Event Notification Report for March 13, 2000
U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
03/10/2000 - 03/13/2000
** EVENT NUMBERS **
36721 36782 36783 36784 36785 36786 36787 36788 36789
!!!!!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!!!
+------------------------------------------------------------------------------+
|Fuel Cycle Facility |Event Number: 36721 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PORTSMOUTH GASEOUS DIFFUSION PLANT |NOTIFICATION DATE: 02/23/2000|
| RXTYPE: URANIUM ENRICHMENT FACILITY |NOTIFICATION TIME: 19:11[EST]|
| COMMENTS: 2 DEMOCRACY CENTER |EVENT DATE: 02/23/2000|
| 6903 ROCKLEDGE DRIVE |EVENT TIME: 08:35[EST]|
| BETHESDA, MD 20817 (301)564-3200 |LAST UPDATE DATE: 03/10/2000|
| CITY: PIKETON REGION: 3 +-----------------------------+
| COUNTY: PIKE STATE: OH |PERSON ORGANIZATION |
|LICENSE#: GDP-2 AGREEMENT: N |JAMES CREED R3 |
| DOCKET: 0707002 |WILLIAM BRACH NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: RICK LARSON | |
| HQ OPS OFFICER: DICK JOLLIFFE | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|NCFR NON CFR REPORT REQMNT | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| - Emergency response procedures implemented due to a smoke head alarm and |
| high pressure vent actuation - |
| |
| At approximately 0835 on 02/23/00, Portsmouth operations personnel received |
| a smoke head alarm for extended range product (ERP) compressor W-3 in Area |
| Control Room (ACR)-4. At the same time, another operator in ACR-4 looked at |
| the TV monitor for the ERP compressors and saw a "puff" of smoke around |
| compressor W-3. The plant's "SEE & FLEE" emergency procedures were |
| immediately implemented. The ERP stations' high pressure vent actuated |
| taking the station below atmospheric pressure. Results of all surveys |
| performed during the emergency response were less than minimum detectable |
| activity. At 0949, the emergency response was cancelled. The ERP station |
| was made inoperable to allow for testing of the system by operations and |
| engineering personnel. Portsmouth is reporting this alarm as a valid |
| actuation of a "Q" Safety System. |
| |
| This event is reportable to the NRC as a valid actuation of a "Q" Safety |
| System in accordance with Safety Analysis Report, Section 6.9 (24-HOUR |
| REPORT). |
| |
| There was no loss of hazardous or radioactive material nor radioactive or |
| radiological contamination exposure as result of this event. |
| |
| The NRC Resident Inspector and DOE site representative were notified of this |
| event. |
| |
| PTS-2000-020 PR-PTS-00-01079 |
| |
| * * * RETRACTION 1501 3/10/2000 FROM HALCOMB TAKEN BY STRANSKY * * * |
| |
| "After further investigation, it has been determined that a 'Q' safety |
| system actuation did not occur. The Pyrotronic smoke detectors and the ERP |
| station high pressure vent are not classified as 'Q' safety systems. The |
| CADP smokeheads, which are classified as 'Q' safety systems, did not |
| actuate. Since there was no safety system actuation, this event has been |
| determined to be not reportable to the NRC and is being retracted." |
| |
| The NRC resident inspector has been informed of this retraction. Notified |
| R3DO (Lanksbury). |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Fuel Cycle Facility |Event Number: 36782 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PORTSMOUTH GASEOUS DIFFUSION PLANT |NOTIFICATION DATE: 03/09/2000|
| RXTYPE: URANIUM ENRICHMENT FACILITY |NOTIFICATION TIME: 17:04[EST]|
| COMMENTS: 2 DEMOCRACY CENTER |EVENT DATE: 03/09/2000|
| 6903 ROCKLEDGE DRIVE |EVENT TIME: 03:32[EST]|
| BETHESDA, MD 20817 (301)564-3200 |LAST UPDATE DATE: 03/11/2000|
| CITY: PIKETON REGION: 3 +-----------------------------+
| COUNTY: PIKE STATE: OH |PERSON ORGANIZATION |
|LICENSE#: GDP-2 AGREEMENT: N |ROGER LANKSBURY R3 |
| DOCKET: 0707002 |THOMAS ESSIG NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: JEFF CASTLE | |
| HQ OPS OFFICER: BOB STRANSKY | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|NBNL RESPONSE-BULLETIN | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| 24-HOUR NRC BULLETIN 91-01 REPORT |
| |
| "At 0332 hours on 03/09/2000, it was identified that an error had been made |
| in connecting a cell calibration test buggy, covered by NCSA-PLANT069, at |
| X-330 building cell 31-3-9. The cell was connected to the HI DAT (HI DATUM) |
| port instead of the PROCESS inlet port. The NCSA identifies installed |
| chemical traps on the test buggy as a passive design characteristic relied |
| upon to prevent an accumulation of uranium in the vacuum pump oil. The |
| installed traps are also a control contingency in preventing back flow of |
| vacuum pump oil to the cell manifold. Connection of the process system to |
| the HI DAT port bypassed these chemical traps and provided a direct flow |
| path between the process gas system and vacuum pump. |
| |
| "Establishing a connection between the vacuum pump and cell process gas |
| manifold constitutes a loss of one control of the double contingency control |
| principle. The cell calibration buggy was disconnected and it was determined |
| that no oil from the vacuum pump had migrated through the buggy to the cell |
| manifold. The potential for a criticality to occur is precluded based on the |
| amount of oil contained in the vacuum pump and by the assay of the material |
| at that point of the cascade. Pull compliance with NCSA-PLANT069 was |
| regained when the calibration buggy was disconnected from the cell |
| manifold. |
| |
| "There was no loss of hazardous/radioactive material or |
| radioactive/radiological contamination exposure as a result of this event. |
| |
| "SAFETY SIGNIFICANCE OF EVENTS: |
| |
| "This event has a low safety significance. Due to operator error, the cell |
| was connected to the wrong inlet port This allowed the possibility that |
| process gas could bypass the chemical trap(s) and then collect in the oil |
| reservoir of the vacuum pump. The oil reservoir (limited to <= one quart) is |
| sized such that it is safe for 100% enriched material. The process gas that |
| may have entered the oil reservoir of the pump is approximately 2.25% |
| enriched. There is insufficient oil In the vacuum pump for a criticality to |
| occur. Additionally the test buggy has been disconnected from the cell. |
| Thus, there is no possibility of adding additional uranium to the oil |
| reservoir of the vacuum pump. |
| |
| "POTENTIAL CRITICALITY PATHWAYS INVOLVED (BRIEF SCENARIO(S) OF HOW |
| CRITICALITY COULD OCCUR): |
| |
| "For a criticality to occur, sufficient UF6 would need to collect in the oil |
| reservoir of the vacuum pump and the vacuum pump would need to be replaced |
| with a different model such that the oil reservoir is large enough (greater |
| than 3 quarts) for a criticality to occur. |
| |
| "CONTROLLED PARAMETERS (MASS, MODERATION, GEOMETRY, CONCENTRATION, ETC.): |
| |
| "This NCSA relied on preventing the accumulation of process gas from |
| collecting in the vacuum pump oil by placing a chemical trap upstream of the |
| vacuum pump. This control was lost. The second control was to limit the |
| amount of oil in the vacuum pump to less than 1 quart, which is less than |
| the minimum volume of oil required for a criticality at 100% U235. |
| |
| "ESTIMATED AMOUNT, ENRICHMENT, FORM OF LICENSED MATERIAL (INCLUDE PROCESS |
| LIMIT AND % WORST CASE OF CRITICAL MASS): |
| |
| "At this time it is not known if any UF6 reached the oil reservoir for the |
| vacuum pump. Enrichment in cell 31-3-9 is estimated to be 2.25% U-235. If |
| process gas reached the oil reservoir, it would be in the form of UF4/oil |
| mixture. |
| |
| "NUCLEAR CRITICALITY SAFETY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION |
| OF THE FAILURES OR DEFICIENCIES |
| |
| "Nuclear criticality safety is maintained by two controls, the first Is the |
| placement of chemical traps upstream of the vacuum pump. This control was |
| lost. The second control was on the allowed volume of oil in the reservoir |
| of the vacuum pump. This control was maintained throughout the event. Thus, |
| one control relied on for double contingency was lost." |
| |
| The NRC resident inspector has been informed of this notification. |
| |
| * * * UPDATE ON 03/11/00 AT 0128 ET FROM ERIC SPAETH TAKEN BY MACKINNON * * |
| * |
| |
| It was discovered that the internal volume of the oil reservoir in the |
| vacuum pump was greater than the NCSA control limit of less than or equal to |
| one quart. The actual volume of the reservoir is approximately one and |
| one third quart (The manufacturer recommends that the vacuum pump oil level |
| should be 1.33 quarts). At the time of the initial notification, it was not |
| known that the oil reservoir was greater than the NCSA controlled limit. |
| However, this volume is still within the three quart volume analyzed in the |
| NCSE for the normal case condition. Therefore, the safety significance of |
| this event remains low. The Certificate holder said that the NCSA limit for |
| oil will be increased. |
| |
| CONTROLLED PARAMETERS (MASS, MODERATION, GEOMETRY, CONCENTRATION, ETC.): |
| |
| The second control parameter was also lost which was to limit the oil level |
| in the vacuum pump to less than one quart. |
| |
| NUCLEAR CRITICALITY SAFETY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION |
| OF THE FAILURES OR DEFICIENCIES: |
| |
| The second control was lost (oil level greater than one quart) when it was |
| discovered that the volume of the oil reservoir exceeded the allowed volume |
| in the NCSA. Even though the volume of the reservoir exceeded the allowed |
| amount it was still within the three quart limit analyzed in the NCSE for |
| the normal case condition. |
| R3DO (Roger Lanksbury) and NMSS (Tom Essig) notified. |
| |
| |
| |
| The NRC Resident Inspector was notified of this event update by the |
| certificate holder. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36783 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: NORTH ANNA REGION: 2 |NOTIFICATION DATE: 03/10/2000|
| UNIT: [1] [2] [] STATE: VA |NOTIFICATION TIME: 13:01[EST]|
| RXTYPE: [1] W-3-LP,[2] W-3-LP |EVENT DATE: 03/09/2000|
+------------------------------------------------+EVENT TIME: 14:30[EST]|
| NRC NOTIFIED BY: RICHARD WESLEY |LAST UPDATE DATE: 03/10/2000|
| HQ OPS OFFICER: BOB STRANSKY +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |EDWARD MCALPINE R2 |
|10 CFR SECTION: | |
|APRE 50.72(b)(2)(vi) OFFSITE NOTIFICATION | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 100 Power Operation |100 Power Operation |
|2 N Y 100 Power Operation |100 Power Operation |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| NOTIFICATION TO FEDERAL ENERGY REGULATORY COMMISSION |
| |
| The licensee notified the Federal Energy Regulatory Commission (FERC) |
| regarding the inoperability of the spillway diesel generator. The diesel |
| generator was inoperable while batteries were being replaced. The diesel |
| generator is now fully operable. |
| |
| The NRC resident inspector has been informed of this notification by the |
| licensee. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36784 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: CATAWBA REGION: 2 |NOTIFICATION DATE: 03/10/2000|
| UNIT: [1] [2] [] STATE: SC |NOTIFICATION TIME: 17:58[EST]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 03/09/2000|
+------------------------------------------------+EVENT TIME: 22:57[EST]|
| NRC NOTIFIED BY: DON BRADLEY |LAST UPDATE DATE: 03/10/2000|
| HQ OPS OFFICER: BOB STRANSKY +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |EDWARD MCALPINE R2 |
|10 CFR SECTION: | |
|NINF INFORMATION ONLY | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|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 95 Power Operation |95 Power Operation |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| COURTESY NOTIFICATION REGARDING ONSITE FATALITY |
| |
| At 2357 on 3/9/2000, a Duke temporary employee was pronounced dead at the |
| Piedmont Medical Center. The individual had collapsed at the site, and |
| attempts to resuscitate the individual were unsuccessful. The licensee made |
| a courtesy notification to the Occupational Safety and Health Administration |
| (OSHA) regarding this occurrence. |
| |
| The NRC resident inspector has been informed of this notification by the |
| licensee. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36785 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: KEWAUNEE REGION: 3 |NOTIFICATION DATE: 03/10/2000|
| UNIT: [1] [] [] STATE: WI |NOTIFICATION TIME: 18:13[EST]|
| RXTYPE: [1] W-2-LP |EVENT DATE: 03/10/2000|
+------------------------------------------------+EVENT TIME: 16:20[CST]|
| NRC NOTIFIED BY: TOM WEBB |LAST UPDATE DATE: 03/10/2000|
| HQ OPS OFFICER: BOB STRANSKY +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |ROGER LANKSBURY R3 |
|10 CFR SECTION: | |
|ADEG 50.72(b)(1)(ii) DEGRAD COND DURING OP | |
|AUNA 50.72(b)(1)(ii)(A) UNANALYZED COND OP | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 100 Power Operation |100 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| SHIELD BUILDING VENTILATION FILTER TRAIN FLOW OUTSIDE OF SPECIFICATION |
| |
| While performing an administrative review of the results of surveillance |
| test SP 24-122, "Shield Building Vent Filter Testing," performed on |
| 7/27/1999, the licensee determined that train "B" was outside of |
| specification for fan flow rate. The required flow rate is 6200 SCFM +/- |
| 10%, while measured flow was 5427 SCFM. The Updated Final Safety Analysis |
| Report (UFSAR) assumes a flow rate of 5000 SCFM. Shield building vent train |
| "B" was declared inoperable at 1620 on 3/10/2000, and the unit entered |
| 7-day Technical Specification Limiting Condition for Operation (LCO) |
| 3.6.b.1. The NRC resident inspector has been informed of this event by the |
| licensee. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36786 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: CLINTON REGION: 3 |NOTIFICATION DATE: 03/10/2000|
| UNIT: [1] [] [] STATE: IL |NOTIFICATION TIME: 20:05[EST]|
| RXTYPE: [1] GE-6 |EVENT DATE: 03/10/2000|
+------------------------------------------------+EVENT TIME: 18:45[CST]|
| NRC NOTIFIED BY: MARSHALL FUNKHOUSER |LAST UPDATE DATE: 03/10/2000|
| HQ OPS OFFICER: BOB STRANSKY +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |ROGER LANKSBURY 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 |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 100 Power Operation |100 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| POTENTIAL TO DAMAGE EDG WHEN PARALLELED TO OFFSITE POWER SOURCE |
| |
| "It has been determined that there exists the potential for an unanalyzed |
| interaction between the Static VAR Compensators and a diesel generator when |
| paralleled with an offsite power source. This unintended interaction has |
| been determined to constitute an event or condition during operation that |
| results in the nuclear power plant being in a condition that is outside the |
| design basis of the plant. This condition has been documented in a Condition |
| Report and entered into the plant corrective action program. |
| |
| "Discussion: |
| |
| "Static VAR compensators (SVCs) are installed on the secondary side of the |
| reserve auxiliary transformer (RAT) and the emergency reserve auxiliary |
| transformer (ERAT) to address potentially inadequate grid voltage conditions |
| assuming a loss of coolant accident (LOCA) and unit trip. One SVC is |
| installed on the secondary (4.16 kV) side of the RAT and one SVC is |
| installed on the secondary side (4.16 kV) side of the ERAT. The RAT is |
| associated with the offsite 345 kV transmission system and the BRAT is |
| associated with the offsite 138 kV transmission system. These two |
| transmission systems constitute the two required offsite electrical power |
| sources for the plant's three Class 1E 4.16 kV buses, Each Class 1B 4.16 kV |
| buses is also capable of being supplied by a dedicated diesel generator. |
| |
| "The design of the SVC is described in a Design Report that was included as |
| Attachment 5 to the CPS license amendment submittal dated May 4, 1998 |
| (letter U602972). During parallel operations between the diesel generator |
| and the offsite power source via either the RAT or the ERAT, the SVCs are |
| designed to 'freeze' such that their design function of regulating voltage |
| on the secondary side of the RAT and the BRAT does not result in detrimental |
| interaction between its control circuits and the control circuits of the |
| diesel generator. The action of freezing the SVC control circuit ensures |
| that the SVCs do not attempt to counter bus voltage transients inherent in |
| the process of synchronizing the diesel generator to its respective Class |
| 113 bus. The freeze signal is derived from auxiliary contacts on die diesel |
| generator output breaker and the Class 1E bus feeder breakers. |
| |
| "During the root cause investigation into the damage incurred by the |
| Division III diesel generator during a routine surveillance (apparently |
| involving a synchronization attempt with an out-of-synchronization |
| condition), it was determined that a small time delay inherent in the |
| electrical interlock circuitry for the freeze signal allowed the SVC to |
| detrimentally interact with the diesel generator since the SVC freeze |
| condition is not immediately effected upon closure of the diesel generator |
| output breaker. The damage to the Division III diesel generator was |
| previously reported in event number 36736. Although the potential for |
| interaction had been recognized and compensated for in the design, the |
| effect of the longer than expected time delay had not been recognized. |
| Therefore, the effects of the unanalyzed time delay introduced an increased |
| probability of malfunction of equipment important to safety resulting in a |
| design configuration that constitutes an unreviewed safety question as |
| defined in 10 CFR 50.59. |
| |
| "This configuration has been determined to constitute a condition that is |
| outside the design basis of the plant. The design requirement for the SVCs |
| is to assist in maintaining secondary side voltages on the RAT and ERAT |
| within acceptable values consistent with the offsite source 'capacity and |
| capability' requirements of General Design Criterion (GDC) 17. A second |
| requirement was that the SVCs not negatively impact plant structures, |
| systems, and components such that an increase in the probability of a |
| malfunction of equipment important to safety exists. Contrary to the above |
| requirements, the current design of thc SVC controls is such that during |
| synchronization of the diesel generator to its Class 1E bus the potential |
| exists for the SVC to negatively impact bus voltage. As a consequence of the |
| adverse voltage control during the diesel generator synchronization, the |
| potential exists for the SVC response to degrade or impair the function of |
| the diesel generator resulting in an increased probability of malfunction of |
| equipment important to safety. |
| |
| "This design deficiency exists only under the limited circumstances during |
| the conduct of diesel generator surveillances in which the diesel is |
| paralleled with the offsite power source. The postulated duration of the |
| unanalyzed condition is a fraction of a second following the closure of the |
| diesel generator output breaker. This is the approximate length of time |
| required for the SVC control circuits to enter the freeze mode after closure |
| of the diesel generator output breaker. Prior to the breaker closure and |
| following the small time period of the unanalyzed condition, the diesel |
| generators and the offsite power circuits are capable of performing their |
| intended design functions during parallel operations. The Onsite AC Sources |
| Limiting Conditions for Operation (LCO) specified in Technical |
| Specifications 3.8.1 and 3.8.2 are satisfied. |
| |
| "AmerGen is currently pursuing a plant modification to resolve this design |
| deficiency as soon as possible to support testing of the diesel |
| generators." |
| |
| The NRC resident inspector will be informed of this notification by the |
| licensee. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36787 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: SOUTH TEXAS REGION: 4 |NOTIFICATION DATE: 03/11/2000|
| UNIT: [1] [] [] STATE: TX |NOTIFICATION TIME: 08:24[EST]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 03/10/2000|
+------------------------------------------------+EVENT TIME: 11:48[CST]|
| NRC NOTIFIED BY: SCOTT HEAD |LAST UPDATE DATE: 03/11/2000|
| HQ OPS OFFICER: JOHN MacKINNON +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |DAVID LOVELESS R4 |
|10 CFR SECTION: | |
|NLTR LICENSEE 24 HR REPORT | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N N 0 Refueling |0 Refueling |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| ELECTRICAL SEPARATION REQUIREMENTS WERE VIOLATED |
| |
| This event notification was called under License Condition 2.G. - 24 hour |
| License Notification |
| |
| On Monday, March 6, 2000 defueling operations were ongoing at South Texas |
| Project Unit 1 to support the current refueling outage. A temporary |
| modification was installed on that date in a Safety Related "B" Train Load |
| center to supply non-class power to one of the two spent fuel pumps. This |
| modification was installed to ensure "B" train related pump remained |
| available while fuel was being loaded into the spent fuel pool. The other |
| pump was supplied with Class 1E power and was backed up by a diesel. |
| |
| The modification involved running non-class cables in and around the Class |
| 1E load center. On March 10th it was determined that since electrical |
| separation requirements [class to non-class] were violated the load center |
| was rendered inoperable as well as the electrically supported equipment. |
| One component affected was the "B" Train containment isolation valve for |
| normal purge. If the valve is inoperable Technical Specification 3.9.9, |
| Containment Ventilation Isolation System, requires that the normal purge |
| penetration be closed during core alterations. |
| |
| During the time frame that the valve was inoperable fuel movement took place |
| with the normal purge in operation (I.e., the penetration was open). This |
| is in violation of Technical Specification requirements. |
| |
| The NRC Resident Inspector will notified of this event by the licensee. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36788 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: TURKEY POINT REGION: 2 |NOTIFICATION DATE: 03/11/2000|
| UNIT: [3] [] [] STATE: FL |NOTIFICATION TIME: 16:07[EST]|
| RXTYPE: [3] W-3-LP,[4] W-3-LP |EVENT DATE: 03/11/2000|
+------------------------------------------------+EVENT TIME: 14:30[EST]|
| NRC NOTIFIED BY: DEAL |LAST UPDATE DATE: 03/11/2000|
| HQ OPS OFFICER: BOB STRANSKY +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |EDWARD MCALPINE R2 |
|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 |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|3 N N 0 Refueling |0 Refueling |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| DEFECT DETECTED DURING EDDY CURRENT INSPECTION OF STEAM GENERATOR |
| |
| "Turkey Point is performing eddy current testing of the Unit 3 steam |
| generators in accordance with Plant Technical Specifications and Industry |
| Guidance. A 20% inspection was planned for the top of tubesheet expansion |
| transition region in the hot leg of each steam generator. This total planned |
| sample was divided into 1st, 2nd and 3rd samples to address the progressive |
| inspection process of the Technical Specifications Table 4.4-2. The 1st |
| sample of 96 tubes in the 3B steam generator resulted in detection of 1 |
| defect near the top of the tubesheet. In accordance with Technical |
| Specification 4.4.5.5c, this defect results in a C-3 Classification for the |
| 1st sample. |
| |
| "A small number of additional indications near the top of the tubesheet have |
| also been detected in this steam generator, and the inspection is ongoing. |
| The inspection will be expanded to 100% of the tubes in the 3B steam |
| generator. |
| |
| "Eddy current results indicate the defects are shallow volumetric or |
| pit-like in nature. A preliminary structural evaluation indicates that all |
| indications detected to date meet the structural and leakage integrity |
| performance criteria of NEI 97-06, 'Steam Generator Program Guidelines.'" |
| |
| The NRC resident inspector has been informed of this notification by the |
| licensee. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36789 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: SURRY REGION: 2 |NOTIFICATION DATE: 03/12/2000|
| UNIT: [1] [] [] STATE: VA |NOTIFICATION TIME: 05:41[EST]|
| RXTYPE: [1] W-3-LP,[2] W-3-LP |EVENT DATE: 03/12/2000|
+------------------------------------------------+EVENT TIME: 04:49[EST]|
| NRC NOTIFIED BY: RAWLEIGH DILLARD |LAST UPDATE DATE: 03/12/2000|
| HQ OPS OFFICER: STEVE SANDIN +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |EDWARD MCALPINE R2 |
|10 CFR SECTION: | |
|ASHU 50.72(b)(1)(i)(A) PLANT S/D REQD BY TS | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 96 Power Operation |93 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| UNIT 1 COMMENCED A TECH SPEC REQUIRED SHUTDOWN AFTER DECLARING THE REFUELING |
| WATER STORAGE TANK (RWST) INOPERABLE DUE TO THE INABILITY TO CLOSE A |
| MANUALLY OPERATED RWST VALVE. |
| |
| "At 0230 hours, on 3/12/00, it was noted that a Surry Power Station Unit 1 |
| Refueling Water Storage Tank (RWST) suction valve (1-CS-27) that was being |
| used for purification purposes, could not be closed. The valve had been |
| opened under administrative control, as allowed by procedure. The Unit 1 |
| RWST was declared inoperable, and a 1 hour LCO was entered in accordance |
| with Technical Specifications. The 1 hour LCO expired at 0330 hours on |
| 3/12/00, and a 6 hour LCO to hot shutdown was entered in accordance with |
| Technical Specifications. |
| |
| "At 0449 hours, Surry Power Station Unit 1 initiated a plant shutdown from |
| 96% power as required by Technical Specifications, in order to meet the 6 |
| hour LCO. The ramp was stopped at 0459 hours at 93% power, after the valve |
| was mechanically closed. An evaluation of the valve is currently being |
| pursued. |
| |
| "This report is being made pursuant to 10CFR50.72(b)(1)(i)(A), Power Mode |
| Reduction Required by Technical Specifications." |
| |
| The valve involved is a manual grinnel diaphragm valve which appears to have |
| the spindle threads stripped. Maintenance personnel closed the valve by |
| jacking it shut. This particular valve is used for sampling and purification |
| of the RWST and does not interfere with the main suction path used for ECCS. |
| |
| |
| The licensee informed the NRC resident inspector. |
| |
| * * * UPDATE 0945EST ON 3/12/00 FROM DARLENE BROCK TO S.SANDIN * * * |
| |
| The licensee provided the following information as an update: |
| |
| "This is an update to Event #36789. At 0551 hours, on 3/12/00, 1-CS-27 was |
| closed using a Jacking device. In addition, 1-CS-31 and 1 -CS-28 (suction |
| isolation valves on the recirculation pumps) were closed. At 0657 hours the |
| Unit 1 Refueling Water Storage Tank was considered operable. The 6 hour LCO |
| to hot shutdown was exited at 0657 hours on 3/12/00 with the Unit stable at |
| 93% power. Surry Power Station Unit 1 was ramped down from 96% power to 93% |
| power during the 6 hour LCO. At 0900 hours, on 3/12/00, a ramp was |
| commenced to return power on Unit 1 to 96%." |
| |
| The licensee informed the NRC resident inspector. Notified R2DO (McAlpine). |
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Page Last Reviewed/Updated Thursday, March 25, 2021