reading rm-doc collections-event status-event-2002-20020513en
U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
05/10/2002 - 05/13/2002
** EVENT NUMBERS **
38912 38915 38916
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|Power Reactor |Event Number: 38912 |
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| FACILITY: WATERFORD REGION: 4 |NOTIFICATION DATE: 05/10/2002|
| UNIT: [3] [] [] STATE: LA |NOTIFICATION TIME: 14:20[EDT]|
| RXTYPE: [3] CE |EVENT DATE: 05/09/2002|
+------------------------------------------------+EVENT TIME: 14:30[CDT]|
| NRC NOTIFIED BY: RONALD WILLIAMS |LAST UPDATE DATE: 05/10/2002|
| HQ OPS OFFICER: RICH LAURA +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: NON EMERGENCY |MARK SHAFFER R4 |
|10 CFR SECTION: | |
|NONR OTHER UNSPEC REQMNT | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE
|
+-----+----------+-------+--------+-----------------+--------+-----------------+
|3 N Y 100 Power Operation |100 Power Operation |
| | |
| | |
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EVENT TEXT
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| LICENSE CONDITION VIOLATION OF THERMAL POWER AT WATERFORD 3 |
| |
| "During initial installation of the LEFM Check-Plus ultrasonic feedwater |
| flow measuring system installed as part of the Appendix K Power Uprate |
| Project at Waterford 3, it was noted that the three available reactor core |
| calorimetric thermal power indications calculated by the Core Operating |
| Limits Supervisory System (COLSS) were not in agreement. This disagreement |
| was noted on 4/20/02. These thermal power indications consist of: Main |
| Steam venturi Secondary Calorimetric (MSBSCAL), Feedwater Flow venturi |
| Secondary Calorimetric (FWBSCAL) and the Calorimetric indication generated |
| by the newly installed LEEM Check-Plus ultrasonic flow measuring system |
| (USBSCAL). The LEFM Check-Plus ultrasonic flow measuring system was |
| certified for use on May 9, 2002. |
| |
| "Preliminary information indicates the mismatch in indication is likely the |
| result of biases factored into the MSBSCAL indication as well as possible |
| degradation of the secondary side of the plant over time since August 1997. |
| These biases, based on the most accurate USBSCAL indication, are |
| approximately 0.22% power beyond the accepted power measurement uncertainty |
| of 1.68% power for the MSBSCAL. indication. Thus, Waterford 3 may have |
| operated at average power levels in excess of the 100% licensed power limit |
| since approximately August 1997. |
| |
| "Waterford 3 is currently in Mode 1 at 99.9% power using FWBSCAL, the most |
| conservative indication of reactor power. This report is being made per |
| Waterford 3 License Condition 2.F for potential violation of License |
| Condition 2.C.1, "Maximum Power Level." The investigation for this condition |
| is ongoing". |
| |
| The NRC resident inspector was notified. |
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|Power Reactor |Event Number: 38915 |
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| FACILITY: COOK REGION: 3 |NOTIFICATION DATE: 05/13/2002|
| UNIT: [] [2] [] STATE: MI |NOTIFICATION TIME: 02:39[EDT]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 05/12/2002|
+------------------------------------------------+EVENT TIME: 23:01[EDT]|
| NRC NOTIFIED BY: BRADDOCK D. LEWIS |LAST UPDATE DATE: 05/13/2002|
| HQ OPS OFFICER: LEIGH TROCINE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: NON EMERGENCY |ANTON VEGEL R3 |
|10 CFR SECTION: | |
|ARPS 50.72(b)(2)(iv)(B) RPS ACTUATION - CRITICA| |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE
|
+-----+----------+-------+--------+-----------------+--------+-----------------+
| | |
|2 A/R Y 100 Power Operation |0 Hot Standby |
| | |
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EVENT TEXT
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| AUTOMATIC REACTOR TRIP DUE TO AN INSTRUMENTATION RACK POWER SUPPLY
FAILURE |
| WHICH CAUSED A STEAM GENERATOR FEEDWATER REGULATING VALVE TO FAIL
CLOSED |
| |
| The following text is a portion of a facsimile received from the licensee: |
| |
| "DC Cook Unit 2 tripped from full power due to an instrumentation rack power |
| supply failure on 05/12/02 [at] 2301. All control rods fully inserted. No |
| Safety Injection was required. The Unit 2 Reactor is stable and |
| subcritical. The Steam Generator Stop Valves were manually closed by the |
| Operating Crew to stabilize RCS Temperature in accordance with Plant |
| operating procedures. Reactor Coolant Temperature is being maintained |
| manually on the Steam Generator Atmospheric relief valves at No-Load T(ave) |
| in accordance with plant procedures." |
| |
| "This instrumentation rack power supply failure caused the #21 Steam |
| Generator Feed Regulating Valve to fail CLOSED. The Unit 2 Reactor |
| subsequently TRIPPED on Low Level in [the] #21 Steam Generator coincident |
| with Low Feedwater Flow. Several control systems were affected by the |
| control rack instrumentation failure: Pressurizer Pressure Control - |
| transferred control to manual and restored Pressurizer pressure, Pressurizer |
| Level Control - transferred control to manual and restored Pressurizer water |
| level, Refueling Water Sequence - Manually transferred Charging Pump Suction |
| to the RWST. Manual Operator response maintained and restored critical |
| plant parameters in MANUAL to normal parameter values." |
| |
| "Unit 2 entered Technical Specification 3.0.3 for 34 minutes because the |
| control system failures and plant system response temporarily caused the |
| Boration Flow paths from both the Refueling Water Storage Tank and Boric |
| Acid Storage Tanks to become INOPERABLE. Both Boration flow paths were |
| subsequently returned to OPERABLE status by manual Operator action." |
| |
| "The Reactor trip is reportable in accordance with 10 CFR 50.72(b)(2)(iv)(B) |
| as an actuation of the Reactor Protection System (RPS) when the Reactor was |
| critical. The Reactor TRIP was the result of an instrumentation rack power |
| supply failure and was not part of any preplanned test or evolution." |
| |
| The licensee stated that the unit is currently stable in Mode 3 (Hot |
| Standby). The reactor coolant pumps are available for primary system |
| transport control. Pressurizer level and pressure control are in manual. |
| Normal charging and makeup are available, but the automatic function of the |
| refueling water sequence is not available. The auxiliary feedwater pumps |
| automatically started as expected and are currently being utilized to supply |
| water to the steam generators. Secondary steam is being dumped to |
| atmosphere. There is no evidence of steam generator tube leakage, and |
| containment parameters are as expected. There were no safety injections and |
| none were required, and none of the primary power-operated relief valves |
| lifted. |
| |
| The licensee notified the NRC resident inspector. |
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|Power Reactor |Event Number: 38916 |
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| FACILITY: CLINTON REGION: 3 |NOTIFICATION DATE: 05/13/2002|
| UNIT: [1] [] [] STATE: IL |NOTIFICATION TIME: 03:21[EDT]|
| RXTYPE: [1] GE-6 |EVENT DATE: 05/13/2002|
+------------------------------------------------+EVENT TIME: 00:16[CDT]|
| NRC NOTIFIED BY: TOM CLAY |LAST UPDATE DATE: 05/13/2002|
| HQ OPS OFFICER: LEIGH TROCINE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: NON EMERGENCY |ANTON VEGEL R3 |
|10 CFR SECTION: | |
|ARPS 50.72(b)(2)(iv)(B) RPS ACTUATION - CRITICA| |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE
|
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 A/R Y 86 Power Operation |0 Hot Shutdown |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
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| AUTOMATIC REACTOR SCRAM ON HIGH WATER LEVEL DUE TO THE FAILURE OF THE
|
| STARTUP LEVEL CONTROLLER DURING EXTENDED POWER UPRATE TESTING ON
FEEDWATER |
| LEVEL CONTROL |
| |
| The following text is a portion of a facsimile received from the licensee: |
| |
| "With the reactor at 86% power, extended power uprate testing on feedwater |
| level control was in progress. A six-inch level step change from 32" to 38" |
| was inputted for testing. Feedwater was being controlled by the startup |
| level controller. The startup level controller failed to respond, allowing |
| reactor water level to go high and cause a reactor scram on level 8, 52". |
| All other systems responded normally." |
| |
| The licensee stated that all control rods fully inserted due to the |
| automatic reactor scram and that the highest reactor water level was 53". |
| There were no safety injections or emergency core cooling system actuations, |
| and none were required. |
| |
| The licensee stated that the unit is currently in Hot Shutdown. Normal |
| feedwater is being supplied to the reactor vessel for level control, and |
| pressure is being controlled via the main steam line drains which go to the |
| condenser (the heat sink). The outboard main steam isolation valves were |
| manually closed. Containment parameters are normal. |
| |
| The licensee notified the NRC resident inspector. |
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