| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
09/26/2011 - 09/27/2011 ** EVENT NUMBERS ** |
 | General Information | Event Number: 46230 | Rep Org: GE HITACHI NUCLEAR ENERGY
Licensee: GE HITACHI NUCLEAR ENERGY
Region: 1
City: WILMINGTON State: NC
County:
License #:
Agreement: Y
Docket:
NRC Notified By: DALE E. PORTER
HQ OPS Officer: ERIC SIMPSON | Notification Date: 09/03/2010
Notification Time: 15:23 [ET]
Event Date: 09/03/2010
Event Time: [EDT]
Last Update Date: 09/27/2011 | Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
| Person (Organization):
RICHARD CONTE (R1DO)
EUGENE GUTHRIE (R2DO)
TAMARA BLOOMER (R3DO)
RICK DEESE (R4DO)
MIKE CHEOK (NRR)
PART 21 GP via email ()
| Event Text PART 21 - FAILURE TO INCLUDE SEISMIC INPUT IN REACTOR CONTROL BLADE CUSTOMER GUIDANCE
The following is text of a facsimile submitted by the vendor:
"GE Hitachi Nuclear Energy (GEH) has identified that engineering evaluations that support the guidance provided in SC 08-05, Revision 1, do not address the potential impact of a seismic event on the ability to scram as it relates to the channel-control blade interference issue. Note that the seismic loads are not a consideration in the scram timing, but rather the ability to insert the control blades. In other words, the control blades must be capable of inserting during the seismic event, but not to the timing requirements of the Technical Specifications. GEH is evaluating the impact of the seismic loads between the fuel channel and the control blade associated with an Operating Basis Earthquake (OBE), and a Safe Shutdown Earthquake (SSE) on BWR/2-5 plants. The scram capability is expected to be affected due to the added seismic loads at low reactor pressures in the BWR/2-5 plants. The ability to scram for the BWR/6 plants is not adversely affected by the seismic events. Additional evaluation is required to determine to what extent the maximum allowable friction limits specified for the BWR/2-5 plants in SC 08-05 Revision 1 is affected by the addition of seismic loads.
"GEH issues this 60-Day Interim Report in accordance with the requirements set forth in 10 CFR 21.21 (a)(2) to allow additional time to for this evaluation to be completed."
Affected US plants previously notified by vendor and recommended for surveillance program include: Nine Mile Point, Units 1 and 2; Fermi 2; Columbia; FitzPatrick; Pilgrim; Vermont Yankee; Grand Gulf; River Bend; Clinton; Oyster Creek; Dresden, Units 2 and 3; LaSalle, Units 1 and 2; Limerick, Units 1 and 2; Peach Bottom, Units 2 and 3; Quad Cities, Units 1 and 2; Perry, Unit 1; Duane Arnold; Cooper; Monticello; Brunswick, Units 1 and 2; Hope Creek; Hatch, Units 1 and 2; and Browns Ferry, Units 1and 2.
Affected US plants previously notified by vendor and provided information include: Susquehanna, Units 1 and 2 and Browns Ferry, Unit 3.
* * * UPDATE FROM DALE PORTER TO ERIC SIMPSON AT 1556 ON 09/27/2010 * * *
The following update was received via fax:
"This letter provides a revision to the information transmitted on September 2, 2010 in MFN 10-245 concerning an evaluation being performed by GE Hitachi Nuclear Energy (GEH) regarding the failure to include seismic input in channel-control blade interference customer guidance. Two changes have been made in Revision 1:
"1) A statement was added regarding the applicability of this issue to the ABWR and ESBWR design certification documentation.
"2) The original MFN 10-245 referenced the Safety Communication SC 08-05 R1 that was transmitted to the US NRC via MFN 08-420. The references to SC 08-05 were changed to MFN 08-420 to prevent possible confusion.
"As stated herein, GEH has not concluded that this is a reportable condition in accordance with the requirements of 10CFR 21.21(d) and continued evaluation is required to determine the impact of a seismic event on the guidance contained in MFN 08-420."
Notified the R1DO (Gray), R2DO (Hopper), R3DO (Orth), R4DO (Farnholtz), NRR EO (Lee) and Part 21 Group (via email).
* * * UPDATE FROM DALE PORTER TO MARK ABRAMOVITZ AT 1723 ON 12/15/2010 * * *
The following update was received via fax:
"This letter provides information concerning an on-going evaluation being performed by GE Hitachi Nuclear Energy (GEH) regarding the failure to include seismic loads in the guidance provided in MFN 08-420. As stated herein, GEH has not concluded that this is a reportable condition in accordance with the requirements of 10CFR21.21(d) and continued evaluation is required to determine the impact of a seismic event on the guidance contained in MFN 08-420.
"GEH has not completed the evaluation of the impact of the seismic loads between the fuel channel and the control blade associated with an Operating Basis Earthquake (OBE), and a Safe Shutdown Earthquake (SSE) on BWR/2-5 plants."
GEH expects the task to be completed by August 15, 2011.
Notified the R1DO (Holody), R2DO (Henson), R3DO (Kozak), R4DO (Werner), NRR EO (Evans) and Part 21 Group (via email).
* * * UPDATE AT 1808 EDT ON 08/11/11 FROM DALE PORTER TO JOE O'HARA * * *
The following was received via fax:
"GE Hitachi Nuclear Energy (GEH) identified, in July 2010, that engineering evaluations did not address the potential impact of a seismic event on the ability to scram as it relates to the channel-control blade interference issue. GEH provided status of the on-going evaluation in [December 2010]. GEH has not completed the evaluation of the impact of the seismic loads between the fuel channel and the control blade associated with a bounding Safe Shutdown Earthquake (SSE) on BWR/2-5 plants. The scram capability is expected to be affected due to the added seismic loads at low reactor pressures [less than 1000 psig] in the BWR/2-5 plants. Additional evaluations are required to determine to what extent the maximum allowable friction limits specified for the BWR/2-5 plants are affected by the addition of SSE seismic loads at low reactor pressures.
"GEH issues this 60-Day Interim Report in accordance with the requirements set forth in 10CFR 21.21 (a)(2) to allow additional time for this evaluation to be completed."
The following sites are noted as having channel-control blade concerns:
Region 1: Nine Mile Point, Fitzpatrick, Pilgrim, Vermont Yankee, Oyster Creek, Limerick, Peach Bottom, Susquehanna, and Hope Creek
Region 2: Browns Ferry, Brunswick, Hatch,
Region 3: Fermi, Clinton, Dresden, LaSalle, Quad Cities, Perry, Duane Arnold, Monticello
Region 4: Columbia, Grand Gulf, River Bend, Cooper.
Notified R1DO (Powell), R2DO (Hopper), R3DO (Dickson), R4DO (Farnholtz) and NRR Part 21 Grp via email.
* * * UPDATE AT 0037 EDT ON 9/27/11 FROM PORTER TO HUFFMAN VIA E-MAIL * * *
The following is a summary of information received from GE Hitachi Nuclear Energy via e-mail of a letter, Reference MFN 10-245 R4, addressed to the NRC and dated September 26, 2011:
"GE Hitachi (GEH) has determined that the scram capability of the control rod drive mechanism in BWR/2-5 plants may not be sufficient to ensure the control rod will fully insert in a cell with channel-control rod friction at or below the friction limits specified in MFN 08-420 with a concurrent Safe Shutdown Earthquake (SSE). The plant condition for which incomplete control rod insertion might occur is when the reactor is below normal operating pressure (<900 psig) and a scram occurs concurrent with the SSE, for Mark I containment plants, and for the SSE with concurrent Loss-of-Coolant Accident (LOCA) and Safety Relief Valve (SRV) events for Mark II containment plants. In this scenario a Substantial Safety Hazard results because the affected control rods might not fully insert to perform the required safety function.
"GEH has determined that when channel-control blade interference is present at reduced reactor pressure and at friction levels considered acceptable in MFN 08-420, a simultaneously occurring Safe Shutdown Earthquake (SSE) may result in control rod friction that inhibits the full insertion of the affected control rods during a reactor scram from these conditions. This scenario was not explicitly considered in MFN 08-420.
"GEH has also quantified maximum allowable control rod friction for channel-control blade interference during the SSE with reactor system pressure greater than or equal to 900 psig. The previous conclusion regarding the scram capability for the BWR/2-5 plants, last communicated in MFN 10-245 R2, was based upon a reactor system pressure of 1000 psig. The updated evaluation at 900 psig has resulted in modifications to the guidance specified in MFN 08-420.
"The GE Hitachi Letter recommends testing with new allowable friction limits that will ensure control rods fully insert at low reactor pressure concurrent with an SSE (for Mark I containment plants) and SSE with concurrent LOCA (for Mark II containment plants). The enclosure in the GEH letter provides a description of the evaluation, with surveillance recommendations for BWR/2-5 plants. The recommended surveillance is intended to augment the surveillance requirements in the plant Technical Specifications and define populations of control rods to be tested, and the method for testing, until other actions that mitigate or limit the potential for channel control blade interference can be identified and implemented.
"Based upon the evaluation, GEH has concluded that a Reportable Condition under 10CFR Part 21 exists for BWR/2-5 plants. This determination does not apply to BWR/6 or ABWR plants or the ABWR/ESBWR Design Control Document's (DCD). The information contained in this document informs the NRC of the conclusions and recommendations derived from GEH's evaluation of this issue."
The list of potentially affected plants has previously been noted in this Part 21 notification and have been previously notified by GE Hitachi of the concern.
Notified R1DO (Doerflein), R2DO (Lesser), R3DO (Passehl), R4DO (Werner) and NRR Part 21 Grp via email. |
| Agreement State | Event Number: 47281 | Rep Org: PA BUREAU OF RADIATION PROTECTION
Licensee: THOMAS JEFFERSON UNIVERSITY HOSPITAL
Region: 1
City: PHILADELPHIA State: PA
County:
License #: PA-0130
Agreement: Y
Docket:
NRC Notified By: JOE MELNIC
HQ OPS Officer: HOWIE CROUCH | Notification Date: 09/20/2011
Notification Time: 10:27 [ET]
Event Date: 09/16/2011
Event Time: 14:14 [EDT]
Last Update Date: 09/20/2011 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
CHRISTOPHER NEWPORT (R1DO)
ANGELA MCINTOSH (FSME)
| Event Text AGREEMENT STATE REPORT - MEDICAL EVENT INVOLVING Y-90 THERASPHERE UNDERDOSE
The following information was received from the Commonwealth of Pennsylvania via facsimile:
"Event Type: A medical event (ME) involving Y-90 Theraspheres where the patient received 51% of the intended dose, which is reportable under 10CFR35.3045(a)(1)(ii).
'Notifications: On September 16, 2011, at 1414 [EDT], the Department's Southeast Regional Office received notification via phone message about the ME.
"Event Description. The patient was being treated with MDS Nordion Y-90 glass Theraspheres for transarterial radioembolization. A suspected defective catheter caused 49% of the intended dose to clog up in the catheter. No harm to the patient is expected. The referring physician and patient have been notified. No more information is available at this time.
"Cause of the event: The cause of the event is suspected to be a defective catheter.
"Actions: After decay, the catheter will be returned to Nordion for inspection and may also be returned to the manufacturer, Terumo Medical, for a defect analysis. The licensee will be submitting a written report within 15 days. The [Pennsylvania] Department [of Environmental Protection] plans to do a reactive inspection."
PA Report ID: PA110025
A Medical Event may indicate potential problems in a medical facility's use of radioactive materials. It does not necessarily result in harm to the patient. |
| Agreement State | Event Number: 47283 | Rep Org: TEXAS DEPARTMENT OF HEALTH
Licensee: CARIBBEAN INSPECTION & NDT SERVICES INC
Region: 4
City: PORT LAVACA State: TX
County:
License #: 06420
Agreement: Y
Docket:
NRC Notified By: ART TUCKER
HQ OPS Officer: JOE O'HARA | Notification Date: 09/20/2011
Notification Time: 16:39 [ET]
Event Date: 09/19/2011
Event Time: 19:43 [CDT]
Last Update Date: 09/21/2011 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
WAYNE WALKER (R4DO)
KEVIN O'SULLIVAN (FSME)
| Event Text POTENTIAL OVEREXPOSURE DUE TO FAULTY RADIOGRAPHY DEVICE
The following was received via e-mail:
"On September 19, 2011, at 1943 hours, the Agency [state] received an email stating that a radiography trainee may have received an over exposure to his right hand and was seeking medical attention. The email stated that the overexposure occurred because the radiography device used on the job was faulty, but did not provide any information on when or how the possible overexposure occurred.
"On September 20, 2011, the Agency received an email from a licensee Radiation Safety Officer (RSO) stating that an overexposure may have occurred to an employee's hands. The email stated that the licensee had not received any information from the individual who was reported to have received the exposure. The RSO was in route to a hospital in Houston, Texas where the radiographer trainee was reported to have gone for treatment. The employee's film badge has been sent for processing, but no results are available at this time. The licensee is reviewing records to determine where and when the trainee worked during the two months he has been employed. Individuals that worked with the trainee are being interviewed.
"Additional information will be provided as it is received in accordance with SA-300."
Texas Incident # I-8886
* * * UPDATE FROM ART TUCKER TO VINCE KLCO ON 9/21/2011 AT 1820 EDT* * *
The following information was received by facsimile:
"The licensee has reported that the trainee stated that on September 12, 2011, while conducting radiography operations in the field, he removed the guide tube from an Amersham 660 D radiography camera containing 73 curies of Iridium - 192 and saw that the source was protruding out of the camera. The licensee stated that they did not know how far the source was protruding or how it was returned to the fully shielded position. The Agency [state] has contacted the trainee and conducted an interviewed with him over the phone.
"The licensee stated that the results of the trainee's film badge indicated that he received 1,410 millirem on the film badge he was wearing at the time of the event. The trainee is in a Houston, Texas hospital. His doctors are conferring with [the] Radiation Emergency Assistance Center/Training Site (REAC/TS) regarding his medical treatment. An on-site investigation will be performed by the Agency at the licensee's location on September 22, 2011."
Notified the R4DO (Walker) and FSME (O'Sullivan). |
| Power Reactor | Event Number: 47289 | Facility: SAN ONOFRE
Region: 4 State: CA
Unit: [ ] [2] [3]
RX Type: [1] W-3-LP,[2] CE,[3] CE
NRC Notified By: DOUGLAS FOOTE
HQ OPS Officer: STEVE SANDIN | Notification Date: 09/23/2011
Notification Time: 00:27 [ET]
Event Date: 09/22/2011
Event Time: 17:30 [PDT]
Last Update Date: 09/26/2011 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
| Person (Organization):
WAYNE WALKER (R4DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | Y | 100 | Power Operation | 100 | Power Operation | | 3 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text OFFSITE NOTIFICATION TO COUNTY AGENCY DUE TO INADVERTENT WASTEWATER DISCHARGE
"This notification is for San Onofre Units 2 and 3, and is being made in accordance with 10CFR50.72(b)(2)(xi) to report a notification to another government agency regarding a spill of approximately 26000 gallons of untreated waste water (sewage).
"A pending report will be made to the San Diego County Dept. of Environmental Health regarding a untreated (sewage) waste water spill that occurred as San Onofre. The details that were communicated to the DEH are that at 1730 PDT a sewage spill was identified at SONGS in the North Industrial yard area. The leak was being collected in the yard drains and being discharged to the ocean. The sump discharge system has been secured stopping the release."
The licensee will inform the NRC Resident Inspector.
* * * UPDATE FROM LEE KELLY TO JOHN KNOKE AT 1325 EDT ON 9/26/11 * * *
"On September 22, 2011, an untreated (sewage) waste water spill was discovered in the San Onofre North Industrial Area. Southern California Edison (SCE) notified the NRC at 1730 PDT on September 22, 2011, (Event No. 47289) in accordance with 10CFR50.72(b)(2)(xi). The California Emergency Management Agency was notified (Control #11-5652) at 2223 PDT and San Diego County Department of Environmental Health (DEH) was notified at 2239 PDT. The initial report to the NRC indicated the spill was approximately 26,000 gallons. The estimated size of the spill has subsequently been revised to approximately 7,000 gallons.
"At the time of the event, Units 2 and 3 were at 100 percent power. SCE notified the NRC Resident Inspectors about this occurrence and will provide them with a copy of this report."
Notified the R4DO (Greg Werner). |
| Power Reactor | Event Number: 47293 | Facility: ROBINSON
Region: 2 State: SC
Unit: [2] [ ] [ ]
RX Type: [2] W-3-LP
NRC Notified By: RICHARD ROGALSKI
HQ OPS Officer: JOHN KNOKE | Notification Date: 09/26/2011
Notification Time: 14:48 [ET]
Event Date: 09/26/2011
Event Time: 11:45 [EDT]
Last Update Date: 09/26/2011 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
MARK LESSER (R2DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | A/R | Y | 100 | Power Operation | 0 | Hot Standby | Event Text AUTOMATIC REACTOR TRIP DUE TO ONE LOOP LOW FLOW SIGNAL
"At 1145 hours EDT on September 26, 2011, with the unit in Mode 1 at 100% power, an automatic reactor trip occurred. The reactor trip signal was based on the One Loop Low Flow reactor protection function. All Reactor Coolant Pumps (RCP) remained running.
"The Auxiliary Feedwater System automatically actuated, as expected due to low steam generator water level, and provided feedwater to the steam generators. The steam generator and pressurizer Power Operated Relief Valves (PORVs) and the Main Steam Safety Valves did not open during the event. All control rods indicated fully inserted following the reactor trip.
"The normal post-trip electrical lineup is providing power to the plant and the electrical system is stable. Decay heat is being removed via the condenser. The reactor is currently stable in Mode 3.
"The cause of the reactor trip is under investigation." |
| Power Reactor | Event Number: 47294 | Facility: THREE MILE ISLAND
Region: 1 State: PA
Unit: [1] [ ] [ ]
RX Type: [1] B&W-L-LP,[2] B&W-L-LP
NRC Notified By: DOUG GORSE
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 09/26/2011
Notification Time: 15:09 [ET]
Event Date: 09/26/2011
Event Time: 15:00 [EDT]
Last Update Date: 09/26/2011 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
LAWRENCE DOERFLEIN (R1DO)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text NEW RIVER HYDRAULIC ANALYSIS RAISES MAXIMUM FLOOD LEVEL
"TMI new river analysis indicates level above existing UFSAR [Updated Final Safety Analysis Report] flood analysis. At about 15:00 [EDT] on September 26, 2011, a revised River Stage Discharge Analysis was completed and concluded that the Probable Maximum Flood (PMF) water level is higher than previously described in the safety analysis report. This unanalyzed condition is reportable in accordance with 10 CFR 50.72(b)(3)(ii)(B).
"Actions to protect the plant for the higher PMF river level have been implemented. The flood barrier gates that protect safety related equipment have been modified to accommodate the revised river levels. No onsite flood water levels have occurred that could potentially challenge the existing flood barrier system.
"The licensee notified the NRC Resident Inspector."
The licensee will be making a courtesy media notification. | |