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 Start dateReporting criterionTitleEvent descriptionSystemLER
ENS 440376 March 2008 02:00:0010 CFR 50.72(b)(3)(ii)(A), Seriously DegradedCircumferential Piping Crack

On March 5, 2008, a volumetric ultrasonic examination of a Reactor Coolant to Decay Heat System pipe weld was analyzed and determined to identify an unacceptable indication. This dissimilar metal weld is located on the nozzle to the Decay Heat System drop line piping, which is the common suction line from the Reactor Coolant System and is a 12 inch outer diameter pipe. The indication is circumferential, is 15 inches in length, and reaches a maximum localized depth of 65 percent through-wall in one location. The weld was previously partially inspected during the November 2007 refueling outage using manual ultrasonic examination and no indications were identified. The current inspection was done using newly qualified phased array ultrasonic examination techniques in response to industry operating experience regarding dissimilar metal weld flaws. The indication has been found unacceptable per paragraph IWB 3514.4 of the 1989 Addenda of ASME Section XI without further fracture mechanics analysis and is therefore considered reportable. Preparations for a weld overlay repair and further confirmatory manual ultrasonic testing examinations are in progress. The licensee notified the NRC Resident Inspector.

  • * * UPDATE FROM J. TAYLOR TO JOE O'HARA AT 1730 ON 3/10/08 * * *

On March 8, 2008, further fracture mechanics evaluation of the circumferential indication on the Decay Heat System Drop Line determined that the requirements of the ASME Section Xl pipe code were maintained. Specifically, the acceptance criteria of the 1989 Edition Section XI, Table IWB-3641-1, -2, were met with an allowable value for flaw depth / wall thickness (a/t) of 0.75. Consequently, the 65% through wall indication would be considered acceptable for operation and the Degraded Condition Reporting Criteria would not be exceeded. Therefore, this event is not reportable under any 10CFR50.72 criterion. However, due to industry operating experience with dissimilar metals welds, this notification is being made voluntarily. Confirmatory manual ultrasonic testing examinations were completed which validated the presence of the indication originally found via phased array UT examination techniques. A full structural weld overlay repair is in progress which will be completed before returning Crystal River 3 to power operation. The repair effort has already successfully deposited the first weld layer over the location of the flaw. No Licensee Event Report will be submitted for this event. The licensee notified the NRC Resident Inspector. Notified R2DO(Hopper).

Reactor Coolant System
ENS 410248 September 2004 05:00:0010 CFR 50.72(b)(3)(ii)(A), Seriously DegradedPotential Reactor Coolant System Pressure Boundary Leak Located

At 0100 on September 8, 2004, while in Mode 3 conducting a Reactor Building walkdown after an unplanned reactor trip during Tropical Storm Frances, Crystal River Unit 3 identified a potential Reactor Coolant System pressure boundary leak located on a weld associated with a pressurizer level sensing line (upstream of reactor coolant isolation valve RCV-75). The cause of the leak is not known at this time, and the leak did not have a significant effect on plant operation. The last unidentified leak rates that were completed prior to the plant shutdown were 0.10 - 0.13 gpm. This defect in the primary coolant system is unacceptable per ASME Section XI. The condition is 8-Hour reportable per 10CFR 50.72 (b)(3)(ii)(A). The plant will be cooled down to Mode 5 and additional inspections and necessary repairs implemented. Dry boron crystal deposit are on the sensing line upstream of RCV-75 . See similar event reported by Crystal River Unit 3 on 10/04/03 (event # 40222) The NRC Resident Inspector was notified of this event by the licensee.


At 0311 on 9/8/04, Crystal River Unit 3 made an 8-Hour ENS notification in accordance with 10CFR50.72 (b)(3)(ii)(A) concerning a potential Reactor Coolant System pressure boundary leak (Event Notification # 41024). The leak was presumed to exist based on the presence of white, crystalline material deposited on the pipe. The material has been sampled and does NOT contain boron. The presumed leak site has been cleaned and visually inspected with no evidence of a defect in the weld joint or pipe. These additional inspections have demonstrated that the material deposited on the pipe does NOT constitute pressure boundary leakage, and therefore EN 41024 is retracted. The licensee has notified the NRC resident inspector.

Reactor Coolant System
ENS 402224 October 2003 14:32:0010 CFR 50.72(b)(3)(ii)(A), Seriously DegradedDegraded Condition Due to Rcs Pressure Boundary Leakage

At 1032 (EDT) on October 4, 2003 while in mode 4 conducting a planned refueling outage reactor coolant system inspection, Crystal River Unit 3 identified two potential RCS Pressure Boundary Leaks located in Pressurizer penetrations associated with upper level sensing lines. Upon closer inspection of the first one inch penetration, indications confirmed a Pressure Boundary Leak. A closer inspection of the second penetration is planned after obtaining access to the location. The last unidentified leak rate that was completed prior to the plant shutdown was 0.15 gpm. This defect in the primary coolant system is not acceptable per ASME XI. This condition is REPORTABLE per 10CFR 50.72 (b)(3)(ii)(A). Licensee reported that a small amount of boric acid residue and residual rust was identified during inspection in the vicinity of the upper level Pressurizer tap near the Pressurizer shell and the nozzle. A closer inspection is planned of the other two upper Pressurizer level taps.

The NRC resident has been notified by the Licensee.
Reactor Coolant System05000302/LER-2003-003