Event Notification Report for August 16, 2002

STEAM GENERATOR BLOWDOWN VALVE 1MS-5959 MAY NOT CLOSE AGAINST FULL STEAM GENERATOR PRESSURE.


"A draft calculation (01109-C-015) and a review of this calculation concluded that 1MS-5959 would not shut against the maximum differential pressure (dp) assumed within the calculation. The actual dp used for the calculation review is based on the SG atmospheric dump valve set point of 1085 psig. This is the SG pressure assumed for aux feed flow requirements to the SGs. At this dp (1085 psig), 1MS-5959 blow down isolation valve would not shut (1MS-5959 HX-1A SG-Blow down isolation ). All other (3) steam generator blowdown valves would shut against this d/p.

"The blowdown isolation valves are required to isolate to support two functions; 1) containment isolation and , 2) steam generator pressure boundary isolation. The containment isolation function is a class 4 containment penetration (FSAR page 5.2-3), which is a normally operating line connected to a closed system inside of containment, provided with at least one manual valve located outside of containment, and missile protected throughout its length. FSAR Figure 5.2-51-2 lists manual valve 1MS-265 as the containment isolation valve, but also lists 1MS-5959 as an isolation valve inside of containment. Although 1MS-5959 may not be able to close at full SG pressure, it is expected to be able to close at the maximum containment design pressure (60 psig), which may occur if a SG is faulted. Since the requirement for class 4 penetration is still met by using 1MS-265, and 1MS-5959 would still function under the conditions of a faulted SG, the containment isolation function of 1MS-5959 is considered to be operable.

"The function to establish the SG pressure boundary is based on the requirements of the Chapter 14 accident analysis for the Loss of Normal Feedwater (LONF, Section 14.1.10) and the Loss of All AC Power to the Station Auxiliaries (LOAC, Section 14.1.11). The accident analyses credit 200 gpm AFW flow delivered to the SGs after a five minute delay. The acceptance criteria for both accidents is that the pressurizer does not overfill. Overfilling of the pressurizer could result in a small break LOCA due to the assumed failure of a pressurizer safety valve or PORV when passing liquid. If the SG blowdown valve cannot perform its isolation, then an undetermined (although significant) portion of the AFW will be diverted from flowing around the u-tube region. Heat removal assumed in the analysis will not meet the minimum credited in the accident analysis based on a full 200 gpm AFW flow.

"1MS-5959 was shut based on engineering judgement and declared inoperable at 1415, 08/16/02. The valve operator on 1MS-5959 was adjusted per routine maintenance procedure (RMP) 9141, Air operated valve testing and adjustment, such that the valve would shut under the analyzed conditions. 1MS-5959 was declared operable at 1925 08/16/02."


The NRC Resident Inspector was notified of this event by the licensee.

**UPDATE CHUCK KRAUSE TO MIKE NORRIS 1409 EDT 8/19/02**

Update to make corrections to initial report. Referenced FSAR figure 5.2-51-2 should be FSAR figure 5.2-50-1, and manual valve 1MS-265 should be 1MS-266.

Notified R3DO (Lanksbury).

INVALID SYSTEM ACTUATION DUE TO LOSS OF RPS BUS "A"

"On August 16, 2002 at approximately 0254 hours, a loss of Reactor Protection System (RPS) Bus A occurred while operators were attempting to reset a half-scram on the A2 logic channel of the RPS. The A2 logic reset attempt was required in accordance with a surveillance test procedure that was in progress, PPM OSP-RPS-Q402, MSIV Closure Scram Functional. Upon loss of RPS Bus A, control room operators entered Abnormal Procedure ABN-RPS, Loss of RPS, and an operator was dispatched to investigate the problem.

"The operator investigating the problem discovered that the Motor Generator Set (RPS-MG-A) that supplies RPS Bus A was running with both of its electrical protection assemblies (EPAs) in the tripped position. In accordance with procedure ABN-RPS, operators transferred RPS Bus A to its alternate power supply and then reset the half-scram and associated containment outboard isolations. The operators then exited PPM OSP-RPS-Q402. The follow-up investigation found that the loss of the RPS Bus A was due to a trip of the RPS-MG-A output breaker (RPS-MG-CB1) on over-voltage. The output breaker trip caused both downstream Electrical Protection Assemblies (RPS-EPA-3A and RPS-EPA-3C) to trip. This is the expected and required response for RPS-EPA-3A and RPS-EPA-3C."

The loss of power to RPS Bus A had the following effects on the plant: An actuation of RPS trip system A (Half-Scram), Main Steam Line Radiation Monitors A and C lost power; a trip of the Balance of Plant isolation logic in control room relay Cabinet RC-1; an outboard primary containment isolation signal was received by the following (valves closed or were already closed): Main Steam Line Drain Valves, Reactor Water Sample Valve, Drywell Floor and Equipment Drain Isolation Valves, Residual Heat Removal (RHR) Discharge to Radwaste and Heat Exchanger Sample Isolation Valves, RHR Shutdown Cooling (SDC) Reactor Head Spray Isolation Valve, SDC Return Valves, Reactor Water Cleanup (RWCU) Suction Valve. All of these plant equipment responses were as expected and per plant design.

The licensee notified the NRC Resident Inspector.