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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 THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED  !!!!!!!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.                         |
+------------------------------------------------------------------------------+