Event Notification Report for December 18, 2001
U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
12/17/2001 - 12/18/2001
** EVENT NUMBERS **
38575 38577 38578 38579
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|Power Reactor |Event Number: 38575 |
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| FACILITY: PERRY REGION: 3 |NOTIFICATION DATE: 12/16/2001|
| UNIT: [1] [] [] STATE: OH |NOTIFICATION TIME: 00:30[EST]|
| RXTYPE: [1] GE-6 |EVENT DATE: 12/15/2001|
+------------------------------------------------+EVENT TIME: 22:28[EST]|
| NRC NOTIFIED BY: ALAN RABENOLD |LAST UPDATE DATE: 12/17/2001|
| HQ OPS OFFICER: LEIGH TROCINE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: NON EMERGENCY |ANNE MARIE STONE R3 |
|10 CFR SECTION: | |
|ACCS 50.72(b)(2)(iv)(A) ECCS INJECTION | |
|ARPS 50.72(b)(2)(iv)(B) RPS ACTUATION - CRITICA| |
|AESF 50.72(b)(3)(iv)(A) VALID SPECIF SYS ACTUAT| |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE
|
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 A/R Y 100 Power Operation |0 Hot Shutdown |
| | |
| | |
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EVENT TEXT
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| THE REACTOR RECIRCULATION PUMPS DOWN SHIFTED TO SLOW SPEED FOR
UNKNOWN |
| REASONS RESULTING IN AN AUTOMATIC REACTOR SCRAM, BALANCE-OF-PLANT
|
| ISOLATIONS, AND EMERGENCY CORE COOLING SYSTEM ACTUATIONS. |
| |
| At 2228 EST on 12/15/01, the reactor recirculation pumps down shifted to |
| slow speed for unknown reasons. This caused reactor pressure vessel (RPV) |
| water level to swell, and an automatic reactor scram from 100% power |
| occurred when Level 8 was reached. All control rods fully inserted. |
| Following the scram, RPV water level started to decrease because the |
| feedwater pumps also tripped (as designed) when Level 8 was reached. When |
| RPV water level dropped to Level 2 (130 inches), balance-of-plant isolations |
| occurred, and all of the applicable valves properly isolated. In addition, |
| the high pressure core spray (HPCS) and reactor core isolation cooling |
| (RCIC) systems automatically actuated when Level 2 was reached. The HPCS |
| and RCIC systems were utilized to restore RPV water level to the normal |
| range. |
| |
| The unit is currently stable in Mode 3 with RPV pressure at 900 psi and RPV |
| water level at 210 inches. Normal feedwater is being utilized to maintain |
| RPV water level within the proper band, and the HPCS and RCIC systems have |
| been secured. The main steam isolation valves remained open, and the |
| condenser is being utilized as the heat sink. Containment parameters are |
| currently normal, the electrical grid is stable, and the emergency diesel |
| generators are available. |
| |
| The licensee stated that all systems functioned as required and that there |
| was nothing unusual or misunderstood other than the cause of the initiating |
| event (the recirculation pumps down shifting to slow speed). The licensee's |
| investigation is underway. |
| |
| The licensee notified the NRC resident inspector. |
| |
| * * * UPDATE AT 1844 EST ON 12/17/01 FROM DAVID GUDGER TO S. SANDIN * * * |
| |
| "This notification is a follow-up to the notification provided on |
| 12/16/01." |
| |
| "The following is a description of the sequence of plant events that |
| occurred as determined by the failure analysis." |
| |
| "Initial conditions: Feedwater level control was on Master Level controller |
| selected to the 'B' narrow range channel and the plant was operating at 100% |
| power. Due to the level summer card 1C34K657 failure, the level signal from |
| the selected channel rapidly decreased to less than 178 inches, Level 3 (L3) |
| signal, which caused the reactor recirculation pumps to receive a |
| fast-to-slow speed downshift. Simultaneously, feedwater flow rapidly |
| increased in response to the low level on the 'B' channel. The feedwater |
| pumps tripped at 219 inches, Level 8 (L8), which occurred due to swell from |
| the recirculation pumps downshift and increased feedwater flow. The [motor |
| feed pump (MFP)] did not automatically start due to the L8 signal (as |
| designed) and level decreased to 130 inches, Level 2 (L2). The |
| recirculation pumps tripped, and [the] HPCS and RCIC systems [automatically] |
| started and restored level. The MFP L8 signals were reset, and level was |
| controlled on the startup controller." |
| |
| "The Reactor Water Cleanup Inboard Containment Isolation Valve failed to |
| close as designed upon [the] L2 isolation signal. The valve was manually |
| closed during containment valve isolation verification following the scram. |
| The cause of the failure to close was determined to be relay failure. Two |
| relays were subsequently replaced and retested satisfactorily." |
| |
| The licensee informed the NRC resident inspector. The NRC operations |
| officer notified the R3DO (Phillips). |
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|Power Reactor |Event Number: 38577 |
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| FACILITY: SUMMER REGION: 2 |NOTIFICATION DATE: 12/17/2001|
| UNIT: [1] [] [] STATE: SC |NOTIFICATION TIME: 09:58[EST]|
| RXTYPE: [1] W-3-LP |EVENT DATE: 11/20/2001|
+------------------------------------------------+EVENT TIME: 03:25[EST]|
| NRC NOTIFIED BY: JIM TURKETT |LAST UPDATE DATE: 12/17/2001|
| HQ OPS OFFICER: LEIGH TROCINE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: NON EMERGENCY |CHARLES R. OGLE R2 |
|10 CFR SECTION: | |
|AINV 50.73(a)(1) INVALID SPECIF SYSTEM A| |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|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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| INADVERTENT START OF A MOTOR-DRIVEN EMERGENCY FEEDWATER PUMP (60-Day
|
| Report) |
| |
| The following text is a portion of a facsimile received from the licensee: |
| |
| "During the performance of a surveillance test, the 'B' Motor-Driven |
| Emergency Feedwater (EFW) pump was inadvertently started from the Main |
| Control Board when the intended action was to place the control switch in |
| the pull-to-lock position. The plant was preparing to test the function of |
| the solid state output relay K633 as it pertains to sending an open signal |
| to the EFW flow control valves IFV03531, IFV03541, and IFV03551. To prevent |
| actually injecting EFW into the [steam generators (S/G)], the pump for the |
| train whose relay is being tested (in this case 'B') must be placed in |
| pull-to-lock since the same relay which opens the IFVs also starts the |
| pump." |
| |
| "When the BOP operator attempted to place the pump in P-T-L the pump |
| started. This was immediately recognized as an unexpected response, and the |
| pump was shut off within a second. Due to the start of the EFW pump, S/G |
| blowdown isolated as per design and had to be subsequently realigned. [Data |
| from the] plant computer and alarm printers [was] reviewed by the Shift |
| Supervisor and [Shift Engineer] to determine if any cold EFW actually |
| entered the S/Gs. The alarm printer showed 1 second between pump start and |
| pump stop, which is consistent with the prompt error recognition observed by |
| the crew. The plant computer data showed no increase in EFW flow during the |
| time period in question. It is believed from this investigation that the |
| pump never came up to speed to the point where sufficient pressure was |
| developed to swing open the discharge check valves, thus no EFW injection |
| occurred. The cause of this event is attributed to human performance |
| error." |
| |
| "This event does not require a telephone notification under |
| 10CFR50.72(b)(3)(iv)(A) because this is being reported as an invalid ECCS |
| actuation of an Emergency Feedwater Pump. However, the plant is making this |
| call under the criteria of 10CFR50.73(a)(1) in lieu of submitting an LER |
| under 10CFR50.73(a)(2)(iv)." |
| |
| The licensee plans to notify the NRC resident inspector. |
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|Power Reactor |Event Number: 38578 |
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| FACILITY: BRAIDWOOD REGION: 3 |NOTIFICATION DATE: 12/17/2001|
| UNIT: [1] [2] [] STATE: IL |NOTIFICATION TIME: 10:39[EST]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 12/17/2001|
+------------------------------------------------+EVENT TIME: 08:00[CST]|
| NRC NOTIFIED BY: GREG BAKER |LAST UPDATE DATE: 12/17/2001|
| HQ OPS OFFICER: LEIGH TROCINE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: NON EMERGENCY |MONTE PHILLIPS R3 |
|10 CFR SECTION: | |
|HFIT 26.73 FITNESS FOR DUTY | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE
|
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 100 Power Operation |100 Power Operation |
|2 N Y 95 Power Operation |95 Power Operation |
| | |
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EVENT TEXT
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| SIGNIFICANT FITNESS-FOR-DUTY EVENT UNDER BRAIDWOOD PROGRAM SEC-1.10
|
| |
| A non-licensed supervisor tested positive based on a for-cause test due to |
| the odor of alcohol. The individual was immediately escorted out of the |
| Protected Area, and the individual's access has been suspended. (Call the |
| NRC operations officer for additional details and for the Security Manager's |
| telephone number.) |
| |
| The licensee notified the NRC resident inspector. |
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|Power Reactor |Event Number: 38579 |
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| FACILITY: HOPE CREEK REGION: 1 |NOTIFICATION DATE: 12/17/2001|
| UNIT: [1] [] [] STATE: NJ |NOTIFICATION TIME: 16:17[EST]|
| RXTYPE: [1] GE-4 |EVENT DATE: 12/17/2001|
+------------------------------------------------+EVENT TIME: 15:30[EST]|
| NRC NOTIFIED BY: DARON ZAKARIAN |LAST UPDATE DATE: 12/17/2001|
| HQ OPS OFFICER: STEVE SANDIN +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: NON EMERGENCY |JOHN KINNEMAN R1 |
|10 CFR SECTION: | |
|NONR OTHER UNSPEC REQMNT | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE
|
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 100 Power Operation |10 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
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| 24-HOUR REPORT DUE TO POTENTIAL OPERATION IN EXCESS OF OPERATING
LICENSE |
| CONDITION 2.C(1) |
| |
| "After restart of Hope Creek from RF10, analysis of plant parameters |
| indicated that changes occurred in the final feedwater temperature since the |
| previous operating cycle. The temperatures as indicated by temperature |
| loops 1AETE-N041A/B/C/D, were approximately 3 degrees lower than when |
| operating at 100 percent power prior to the outage. This indicates that |
| actual power may have been below 100 percent of last cycle operation. Other |
| parameters, such as First Stage pressure, main turbine control valve |
| position, #6 Feedwater Heater shell pressure, feed pump discharge flow, |
| condensate pump flow, were also lower, substantiating that power was lower |
| than previous cycle. Performance Engineering notified Design Engineering of |
| this condition. Design Engineering then initiated a complete review of |
| related documentation. |
| |
| "During Hope Creek RF9 (May 2000), the main feedwater temperature loops |
| required recalibration to support power up-rate. The traditional method to |
| determine the RTD curve introduced a +1 degree F bias. This was |
| non-conservative, in that power level was higher than calculated by the |
| plant computer. In January of 2001, crossflow was implemented to correct |
| for fouling of the venturis. In October 2001, a non-conservative moisture |
| carryover fraction was used in the core thermal power calculation as |
| reported in Hope Creek Special Report 354/2001-003-00. As a result of these |
| conditions, Hope Creek Generating Station has potentially operated at power |
| levels in excess of Operating License Condition 2.C(1), which requires that |
| the facility be operated at reactor core power levels not in excess of 3339 |
| MWt. The upper limit may have been exceeded by significantly less than |
| 0.1%. This potential overpower condition existed between the time of |
| crossflow implementation and RF10. |
| |
| "This notification is being made in accordance with Hope Creek Operating |
| License Condition 2.F, as a potential violation of Hope Creek Operating |
| License Condition 2.C(1). |
| |
| "The moisture carryover fraction was corrected during RF10. During the |
| recent forced outage, the 4 RTDs were checked and all four loops were |
| calibrated, therefore, eliminating the potential for operating above 100% |
| power." |
| |
| The period of time during which the unit may have operated in this condition |
| occurred between August 1, 2001, and October 10, 2001. The corrective |
| action involving calibration of the 4 RTDs was completed last week. |
| |
| The licensee informed the NRC Resident Inspector. |
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