Event Notification Report for January 28, 2005
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
01/27/2005 - 01/28/2005
EVENT NUMBERS
41547
General Information or Other
Event Number: 41547
Rep Org: GENERAL ELECTRIC NUCLEAR ENERGY
Licensee: GENERAL ELECTRIC NUCLEAR ENERGY
Region: 1
City: WILMINGTON State: NC
County:
License #:
Agreement: Y
Docket:
NRC Notified By: JASON S. POST
HQ OPS Officer: STEVE SANDIN
Licensee: GENERAL ELECTRIC NUCLEAR ENERGY
Region: 1
City: WILMINGTON State: NC
County:
License #:
Agreement: Y
Docket:
NRC Notified By: JASON S. POST
HQ OPS Officer: STEVE SANDIN
Notification Date: 03/29/2005
Notification Time: 19:17 [ET]
Event Date: 01/28/2005
Event Time: 00:00 [EST]
Last Update Date: 03/29/2005
Notification Time: 19:17 [ET]
Event Date: 01/28/2005
Event Time: 00:00 [EST]
Last Update Date: 03/29/2005
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
Person (Organization):
JAMES TRAPP (R1)
MALCOLM WIDMANN (R2)
KENNETH RIEMER (R3)
GREG PICK (R4)
OMID TABATABAI (NRR)
JAMES TRAPP (R1)
MALCOLM WIDMANN (R2)
KENNETH RIEMER (R3)
GREG PICK (R4)
OMID TABATABAI (NRR)
PART 21 REPORT INVOLVING POTENTIAL TO EXCEED LOW PRESSURE TECHNICAL SPECIFICATION SAFETY LIMIT
The following is a portion of a facsimile submitted by GE Energy-Nuclear:
"The defect is a calculation of an anticipated operational occurrence (AOO), which predicts that the Pressure Regulator Failure Maximum Demand (Open) (PRFO) transient will be terminated by a high water level trip as a result of level swell in the reactor. An improved (and approved) model predicts that MSIV closure will occur when steam line pressure reaches the low-pressure isolation setpoint (LPIS), rather than terminate due to a high water level trip. Depending upon the plant-specific response to a PRFO, including the value of the LPIS, reactor steam dome pressure could decrease to below 785 psig while thermal power exceeds 25% of rated, which would be a violation of SL 2.1.1.1. This constitutes a Defect as defined in 10 CFR 21.3, even though there is no safety hazard created. SL 2.1.1.1 was intended to protect fuel cladding integrity during startup conditions without the need to perform a Critical Power Ratio (CPR) calculation. The AOO of concern is a transient from normal operating conditions that causes CPR to increase, so the event produces additional margin to the Minimum Critical Power Ratio Safety Limit (SLMCPR) and does not threaten fuel cladding integrity. The LPIS should not be considered as a Limiting Safety System Setting (LSSS) for SL 2.1.1.1 since it does not provide a 'significant safety function' with regard to protecting fuel cladding integrity. This indicates that SL 2.1.1.1 is overly conservative because an event that causes CPR to increase and does not threaten fuel cladding integrity, may result in exceeding a reactor core SL."
Notified Plants:
AFFECTED: Nine Mile Point 2, Fermi 2, Pilgrim, Vermont Yankee, Limerick 1 & 2, Peach Bottom 2 & 3, Perry 1, and Hope Creek.
POTENTIALLY AFFECTED: Clinton, Oyster Creek, Brunswick 1 & 2, Nine Mile Point 1, FitzPatrick, Grand Gulf, River Bend, Dresden 2 & 3, LaSalle 1 & 2, Quad Cities 1 & 2, Cooper, Duane Arnold, Monticello, Hatch 1 & 2, Browns Ferry 1, 2 & 3.
The following is a portion of a facsimile submitted by GE Energy-Nuclear:
"The defect is a calculation of an anticipated operational occurrence (AOO), which predicts that the Pressure Regulator Failure Maximum Demand (Open) (PRFO) transient will be terminated by a high water level trip as a result of level swell in the reactor. An improved (and approved) model predicts that MSIV closure will occur when steam line pressure reaches the low-pressure isolation setpoint (LPIS), rather than terminate due to a high water level trip. Depending upon the plant-specific response to a PRFO, including the value of the LPIS, reactor steam dome pressure could decrease to below 785 psig while thermal power exceeds 25% of rated, which would be a violation of SL 2.1.1.1. This constitutes a Defect as defined in 10 CFR 21.3, even though there is no safety hazard created. SL 2.1.1.1 was intended to protect fuel cladding integrity during startup conditions without the need to perform a Critical Power Ratio (CPR) calculation. The AOO of concern is a transient from normal operating conditions that causes CPR to increase, so the event produces additional margin to the Minimum Critical Power Ratio Safety Limit (SLMCPR) and does not threaten fuel cladding integrity. The LPIS should not be considered as a Limiting Safety System Setting (LSSS) for SL 2.1.1.1 since it does not provide a 'significant safety function' with regard to protecting fuel cladding integrity. This indicates that SL 2.1.1.1 is overly conservative because an event that causes CPR to increase and does not threaten fuel cladding integrity, may result in exceeding a reactor core SL."
Notified Plants:
AFFECTED: Nine Mile Point 2, Fermi 2, Pilgrim, Vermont Yankee, Limerick 1 & 2, Peach Bottom 2 & 3, Perry 1, and Hope Creek.
POTENTIALLY AFFECTED: Clinton, Oyster Creek, Brunswick 1 & 2, Nine Mile Point 1, FitzPatrick, Grand Gulf, River Bend, Dresden 2 & 3, LaSalle 1 & 2, Quad Cities 1 & 2, Cooper, Duane Arnold, Monticello, Hatch 1 & 2, Browns Ferry 1, 2 & 3.