Event Notification Report for February 21, 2012

DATA PROCESSING NETWORK FAILURE IMPACTED EMERGENCY RESPONSE DATA SYSTEM

"At 0055 PST on 2/21/12, it was discovered that portions of the data processing network (eDNA) had failed. This failure resulted in the Emergency Response Data System (ERDS) entering a mode of operation that caused an interruption of the ability of the ERDS transmission capabilities. At approximately 0213 PST on 2/21/12, actions were taken and all transmission capabilities were restored. The total outage time was determined to be 78 minutes. During the period that ERDS was not available, backup communications were available,

"This is determined to be a partial, but major loss of emergency assessment capability, and is reportable under 10 CFR 50,72(b)(3)(xiii) as 'Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portions of control room indication, Emergency Notification System, or offsite notification system)'."

"The cause of the data processing network (eDNA) and ERDS failure is under investigation."

The licensee will notify the NRC Resident Inspector.

* * * RETRACTION FROM RICHARD GARCIA TO PETE SNYDER AT 1950 EDT ON 5/3/12 * * *

"This event was revisited for a formal reportability assessment and concluded to be not reportable since the capability to transmit data of interest to the NRC was available via the ENS at all times during the ERDS outage. In addition the ERDS system is noted as a supplement to the ENS in 10 CFR 50, Appendix B. Given the limited duration of the event and availability of ENS for data transmission, no major loss of emergency assessment capability occurred."

The licensee notified the NRC Resident Inspector.

TRITIUM IDENTIFIED ONSITE ABOVE THE VOLUNTARY REPORTING THRESHOLD

"On February 17, 2012, North Anna Power Station (NAPS) was notified by its vendor laboratory that a water sample, taken from an onsite ground water sample point, was confirmed to contain tritium above the voluntary reporting threshold of 20,000 picocuries per liter(pCi/L). The water sample, measuring 53,300 pCi/L, was obtained as a part of ongoing activities to determine the source of tritium previously reported to the state and NRC on October 29, 2010 (Event Notification - 46377). Current hydrological studies have determined the ground water in the area migrates to the station power block which is in the opposite direction from the lake. The ground water at the power block is collected in building subsurface drains and transported to a clarifier for processing. Clarifier discharge is accounted for as a monitored liquid effluent release pathway under the radiological effluent control program in accordance with the station's Offsite Dose Calculation Manual. As such, there is no increase to the projected annual dose to a member of the public. There are also no sources of drinking water in this area. Sampling of eight (8) ground water sample points outside the station protected area show no detectable levels of tritium confirming there is no migration offsite.

"The NRC Resident Inspector has been notified. A 30 day written report will be submitted to the NRC in accordance with NEI 07-07, Industry Ground Water Protection Initiative - Final Guidance Document."

The licensee will inform both state and local agencies.

MANUAL REACTOR TRIP DUE TO FEEDWATER HEATER HI HI ALARM

"During a normal shutdown in preparation for refueling outage 2R27, with Unit 2 at approximately 11.42% power, Unit 2 was manually tripped on 2/21/2012 at 2342 CST. The manual reactor trip was in response to a 21/22/23 Feedwater Heater Hi Hi alarm and was directed by the alarm response. Procedure 2E-0, 'Reactor Trip or Safety Injection,' was completed at 2345 CST. No Safety Injection was required. 2ES-0.1, 'Reactor Trip Recovery,' is in progress. Offsite power remains on all safeguards buses for both units. Unit 2 decay heat is via forced circulation and condenser steam dump with main feedwater providing flow to 21/22 steam generators. Auxiliary Feedwater start was not required and Unit 2 AFW remains in its safeguards alignment. No emergency event was declared as a result of this trip. Unit 1 remains at 100% power in Mode 1. Reportable actuations are: Unit 2 reactor protection (scram). The NRC Resident Inspector was notified. State [State of Minnesota] / local [Goodhue county] / Press release will be made. Other government agencies will not be notified. Nothing unusual / not understood. Unit 2 will continue to mode 5."

STEAM GENERATOR TUBE TO TUBE WEAR

The following information was received via email:

"Mitsubishi Heavy Industries, LTD (MHI) has identified steam generator tube wear for San Onofre Nuclear Generating Station.

"Steam Generator tube wear has been identified in areas of the U-bend between anti-vibration bars, where the tube is not supported by the anti-vibration bars. This tube wear occurred due to the contact of a tube with an adjacent tube, and resulted in a leak from a Unit 3 tube. The cause of the tube leak was determined to be the in-plane direction fluid elastic instability under high localized thermal-hydraulic conditions (steam quality (void fraction), flow velocity and hydro-dynamic pressure), and insufficient contact force of the tube to anti-vibration bar. This type Mitsubishi Heavy Industries has identified steam generator tube wear for San Onofre Nuclear Generating Station. The Steam Generator tube wear adjacent to the retainer bars was identified as creating a potential safety hazard.

"The plugging of the tubes that have the possibility of the fluid elastic instability and thermal power output reductions were identified as potential corrective actions. SCE [Southern California Edison] will run Unit 2 at 70% power for a short duration as a corrective action. In addition, SCE and MHI will continue a detailed analysis and investigation to this problem. Additional corrective actions may be required as the analysis and investigation continue. MHI has recommended to the purchaser that it perform eddy current inspection of tubes after Unit 2 operation resumes."

PART 21 INTERIM REPORT - STEAM GENERATOR TUBE WEAR

This interim Part 21 is in regard to San Onofre Nuclear Generating Station, Unit 2, Steam Generator replacement.

"During the first refueling outage following steam generator replacement, eddy current testing identified ten total tubes with depths of 90 to 28 percent of the tube wall thickness. Some of the affected tubes were located adjacent to retainer bars. The retainer bars are part of the floating anti-vibration bar (AVB) structure that stabilizes the u-bend region of the tubes.

"Other tubes in the two steam generators had detectable wear associated with support points elsewhere in the AVB structure. Each steam generator has 9727 tubes with an 8 percent (778 tubes) design margin for tube plugging.

"Discovery Date: February 13, 2012

"Evaluation completion schedule date: May 31, 2012"

"Those Mitsubishi Heavy Industries customers potentially affected by this issue have been notified and will receive a copy of this interim report."

Reference Document: UET-20120089
Interim Report No: U21-018-IR (0)

Notified R1DO (Joustra), R2DO (Nease), R3DO (Peterson), R4DO (O'Keefe), and Part 21 Group via email.

* * * UPDATE FROM MITSUBISHI HEAVY INDUSTRIES, LTD VIA FAX ON 6/4/12 AT 1145 EDT * * *

The vendor changed the number of tubes identified with wear depths of 90 to 28 percent from ten tubes to six tubes and only some of the tubes were adjacent to retainer bars.

Notified R1DO (Cahill), R2DO (Vias), R3DO (Passehl), R4DO (Gepford) and Part 21 Groups via email.

* * * UPDATE FROM MITSUBISHI HEAVY INDUSTRIES, LTD VIA FAX ON 9/7/12 AT 1539 EDT * * *

"MHI evaluated that the deviation contains a reportable defect for San Onofre Nuclear Generating Station Unit 2 and 3."

Notified R4DO (Gaddy) and Part 21 Reactor Group via email.