Event Notification Report for October 04, 2000
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
10/03/2000 - 10/04/2000
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor
Event Number: 37410
Facility: POINT BEACH
Region: 3 State: WI
Unit: [1] [2] []
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: MIKE MEYER
HQ OPS Officer: LEIGH TROCINE
Region: 3 State: WI
Unit: [1] [2] []
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: MIKE MEYER
HQ OPS Officer: LEIGH TROCINE
Notification Date: 10/04/2000
Notification Time: 19:50 [ET]
Event Date: 10/04/2000
Event Time: 18:20 [CDT]
Last Update Date: 10/10/2000
Notification Time: 19:50 [ET]
Event Date: 10/04/2000
Event Time: 18:20 [CDT]
Last Update Date: 10/10/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(1)(ii)(B) - OUTSIDE DESIGN BASIS
10 CFR Section:
50.72(b)(1)(ii)(B) - OUTSIDE DESIGN BASIS
Person (Organization):
JOHN MADERA (R3)
JOHN MADERA (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 100 | Power Operation |
| 2 | N | Y | 100 | Power Operation | 100 | Power Operation |
POTENTIAL OUTSIDE DESIGN BASIS ISSUE DUE TO LACK OF DOCUMENTATION TO SHOW THAT ATMOSPHERIC STEAM DUMP VALVE CAPACITY WILL SUPPORT THE COOLDOWN ASSUMPTIONS FOR A STEAM GENERATOR TUBE RUPTURE EVENT WHEN OFFSITE POWER IS NOT AVAILABLE
The following text is a portion of a facsimile received from the licensee:
"During a main steam and steam dump system validation, no documentation could be found to show that the atmospheric steam dump (ADV) valve capacity will support the 44-minute cooldown assumptions for a steam generator tube rupture (SGTR) event as described in the FSAR when offsite power is not available. FSAR Chapter 14 and plant procedures include assumptions and procedure steps that require use of the [ADVs] to cool down the reactor coolant system (RCS) within 44 minutes of a SGTR event to allow the RCS pressure to equalize with the faulted steam generator pressure to terminate primary to secondary break flow. The SGTR radiological dose assessment is based upon this 44-minute time frame for ending flow to the ruptured steam generator."
"The current licensing basis calculation for the SGTR event calculates steam releases from the intact and the faulted steam generators for the purpose of determining offsite dose releases. However, this calculation does not appear to support a demonstration of the capability of the ADV on the intact steam generator to reduce the primary system temperature (with sufficient subcooling margin) to allow depressurizing the RCS to equilibrium pressure with the faulted steam generator. No calculation or other documentation has been located that provides an evaluation of the amount of heat in the form of steam to be removed by the intact steam generator's ADV to provide the cooldown within the 44-minute time assumed in the accident analysis. An operability determination is being performed per [Generic Letter] 91-18."
The licensee stated that there was nothing unusual or not understood and that all systems functioned as required. The licensee also stated that the units were not in any limiting conditions for operation as a result of this issue.
The licensee notified the NRC resident inspector.
* * * RETRACTION on 10/10/00 at 0934 ET by John Sell taken by MacKinnon * * *
The licensee's engineering staff has completed an Operability Determination (OD) in accordance with GL 91-18, which investigated the applicability of the existing SGTR analysis for Point Beach Nuclear Plant (PBNP). This evaluation concluded that the actual installed capacity of the ADV supports the 44 minute cooldown assumption and that there exists ample assurance that under design and licensing conditions, the PBNP physical plant can actually support a significantly shorter duration mitigation. The limiting factor for the SGTR mitigation remains the operator response and the procedures driving that response. The approved OD concluded that the ADV system is operable and fully meets performance requirements. R3DO (John Madera) notified.
The NRC Resident Inspector was notified of this event retraction by the licensee.
The following text is a portion of a facsimile received from the licensee:
"During a main steam and steam dump system validation, no documentation could be found to show that the atmospheric steam dump (ADV) valve capacity will support the 44-minute cooldown assumptions for a steam generator tube rupture (SGTR) event as described in the FSAR when offsite power is not available. FSAR Chapter 14 and plant procedures include assumptions and procedure steps that require use of the [ADVs] to cool down the reactor coolant system (RCS) within 44 minutes of a SGTR event to allow the RCS pressure to equalize with the faulted steam generator pressure to terminate primary to secondary break flow. The SGTR radiological dose assessment is based upon this 44-minute time frame for ending flow to the ruptured steam generator."
"The current licensing basis calculation for the SGTR event calculates steam releases from the intact and the faulted steam generators for the purpose of determining offsite dose releases. However, this calculation does not appear to support a demonstration of the capability of the ADV on the intact steam generator to reduce the primary system temperature (with sufficient subcooling margin) to allow depressurizing the RCS to equilibrium pressure with the faulted steam generator. No calculation or other documentation has been located that provides an evaluation of the amount of heat in the form of steam to be removed by the intact steam generator's ADV to provide the cooldown within the 44-minute time assumed in the accident analysis. An operability determination is being performed per [Generic Letter] 91-18."
The licensee stated that there was nothing unusual or not understood and that all systems functioned as required. The licensee also stated that the units were not in any limiting conditions for operation as a result of this issue.
The licensee notified the NRC resident inspector.
* * * RETRACTION on 10/10/00 at 0934 ET by John Sell taken by MacKinnon * * *
The licensee's engineering staff has completed an Operability Determination (OD) in accordance with GL 91-18, which investigated the applicability of the existing SGTR analysis for Point Beach Nuclear Plant (PBNP). This evaluation concluded that the actual installed capacity of the ADV supports the 44 minute cooldown assumption and that there exists ample assurance that under design and licensing conditions, the PBNP physical plant can actually support a significantly shorter duration mitigation. The limiting factor for the SGTR mitigation remains the operator response and the procedures driving that response. The approved OD concluded that the ADV system is operable and fully meets performance requirements. R3DO (John Madera) notified.
The NRC Resident Inspector was notified of this event retraction by the licensee.
Power Reactor
Event Number: 37411
Facility: PALO VERDE
Region: 4 State: AZ
Unit: [] [2] []
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: DAN MARKS
HQ OPS Officer: STEVE SANDIN
Region: 4 State: AZ
Unit: [] [2] []
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: DAN MARKS
HQ OPS Officer: STEVE SANDIN
Notification Date: 10/05/2000
Notification Time: 15:48 [ET]
Event Date: 10/04/2000
Event Time: 23:00 [MST]
Last Update Date: 10/05/2000
Notification Time: 15:48 [ET]
Event Date: 10/04/2000
Event Time: 23:00 [MST]
Last Update Date: 10/05/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(i) - DEG/UNANALYZED COND
10 CFR Section:
50.72(b)(2)(i) - DEG/UNANALYZED COND
Person (Organization):
DAVE LOVELESS (R4)
DAVE LOVELESS (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | N | 0 | Hot Shutdown | 0 | Cold Shutdown |
EVIDENCE OF RCS PRESSURE BOUNDARY LEAKAGE DISCOVERED DURING INSERVICE INSPECTION OF PRESSURIZER HEATER NOZZLE SLEEVE
"The following event description is based on information currently available. If through subsequent reviews of this event, additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73.
"On October 4, 2000 at approximately 23:00 MST, Palo Verde Nuclear Generating Station (PVNGS) Unit 2 inservice inspection personnel discovered evidence of reactor coolant system (RCS) pressure boundary leakage. PVNGS Unit 2 was shutdown in Mode 4 conducting cooldown and depressurization into its ninth refueling outage at the time of discovery. Currently, PVNGS Unit 2 is in Mode 5. RCS temperature is approximately 170 degrees Fahrenheit and RCS pressure is approximately 350 psia.
"The leakage was discovered at pressurizer heater nozzle sleeve A06 during inservice inspection (ISI) activities. The leakage was detected in the form of a small deposit of boron accumulation at the sleeve. PVNGS has conducted inspections of these heater sleeves during each refueling outage since the discovery that Inconel alloy 600 heater sleeves are susceptible to cracking. No evidence of leakage was detected when inspected during the last outage approximately 18 months ago. The apparent cause is primary water stress corrosion cracking (PWSCC) from the inside diameter of the sleeve.
"The timing of this ENS report was based on the determination at 10:30 MST on October 5, 2000 that the boron accumulation represented a serious degradation of a principal safety barrier. PVNGS Unit 2 Technical Specification Limiting Condition for Operation (LCO) 3.4.14 (RCS Operational Leakage) permits no reactor coolant system pressure boundary leakage. It was therefore conservatively concluded that any evidence of pressure boundary leakage, regardless of magnitude, represents serious degradation of a principal safety barrier. Technical Specification Limiting Condition for Operation 3.4.14 is applicable in Modes 1, 2, 3 and 4. Unit 2 entered Mode 5 at 01:50 MST on October 5, 2000, in compliance with LCO 3.4.14 ACTION B.2. The sleeve will be repaired or replaced prior to re-entering Mode 4.
"No ESF actuations occurred and none were required. No structures, systems or components were inoperable that contributed to this event, particularly the fuel cladding and the containment fission product barriers. The event did not result in the release of radioactivity to the environment and did not adversely affect the safe operation of the plant or the health and safety of the public."
The licensee informed the NRC resident inspector.
"The following event description is based on information currently available. If through subsequent reviews of this event, additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73.
"On October 4, 2000 at approximately 23:00 MST, Palo Verde Nuclear Generating Station (PVNGS) Unit 2 inservice inspection personnel discovered evidence of reactor coolant system (RCS) pressure boundary leakage. PVNGS Unit 2 was shutdown in Mode 4 conducting cooldown and depressurization into its ninth refueling outage at the time of discovery. Currently, PVNGS Unit 2 is in Mode 5. RCS temperature is approximately 170 degrees Fahrenheit and RCS pressure is approximately 350 psia.
"The leakage was discovered at pressurizer heater nozzle sleeve A06 during inservice inspection (ISI) activities. The leakage was detected in the form of a small deposit of boron accumulation at the sleeve. PVNGS has conducted inspections of these heater sleeves during each refueling outage since the discovery that Inconel alloy 600 heater sleeves are susceptible to cracking. No evidence of leakage was detected when inspected during the last outage approximately 18 months ago. The apparent cause is primary water stress corrosion cracking (PWSCC) from the inside diameter of the sleeve.
"The timing of this ENS report was based on the determination at 10:30 MST on October 5, 2000 that the boron accumulation represented a serious degradation of a principal safety barrier. PVNGS Unit 2 Technical Specification Limiting Condition for Operation (LCO) 3.4.14 (RCS Operational Leakage) permits no reactor coolant system pressure boundary leakage. It was therefore conservatively concluded that any evidence of pressure boundary leakage, regardless of magnitude, represents serious degradation of a principal safety barrier. Technical Specification Limiting Condition for Operation 3.4.14 is applicable in Modes 1, 2, 3 and 4. Unit 2 entered Mode 5 at 01:50 MST on October 5, 2000, in compliance with LCO 3.4.14 ACTION B.2. The sleeve will be repaired or replaced prior to re-entering Mode 4.
"No ESF actuations occurred and none were required. No structures, systems or components were inoperable that contributed to this event, particularly the fuel cladding and the containment fission product barriers. The event did not result in the release of radioactivity to the environment and did not adversely affect the safe operation of the plant or the health and safety of the public."
The licensee informed the NRC resident inspector.
Hospital
Event Number: 37412
Rep Org: WILLS EYE HOSPITAL
Licensee: WILLS EYE HOSPITAL
Region: 1
City: PHILADELPHIA State: PA
County:
License #: 37-00783-05
Agreement: N
Docket:
NRC Notified By: BEVERLY DOWNES
HQ OPS Officer: STEVE SANDIN
Licensee: WILLS EYE HOSPITAL
Region: 1
City: PHILADELPHIA State: PA
County:
License #: 37-00783-05
Agreement: N
Docket:
NRC Notified By: BEVERLY DOWNES
HQ OPS Officer: STEVE SANDIN
Notification Date: 10/05/2000
Notification Time: 16:48 [ET]
Event Date: 10/04/2000
Event Time: 00:00 [EDT]
Last Update Date: 10/05/2000
Notification Time: 16:48 [ET]
Event Date: 10/04/2000
Event Time: 00:00 [EDT]
Last Update Date: 10/05/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
35.33 - MED MISADMINISTRATION
10 CFR Section:
35.33 - MED MISADMINISTRATION
Person (Organization):
JAMES LINVILLE (R1)
M. WAYNE HODGES (NMSS)
JAMES LINVILLE (R1)
M. WAYNE HODGES (NMSS)
POTENTIAL MEDICAL MISADMINISTRATION INVOLVING MISTARGETED GAMMA-KNIFE TREATMENT
A PATIENT UNDERGOING GAMMA-KNIFE TREATMENT FOR AN ACOUSTIC TUMOR, RECEIVED APPROXIMATELY 4 GRAY TO TISSUE 8 MILLIMETERS ABOVE THE TARGET DUE TO AN ERROR INPUTTING ONE OF THREE COORDINATES DURING THE FIRST OF THREE SHOTS ON 10/4/00. THE ERROR WAS RECOGNIZED WHILE SETTING UP FOR THE REMAINING TWO SHOTS. THE PHYSICIAN WAS INFORMED AND THE TREATMENT PLAN MODIFIED SO THAT THE TARGET WOULD RECEIVE THE CORRECT DOSE. THE LICENSEE PLANS ON CONTACTING NRC REGION I TO DISCUSS REPORTABILITY OF THIS EVENT.
A PATIENT UNDERGOING GAMMA-KNIFE TREATMENT FOR AN ACOUSTIC TUMOR, RECEIVED APPROXIMATELY 4 GRAY TO TISSUE 8 MILLIMETERS ABOVE THE TARGET DUE TO AN ERROR INPUTTING ONE OF THREE COORDINATES DURING THE FIRST OF THREE SHOTS ON 10/4/00. THE ERROR WAS RECOGNIZED WHILE SETTING UP FOR THE REMAINING TWO SHOTS. THE PHYSICIAN WAS INFORMED AND THE TREATMENT PLAN MODIFIED SO THAT THE TARGET WOULD RECEIVE THE CORRECT DOSE. THE LICENSEE PLANS ON CONTACTING NRC REGION I TO DISCUSS REPORTABILITY OF THIS EVENT.