Event Notification Report for September 3, 2002
U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
08/30/2002 - 09/03/2002
** EVENT NUMBERS **
39155 39156 39159 39161 39162 39163 39164
+------------------------------------------------------------------------------+
|General Information or Other |Event Number: 39155 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| REP ORG: MA RADIATION CONTROL PROGRAM |NOTIFICATION DATE: 08/27/2002|
|LICENSEE: NITON CORPORATION |NOTIFICATION TIME: 16:06[EDT]|
| CITY: BILLERICA REGION: 1 |EVENT DATE: 08/27/2002|
| COUNTY: STATE: MA |EVENT TIME: [EDT]|
|LICENSE#: 55-0238 AGREEMENT: Y |LAST UPDATE DATE: 08/27/2002|
| DOCKET: |+----------------------------+
| |PERSON ORGANIZATION |
| |JAMES NOGGLE R1 |
| |E. WILLIAM BRACH NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: JOSH DAHLER | |
| HQ OPS OFFICER: FANGIE JONES | |
+------------------------------------------------+ |
|EMERGENCY CLASS: NON EMERGENCY | |
|10 CFR SECTION: | |
|NAGR AGREEMENT STATE | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| AGREEMENT STATE REPORT - REPORT OF LEAKING SEALED SOURCE |
| |
| "On August 27, 2002, Niton Corporation provided a report of a leaking sealed |
| source that was discovered by Niton on July 5, 2002 during removal of a |
| source from a Niton XLi Series X-ray fluorescence device source holder. The |
| seal source is AEA Model IEC.A1, S/N 8740LG containing Fe-55. The estimated |
| leakage was 130,000 dpm (0.059 microcuries). |
| |
| "The leaking source was placed in a lead pig. The source holder was |
| decontaminated. The workbench and surrounding area were surveyed and found |
| to be free of contamination. The leaking source and all decontamination |
| material were placed into the licensee's waste stream." |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|General Information or Other |Event Number: 39156 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| REP ORG: CALIFORNIA RADIATION CONTROL PRGM |NOTIFICATION DATE: 08/28/2002|
|LICENSEE: KLEINFELDER, INC. |NOTIFICATION TIME: 00:35[EDT]|
| CITY: DIAMOND BAR REGION: 4 |EVENT DATE: 08/26/2002|
| COUNTY: STATE: CA |EVENT TIME: [PDT]|
|LICENSE#: 6852-19 AGREEMENT: Y |LAST UPDATE DATE: 08/28/2002|
| DOCKET: |+----------------------------+
| |PERSON ORGANIZATION |
| |DAVID GRAVES R4 |
| |E. WILLIAM BRACH NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: ROBERT GREGER | |
| HQ OPS OFFICER: MIKE NORRIS | |
+------------------------------------------------+ |
|EMERGENCY CLASS: NON EMERGENCY | |
|10 CFR SECTION: | |
|NAGR AGREEMENT STATE | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| AGREEMENT STATE REPORT INVOLVING A LOST/STOLEN TROXLER GAUGE |
| |
| A Troxler gauge, model 3411, was found missing from a storage building at a |
| Redlands, CA location. The gauge, serial number 12878, contains Cs-137 and |
| Am-241/Be. Local law enforcement has been notified and a reward has been |
| offered for it's return. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|General Information or Other |Event Number: 39159 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| REP ORG: ROSEMOUNT NUCLEAR INSTRUMENTS, INC |NOTIFICATION DATE: 08/30/2002|
|LICENSEE: ROSEMOUNT NUCLEAR INSTRUMENTS, INC |NOTIFICATION TIME: 14:05[EDT]|
| CITY: EDEN PRAIRIE REGION: 3 |EVENT DATE: 08/30/2002|
| COUNTY: STATE: MN |EVENT TIME: [CDT]|
|LICENSE#: AGREEMENT: N |LAST UPDATE DATE: 08/30/2002|
| DOCKET: |+----------------------------+
| |PERSON ORGANIZATION |
| |RONALD GARDNER R3 |
| |JAMES NOGGLE R1 |
+------------------------------------------------+STEPHEN CAHILL R2 |
| NRC NOTIFIED BY: JEFFREY W. SCHMITT |DAVID GRAVES R4 |
| HQ OPS OFFICER: STEVE SANDIN |VERN HODGE NRR |
+------------------------------------------------+ |
|EMERGENCY CLASS: NON EMERGENCY | |
|10 CFR SECTION: | |
|CCCC 21.21 UNSPECIFIED PARAGRAPH | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| PART 21 INVOLVING NON-CONFORMING ROSEMOUNT PRESSURE TRANSMITTERS AND SPARE |
| PARTS |
| |
| The following information was received via fax: |
| |
| "Re: Notification under 10 CFR Part 21 for Model 1153 and 1154 pressure |
| transmitters with optional adjustable damping circuit boards, and spare |
| adjustable damping circuit boards |
| |
| "Pursuant to 10 CFR Part 21, section 21.21(b), Rosemount Nuclear |
| Instruments, Inc. (RNII) is writing to inform you that: |
| |
| (a) certain Model 1153 and 1154 pressure transmitters with special |
| adjustable damping options N0037, N0039, N0058, N0085, N0102, N0108, N0120 |
| shipped between December 2000 and July 2002, and |
| |
| (b) certain spare adjustable damping circuit boards with part number |
| 01154-0021-0004 shipped between December 2000 and July 2002, may not meet |
| RNII's published Nuclear Steam Pressure/Temperature specification and/or |
| RNII's published Post DBE Operation specification. Model 1153 and 1154 |
| transmitters which do not contain the special optional adjustable damping |
| circuit board are not affected. The adjustable damping feature establishes |
| the time constant of the transmitter's output. A list of affected Model |
| 1153 and 1154 transmitters and spare adjustable damping circuit boards |
| shipped to the end user's facility is provided in Attachment A. |
| |
| "1.0 Name and address of the individual providing the information: |
| |
| Mr. Jeffrey W. Schmitt |
| Vice President & General Manager |
| Rosemount Nuclear Instruments, Inc. |
| 12001 Technology Drive |
| Eden Prairie, MN 55344 |
| |
| "2.0 Identification of items supplied: |
| |
| Model 1153 and 1154 Pressure Transmitters with special adjustable damping |
| options N0037, N0039, N0058, N0085, N0102, N0108, N0120 as identified by the |
| appropriate model code. The applicable model codes begin with either '1153' |
| or '1154' and end with the applicable N option number, for example |
| '1154DP5RAN0037.' |
| |
| and; |
| |
| Spare adjustable damping circuit boards with part number 01154-0021-0004. |
| |
| See Attachment A for complete listing of affected equipment shipped to the |
| end user's facility. |
| |
| "3.0 Identification of firm supplying the item: |
| |
| Rosemount Nuclear Instruments, Inc. |
| 12001 Technology Drive |
| Eden Prairie, MN 55344 |
| |
| "4.0 Nature of the failure and potential safety hazard: |
| |
| This notification impacts: |
| |
| (a) certain Model 1153 and 1154 pressure transmitters with special |
| adjustable damping options N0037, N0039, N0058, N0085, N0102, N0108, N0120 |
| shipped between December 2000 and July 2002 and |
| |
| (b) certain spare adjustable damping circuit boards with part number |
| 01154-0021-0004 shipped between December 2000 and July 2002, which may not |
| meet RNII's published Nuclear Steam Pressure/Temperature specification |
| and/or RNII's published Post DBE Operation specification. |
| |
| American Capacitor Corporation, the manufacturer of a capacitor used on the |
| adjustable damping circuit board recently informed RNII that capacitors |
| shipped to RNIIs adjustable damping circuit board supplier after June 2000 |
| were inadequately tested for conformance to the required high temperature |
| insulation resistance (IR) as specified by RNII's procurement documents. |
| Subsequent RNII testing confirmed that the capacitors were non-conforming to |
| RNII's IR specifications. These non-conforming capacitors were assembled |
| into certain transmitters and spare circuit boards that were shipped from |
| RNII between December 2000 and July 2002. Further RNII testing demonstrated |
| that model 1153 and 1154 transmitters with the optional adjustable damping |
| circuit boards containing non-conforming capacitors would not meet RNIIs |
| published Nuclear Steam Pressure/Temperature and Post DBE Operation |
| specifications for the published temperature profile. Due to the transmitter |
| performance variation present in the test results, RNII cannot provide an |
| interim specification. |
| |
| Testing has demonstrated that affected adjustable damping circuit boards do |
| meet product specifications at normal operating temperatures (40-200 degrees |
| F) |
| |
| RNII does not have sufficient information relative to the end user's |
| specific applications to determine the potential safety-related impact to |
| its plant. The end user must determine the impact on its plant operations |
| and plant safety and take action as deemed necessary. |
| |
| "5.0 The corrective action which is taken, the name of the individual or |
| organization responsible for that action, and the length of time taken to |
| complete that action: |
| |
| (a) Upon notification from the capacitor manufacturer of this issue, RNII |
| placed holds on all shipments containing potentially affected equipment |
| (model 1153 and 1154 transmitters with the special adjustable damping |
| options N0037, N0039, N0058, N0085, N0102, N0108, N0120 and spare adjustable |
| damping circuit boards with part number 01154-0021-0004). |
| |
| (b) The capacitor manufacturer identified the specific capacitor lots that |
| were inadequately tested. Only adjustable damping circuit boards built with |
| capacitors from these lots are affected. |
| |
| (c) RNII performed subsequent testing and established that Model 1153 and |
| 1154 transmitters with special adjustable damping options N0037, N0039, |
| N0058, N0085, N0102, N0108, N0120 having adjustable damping circuit boards |
| containing capacitors that do not meet RNII IR specifications would not meet |
| the published Nuclear Steam Pressure/Temperature specification and/or Post |
| DBE Operation specification for the published temperature profile. Testing |
| has demonstrated that affected adjustable damping circuit boards do meet |
| product specifications at normal operating temperatures (40-200 degrees F), |
| |
| (d) RNII has initiated actions with the adjustable damping circuit board |
| supplier to ensure that all capacitors received from sub-suppliers are |
| tested to demonstrate compliance with the IR specifications prior to |
| installation on adjustable damping circuit boards. |
| |
| (e) Replacement capacitors have been identified and tested which conform to |
| RNII's procurement specifications. RNII will audit the quality system of the |
| new supplier to assure its quality system meets the requirements of 10CFR |
| Part 50 Appendix B and all requirements of a qualified supplier to RNII. |
| Completion: 6 September 2002. |
| |
| (f) RNII will repair or replace any affected adjustable damping circuit |
| board at the end user's request. Completion: As required. |
| |
| "6.0 Any advice related to the potential failure of the item: |
| |
| "The end user must determine the impact of this deviation upon its plant's |
| operation and safety and take action as deemed necessary. If the end user |
| determines that replacement of the affected special adjustable damping |
| circuit boards is required, please contact Rosemount Nuclear Instruments, |
| Inc. to obtain a replacement adjustable damping circuit board." |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Fuel Cycle Facility |Event Number: 39161 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PORTSMOUTH GASEOUS DIFFUSION PLANT |NOTIFICATION DATE: 08/30/2002|
| RXTYPE: URANIUM ENRICHMENT FACILITY |NOTIFICATION TIME: 17:02[EDT]|
| COMMENTS: 2 DEMOCRACY CENTER |EVENT DATE: 08/30/2002|
| 6903 ROCKLEDGE DRIVE |EVENT TIME: 01:05[EDT]|
| BETHESDA, MD 20817 (301)564-3200 |LAST UPDATE DATE: 08/30/2002|
| CITY: PIKETON REGION: 3 +-----------------------------+
| COUNTY: PIKE STATE: OH |PERSON ORGANIZATION |
|LICENSE#: GDP-2 AGREEMENT: N |RONALD GARDNER R3 |
| DOCKET: 0707002 |E. WILLIAM BRACH NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: ERIC SPAETH | |
| HQ OPS OFFICER: STEVE SANDIN | |
+------------------------------------------------+ |
|EMERGENCY CLASS: NON EMERGENCY | |
|10 CFR SECTION: | |
|NONR OTHER UNSPEC REQMNT | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| REPORT INVOLVING A VALID ACTUATION OF A "Q" SAFETY SYSTEM |
| |
| "At 0105 hrs, Autoclave (AC) #4 in the X-344 facility experienced a steam |
| shutdown due to condensate level alarms A and B actuating. AC #4 was in an |
| applicable TSR mode II (heating) when the alarms actuated. This is |
| considered a valid actuation of a 'Q' safety system. The AC was placed in |
| mode 7 (shutdown). An investigation is underway to determine the cause of |
| the actuation." |
| |
| Operations informed both the DOE Site Representative and the NRC Resident |
| Inspector. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Fuel Cycle Facility |Event Number: 39162 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PADUCAH GASEOUS DIFFUSION PLANT |NOTIFICATION DATE: 08/31/2002|
| RXTYPE: URANIUM ENRICHMENT FACILITY |NOTIFICATION TIME: 16:52[EDT]|
| COMMENTS: 2 DEMOCRACY CENTER |EVENT DATE: 08/31/2002|
| 6903 ROCKLEDGE DRIVE |EVENT TIME: 01:20[CDT]|
| BETHESDA, MD 20817 (301)564-3200 |LAST UPDATE DATE: 08/31/2002|
| CITY: PADUCAH REGION: 3 +-----------------------------+
| COUNTY: McCRACKEN STATE: KY |PERSON ORGANIZATION |
|LICENSE#: GDP-1 AGREEMENT: Y |RONALD GARDNER R3 |
| DOCKET: 0707001 |E. WILLIAM BRACH NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: K. A. BEASLEY | |
| HQ OPS OFFICER: MIKE NORRIS | |
+------------------------------------------------+ |
|EMERGENCY CLASS: NON EMERGENCY | |
|10 CFR SECTION: | |
|NBNL RESPONSE-BULLETIN | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| NRC BULLETIN 91-01 24 HOUR REPORT - IMPROPER ISOLATION OF PUMP PRESSURE |
| ALARMS |
| |
| The following is taken from a facsimile report. |
| |
| At 0120, on 8-31-02 the Plant Shift Superintendent (PSS) determined that a |
| loss of double contingency had occurred in the C-337 process building. In |
| preparation for maintenance work on the #2 dual Speed Purge and Evacuation |
| (P&E) pump Instrument Maintenance had been requested to disable the two |
| delta pressure alarms associated with this pump. The delta pressure alarms |
| are safety related items (SRI). The SRI requires that both delta pressure |
| alarms be functional while the P&E pump is isolated from the cascade. |
| Building Operations personnel later discovered that one of the delta |
| pressure alarms that had been disabled was an alarm for the #3 P&E pump |
| which was not undergoing maintenance. Both delta-pressure alarms for the #3 |
| P&E pump were required to be in service at the time. The C-337 Front Line |
| Manager and Instrument Maintenance responded to the alarm panel and restored |
| the alarm to the #3 P&E pump. At that time, double contingency was |
| restored. |
| |
| This event resulted in a loss of one leg of double contingency in which |
| double contingency was restored within 4 hours. |
| |
| The NRC Resident Inspector has been notified. |
| |
| SAFETY SIGNIFICANCE OF EVENTS: |
| |
| While the R-114 was demonstrated to be dry, an SRI for double contingency |
| was rendered not functional without the knowledge of the operator. |
| |
| POTENTIAL CRITICALITY PATHWAYS INVOLVED (BRIEF SCENARIO(S) OF HOW |
| CRITICALITY COULD OCCUR) |
| |
| In order for a criticality to occur, an unsafe mass uranium deposit would |
| have to be present in the process gas system, an R-114 leak to the process |
| gas system would have to occur, the R-114 would have to contain an |
| unacceptable amount of moisture and moderate the deposit. |
| |
| CONTROLLED PARAMETERS (MASS, MODERATION, GEOMETRY. CONCENTRATION, ETC |
| |
| Double contingency is maintained by implementing two controls on |
| moderation. |
| |
| ESTIMATED AMOUNT, ENRICHMENT, FORMS OF LICENSED MATERIAL (INCLUDES PROCESS |
| LIMIT AND % WORST CASE CRITICAL MASS): |
| |
| No known mass of licensed material exists in the #3 Dual Speed P & E pump. |
| |
| NUCLEAR CRITICALITY SAFETY CONTROL SYSTEM(S) OR CONTROL SYSTEM(S) AND |
| DESCRIPTION OF THE FAILURES OR DEFICIENCIES |
| |
| The first leg of double contingency is based on the delta pressure switch |
| alarming to alert the operator to take action to either verify the R-114 is |
| dry or to isolate the RCW and remove the R-114 from the system in order to |
| prevent water from leaking into the process gas system. This delta pressure |
| alarm was maintained as operable, the SRI was maintained and the control was |
| not violated. |
| |
| The second leg of double contingency is based on the independent delta |
| pressure switch alarming to alert the operator to take action to either |
| verify the R-114 is dry or to isolate the RCW and remove the R-114 from the |
| system in order to prevent water from leaking into the process gas system. |
| The independent delta pressure alarm was disconnected without first |
| isolating the RCW system violating the SRI. The Independent delta pressure |
| switch was reconnected immediately upon discovery. This re-established |
| double contingency. |
| |
| Since double contingency is based on two controls (i.e. SRIs) on one |
| parameter, double contingency was not maintained. |
| |
| CORRECTIVE ACTIONS TO RESTORE SAFETY SYSTEMS AND WHEN EACH WAS IMPLEMENTED: |
| |
| 1. Make the #3 dual speed P&E alarm operable. |
| 2. Independently sample the #3 dual speed P&E R-114 system for moisture. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|General Information or Other |Event Number: 39163 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| REP ORG: U. S. STEEL-GARY WORKS |NOTIFICATION DATE: 09/01/2002|
|LICENSEE: U. S. STEEL |NOTIFICATION TIME: 14:42[EDT]|
| CITY: GARY REGION: 3 |EVENT DATE: 09/01/2002|
| COUNTY: LAKE STATE: IN |EVENT TIME: 05:02[CST]|
|LICENSE#: 13-079664-08 AGREEMENT: N |LAST UPDATE DATE: 09/01/2002|
| DOCKET: |+----------------------------+
| |PERSON ORGANIZATION |
| |RONALD GARDNER R3 |
| |E. WILLIAM BRACH NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: BRIAN KUNKLE | |
| HQ OPS OFFICER: GERRY WAIG | |
+------------------------------------------------+ |
|EMERGENCY CLASS: NON EMERGENCY | |
|10 CFR SECTION: | |
|NONR OTHER UNSPEC REQMNT | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| DAMAGED SOURCE HOLDER ON STEEL CASTER LEVEL GAUGE |
| |
| A radioactive source holder on a steel caster unit was damaged from metal |
| overflow at the U. S. Steel - Gary Works, Steel North Facility #2 Caster, |
| located in Gary, Indiana. The gauge is a Berthold Mold Level Gauge, Model |
| LB300ML, Serial Number 2849-12-96. Originally a 10 millicurie Cobalt 60 (Dec |
| 1996) source, it is currently estimated by the licensee to be about 5 |
| millicuries. The damage was limited to the source holder outer casing. |
| Neither the shutter closing mechanism nor the source shielding (tungsten) |
| were damaged. The damaged area involved a top plate where a lock can be |
| applied to the shutter mechanism and steel lifting lugs. The source was |
| verified in the source holder and the shutter closed. A radiation survey of |
| the area was conducted and no contamination or abnormal readings were |
| detected. The licensee has secured the shielded source and is making |
| arrangements for repair. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 39164 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: DUANE ARNOLD REGION: 3 |NOTIFICATION DATE: 09/02/2002|
| UNIT: [1] [] [] STATE: IA |NOTIFICATION TIME: 22:22[EDT]|
| RXTYPE: [1] GE-4 |EVENT DATE: 09/02/2002|
+------------------------------------------------+EVENT TIME: 21:00[CDT]|
| NRC NOTIFIED BY: CLARK |LAST UPDATE DATE: 09/02/2002|
| HQ OPS OFFICER: CHAUNCEY GOULD +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: NON EMERGENCY |RONALD GARDNER R3 |
|10 CFR SECTION: | |
|ASHU 50.72(b)(2)(i) PLANT S/D REQD BY TS | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 9 Power Operation |7 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| PLANT ENTERED TS 3.5.3 DUE TO THE RCIC TURBINE SYSTEM BEING DECLARED |
| INOPERABLE |
| |
| The plant commenced reactor shutdown at 2100 CDT due to the RCIC turbine |
| bearing oil system not maintaining the proper oil level. They are currently |
| troubleshooting the problem but have not yet determine the cause. The TS |
| 3.5.3 requires the plant to be in mode 3 by 0028 on 9/3/02 and reduce |
| reactor steam pressure to less than or equal to 150 lbs. By 0028 on 9/4/02. |
| |
| The NRC Resident Inspector was notified. |
+------------------------------------------------------------------------------+
Page Last Reviewed/Updated Wednesday, March 24, 2021
Page Last Reviewed/Updated Wednesday, March 24, 2021