United States Nuclear Regulatory Commission - Protecting People and the Environment
Home > NRC Library > Document Collections > Reports Associated With Events > Event Notification Reports > 2000

Event Notification Report for April 14, 2000

                    U.S. Nuclear Regulatory Commission
                              Operations Center

                              Event Reports For
                           04/13/2000 - 04/14/2000

                              ** EVENT NUMBERS **

36873  36885  36886  36887  36888  36889  

+------------------------------------------------------------------------------+
|Power Reactor                                    |Event Number:   36873       |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: FERMI                    REGION:  3  |NOTIFICATION DATE: 04/07/2000|
|    UNIT:  [2] [] []                 STATE:  MI |NOTIFICATION TIME: 17:15[EDT]|
|   RXTYPE: [2] GE-4                             |EVENT DATE:        04/07/2000|
+------------------------------------------------+EVENT TIME:        14:00[EDT]|
| NRC NOTIFIED BY:  S. MAREK                     |LAST UPDATE DATE:  04/13/2000|
|  HQ OPS OFFICER:  JOHN MacKINNON               +-----------------------------+
+------------------------------------------------+PERSON          ORGANIZATION |
|EMERGENCY CLASS:          N/A                   |THOMAS KOZAK         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   |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|2     N          N       0        Refueling        |0        Refueling        |
|                                                   |                          |
|                                                   |                          |
+------------------------------------------------------------------------------+
                                   EVENT TEXT                                   
+------------------------------------------------------------------------------+
| "C" MAIN STEAM LINE FAILS LOCAL LEAK RATE TEST DUE TO EXCESSIVE LEAKAGE      |
|                                                                              |
| During Local Leak Rate Testing (LLRT) of the "C" Main Steam Line, the        |
| as-found leakage could not be quantified.  Technical Specification           |
| surveillance requirement, SR 3.6.1.3.12 limit of less than 100 scfh combined |
| MSIV leakage rate for all four main steam lines when tested at greater than  |
| 25 psig was exceeded.  Fermi 2 Main Steam Lines are equipped with a Main     |
| Steam Line Isolation Valve Leakage Control System (MSIVLCS) which is         |
| designed to maintain pressure between the MSIVs slightly above that of       |
| primary containment.  Since the leakage could not be quantified, it could    |
| not be demonstrated that the leakage did not exceed the capacity of the      |
| MSIVLCS.                                                                     |
|                                                                              |
| The NRC Resident Inspector was notified of this event by the licensee.       |
|                                                                              |
| * * * UPDATE ON 04/07/00 AT 1836 EDT BY S. MAREK TAKEN BY MACKINNON * * *    |
|                                                                              |
| During LLRT of the "D" Main Steam Line, the as-found leakage could not be    |
| quantified.  Technical Specification surveillance requirement, SR 3.6.1.3.12 |
| limit of less than 100 scfh combined MSIV leakage rate for all four main     |
| steam lines when tested at greater than 25 psig was exceeded.  Fermi 2 Main  |
| Steam Lines are equipped with a MSIVLCS which is designed to maintain        |
| pressure between the MSIVs slightly above that of primary containment.       |
| Since the leakage could not be quantified, it could not be demonstrated that |
| the leakage did not exceed the capacity of the MSIVLCS.  The "A" & "B" Main  |
| Steam Lines passed their LLRT.                                               |
|                                                                              |
| The NRC Resident Inspector was notified of this update by the licensee.      |
| R3DO (T. Kozak) was notified.                                                |
|                                                                              |
| * * * UPDATE AT 0931 ON 4/12/2000 FROM S. MAREK TAKEN BY STEVE SANDIN * * *  |
|                                                                              |
| "During Local Leak Rate Testing (LLRT) of Main Steam Line drain valve,       |
| B2103F019, the as found leakage was 232.7 SCFH, the allowable leakage is     |
| 1.00 SCFH.                                                                   |
|                                                                              |
| "The leakage through this penetration X-8 of 232.7 SCFH exceeds Technical    |
| Specification surveillance requirement, SR 3.6.1.3.11 Bypass Leakage Total   |
| limit of less than 0.04 La, (11.9) SCFH. The total containment leakage is    |
| 274.62 SCFH. which is less than 1.0 La of 297.30 SCFH.                       |
|                                                                              |
| "This report is being made in accordance with 10CFR 50.72(b)(2)(i), any      |
| event or condition found while the reactor is shutdown that, had it been     |
| found while the reactor was in operation, would have resulted in the nuclear |
| plant, including its principal safety barriers, being seriously degraded or  |
| being in an unanalyzed condition that significantly compromises plant        |
| safety."                                                                     |
|                                                                              |
| The NRC resident inspector has been informed of this update by the licensee. |
| Notified R3DO (Hiland).                                                      |
+------------------------------------------------------------------------------+

+------------------------------------------------------------------------------+
|Fuel Cycle Facility                              |Event Number:   36885       |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PADUCAH GASEOUS DIFFUSION PLANT      |NOTIFICATION DATE: 04/13/2000|
|   RXTYPE: URANIUM ENRICHMENT FACILITY          |NOTIFICATION TIME: 03:08[EDT]|
| COMMENTS: 2 DEMOCRACY CENTER                   |EVENT DATE:        04/12/2000|
|           6903 ROCKLEDGE DRIVE                 |EVENT TIME:        13:47[CDT]|
|           BETHESDA, MD 20817    (301)564-3200  |LAST UPDATE DATE:  04/13/2000|
|    CITY:  PADUCAH                  REGION:  3  +-----------------------------+
|  COUNTY:  McCRACKEN                 STATE:  KY |PERSON          ORGANIZATION |
|LICENSE#:  GDP-1                 AGREEMENT:  Y  |PATRICK HILAND       R3      |
|  DOCKET:  0707001                              |SUSAN SHANKMAN       NMSS    |
+------------------------------------------------+                             |
| NRC NOTIFIED BY:  CALVIN PITTMAN               |                             |
|  HQ OPS OFFICER:  FANGIE JONES                 |                             |
+------------------------------------------------+                             |
|EMERGENCY CLASS:          N/A                   |                             |
|10 CFR SECTION:                                 |                             |
|NBNL                     RESPONSE-BULLETIN      |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
+------------------------------------------------------------------------------+

                                   EVENT TEXT                                   
+------------------------------------------------------------------------------+
| NRC BULLETIN 91-01, 24 HOUR REPORT                                           |
|                                                                              |
| The following is quoted from the licensee's report:                          |
|                                                                              |
| 1-kg cylinders were discovered in the C-710 Isotopic Lab that violate the    |
| wall thickness design specification of NCSE 1493-03. The wall thickness      |
| credited in the NCSE is 0.109".  Wall thicknesses of some cylinders were     |
| discovered as low as 0.065". The wall thickness is credited in the           |
| criticality safety calculations to demonstrate double contingency.           |
|                                                                              |
| SAFETY SIGNIFICANCE OF EVENTS:                                               |
|                                                                              |
| A design feature limitation credited to ensure double contingency was        |
| exceeded. Calculations demonstrate that greater than 240 cylinders using a   |
| wall thickness of 0.065" of optimally moderated UO2F2 solution are safe.     |
| There are a total of 95 1-kg cylinders in the three storage cabinets in the  |
| Isotopic Lab.                                                                |
|                                                                              |
| POTENTIAL CRITICALITY PATHWAYS INVOLVED (BRIEF SCENARIO(S) OF HOW            |
| CRITICALITY COULD OCCUR:                                                     |
|                                                                              |
| In order for a criticality to be possible, the batch limitation would have   |
| to be exceeded by more than a factor of three. Additionally, the 1-kg        |
| cylinders would have to be filled with optimally moderated UO2F2 solution    |
| instead of the existing UF6.                                                 |
|                                                                              |
| CONTROLLED PARAMETERS (MASS, MODERATION, GEOMETRY, CONCENTRATION, ETC.):     |
|                                                                              |
| Double contingency for this scenario is established by implementing          |
| interaction and geometry controls.                                           |
|                                                                              |
| ESTIMATED AMOUNT, ENRICHMENT, FORM OF LICENSED MATERIAL (INCLUDE PROCESS     |
| LIMIT AND % WORST CASE CRITICAL MASS):                                       |
|                                                                              |
| There are 95 1-kg cylinders in C-710 only some of which have been determined |
| to have inadequate wall thickness. The assay of these cylinders varies from  |
| less than 1% U235 to approximately 4.6% U235. The material contained in      |
| these cylinders is UF6.                                                      |
|                                                                              |
| NUCLEAR CRITICALITY SAFETY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION   |
| OF THE FAILURES OR DEFICIENCIES:                                             |
|                                                                              |
| The first leg of double contingency relies on interaction control through    |
| the application of batch limits. This control was not violated and the first |
| leg of double contingency was maintained.                                    |
|                                                                              |
| The second leg of double contingency is based on geometry control. This is   |
| controlled through implementation of design specifications for the 1-kg      |
| cylinder. The actual wall thickness was discovered to be less than that      |
| credited in the design features. Therefore, the geometry process parameter   |
| limit was exceeded.                                                          |
|                                                                              |
| The geometry process parameter was violated, therefore double contingency    |
| was not maintained.                                                          |
|                                                                              |
| CORRECTIVE ACTIONS TO RESTORE SAFETY SYSTEMS AND WHEN EACH WAS IMPLEMENTED:  |
|                                                                              |
| This area is being controlled to ensure that no fissile material is moved    |
| within two feet of this storage area. NCS is in the process of developing a  |
| remediation plan to correct this condition.                                  |
|                                                                              |
| NUCLEAR CRITICALITY SAFETY CONTROLS INVOLVED AND THEIR IMPACT ON DOUBLE      |
| CONTINGENCY:                                                                 |
|                                                                              |
| Double contingency for this scenario is established by implementing          |
| interaction and geometry controls.                                           |
|                                                                              |
| The first leg of double contingency relies on interaction control through    |
| the application of batch limits. This control was not violated and the first |
| leg of double contingency was maintained.                                    |
|                                                                              |
| The second leg of double contingency is based on geometry control, This is   |
| controlled through implementation of design specifications for the 1-kg      |
| cylinder. The actual wall thickness was discovered to be less than that      |
| credited in the design features. Therefore, the geometry process parameter   |
| limit was exceeded.                                                          |
|                                                                              |
| The geometry process parameter was violated therefore double contingency was |
| not maintained.                                                              |
|                                                                              |
| POTENTIAL CRITICALITY PATHWAYS INVOLVED;                                     |
|                                                                              |
| In order for a criticality to be possible, the batch limits would have to be |
| exceeded by more than a factor of three. Additionally, the 1-kg cylinders    |
| would have to be filled with optimally moderated UO2F2 solution instead of   |
| the existing UF6.                                                            |
|                                                                              |
| SAFETY SIGNIFICANCE OF INCIDENT:                                             |
|                                                                              |
| A design feature limitation credited to ensure double contingency was        |
| exceeded. Calculations demonstrate that greater than 240 cylinders using a   |
| wall thickness of 0.065" of optimally moderated UO2F2 solution are safe.     |
| There are a total 95 1-kg cylinders in the three storage cabinets in the     |
| Isotopic Lab.                                                                |
|                                                                              |
| EXCLUSION ZONE AND POSTINGS:                                                 |
|                                                                              |
| Post the area as follows in accordance with CP2-EG-NS1031. Ensure all four   |
| sides including areas on opposite sides of adjacent walls less than 2-feet   |
| from the storage cabinets.                                                   |
|                                                                              |
| The licensee notified the NRC Resident Inspector.                            |
+------------------------------------------------------------------------------+

+------------------------------------------------------------------------------+
|Fuel Cycle Facility                              |Event Number:   36886       |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: MALLINCKRODT INC.                    |NOTIFICATION DATE: 04/13/2000|
|   RXTYPE: NUCLEAR PHARMACY                     |NOTIFICATION TIME: 13:59[EDT]|
| COMMENTS: RADIOPHARMACEUTICAL OPERATIONS       |EVENT DATE:        03/31/2000|
|           MEDICAL R&D                          |EVENT TIME:        12:00[CDT]|
|                                                |LAST UPDATE DATE:  04/13/2000|
|    CITY:  MARYLAND HTS.            REGION:  3  +-----------------------------+
|  COUNTY:  ST. LOUIS                 STATE:  MO |PERSON          ORGANIZATION |
|LICENSE#:  24-4206-01            AGREEMENT:  N  |PATRICK HILAND       R3      |
|  DOCKET:  03000001                             |BRAIN SMEITH         NMSS    |
+------------------------------------------------+FRANK CONGEL         IRO     |
| NRC NOTIFIED BY:  JIM SCHUH                    |                             |
|  HQ OPS OFFICER:  JOHN MacKINNON               |                             |
+------------------------------------------------+                             |
|EMERGENCY CLASS:          N/A                   |                             |
|10 CFR SECTION:                                 |                             |
|BAD1 20.2202(a)(1)       PERS OVEREXPOSURE      |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
+------------------------------------------------------------------------------+

                                   EVENT TEXT                                   
+------------------------------------------------------------------------------+
| EVENT THREATENS TO CAUSE A SHALLOW-DOSE EQUIVALENT TO THE LEFT HAND OF 250   |
| RADS (2.5 Gy).                                                               |
|                                                                              |
| The Mallinckrodt Radiation Safety Officer called in an immediate             |
| notification under 10 CFR 20.2202(a)(1)(iii).                                |
|                                                                              |
| A Ring Badge from the right index finger of an employee had a reading of     |
| 5685 mrem.  On 03/31/00 a Mallinckrodt employee working in the generator     |
| manufacturing line facility handled a column containing 19 curies of Mo-99   |
| with his left hand.  The individual was supposed to use forceps to           |
| manipulate needles inside the generator but instead used his fingers.        |
|                                                                              |
| The calculated dose to his right index finger tip is 31 rem at 1.5" from the |
| source of activity. The employee recreated the event from which the licensee |
| concluded that the finger tips of the left hand were intermittently in       |
| contact with the Mo-99 generator column over a span of 30 minutes. At this   |
| time the Radiation Safety Officer postulates that the dose to the fingers of |
| the individual's left hand may exceed 250 rads. The individuals whole body   |
| dose has not been calculated.                                                |
+------------------------------------------------------------------------------+

+------------------------------------------------------------------------------+
|Power Reactor                                    |Event Number:   36887       |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: RIVER BEND               REGION:  4  |NOTIFICATION DATE: 04/13/2000|
|    UNIT:  [1] [] []                 STATE:  LA |NOTIFICATION TIME: 14:19[EDT]|
|   RXTYPE: [1] GE-6                             |EVENT DATE:        04/13/2000|
+------------------------------------------------+EVENT TIME:        10:04[CDT]|
| NRC NOTIFIED BY:  GLENN KRAUSE                 |LAST UPDATE DATE:  04/13/2000|
|  HQ OPS OFFICER:  STEVE SANDIN                 +-----------------------------+
+------------------------------------------------+PERSON          ORGANIZATION |
|EMERGENCY CLASS:          N/A                   |CHARLES PAULK        R4      |
|10 CFR SECTION:                                 |                             |
|AESF 50.72(b)(2)(ii)     ESF ACTUATION          |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR|   INIT RX MODE  |CURR PWR|  CURR RX MODE   |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1     N          Y       78       Power Operation  |78       Power Operation  |
|                                                   |                          |
|                                                   |                          |
+------------------------------------------------------------------------------+
                                   EVENT TEXT                                   
+------------------------------------------------------------------------------+
| UNEXPECTED REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM ISOLATION DURING     |
| SURVEILLANCE TESTING                                                         |
|                                                                              |
| "At 1004 on 04/13/00 during the performance of I&C STP 207-4539 (RCIC        |
| Isolation - RCIC steam supply pressure low channel functional test on E31    |
| N685B) received a Division II isolation of the RCIC system.  The isolation   |
| appears to be invalid.  E51-F063 (RCIC Steam Supply Inboard Isolation Valve) |
| and RCIC Trip and Throttle Valve went from open to closed as designed.       |
| Investigation is ongoing to determine the cause."                            |
|                                                                              |
| RCIC was declared inoperable placing the unit in a 14-day LCO A/S.           |
|                                                                              |
| The licensee informed the NRC Resident Inspector.                            |
+------------------------------------------------------------------------------+

+------------------------------------------------------------------------------+
|Power Reactor                                    |Event Number:   36888       |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: WATERFORD                REGION:  4  |NOTIFICATION DATE: 04/13/2000|
|    UNIT:  [3] [] []                 STATE:  LA |NOTIFICATION TIME: 17:13[EDT]|
|   RXTYPE: [3] CE                               |EVENT DATE:        04/13/2000|
+------------------------------------------------+EVENT TIME:        15:35[CDT]|
| NRC NOTIFIED BY:  E. LEMKE                     |LAST UPDATE DATE:  04/13/2000|
|  HQ OPS OFFICER:  JOHN MacKINNON               +-----------------------------+
+------------------------------------------------+PERSON          ORGANIZATION |
|EMERGENCY CLASS:          N/A                   |CHARLES PAULK        R4      |
|10 CFR SECTION:                                 |                             |
|AINC 50.72(b)(2)(iii)(C) POT UNCNTRL RAD REL    |                             |
|                                                |                             |
|                                                |                             |
|                                                |                             |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR|   INIT RX MODE  |CURR PWR|  CURR RX MODE   |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|3     N          Y       100      Power Operation  |100      Power Operation  |
|                                                   |                          |
|                                                   |                          |
+------------------------------------------------------------------------------+
                                   EVENT TEXT                                   
+------------------------------------------------------------------------------+
| FEEDWATER ISOLATION VALVES MAY CLOSE FASTER THAN 1.5-SECOND DESIGN BASIS     |
| LIMIT.                                                                       |
|                                                                              |
| On March 21, 2000, Waterford Unit 3 determined that, based on a new          |
| calculation methodology and the latest stroke time data, the Feedwater       |
| Isolation Valves (FWIVs) FW-184 A (B) may close faster than the 1.5-second   |
| design basis limit.                                                          |
|                                                                              |
| The physical plant was determined to be operable and the FWIVs would have    |
| performed their intended safety function at the time the condition was       |
| identified.  An initial operability evaluation was made in accordance with   |
| procedure W4.101, which determined that the valves were operable at the time |
| the evaluation was conducted.  This was based on an engineering evaluation   |
| that determined that the faster closure of the FWIVs would not result in     |
| water hammer loads that would prevent the FWIVs and their associated         |
| penetrations from performing their required safety function.  The            |
| engineering evaluation determined that the analyzed increase in fast valve   |
| closure (FVC) load is 57% for FW-184A and 52.9% for FW-184B.                 |
|                                                                              |
| A subsequent evaluation was performed for FW-184A (B) to determine if at any |
| time in the last two years the increase in FVC load may have exceeded the    |
| allowable loads determined by the engineering evaluation for W4.101.  That   |
| evaluation determined on one occasion for FW-184A and six occasions for      |
| FW-184B, the percent increases for the FWIVs and subsequent stroke times     |
| exceeded the values provided in the operability determination                |
|                                                                              |
| On these occasions in question, closure of FW-184A (B) in response to the    |
| most adverse accident scenario could have potentially produced water hammer  |
| that may have exceeded the capability of  piping supports in the FW system   |
| between the SGs and FW-184A (B). This may have resulted in the subsequent    |
| loss of containment isolation function of FW-184A (B).                       |
|                                                                              |
| The NRC Resident Inspector was notified of this event by the licensee.       |
+------------------------------------------------------------------------------+

+------------------------------------------------------------------------------+
|Power Reactor                                    |Event Number:   36889       |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: POINT BEACH              REGION:  3  |NOTIFICATION DATE: 04/13/2000|
|    UNIT:  [] [2] []                 STATE:  WI |NOTIFICATION TIME: 17:54[EDT]|
|   RXTYPE: [1] W-2-LP,[2] W-2-LP                |EVENT DATE:        04/13/2000|
+------------------------------------------------+EVENT TIME:        16:10[CDT]|
| NRC NOTIFIED BY:  RANDY HASTINGS               |LAST UPDATE DATE:  04/13/2000|
|  HQ OPS OFFICER:  JOHN MacKINNON               +-----------------------------+
+------------------------------------------------+PERSON          ORGANIZATION |
|EMERGENCY CLASS:          N/A                   |PATRICK HILAND       R3      |
|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   |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|                                                   |                          |
|2     N          Y       100      Power Operation  |100      Power Operation  |
|                                                   |                          |
+------------------------------------------------------------------------------+
                                   EVENT TEXT                                   
+------------------------------------------------------------------------------+
| BOTH STEAM GENERATOR PRESSURE TRANSMITTERS WOULD NOT BE AVAILABLE AS A       |
| RESULT OF A POSTULATED FIRE.                                                 |
|                                                                              |
|                                                                              |
| It was discovered that the 2A Steam Generator pressure transmitter 2PT-469   |
| would not be available as a result of a postulated fire event in the North   |
| Section of the Primary Auxiliary Building, 26 foot elevation.  2PT-469 is a  |
| redundant Appendix R instrument to 2PT-483,  2B Steam Generator pressure     |
| transmitter.  2PT-483 was already known to be lost due to this postulated    |
| fire event.  The loss of both pressure transmitters places the plant outside |
| the design basis for Appendix R.                                             |
|                                                                              |
| This was discovered during the Appendix R Rebaselining Project review.       |
| Compensatory actions include a fire watch in the appropriate fire zone       |
| within one hour of discovery and twice per shift thereafter.  As a long term |
| corrective action Point Beach is already pursuing a plant modification to    |
| re-route pressure transmitter 2PT-483 cables.                                |
|                                                                              |
| The NRC Resident Inspector was notified of this event by the licensee.       |
+------------------------------------------------------------------------------+