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
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|Power Reactor |Event Number: 36873 |
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+------------------------------------------------------------------------------+
| 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
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| "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). |
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|Fuel Cycle Facility |Event Number: 36885 |
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| 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
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| 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 |
+------------------------------------------------------------------------------+
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| 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
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| 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
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| 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. |
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