Semantic search
| Entered date | Event description | |
|---|---|---|
| ENS 40714 | 29 April 2004 01:40:00 | During post maintenance testing following a High Pressure Coolant Injection (HPCI) System outage, the HPCI System was not declared operable due to unstable operation - oscillations of turbine speed (300-400 RPM), pump flow (600 GPM) and discharge pressure (300-550 psig) were seen in both automatic and manual flow control during the System Operability Periodic Test. HPCI had been declared inoperable at 0410 on 4/28/04 and placed under clearance to support planned maintenance on the Flow Controller, Flow Transmitter, system valves and condensate pump. The cause of the unstable operation is currently being investigated. The licensee notified the NRC Resident Inspector.
On April 28, 2004, at 0410 hours, the High Pressure Coolant Injection (HPCI) system was declared inoperable to support scheduled maintenance on the HPCI system. To satisfy post maintenance test requirements and support restoring the HPCI system to an operable status, surveillance test, OPT-09.2, "HPCI System Operability Test," was performed. During this testing at 2230 hours, oscillations in pump flow, pressure, and turbine speed were observed in both the automatic and manual flow control modes of operation. Based on the test results, HPCI remained inoperable until the cause of the oscillations could be identified, corrective actions implemented, and the system satisfactorily tested. On April 29, 2004, at 0140 hours, the NRC was conservatively notified (Event Number 40714), in accordance with 10 CPR 50.72(b)(3)(v)(D), of a condition that at the time of discovery could have prevented the fulfillment of the HPCI safety function. Troubleshooting determined that the HPCI flow controller to the 2-E41-C002-CNV Ramp Generator Signal Converter (RGSC) was subject to spurious deviations. The RGSC was removed from its installed position and bench tested. Circuit review and testing determined that electronic component degradation was the most likely cause of the RGSC output signal perturbations. A new RGSC was calibrated and installed. After installation, in place testing showed that the new ramp generator did not exhibit signal variations. A final HPCI system operability test was performed and verified that the HPCI system was responding normally. On May 1, 2004, at 1220 hours, the HPC1 system was restored to service. Reportability Discussion: NUREG-1022, Rev. 2, Section 3.2.7 (page 56) lists types of events or conditions that are generally not reportable under 10 CFR 50.72(b)(3)(v) and 10 CFR 50.73(a)(2)(v) criteria. The list of not-reportable conditions includes: Removal of a system or part of a system from service as part of a planned evolution for maintenance or surveillance testing when done in accordance with an approved procedure and the plant's TS (unless a condition is discovered that could have prevented the system from performing its function). On April 28, 2004, the HPCI system was removed from service to support a planned maintenance system outage. In addition, surveillance testing was performed to support system restoration following the maintenance outage for testing in accordance with an approved surveillance procedure. Based on the post maintenance test results, the HPCI system was declared inoperable and since the HPCI system is a single train safety system, an ENS notification was made. However, further evaluation of the condition determined that no condition was discovered that could have prevented the HPCI system from performing its functions. The following information provides the basis for that determination: Review of all applicable operating data collected during HPCI system testing performed from the time of discovery of the oscillation concern indicates that (1) the resulting HPCI speed spikes were in the positive direction and therefore no concern related to the inability of HPCI to provide adequate vessel level makeup existed, (2) the RGSC output perturbations followed a consistent pattern and the magnitude of the associated control problem would not have become more severe for a period longer than the assumed HPCI mission time, and (3) none of the excursions experienced were high enough to cause a system trip at the 5000 plus/minus 100 rpm overspeed trip setpoint. Even if a spurious speed spike resulted in a HPCI overspeed trip, the HPCI overspeed trip is designed to automatically reset and allow the system to ramp back up to operating speed. Given these facts, there was no observed performance that represented a loss of system function. Had HPCI been called upon to inject during the time that the condition resulting in HPCI flow oscillations existed, the system would have met all functional requirements. Carolina Power & Light Company, doing business as Progress Energy Carolinas, Inc., has determined that this event does not meet the 10 CFR 50.72 or 10 CFR 50.73 reporting criteria and the notification for Event Number 40714, is retracted. The resident inspector has been notified. Notified R2 DO (R. Ayres). |
| ENS 40715 | 29 April 2004 11:03:00 | The licensee reported the discovery of minor leaks detected in two of four surface evaporator waste ponds used as a portion of an uranium milling process. Each pond has a leak detection system which is checked once per week. The minor leaks were detected on 4/28/04 indicating some minor leakage through the pool liner. The licensee stated there was no release of liquid to the environment. Corrective actions planned are to pump down the two ponds and inspect and repair the liner as needed. This report was made pursuant to License Condition 11.4 which specifies to notify the NRC of any spills or leaks. The licensee also notified the Wyoming Department of Environmental Quality.
The licensee determined this event was not reportable because 10 CFR40.60 and 10 CFR20.2202 did not apply. The leakage was very minor and none was released to the environment. Notified NMSS (J. Hickey) and R4DO (C. Marschall). |
| ENS 40619 | 28 March 2004 21:25:00 | At about 1349 (PST) on March 28, 2004, while a Security Offer was receiving his weapon in the armory, the weapon accidentally discharged as it was being holstered. The discharge grazed the officer's leg and the officer was treated for the resulting abrasion. At about 1425 PST, Southern California Edison notified the FBI (SONGS LLEA). SCE is reporting this occurrence in accordance with 10CFR50.72(b)(2)(xi). The NRC Resident Inspector was notified. |
| ENS 40614 | 27 March 2004 21:22:00 | The item was found adjacent to a house in a wooded area in East Lyme, CT. It was a cylinder measuring 6 inches in length and 2 inches in diameter. The bottom of the cylinder had the following serial number: M2477. It was a general licensed strontium-90 source. The source was contained inside a metal box with a radioactive material symbol on the outside. The State response personnel conducted a radiological survey. The source read 250 millirem per hour on contact for gamma. The source read 3.2 rem per hour on contact for beta. At 12 inches, the source measured 5 millirem per hour. At one meter the source measured less than 1 millirem per hour. The State took the source to a secure locked location for followup on Monday to try to determine the owner.
The item was identified as a component to a helicopter In-flight Blade Inspection System. Notified Mark Evetts at the Homeland Security Operations Center.
U.S. NRC, Region I (Mr. Jim Schmidt) informed the U. S. Navy (NAVSEADET RASO) on March 31, 2004 that an IBIS device (serial number 2397) containing 500 microCi of Sr-90 was found on private property in East Lyme, Connecticut, on March 28, 2004. The NRC also informed the Navy that the IBIS device was confiscated by the State of Connecticut, and placed in safe and secure storage. Early results of NRC inquiries pointed to the U.S. Navy as the probable owner of the IBIS device. A preliminary investigation by the Navy has found that the these devices are routinely returned to the manufacturer (General Nucleonics) for refurbishing, where they may be resold, through the Defense Logistic Agency to the Navy, the Marine Corp, the Army, or the Coast Guard. The Navy's preliminary investigation also found that these devices are also used on some commercial aircraft. The Navy's preliminary investigation verified that the Navy possessed the device until at least 1989, but found no conclusive data or documentation to show that the device was returned to the manufacturer after that time. Therefore, the Navy has accepted responsibility for the device. The Navy's preliminary investigation was unable to determine how the IBIS device came to be located on private property. The Navy has contacted the State of Connecticut, and has initiated actions to retrieve the device for proper disposal or facilitate its return to the manufacturer. The Navy will follow-up with a 30 day report as required by 10 CFR 20.2201. United States Navy General Radioactive Material license number is 45-23645-01NA. R1DO (Della Greca) & NMSS (Roberto Torres) notified. |
| ENS 40601 | 20 March 2004 15:25:00 | At 1340 on 3/20/2004, Calvert Cliffs Unit 1 Reactor automatically shutdown due to low Steam Generator Water Level. The low water level was caused by a loss of at least one Steam Generator Feed Pump. The loss was initially caused by a short or ground from Chart Recorder maintenance in Panel 1C29. 1C29 is a control panel in the control room. The chart recorder was a 500KV Bus Voltage Monitor. The post trip primary indications responded normally. The auto steam dump operation responded normally until the quick open signal was cleared at which time the Turbine Bypass Valves failed shut. It is unclear at this time why they failed shut. Steam dump continues through the use of the Atmospheric dump valves and feedwater is supplied via the Auxiliary Feedwater System with the use of 11 (Steam Driven Pump) & 13 AFW Pump (electric driven pump). Lowest Steam Generator Level was -210" in 11 S/G and -115" in 12 S/G level. Current Conditions are RCS pressure is 2250 PSIA and temperature is 532�F. All control rods properly inserted into the reactor core. The NRC Resident Inspector was notified. |
| ENS 40600 | 19 March 2004 21:00:00 | On 3/19/04 at 14:08 PM EST, the Unit-1. HPCI system was declared inoperable due to a hand-switch failure, which prevented main control room operation of the HV-055-1F0O1 HPCI steam admission valve. The system had just successfully completed its functional surveillance test and the switch broke resulting in the operators using an, alternate means to shutdown the HPCI system. The steam admission valve is not a PCIV. The valve was open all the time; the system was shutdown using an alternate procedure. The system is now blocked for hand-switch replacement. This report is being made pursuant to 10CFR50.72(b)(3)(v) for failure of a single train accident mitigation system. The NRC Resident inspector was notified. Operators entered the unit into a 14 day LCO for declaring HPCI system inoperable.
This is a retraction of the event notification made on 3/19/04 at 21:00 hours. This event (#40600) was initially reported as a safety system functional failure under the requirement of 10CFR50.72(b)(3)(v)(D). Unit 1 High Pressure Coolant Injection (HPCI) system was declared inoperable due to a handswitch failure that prevented operation of the HV-055-1F001 HPCI steam admission valve from the main control room. The handswitch failed while attempting to close the steam admission valve during system shutdown at the conclusion of surveillance testing. The operator then closed the outboard steam line isolation valve to complete the system shutdown. HPCI is automatically initiated by low reactor level or high drywell pressure signals. Manual HPCI initiation for inventory makeup is performed by depressing the initiation pushbutton. Both automatic initiation and manual initiation using the pushbutton open the steam admission valve. The failed handswitch did not adversely affect the initiation of the HPCI system in the inventory makeup mode. The manual startup of the system in the test mode of operation is unavailable when the handswitch is failed. The test mode can be used to remove decay heat following an isolation of the main steam lines but it is not a credited safety function. The steam admission valve handswitch would not be used to shutdown the system when an automatic initiation signal is present; the outboard steam line isolation valve would be closed to secure the system. This is the method that was used to secure the system during the surveillance test. The handswitch is mainly used to startup and secure the system during surveillance testing. The system was removed from service as part of a planned evolution to conduct surveillance testing in accordance with an approved procedure and the plant's Technical Specifications. The system was secured for the purpose of concluding the surveillance test and replacing the handswitch. The failed handswitch did not adversely affect the systems capability of performing its safety function. A condition did not exist at the time of discovery that could have prevented the fulfillment of the HPCI safety function. Notified R1DO (B. MCDERMOTT). |
| ENS 40599 | 19 March 2004 17:45:00 | On 03/19/04 at 0910 hours, Autoclave #5 in the X-343 Facility experienced a Steam Shutdown due to high condensate level alarms (A) and (B) actuating. The autoclave was in an applicable TSR mode (Mode II, Heating) at the time of the alarm actuations. The autoclave was placed in Mode VII (Shutdown) and declared INOPERABLE by the Plant Shift Superintendent (PSS). An investigation is underway to determine the cause of the actuations. No release of radioactive material occurred as a result of this incident. This event is being reported in accordance with UE2-RA-RE1030, Appendix D. J. 2. Safety System Actuations. |
| ENS 40591 | 15 March 2004 20:57:00 | The licensee reported that an automatic reactor trip occurred on 3/15/04 at 2020 EST due to the spurious trip of one main feed pump which caused a low steam generator water level trip signal. The plant has two main feed pumps. Operators reset the tripped main feed pump but steam generator levels didn't recover in time. The lowest level observed was in the "B" steam generator at 15% level as compared to the normal level of 65%. The trip setpoint is at 50% level. All control rods properly inserted into the core. The auxiliary feed water system automatically initiated as designed and expected. The plant remains stable in mode 3 while the licensee commenced the post trip review to determine the cause of the main feed pump trip. Decay heat removal was established using the AFW system to feed steam generators and bleed steam to the main condenser through the condenser dump valves. The licensee notified the NRC Resident Inspector. |
| ENS 40571 | 6 March 2004 08:20:00 | On 3/06/04 at 0528 Plant Security was notified of an accident at the entrance to the site involving an employee leaving work and a south bound vehicle on PA Route 11. There were no reported injuries. Local law enforcement was contacted and investigated the incident. Because of the involvement of a LLEA and potential media or general public interest in the event, the Pennsylvania Emergency Management Agency (PEMA) was notified of the incident at 0812 hours. Based on the notification to a government agency and possible public interest, this event was determined to be reportable under 10CFR50.72(b)(2)(xi). The NRC Resident Inspector was notified. |
| ENS 40570 | 6 March 2004 02:02:00 | The licensee reported that the "B" Steam Generator Feed Pump tripped unexpectedly and would not reset causing lowering steam generator water levels. Operators manually tripped the reactor and all control rods properly inserted. The auxiliary feed water (AFW) system automatically initiated to restore steam generator water levels. The lowest steam generator water level observed during the event was 55% level as opposed to the normal level of 70%. No primary relief valves lifted. Operators established decay heat removal capability using AFW system and the atmospheric steam dump valves. The licensee initiated a post trip review to determine the cause of the feed pump trip. The NRC Resident inspector has been notified by the licensee. |
| ENS 40545 | 25 February 2004 14:05:00 | Made 2 hour notification to the New Jersey Department of Environmental Protection of a New Jersey pollutant discharge elimination system permit violation. The limit for petroleum hydrocarbons is a daily average of 10 ppm over a month period. Current projected amount is 40 ppm. The sample is taken at the oil water separator outfall for liquid going to the storm drains. The NRC Inspector was notified. |
| ENS 40537 | 22 February 2004 18:42:00 | Peach Bottom Unit 2 reactor was manually scrammed due to degrading main condenser vacuum. The reactor was manually scrammed prior to reaching the automatic scram setpoint. All plant systems responded as expected with no significant issues noted. A Group II and Group III Primary Containment Isolation was received due to reactor water level passing through 1 inch. All isolation systems responded as required and repositioned to their expected positions. The licensee also indicated that all control rods properly inserted into the core. The method of decay heat removal was using the main condenser. The licensee initiated a post scram review to identify and correct the source of degrading vacuum. The licensee also indicated the manual scram was initiated at 25 inches and lowering of condenser vacuum. The licensee notified the NRC Resident Inspector. |
| ENS 40598 | 19 March 2004 16:17:00 | This report is being made in accordance with 10CFR50.73 (a)(1), which states, in part, "in the case of an, invalid actuation reported under 10 CFR50.73(a)(2)(iv), other than actuation of the reactor protection system (RPS) when the reactor is critical, the licensee may, at its option, provide a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER." These invalid actuations are being reported under 10CFR50.73(a)(2)(iv)(A). On February 22, 2004, at 1016 hours, while operating at 100 percent power, the reactor protection system manual scram channel functional test was performed. When a division 1 manual scram pushbutton was depressed, in addition to the expected half scram, the reactor protection system (RPS) A motor-generator electrical protection assembly circuit breakers tripped open. Loss of power from the RPS A motor-generator caused the loss of electrical power to the RPS instrumentation. This resulted in the invalid actuation of the logic that is powered by RPS A. The actuations resulted in isolation signals that closed one or more valves in each of the following division 1 subsystems: main steam line drains, containment and drywell radiation monitors, reactor water cleanup, fuel pool cooling, suppression pool cleanup, containment and drywell radwaste sumps and containment chilled water. All systems and components responded as designed for the signals that resulted from the loss of RPS A buss. This is considered a partial actuation since only division 1 components were effected. Following restoration of power to the RIPS A buss, the actuation logic was reset, and the equipment/systems were returned to the status required by plant conditions. Discussion of the causes and corrective actions associated with this event are documented in the corrective action program in condition report 04-00901. The resident inspector has been notified. |
| ENS 40535 | 21 February 2004 12:06:00 | At 10:05 AM PST, the Susquehanna LLC Shift Manager was notified that a member of the general public required medical assistance. The individual was at a company owned public recreation area adjacent to the Susquehanna LLC River Intake access road. The recreation area is outside the protected area. An ambulance was called to the scene, arriving at 10:28 and leaving at 10:52. The individual was transported to a local hospital. The LLEA and PEMA were notified of the incident. The NRC Resident Inspector was notified. |
| ENS 40540 | 23 February 2004 14:20:00 | The licensee at University of Virginia Hospital reported an event where a radioactive medical source was missing for approximately 2 hours. The patient was being treated for uterine cancer. At the end of the treatment, the licensee removed 8 catheters from the patient. Unknown at the time, one ribbon with 8 seeds of Ir-192, with an approximate activity of 5 millicuries, fell onto the floor. The licensee performed a search and radiological surveys, and the missing ribbon was located 2 hours later in a trash compactor. The licensee is performing an assessment of any unplanned exposures that resulted from this event.
The licensee has concluded their assessment of any unplanned exposures from this incident, along with determining its root cause. In the unlikely case that the patient was laying directly on top of the source (on contact) for the entire 30 minutes from the time the physicians removed the sources until the missing ribbon was discovered, her skin exposure would have been 662 rad. This exposure is less than her skin exposure from other treatments of the tumor thus far, and would result in minimal adverse effects. If the 30 minute exposure occurred at a distance of 1.5 millimeters from the patient, her exposure would have been 41 rad to the skin. It is highly unlikely that the patient received anywhere near these exposure levels, as the missing ribbon was most likely picked up with other trash shortly after the room was de-posted, and prior to the physicians discovering the loss. In the brief time (1-2 minutes) it took to transport the ribbon with other trash to the dumpster, and during the time the ribbon was in the dumpster, any exposures to additional personnel would have been negligible. The root cause of the lost ribbon is that the meter used to survey the room following the procedure was defective. Another meter was used to locate the ribbon in the trash compactor. Notified R1DO (Shanbaky) and NMSS (Essig)
Due to skin reddening on the patient, the patient may have received an over exposure to the thigh. It is unknown how long the source was stuck to the patient's skin or the exposure. The licensee is continuing to investigate. Notified R1DO (Cobey) and NMSS (Brown) |
| ENS 40510 | 10 February 2004 11:02:00 | The licensee reported a damaged/missing Humboldt Density Gauge at a demolition site located in East Orange, New Jersey near the intersection of Gould Avenue and Steuben Street. The gauge was a model HS-5001 with a serial number of 2822. The source contains less than 11 millicuries of Cs-137 and less than 44 millicuries of Am-241:Be. Only the gauge faceplate was found and no other parts including the radioactive source. The licensee performed a radiological survey and did not detect any levels greater than background. The licensee started interviews to determine what happened and where the remaining parts are located. The licensee notified the NRC Region I office on 2/9/04 and two NRC inspectors were arriving onsite at the time of this report. |
| ENS 40459 | 19 January 2004 20:07:00 | A non-licensed contract supervisor refused to take a fitness-for-duty test resulting in the termination of his site access. Contact the HOO for additional details. The licensee indicated they will notify the NRC Resident Inspector. |
| ENS 40458 | 19 January 2004 16:38:00 | At approximately 10:30 CSTon 1/19/04, the Doe Run Company, located in Bunker, MO, reported that the source head of a moisture density gauge broke and fell to the floor. The gauge was being used in a slurry density line for mining and milling operations. The gauge is a Texas Nuclear, model SG-5191, gauge containing 500 millicuries of CS-137. The source serial number is B-1319. The licensee indicated that the area was isolated and the gauge manufacturer was contacted for corrective action. The source remained within the gauge head. |
| ENS 40460 | 21 January 2004 16:20:00 | A medical event occurred on 1/19/04 at the Bowling Green Medical Center located in Bowling Green, Kentucky. Specifically, an inner vascular Brachytherapy treatment was planned. The catheter ran outside the body through an external valve which was inadvertently partially shut resulting in no dose to the target area. As a result, the doctor administering the treatment received .736 gray at his fingertips. Also, the patient received .736 gray to the thigh area. The source involved was 43.14 curies of strontium-90. There was no significant adverse health effects from this event. A review was initiated by the licensee to determine the cause and to initiate corrective actions. |
| ENS 40461 | 21 January 2004 16:28:00 | On 1/16/04, the radiation detectors alarmed at the North American Stainless scrap yard, located in Ghent, Kentucky, when a truck entered the scrap yard. The scrap yard radiation detectors indicated 42 microrem per hour. The truck was not accepted and sent back to its origin. The scrap material in the truck came from Industrial Services of America. Investigation determined that a piece of a fixed gauge caused the radiation alarms to set off. Apparently, the gauge went through a shredder. On-contact radiation readings were 35 millirem/hour with the shutter closed and 58 millirem/hour with the shutter open. This was considered a loss of radioactive material and further review will be performed to try to determine the owner of the gauge. |
| ENS 40436 | 10 January 2004 15:22:00 | During the performance of a normal orderly planned shutdown for maintenance a packing leak was discovered on Main Steam Inboard Isolation Valve 01-02 (01-02). Drywall inspection results revealed MSIV 01-02 operator gear housing degraded allowing grease to be lost. Engineering assessment of the condition of the limitorque operator of MSIV 01-02 is not able to provide enough information to support operability of MSIV 01-02. Declaring MSIV 01-02 inoperable. Entering T.S. 3.2.7(b and c). Action required to close one valve in the line having an inoperable valve or initiate a normal orderly shutdown with in one hour and be in cold shutdown within ten hours. Entering T.S. 3.3.4(b and c). Action required to close one valve in the line having an inoperable valve within four hours or reduce reactor coolant temperature to a value less than 215 degrees F within ten hours. Most limiting action required in current condition is 10 hours to be in cold shutdown. 01-02 will be closed as soon as all rods are in. The NRC Resident Inspector was notified. |
| ENS 40433 | 8 January 2004 19:12:00 | Event Description: Reactor Coolant System, (RCS) Leakage in excess of TS limits. RCS leakage calculations have shown a gradual increase in leakage following Unit l start-up from a refueling outage. On 1/7/2004 @ 1800 hours an RCS leakage calculation was performed with the calculated leak rate at 0.300 gpm. On 1/8/2004 @ 1225 hours, with Reactor power at approximately 26% Rated Thermal Power (RTP), leakage had increased to 0.778 gpm. Personnel have entered containment and have determined the source to be within the A steam generator cavity. At 1312 hours a power reduction to 17% RTP was initiated to reduce dose sufficiently to allow entry into the cavity in an attempt to identify the exact leak source. On 1/8/2004 @ 1502 hours, the leak rate was calculated to be one gpm which exceeds the allowed leakage for unidentified leakage. Units 2 and 3 are unaffected and remain at 100% RTP. Initial Safety Significance: The measured leakage is well within the capacity of the normal RCS make-up system. The increase in leakage rate has been slow, and is not expected to increase significantly prior to completion of the TS required shutdown. The leakage is confined inside containment. At this time, there has been NO RADIOLOGICAL RELEASE associated with this event. NOTE: Entry into the emergency plan is not required unless the leakage is => 10 gpm unidentified or => 25 gpm identified, the current leakage is significantly below those limits. Corrective Action(s)= Unit shutdown per TS is in progress. Isolation or repair of the leak will be evaluated once the location of the leak source is determined. TS 3.4.13 allows 4 hours to reduce leakage within limits. If the source remains unidentified and leakage remains above limits, the TS requires shutdown to Mode 3 within 12 hours and to Mode 5 within 36 hours. The NRC resident inspector was notified. |
| ENS 40432 | 8 January 2004 18:45:00 | The licensee reported that the upper lights on the meteorological tower were not working. The licensee made an offsite notification to the FAA and will initiate corrective actions. The licensee notified the NRC resident inspector.
The licensee is retracting this notification after a further review of the reporting requirements determined that the notification was unnecessary. The condition reported to other government agencies was not related to the radiological health and safety of the public or onsite personnel or the protection of the environment. The NRC Resident Inspector was informed. Notified Reg 1 RDO (Noggle) |
| ENS 40301 | 7 November 2003 04:15:00 | Inserted a Manual Reactor Scram due to high reactor coolant conductivity. Source of high reactor coolant conductivity is not certain at this time. In addition, this report is being made under 50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any systems listed in paragraph (b)(3)(iv)(B) of this section' due to PCIS groups 2, 3, and 4 being received when reactor water level dropped below 170". All isolations went to completion. The reactor water level drop is normal following a scram from 45% power due to void collapse in the reactor vessel. Reactor water level was restored to normal and the PCIS group isolations were reset. The current plan is to cool down the plant and enter Mode 4, plant cool down is being achieved using the RCIC system until the source of the conductivity transient is identified and isolated. The licensee indicated shortly prior to the scram a coolant chemistry transient was in progress. The limit for reactor coolant conductivity is 5 micro mhos per centimeter and the highest conductivity reading reached was 8 micro mhos per centimeter. The licensee speculates the most likely source of the high conductivity is related to an off service condensate demineralizer. There were no automatic ECCS initiations. No SRVs lifted. All control rods properly inserted. The licensee notified the NRC Resident Inspector. |
| ENS 40298 | 5 November 2003 03:45:00 | Unit 2 auxiliary feedwater support systems actuated during scheduled ATWS testing when an unrelated clearance order placement de-energized two 6.9 kv busses (256 and 258). The 2 of 4 6.9 kv bus undervoltage coincidence initiated a valid auto-start signal causing lube oil pumps 2AF01PA-A, 2AF01PB-A and 2AF01PB-C to start. Auxiliary Feedwater Pump AOV discharge valves 2AF004A and 2AF004B auto opened. 2AF01PA 4kv breaker, which was in the equipment test position. Neither auxiliary feedwater pump started and no water transferred to the steam generators. The licensee notified the NRC resident inspector. |
| ENS 40279 | 27 October 2003 11:59:00 | Radiographic exposure device (Camera) was left unsecure on the tailgate of a company pickup truck before departing Licensee's local site enroute to the work site at: Huntsman Polymers, 2400 South Grandview Avenue, Odessa, Texas 79766. The Radiographer Trainer, (deleted), and his Trainee enter into conversation with another company employee and failed to block and brace the device or to secure the device to their truck. They drove out of the shop area and onto the street in front of the office. Approximately 100 yards from the office, the device fell off the tailgate and onto the service road. Approximately 10 minutes later, an NDT customer enroute to the Licensee's facility came by and noticed the camera in the road. He recovered the device and transported it to the Licensee's Odessa office. The crew was notified by cell phone and returned to the shop. The camera is an Industrial Nuclear Company, Model IR-100 exposure device, Serial No. 4318, containing 80 curies of iridium 192. The camera was surveyed for external damage and radiation at the shop. No external damage was noted and the results of the radiation survey revealed radiation levels that were the same as when the device had initially been surveyed that day. The exposure device was taken to a safe location, attached to associated equipment and operated to determine if the device had suffered internal damage. No damage was noted as the device operated perfectly. The device was leak tested. Results of this test have not been returned to the company as of October 27, 2003. As corrective action both the Radiographer Trainer and the Trainee have received a written warning under the Licensee's disciplinary policy. Both individuals will be required to participate in several radiation safety programs and be re-tested. The Trainee will be require to attend another 40-hour radiation safety class in December 2003. The Licensee is being cited for: failure to physically secure radioactive material; and failure to make an immediate (24-hour) notice to this Agency. In addition, the Radiographer Trainer will also be cited for failure to secure the device. Escalated Enforcement actions have been recommended which will include assessment of an Administrative penalty for the Licensee. |
| ENS 40256 | 17 October 2003 11:54:00 | System Affected : Steam and Feedwater Rupture Control System (SFRCS) Actuation / Initiator Affected: Main Steam Line Rupture on Steam Generator Effect of event on plant in current Mode - None Actions Taken or planned: Declared SFRCS Actuation Channels 1 and 2 inoperable. Will track as a Mode 3 restraint against T.S. 3.3.2.2 (SFRCS Instrumentation). Will resolve, correct and retest prior to entry into Mode 3. Discovered SFRCS Logic Channel 1 block upon restoration and repowering after maintenance activities on 10/16/03. After discussions with Engineering Department representatives it has been determined that a design issue exists with the SFRCS system where it would be possible that SFRCS Channels 1 and 2 low pressure trips could be blocked during a scenario where a Main Steam Line Rupture would be followed by a loss of off-site power. This would result in the potential that the OTSG (once through steam generators) with the main steam line rupture would be fed by the AFW system. This requires an 8 hour notification to the NRC per 10CFR50.72(b)(3)(ii)(B). The licensee notified the NRC resident inspector. |
| ENS 40255 | 16 October 2003 17:53:00 | Double Contingency Protection: Double contingency protection for the ADU Bulk Blending System is assured by (1) preventing moderator from becoming available to a bulk container, and (2) preventing moderator from entering a bulk container. The first contingency did not occur because the moderator was never available to a bulk container. Moderator is prevented from entering a bulk container by preventing high moisture polypaks from being dumped into the bulk container. The polypaks are processed through a scan and dump interlock, which prevents unacceptable polypaks from being dumped. The software malfunction left less than previously documented double contingency protection for the system. In accordance with Westinghouse Operating License (SNM-1107), paragraph 3.7.3 (c.5a), this event satisfies the criteria for a 24-hour notification. As Found Condition: See Reason for Notification above. Summary of Activity: Immediately after all packs were dumped into the bulk container to complete the blend of material, Operations noticed that the packs had not been denoted as "consumed" by the data base, and notified the computer system administrator and Nuclear Criticality Safety. The computer system administrator stopped all dumping operations. The computer system administrator immediately checked all packs that had been dumped into the blend. All moisture values were acceptable. All operations that use the same PLC interface program for criticality controls were stopped. Conclusions: Less than previously documented double contingency protection remained. All moisture values for the material involved were acceptable, and the total amount of moderator in the blend was very low, far less than criticality limits. At no time was the health or safety to any employee or member of the public in jeopardy. No exposure to hazardous material was involved. The Incident Review Committee (IRC) determined that this is a safety significant incident in accordance with governing procedures. A causal analysis will be performed. |
| ENS 40254 | 16 October 2003 13:47:00 | The licensee declared an Unusual Event at 12:30 CDT pertaining to a small fire, located at the high pressure turbine number 1 bearing, which was burning greater than 10 minutes. There were no injuries and the turbine continues to operate at full power while the fire brigade attempts to extinguish the fire. Subsequently, the licensee extinguished the fire; however, several minutes later, the fire restarted. The licensee decided to lower power to approximately 65% power to lower general area radiation levels for the fire brigade personnel. The fire was extinguished and reflashed a few times. The fire brigade used water, dry chemicals and foam when fighting the fire. The source of the fire at the bearing is under investigation by the licensee.
The licensee reports that the fire is out and the investigation determined that the most likely source of the fire was oil pooled under the bearing pedestal. No current oil leakage was found. The licensee planned on contacting the turbine vendor prior to returning the plant to full power. The Unusual Event was terminated at 1541 CDT. Notified R4DO (G. Sanborn), DIRO (D. Weaver), NRR EO (T. Reis), DHS (J. Clardy) and FEMA (J. Dunker). |
| ENS 40250 | 15 October 2003 19:30:00 | At 1603 Salem Unit 1 commenced a down power to comply with Technical Specifications. Tech. Spec. 3.0.3. was entered at 1505 when it was determined that the feed regulation valve for the 14 Steam generator (14BF19) was bound at approximately 71% open. This valve is required to close on a feedwater isolation signal. Currently the reactor is at 40% power and lowering with plans to stabilize at 25% prior to taking the unit off line (mode 3) by 2205 tonight. Unit 1 has no other tech spec equipment out of service. Unit 1 is in two additional shutdown Technical Specifications, both are related to the control room ventilation system which is shared with Unit 2. Unit 2 is in a scheduled refueling outage and one of the Unit 2 emergency intake dampers is disabled (T.S. 3.7.6.1.a). This is a 7 day LCO that expires on 10/22/03. Additionally Unit 2's emergency filtration system is out of service for planned maintenance (T.S 3.7.6.1.a). This is a 30 day LCO that expires on 11/13/03. All other systems on Unit 1 are operating as expected for current plant conditions. The licensee notified the NRC Resident Inspector. |
| ENS 40227 | 7 October 2003 06:20:00 | The Unit 2 HPCI system has been declared INOPERABLE due to the failure of a pressure suppression pool high-level switch during surveillance and being unable to maintain the HPCI pump suction lined up to the pressure suppression pool in standby lineup. Per technical specification 3.3.5.1. required action D.2.2. with the high level switch INOPERABLE the HPCI system must have the pump suction aligned to the suppression pool with in 24 hours or declare the HPCI system INOPERABLE when the required action completion time can not be met. The function of the suppression pool high-level switch causes the HPCI pump suction to automatically realign to the suppression pool when the suppression pool reaches the high level set point during an accident. When the pump suction swap to the suppression pool was performed the HPCI gland seal leak off pump began to run automatically. Further investigation determined that there was an unknown input into the HPCI gland seal leak off condenser (GSLO). Normally the GSLO does not have an input while in standby lineup and therefore does not normally operate while the pump suction is aligned to the Pressure Suppression Pool. The GSLO pump would have a discharge flow path with the HPCI system running and therefore would have the ability to maintain its design function. However in the standby line (-up) with the pump suction aligned to the suppression pool the GSLO pump did not have a discharge flow path. This inability to pump the GSLO condenser required the suction be realigned to the condensate storage tank and with the combination of a failed suppression pool high-level switch requires declaration of INOPERABITY of the HPCI system. This is a 14 day LCO. This is a single train system and reportable under SAF 1.8 Event or Condition That Could Have Prevented Fulfillment of a Safety Function. The INOPERABLE HPCI system function can be manually initiated from the main control room. Investigation and troubleshooting into the unknown HPCI GSLO input has been, initiated in parallel with the repairs to the failed suppression high level switch. The licensee intended on notifying the NRC Resident Inspector. |
| ENS 40225 | 5 October 2003 06:26:00 | While performing common mode failure testing of the Emergency Diesel Generators (EDG), the 'C' EDG was declared INOPERABLE for planned installation of required test equipment. Concurrent with the inoperability of the 'C' EDG, the 'B' Control Room Emergency Filtration (CREF) System has been INOPERABLE for emergent corrective maintenance since 10/2/03 at 0502. Because the 'C' EDG is the emergency power supply for the 'A' CREF train, 'A' CREF was also declared INOPERABLE and Technical Specification 3.0.3 was entered as of 0300 hrs on 10/05/03. At 0430 hrs on 10/05/03, the test equipment was removed from 'C' EDG, thereby restoring it and 'A' CREF to an operable status, and Technical Specification 3.0.3 was exited. Testing did verify the absence of a common mode failure and all EDG's are operable. The Control Room Ventilation System provides heating, cooling, ventilation, and environmental control for the control room and adjacent areas. Under accident conditions, CREF ensures that the control room will remain habitable during and following all design basis accidents. Because the CREF system is required to automatically respond in the event of a design basis accident, having both trains of CREF inoperable at the same time impacted the ability to mitigate the consequences of an accident. Therefore, this event is being reported in accordance with 10CFR50.72(b)(3)(v)(D). The plant is currently in HOT SHUTDOWN for repair of an emergent turbine hydraulic fluid leak, with decay heat removal to the main condenser via turbine bypass valves. The NRC resident inspector was notified by the licensee.
At the time of the original notification, both trains of Control Room Emergency Filtration (CREF) were declared inoperable impacting the ability of CREF to mitigate the consequences of an accident. The "B" train was inoperable for emergent corrective maintenance and the "A" train was declared inoperable when test equipment was connected to the "C" emergency diesel generator (EDG). The "C" EDG provides emergency power to the "A" train of CREF. Subsequent to this event, an evaluation of the test equipment impact to the "C" EDG was performed and determined that the "C" EDG would still be capable of providing emergency power to the "A" CREF train in the event offsite power is lost. Therefore, the "A" CREF train remained available to respond to a design basis accident. Thus, the safety function would have been fulfilled." R1DO (Brain McDermott) notified. The NRC Resident Inspector will be notified of this retraction by the licensee. |
| ENS 40198 | 26 September 2003 01:25:00 | On 09/25/03 @ 2210, while restoring reactor vessel instrumentation to service on Unit 3, an invalid ECCS actuation, (reactor vessel lo-lo-lo) occurred which caused all four (4) EDG to automatically start. All eight (8) 4 KV buses remained supplied from offsite sources. The Unit 3 ECCS pump auto starts were previously defeated per plant procedure and were not required to be operable per Tech Specs. The invalid ECCS initiation signal caused the offsite power Loss-of-Power instrumentation setpoints to transfer to the degraded voltage LOCA setpoints. With the degraded voltage LOCA setpoints initiated, the 4 KV E-bus fast transfer capability on a degraded voltage-NON-LOCA condition is defeated. With the inability to fast transfer the 4 KV E-buses on a degraded voltage NON-LOCA condition, both offsite sources are inoperable. Prior to restoring the reactor vessel level instrumentation, the E-2 diesel generator was inoperable. With both offsite source.; and one EDG inoperable, LCO 3.8.1 Required Action H.1 requires an entry into LCO 3.0.3 for Unit 2. With Unit 3 in MODE 5, LCO 3.0.3 is not applicable. After one (1) hour, actions were taken to initiate a plant shutdown per LCO 3.0.3. No negative reactivity was added as the preparations were administrative in nature. At 09/26/03 @ 0006, the invalid ECCS initiation was removed. And all offsite Loss-of-Power instrumentation has been returned to OPERABLE status. Following the automatic start of all four (4) EDGs, E-1, E-3, and E-4 were successfully shutdown per plant procedures. E-2 could not be shutdown per plant procedures and was shutdown locally. The cause of the inability to shutdown E-2 per plant procedures is under investigation. E-2 EDG remains inoperable. The NRC Resident Inspector was notified by the licensee.
This 60-day optional report, as allowed by 10 CFR 50.73(a)(1), is being made under the reporting requirement in 10CFR50.73(a)(2)(iv)(A) to describe an unplanned, invalid actuation of a specified system, specifically the Units 2 and 3 Emergency Diesel Generators (EDGs).' 'On 9/25/03 at 2210 while restoring reactor vessel instrumentation to service on Unit 3, an invalid EDG start actuation signal on Reactor Vessel Lo-Lo-Lo level occurred which caused all four Emergency Diesel Generators to automatically start. Investigation into the cause of the event identified that the sequence of operating instrument valves associated with the instrument rack for level transmitters LT 72A and LT 72C resulted in the invalid reactor vessel lo-lo-lo reactor water level signal and subsequent EDG logic actuation. The EDGs initiated as expected for the given conditions. Off-site power was not affected and continued to supply power to the emergency busses. Once the reason for the EDG initiation was determined, the E-1, E-3, and E-4 EDGs were shut down per plant procedures from the Main Control Room by approximately 2255 hours. The E-2 EDG could not be shut down normally from the Main Control Room control switch. It was later shut down at the local control panel in the E-2 EDG bay at about 0030 hours on 9/26/03. Repairs to the E-2 EDG control switch were completed in accordance with the Corrective Action Program (CR 177605).' 'Notification of the event to the NRC was initially made at 0125 on 9/26/03 (EN # 40198). Since the initial report, it was determined that a Technical Specification LCO 3.0.3 condition did not exist since the off-site sources were operable. This event has been entered into the site-specific corrective action program for resolution (CR 1776 10). The licensee has informed the NRC Resident Inspector. |
| ENS 40196 | 24 September 2003 03:20:00 | At 0053 hours on September 24, 2003 with Susquehanna Unit 1 operating at 100% power an automatic reactor scram occurred due to low water level. At the time of the scram, reactor feed pump testing was in progress and the 'C' reactor feed pump tripped. The reactor recirc pumps runback initiated as expected when water level reached 30" with the feed pump tripped. Level continued to drop and reached the Level 3 auto scram setpoint. Level continued to drop and reached a low level of approximately -48" wide range. Reactor Core Isolation Cooling and High Pressure Coolant Injection auto started at their initiation setpoints and injected to the vessel to recover level. All level 2 and 3 containment isolations occurred as expected. The reactor recirc pumps tripped as expected when level 2 was reached. Reactor Pressure was controlled with bypass valves, there were no Safety Relief Valve lifts. There are no challenges to containment. Unit 1 is currently stable in Mode 3 with both reactor recirc pumps restarted. A human performance error was the cause of the reactor feed pump trip. Investigation is continuing into the plant response to the reactor feed pump trip. The NRC Resident Inspector was notified of this event. |
| ENS 40193 | 23 September 2003 13:00:00 | On 23 September 2003, the licensee notified the Agency that a Humboldt Scientific moisture/density gauge, serial number 1136, had been run over by a roller at a construction site in the vicinity of interstate 8 and Avenue 3E in Yuma. The source was in the backscatter position at the time; however, the licensee was able to retract the source into its shielded position. A leak test is being performed. The Agency will continue to investigate. The gauge contains approximately 10 millicuries of cesium 137 and 40 millicuries of americium-beryllium. The NRC and Governor's office are being notified of this event. |
| ENS 40192 | 23 September 2003 11:15:00 | During a scheduled bare metal visual inspection of the Unit 1 reactor vessel head prior to RV head retirement, evidence of possible through wall leakage was observed on two control rod drive mechanism (CRDM) and one thermocouple (T/C) penetrations (nozzles 6 and 16 and T/C nozzle 7). Of these locations, only the T/C had been previously repaired (plugged) in December 2000. Initial Safety Significance: Any RCS leakage from these penetrations would have been below the threshold of measurability by the reactor coolant system leakage measurement process. Total measured RCS leakage prior to unit shutdown was varying between 0.15 gallons per minute and .24 gallons per minute. Corrective Action: The reactor vessel head is scheduled for replacement during the present refueling outage. Therefore, there are no plans at this time to perform additional inspections or repairs on the current head. The NRC Resident Inspector was notified. |
| ENS 40164 | 16 September 2003 23:17:00 | Hurricane warning issued. There is no significant impact to the safety of the plant at this time. The plant area is not currently experiencing any unusual winds. Unit 1 is at 94% and Unit 2 is at 96% power. The wind speed at the site is 14 miles per hour at the time of this report. The NRC Resident Inspector was notified by the licensee.
Brunswick has terminated the Unusual Event at 1740 EDT. This is based upon the hurricane warning being lifted. The licensee has notified the NRC Resident Inspector. Notified Wessman (DIRO), Boland (R2DO), FEMA (C. Bagwell), DHS (C. Wilson), Calvo (NRR EO). |
| ENS 40162 | 16 September 2003 12:35:00 | A contract employee, apparently suffered a heart attack, while in processing for Refueling Outage 09. The employee was treated by the site EMTs and transported to the hospital. It was subsequently reported that the employee passed away. Notification that the employee expired was at 11:03 EST. The licensee notified OSHA as required for a heart attack fatality. The NRC Resident Inspector was notified. |
| ENS 40160 | 15 September 2003 18:25:00 | A patient was administered 15 millicuries of I-131 sodium iodide rather than 4 millicuries for a post-ablation scan. The administrative dosage differed from the prescribed dosage by more than 20% of the prescribed dosage. |
| ENS 40159 | 15 September 2003 21:40:00 | The licensee reported an apparent abnormality involving Boral spent fuel pool test coupons. Westinghouse built the spent fuel pool racks using Boral material manufactured by ARR Inc. Specifically, inspection of test coupons revealed bulging or blistering of the aluminum cladding. The information provided by the licensee does not include any other affected plants. |
| ENS 40139 | 7 September 2003 23:15:00 | At 0750 on 9/7/03, it was discovered that the Recirculating Cooling Water (RCW) Supply valve for C333 Unit 6 Cell 2 was not positioned correctly for the current condition of the cell, in violation of NCSA CAS-011. On 9/3/03, the cell was in a fluorinating environment with the Odd R-114 system drained and evacuated for leak repairs, Odd RCW valves closed, the Even R-114 system was not drained with the Even RCW valves open satisfying the conditions of CAS-002. It was determined that the cell needed to have a UF6 negative obtained for maintenance work. The UF6 negative was initiated without closing the Even RCW Supply valve, tagging both the Supply and Return valve, and without performing the independent checks for valve position, which violates NCSA CAS-002. Once the UF6 negative was obtained, the cell transitioned to NCSA CAS-011 without satisfying the RCW isolation controls of that NCSA. Both RCW isolation controls require that the RCW Supply valve be tagged closed and that the RCW Return valve be tagged open and both valves independently verified to be positioned correctly. The NRC Senior Resident Inspector has been notified of this event. SAFETY SIGNIFICANCE OF EVENTS: Since the process condition (lack of moisture in the coolant, and therefore in the process gas system) was maintained, the safety significance of this incident is low. POTENTIAL CRITICALITY PATHWAYS INVOLVED (BRIEF SCENARIO OF NOW CRITICALITY OF HOW CRITICALITY COULD OCCUR): Once the fluorinating environment is removed from a process cell, moisture that may leak into the process gas system could potentially moderate any uranium that may be present. Sufficient water would have to leak into the process gas system and moderate a critical mass of uranium. CONTROLLED PARAMETERS (MASS, MODERATION, GEOMETRY, CONCENTRATION, ETC Double contingency is maintained by implementing two controls on mass. NUCLEAR CRITICALITY SAFETY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION OF THE FAILURES OR DEFICIENCIES: The first leg of double contingency is based on ensuring that a coolant condenser leak will not introduce RCW (moderator) into the process gas side of a cell through the coolant. This is accomplished by either maintaining a fluorinating environment in the cell or by restricting /isolating the RCW supply prior to removing the fluorinating environment. This restriction/isolation is accomplished by either removing a supply spool piece or closing and tagging the manual supply valve and tagging open the return valve. Neither of these two RCW alignments was maintained. The second leg of double contingency is based on an independent verification of the RCW alignment relied upon for the first leg of double contingency. The independent verification was not performed. Therefore, this control was lost, Since double contingency is based on two controls on one parameter, double contingency was not maintained. CORRECTIVE ACTIONS TO RESTORE SAFETY SYSTEMS AND WHEN EACH WAS IMPLEMPNTED: At 1010 on 9-7-03, the Even RCW Supply valve was tagged closed, the Odd RCW Return valve was tagged open, and both valves were independently verified. The coolant moisture content was checked and was less than minimum detectible moisture. These actions have placed the system back in compliance with NCSA CAS-011. |
| ENS 40135 | 5 September 2003 17:10:00 | Transport of a potentially radioactively contaminated person to Carolinas Medical Center due to head and neck injury. This is a conservative classification. Initial frisks of the injured individual did not reveal any contamination. His back could not be frisked because he was on a stretcher with potential head and neck injuries. The licensee notified the NRC Resident Inspector.
Further radiological surveys determined the individual was not contaminated. |
| ENS 40136 | 5 September 2003 19:44:00 | On September 4, 2003, two patients were scheduled for prostate brachytherapy implants. The first was suppose to be implanted with Iodine 125 and the second with Palladium 103. In anticipation of these procedures both the Palladium 103 and Iodine 125 were brought to the surgical suite. The first implant procedure was started and after two strands of Palladium 103 had been implanted it was discovered that an error had been made because the patient was scheduled to be implanted with Iodine 125. The treatment plan had been performed for Iodine 125. The decision was made to continue with the implanting of the Palladium 103 after a new treatment plan was performed. Based on the new plan the appropriate number of Palladium 103 seeds were implanted delivering the prescribed dose. Initial report from (DELETED) was that the patient was scheduled for 7:30 AM, so the event most likely occurred around 8:00 AM. Specific details will be included in his full report to follow. Because the treatment planning equipment was already in the surgical suits, the new plan could be developed with little delay. Specific time frame delays to follow in the full report. The second patient, scheduled for Palladium 103 was in the same day surgery suite and had not been prepped nor given an IV. He was rescheduled for another day, as they did not have enough Palladium 103 for both patients. |
| ENS 40137 | 5 September 2003 23:47:00 | A nuclear gauge was stolen during the night from a pickup truck parked at a residence. The chain and lock were cut and the gauge was taken. We are unaware of whether other items were taken. The technician became aware the gauge was missing when he arrived at a jobsite at about 5:30 am on 09/05/2003. The field service manager looked around the neighborhood and at local pawn shops but did not locate the gauge. The Licensee is considering a public announcement and the offering of a reward. The Agency is investigating the incident. The gauge is a Troxler 3430, serial # 17272, with 39.2 milliCuries of Am-241 (serial # 47-12694) and 6 milliCuries of Cs-137 (serial # 50-6643). |
| ENS 40099 | 25 August 2003 13:18:00 | The licensee reported five I-123 capsules that were lost when they fell out of the back doors of a panel van during shipment by Associates Couriers International (ACI). This occurred in Bridgeton, MO at Lindberg Boulevard at Blake. The I-123 capsules each contained 600 microcuries during precalibration and were calibrated for noon at 8/26/03 for 100 microcuries. The capsules are used for diagnostic thyroid disorders. The licensee indicated there was no significant health aspect of this event due to the short half life (i.e., 13 hours) of I-123 and the relatively low activity level. |
| ENS 40095 | 22 August 2003 09:05:00 | The licensee reported that an automatic reactor scram occurred at 0259 due to a turbine generator trip from a hi-hi moisture separator water level. One control rod indicated position 02; however, the remainder of the control rods indicated full in. All other systems and components functioned as designed. The licensee is cooling the plant down to cold shutdown. The licensee notified the NRC Resident Inspector. |
| ENS 40076 | 15 August 2003 15:27:00 | The US Navy reported a lost device involving an IBIS (in-flight blade inspection system) used on helicopters. The device uses 500 microcuries of strontium-90. The device sealed source and registry number is CA321D103G. The master materials license number is 4523645. The US Navy reported that the device most likely fell off during flight approximately 17 miles northwest of New River, NC. A search for the device was unsuccessful. |
| ENS 40431 | 8 January 2004 17:00:00 | Associates and Medical Physics reported 6 related medical events that occurred between 3/22/01 thru 5/14/03 at Douglas City Hospital located in Alexandria, Minnesota. The licensee used Samarium 153 to treat patients for bone pain. The highest intended source strength was 115 milli-curies. The licensee discovered that the patients were actually under-dosed by 34% due to an incorrect dose calibrator setting. There was no significant adverse effects to the patients. Corrective actions were taken to calibrate the dose calibrator. |