Event Notification Report for August 10, 2001
U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
08/09/2001 - 08/10/2001
** EVENT NUMBERS **
38173 38186 38193 38194 38195
!!!!!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!!!
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|Power Reactor |Event Number: 38173 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: SOUTH TEXAS REGION: 4 |NOTIFICATION DATE: 07/25/2001|
| UNIT: [1] [2] [] STATE: TX |NOTIFICATION TIME: 19:02[EDT]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 07/25/2001|
+------------------------------------------------+EVENT TIME: 10:11[CDT]|
| NRC NOTIFIED BY: WAYNE HARRISON |LAST UPDATE DATE: 08/09/2001|
| HQ OPS OFFICER: BOB STRANSKY +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: NON EMERGENCY |GARY SANBORN R4 |
|10 CFR SECTION: | |
|AUNA 50.72(b)(3)(ii)(B) UNANALYZED CONDITION | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|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 100 Power Operation |100 Power Operation |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
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| POTENTIAL TO DRAIN AFW STORAGE TANK DURING PLANT FLOODING SCENARIO |
| |
| "STPNOC Engineering identified a condition outside the station's design |
| basis that has been determined to be reportable under 10CFR50.73(a)(2)(ii) |
| as an unanalyzed condition that significantly degraded plant safety. |
| Notification is required by 10CFR50.72(b)(3)(ii)(B). |
| |
| "During review of design calculations, a new internal flooding condition was |
| identified that could have resulted in depletion of the AFW Storage Tank to |
| the point where the plant would not be able to transition from AFW to RHR as |
| designed. Each of STP's four trains of AFW is enclosed in its own |
| water-tight compartment directly beneath its associated MFW line RCB |
| penetration. Operator response to a main feedline break includes isolation |
| of the faulted steam generator, including AFW. If the MFW break is |
| postulated to occur in the MFW penetration area above AFW, the AFW cubicle |
| beneath the break will flood. In the case of the D train steam-driven AFW, |
| the water level will submerge the turbine-driven AFW pump, its trip/throttle |
| valve and AFW isolation valves in about 6 to 30 minutes, depending on break |
| size. As a consequence of the accident, the submerged motor-operated valves |
| are assumed to fail as-is, supplying steam to the turbine-driven pump and |
| allowing AFW flow. In addition, the analysis assumes the single failure of |
| one unsubmerged steam supply isolation valve. The steam-driven pump will |
| continue to function while submerged and continue to take suction from the |
| AFWST and expel it out the break. Unless the pump is secured, its continued |
| operation could accelerate the depletion of the AFWST to the point that the |
| plant would not be able to transition from AFW to RHR as designed. This |
| condition affects only the D train steam-driven AFW. Trains A, B, and C are |
| motor driven and can be readily secured at their power source if necessary. |
| |
| "Compensatory action is being taken to implement a temporary modification to |
| allow operator action to isolate the AFW supply to the steam-driven AFW pump |
| at the AFWST." |
| |
| The NRC resident inspector has been informed of this condition by the |
| licensee. |
| |
| ***** RETRACTION FROM WAYNE HARRISON TO L. TROCINE RECEIVED AT 1507 EDT ON |
| 08/09/01 ***** |
| |
| The following text is a portion of a facsimile received from the licensee: |
| |
| "[...] This notification is RETRACTED. The break described above is in a |
| break exclusion zone and is the non-mechanistic break required by the |
| Standard Review Plan to be postulated for the purpose of defining the |
| environmental conditions for qualification of equipment. Consideration of a |
| single failure with this non-mechanistic break is not required. |
| Consequently, the unsubmerged steam supply valve may be assumed to function |
| to isolate steam to the turbine-driven AFW pump and there is not a condition |
| that significantly degrades plant safety." |
| |
| The licensee plans to notify the NRC resident inspector. The NRC operations |
| officer notified the R4DO (Charles Marschall). |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 38186 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: DIABLO CANYON REGION: 4 |NOTIFICATION DATE: 08/04/2001|
| UNIT: [1] [2] [] STATE: CA |NOTIFICATION TIME: 19:55[EDT]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 08/04/2001|
+------------------------------------------------+EVENT TIME: 15:22[PDT]|
| NRC NOTIFIED BY: DAVID BAHNER |LAST UPDATE DATE: 08/09/2001|
| HQ OPS OFFICER: STEVE SANDIN +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: NON EMERGENCY |CHUCK CAIN R4 |
|10 CFR SECTION: |HERB BERKOW NRR |
|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 N Y 100 Power Operation |100 Power Operation |
|2 N Y 100 Power Operation |100 Power Operation |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| AUTOMATIC START OF BOTH UNITS DIESEL GENERATORS DUE TO A LOSS OF START-UP |
| POWER |
| |
| "Electrical fault caused loss of Start-up Power to both units. This caused |
| all six [6] diesel generators on both units to automatically start." |
| |
| Circuit breaker CB212 which supplies 230KV start-up power to both units |
| opened due to an unidentified electrical fault. Initial indication was a |
| loss of power to outbuildings onsite followed about twenty-two (22) minutes |
| later with a loss of power to the Start-up transformers. Investigation |
| found the grounding fuses on the Unit 1 side open. Should a unit trip occur |
| while in this condition, RCS cooldown by natural circulation would be |
| required. |
| |
| Both units are in Tech Spec LCO 3.8.1 Condition 'A' which requires |
| restoration of start-up power within 72-hours or unit shutdown; mode 3 (hot |
| standby) in the following 6-hours followed by mode 5 (cold shutdown) |
| 36-hours later. The licensee is performing a maintenance and engineering |
| evaluation to determine if start-up can be restored by closing a cross-tie |
| breaker from Unit 2. |
| |
| None of the diesel generators loaded and all have been secured and returned |
| to standby. |
| |
| The licensee informed the NRC resident inspector. |
| |
| ***** UPDATE FROM JEFF KNISLEY TO LEIGH TROCINE RECEIVED AT 1759 EDT ON |
| 08/09/01 ***** |
| |
| The following text is a portion of a facsimile received from the licensee: |
| |
| "[...] FOLLOW-UP INFORMATION" |
| |
| "The event, as described in the original event notification number 38186, |
| caused all six diesel generators on both units to automatically start. The |
| diesel generators started, as designed, upon receiving an anticipatory |
| signal due to the loss of Start-Up Power; and the diesel generators did not |
| load, as designed, because vital and non vital power was still available |
| from the Auxiliary Transformers, back-fed from the generator output of each |
| unit." |
| |
| "As stated in the original event notification, the grounding fuses for the |
| Unit 1 Start-Up Transformer were found open. The immediate cause of the |
| event was a fault in the 12kV Fuse Cabinet for the Unit 1 Start-Up |
| Transformer ground resistor. The ground resistor was observed to be |
| radiating heat just prior to the event witnessed by the damage to the Fuse |
| Cabinet. The cause of this failure is being investigated within the |
| corrective action program." |
| |
| "As stated in the original event notification, there was a loss of power to |
| outbuildings onsite. The 12kV underground distribution breaker tripped [on |
| 08/04/01 at] 1500 PDT, causing the loss of non-essential site loads; such as |
| the Administration, Training and Maintenance Buildings, and support |
| facilities -supporting power production including water treatment for |
| Condensate make-up, back-up Service Air Compressors, and Radwaste Laundry |
| Building." |
| |
| "Start-Up power to Unit 2 was restored [on 08/05/01 at] 0513 PDT. After |
| assuring there was no damage to the Unit 1 Start-Up Bus, power was restored |
| to the Unit 1 Start-Up Bus [on 08/06/01 at] 1122 PDT, via a cross-tie |
| breaker from the Unit 2 Start-Up Transformer. The 12kV underground loop was |
| reenergized [on 08/07/01], thus restoring power to the non-essential site |
| loads. Repair of the Fuse Cabinet is in progress." |
| |
| The licensee notified the NRC resident inspector. The NRC operations |
| officer notified the R4DO (Charles Marschall). |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|General Information or Other |Event Number: 38193 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| REP ORG: ARKANSAS DEPARTMENT OF HEALTH |NOTIFICATION DATE: 08/09/2001|
|LICENSEE: GREEN BAY PACKAGING |NOTIFICATION TIME: 14:22[EDT]|
| CITY: MORRILTON REGION: 4 |EVENT DATE: 08/07/2001|
| COUNTY: STATE: AR |EVENT TIME: [CDT]|
|LICENSE#: ARK197BP01-98 AGREEMENT: Y |LAST UPDATE DATE: 08/09/2001|
| DOCKET: |+----------------------------+
| |PERSON ORGANIZATION |
| |CHARLES MARSCHALL R4 |
| |FRED BROWN NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: JARED THOMPSON (fax) | |
| HQ OPS OFFICER: LEIGH TROCINE | |
+------------------------------------------------+ |
|EMERGENCY CLASS: NON EMERGENCY | |
|10 CFR SECTION: | |
|NAGR AGREEMENT STATE | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| AGREEMENT STATE REPORT REGARDING A DAMAGED NDC GAUGE CONTAINING |
| AMERICIUM-241 AT |
| GREEN BAY PACKAGING IN MORRILTON, ARKANSAS |
| |
| The following text is a portion of a facsimile received from the Arkansas |
| Department of Health: |
| |
| "The Department received notified on August 8, 2001, that a gauge containing |
| radioactive material was damaged during maintenance operations at Green Bay |
| Packaging in Morrilton, Arkansas. The gauge is licensed under Arkansas |
| Radioactive Material License Number ARK-197-BP-01-98." |
| |
| "The incident occurred on August 7, 2001, when the gauge became caught in |
| the conveyor wire and was pulled away from the mounting bracket, falling and |
| striking a roller. The impact tore the shutter from the gauge. The gauge |
| was retrieved without exposure to personnel. Temporary shielding was added |
| to the gauge, and the gauge was placed in storage by the Radiation Safety |
| Officer." |
| |
| "It was determined that the source was intact, and a leak test was |
| performed. The gauge is to be returned to the manufacturer." |
| |
| "The gauge is [an] NDC Model Number 104F, Serial Number 12605, containing |
| 0.93 gBq (21 mCi) of Americium-241." |
| |
| "The Department is conducting an investigation of the incident." |
| |
| (Call the NRC operations officer for contact information.) |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 38194 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: OYSTER CREEK REGION: 1 |NOTIFICATION DATE: 08/09/2001|
| UNIT: [1] [] [] STATE: NJ |NOTIFICATION TIME: 18:18[EDT]|
| RXTYPE: [1] GE-2 |EVENT DATE: 08/09/2001|
+------------------------------------------------+EVENT TIME: 16:01[EDT]|
| NRC NOTIFIED BY: ERIC DeMONCH |LAST UPDATE DATE: 08/09/2001|
| HQ OPS OFFICER: LEIGH TROCINE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: NON EMERGENCY |DANIEL HOLODY R1 |
|10 CFR SECTION: | |
|APRE 50.72(b)(2)(xi) OFFSITE NOTIFICATION | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 100 Power Operation |93 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| OFFSITE NOTIFICATION REGARDING APPROVAL TO EXCEED THE NORMAL OPERATING |
| DISCHARGE TEMPERATURE PERMIT LIMIT DUE TO THE EMERGENCY NEED FOR POWER AND |
| CONDITIONS ON THE GRID |
| |
| At approximately 1500, the licensee began to reduce reactor power from 100% |
| because main condenser discharge temperature began to approach the discharge |
| permit temperature limit of 106�F. At 1601, the system dispatcher notified |
| the licensee of an emergency need for power and informed the licensee that |
| the criteria to exceed the discharge permit temperature limit was |
| authorized. In this condition, a discharge temperature of 110�F is allowed. |
| As a result, the licensee ceased the power reduction with the unit at |
| approximately 98% reactor power and returned reactor power to 100%. |
| |
| The emergency need for power was cancelled at 1612. As a result, the |
| licensee commenced another power reduction in order to reduce the main |
| condenser discharge temperature and comply with the 106�F discharge permit |
| limit. |
| |
| The highest main condenser discharge temperature attained was 106.36�F. |
| Reactor power is currently stable at 93%. |
| |
| The licensee notified the New Jersey Department of Environmental Protection |
| and plans to notify the NRC resident inspector. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 38195 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PALISADES REGION: 3 |NOTIFICATION DATE: 08/09/2001|
| UNIT: [1] [] [] STATE: MI |NOTIFICATION TIME: 21:37[EDT]|
| RXTYPE: [1] CE |EVENT DATE: 08/09/2001|
+------------------------------------------------+EVENT TIME: 20:00[EDT]|
| NRC NOTIFIED BY: MARK HOLBEIN |LAST UPDATE DATE: 08/09/2001|
| HQ OPS OFFICER: LEIGH TROCINE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: NON EMERGENCY |ANNE MARIE STONE R3 |
|10 CFR SECTION: | |
|AUNA 50.72(b)(3)(ii)(B) UNANALYZED CONDITION | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N N 0 Cold Shutdown |0 Cold Shutdown |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| DISCOVERY THAT SUPPORTS MAY NOT BE CAPABLE OF PROPERLY RESTRAINING THE |
| MISSILE SHIELD DURING A DESIGN BASIS SEISMIC EVENT |
| |
| The following text is a portion of a facsimile received from the licensee: |
| |
| "Evaluation of the Missile Shield, over the Reactor Vessel Head, has |
| revealed that the supports for the Missile Shield may not be capable of |
| properly restraining the Missile [Shield] in the horizontal direction during |
| a Design Basis Seismic Event. Consequently the Missile Shield is considered |
| inoperable pending further evaluation. This is being reported as an |
| unanalyzed condition that could significantly degrade Plant safety." |
| |
| The licensee notified the NRC resident inspector. |
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