Event Notification Report for November 05, 2021

EN Revision Imported Date: 11/5/2021

EN Revision Text: HIGH PRESSURE COOLANT INJECTION INOPERABLE DUE TO LOW LEVEL INSTRUMENTATION

"At 0241 [EDT] on June 3, 2021, during performance of a High Pressure Coolant Injection (HPCI) Condensate Storage Tank (CST) level functional surveillance, the HPCI torus suction inboard isolation valve was slow to open during swap of suction from the CST to the Torus. On June 9, 2021, it was determined that as a result of the June 3, 2021, slow swap condition, TS 3.3.5.1 Required Action D.1 to declare HPCI inoperable within 1 hour was applicable due to inoperable CST low level instrumentation channels. At 1817 [EDT] on June 3, 2021, HPCI suction was swapped to the torus, making TS Required Action D.1 no longer applicable. Reactor Core Isolation Cooling (RCIC) was available throughout this condition.

"At 0900 [EDT] on November 3, 2021, it was determined that an NRC event report due to HPCI inoperability should have been made. This event is being reported as a late 8-hour non-emergency notification pursuant to 10CFR50.72(b)(3)(v)(D) based on an unplanned HPCI inoperability.

"The cause of the slow valve opening was later determined to be corrosion products on contacts of a relay in the CST low level instrumentation logic. On June 4, 2021 at 1451 [EDT], the HPCI CST Level Functional Test was completed Satisfactorily, restoring HPCI Instrumentation to Operable.

"There was no impact on the health and safety of the public or plant personnel.

"The NRC Resident Inspector has been notified."

EN Revision Imported Date: 11/5/2021

EN Revision Text: INCORRECT LOCAL LEAK RATE

"On 11/03/2021 Columbia Generating Station concluded the results for refueling outage 24 (R24) and 25 (R25) Local Leak Rate Testing as-found data was incorrect.

At 1231 PDT on November 3rd, 2021 Columbia Generating Station determined the local leak rate tests (LLRT) for the X- 25B containment penetration did not meet Technical Specification requirements for LLRT acceptance criteria. The incorrect LLRT data identified for residual heat removal (RHR) B Suppression Pool Spray containment isolation valve (RHR-V-27B) was from the previous two refueling outages (R24 on 5/22/2019 & R25 on 6/512021) at which time primary containment was not required to be operable. The corrected leakage assigned to the X-25 penetration also resulted in total Type B and C leakage summation exceeding the maximum allowable leakage rate for the primary containment (1.0 La) for R24 and exceeding 0.6La in R25.

The valve was flushed and retested satisfactory prior to entering the mode of applicability. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A).

There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified."

EN Revision Imported Date: 11/8/2021

EN Revision Text: FITNESS-FOR-DUTY REPORT

A non-licensed employee had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the plant has been terminated.

The NRC Resident Inspector has been notified.

DOSE TO AN EMBRYO

The following is a summary of a phone call with Community Health Network:

A patient came in to receive a radiotherapy treatment of 148 mCi I-131 on September 30, 2021. The patient received a negative pregnancy test prior to the treatment. At a later date, the patient realized that she was pregnant and estimated the date of conception to be either September 29, 2021 or October 10, 2021. The radiation safety officer estimated the dose to the embryo to be 394 mSv (39.4 rem) based on a conception date of September 29. The estimated dose to the embryo would be 1.7 mSv (0.17 rem) based on a conception date of October 10. The licensee is reporting this event based on the conception date of September 29 because of the potential for higher dose to the embryo.

LOSS OF HIGH RADIATION AUDIBLE ALARM

"On November 5th, 2021, [1207 MDT] it was identified that a portable Area Radiation Monitor (ARM), which was performing the 10 CFR 70.24 safety function of the Criticality Accident Alarm System (CAAS) to energize clearly audible alarm signals if accidental criticality occurs, had been removed from the area in error.

"On July 20th, 2021, during routine CAAS maintenance, UUSA [Urenco USA] staff identified an area in which a CAAS alarm was not clearly audible. UUSA arranged the ARM as a compensatory measure which achieves an equivalent 10 CFR 70.24 safety function in the affected area. UUSA reported this event to the NRC under Event Notification 55480 in accordance with 10 CFR 70.50(b)(2) in which equipment is disabled or fails to function as designed when required by regulation.

"Removal of this ARM resulted in an inability for radiation detectors to energize clearly audible alarm signals if accidental criticality occurs in the affected area. UUSA hereby reports this event in accordance with 10 CFR 70.50(b)(2), in which equipment is disabled or fails to function as designed when required by regulation (10 CFR 70.24)."

The licensee will notify the NRC Region 2 office.

STEAM SUPPLY CHECK FAILURE TO TURBINE DRIVEN AUXILLARY FEEDWATER

"During a Unit 2 refueling outage valve overhaul activity on the steam supply check valve from the number 2 steam generator to the turbine driven auxiliary feedwater pump, 2-MS-4B, the check valve was found with its disc separated from the disc arm. This failure would have resulted in the blowdown of both steam generators during a main steam line break in the steam generator number 2 main steam line upstream of the main steam isolation valves until the operators could isolate the faulted steam generator. On November 6, at approximately 1100 EDT evaluation determined that this condition constituted an unanalyzed condition that could impact containment pressure.

"There has been no radioactive release to the environment. The valve has been repaired. The check valve in the steam supply from the number 1 steam generator to the turbine driven auxiliary feedwater pump was inspected and found to be satisfactory. There has been no impact to Unit 3 which remains at 100% power.

"The Senior Resident has been notified.

"This condition is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(B) as a condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety."