Event Notification Report for November 25, 2014

NON-CONSERVATISMS IN METHODOLOGY USED FOR EMERGENCY CORE COOLING SYSTEM PERFORMANCE REQUIREMENTS

"On Tuesday, November 25, 2014, at 1200 EST, the FirstEnergy Nuclear Operating Company (FENOC) reviewed AREVA 10CFR50.46 Notification Letter FAB14-00625 for the Davis-Besse Nuclear Power Station (DBNPS). This letter indicates that certain non-conservatisms were discovered in the methodology application and inputs used by AREVA for nuclear fuel core configurations with Mark-B-HTP fuel when operated under certain conditions. When corrected, this increases the Peak Cladding Temperature (PCT) in excess of the value prescribed in 10CFR50.46(b)(1) under Loss of Coolant Accident (LOCA) conditions. The DBNPS reactor core contains Mark-B-HTP fuel. 10CFR50.46 paragraph (b) defines the acceptance criteria for the LOCA analysis process. The DBNPS licensing basis PCT is evaluated for compliance with and must not exceed the criterion prescribed in 10CFR50.46(b)(1).

"AREVA had provided compensatory measures in terms of plant axial imbalance limits and Fq linear heat rate limits associated with reductions in LOCA linear heat rates so that the DBNPS operates within 50.46 limits. FENOC implemented the compensatory measures at the DBNPS on October 23, 2014, per AREVA recommendations, and as a result the errors reported have no impact on current plant operation or public health and safety. Preliminary analysis of past operating conditions indicate that the DBNPS did not exceed the 50.46(b)(1) criteria for PCT. This 8-hour notification is being reported in accordance with 10CFR50.72(b)(3)(ii)(B).

"Based on 50.46(a)(3)(ii) criteria, FENOC will submit a report within 30 days for the DBNPS.

"FENOC has notified the DBNPS NRC Senior Resident Inspector."

DEFICIENCY IN METHODOLOGY USED FOR EMERGENCY CORE COOLING SYSTEM PERFORMANCE REQUIREMENTS

"On Tuesday, November 25, 2014, at 1212 EST, Duke Energy Carolinas, LLC (Duke Energy) reviewed AREVA 10CFR50.46 Notification Letter FAB 14-00631. This letter indicates that a deficiency was discovered in the uranium thermal conductivity models used in the Oconee Nuclear Station Loss of Coolant Accident (LOCA) analysis of record. When the deficiency is corrected, the LOCA Peak Cladding Temperature (PCT) limits may be in excess of 2200 degrees F. 10CFR50.46 paragraph (b) defines the acceptance criteria for the LOCA analysis process. The Oconee licensing basis PCT is evaluated for compliance with the criterion 10CFR50.46(b)(1) and must not exceed a PCT of 2200 degrees F.

"On October 20, 2014, AREVA recommended actions in the form of reductions in LOCA linear heat rates, which were then translated into reduced axial imbalance limits for the excore and backup incore detector systems. This was done to ensure that Duke Energy operated within 10CFR50.46 limits in the event of a loss of the full incore detector system. The full incore detector system is the primary method for evaluating imbalance and the imbalance limits are unaffected by the reduction in LOCA linear heat rates. The full incore detector system is operable and meets Technical Specification 3.2.2. In addition, the Reactor Protection System (RPS) trip limits for imbalance are not derived from the LOCA analysis and are not affected.

"When AREVA notified Duke Energy of the deficiency, Duke Energy confirmed that existing administrative limits bound the AREVA recommended actions and as a result, the errors reported have no impact on plant operation or public health and safety.

"This event affects all three (3) units and is being reported in accordance with 10CFR50.72(b)(3)(ii)(B).

"Based on 10CFR50.46(a)(3)(ii) criteria, Duke Energy will submit a written report within 30 days.

"Duke Energy has notified the NRC Senior Resident Inspector."

DEFICIENCY IN METHODOLOGY USED FOR EMERGENCY CORE COOLING SYSTEM PERFORMANCE REQUIREMENTS

"On Tuesday November 25, 2014, at 1211 CST, Arkansas Nuclear One, Unit 1 (ANO-1) reviewed AREVA 10 CFR 50.46 Notification Letter FAB14-00632. This letter indicates that a deficiency was discovered in the uranium thermal conductivity models used in the ANO-1 Loss of Coolant Accident (LOCA) analysis of record. When the deficiency is corrected, the LOCA Peak Cladding Temperature (PCT) limits may be in excess of 2200 degrees Fahrenheit (F). 10 CFR 50.46 paragraph (b) defines the acceptance criteria for the LOCA analysis process. The ANO-1 licensing basis PCT is evaluated for compliance with the criterion 10 CFR 50.46(b)(1) and must not exceed a PCT of 2200 degrees F.

"During AREVA's review of the issue, AREVA had provided compensatory measures in the form of reductions in LOCA linear heat rates as a contingency in case the errors were found to be substantiated, which were then translated into reduced axial imbalance limits so that ANO-1 would operate within 10 CFR 50.46 limits. As a precautionary measure pending the completed analysis, ANO-1 implemented the compensatory measures on October 20, 2014, and as a result, the errors reported have no impact on current plant operation or public health and safety. This event is being conservatively reported in accordance with 10 CFR 50.72(b)(3)(ii)(B).

"Based on 50.46(a)(3)(ii) criteria, ANO-1 will submit a written report within 30 days.

"ANO-1 has notified the NRC Senior Resident Inspector."

REACTOR CONSOLE KEY LEFT IN CONTROL CONSOLE UNATTENDED

"On Tuesday, November 25, 2014, a violation of Technical Specification A.1, 'Shutdown,' occurred at the USGS TRIGA research reactor. Contrary to the specification, the key to the console key switch was left in the control console, in the OFF position, while the control room was unoccupied. This condition existed intermittently for approximately nine minutes during the time period from 1421 MST to 1436 MST, when the Reactor Supervisor discovered the key. The reactor was shutdown during this time period, and the staff members were performing routine maintenance. The facility shutdown checklist had been completed, and the key had been locked in a lock box, but the key was subsequently removed from the lock box by a Senior Reactor Operator and put in the console switch for maintenance checks. The reactor was continuously occupied by reactor staff during this event, and there were no non-staff personnel present. This event is being reported to the NRC's facility inspector, NRC Operations Center, the USGS Reactor Administrator, and the USGS Reactor Operations Committee."