This letter discusses the key elements of Omaha Public Power District's (0 PPD)

. planned installation of a third auxiliary feedwater (AFW) pump to resolve Generic Issue No. 124, auxiliary feedwater system reliability.

In addition, OPPD intends to modify the AFW system in several other areas to enhance the overall system reliability.

OPPD is proceeding with the final design shown in the attached Figure 1.

Features included in the proposed design are:

e A third AFW pump e A third AFW power supply (diesel) e A second AFW water source m Less dependence on local manual actions by operators s Physical separation from the existing AFW pumps e Enhanced ability to respond to a loss of all alternating current power Completion of this modification will increase the level of AFW system reliability.

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U. S. Nuclear Regulatory Commission LIC-89-955 Page 2 Third Auxiliary feedwater Pump Desian The final design of the proposed modification includes an AFW pump which is diesel driven, and thereby diverse from the power supplies of the existing AFW pumps. The fuel oil supply for the diesel will utilize portions of the existing auxiliary boiler fuel oil system.

The pump will be sized to meet the flow and pressure conditions of the existing AFW pumps and can also be used as a start-up pump. The new AFW subsystem will primarily be located in the turbine building and will not be classified as safety related or be seismically qualified. The new AFW pump suction will be taken from the condensate storage tank, thus providing a diverse water supply for the AFW system. The pump will be aligned to the AFW system to allow flow to both steam generators through the main feedwater nozzles.

Remote operation caphbility from the control room will be a design feature of the new AFW pump.

In addition, a capability will be provided to transfer water from the condensate storage tank to the emergency feedwater storage tank.

Remote actuation capability from the control room will be provided for the feedwater isolation valves (HCV-1105, HCV-1106, HCV-1385 and HCV-1386) to reduce the dependency on local operator action.

Containment Isolation Valve Confiouration In Reference 3, the NRC requested a re-evaluation of the configuration of the AFW containment isolation valves.

Specifically:

"The four air-operated valves (HCV-1107A and B, and HCV-Il08A and B) that permit the AFW pumps flow to the steam generators are normally closed and should open upon receipt of an AFWS initiation signal.

Since there is a potential that the closed valves may fail to open on demand, the licensee is requested to evaluate the merits of both the normally opea and the normally closed valve configuration and adopt the more reliable configuration. A high degree of reliability of these discharge valves would alleviate the staff concerns raised in generic issues GI-122.1.a, b, and c, with respect to isolation valve failure, and interruption and recovery of AFW flow."

These valves are required to be in different positions during various accident scenarios.

Also, during events which require operation of the AFW system, the valves are required to open and close on low and high water level respectively to control water level in the steam generators. The valves are required to provide a high reliability of opening on low steam generator level to ensure availability of a reactor core decay heat removal flow path when the AFW system is required to operate.

To satisfy this function, the valves are designed to i

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'LIC-89-955 Page 3 fail open. Additionally, there are system requirements that require the valves be closed during various accident scenarios.

These requirements are:

Main Steam Line Breaks - The AFW containment isolation valves are required to be initially closed to ensure that the AFW flow is delivered only to the intact steam generator to prevent the potential for containment overpressurization or reactor coolant system rapid cooldown.

Main Feedwater Line Breaks - The valves are required to be initially closed to ensure that the AFW flow is delivered only to the intact steam generator to prevent the depletion of the AFW primary water source without providing any decay heat removal.

Loss of Coolant Accidents - The outside containment isolation valves (HCV-11078; HCV-Il088) may be required to be closed post LOCA to maintain containment boundary integrity.

The present valve arrangement (normally closed / fail open) best satisfies all of the above requirements.

The proposed design change includes adding seismic I safety related air accumulators on the four containment isolation valves in question (HCV-Il07A, B, and HCV-Il08A, B).

This will enhance the reliability of the present v:lving arrangement and reduce reliance on operator actions during various accident scenarios.

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The completion of the system modification and acceptance testing is dependent on scheduled material deliveries and their associated construction activities.

Delivery of the pump and driver assembly is currently scheduled for March 1, 1990; this is critical to meet the target completion date. The schedule is to complete outage dependent activities during the 1990 refueling outage. The remaining installation and testing activities shall be completed by July 30, 3

1990 consistent with OPPD's Safety Enhancement Program schedule provided in i

l Reference 4.

These actions address the reliability issues discussed in Reference 3.

If you should have any questions, please contact me.

Sincerely, l

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D vision Manager L

Nuclear Operations l

KJM/pjc Attachment l

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LeBoeuf, Lamb, Leiby & MacRae l

R. D. Martin, NRC Regional Administrator l

A. Bournia, NRC Project Manager P. H. Harrell, NRC Senior Resident Inspector 1

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