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F ACALITY HAM 6 41) PAGE a |DOCatT asuusta 131 CRYSTAL R1VER UNIT 3 0 !5 I o 1o 10 l 3!012 1loFl015 ' " ' " * Failure of RCP "A" Results in Reactor Irip and Emergency Feedwater Initiation IV4NT DATE(Gl Lin huMatt tot alPOAT DATE 011 OTMAR F ACILITIEB Nv0LytD 186

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l I ! I I I I f 1 1 I I I I I _ SUPPLEMENTAL ASPOmf E mPECTED Hel E nPt ctf D 'nft'N|i" f '~~] Tts tier ., u M eAPacreo svowwom oan> D ho I | auf a ACT ,ww, sm - ...... ,,. . P mM.we . .. m--~~a aa- o S j On January 1, 1986, Crystal River Unit 3 was operating at 92*5 reactor ' power while generating 830 MWe, At 2334, a number of alarms relating to the "A" reactor (RCp) were received within a 1.5 second interval. Three seconds after the first alarms were received, the reactor tripped on nuclear overpower based on Reactor Coolant System (RCS) flow and axial power imbalance (flux / delta flux / flow) The motor for "A" RCp continued to run for approximately two minutes until manually secured by the control board operator. The "A" RCp motor indications, along with the rapid Reactor Coolant System flow degradation, are indicative of a separation between the "A" RCp motor and the pump.

Following the expected main turbine automatic trip, the turbine stop | valve closure caused a pressure spike which resulted in a j spurious actuation of the Emergency Feedwater (EFW) System. The cause of the separation between the "A" RCp motor and the pump was a failed reactor coolant pump shaft. The failed reactor coolant pump shaft has been replaced. The cause of the spurious EFW actuation was the rapid spiking of the level transmitters an response to an oscillatory pressure wave phenomenon following turbine stop valve closure FpC has taken corrective action which should prevent the spurious start of EFW

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| EVENT DESCRTOTIOr{ On January 1, 1986, Crystal River Unit 3 was operating at 9 2 $. reactor power while generating 830 MWe. At 2334, a number of alarms relating

| to the "A" reactor coolant pump (RCP) (AB, P) were received. These alarms, received within a 1.5 second interval, included loose parts ( monitoring, RCP-1A motor vibration high, RCP-1A air cooler leakage high, RCP-1A bottom oil level high, reactor coolant loop A flow low and total reactor coolant flow low. Three seconds after the first alarms were received, the reactor tripped on nuclear overpower based on Reactor Coolant System (RCS) flow and axial power imbalance (flux / delta flux / flow) The motor for "A" RCP continued to run for approximately two minutes until manually secured by the control board

operator. Indication of "A" RCP motor current dropped from normal - full load amperage to 309 of that value at the time of the initial event and remained there until the motor was secured. The "A" RCP cmperage indications, along with the rapid RCS flow degradation, are indicative of a separation between the "A" RCP motor and the pump. Following the expected main turbine automatic trip, the turbine stop valve (SB,V) closure caused a steam pressure spike resulting in a spurious low steam generator level indication. The low level indication caused an actuation of the Emergency Feedwater (EFW) System (Bt.). I At 2335 both main condensate (SD, P) tripped on high deaerator tank (SJ, DEA) level. The main condensate pumps were restarted and feedwater continued to be supplied via the Main Feedwater System. Also at 2335, one high pressure injection valve (BQ, INV) was n'anually opened for four and one half minutes to aid in maintaining pressurizer level. Three main steam safety valves (MSSV) (SB, RV) failed to reseat properly. One of these valves shut after being manually lifted several times. The other two valves reseated when main steam pressure was lowered. The failure of these valves to properly reseat had no significant effect on RCS temperature.

At 2338, the control board operator attempted to secure the Emergency Feedwater System. This action was taken because the Main Feedwater System was providing satisfactory steam generator level control and emergency feedwater was not needed. When the control board operator placed the control stations for the emergency feedwater control valves in " hand" in preparation for resetting the EFW actuation, one of the control valves immediately went from fully shut to the fully open position. This valve opening caused emergency feedwater flow to be initiated to the "B" OTSG. Flow continued for approximately 23 ?,econds while the EFW control valve was being shut from the control 3tation. The improper operation of the EFW control valve had no significant effect on RCS temperature.

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The "A" RCp shaft experienced cracking in both axial and circumferen- tj al directions. Fo]Iowing are the observations and conclusions related to the pump evaluation.

| AXIAL CRACKING: Axial cracks were located in the lower 3/4 inches of the thermal barrier region where relatively cool seal injection mixes with reactor coolant. Cyclic thermal stresses were thereby produced and most likely caused shallow axial cracking. The fracture mechanics analysis showed that these cracks have arrested. The pre- dicted crack depths have been confirmed to be conservative by field measurements. This shallow axial cracking is benign and was not a factor in the actual shaft failure.

CIRCUMFERENTIAL CRACKING: The "A" RCp shaft circumferential crack was located in an unused split ring groove directly below the ACME threaded region. Initiation and propagation of the crack was ascribed to the mechanism of fatigue. The major factors explaining the crack initiations in the split ring groove area are high residual stress at the groove and possible mechanical and thermal stresses in the area prior to service. This crack growth came to a stop due to the lack of crack driving force at a certain crack depth. After a long pause, crack growth resumed due to additional loads from broken impeller-to-shaft bolts plus normal bending loads. Crack propagation beyond this point could have occurred from normal operation loads.

PUMP COVERS: The "B", "C", and "D" pump covers experienced fine axial cracking in the thermal barrier region. These cracks were shallow and fracture mechanics analysas shows that they have arrested and will not propagate beyond their current depth. The "A" pump cover was not examined; however, similar cracking is suspected.

BOLTS / PINS: Shaft-to-impeller bolts and drive pins experienced cracking in FCp "A" and RCp "B" Based on the mixed fracture modes, corroded surfaces, and previous work with the alloy and environment, it is probable that most of the bolt and pin cracking was initiated by fatigue and propagated by corrosion fatigue and/or stress corrosion cracking. The cap screws failed in areas of high stress concentra- tion. Increased stresses due to "short" holes may have contributed to the failures. It is possible that bending stresses from nonperpen- dicular holes may have also been a factor.

The cause of the spericus EFW actuation was the rapid spiking of the level transmitters in response to an oscillatory pressure wave phenomenon following turbine stop valve closure.

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CRYSTAL RIVER UNIT 3 - - o|s;ojolo1310|2 8| 6 0 |011 0 13 Of4 0F 0 |5 , m .~. . . .~ ..e ,.- u.., n n The condensate pumps tripped on high deaerator tank level. When the maj n turbine tripped, extraction steam to the feedwater and condensate heaters (including the deaerator) was terminated. This loss of steam caused the pressure in the deaerator to drop below the saturation pressure for water already stored there, resulting in spontaneous boiling of the water which momentarily raised the level in the tank above the condensate pump trip setpoint.

The cause of the failure of the MSSVs to fully reseat until the valve was manually lifted or until steam pressure was manually lowered could not be determined. The cause of the improper operation of the EFW control valve has not been determ;ned. All attempts to reproduce the improper operation were unsuccessful.

ANALYSIS OF EVEfjI

The Reactor protection System (JC) functioned as designed to automatically shutdown the reactor following the reduction in RCS flow.

| The Emergency Feedwater System automatically actuated on a low steam generator level indication although EFW was not required. All RCS temperatures and pressures remained within the normal operating and post-trip windows through use of the Main Feedwater System. The delays in MSSV reseating were minor and are not considered to be a control or safety problem. Therefore, the reactor core was adequately cooled throughout the transient. This event in no way compromised the safety related function of the reactor coolant pump. That function is to act as part of the reactor coolant boundary.

With the exception of the EFW control valve, no safety related equipment failed to perform as required during this event. Therefore, this event has no adverse safety consequences.

CORRECTIVE ACTIONS All RCp shafts have been replaced. One replacement shaft is of the design which failed and is constructed of the same material (Alloy A-286). The second replacement shaft is of a design which does not have the groove located below the ACME threaded region and it is also made of Alloy A-286. The other two shafts are designed without the groove and are made of Alloy A-479 XM-19 (Nitronic 50)

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the impeller-to-shaft bolt material was changed from Alloy A-206 to | Inconel .' X-7 50 HTH. This alloy in widely used for reactor internals ! colting applications and it possesses excellent resistance to j intergranular stress corrosion cracking. In addition to the change in material, the bolts were redesigned with an elliptical radius under the head and a controlled thread root radius.. These changes reduce the peak stress in the bolt thus reducing the possibility of stress corrosion and fatigue. Additionally, .the locking device was changed i from lockwire to locking pins to facilitate the ultrasonic testing j method for verifying preload.

| Drive pin- material 'was changed from Alloy A-286 to A-479 XM-19 (Nitronic 50). This material provides enhanced stress corrosion cracking resistance while providing adequate strength.

| Florida power Corporation is considering a design change to the condensate pump trip circuit to avoid the trip on the short duration deaerator tank level increase following a main turbine trip. ] The MSSV which required manual lifting to reseat was disassembled and

lapped. That . valve and the two MSSVs which required steam pressure ; reduction to fully reseat have been checked for proper setpoints. FPC has installed a modification which provides a 2-second time delay in the EFW actuation circuit. This time delay has been analyzed to be of sufficient duration to prevent the spurious actuation of EFW following a main turbine trip, yet is also short enough not to degrade the EFW actuation response time. Extensive troubleshooting could neither reproduce nor determine the cause of improper operation of the EFW control valve. No further actions are planned.

PREVIOUS SIMILAR EVENTS

The trip, " nuclear overpower based on RCS flow and axial power imbalance", has initiated an automatic reactor trip two times. This is the first actuation due to RCS flow reduction. Spurious low steam generator level indication following a turbine trip has occurred three times. These actuations were reported in LERs 35-20, 85-23, and 85-28.

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December 9, 1987 3F1287-10

U. S. Nuclear Regulatory Costission Attention: Document Control Desk Washington, D. C. 20555 Subject: Crystal River Unit No. 3 Docket No. 50-302 Operating License No. DPR-72 Licensee Event Report 86-001-03 Dear Sir: Enclosed is Licensee Event Report (IER) 86-001-03 which is being submitted in accordance with 10 CFR 50.73. Should there be any questions, please contact this office.

| Sincerely,

. paqucy) E C. Simpson, Director Nuclear Operations Site Support

WLR: mag xc: Dr. J. Nelson Grace Regional Administrator, Region II

Mr. T. F. Stetka Senior Resident Inspector

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I A Florida Progress Company