Event Notification Report for May 16, 2001

OFFSITE NOTIFICATION/INFORMATION REGARDING A RECYCLE FACILITY SHIPMENT CONTAINING RADIOACTIVE MATERIALS

The following text is a portion of a facsimile received from the licensee:

"Indian Point is making the following notification pursuant to 10CFR50.72(b)(2)(xi) and in accordance with Station Administrative Order 124, Item 47."

"This report is being made pursuant to 10CFR50.72(b)(2)(xi), 'Any event or situation, related to the health and safety of the public or on-site personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made.'"

"At approximately 1230 hours today[,] IP-2 was notified by KARTA Industries, a recycle facility, that a shipment made by KARTA to Ohio was returned because it contained radioactive material. KARTA plans to notify the New York State Department of Environmental Conservation (DEC) of this event. This report is being made for information only since the waste does not appear to be associated with Indian Point operations. When notified by KARTA, IP-2 dispatched radiation protection personnel to the KARTA facility in Peekskill, NY. By 1330 [hours,] IP-2 RP personnel surveyed the returned bale and determined it showed [approximately] 2.5 mrem/hr on contact. Since IP-2 has expertise in radiation protection[,] it was decided to bring the bale to the IP-2 site to ensure proper control and disposal. Isotopic analysis of a sample of the bale material indicates mostly I-131[,] which indicates a medical treatment, not material from a nuclear power plant. KARTA does handle recyclable material from hospitals in the Peekskill vicinity. IP-2 will assay the entire bale and ensure proper disposal. IP-2 has been informed by KARTA that KARTA plans to notify the New York State Department of Environmental Conservation (DEC)."

The licensee plans to notify the NRC resident inspector.

DISCOVERY THAT THE CONTROL ROOM HABITABILITY SYSTEM MAY NOT MEET DESIGN REQUIREMENTS UNDER CERTAIN CONDITIONS

The following text is a portion of a facsimile received from the licensee:

"The control room habitability system may not meet design requirements under certain conditions. Specifically, any design accident where the control room bottled air pressurization system actuates and the exhaust fan of the cable vault and tunnel area trips with the supply fan to the cable vault and tunnel continuing to run, the minimum delta pressure between these areas may not be maintained."

"The control room bottled air systems of both Unit 1 and 2 have been declared inoperable, and a 7-day [technical specification action statement has] been entered."

The licensee plans to notify the NRC resident inspector.

* * * RETRACTED AT 1141 EDT ON 6/6/01 BY JAMES CROSSMAN TO FANGIE JONES * * *

After subsequent engineering evaluation, the licensee has determined that this event was not reportable and is retracting the event notification. The evaluation concluded that the effect of a failure of a non-safety related component on the control room envelope and bottle air system is not required during a design basis accident. Therefore, the proposed (not actual) failure assumptions are not considered reasonable and there is no need to report this condition.

The licensee notified the NRC Resident Inspector. The R2DO (Ken Barr) was notified.

REACTOR WATER CLEANUP SYSTEM ISOLATION DURING APPROACH TO CRITICALITY AND MANUAL REACTOR SCRAM

The following text is a portion of a facsimile received from the licensee:

"During approach to criticality[,] the reactor water cleanup (RWCU) system isolated on high temperature out of the non-[regenerative] heat exchanger. The cause of the high temperature was due to air entering the system from an ongoing local leak rate test. The RWCU system was providing the letdown path to control [reactor] water level from control rod drive cooling [water] flow. The control room was unable to recover the RWCU system and made the decision to manually scram the [reactor]. A manual actuation of [the reactor protection system] was inserted by placing the [reactor] mode switch to shutdown. All rods [fully] inserted[,] and all systems responded as expected."

The NRC resident inspector was in the control room at the time of the event.

NRC BULLETIN 91-01 RESPONSE - CRITICALITY CONTROL - (4-Hour Report)

During the revision to NCSE 052 and 039, it was discovered that the condenser could potentially be pressurized greater than 35.5 psia when the condenser supply and return valve are both closed without a fluorinating environment in the process gas system. NCSA CAS-002/011 allow both the supply and return valves to be closed when draining the condenser without a fluorinating environment. NCSEs 052 and 039 rely on the condenser supply valve to be closed enough to prevent the condenser from being pressurized above 35.5 psia during the time the return valve is closed and maintenance personnel relieve the pressure on the condenser. Based on recent discussions, it was discovered that closure of the supply valve may not provide sufficient isolation to prevent the condenser from being pressurized above 35.5 psia with the return valve closed because of potential seat leakage of the supply valve.

The licensee has stopped all maintenance activities which could create this condition. Also, samples will be taken where this maintenance has occurred to check that the freon still complies with the water content limits.

Paducah personnel notified the NRC resident inspector.

***** UPDATE RECEIVED AT 1500 EDT ON 06/21/01 FROM CALVIN PITTMAN TO LEIGH TROCINE *****

The following text is a portion of a facsimile received from Paducah personnel:

"At 1900 [CDT] on [05/16/01], the Plant Shift Superintendent (PSS) was notified of a deficient Nuclear Criticality Safety Evaluation. NCSAs CAS 002 and CAS 011 allow both the Recirculating Cooling Water (RCW) supply and return valves to be closed when draining the R-114 condenser without a fluorinating environment. The NCS evaluations rely on the RCW supply valve to be closed enough to prevent the condenser from being pressurized above 35.5 psia during the time the RCW return valve is closed and maintenance relieves the pressure on the condenser. Based on recent discussions between NCS and maintenance personnel, it was discovered that closure of the RCW supply valve might not provide sufficient isolation to prevent the condenser from being pressurized above 35.5 psia with the return valve closed. The RCW system provides cooling water to the R-114 condenser, which removes heat from the freon system. The freon is used to remove the heat of compression from the process gas system. The R-114 pressure is maintained above the RCW pressure to form a barrier between the process gas and the RCW cooling water, thus preventing moderation from occurring should a leak in the condenser occur. If the RCW pressure is allowed to rise above 35.5 psia, the potential exists that the RCW pressure could exceed that of the R-114 pressure. At this point, an actual condition of RCW pressure greater than 35.5 psia has not been identified."

"UPDATE [on 06/20/01] ATRC 01-3248: During the Engineering review to determine permanent corrective actions, additional immediate corrective actions have become necessary. It has been determined that when the condenser return valve is open, condenser pressure can also be affected when the return header alignment is changed (i.e., closure of RCW loop isolation valves, RCW building header valves, cooling tower riser valves, etc.). Controls have been implemented to ensure that changes to the RCW system do not impact our ability to maintain R-114 pressure above RCW pressure."

"SAFETY SIGNIFICANCE OF EVENTS: The controls credited for the isolation of the condenser to ensure the pressure at the condenser does not exceed 35.5 psia do not meet the intent of the NCSE for isolation. Therefore, the NCSE analysis is deficient and double contingency is not maintained. However, the probability for a criticality is unlikely due to the number of conditions that must be met in order for a criticality to be possible."

"POTENTIAL CRITICALITY PATHWAYS INVOLVED (BRIEF SCENARIO(S) OF HOW CRITICALITY COULD OCCUR): In order for a criticality to be possible, the following conditions must exist. With the condenser supply and return valves closed, the supply or return valve must be leaking to allow the condenser pressure to exceed the minimum coolant pressure of 35.5 psia. With the configuration of the supply valve closed and the return valve opened, the RCW return system must be changed (i.e. closure of loop isolation valves, RCW building header valves, cooling tower riser valves, etc.) enough that the affected condenser pressure increased greater than 35.5 psia. The process gas equipment must be at a UF6 negative and contain a UO2F2 deposit greater than a critical mass. The condenser must have a leak of sufficient rate and duration to allow enough water to overcome the down comer allowing liquid water to enter the cooler. The cooler must also have a simultaneous leak, which would allow wet coolant to leak into the process gas side of the equipment at a location, which would allow moderation of the deposit."

"CONTROLLED PARAMETERS (MASS, MODERATION, GEOMETRY, CONCENTRATION, ETC.): Double contingency for this scenario is established by implementing two controls for moderation."

"ESTIMATED AMOUNT, ENRICHMENT, FORM OF LICENSED MATERIAL (INCLUDE PROCESS LIMIT AND % WORST CASE CRITICAL MASS): Maximum assay of 5.5 wt. % U235"

"NUCLEAR CRITICALITY SAFETY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION OF THE FAILURES OR DEFICIENCIES: The first leg of [the] double contingency is based on isolation of the RCW condenser by closing the supply valve to maintain the condenser pressure less than 35.5 psia. Based on discussions with maintenance personnel, the current leak rates of the supply valve may be greater than the amount assumed to ensure the condenser pressure does not exceed 35.5 psia when the return valve is closed. Also, when the return valve is open, the RCW return system could be changed (i.e., closure of loop isolation valve, RCW building header valves, cooling tower riser valves, etc.). Therefore, the intent of this control has not been met, and this leg of double contingency was not maintained."

"The second leg of [the] double contingency is based on the independent verification of the isolation of the supply valve and when the return valve is required to be open, verification that the return valve is open. Since adequate isolation of the supply valve cannot be ensured based on verification of supply valve closure or return valve open, when required, the intent of this independent verification control was not met. Therefore, this leg of double contingency was not maintained."

"Since isolation of the RCW condenser with the supply valve closed and the return valve closed or opened may not meet the intent of the isolation requirement to maintain the condenser pressure less than 35.5 psia, these controls are deficient and double contingency has not been maintained."

"CORRECTIVE ACTIONS TO RESTORE SAFETY SYSTEMS AND WHEN EACH WAS IMPLEMENTED: Until the NCSEs and NCSAs can be revised and additional controls established:"

"Condenser Return Valve Closed"

"1. All maintenance activities related to closing the condenser return valve on a system, which is at a UF6 negative, have been stopped. This does not apply to systems that are not connected to the supply line or the C-310 TOPS Boosters."

"2. The coolant systems for condensers, which had the return valve closed without a fluorinating environment present, will be sampled in accordance with NCSA CAS-002 to verify the CFC-114 still complies with the water content limits."

"3. The return [valves] on condensers, which currently have the return valve closed, fluorinating environment removed (at a UF6 negative), and is connected to the supply line, will be opened."

"RCW Building Return System (based on ATRC-01-3248)"

"4. The plant power level will be maintained below 400 MW until NCSE changes are completed and/or necessary controls are implemented or specific evolutions are evaluated by NCS and approval is given."

"5. Both building RCW return header valves to a given header will not be closed."

"6. At least 3 cooling tower risers for every large pump and 2 risers for every small pump will be kept in service."

"7. Respective building RCW bypasses will be closed before isolating an RCW system loop."

"8. For any auxiliary systems not protected by a delta P alarm (i.e., A-310 Booster, A-335 Booster, C-337 B Booster) except the C-310 TOPS Booster, any activity that requires verification that the condenser return valve is open will also verify other valves in the return path to the applicable RCW return header are also open."

Paducah personnel notified the NRC resident inspector. The NRC operations officer notified the R3DO (Bruce Jorgensen) and NMSS EO (Patricia Holahan).]

* * * UPDATE 1705EDT ON 8/21/01 FROM CALVIN PITTMAN TO S. SANDIN * * *

The following information was provided as an update:

"The plant has completed and approved a USQD [Unreviewed Safety Question Determination] to allow the plant power level to be increased to 850 MW."

The NRC resident inspector has been informed. Notified R3DO(Gardner) and NMSS(Essig).

* * * UPDATE 1300 9/10/2001 FRIN BEASLEY TAKEN BY STRANSKY * * *

"NCSE 039/NCSA CAS-011 and NCSE 052/NCSA CAS-002 have been revised, approved, and implemented by the plant. These analyses are no longer considered deficient. The analysis incorporates additional controls associated with the cascade RCW system that are necessary for double contingency above a power level of 850 MW."

The NRC resident inspector has been informed. Notified R3DO (Burgess) & NMSS (Wayne Hodges)

SAFETY PARAMETER DISPLAY SYSTEM (SPDS) INOPERABLE

At 2259 hours on 5/16/01, the SPDS portion of the Emergency Response Facilities Computer System (ERFCS) was noted to be inoperable because the system time was not updating. The Technical Support Center (TSC) keyboard is not available nor is the Control Room Unit keyboard available for either Unit. Both Local Emergency Operations Facility (LEOF) and Central Emergency Operations Facility (CEOF) SPDS data links were unavailable. Initial indication of loss of ERFCS function was computer linkage failure at 2159 hours on 5/16/01. The ERFCS was rebooted satisfactorily with a new power supply. All portions of the system were noted to be functioning correctly at 0120 hours on 5/17/01.

The licensee notified the NRC resident inspector.

TURBINE TRIP FOR UNKNOWN REASONS AND SUBSEQUENT REACTOR SCRAM

The following text is a portion of a facsimile received from the licensee:

"Nine Mile Point Unit 2 experienced a turbine trip and reactor scram (Reactor Protection System (RPS) actuation) from approximately 90% power. All control rods fully inserted. Reactor water level lowered to below Level 3 (159.3 inches indicated, RPS actuation level) following the scram due to high power scram. Level recovered to above Level 8. (203.8 inches indicated, high level feed pump trip level). The operating feed pumps tripped as designed, and the 'A' [feed pump] was restarted. Level was restored to the normal band. Reactor pressure control continues to be controlled using the main condenser and a normal cooldown is in progress. Cause of the turbine trip is under investigation."

The licensee stated that there were no related maintenance or surveillance activities ongoing at the time of the turbine trip. None of the relief valves lifted, and all containment parameters are currently normal. All emergency core cooling systems (ECCS) are available, and no ECCS actuations were required during this event. Other than the reactor scram, there were no engineered safety features actuations, and none were required. There was nothing unusual or misunderstood except for the cause of the turbine trip, and all systems functioned as required in response to the turbine trip and subsequent reactor scram.

The licensee notified the NRC resident inspector and the State. The licensee also plans to issue a press release.