Event Notification Report for March 30, 2001
U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
03/29/2001 - 03/30/2001
** EVENT NUMBERS **
37870 37871 37872 37873 37874
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|Power Reactor |Event Number: 37870 |
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| FACILITY: WOLF CREEK REGION: 4 |NOTIFICATION DATE: 03/29/2001|
| UNIT: [1] [] [] STATE: KS |NOTIFICATION TIME: 07:55[EST]|
| RXTYPE: [1] W-4-LP |EVENT DATE: 03/29/2001|
+------------------------------------------------+EVENT TIME: 03:20[CST]|
| NRC NOTIFIED BY: JOE LARUE |LAST UPDATE DATE: 03/29/2001|
| HQ OPS OFFICER: FANGIE JONES +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |DAVE LOVELESS R4 |
|10 CFR SECTION: | |
|*COM 50.72(b)(3)(xiii) LOSS COMM/ASMT/RESPONSE| |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 100 Power Operation |100 Power Operation |
| | |
| | |
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EVENT TEXT
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| LOSS OF 6 EMERGENCY PLAN SIRENS |
| |
| The licensee had a problem with the 69 kV power grid in southern Coffey |
| County, which in turn caused the loss of 6 emergency plan sirens at 0220 CST |
| on 3/29/01. At 0320 CST, the criteria for making a 10 CFR 50.72 report on a |
| major loss of emergency communications was met. At 0432 CST the sirens were |
| restored to normal operation. |
| |
| The licensee notified the NRC Resident Inspector. |
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|Power Reactor |Event Number: 37871 |
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| FACILITY: MILLSTONE REGION: 1 |NOTIFICATION DATE: 03/29/2001|
| UNIT: [] [2] [] STATE: CT |NOTIFICATION TIME: 13:43[EST]|
| RXTYPE: [1] GE-3,[2] CE,[3] W-4-LP |EVENT DATE: 01/31/2001|
+------------------------------------------------+EVENT TIME: 15:59[EST]|
| NRC NOTIFIED BY: HUFF |LAST UPDATE DATE: 03/29/2001|
| HQ OPS OFFICER: CHAUNCEY GOULD +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |MICHELE EVANS R1 |
|10 CFR SECTION: |ED GOODWIN EO |
|*INV 50.73(a)(1) INVALID SPECIF SYSTEM A| |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
| | |
|2 N Y 100 Power Operation |100 Power Operation |
| | |
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EVENT TEXT
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| INVALID ACTUATION OF THE "A" EMERGENCY DIESEL GENERATOR |
| |
| On January 31, 2001 at 1559 while performing an air roll of the "A" |
| Emergency Diesel Generator (EDG) as part of the restoration of the EDG to an |
| operable status, the EDG started. This occurred because the fuel racks were |
| not properly tripped as required by the operating procedure. The operator |
| had attempted to trip the fuel racks, but did not hit the trip button with |
| sufficient force to trip the racks. As a result the EDG started and came up |
| to speed, but the EDG did not progress through the loading sequence since |
| there was no loss of power. Control Room personnel were immediately |
| notified and the EDG was properly shutdown by initiating an emergency trip |
| from the Control Room. |
| |
| The actuation of the EDG occurred due to human error and procedural |
| inadequacy and was not due to actual plant conditions. Therefore, the |
| actuation was not a valid actuation. The "A" EDG had been declared |
| inoperable under Technical Specification 3.8.1.1.b to perform a six hour |
| loaded run. 10CFR50.73(a)(2)(iv) excludes reporting an invalid actuation if |
| the equipment is properly taken out of service. However, the EDG was "not |
| properly taken out of service" at the time as described in NUREG-1022 to |
| qualify for the exclusion to reporting the invalid actuation. Although the |
| EDG had been declared inoperable, it was still "available". If an |
| Engineered Safeguards Actuation System (ESAS) signal had occurred, the EDG |
| would have been available to perform its function. In order to qualify for |
| the exclusion from reporting this invalid actuation, the undervoltage start |
| capability from ESAS would have to have been defeated, the flow of fuel to |
| the EDG stopped, or other appropriate action taken. Therefore, this event |
| is reportable as a 60-day ENS phone call. |
| |
| The NRC Resident Inspector was notified. |
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|Fuel Cycle Facility |Event Number: 37872 |
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| FACILITY: PORTSMOUTH GASEOUS DIFFUSION PLANT |NOTIFICATION DATE: 03/29/2001|
| RXTYPE: URANIUM ENRICHMENT FACILITY |NOTIFICATION TIME: 14:22[EST]|
| COMMENTS: 2 DEMOCRACY CENTER |EVENT DATE: 03/29/2001|
| 6903 ROCKLEDGE DRIVE |EVENT TIME: 10:45[EST]|
| BETHESDA, MD 20817 (301)564-3200 |LAST UPDATE DATE: 03/29/2001|
| CITY: PIKETON REGION: 3 +-----------------------------+
| COUNTY: PIKE STATE: OH |PERSON ORGANIZATION |
|LICENSE#: GDP-2 AGREEMENT: N |BRUCE BURGESS R3 |
| DOCKET: 0707002 |FRITZ STURZ NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: SPAETH | |
| HQ OPS OFFICER: CHAUNCEY GOULD | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|NONR OTHER UNSPEC REQMNT | |
| | |
| | |
| | |
| | |
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EVENT TEXT
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| HIGH LEVEL CONDENSATE PROBES ACTIVATED ON AUTOCLAVE #4 |
| |
| At 1045 hours on 03/29/01, both the "A" and "B" High Level Condensate probes |
| activated on Autoclave #4 in the X-344 UF6 Sampling Facility causing a High |
| Condensate Alarm. The activation of either of these probes will cause the |
| autoclave steam supply valves ( "Q" components) to actuate (close). The |
| steam supply valves closed, as designed, autoclave #4 was in TSR Mode II |
| (HEATING) at the time of the alarm activation. Operations personnel |
| immediately responded to the alarm in accordance with the approved Alarm |
| Response Procedures. The alarm reset when operations personnel drained |
| excess condensate from autoclave drain line, and then acknowledged the alarm |
| at the local autoclave control panel. All autoclave (and cylinder) |
| operating parameters were determined to be within normal ranges. Autoclave |
| #4 was placed in TSR mode VII (Shutdown) by operations personnel and |
| declared inoperable by the Plant Shift Superintendent. Based on the |
| actuation of both High Level Condensate probes, this event is being reported |
| as a valid safety system actuation, a 24 hour NRC event. An Engineering |
| Evaluation has been requested to investigate the circumstances surrounding |
| this safety actuation. |
| |
| The DOE Representative will be informed. Reg 3 (Monty Phillips) was |
| notified. |
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|Power Reactor |Event Number: 37873 |
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| FACILITY: PERRY REGION: 3 |NOTIFICATION DATE: 03/29/2001|
| UNIT: [1] [] [] STATE: OH |NOTIFICATION TIME: 14:42[EST]|
| RXTYPE: [1] GE-6 |EVENT DATE: 03/29/2001|
+------------------------------------------------+EVENT TIME: 11:00[EST]|
| NRC NOTIFIED BY: RABENOLD |LAST UPDATE DATE: 03/29/2001|
| HQ OPS OFFICER: CHAUNCEY GOULD +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |BRUCE BURGESS R3 |
|10 CFR SECTION: | |
|*DEG 50.72(b)(3)(ii)(A) DEGRAD COND DURING OP | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 84 Power Operation |84 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
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| SECONDARY CONTAINMENT BYPASS LEAKAGE FAILED TO MEET ACCEPTANCE CRITERIA |
| |
| During the reportability review of Condition Report 01-1582 it was |
| identified that a degraded condition had existed that is required to be |
| reported per 10 CFR 50.72 (b)(3)(ii). It was confirmed that Secondary |
| Containment Bypass leakage did not meet the acceptance criteria of less than |
| or equal to 0.0504 La for Type B and C LLRTs. Specifically the minimum |
| pathway leakage limit is 4340 sccm, however, the as found value was 20,653 |
| sccm (adjusted for instrument accuracy). The dominant leakage (18,110 sccm) |
| was from the High Pressure Core Spray Valve 1 E22-F0012, packing. The |
| penetration for this system credits the 1 E22-F0012 (located outside |
| Containment) as the inboard barrier and the closed system as the outboard |
| barrier, therefore, packing leakage in this case results in minimum pathway |
| leakage. The plant was in Refuel Outage 8 when the leakage was identified. |
| The valve has since been repacked and the current leakage is within limits. |
| |
| The NRC Resident Inspector was notified. |
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|General Information or Other |Event Number: 37874 |
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| REP ORG: FRAMATOME ANP RICHLAND INC |NOTIFICATION DATE: 03/29/2001|
|LICENSEE: FRAMATOME ANP RICHLAND INC |NOTIFICATION TIME: 17:58[EST]|
| CITY: RICHLAND REGION: 4 |EVENT DATE: 03/29/2001|
| COUNTY: STATE: WA |EVENT TIME: [PST]|
|LICENSE#: AGREEMENT: Y |LAST UPDATE DATE: 03/29/2001|
| DOCKET: |+----------------------------+
| |PERSON ORGANIZATION |
| |DAVE LOVELESS R4 |
| |BRUCE BURGESS R3 |
+------------------------------------------------+MICHELE EVANS R1 |
| NRC NOTIFIED BY: MALLAY |VERN HODGE NRR |
| HQ OPS OFFICER: CHAUNCEY GOULD | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|CCCC 21.21 UNSPECIFIED PARAGRAPH | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
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| PART 21 - AN INCORRECT COMPUTER CODE WAS SUPPLIED BY FRAMATOME TO FOUR |
| LICENSEES |
| FOR CALCULATION OF THE MINIMUM CRITICAL POWER RATIO (MCPR) |
| |
| In the performance of the analysis used to establish the MCPR operating |
| limits, the use of an inappropriate reference temperature resulted in an |
| overprediction of the thermal conductivity of the fuel. This overprediction |
| produced MCPR operating limits that were lower than they should have been. |
| Therefore, the MCPR limits must be raised by up to 0.01 or 0.02, depending |
| on the affected power plants. |
| |
| The MCPR must be raised by up to 0.01 from the limits previously provided |
| for Dresden Units 2 and 3, and for Quad Cities Units 1 and 2. The MCPR |
| limits for LaSalle Units 1 and 2 must be revised by up to 0.01 for power |
| levels of 60 percent and above and by up to 002 below 60 percent. |
| Calculations for Susquehanna Units 1 and 2 were affected but the MCPR limits |
| were unchanged. Continued compliance with the over-pressurization criteria |
| specified by the ASME code has been demonstrated. |
| |
| All units, including Susquehanna, have been notified. Compensatory measures |
| have been provided to the affected plants until confirmatory analyses have |
| been completed. The appropriate reference temperature has been installed in |
| the computer code used for the MCPR analysis. |
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