Event Notification Report for February 15, 2000

THREE MILE ISLAND UNIT 2 SPENT FUEL STORAGE CONFINEMENT SYSTEM IN USE HAS A POTENTIAL FOR A SIGNIFICANT REDUCTION IN EFFECTIVENESS. THIS IS NOT A SAFETY ISSUE.


At 1325 MST, 15 FEB 2000, a determination was made that a spent fuel storage confinement system in use has a potential for a significant reduction in effectiveness.

This conclusion is based on an engineering evaluation of plugs in the Three Mile Island Unit 2 (TMI-2) core debris Canisters. These plugs are approximately 2" diameter and there are two plugs in each TMI-2 Canister. There are eight TMI-2 Canisters in the Dry Storage Canister (DSC) currently in storage at the TMI-2 ISFSI. The TMI-2 ISFSI Safety Analysis Report takes credit for the plugs being in place for the design function of the confinement of radioactive material.

The DSC containing the eight TMI-2 Canisters continues to provide an intact outer confinement boundary.

The condition described above exists at the TMI-2 ISFSI at the Idaho National Engineering and Environmental Laboratory.

The affected DSC contains eight TMI-2 Canisters containing TMI-2 core debris. The U-235 inventory is conservatively estimated to be significantly less than 1 kilogram for the entire DSC.

There are no potential for personnel radiation exposure associated with this condition. There is no potential for release of radioactive material from the DSC associated with this condition.

NRC Region IV has been notified of this event.

REACTOR MANUALLY TRIPPED DUE TO INDICATION OF A STEAM GENERATOR TUBE RUPTURE ON STEAM GENERATOR # 24.

The licensee was at 99% power when they received indication of a steam generator tube rupture on steam generator #24. The steam generator tube rupture was initially estimated to be approximately 90 gallons per minute. At 1929, the reactor was manually tripped, and all rods fully inserted into the core. The faulted steam generator was isolated. Currently, reactor coolant system pressure is 2243 psig, temperature is 547 °F and pressurizer level is being maintained at 35%. The licensee declared an Alert classification at 1929 per EAL 3.1.2 (large amount of water out of the reactor coolant system). The NRC entered Standby at 2049 EST.

The NRC response was deescalated to Monitoring at 2335 EST.

The NRC Resident Inspector was notified of this event by the licensee.

* * * UPDATE AT 1000 @2/16/00 * * * COMMISSIONER'S ASSISTANTS BRIEFING GIVEN BY RANDY BLOUGH ON THE INDIAN PT 2 ALERT.

The following persons participated in the briefing: Region 1 RA (Hub Miller), R1 PAO(Diane Screnci), R1 (Jennifer England), RI (Pete Habighorst), R1DO (F. Costello), IRO (F. Congel), EDO (Bill Travers), DEDR (Frank Miraglia), NRR (Sam Collins), EDO Staff (Mike Tschiltz), Chairman's Office (Terence Chan), Comm. Dicus' (Thomas Hiltz), Comm. McGaffigan's (Jeff Sharkey), Comm. Diaz's (Roger Davis), Comm. Merrifield's (Brain McCabe), NRR PD (Jim Clifford), PM (Jeff Harold), NRR EO (Tad Marsh), PA (Sue Gagner), State Programs (Paul Lohaus), IP (Howard Faulkner), and CAO (Tom Madden). DEDMRS Staff and SECY did not participate.

* * * UPDATE AT 1922 @ 2/16/00 BY SPALL TAKEN BY MACKINNON * * *

As of 1657 hours, Unit 2 attained a cold shutdown condition. Leakage from the reactor coolant system into # 24 steam generator has stopped. Indian Point 2 is terminating the alert classification as of 1850 hours and will be entering the Recovery Phase.


At 1915 hours NRC exited Monitoring mode and returned to normal operations.

NRC Resident Inspectors notified.

R1DO (F Costello), NRR EO(D Wessman) and IRO (Joe Giitter) notified.

FEMA (Eaches), DOE (Bob Young), USDA (Corley), HHS (Pamela), and EPA (Garneau) notified that Indian Point Unit 2 exited their Alert classification at 1850 hour and entered Recovery Phase and were also notified that the Nuclear Regulatory Commission exited Monitoring Mode at 1915 hours.

UNIDENTIFIED LEAKAGE 1.1 GALLONS PER MINUTE

"Elevated reading noted on Unit 2 Plant Vent Radiation Monitor (2R41D) concurrent with slowly lowering Volume Control Tank (VCT) level. Plant was operating at 100% reactor power. Investigation by on-shift personnel estimated unidentified RCS leakage at 1.1 gpm. This value exceeds the Tech Spec limit of 1 gpm unidentified (Reference Tech Spec 3.4.7.2). The Unit 2 Reactor Coolant Filter had been replaced on dayshift. Operators isolated the Reactor Coolant Filter at 2010 and the leakage appears to be stopped. This report is being made for RCS leakage outside containment which may result in exceeding GDC-19 criteria for Control Room habitability."

The leak appears to have started at 1638 EST and was stopped at 2010 EST. The volume that was lost is estimated at 160 gallons, the leak was through a hard pipe directly to a radiation waste tank.

The licensee notified the NRC Resident Inspector.

* * * RETRACTION ON 02/16/00 AT 1647ET BY S. SAUER TAKEN BY MACKINNON * * *

Further review of the above event report for reactor coolant leakage outside containment, which may have resulted in exceeding GDC-19 criteria for control room habitability has identified that this leakage flow path is not part of the post accident recirculation flow path in communication with the containment sump. Leakage from systems that are isolated from the recirculation flow path or that can be isolated from it, without degrading the ability to recirculate the reactor coolant system are not included in leakage requirement of the USFAR.
R1DO (F. Costello) notified.

The NRC Resident inspector was notified of this event retraction by the licensee.

FLAWED WELD DISCOVERED DURING PLANT OUTAGE

"At 2058 CST on 2/15/2000, a flawed weld was identified on an instrument connection to the Reactor Coolant System Loop 'A' Hot Leg Piping. The subject flaw was identified by Non-Destructive Examination (NDE) at the connection of Level Instrumentation to the RCS Hot Leg Piping. Unit 1 was taken offline at 0010 on 2/5/2000 for scheduled maintenance. Boron buildup was identified on 2/15/00 at the subject connection while the plant was at cold shutdown condition. The accumulated boron indicates the leakage through the weld flaw is classified as very small weepage. The unidentified RCS leak rate prior to thc scheduled outage was 0.093 GPM. No unexplainable changes in RCS leak rate exist for cycle 16 operation. Due to the physical location of the connection, major scaffolding was required to access the area. At 2058 on 2/15/2000, NDE was complete identifying the reportable condition. Further investigation by Plant Staff is in-progress to characterize the flaw morphology and cause. This report is being performed to comply with the requirements of 10CFR50.72(b)(2)(i)."

The NRC resident inspector has been informed of this event by the licensee.

* * * UPDATE AT 1352 ON 2/18/2000, BY DAVENPORT RECEIVED BY WEAVER * * *

Unit 1 was taken off-line at 0010 CST on February 5, 2000, for scheduled maintenance and is currently in a cold shutdown condition. Boron buildup was identified at 5 additional level instrumentation connections on the RCS A and B Hot Leg piping. The accumulated boron indicates the leakage through the weld flaws is very small weepage. Due to the physical locations of the connections, major scaffolding was required to access the areas. NDE was performed to confirm the initial nozzle leak found on 2/15/2000. Due to radiological conditions, as-found NDE will not be performed on the remaining leaking nozzles as planned. The visual inspections already performed are sufficient to determine past leakage. None of the identified flaws are actively leaking with the reactor in cold shutdown. The unidentified RCS leak rate prior to the scheduled outage was 0.093 gpm. No unexplainable changes in the RCS leak rate existed for the Cycle 16 operation. Further investigation by Plant Staff is in-progress to characterize the flaw morphology and cause.

The licensee notified the NRC resident inspector, the Operations Center notified the R4DO (Tapia).