|
U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
05/01/2015 - 05/04/2015
** EVENT NUMBERS **
|
| Part 21 |
Event Number: 50900 |
Rep Org: CURTISS WRIGHT FLOW CONTROL CO.
Licensee: CURTISS WRIGHT FLOW CONTROL CO.
Region: 1
City: EAST FARMINGDALE State: NY
County:
License #:
Agreement: Y
Docket:
NRC Notified By: JOHN DeBONIS
HQ OPS Officer: STEVE SANDIN |
Notification Date: 03/17/2015
Notification Time: 09:59 [ET]
Event Date: 03/17/2015
Event Time: [EDT]
Last Update Date: 05/01/2015 |
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(a)(2) - INTERIM EVAL OF DEVIATION |
Person (Organization):
GLENN DENTEL (R1DO)
BINOY DESAI (R2DO)
PART 21/50.55 REACT (EMAI) |
Event Text
INTERIM PART 21 REPORT - POTENTIAL TEST INDUCED DEFECT IN A 0867F MAIN STEAM SAFETY RELIEF VALVES
The following report was received from Curtiss - Wright via email:
"This letter provides interim notification of a potential test induced defect in a 0867F Series Main Steam Safety Relief Valves (MS-SRVs) manufactured and supplied by Target Rock (TR). The information required for this notification is provided below:
"(i) Name and address of the individual or individuals informing the Commission.
William Brunet
Director of Quality Assurance
James White
General Manager
Target Rock, Business Unit of Curtiss-Wright Flow Control Corporation
1966E Broadhollow Road
East Farmingdale, NY 11735
"(ii) Identification of the basic component supplied for such facility or such activity within the United States which may fail to comply or contains a potential defect.
Target Rock 0867F Series of Main Steam-Safety Relief Valves Manufactured by Target Rock. This is a 3-stage piloted valve consisting of a main valve (the 'Main') with an actuator mounted to it (the 'Topworks'). The 0867F is the latest generation of the 67F line of MS-SRVs, including the original 3-Stage and 2-Stage designs, and this product line has over 40 years of plant operational experience. Only the 0867F is under investigation. This is due to the differences between the 0867F design and the other designs.
"(iii) Identification of the firm supplying the basic component which fails to comply or contains a defect.
Target Rock, Business Unit of Curtiss-Wright Flow Control Corporation
1966E Broadhollow Road
East Farmingdale, NY 11735
"(iv) Nature of the defect or failure to comply and the safety hazard which is created or could be created by such defect or failure to comply.
As we understand it, the Pilgrim Station recently manually opened the Target Rock Main Steam Safety Relief Valves (MS-SRVs) as part of cooling down the reactor following a loss of offsite power. One of the four installed MS-SRVs may not have fully opened. As-found steam testing of the affected MS-SRV did not duplicate this failure; the valve opened on demand. However, the valve did not re-close as expected. Internal inspections found damaged parts in the main stage subassembly that could potentially affect the ability of the MS-SRV to operate as designed.
We are investigating potential root causes for this damage. However, we are still unable to determine if a specific defect exists. GE SIL-196, Supplement 17 determined Main Spring relaxation was caused by 'extreme dynamics encountered during limited flow testing . Valve dynamics under full flow conditions (i.e. discharge not gagged) are much less severe than those under limited flow conditions.' These extreme dynamics, under limited flow test conditions, are the focus of our investigation. Specific areas of investigation include;
a) Testing of materials to verify they are consistent with our material specifications,
b) evaluation of differences between the 0867F and earlier designs, and
c) evaluation of the differences between different limited flow test loop configurations and test procedures
"(v) The date on which the information of such defect or failure to comply was obtained.
The Pilgrim event occurred on January 27, 2015. As-found testing occurred on February 2, 2015.
"(vi) In the case of a basic component which contains a defect or fails to comply, the number and location of these components in use at, supplied for, being supplied for, or may be supplied for, manufactured, or being manufactured for one or more facilities or activities subject to the regulations in this part.
While we have yet to determine if a specific defect exists, the following plants were supplied 0867F MS-SRVs:
- Pilgrim (Model 09J-001) Quantity Shipped = 8
- Fitzpatrick (Model 09H-001) Quantity Shipped = 4, Quantity on order= 8
- Hatch 1 and 2 (Model 09G-001) Quantity Shipped= 24, Quantity on order= 12
The following plants will be supplied 0867F MS-SRVs:
- Hope Creek (Models 14J-001, 14J-002) Quantity on order = 7
"(vii) The corrective action which has been, is being, or will be taken; the name of the individual or organization responsible for the action; and the length of time that has been or will be taken to complete the action.
The root cause of the potential test induced defect has not yet been confirmed as of the date of this report. Therefore, no specific corrective actions have been initiated. Target Rock Problem Report 080 will document the corrective actions when they are determined and complete the 10 CFR Part 21 evaluation of the potential test induced defect. This determination will be based on further mechanical and material evaluations. TR anticipates completing these evaluations within 45 days; however, in the event the evaluations are not completed, TR will forward another interim report within 45 days.
"(viii) Any advice related to the defect or failure to comply about the facility, activity, or basic component that has been, is being, or will be given to purchasers or licensees.
We are working with all three (4) sites to identify appropriate precautions.
"(ix) In the case of an early site permit, the entities to whom an early site permit was transferred.
Not applicable.
"Should you have any questions regarding this matter, please contact Michael Cinque, Director of Program Management at (631 ) 293-3800."
* * * UPDATE FROM JOHN DeBONIS (VIA EMAIL) TO HOWIE CROUCH AT 1355 EDT ON 5/1/15 * * *
Curtiss-Wright provided an update to state that their root cause analysis is still in progress and they anticipate completion within 60 days.
Notified NRR Part 21 Group (via email), R1DO (Gray), and R2DO (Ehrhardt). |
| Agreement State |
Event Number: 51007 |
Rep Org: TEXAS DEPT OF STATE HEALTH SERVICES
Licensee: RRC POWER AND ENERGY LLC
Region: 4
City: ROUND ROCK State: TX
County:
License #: 06105
Agreement: Y
Docket:
NRC Notified By: ART TUCKER
HQ OPS Officer: JEFF ROTTON |
Notification Date: 04/23/2015
Notification Time: 23:13 [ET]
Event Date: 04/23/2015
Event Time: 20:45 [CDT]
Last Update Date: 04/24/2015 |
Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE |
Person (Organization):
JAMES DRAKE (R4DO)
NMSS_EVENTS_NOTIFICA (EMAI)
ILTAB (EMAI)
MEXICO (EMAI) |
| This material event contains a "Less than Cat 3 " level of radioactive material. |
Event Text
AGREEMENT STATE REPORT - LOST MOISTURE DENSITY GAUGE
The following information was provided by the State of Texas via email:
"On April 23, 2015, at 2117 [CDT], the Agency [Texas Department of State Health Services] was contacted by the licensee's radiation safety officer (RSO) and informed that one of their InstroTek model 3500 moisture/density gauges containing an 8 millicurie cesium - 137 source and a 40 millicurie americium/beryllium source was lost by a technician at a work site [Midland, Texas]. The technician had placed the gauge on the back of his pickup truck. He walked around the job site with the work area supervisor for a few minutes, then returned to his truck and drove to the new work site. When he arrived at the new site he notice the gauge was missing. The technician notified the RSO and local law enforcement. He then began looking for the gauge. The RSO stated he was responding to the location to help look for the gauge. The RSO stated the operating rod for the cesium source was locked in the shielded position. Additional information will be provided as it is received in accordance with SA-300."
Texas Incident #: I-9307
* * * UPDATE ON 4/24/15 AT 1508 EDT FROM ART TUCKER TO DONG PARK * * *
The following information was provided by the State of Texas via email:
"On April 24, 2015, at 0929 CDT the Agency [Texas Department of State Health Services] was notified by the licensee's radiation safety officer (RSO) that the gauge had been recovered. The RSO stated they had been contacted by a contractor who had found the gauge yesterday and taken it to their [location] where they stored it over night. The RSO stated the operating handle for the cesium source was still locked. The gauge did not appear to be damaged. The RSO stated a leak test was performed of the gauge sources and the gauge has been placed in storage until the results of the leak test are received. Additional information will be provided as it is received in accordance with SA-300."
Notified R4DO (Drake), NMSS Events Notifications, ILTAB, and Mexico via email.
THIS MATERIAL EVENT CONTAINS A "LESS THAN CAT 3" LEVEL OF RADIOACTIVE MATERIAL
Sources that are "Less than IAEA Category 3 sources," are either sources that are very unlikely to cause permanent injury to individuals or contain a very small amount of radioactive material that would not cause any permanent injury. Some of these sources, such as moisture density gauges or thickness gauges that are Category 4, the amount of unshielded radioactive material, if not safely managed or securely protected, could possibly - although it is unlikely - temporarily injure someone who handled it or were otherwise in contact with it, or who were close to it for a period of many weeks. For additional information go to http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf |
| Agreement State |
Event Number: 51009 |
Rep Org: MA RADIATION CONTROL PROGRAM
Licensee: BAYSTATE HEALTH
Region: 1
City: SPRINGFIELD State: MA
County:
License #: 60-0095
Agreement: Y
Docket:
NRC Notified By: JOSHUA DAEHLER
HQ OPS Officer: VINCE KLCO |
Notification Date: 04/24/2015
Notification Time: 09:22 [ET]
Event Date: 04/22/2015
Event Time: [EDT]
Last Update Date: 04/24/2015 |
Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE |
Person (Organization):
JAMES NOGGLE (R1DO)
NMSS EVENTS NOTIFICA (EMAI) |
Event Text
AGREEMENT STATE REPORT - BULK DOSE ADMINISTERED TO PATIENT
The following information was received from the Commonwealth of Massachusetts via email:
"The licensee's Radiation Safety Officer (RSO) reported on April 23, 2015 that, on the morning of April 22, 2015, the licensee mistakenly administered to a patient the wrong radioactive drug, a 118 mCi Tc-99m bulk dose instead of the prescribed 12.9 mCi Tc-99m Sestamibi dose, at the licensee's Baystate Franklin Medical Center facility.
"The wrong radioactive drug administered was reported by the licensee's RSO to have resulted in 5.6 rem effective dose equivalent to the patient, a reportable medical event in accordance with 105 CMR 120.594(A)(1)(b)1.
"The licensee's RSO reported that the patient and the referring physician have been notified and that the RSO did not expect any harm to the patient.
"The RSO reported the cause included that proper procedures were not followed.
"The Agency (Massachusetts Radiation Control Program) plans to perform a special inspection and considers this event to be open."
A Medical Event may indicate potential problems in a medical facility's use of radioactive materials. It does not necessarily result in harm to the patient. |
| Power Reactor |
Event Number: 51028 |
Facility: RIVER BEND
Region: 4 State: LA
Unit: [1] [ ] [ ]
RX Type: [1] GE-6
NRC Notified By: ROBERT MELTON
HQ OPS Officer: DONALD NORWOOD |
Notification Date: 05/01/2015
Notification Time: 08:36 [ET]
Event Date: 04/30/2015
Event Time: 23:44 [CDT]
Last Update Date: 05/01/2015 |
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION |
Person (Organization):
GEOFFREY MILLER (R4DO) |
| Unit |
SCRAM Code |
RX CRIT |
Initial PWR |
Initial RX Mode |
Current PWR |
Current RX Mode |
| 1 |
N |
Y |
100 |
Power Operation |
100 |
Power Operation |
Event Text
HIGH PRESSURE CORE SPRAY SYSTEM INOPERABLE
"River Bend Station personnel declared the High Pressure Core Spray (HPCS) system inoperable at 2344 CDT on 4/30/2015.
"The HPCS system at River Bend Station includes a test return line to the Condensate Storage Tank (CST). The test return line is isolated by two motor operated valves (E22-MOVF010 and E22-MOVF011), with both having a safety function to close on an ECCS initiation signal to ensure that injection flow is directed to the reactor vessel. There is currently a blind flange installed downstream of these two valves. While the HPCS pump is normally aligned to the CST, the credited source of water for the pump is the suppression pool. Accordingly, the pump suction is realigned to the suppression pool on low level in the CST or when suppression pool level rises to a certain point. While performing maintenance on the downstream test return valve (E22-MOVF011), station personnel identified leakage past the upstream test return valve (E22-MOVF010) which was being used as an isolation boundary. In evaluating this condition, engineering personnel noted that the observed leakage past the upstream isolation MOV might be sufficient to deplete suppression pool inventory such that it would not be capable of performing its specified function for the duration of the 30-day mission time. The issue of concern is that once HPCS is aligned to the suppression pool post-LOCA, pool inventory would be lost due to the leaking upstream isolation valve (E22-MOVF010) and out the disassembled downstream isolation valve (E22-MOVF011).
"Based on that concern, the HPCS pump suction valve from the suppression pool was disabled in the closed position to preserve pool inventory. This action caused the HPCS system to be declared inoperable at 2344 CDT. This action results in a 14 day shutdown LCO and is reportable to the NRC in accordance with 10CFR50.72(b)(3)(v)D.
"The HPCS pump remained available with its suction aligned to the CST.
"Message has been left with NRC Senior Resident Inspector." |
| Power Reactor |
Event Number: 51029 |
Facility: ARKANSAS NUCLEAR
Region: 4 State: AR
Unit: [1] [2] [ ]
RX Type: [1] B&W-L-LP,[2] CE
NRC Notified By: THERON ROWBOTHAM
HQ OPS Officer: DANIEL MILLS |
Notification Date: 05/01/2015
Notification Time: 16:14 [ET]
Event Date: 05/01/2015
Event Time: 16:00 [CDT]
Last Update Date: 05/01/2015 |
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE |
Person (Organization):
GEOFFREY MILLER (R4DO) |
| Unit |
SCRAM Code |
RX CRIT |
Initial PWR |
Initial RX Mode |
Current PWR |
Current RX Mode |
| 1 |
N |
Y |
100 |
Power Operation |
100 |
Power Operation |
| 2 |
N |
Y |
100 |
Power Operation |
100 |
Power Operation |
Event Text
LOSS OF SEISMIC MONITORING DUE TO PLANNED MAINTENANCE
"This notification is conservatively being made in accordance with 10 CFR 50.72(b)(3)(xiii) as an event that will result in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g. significant portion of control room indication, Emergency Notification System or offsite notification system.)
"The emergency preparedness plan requires seismic monitoring instruments to diagnose an earthquake for emergency action levels (EAL) HU6 (Natural or destructive phenomena affecting PROTECTED AREA) and HA6 (Natural and destructive phenomena affecting VITAL AREAS).
"At approximately 1600 [CDT] on May 1, 2015, ANO plans to remove Motor Control Center B33 from service for maintenance. This will render the alarm functions for the seismic monitors nonfunctional. It is expected that this maintenance will take approximately 72 hours to complete. ANO procedures provide compensatory measures of using offsite sources to obtain seismic data."
The NRC Resident Inspector has been notified. |
| Part 21 |
Event Number: 51030 |
Rep Org: AZZ/NLI NUCLEAR LOGISTICS, INC
Licensee: ALLEN BRADLEY
Region: 4
City: Fort Worth State: TX
County:
License #:
Agreement: Y
Docket:
NRC Notified By: TRACY BOLT
HQ OPS Officer: DANIEL MILLS |
Notification Date: 05/01/2015
Notification Time: 13:32 [ET]
Event Date: 04/30/2015
Event Time: [CDT]
Last Update Date: 05/01/2015 |
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(d)(3)(i) - DEFECTS AND NONCOMPLIANCE |
Person (Organization):
MEL GRAY (R1DO)
FRANK EHRHARDT (R2DO)
ROBERT ORLIKOWSKI (R3DO)
GEOFFREY MILLER (R4DO)
PART 21/50.55 REACT (EMAI) |
Event Text
POTENTIALLY UNQUALIFIED COMPONENT IN CERTAIN ALLEN BRADLEY TIMING RELAYS
The following is an excerpt from a document received from the licensee via email:
"Report of potential 10 CFR Part 21, Allen Bradley Timing Relay Model 700RTC
"Pursuant to 10 CFR 21.21(d)(3)(ii), AZZ/NLI is providing written notification of the identification of a potential failure to comply.
"On the basis of our evaluation, it is determined that AZZ/NLI does not have sufficient information to determine if the subject condition would, or has, created a Substantial Safety Hazard or would have created a Technical Specification Safety Limit violation as it relates to the subject plant applications.
"The specific part which fails to comply or contains a defect:
"As of 2009-2010, Allen Bradley relays base model 700RTC, contain an unevaluated CPLD (Complex Programmable Logic Device). This was an unpublished design change that was implemented to replace an obsolete integrated circuit chip. The undocumented design change did not result in a part number change from Allen-Bradley. There was no change to the appearance of the relay that would identify any design changes were made to the relay configuration. Therefore, NLI qualification/dedication of the relays after 2009 have not included additional testing for the new CPLD component.
"The timing relay model 700RTC has been dedicated/qualified for multiple applications for various plants.
"Between 2009-2010 Allen Bradley made a design change without changing the part number of the commercial relay or providing any documented evidence of a design change. The manufacturer specification data sheets maintain the classification that the relays are 'solid state', which would imply that there are no digital devices installed in the relay. However, after inspection of the internals of the timing relay (Figure 2), it has been identified that the unit does contain a CPLD which meets the definition of a digital device under the guidance of NEI 01-01."
Potentially affected plants include Browns Ferry, Ginna, Millstone, Nine Mile Point, North Anna, Ft. Calhoun, Perry, River Bend, South Texas Project, and St. Lucie. |
| Power Reactor |
Event Number: 51033 |
Facility: PALISADES
Region: 3 State: MI
Unit: [1] [ ] [ ]
RX Type: [1] CE
NRC Notified By: JC RANEY
HQ OPS Officer: DANIEL MILLS |
Notification Date: 05/02/2015
Notification Time: 12:58 [ET]
Event Date: 05/02/2015
Event Time: 12:23 [EDT]
Last Update Date: 05/02/2015 |
Emergency Class: UNUSUAL EVENT
10 CFR Section:
50.72(a) (1) (i) - EMERGENCY DECLARED |
Person (Organization):
ROBERT ORLIKOWSKI (R3DO)
BERNARD STAPLETON (IRD)
SCOTT MORRIS (NRR)
BILL DEAN (NRR)
CINDY PEDERSON (R3RA) |
| Unit |
SCRAM Code |
RX CRIT |
Initial PWR |
Initial RX Mode |
Current PWR |
Current RX Mode |
| 1 |
N |
Y |
100 |
Power Operation |
100 |
Power Operation |
Event Text
UNUSUAL EVENT DECLARED DUE TO SEISMIC ACTIVITY FELT ON SITE
At 1241 EDT, Operations staff at Palisades declared an Unusual Event under EAL HU1.1 due to seismic activity felt on site. No seismic alarms were initiated. No plant equipment was affected. The epicenter of the 4.2 magnitude earthquake was located south of Galesburg, MI. Palisades continues to operate at 100% power.
The licensee has notified the NRC Resident Inspector.
* * * UPDATE FROM JC RANEY TO DANIEL MILLS AT 1601 EDT ON 5/2/15 * * *
The licensee terminated the Unusual Event at 1541 EDT on 5/2/15.
The licensee has notified the NRC Resident Inspector and the state and local government.
Notified R3DO (Orlikowski), IRD MOC (Stapleton), NRR EO (Morris), NRR ET (Dean), and R3RA (Pederson).
Notified other Federal Agencies (DHS SWO, FEMA Ops, FEMA NWC, NICC Watch Officer and NuclearSSA). |
| Power Reactor |
Event Number: 51034 |
Facility: MONTICELLO
Region: 3 State: MN
Unit: [1] [ ] [ ]
RX Type: [1] GE-3
NRC Notified By: LT. DANA ANTON
HQ OPS Officer: HOWIE CROUCH |
Notification Date: 05/02/2015
Notification Time: 20:57 [ET]
Event Date: 05/02/2015
Event Time: 12:47 [CDT]
Last Update Date: 05/02/2015 |
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(B) - POT RHR INOP |
Person (Organization):
ROBERT ORLIKOWSKI (R3DO) |
| Unit |
SCRAM Code |
RX CRIT |
Initial PWR |
Initial RX Mode |
Current PWR |
Current RX Mode |
| 1 |
N |
N |
0 |
Refueling |
0 |
Refueling |
Event Text
LOSS OF SHUTDOWN COOLING DUE TO LOSS OF 4KV BUS
"On 5/2/2015 at 1247 CDT, while the plant was in MODE 5 for a refueling outage with the vessel cavity flooded, during logic testing of the non-credited essential 4kV bus, MNGP [Monticello Nuclear Generating Plant] experienced a human performance error that caused a loss of the 4kV bus and essential load center. Loss of the load center de-energized valve position indication on one shutdown cooling isolation valve. This tripped the shutdown cooling pump on a pump suction interlock, resulting in the loss of shutdown cooling. This is being reported under 10 CFR 50.72(b)(3)(v)(B) as an event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to remove residual heat.
"Following the event, the safety related load centers were cross-tied to allow restoration of shutdown cooling. Shutdown cooling was restored at 1603. While shutdown cooling was out of service, vessel water temperature rose approximately 10øF to 91.1øF, which is within the allowed temperature band.
"This event did not result in the release of any radioactive material and did not challenge the health and safety of the public.
"The NRC Resident Inspector has been notified." |
| Power Reactor |
Event Number: 51035 |
Facility: MILLSTONE
Region: 1 State: CT
Unit: [ ] [ ] [3]
RX Type: [1] GE-3,[2] CE,[3] W-4-LP
NRC Notified By: MICHAEL FRECHETTE
HQ OPS Officer: VINCE KLCO |
Notification Date: 05/03/2015
Notification Time: 00:05 [ET]
Event Date: 05/02/2015
Event Time: 21:30 [EDT]
Last Update Date: 05/03/2015 |
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE |
Person (Organization):
MEL GRAY (R1DO) |
| Unit |
SCRAM Code |
RX CRIT |
Initial PWR |
Initial RX Mode |
Current PWR |
Current RX Mode |
| 3 |
N |
Y |
100 |
Power Operation |
100 |
Power Operation |
Event Text
NORMAL AND HI RANGE VENTILATION PROCESS RADIATION MONITORS OUT OF SERVICE
"Loss of assessment capability due to unplanned removal from service of a radiation monitor due to process flow monitor indication failing hi. The normal and hi range ventilation vent process radiation monitors (3HVR*RE10A/B are out of service. This condition was discovered during control room rounds. The condition is reportable per 10CFR50.72(b)(3)(xiii)."
Compensatory measures are in place.
The licensee notified the NRC Resident Inspector and applicable State and Local authorities. |
| Power Reactor |
Event Number: 51036 |
Facility: WOLF CREEK
Region: 4 State: KS
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP
NRC Notified By: LARRY HAUTH
HQ OPS Officer: DANIEL MILLS |
Notification Date: 05/03/2015
Notification Time: 12:56 [ET]
Event Date: 05/03/2015
Event Time: 10:22 [CDT]
Last Update Date: 05/03/2015 |
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION |
Person (Organization):
GEOFFREY MILLER (R4DO) |
| Unit |
SCRAM Code |
RX CRIT |
Initial PWR |
Initial RX Mode |
Current PWR |
Current RX Mode |
| 1 |
M/R |
Y |
22 |
Power Operation |
0 |
Hot Standby |
Event Text
FEEDWATER ISOLATION, TURBINE TRIP AND MANUAL REACTOR TRIP DURING POWER ASCENSION
"On 5/3/2015 during power ascension following Refueling Outage 20, Steam Generator 'C' water level increased rapidly, causing a Feedwater isolation on high Steam Generator water level and an associated Turbine trip. The reactor was subsequently manually tripped.
"At the start of the event, reactor power was approximately 22%. Plant staff was in the process of transferring from Main Feedwater Bypass Feed Regulating Valve control, used for low power control, to Main Feedwater Regulating Valve control as part of power ascension. When the Main Feedwater Regulating Valve for 'C' Steam Generator (AEFCV-530) was opened, it went to about 80% open, causing an overfeed of the 'C' Steam Generator. High Steam Generator water level in 'C' Steam Generator initiated an automatic Feedwater Isolation Signal, automatic Turbine Trip and automatic trip of the operating main feed pump. The operating crew initiated a manual reactor trip.
"The Auxiliary Feedwater System automatically initiated as part of the plant response to the feedwater system transient.
"The plant is presently stable in Mode 3. All equipment functioned normally, except the 'C' Main Feedwater Regulating Valve (AEFCV0530) which did not function to properly control Steam Generator level. This valve did function as designed to close on the Feedwater Isolation Signal.
"NRC Resident Inspector has been contacted." |
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