10/14/14 Hot Standby ~ Power Reactor Event Number: 50529 Facility: SURRY Region: 2 State: VA Unit: [ ] [2] AUTOMATIC REACTOR TRIP DUE TO A SPURIOUS OVERPOWER / DELTA TEMPERATURE SIGNAL Unit 2 reactor automatically tripped at 0758 [EDT] hours on 10/13/2014, due to a spurious overpower/delta temperature signal on all three channels. The cause of the spurious signal is unknown at this time. Currently, reactor coolant system temperature is being maintained stable at 546 [F] degrees. All three auxiliary feedwater pumps automatically initiated as designed on low-low steam generator level following the trip. All systems responded as expected with the exception [both] of the intermediate range neutron indication[s], which was determined to be under-compensated. The source range indication did not automatically energize and was energized manually. All other systems operated as required. This notification is being made pursuant to 10 CFR 50.72(b)(2)(iv)(B) for 4-hour notification of reactor protection system activation and 10 CFR 50. 72(b )(3)(iv)(A) for 8-hour notification of automatic actuation of auxiliary feedwater. The NRC resident has been notified of this event and is on site. There were no radiation releases due to this event, nor were there any personnel injuries or contamination events. There was no testing in progress when the reactor trip occurred. The reactor trip was considered uncomplicated. All control rods fully inserted. Decay heat is being released via main feedwater and the condenser steam dumps. Normal offsite power is available. There was no effect on Surry Unit 1 which continues to operate at 100% power. The licensee is investigating the cause of the overpower/delta temperature actuation. Hot Standby ~ Power Reactor Event Number: 50526 Facility: VOGTLE Region: 2 State: GA Unit: [ ] [2] MANUAL REACTOR TRIP DURING REACTOR STARTUP VEGP [Vogtle Electric Generating Plant] Unit 2 was performing startup and had taken reactor critical at 0929 EDT. When attempting to stabilize power to collect critical data, control rods were inserted with Control Bank D the expected group to insert. Control Bank A inserted instead of Control Bank D. Power had reached 6 E-2 percent as indicated by IR [intermediate range] indication when control room crew performed a manual reactor trip. AFW [auxiliary feed water] was in service to support plant conditions prior to the trip and did not receive any actuation signal. All equipment operated as expected. Unit 2 is currently stable in Mode 3 at normal operating temperature and pressure. The licensee has notified the N RC Resident Inspector. Hot Shutdown ~ Power Reactor Event Number: 50524 Facility: OYSTER CREEK Region: 1 State: NJ Unit: [1] MANUAL REACTOR SCRAM DUE TO DECREASING REACTOR WATER LEVEL Today at approximately 0250 (EDT) [on 10/12/14], during a planned reactor power ascension with reactor power at approximately 1% of rated thermal power, reactor water level began lowering. Operators inserted a manual SCRAM at 0251 (EDT) in accordance with station procedures. The cause of the lowering reactor level is currently under investigation. All rods inserted into the core and all systems functioned as expected during the scram. No electromatic (EMRVs) or safety relief valves lifted during the transient. The plant is currently shutdown and plant parameters are stable. This event is reportable per 10CFR50.72(b)(2)(iv)(B) - any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. Decay heat is being released to the main condenser and normal offsite power is being maintained. The licensee has notified the NRC Resident Inspector. Power Reactor Event Number: 50527 Facility: SALEM Region: 1 State: NJ Unit: [ ] [2] SERVICE WATER PUMPS INOPERABLE At 0226 [EDT on 10/12/14], the 25 Service Water Pump Traveling Screen differential pressure transmitter high side valve, 2LD2729-HIV was discovered closed while performing the monthly bubbler blow down activity. The associated 25 Service Water Pump was operable at this time. The differential pressure transmitter high side valve, 2LD2729-HIV, in the closed/discovered position would have prevented the operation of the 25 Service Water Traveling Screen due to high differential pressure. The 25 Service Water Traveling Screen needs to be operable to support 25 Service Water Pump operability. 25 Service Water Pump Traveling Water Screen was restored to operable after differential pressure transmitter high side valve, 2LD2729-HIV was reopened. The station subsequently verified all Unit 1 and 2 high side and low side differential pressure transmitter valves positions were correct. At 0600 [EDT on 10/12/14], it was identified that the last manipulation of differential pressure transmitter high side valve, 2LD2729-HIV was on 9/7/14. Based on the last known manipulation it is assumed that differential pressure transmitter high side valve, 2LD2729-HIV remained closed from that time until the condition was discovered. Review of other activities performed from 9/7/2014 to present determined that surveillance testing of 21 Service Water Pump resulted in 21, 22, and 23 Service Water Pumps being inoperable on 9/18/2014 for several hours. During that surveillance, combined with the mis-positioned instrument valve on 25 Service Water Pump, five of the six Service Water Pumps would have been inoperable which may have prevented the fulfillment of a safety function. This event is being reported under the requirements of 10CFR50.72(b)(3)(v)(B) as Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of systems that are needed to remove residual heat. The licensee has notified the NRC Resident Inspector. No one was injured as a result of the failure of 25 Service Water Traveling Screen inoperability. Refueling ~ Power Reactor Event Number: 50530 Facility: COOPER Region: 4 State: NE Unit: [1] POTENTIALLY CONTAMINATED INDIVIDUAL TRANSPORTED OFFSITE FOR MEDICAL TREATMENT At 0426 CST [on 10/12/14], a potentially contaminated individual [contract employee] was transported off-site for medical attention at Nemaha County Hospital. The individual had been working in a contaminated area in the main condenser. When the Incident Commander (IC) and Emergency Medical Technicians (EMT) arrived on station the individual was no longer in the contaminated area but was still in the Radiological Controlled Area (RCA). Radiological Protection personnel were dispatched with the individual in the ambulance and surveyed him in route. At 0445 CST, prior to arrival at the hospital, it was confirmed that the individual was not contaminated. The individual was suffering from a heat related medical condition. The licensee has notified the NRC Resident Inspector. AND Power Reactor Event Number: 50530 Facility: COOPER Region: 4 State: NE Unit: [1 OFFSITE NOTIFICATION - DROPPED CONTROL ROD DURING REFUELING OPERATIONS At 1530 CDT on 10/13/14, Cooper Nuclear Station will make a press release to the local media. This press release is with regards to the control rod blade which was dropped over the core during refueling operations when the control rod blade fell from the lifting tool and came to rest on the reactor vessel top guide in a section that contained no fuel. This press release was authorized at 1327 CDT. The control rod was dropped on 10/11/2014. There was no damage to the reactor fuel. The control rod is being replaced and is in the spent fuel pool. The licensee notified the NRC Resident Inspector. Power Reactor Event Number: 50523 Facility: PRAIRIE ISLAND Region: 3 State: MN Unit: [1] [2] MISSING FIRE BARRIER During a Fire Penetration walkdown, an opening less than 1[inch] tall under a duct was identified between the Train B Aux Feedwater Pump (AFWP) Room (Unit 1 side) and the Bus150/160 room. This constitutes a missing fire barrier between Fire Area (FA) 32 and FA 37 such that the required degree of separation for redundant safe shutdown trains is lacking. A firewatch has been established on the Unit 1 side of the AFWP Room and Bus 150/160 room. The compensatory fire watch will remain in place per F5 Appendix K until the fire barrier is returned to full functional status. The discovery of this non-compliance is being reported as an unanalyzed condition as defined by 10CFR50.72(b)(3)(ii)(B). The protection of the health and safety of the public was not affected by this issue. The licensee has notified the NRC Resident Inspector of this event. AND Power Reactor Event Number: 50531 Facility: PRAIRIE ISLAND Region: 3 State: MN Unit: [1] UNANALYZED CONDITION DUE TO A MISSING FIRE BARRIER As part of a fire modeling analysis evaluating a missing fire barrier reported in ENS notification 50475 (a gap where the ventilation duct passes through the wall from the Bus 150/160 room to the Auxiliary Feedwater (AFW) pump room), an additional missing fire barrier was identified. The barrier is related to separation of redundant pressurizer heater banks credited for safe shutdown. For Fire Area 32 (AFW pump room), Group E Pressurizer Heaters are credited for safe shutdown because Group A and Group B Pressurizer Heater cables could be affected by a fire in this area. It was determined that a cable associated with the Group C, D, and E Pressurizer Heaters is routed in Fire Area 32. Therefore, a fire in Fire Area 32 could affect all five Pressurizer Heater Groups. An evaluation has previously demonstrated that Mode 3 Hot Standby could be maintained with no charging pumps or pressurizer heaters available, but it has not been determined if Mode 5 could be achieved. Therefore, this missing fire barrier meets the reporting criteria for 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. The protection of the health and safety of the public was not affected by this issue. Unit 1 is in mode 6 and no fire has occurred. Compensatory measures (fire watches) are in place for Fire Area 32. The licensee has notified the NRC Senior Resident Inspector. Unit 2 has a different cable routing and is not affected. Power Reactor Event Number: 50532 Facility: FITZPATRICK Region: 1 State: NY Unit: [1] HIGH PRESSURE CORE INJECTION DEGRADED ACCIDENT MITIGATION CAPABILITY During the plant response to the trip of the B Recirculating water pump, reactor water level rose to the HPCI [High Pressure Core Injection] high water level trip setpoint as indicated on the associated instrumentation. With this high water level trip actuated, the HPCI high drywell pressure initiation signal would not have allowed the HPCI system to perform its intended safety function if required. If the HPCI system received the low water level initiation signal, the system would have been able to perform Its intended safety function. This high water level signal was actuated from 1935 [EDT] until reset at 1940 [EDT]. This is reportable under 50.72(b)(3)(v). The licensee notified NRC Resident Inspector. Agreement State Event Number: 50505 Rep Org: TENNESSEE DIV OF RAD HEALTH Licensee: SIEMENS MOLECULAR IMAGING Region: 1 City: KNOXVILLE State: TN AGREEMENT STATE REPORT - PRELIMINARY REPORT ON A CONTAMINATED EMPLOYEE The following was received from the State of Tennessee via email: Not much information is known at present, but on September 30th, an employee at Siemens was involved in some type of exposure that resulted in a skin dose contamination. No other spread of contamination has been noted. REACT/s was notified via Siemens and is being used for additional consultation. Awaiting more information from the licensee on [the type of] isotopes involved, a full detailed report of the incident, and their plan of action for the employee. Will update accordingly. TN ID NO.: TN-14-186 * * * UPDATE ON 10/6/14 AT 1539 EDT FROM SABRINA ROBERTSON TO DONG PARK * * * The following report from the licensee was received from the State of Tennessee via email: Description of the Event On 9/30/2014 at approximately 2:35pm, [an employee] was drawing activity to manufacture line sources out of a stock vial of Ge-68. She drew up the required activity into a syringe and when she removed the syringe from the rubber septum of the Ge-68 DTPA stock solution vial, the vial evidently became pressurized and sprayed liquid on to her nasal facial area. Another employee in the area immediately notified the Radiation Safety Officer (RSO). Mitigation The Radiation Safety Department responded and assisted with decontamination of the affected areas. Decontamination was started approximately 15 minutes after notification of the incident. A water lavage of the facial area was performed and washing of the affected area with hand soap and shaving cream. Wet nasal swabs were used to remove contamination from the nasal passage. After decontamination it was determined that the contamination was confined to the area around her nose and mouth, including inside her nasal passage. Since that time she has collected all nasal excretions and any sputum, as well as wipes of her skin in the affected area. Radiation surveys are done several times per day to ascertain activity reduction. We are continuing to perform surveys, nasal swabs and urine bioassays until all contamination is removed. To date all urine bioassay samples have been within the range of background. REAC/TS was contacted on October 2, 2014 to determine if they had any further ideas for contamination removal. They recommended the use of make-up removal pads on the face, but did not have any further ideas for removal of contamination from the nasal passage. Dose Estimation Using the surveys and an initial estimated activity for the contamination event of 76 microCuries in the highest affected area, VARSKIN was used to calculate dose to the skin. Initial calculations estimated the dose to remain below regulatory limits. On October 2, 2014 the Health Physicist running the VARSKIN program noticed that it was not calculating the dose from both Ge-68 and its daughter Ga-68. The calculations were revised and the results exceeded the regulatory limits. The RSO notified the State of Tennessee Division of Radiological Health on October 3, 2014. As of October 3, the calculations showed that skin dose is approximately 132 rem, which exceeds the regulatory limit of 50 rem. [The employee] has been notified of her estimated dose levels and has been involved with decontamination procedures. We will continue to adjust the total dose daily as surveys are performed. We will submit the final dose number when all contamination is removed and calculations are completed. Our current estimated dose at 10 days from today with no further removal of contamination is 240 rem to the skin. Notified R1DO (Bower) and FSME Events Resource via email. * * * UPDATE FROM SABRINA ROBERTSON TO HOWIE CROUCH (VIA EMAIL) AT 0825 EDT ON 10/8/14 * * * The following update was received from the State of Tennessee via email: Summary of the long-term plan for monitoring the employee We will continue to monitor the employees nasal facial area daily, collect and count wipe samples from skin in the affected area and nasal secretions, collect and count urine bioassay samples for at least the next week (October 14, 2014) unless all measurements are within the range of background before that date. We will assess the value of continued measurements at that time. Contamination and projected 10-day calculated skin dose for Monday, October 6 and Tuesday, October 7, 2014. The two areas from which skin dose is being estimated are the bridge of the nose (#1) and the tip of the nose (#2). Measurements are taken using a pancake GM detector (Ludlum 44-9). Skin dose is calculated using VARSKIN. October 6 - contamination readings: #1 = 840 cpm; #2 = 4940; calculated skin dose @10 days = 229, 934 mrad October 7 - contamination readings: #1 = 740 cpm; #2 = 6040 cpm; calculated skin dose @ 10 days = 239, 070 mrad [Licensee staff] are trying to determine a more realistic efficiency for the Ludlum 44-9 GM detector, as we believe it is closer to that of P-32 than to F-18. The use of a higher efficiency will substantially reduce the calculated dose. As you suggested, we have contacted REAC/TS about conduct an independent verification of our skin dose calculation. They informed us they are not able to conduct independent verification. However, they did provide us with consultants who are able to provide this service. Finally, as we discussed [with the State of Tennessee], Siemens will provide you with a daily update of the contamination readings and 10-day projected skin dose. Notified R1DO (Bower) and FSME Events Resource email. And other reportable events.... nrc.gov/reading-rm/doc-collections/event-status/event/en.html
Posted on: Tue, 14 Oct 2014 10:32:13 +0000