Event Notification Report for March 05, 2024

AGREEMENT STATE REPORT - FAILED SOURCE LEAK TEST

The following is a summary of information received from the Georgia Radioactive Material Program (the Department) via email:

On January 6, 2024, the Department was notified by the licensee, that on January 3, 2024, a nuclear medicine technologist was performing routine leak testing of sealed sources in preparation of returning the sources to the manufacturer. The leak test indicated one of the sealed sources (Co-57, 21.12 microcuries as of 2/1/2023, manufacturer: Eckert and Ziegler, model: PHI-0124, serial number: V6-599) had more than 0.005 microcuries of removable Co-57 contamination. The sealed source was secured, the radiation safety officer (RSO) was notified, and decontamination protocol was followed.

Post-decontamination surveys and wipe tests of the staff and department indicated that there was no detectable contamination in the department or on staff members. The sealed source and the waste generated during the decontamination process were placed in leakproof containers and marked as containing Co-57. All items are currently stored in the nuclear medicine hot lab. Disposal with a waste disposal company has been arranged.

Georgia NMED event number: 76

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

AUTOMATIC REACTOR TRIP DUE TO LOSS OF MAIN FEEDWATER PUMP SUCTION

The following information was provided by the licensee via email:

"At 1142 CST on 3/3/2024, with Unit 2 in Mode 1 at 29 percent power, the reactor automatically tripped due to a turbine trip caused by a loss of suction to the 22 main feedwater pump. All systems responded normally post trip. Decay heat is being removed via the auxiliary feedwater water system. Secondary steam control mechanism is the steam generator PORVs [power operated relief valves]. Unit 1 remains at 100 percent power and is unaffected.

"This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The resident NRC inspector has been notified."

The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance:

The trip occurred while the licensee was returning to power operations after a refueling outage. During the trip, all rods inserted into the core. The plant is in a normal shutdown electrical lineup with offsite power available. The plant will be maintained at normal operating temperature and pressure. There is no known primary to secondary leakage. The cause of the loss of 22 main feedwater pump suction is under investigation.

AUTOMATIC REACTOR SCRAM DUE TO MAIN TURBINE TRIP ON LOW CONDENSER VACUUM

The following information was provided by the licensee via email:
"On 3/3/24 at 1942 EST, while performing a plant shutdown in preparation for a refuel outage, Nine Mile Point Unit 2 experienced a reactor scram due to a main turbine trip on low condenser vacuum. The plant was at approximately 55 percent power at the time of the reactor scram.

"Additionally, following the scram a low RPV [reactor pressure vessel] level scram and containment isolation signal on level 3 was received, as expected. The containment isolation signal impacted RHR [residual heat removal] shutdown cooling, RHR letdown to radwaste, and RHR sampling. All impacted valves were closed at the time the isolation occurred.

"All control rods were fully inserted. Plant response was as expected. Post scram, the main turbine bypass valves are being used to control decay heat, and normal post scram level control is via the feed / condensate system.

"This is being report under 10 CFR 50.72(b)(2)(iv)(B), 'RPS Actuation', and 10 CFR 50.72(b)(3)(iv)(A), 'Specified System Actuation'.

"Unit 1 is not affected. There was no impact on the health and safety of the public or plant personnel.

"The NRC Resident Inspector has been notified."

The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance:

The cause of the low condenser vacuum was a momentary loss of sealing steam. The condenser remained viable for decay heat removal. All safety equipment is available. The grid is stable with the plant in its normal shutdown electrical configuration.