Event Notification Report for November 10, 1999
U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
11/09/1999 - 11/10/1999
** EVENT NUMBERS **
36290 36331 36338 36411 36415 36416 36417
!!!!!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!!!
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36290 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: FITZPATRICK REGION: 1 |NOTIFICATION DATE: 10/14/1999|
| UNIT: [1] [] [] STATE: NY |NOTIFICATION TIME: 12:29[EDT]|
| RXTYPE: [1] GE-4 |EVENT DATE: 10/14/1999|
+------------------------------------------------+EVENT TIME: 11:30[EDT]|
| NRC NOTIFIED BY: MARK ABRAMSKI |LAST UPDATE DATE: 11/09/1999|
| HQ OPS OFFICER: JOHN MacKINNON +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |JAMES NOGGLE R1 |
|10 CFR SECTION: | |
|AOUT 50.72(b)(1)(ii)(B) OUTSIDE DESIGN BASIS | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 100 Power Operation |100 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| The Charcoal Absorption Efficiency of Train 'B" Standby Gas Treatment |
| System was discovered to be less than 99.8%. |
| |
| An engineering review of the absorption capability of the Standby Gas |
| Treatment charcoal filters has concluded that the "B" Division of Standby |
| Gas Treatment has been inoperable since 03/30/99. On 04/10/99, samples of |
| the charcoal of the Standby Gas Treatment system were sent offsite to check |
| the absorption efficiency of the charcoal, and other properties of the |
| charcoal. The results of the analysis were received by the licensee on |
| 05/16/99. Today, it was discovered that the absorption efficiency of the |
| charcoal is 99.3%. The licensee has a commitment that the absorption |
| efficiency of the charcoal will be at least 99.8%. The "B" Division of |
| Standby Gas Treatment System is inoperable at this time for another reason. |
| |
| The NRC Resident Inspector was notified of this event by the licensee. |
| |
| * * * RETRACTION 0801 11/9/1999 FROM COSTEDIO TAKEN BY STRANSKY * * * |
| |
| The LOCA dose evaluations assume a standby gas treatment (SBGT) system |
| charcoal efficiency of 99%. Based on the SBGT train 'B' charcoal efficiency |
| test results of 99.37%, the licensee believes that reasonable assurance |
| exists to conclude that the SBGT system would have performed its intended |
| safety function. Based upon this conclusion, the plant did not operate |
| outside of its design basis. The NRC resident inspector has been informed of |
| this event by the licensee. Notified R1DO (Holody). |
+------------------------------------------------------------------------------+
!!!!!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!!!
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36331 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: FT CALHOUN REGION: 4 |NOTIFICATION DATE: 10/22/1999|
| UNIT: [1] [] [] STATE: NE |NOTIFICATION TIME: 13:04[EDT]|
| RXTYPE: [1] CE |EVENT DATE: 10/22/1999|
+------------------------------------------------+EVENT TIME: 09:22[CDT]|
| NRC NOTIFIED BY: ERICK MATZKE |LAST UPDATE DATE: 11/09/1999|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |LINDA SMITH R4 |
|10 CFR SECTION: | |
|ADAS 50.72(b)(2)(i) DEG/UNANALYZED COND | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|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
+------------------------------------------------------------------------------+
| - Steam Generator #RC-2A in Tech Spec Category C-3; Steam Generator |
| inspections continuing - |
| |
| At 0922 CDT on 10/22/99, during eddy current testing of the steam generators |
| (SG), it has been determined that 50 tubes in SG #RC-2A require plugging. |
| This places SG #RC-2A in Technical Specification category C-3 per 3.17(2). |
| Forty-four tubes have been determined to require plugging in SG #RC-2B at |
| this time. Eddy current testing is continuing on the SGs. A 100% full |
| length bobbin coil inspection program has been completed in both SGs. A |
| rotating pancake coil probe (Plus Point) is being used to inspect 100% of |
| the top of the hot leg tube sheets for both SGs. One hundred percent of |
| these inspections for the 'A' SG are complete with about 99% evaluated. |
| About 85% are complete on the 'B' SG with the rest expected to be completed |
| on 10/22/99. In addition, a large number of rotating pancake coil probe |
| inspections are being conducted at other locations in the SGs. In-situ |
| pressure testing is being completed where needed. To date, 4 tubes in the |
| 'A' SG and 2 tubes in the 'B' SG have been pressure tested. All 6 of these |
| tubes have passed at 3 times normal operating differential pressure with |
| zero leakage. |
| |
| This report is conservatively being made prior to completing the SG testing |
| and before completely evaluating the effect on the plant. Further |
| evaluation of reportability will be completed following the completion of |
| the eddy current and in-situ pressure testing of the SGs. |
| |
| The licensee notified the NRC Resident Inspector. |
| |
| NOTE: Refer to related Event #36338. |
| |
| * * * RETRACTION 0923 EST 11/9/1999 FROM MATZKE TAKEN BY STRANSKY * * * |
| |
| "Both steam generators were declared in category C-3 per Technical |
| Specification 3.17 due to having greater than 1% of the inspected tubes |
| being found defective. In steam generator RC-2A, 63 tubes were found |
| defective out of 4901 inspected tubes. In steam generator RC-2B, 57 tubes |
| were found defective out of 4905 inspected tubes. |
| |
| "The technical specifications contain provisions for plugging tubes when |
| they are found to contain defects that penetrate greater than 40% |
| through-wall. Under this technical specification, it is expected that the |
| plant may operate for a period of time with defects greater than 40% |
| through-wall prior to being found and plugged. The tubes that were found to |
| contain defects were plugged in accordance with technical specifications. |
| The plugging criteria currently in use at Fort Calhoun Station requires that |
| all indications of corrosion detected by eddy current are plugged on |
| detection due to the absence of a qualified technique for sizing |
| indications. |
| |
| "Following eddy current testing of the steam generator tubes, in-situ |
| pressure testing was performed on certain defects which exceeded the |
| screening criteria. The criteria is based on the potential to exceed the |
| performance criteria for leakage or structural integrity. The leakage |
| performance criterion requires that leakage from all defects within a steam |
| generator shall not exceed 1 gallon per minute under worst case accident |
| differential pressure and the structural integrity performance criterion |
| states that the tubes shall withstand pressure of up to three times normal |
| operating differential pressure without burst. Selected indications which |
| are representative of the worst of the population of indications found in |
| the steam generators successfully passed in-situ pressure tests with no |
| leakage at worst case accident differential pressure and no leakage at three |
| times normal operating differential pressure. There was no detectable |
| primary-to-secondary leakage during operation prior to shutdown for the |
| current refueling outage. Therefore, the steam generators were both |
| available to perform their required safety functions as verified through |
| in-situ pressure testing. Based on the testing performed, the tubes are not |
| considered to have been seriously degraded, the plant was not in an |
| unanalyzed condition, the steam generators would have performed their design |
| basis functions during accident conditions, and this did not constitute a |
| condition outside the plant's operating and emergency procedures. |
| |
| "The original reports were made conservatively awaiting the completion of |
| the inspection and testing program. The reports are now being retracted |
| based on the complete testing results. The Fort Calhoun Station Technical |
| Specification 30-day plugging report and 6-month inspection report will be |
| submitted as required." |
| |
| The NRC resident inspector has been informed of this retraction by the |
| licensee. Notified R4DO (Smith). |
+------------------------------------------------------------------------------+
!!!!!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!!!
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36338 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: FT CALHOUN REGION: 4 |NOTIFICATION DATE: 10/23/1999|
| UNIT: [1] [] [] STATE: NE |NOTIFICATION TIME: 20:59[EDT]|
| RXTYPE: [1] CE |EVENT DATE: 10/23/1999|
+------------------------------------------------+EVENT TIME: 16:45[CDT]|
| NRC NOTIFIED BY: KEVIN BOSTON |LAST UPDATE DATE: 11/09/1999|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |LINDA SMITH R4 |
|10 CFR SECTION: | |
|ADAS 50.72(b)(2)(i) DEG/UNANALYZED COND | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N N 0 Refueling |0 Refueling |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| - Steam Generator #RC-2B is in Tech Spec Category C-3 - |
| |
| In accordance with Tech Spec Section 3.17 (5), Reporting Requirements, the |
| following 4-hour non-emergency report is being made. |
| |
| During Eddy Current Testing tube inspections on Steam Generator (SG) #RC-2B, |
| greater than 1% of the tubes tested were found to be defective. The number |
| of inspected tubes during the 1999 refueling outage is 4905 in SG #RC-2B. |
| The number of tubes considered defective and require plugging exceeded 49 |
| tubes. |
| |
| SG #RC-2B was declared in Tech Spec 3.17, Table 3-13, Category C-3, at 1645 |
| CDT on 10/23/99. Tube testing is being conducted under procedure |
| SE-ST-RC-0003, Inservice Testing of Steam Generator Tubes. |
| |
| The licensee notified the NRC Resident Inspector. |
| |
| Note: Refer to related Event #36331. |
| |
| * * * RETRACTION 0923 11/9/1999 FROM MATZKE TAKEN BY STRANSKY * * * |
| |
| "Both steam generators were declared in category C-3 per Technical |
| Specification 3.17 due to having greater than 1% of the inspected tubes |
| being found defective. In steam generator RC-2A, 63 tubes were found |
| defective out of 4901 inspected tubes. In steam generator RC-2B, 57 tubes |
| were found defective out of 4905 inspected tubes. |
| |
| "The technical specifications contain provisions for plugging tubes when |
| they are found to contain defects that penetrate greater than 40% |
| through-wall. Under this technical specification, it is expected that the |
| plant may operate for a period of time with defects greater than 40% |
| through-wall prior to being found and plugged. The tubes that were found to |
| contain defects were plugged in accordance with technical specifications. |
| The plugging criteria currently in use at Fort Calhoun Station requires that |
| all indications of corrosion detected by eddy current are plugged on |
| detection due to the absence of a qualified technique for sizing |
| indications. |
| |
| "Following eddy current testing of the steam generator tubes, in-situ |
| pressure testing was performed on certain defects which exceeded the |
| screening criteria. The criteria is based on the potential to exceed the |
| performance criteria for leakage or structural integrity. The leakage |
| performance criterion requires that leakage from all defects within a steam |
| generator shall not exceed 1 gallon per minute under worst case accident |
| differential pressure and the structural integrity performance criterion |
| states that the tubes shall withstand pressure of up to three times normal |
| operating differential pressure without burst. Selected indications which |
| are representative of the worst of the population of indications found in |
| the steam generators successfully passed in-situ pressure tests with no |
| leakage at worst case accident differential pressure and no leakage at three |
| times normal operating differential pressure. There was no detectable |
| primary-to-secondary leakage during operation prior to shutdown for the |
| current refueling outage. Therefore, the steam generators were both |
| available to perform their required safety functions as verified through |
| in-situ pressure testing. Based on the testing performed, the tubes are not |
| considered to have been seriously degraded, the plant was not in an |
| unanalyzed condition, the steam generators would have performed their design |
| basis functions during accident conditions, and this did not constitute a |
| condition outside the plant's operating and emergency procedures. |
| |
| "The original reports were made conservatively awaiting the completion of |
| the inspection and testing program. The reports are now being retracted |
| based on the complete testing results. The Fort Calhoun Station Technical |
| Specification 30-day plugging report and 6-month inspection report will be |
| submitted as required." |
| |
| The NRC resident inspector has been informed of this retraction by the |
| licensee. Notified R4DO (Smith). |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Fuel Cycle Facility |Event Number: 36411 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PORTSMOUTH GASEOUS DIFFUSION PLANT |NOTIFICATION DATE: 11/07/1999|
| RXTYPE: URANIUM ENRICHMENT FACILITY |NOTIFICATION TIME: 22:36[EST]|
| COMMENTS: 2 DEMOCRACY CENTER |EVENT DATE: 11/07/1999|
| 6903 ROCKLEDGE DRIVE |EVENT TIME: 04:00[EST]|
| BETHESDA, MD 20817 (301)564-3200 |LAST UPDATE DATE: 11/09/1999|
| CITY: PIKETON REGION: 3 +-----------------------------+
| COUNTY: PIKE STATE: OH |PERSON ORGANIZATION |
|LICENSE#: GDP-2 AGREEMENT: N |JAMES CREED R3 |
| DOCKET: 0707002 |JOHN HICKEY NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: KURT SISLER | |
| HQ OPS OFFICER: STEVE SANDIN | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|NBNL RESPONSE-BULLETIN | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| 24-HOUR NRC BULLETIN 91-01 NOTIFICATION INVOLVING LOSS OF CRITICALITY |
| CONTROL (MODERATION) |
| |
| "On 11/7/99 at 0400 hours, it was discovered that the standard solution used |
| to calibrate X-344, Autoclave #2 conductivity system probes on 11/1/99 was |
| past its shelf life expiration date as stated on the certificate of NIST |
| traceability. This brings into question the operability (AQ-NCS boundary |
| item) of the conductivity system. Autoclave #2 was operated in Mode II |
| (Cylinder Heating) for approximately 30 minutes on 11/2/99. Conductivity |
| system as-found readings were performed on 11/2/99 with a standard solution |
| in date according to NIST requirements, The as-found results were within |
| tolerance indicating that the system would have performed its intended |
| safety function. |
| |
| "Nuclear Criticality Engineering has determined that the operation of the |
| autoclave, after calibration of the conductivity system with out-dated |
| conductivity standard solution, constitutes the loss of one (1) NCS control |
| (moderation). The other control (Mass) was maintained throughout the |
| duration of this event. The loss of one NCS control is reportable to the NRC |
| as a 24-hour event. |
| |
| "THERE WAS NO LOSS OF HAZARDOUS/RADIOACTIVE MATERIAL OR |
| RADIOACTIVE/RADIOLOGICAL CONTAMINATION EXPOSURE AS A RESULT OF THIS EVENT. |
| |
| "SAFETY SIGNIFICANCE OF EVENTS: |
| |
| "The safety significance of this event is low. The conductivity probes are |
| required to be operable and are tested semi-annually to verify this. The |
| probes were tested on 11/1/99 however, the solution used to calibrate the |
| probes was out of date. If the probes completely failed a UF6 release in |
| quantities greater than the minimum critical mass would have to occur or a |
| slow release could allow a dilute UO(2)F(2) solution to reach unfavorable |
| geometry storm drains. There was no UF6 release during this event and the |
| autoclave was only operated for 30 minutes in this condition (As-found |
| testing with in-date standard on 11/2/99 revealed that the conductivity |
| system was within allowable tolerance). |
| |
| "POTENTIAL CRITICALITY PATHWAYS INVOLVED (BRIEF SCENARIO[S] OF HOW |
| CRITICALITY COULD OCCUR): |
| |
| "The potential pathway to criticality is that a slow UF6 release occurs |
| (less than 2 pounds per minute) and the conductivity probes fail to detect |
| it. A release with greater than 2 pounds per minute would isolate the |
| autoclave due to high pressure. This slow release could allow a dilute |
| solution to reach unfavorable geometry storm drains. |
| |
| "CONTROLLED PARAMETERS (MASS, MODERATION, GEOMETRY, CONCENTRATION, ETC.): |
| |
| "The controlled parameters are mass and moderation. |
| |
| "ESTIMATED AMOUNT, ENRICHMENT, FORM OF LICENSED MATERIAL (INCLUDE PROCESS |
| LIMIT AND % WORST CASE OF CRITICAL MASS): |
| |
| "The estimated amount of material is zero because there was no release, the |
| maximum enrichment is 5% U235 and the form of material is UF6. |
| |
| "NUCLEAR CRITICALITY SAFETY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION |
| OF THE FAILURES OR DEFICIENCIES |
| |
| "The control that was lost was moderation. The conductivity probes were |
| calibrated with a solution that was out of date and therefore the probes |
| were inoperable. The autoclave was operated for 30 minutes in this |
| condition. |
| |
| "CORRECTIVE ACTIONS TO RESTORE SAFETY SYSTEM AND WHEN EACH WAS IMPLEMENTED: |
| |
| "As found readings were completed with an in-date batch of conductivity |
| standard solution. All as-found readings were in acceptable limits. |
| Autoclave #2 is inoperable (since 11/4/99) for reasons other than this |
| event." |
| |
| The standard solution used to calibrate the conductivity system expired on |
| 10/12/99. Operations has the cause for this incident report under review; |
| however, the preliminary investigation attributes the failure to personnel |
| error. |
| |
| The NRC Resident Inspector was informed. |
| |
| * * * UPDATE AT 1434 ON 11/9/99 BY SPAETH TAKEN BY WEAVER * * * |
| |
| Further evaluation has determined that although the autoclave was operated |
| with a conductivity system that was calibrated with an out-of-date buffer |
| solution, as-found data show that the system was operable and capable of |
| performing its safety function as required by the NCSA. Therefore, double |
| contingency controls remained in place throughout this incident. |
| |
| It should also be noted that the original event description incorrectly |
| stated that moderation and mass were the controlled parameters. The |
| controlled parameters in this case should have been reported as moderation |
| and geometry. Moderation control is based on the integrity of the UF6 |
| cylinder, and geometry control is based on the conductivity system |
| preventing UF6 from entering the unfavorable geometry floor drains. |
| |
| HOO notified R3DO (Parker) and NMSS (Piccone). |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Hospital |Event Number: 36415 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| REP ORG: FOREST PARK HOSPITAL |NOTIFICATION DATE: 11/09/1999|
|LICENSEE: FOREST PARK HOSPITAL |NOTIFICATION TIME: 10:30[EST]|
| CITY: ST. LOUIS REGION: 3 |EVENT DATE: 11/05/1999|
| COUNTY: STATE: MO |EVENT TIME: 18:30[CST]|
|LICENSE#: 24-00752-01 AGREEMENT: N |LAST UPDATE DATE: 11/09/1999|
| DOCKET: |+----------------------------+
| |PERSON ORGANIZATION |
| |MICHAEL PARKER R3 |
| |JOHN HICKEY NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: DAVID KEYS | |
| HQ OPS OFFICER: BOB STRANSKY | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|LADM 35.33(a) MED MISADMINISTRATION | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| MEDICAL MISADMINISTRATION |
| |
| The licensee reported that a patient received treatment of an incorrect site |
| as the result of an error in setting up the Nucletron HDR (high dose rate) |
| afterloader device. When the treatment simulation was run, a dwell setting |
| of 1.0 cm was used; however, when the actual treatment was administered, a |
| dwell setting of 0.5 cm was selected. This resulted in the actual treatment |
| site being displaced 5 cm from the intended site. The intended site received |
| less than 10% of the prescribed dose. The misadministration was discovered |
| at approximately 1600 CST on 11/8/1999. The licensee plans to revise |
| treatment procedures to ensure that the dwell setting used during treatment |
| planning is the same as that used during the administration of treatment. |
| |
| The licensee intends to continue treatment of the patient at a later date. |
| The licensee has contacted NRC Region III (Null) regarding this event. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36416 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: DUANE ARNOLD REGION: 3 |NOTIFICATION DATE: 11/09/1999|
| UNIT: [1] [] [] STATE: IA |NOTIFICATION TIME: 12:15[EST]|
| RXTYPE: [1] GE-4 |EVENT DATE: 11/09/1999|
+------------------------------------------------+EVENT TIME: 10:30[CST]|
| NRC NOTIFIED BY: BOB MURRELL |LAST UPDATE DATE: 11/09/1999|
| HQ OPS OFFICER: DOUG WEAVER +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |MICHAEL PARKER R3 |
|10 CFR SECTION: | |
|ADAS 50.72(b)(2)(i) DEG/UNANALYZED COND | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N N 0 Refueling |0 Refueling |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| RECIRCULATION RISER WELD FOUND CRACKED |
| |
| While performing ultrasonic exam of recirc riser weld RRD-F002 (nozzle to |
| safe-end weld) indications of Intergranular Stress Corrosion Cracking |
| (IGSCC) were identified. Specifically, an approximately 65% through-wall |
| crack was found on the 'D' RECIRC riser nozzle to safe-end weld. This |
| nozzle was being inspected as a part of an expanded inspection scope as a |
| result of a similar indication found on the 'B' RECIRC riser nozzle to |
| safe-end weld. To date, 5 out of 10 welds have been inspected. |
| |
| The licensee notified the NRC resident inspector. |
| |
| SEE RELATED EVENT: #36402. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36417 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: INDIAN POINT REGION: 1 |NOTIFICATION DATE: 11/09/1999|
| UNIT: [2] [] [] STATE: NY |NOTIFICATION TIME: 17:50[EST]|
| RXTYPE: [2] W-4-LP,[3] W-4-LP |EVENT DATE: 11/09/1999|
+------------------------------------------------+EVENT TIME: 12:40[EST]|
| NRC NOTIFIED BY: KEVIN DONNELLY |LAST UPDATE DATE: 11/09/1999|
| HQ OPS OFFICER: DOUG WEAVER +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |DAN HOLODY R1 |
|10 CFR SECTION: | |
|APRE 50.72(b)(2)(vi) OFFSITE NOTIFICATION | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|2 N Y 99 Power Operation |99 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| OFFSITE NOTIFICATION - WORKERS EXPOSED TO MERCURY |
| |
| The licensee notified the National Response Center, Environmental Protection |
| Agency RIV and the Tennessee Hotline of an incident involving mercury. A |
| radwaste shipment from Indian Point 2 was sent to GTS - Duratek in |
| Tennessee. A worker at GTS - Duratek was exposed to a small amount of |
| mercury while opening a bag. The shipment was marked as only radioactive |
| and was not supposed to contain any mercury. The licensee is investigating |
| to determine the source of the mercury. |
| |
| The licensee will notify the NRC resident inspector. |
+------------------------------------------------------------------------------+
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Page Last Reviewed/Updated Thursday, March 25, 2021