Event Notification Report for November 12, 1999
U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
11/10/1999 - 11/12/1999
** EVENT NUMBERS **
36331 36336 36338 36418 36419 36420 36421 36422 36423 36424 36425 36426
36427
+------------------------------------------------------------------------------+
|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: 11/11/1999|
| 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). |
| |
| * * * UPDATE AT 1605 ON 11/11/99 MATZKE TO GOULD * * * UNRETRACT A |
| RETRACTION |
| |
| As a result of discussions with NRC personnel on 11/10/99, regarding |
| reporting requirements, OPPD is withdrawing the retraction notification of |
| 11/09/99, and changing the basis of the original notification to |
| "voluntary". This is because the reporting requirements of 10CFR50.72 were |
| determined to be not applicable, and thus it is not clear that reporting |
| pursuant to the technical specifications is required. However, due to NRC |
| desire for formal, prompt notification of this type of event and the absence |
| of explicit regulatory guidance, reporting on a voluntary basis appears |
| appropriate. |
| |
| The 30 day steam generator tube plugging report to the NRC pursuant to |
| Technical Specification 3.17(5)(i) has been sent. The six month steam |
| generator inspection report pursuant to Technical Specification 3.17(5)(ii) |
| will be sent as required and will include the information requested in the |
| second sentence of Technical Specification 3.17(5)(iii). |
| |
| The licensess notified the NRC Resident Inspector. The NRC Operations |
| Officer notified R4DO (Linda Smith). |
+------------------------------------------------------------------------------+
!!!!!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!!!
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36336 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: NINE MILE POINT REGION: 1 |NOTIFICATION DATE: 10/22/1999|
| UNIT: [1] [] [] STATE: NY |NOTIFICATION TIME: 20:55[EDT]|
| RXTYPE: [1] GE-2,[2] GE-5 |EVENT DATE: 10/22/1999|
+------------------------------------------------+EVENT TIME: 19:55[EDT]|
| NRC NOTIFIED BY: DON SHEEHAN |LAST UPDATE DATE: 11/10/1999|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |WILLIAM COOK R1 |
|10 CFR SECTION: | |
|AOUT 50.72(b)(1)(ii)(B) OUTSIDE DESIGN BASIS | |
|NLCO TECH SPEC LCO A/S | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 98 Power Operation |98 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| - UNIT 1 IS OPERATING OUTSIDE ITS DESIGN BASIS IN ITS CURRENT OPERATING |
| CONDITION - |
| |
| Unit 1 received information from its Engineering Department that Unit 1 is |
| operating outside of design basis in its current operating condition. |
| Specifically, with #11 Reactor Recirc Pump fully isolated, an analysis for |
| thermal shock caused by initiation of #12 Emergency Cooling Loop injecting |
| through #11 Reactor Recirc Loop suction nozzle has not been performed. |
| Mitigating actions include isolating #12 Emergency Cooling Loop, in order to |
| return the plant to an analyzed condition, which requires that the plant |
| enter a 7-Day Technical Specification Shutdown LCO until such time that an |
| analysis for thermal shock has been performed. Engineering Department has |
| reasonable assurance that this analysis will be completed within the 7 days |
| required by Unit 1 Tech Specs. |
| |
| All other Emergency Core Cooling System equipment is operable. |
| |
| This event has no effect on Unit 2. |
| |
| The licensee notified the NRC Resident Inspector. |
| |
| * * * UPDATE AT 1349 ON 11/10/99 BY KIRCHNER TAKEN BY WEAVER * * * |
| |
| A licensee engineering supporting analysis has been performed which |
| demonstrated that the plant has been operating within the design basis. |
| Therefore, this event has been retracted. The licensee notified the NRC |
| resident inspector. The NRC Operations Officer notified R1DO (Dan Holody). |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|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: 11/11/1999|
| 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 of tubes of Steam Generator (SG) #RC-2B, greater |
| than 1% of the tubes tested were found to be defective. The number of |
| tested 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 AT 0923 ON 11/09/99 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 licensee notified the NRC Resident Inspector. The NRC Operations Officer |
| notified R4DO (Linda Smith). |
| |
| * * * UPDATE AT 1605 ON 11/11/99 BY MATZKE TO GOULD * * * UNRETRACT A |
| RETRACTION |
| |
| As a result of discussions with NRC personnel on 11/10/99 regarding |
| reporting requirements, OPPD is withdrawing the retraction notification of |
| 11/09/99, and changing the basis of the original notification to |
| "voluntary". This is because the reporting requirements of 10CFR50.72 were |
| determined to be not applicable, and thus it is not clear that reporting |
| pursuant to the technical specifications is required. However, due to NRC |
| desire for formal, prompt notification of this type of event and the absence |
| of explicit regulatory guidance, reporting on a voluntary basis appears |
| appropriate. |
| |
| The 30 day steam generator tube plugging report to the NRC pursuant to |
| Technical Specification 3.17(5)(i) has been sent. The six month steam |
| generator inspection report pursuant to Technical Specification 3.17(5)(ii) |
| will be sent as required and will include the information requested in the |
| second sentence of Technical Specification 3.1 7(5)(iii). |
| |
| The licensee notified the NRC Resident Inspector. The NRC Operations |
| Officer notified R4DO (Linda Smith). |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Fuel Cycle Facility |Event Number: 36418 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PORTSMOUTH GASEOUS DIFFUSION PLANT |NOTIFICATION DATE: 11/10/1999|
| RXTYPE: URANIUM ENRICHMENT FACILITY |NOTIFICATION TIME: 14:33[EST]|
| COMMENTS: 2 DEMOCRACY CENTER |EVENT DATE: 11/10/1999|
| 6903 ROCKLEDGE DRIVE |EVENT TIME: [EST]|
| BETHESDA, MD 20817 (301)564-3200 |LAST UPDATE DATE: 11/10/1999|
| CITY: PIKETON REGION: 3 +-----------------------------+
| COUNTY: PIKE STATE: OH |PERSON ORGANIZATION |
|LICENSE#: GDP-2 AGREEMENT: N |MICHAEL PARKER R3 |
| DOCKET: 0707002 |JOHN HICKEY NMSS |
+------------------------------------------------+JOSEPH GIITTER IRO |
| NRC NOTIFIED BY: ERIC SPAETH | |
| HQ OPS OFFICER: DOUG WEAVER | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|NBNL RESPONSE-BULLETIN | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| 4 HOUR BULLETIN 91-01 REPORT |
| |
| On November 11, 1999 at 1020, the Nuclear Criticality Safety (NCS) staff |
| determined that a Nuclear Criticality Safely Approval (NCSA) for X-705 was |
| deficient. The compressor turnover pit located in the equipment disassembly |
| north tear down area, has a sump pump which activates if solution should |
| accumulate above 2.3 inches. NCSA-0705_031.AOO analyzed the turnover pit, |
| assuming a maximum enrichment of 3%. The Nuclear Criticality Safety |
| Evaluation (NCSE) for this operation failed to consider the contingency that |
| the Geometrically Safe Storage (GSS) located overhead, which was analyzed at |
| 100% enrichment, could leak into the compressor turnover pit. |
| |
| The licensee notified the NRC resident inspector. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|General Information or Other |Event Number: 36419 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| REP ORG: STATE OF LOUISIANA |NOTIFICATION DATE: 11/10/1999|
|LICENSEE: MOORE & ASSOCIATES, INC. |NOTIFICATION TIME: 16:00[EST]|
| CITY: REGION: 4 |EVENT DATE: 11/10/1999|
| COUNTY: E. BATON ROUGE PAR. STATE: LA |EVENT TIME: 07:00[CST]|
|LICENSE#: AGREEMENT: Y |LAST UPDATE DATE: 11/10/1999|
| DOCKET: |+----------------------------+
| |PERSON ORGANIZATION |
| |LINDA SMITH R4 |
| | |
+------------------------------------------------+ |
| NRC NOTIFIED BY: TORY MEAUX | |
| HQ OPS OFFICER: DOUG WEAVER | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|NAGR AGREEMENT STATE | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| AGREEMENT STATE REPORT - STOLEN MOISTURE/DENSITY GAUGE |
| |
| The state of Louisiana reported that a Campbell moisture/density gauge, |
| model MC3, serial number 6368, was stolen on 11/10/99. The gauge was |
| reported missing at 0700 CST on 11/10/99. The gauge contains 10 �Ci of |
| Cs-137 and a 50 �Ci AmBe source. The gauge was stolen from the back of a |
| truck at the offices of Moore & Associates. The lock on the gauge had been |
| cut. The theft was reported to the local and state police departments. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36420 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: HATCH REGION: 2 |NOTIFICATION DATE: 11/10/1999|
| UNIT: [1] [2] [] STATE: GA |NOTIFICATION TIME: 17:39[EST]|
| RXTYPE: [1] GE-4,[2] GE-4 |EVENT DATE: 11/10/1999|
+------------------------------------------------+EVENT TIME: 15:49[EST]|
| NRC NOTIFIED BY: BARRY COLEMAN |LAST UPDATE DATE: 11/10/1999|
| HQ OPS OFFICER: DOUG WEAVER +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |CHARLES OGLE R2 |
|10 CFR SECTION: | |
|HFIT 26.73 FITNESS FOR DUTY | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 100 Power Operation |100 Power Operation |
|2 N Y 100 Power Operation |100 Power Operation |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| FITNESS FOR DUTY REPORT |
| |
| A supervisory contract employee was found to have used two illegal drugs |
| during a random drug screening. |
| The NRC resident inspector has been notified. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36421 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: BRUNSWICK REGION: 2 |NOTIFICATION DATE: 11/10/1999|
| UNIT: [1] [2] [] STATE: NC |NOTIFICATION TIME: 19:09[EST]|
| RXTYPE: [1] GE-4,[2] GE-4 |EVENT DATE: 11/10/1999|
+------------------------------------------------+EVENT TIME: 18:15[EST]|
| NRC NOTIFIED BY: DAN HARDIN |LAST UPDATE DATE: 11/10/1999|
| HQ OPS OFFICER: DOUG WEAVER +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |CHARLES OGLE R2 |
|10 CFR SECTION: | |
|DDDD 73.71 UNSPECIFIED PARAGRAPH | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 100 Power Operation |100 Power Operation |
|2 N Y 100 Power Operation |100 Power Operation |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| SECURITY REPORT |
| |
| UNESCORTED ACCESS GRANTED INAPPROPRIATELY. IMMEDIATE CORRECTIVE ACTIONS |
| TAKEN UPON DISCOVERY. THE LICENSEE NOTIFIED THE NRC RESIDENT INSPECTOR. |
| SEE THE HOO LOG FOR ADDITIONAL DETAILS. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36422 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: DUANE ARNOLD REGION: 3 |NOTIFICATION DATE: 11/10/1999|
| UNIT: [1] [] [] STATE: IA |NOTIFICATION TIME: 21:27[EST]|
| RXTYPE: [1] GE-4 |EVENT DATE: 11/10/1999|
+------------------------------------------------+EVENT TIME: 19:10[CST]|
| NRC NOTIFIED BY: DICK FOWLER |LAST UPDATE DATE: 11/10/1999|
| 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
+------------------------------------------------------------------------------+
| INTERGRANULAR STRESS CORROSION CRACKING IDENTIFIED ON RECIRCULATION RISER |
| WELD |
| |
| While performing ultrasonic examination of recirculation riser weld |
| #RRF-F002, nozzle to safe end weld, indications of Intergranular Stress |
| Corrosion Cracking (IGSCC) were identified. Specifically, a 0.3 inch X 13 |
| inch long crack on the nozzle to safe end weld. This nozzle was being |
| inspected as part of an expanded inspection scope as a result of similar |
| indications found on the 'B' recirc riser to safe end weld. To date, |
| indications have been found on the 'B', 'D', and 'F' nozzles with no |
| indications on the 'A' and 'C' nozzles. The 'E', 'G' and 'H' inspections |
| have not been completed. |
| |
| The licensee notified the NRC resident inspector. |
| |
| See events #36402 and #36416. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Fuel Cycle Facility |Event Number: 36423 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PORTSMOUTH GASEOUS DIFFUSION PLANT |NOTIFICATION DATE: 11/10/1999|
| RXTYPE: URANIUM ENRICHMENT FACILITY |NOTIFICATION TIME: 22:02[EST]|
| COMMENTS: 2 DEMOCRACY CENTER |EVENT DATE: 11/09/1999|
| 6903 ROCKLEDGE DRIVE |EVENT TIME: 23:15[EST]|
| BETHESDA, MD 20817 (301)564-3200 |LAST UPDATE DATE: 11/10/1999|
| CITY: PIKETON REGION: 3 +-----------------------------+
| COUNTY: PIKE STATE: OH |PERSON ORGANIZATION |
|LICENSE#: GDP-2 AGREEMENT: N |MICHAEL PARKER R3 |
| DOCKET: 0707002 |ROBERT PIERSON NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: JEFF CASTLE | |
| HQ OPS OFFICER: DOUG WEAVER | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|NBNL RESPONSE-BULLETIN | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| 24 HOUR BULLETIN 91-01 REPORT |
| |
| At 2315 on 11/09/99, a maintenance mechanic in the X-330 building identified |
| four pieces of equipment which violated NCSA-PLANT062.A02 requirement #4 |
| which states "Openings / penetrations made during maintenance activities |
| shall be covered to minimize the potential for moderator collection and |
| moist air exposure when unattended". The violations were corrected by |
| repairing or replacing the damaged covers by 0218 on 11/10/99. One |
| additional violation of the NCSA requirement was identified during followup |
| facility walkdowns. The additional piece of equipment was discovered in |
| the X-330 building at 1030 on 11/10/99. This deficiency was corrected at |
| 1259 on 11/10/99. This constitutes a loss of control such that only one |
| double contingency control remains in place. |
| |
| The licensee notified the NRC resident inspector. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36424 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: BYRON REGION: 3 |NOTIFICATION DATE: 11/11/1999|
| UNIT: [] [2] [] STATE: IL |NOTIFICATION TIME: 03:06[EST]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 11/10/1999|
+------------------------------------------------+EVENT TIME: 22:20[CST]|
| NRC NOTIFIED BY: MARRI MARCHIONDA |LAST UPDATE DATE: 11/11/1999|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |MICHAEL PARKER R3 |
|10 CFR SECTION: | |
|ARPS 50.72(b)(2)(ii) RPS ACTUATION | |
|AESF 50.72(b)(2)(ii) ESF ACTUATION | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
| | |
|2 N N 0 Cold Shutdown |0 Cold Shutdown |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| - S/G LEVEL DROPPED TO REACTOR TRIP/AFW AUTO START SETPOINT DURING |
| MAINTENANCE - |
| |
| At 2220 CST on 11/10/99 with Unit 2 in Mode 5 (Cold Shutdown), priming of |
| the feedwater isolation valves to support maintenance activities was in |
| progress. While priming the 'B' steam generator (S/G) feedwater isolation |
| valve #2FW009B, the valve was opened and water back-flowed from the 'B' S/G |
| into the feedwater piping and the S/G level decreased to the low level 2 |
| reactor trip setpoint of 36.3% which is also the setpoint for the autostart |
| of the auxiliary feedwater (AFW) system. The initial level in the 'B' S/G |
| was 50%. A reactor protection system actuation and engineered safety |
| features actuation occurred and all plant systems responded normally for |
| Mode 5. The reactor trip breakers opened but the control rod drive system |
| was out of service and not capable of control rod withdrawal. The AFW |
| auxiliary lube oil pumps auto started but the AFW pumps were in pull-to-lock |
| due to current plant conditions so the AFW pumps did not start. The cause |
| of the level decrease is due to unfilled feedwater piping following outage |
| restoration. Corrective action is being determined. This event is |
| reportable under the requirements of 10CFR50.72(b)(2)(ii). |
| |
| The licensee notified the NRC Resident Inspector. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36425 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: FERMI REGION: 3 |NOTIFICATION DATE: 11/11/1999|
| UNIT: [2] [] [] STATE: MI |NOTIFICATION TIME: 09:59[EST]|
| RXTYPE: [2] GE-4 |EVENT DATE: 11/11/1999|
+------------------------------------------------+EVENT TIME: 07:15[EST]|
| NRC NOTIFIED BY: MIKE PHILIPPON |LAST UPDATE DATE: 11/11/1999|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |MICHAEL PARKER R3 |
|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 |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|2 N Y 97 Power Operation |97 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| - REACTOR WATER CLEANUP SYSTEM ISOLATION DUE TO HIGH DIFFERENTIAL |
| TEMPERATURE - |
| |
| A Reactor Water Cleanup (RWCU) System steam leak detection high differential |
| temperature alarm was received and the subsequent isolation of the outboard |
| isolation valves #G33F004 and #G33F220 occurred. The reactor building was |
| immediately evacuated and access was restricted as a precaution until the |
| cause of the high temperature alarm and RWCU System isolation could be |
| determined. Although it was confirmed that there was no steam leak, the |
| RWCU System isolation signal was considered valid because an actual high |
| differential temperature condition existed. The valid isolation signal was |
| a result of a trip of the Reactor Building HVAC System and the failure of |
| the RWCU pump 'B' room cooler to automatically start on high room |
| temperature. Investigation revealed that the thermal overloads on the RWCU |
| room cooler were tripped. The Reactor Building HVAC System had tripped on |
| low freezstat (temperature switch) temperature. All other systems operated |
| as expected, and the RWCU System remains shut down pending investigation of |
| the failure of the 'B' room cooler to start. |
| |
| The NRC Resident Inspector was notified by the licensee. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36426 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PALISADES REGION: 3 |NOTIFICATION DATE: 11/11/1999|
| UNIT: [1] [] [] STATE: MI |NOTIFICATION TIME: 11:18[EST]|
| RXTYPE: [1] CE |EVENT DATE: 11/10/1999|
+------------------------------------------------+EVENT TIME: 14:00[EST]|
| NRC NOTIFIED BY: STEVE ELLIS |LAST UPDATE DATE: 11/11/1999|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |MICHAEL PARKER R3 |
|10 CFR SECTION: | |
|HFIT 26.73 FITNESS FOR DUTY | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|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
+------------------------------------------------------------------------------+
| - TWO EMPTY BEER CANS FOUND IN A TRAILER INSIDE THE PLANT PROTECTED AREA - |
| |
| Plant janitors found two empty beer cans in a shower trailer in a |
| maintenance workers area inside the plant protected area. The licensee is |
| interviewing the janitors and searched the trailer for additional cans; none |
| were found. The licensee sent the cans to the local police for possible |
| identification of fingerprints; results are pending. The licensee searched |
| the hand held items of people exiting the plant protected area; nothing |
| unusual was found. The licensee notified the NRC Resident Inspector and is |
| continuing to investigate this incident. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36427 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: ARKANSAS NUCLEAR REGION: 4 |NOTIFICATION DATE: 11/11/1999|
| UNIT: [] [2] [] STATE: AR |NOTIFICATION TIME: 17:53[EST]|
| RXTYPE: [1] B&W-L-LP,[2] CE |EVENT DATE: 11/11/1999|
+------------------------------------------------+EVENT TIME: 16:45[CST]|
| NRC NOTIFIED BY: SCOTT |LAST UPDATE DATE: 11/11/1999|
| HQ OPS OFFICER: CHAUNCEY GOULD +-----------------------------+
+------------------------------------------------+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 |
+-----+----------+-------+--------+-----------------+--------+-----------------+
| | |
|2 N N 0 Cold Shutdown |0 Cold Shutdown |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| FAILURE OF INSIDE AND OUTSIDE CCW CONTAINMENT ISOLATION VALVES. |
| |
| ON 11/09/99, WHILE IN MODE 5, COLD SHUTDOWN, FOR MID-CYCLE OUTAGE, THE |
| OUTSIDE CONTAINMENT ISOLATION VALVE FOR COMPONENT COOLING WATER (CCW) RETURN |
| (2CV-5255-1) FAILED TO FULLY CLOSE BASED ON FLOW NOISE DURING THE INITIAL |
| STROKE. THE VALVE WAS STROKED 10 TIMES AND APPEARED TO CLOSE ON ALL STROKES |
| FOLLOWING THE INITIAL ATTEMPT. ON 11/10/99, THE INSIDE CONTAINMENT |
| ISOLATION VALVE FOR CCW RETURN (2CV-5254-2) FAILED TO REACH THE TRAVEL LIMIT |
| ON ITS FIRST STROKE. AN IMPROMPTU LEAK TEST USING WATER WITH 2CV-5254-2 IN |
| ITS INITIAL POSITION INDICATED 300-500 ML/MIN LEAKAGE VIA THE VALVE SEAT. |
| BOTH VALVES WERE OVERHAULED DURING THE REFUELING OUTAGE IN EARLY 1999 AND |
| SUCCESSFULLY PASSED POST-MAINTENANCE TESTING. DETERMINATION OF THE CAUSE(S) |
| AND RESTORATION OF BOTH VALVES TO AN OPERABLE STATUS WILL BE COMPLETED |
| BEFORE PLANT STARTUP. THERE DOES NOT APPEAR TO BE FIRM EVIDENCE CONCERNING |
| A TIME OF FAILURE BEFORE THE CONDITIONS WERE DISCOVERED. ON NOVEMBER 11, |
| 1999, DURING A REVIEW OF THESE CONDITIONS AFFECTING THE SAME CONTAINMENT |
| PENETRATION, IT WAS DETERMINED THAT THEY MET THE REPORTING CRITERION OF |
| 10CFR5O.72(b)(2)(I) AS A SERIOUS DEGRADATION OF A PRINCIPAL SAFETY BARRIER |
| (REACTOR CONTAINMENT) FOUND WHILE SHUTDOWN. |
| |
| THE RESIDENT INSPECTOR WAS NOTIFIED. |
+------------------------------------------------------------------------------+
Page Last Reviewed/Updated Thursday, March 25, 2021