Degraded Ability of Steam Generators to Remove Decay Heat by Natural CirculationML031060293 |
Person / Time |
---|
Site: |
Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant |
---|
Issue date: |
08/28/1995 |
---|
From: |
Crutchfield D Office of Nuclear Reactor Regulation |
---|
To: |
|
---|
References |
---|
IN-95-035, NUDOCS 9508220031 |
Download: ML031060293 (11) |
|
Similar Documents at Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant |
---|
Category:NRC Information Notice
MONTHYEARInformation Notice 2020-02, Flex Diesel Generator Operational Challenges2020-09-15015 September 2020 Flex Diesel Generator Operational Challenges ML20225A0322020-09-0303 September 2020 NRC Choice Letter to NAC International with Attached Safety Inspection Report, IR 0721015/2020201, February 24-27, 2020 and July 22, 2020, Inspection of NAC International in Norcross, Georgia Information Notice 2012-09, PWROG-16043-NP-A, Revision 2, PWROG Program to Address NRC Information Notice 2012-09: Irradiation Effects on Fuel Assembly Spacer Grid Crush Strength for Westinghouse and CE PWR Fuel Designs.2019-11-30030 November 2019 PWROG-16043-NP-A, Revision 2, PWROG Program to Address NRC Information Notice 2012-09: Irradiation Effects on Fuel Assembly Spacer Grid Crush Strength for Westinghouse and CE PWR Fuel Designs. Information Notice 2011-20, NRC060 - NRC Information Notice 2011-20: Concrete Degradation by Alkali-Silica Reaction (Nov. 18, 2011)2019-07-24024 July 2019 NRC060 - NRC Information Notice 2011-20: Concrete Degradation by Alkali-Silica Reaction (Nov. 18, 2011) ML19196A2452019-07-15015 July 2019 Public Notice - Sequoyah Nuclear Plant, Unit 2 - Exigent Amendment to Facility Operating License Information Notice 2019-01, Inadequate Evaluation of Temporary Alterations2019-03-12012 March 2019 Inadequate Evaluation of Temporary Alterations ML16028A3082016-04-27027 April 2016 NRC Information Notice; IN 2016-05: Operating Experience Regarding Complications From a Loss of Instrumentation Air Information Notice 2015-05, Inoperability of Auxiliary and Emergency Feedwater Auto Start Circuits on Loss of Main Feedwater Pumps2015-05-12012 May 2015 Inoperability of Auxiliary and Emergency Feedwater Auto Start Circuits on Loss of Main Feedwater Pumps Information Notice 2015-05, Inoperability Of Auxiliary And Emergency Feedwater Auto Start Circuits On Loss Of Main Feedwater Pumps2015-05-12012 May 2015 Inoperability Of Auxiliary And Emergency Feedwater Auto Start Circuits On Loss Of Main Feedwater Pumps Information Notice 2013-20, OFFICIAL EXHIBIT - NYS000538-00-BD01 - NRC Information Notice 2013-20: Steam Generator Channel Head and Tubesheet Degradation (October 3, 2013) (ML13204A143)2013-10-0303 October 2013 OFFICIAL EXHIBIT - NYS000538-00-BD01 - NRC Information Notice 2013-20: Steam Generator Channel Head and Tubesheet Degradation (October 3, 2013) (ML13204A143) Information Notice 2013-20, Official Exhibit - NYS000538-00-BD01 - NRC Information Notice 2013-20: Steam Generator Channel Head and Tubesheet Degradation (October 3, 2013) (ML13204A143)2013-10-0303 October 2013 Official Exhibit - NYS000538-00-BD01 - NRC Information Notice 2013-20: Steam Generator Channel Head and Tubesheet Degradation (October 3, 2013) (ML13204A143) Information Notice 2013-11, OFFICIAL EXHIBIT - NYS000551-00-BD01 - NRC Information Notice 2013-11: Crack-Like Indication at Dents/Dings and in the Freespan Region of Thermally Treated Alloy 600 Steam Generator Tubes (July 3, 2013)2013-07-0303 July 2013 OFFICIAL EXHIBIT - NYS000551-00-BD01 - NRC Information Notice 2013-11: Crack-Like Indication at Dents/Dings and in the Freespan Region of Thermally Treated Alloy 600 Steam Generator Tubes (July 3, 2013) Information Notice 2013-11, Official Exhibit - NYS000551-00-BD01 - NRC Information Notice 2013-11: Crack-Like Indication at Dents/Dings and in the Freespan Region of Thermally Treated Alloy 600 Steam Generator Tubes (July 3, 2013)2013-07-0303 July 2013 Official Exhibit - NYS000551-00-BD01 - NRC Information Notice 2013-11: Crack-Like Indication at Dents/Dings and in the Freespan Region of Thermally Treated Alloy 600 Steam Generator Tubes (July 3, 2013) Information Notice 2010-12, Intervenors' Fifth Motion to Amend and/or Supplement Proposed Contention No. 5 (Shield Building Cracking). Appendix VI: NRC FOIA Responses (B-51 Through B-53); Turkey Point Event Report; NRC Information Notice 2010-12: Contain2012-08-17017 August 2012 Intervenors' Fifth Motion to Amend and/or Supplement Proposed Contention No. 5 (Shield Building Cracking). Appendix VI: NRC FOIA Responses (B-51 Through B-53); Turkey Point Event Report; NRC Information Notice 2010-12: Containment Liner Cor Information Notice 2010-12, Intervenors' Fifth Motion to Amend and/or Supplement Proposed Contention No. 5 (Shield Building Cracking). Appendix VI: NRC FOIA Responses (B-51 Through B-53); Turkey Point Event Report; NRC Information Notice 2010-12: Con2012-08-17017 August 2012 Intervenors' Fifth Motion to Amend and/or Supplement Proposed Contention No. 5 (Shield Building Cracking). Appendix VI: NRC FOIA Responses (B-51 Through B-53); Turkey Point Event Report; NRC Information Notice 2010-12: Containment Liner Cor Information Notice 2010-12, Intervenors' Fifth Motion to Amend And/Or Supplement Proposed Contention No. 5 (Shield Building Cracking). Appendix VI: NRC FOIA Responses (B-51 Through B-53); Turkey Point Event Report; NRC Information Notic2012-08-17017 August 2012 Intervenors' Fifth Motion to Amend And/Or Supplement Proposed Contention No. 5 (Shield Building Cracking). Appendix VI: NRC FOIA Responses (B-51 Through B-53); Turkey Point Event Report; NRC Information Notice 2010-12: Containment Liner Cor Information Notice 2012-13, Boraflex Degradation Surveillance Programs and Corrective Actions in the Spent Fuel Pool2012-08-10010 August 2012 Boraflex Degradation Surveillance Programs and Corrective Actions in the Spent Fuel Pool Information Notice 2012-13, Boraflex Degradation Surveillance Programs And Corrective Actions In The Spent Fuel Pool2012-08-10010 August 2012 Boraflex Degradation Surveillance Programs And Corrective Actions In The Spent Fuel Pool Information Notice 2012-11, Age Related Capacitor Degradation2012-07-23023 July 2012 Age Related Capacitor Degradation ML12031A0132012-02-0606 February 2012 U.S. Nuclear Regulatory Commission Investigation Report No. 2-2010-058, Cpn International, Inc Information Notice 2011-19, Licensee Event Reports Containing Information Pertaining to Defects to Basic Components2011-09-26026 September 2011 Licensee Event Reports Containing Information Pertaining to Defects to Basic Components Information Notice 2011-15, Steel Containment Degradation and Associated License Renewal Aging Management Issues2011-08-0101 August 2011 Steel Containment Degradation and Associated License Renewal Aging Management Issues Information Notice 2011-17, Calculation Methodologies for Operability Determinations of Gas Voids in Nuclear Power Plant Piping2011-07-26026 July 2011 Calculation Methodologies for Operability Determinations of Gas Voids in Nuclear Power Plant Piping Information Notice 2011-13, Official Exhibit - NYS000329-00-BD01 - NRC Information Notice 2011-13, Control Rod Blade Cracking Resulting in Reduced Design Lifetime (Jun 29, 2011) (NRC in 2011-13)2011-06-29029 June 2011 Official Exhibit - NYS000329-00-BD01 - NRC Information Notice 2011-13, Control Rod Blade Cracking Resulting in Reduced Design Lifetime (Jun 29, 2011) (NRC in 2011-13) Information Notice 2011-13, Official Exhibit - Nys000329-00-Bd01 - NRC Information Notice 2011-13, Control Rod Blade Cracking Resulting in Reduced Design Lifetime (Jun 29, 2011) (Nrc in 2011-13)2011-06-29029 June 2011 Official Exhibit - Nys000329-00-Bd01 - NRC Information Notice 2011-13, Control Rod Blade Cracking Resulting in Reduced Design Lifetime (Jun 29, 2011) (Nrc in 2011-13) Information Notice 2011-13, OFFICIAL EXHIBIT - NYS000329-00-BD01 - NRC Information Notice 2011-13, Control Rod Blade Cracking Resulting in Reduced Design Lifetime (Jun 29, 2011) (NRC in 2011-13)2011-06-29029 June 2011 OFFICIAL EXHIBIT - NYS000329-00-BD01 - NRC Information Notice 2011-13, Control Rod Blade Cracking Resulting in Reduced Design Lifetime (Jun 29, 2011) (NRC in 2011-13) Information Notice 2011-04, IN: Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking in Stainless Steel Piping in Pressurized Water Reactors2011-02-23023 February 2011 IN: Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking in Stainless Steel Piping in Pressurized Water Reactors Information Notice 2011-04, In: Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking in Stainless Steel Piping in Pressurized Water Reactors2011-02-23023 February 2011 In: Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking in Stainless Steel Piping in Pressurized Water Reactors Information Notice 2011-04, in: Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking in Stainless Steel Piping in Pressurized Water Reactors2011-02-23023 February 2011 in: Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking in Stainless Steel Piping in Pressurized Water Reactors Information Notice 2010-26, New England Coalition'S Motion for Leave to Reply to NRC Staff'S Objection to Nec'S Notification of Information Notice 2010-26 and Entergy'S Response to the Supplement to Nec'S Petition for Commission Review of LBP-10-2010-12-30030 December 2010 New England Coalition'S Motion for Leave to Reply to NRC Staff'S Objection to Nec'S Notification of Information Notice 2010-26 and Entergy'S Response to the Supplement to Nec'S Petition for Commission Review of LBP-10-19 Information Notice 2010-26, New England Coalition'S Motion for Leave to Reply to NRC Staff'S Objection to Nec'S Notification of Information Notice 2010-26 and Entergy'S Response to the Supplement to Nec'S Petition for Commission Review2010-12-30030 December 2010 New England Coalition'S Motion for Leave to Reply to NRC Staff'S Objection to Nec'S Notification of Information Notice 2010-26 and Entergy'S Response to the Supplement to Nec'S Petition for Commission Review of LBP-10-19 Information Notice 2010-26, 2010/12/21-NRC Staff'S Objection to Nec'S Notification of Information Notice 2010-262010-12-21021 December 2010 2010/12/21-NRC Staff'S Objection to Nec'S Notification of Information Notice 2010-26 ML13066A1872009-12-16016 December 2009 Draft NRC Information Notice 2009-xx - Underestimate of Dam Failure Frequency Used in Probabilistic Risk Assessments ML1007804482009-11-23023 November 2009 Email from Peter Bamford, NRR to Pamela Cowan, Exelon on TMI Contamination Control Event Information Notice 2009-11, NSP000059-Revised Prefiled Testimony of Northard/Petersen/Peterson-NRC Information Notice 2009-112009-07-0707 July 2009 NSP000059-Revised Prefiled Testimony of Northard/Petersen/Peterson-NRC Information Notice 2009-11 Information Notice 2009-10, Official Exhibit - NYS000019-00-BD01- NRC Information Notice 2009-10, Transformers Failures - Recent Operating Experience (Jul. 7, 2009) (NRC in 2009-10)2009-07-0707 July 2009 Official Exhibit - NYS000019-00-BD01- NRC Information Notice 2009-10, Transformers Failures - Recent Operating Experience (Jul. 7, 2009) (NRC in 2009-10) Information Notice 2009-09, Improper Flow Controller Settings Renders Injection Systems Inoperable and Surveillance Did Not Identify2009-06-19019 June 2009 Improper Flow Controller Settings Renders Injection Systems Inoperable and Surveillance Did Not Identify Information Notice 2008-12, Reactor Trip Due to Off-Site Power Fluctuation2008-07-0707 July 2008 Reactor Trip Due to Off-Site Power Fluctuation Information Notice 2008-11, Service Water System Degradation at Brunswicksteam Electric Plant Unit 12008-06-18018 June 2008 Service Water System Degradation at Brunswicksteam Electric Plant Unit 1 Information Notice 2008-04, Counterfeit Parts Supplied to Nuclear Power Plants2008-04-0707 April 2008 Counterfeit Parts Supplied to Nuclear Power Plants Information Notice 1991-09, Counterfeiting of Crane Valves2007-09-25025 September 2007 Counterfeiting of Crane Valves Information Notice 2007-28, Potential Common Cause Vulnerabilities in Essential Service Water Systems Due to Inadequate Chemistry Controls2007-09-19019 September 2007 Potential Common Cause Vulnerabilities in Essential Service Water Systems Due to Inadequate Chemistry Controls Information Notice 2007-29, Temporary Scaffolding Affects Operability of Safety-Related Equipment2007-09-17017 September 2007 Temporary Scaffolding Affects Operability of Safety-Related Equipment Information Notice 2007-14, Loss of Offsite Power and Dual-Unit Trip at Catawba Nuclear Generating Station2007-03-30030 March 2007 Loss of Offsite Power and Dual-Unit Trip at Catawba Nuclear Generating Station Information Notice 2007-06, Potential Common Cause Vulnerabilities in Essential Service Water Systems2007-02-0909 February 2007 Potential Common Cause Vulnerabilities in Essential Service Water Systems Information Notice 2007-05, Vertical Deep Draft Pump Shaft and Coupling Failures2007-02-0909 February 2007 Vertical Deep Draft Pump Shaft and Coupling Failures Information Notice 2006-31, Inadequate Fault Interrupting Rating of Breakers2006-12-26026 December 2006 Inadequate Fault Interrupting Rating of Breakers Information Notice 2006-29, Potential Common Cause Failure of Motor-operated Valves as a Result of Stem Nut Wear2006-12-14014 December 2006 Potential Common Cause Failure of Motor-operated Valves as a Result of Stem Nut Wear Information Notice 2006-29, Potential Common Cause Failure of Motor-operated Valves As a Result of Stem Nut Wear2006-12-14014 December 2006 Potential Common Cause Failure of Motor-operated Valves As a Result of Stem Nut Wear Information Notice 2006-13, E-mail from M. Mclaughlin on NRC, Regarding NRC Information Notice 2006-13: Groundwater Contamination2006-07-13013 July 2006 E-mail from M. Mclaughlin on NRC, Regarding NRC Information Notice 2006-13: Groundwater Contamination 2020-09-03 The following query condition could not be considered due to this wiki's restrictions on query size or depth: <code> [[:Beaver Valley]] OR [[:Millstone]] OR [[:Hatch]] OR [[:Monticello]] OR [[:Calvert Cliffs]] OR [[:Dresden]] OR [[:Davis Besse]] OR [[:Peach Bottom]] OR [[:Browns Ferry]] OR [[:Salem]] OR [[:Oconee]] OR [[:Mcguire]] OR [[:Nine Mile Point]] OR [[:Palisades]] OR [[:Palo Verde]] OR [[:Perry]] OR [[:Indian Point]] OR [[:Fermi]] OR [[:Kewaunee]] OR [[:Catawba]] OR [[:Harris]] OR [[:Wolf Creek]] OR [[:Saint Lucie]] OR [[:Point Beach]] OR [[:Oyster Creek]] OR [[:Watts Bar]] OR [[:Hope Creek]] OR [[:Grand Gulf]] OR [[:Cooper]] OR [[:Sequoyah]] OR [[:Byron]] OR [[:Pilgrim]] OR [[:Arkansas Nuclear]] OR [[:Three Mile Island]] OR [[:Braidwood]] OR [[:Susquehanna]] OR [[:Summer]] OR [[:Prairie Island]] OR [[:Columbia]] OR [[:Seabrook]] OR [[:Brunswick]] OR [[:Surry]] OR [[:Limerick]] OR [[:North Anna]] OR [[:Turkey Point]] OR [[:River Bend]] OR [[:Vermont Yankee]] OR [[:Crystal River]] OR [[:Haddam Neck]] OR [[:Ginna]] OR [[:Diablo Canyon]] OR [[:Callaway]] OR [[:Vogtle]] OR [[:Waterford]] OR [[:Duane Arnold]] OR [[:Farley]] OR [[:Robinson]] OR [[:Clinton]] OR [[:South Texas]] OR [[:San Onofre]] OR [[:Cook]] OR [[:Comanche Peak]] OR [[:Yankee Rowe]] OR [[:Maine Yankee]] OR [[:Quad Cities]] OR [[:Humboldt Bay]] OR [[:La Crosse]] OR [[:Big Rock Point]] OR [[:Rancho Seco]] OR [[:Zion]] OR [[:Midland]] OR [[:Bellefonte]] OR [[:Fort Calhoun]] OR [[:FitzPatrick]] OR [[:McGuire]] OR [[:LaSalle]] OR [[:Fort Saint Vrain]] OR [[:Shoreham]] OR [[:Satsop]] OR [[:Trojan]] OR [[:Atlantic Nuclear Power Plant]] </code>.
[Table view]The following query condition could not be considered due to this wiki's restrictions on query size or depth: <code> [[:Beaver Valley]] OR [[:Millstone]] OR [[:Hatch]] OR [[:Monticello]] OR [[:Calvert Cliffs]] OR [[:Dresden]] OR [[:Davis Besse]] OR [[:Peach Bottom]] OR [[:Browns Ferry]] OR [[:Salem]] OR [[:Oconee]] OR [[:Mcguire]] OR [[:Nine Mile Point]] OR [[:Palisades]] OR [[:Palo Verde]] OR [[:Perry]] OR [[:Indian Point]] OR [[:Fermi]] OR [[:Kewaunee]] OR [[:Catawba]] OR [[:Harris]] OR [[:Wolf Creek]] OR [[:Saint Lucie]] OR [[:Point Beach]] OR [[:Oyster Creek]] OR [[:Watts Bar]] OR [[:Hope Creek]] OR [[:Grand Gulf]] OR [[:Cooper]] OR [[:Sequoyah]] OR [[:Byron]] OR [[:Pilgrim]] OR [[:Arkansas Nuclear]] OR [[:Three Mile Island]] OR [[:Braidwood]] OR [[:Susquehanna]] OR [[:Summer]] OR [[:Prairie Island]] OR [[:Columbia]] OR [[:Seabrook]] OR [[:Brunswick]] OR [[:Surry]] OR [[:Limerick]] OR [[:North Anna]] OR [[:Turkey Point]] OR [[:River Bend]] OR [[:Vermont Yankee]] OR [[:Crystal River]] OR [[:Haddam Neck]] OR [[:Ginna]] OR [[:Diablo Canyon]] OR [[:Callaway]] OR [[:Vogtle]] OR [[:Waterford]] OR [[:Duane Arnold]] OR [[:Farley]] OR [[:Robinson]] OR [[:Clinton]] OR [[:South Texas]] OR [[:San Onofre]] OR [[:Cook]] OR [[:Comanche Peak]] OR [[:Yankee Rowe]] OR [[:Maine Yankee]] OR [[:Quad Cities]] OR [[:Humboldt Bay]] OR [[:La Crosse]] OR [[:Big Rock Point]] OR [[:Rancho Seco]] OR [[:Zion]] OR [[:Midland]] OR [[:Bellefonte]] OR [[:Fort Calhoun]] OR [[:FitzPatrick]] OR [[:McGuire]] OR [[:LaSalle]] OR [[:Fort Saint Vrain]] OR [[:Shoreham]] OR [[:Satsop]] OR [[:Trojan]] OR [[:Atlantic Nuclear Power Plant]] </code>. |
K-,
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-0001 August 28, 1995 NRC INFORMATION NOTICE 95-35: DEGRADED ABILITY OF STEAM GENERATORS
TO REMOVE DECAY HEAT BY NATURAL CIRCULATION
Addressees
All holders of operating licenses or construction permits for pressurized- water reactors (PWRs).
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to conditions that may degrade the ability of the
steam generators (SGs) to remove decay heat by natural circulation of the
reactor coolant in Mode 5, cold shutdown. It is expected that recipients will
review the information for applicability to their facilities and consider
actions, as appropriate, to avoid similar problems. However, suggestions
contained in this information notice are not NRC requirements; therefore, no
specific action or written response is required.
DescriDtion of Circumstances
During a refueling outage in September 1994 on Vogtle Unit 1, the licensee
found that the ability to remove decay heat by natural circulation through the
steam generators was degraded. On September 11, 1994, Vogtle Unit 1 entered a
refueling outage. On September 16, 1994, the unit was in Mode 5 with the
reactor coolant system (RCS) being drained down. The reactor coolant pumps
were tagged out of service and residual heat removal (RHR) system Train B was
in service removing decay heat. RHR Train A was operating intermittently but
had been administratively removed from service to adjust the suction valve
limit switch and allow stroking of local valves. During this period, the SGs
were relied upon to meet the technical specification requirement for a second
source of decay heat removal. While taking credit for the steam generators as
a heat sink, the RCS was vented to the containment atmosphere when the
pressurizer code safety valves were removed and a conoseal on the reactor
vessel head was disassembled. The licensee later determined that the heat
removal capability by natural circulation through the SGs was degraded. When
the RCS was vented, the inability to pressurize the RCS reduced the natural
circulation cooling capacity.
On February 13, 1995, the licensee for Turkey Point Units 3 and 4 reported
that during previous refueling outages it had relied upon the SGs for decay
heat removal when the SGs may not have been able to perform that function.
Specifically, the licensee relied upon the SGs as one of the means of decay
heat removal while testing one of the RHR loops with the RCS vented. The
PD& If E IUo4K. 7S-o3s- 5¢g
9508220031 QII
<L
-' IN 95-35 August 28, 1995 licensee later concluded that the RCS cannot support subcooled natural
circulation decay heat removal through the SGs while the RCS is vented.
Discussion
These two examples illustrate plant conditions that were not adequate to fully
support natural circulation through the SGs as a method of decay heat removal
during operation in Mode 5 with loops filled and the reactor coolant pumps out
of service.
Technical specifications generally require two methods of decay heat removal
in Mode 5 with loops filled. When this is the case, they generally go on to
indicate that this requirement can be satisfied by two loops of RHR or one
loop of RHR and a minimum water level in the SGs. Decay heat can be removed
either through the RHR system or through the SGs by natural circulation after
the reactor coolant pumps are secured. The heat removal mechanism with
residual heat removal is through forced circulation through the RHR heat
exchanger. Heat removal with natural circulation of reactor coolant through
the SGs occurs because of the differential pressure created between the heated
water in the reactor core and the cooler water in the SG tubes. This
differential pressure is created through temperature differences that in turn
create fluid density differences between these two locations.
When the RCS is being depressurized and cooled down, the reactor coolant pumps
are stopped, the RCS is depressurized and vented, and level is decreased in
preparation for Mode 6 (refueling) entry. In Mode 6, both RHR trains must be
operable. During the transition from Mode 5, with no reactor coolant pumps
running, to Mode 6, plant conditions may exist that are not adequate to
support natural circulation. The second train of RHR may need to be operable
before proceeding with plant cooldown and depressurization to provide a second
method for RCS cooling.
During natural circulation, the SG secondary side water boils and steams off
through the atmospheric relief valves or other openings that may exist during
shutdown conditions. The minimum temperature at which boiling will begin in
the SG is 100lC [2120 F]. A minimum temperature differential of 28- C
[500 F] between the RCS and the SG secondary water is routinely used for
evaluating conditions that would ensure sufficient natural circulation flow to
prevent boiling in the core. The heat transfer rate across the steam
generator tubes is less for lower RCS-to-SG secondary temperature
differentials but still may be adequate to promote sufficient natural
circulation and prevent core boiling. Adding the differential temperature of
280 C [50 F] to 1000 C [2120 F] results in a minimum RCS temperature of
128- C [262 F] to maintain sufficient natural circulation flow. The lowest
pressure point in the RCS, at the top of the SG tubes, should therefore be
maintained above the saturation pressure for 128' C [262 F]. If the RCS
pressure at the top of the SG tubes is allowed to fall below the primary fluid
saturation temperature, flashing and steam voiding may occur, interrupting or
degrading the natural circulation flow path. Additionally, when system
pressure is dropped with elevated water temperatures, gases may come out of
solution.
I i I
<-' IN 95-35 August 28, 1995 When relying on the ability of the SGs to remove decay heat by natural
circulation of reactor coolant in Mode 5, the following factors are worthy of
consideration: (1)the ability to pressurize and control pressure in the RCS,
(2)secondary side water level in the SGs relied upon for decay heat removal,
(3)availability of a supply of feedwater, and (4) availability of an
auxiliary feedwater pump capable of injecting into the relied-upon SGs.
Consideration should also be given to avoiding the potential for
pressurization of the SG secondary side. It is also important to note that
during the decay heat removal scenario for the natural circulation process, a
mode change (Mode 5 to Mode 4) could occur due to heat up of the RCS.
This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact
the technical contact listed below or the appropriate Office of Nuclear
Reactor Regulation (NRR) project manager.
Crutch el Director
D sM.
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical contact: Brian R. Bonser, RII
(706) 554-9901 Attachment:
List of Recently Issued NRC Information Notices
47/TW%4~ fSjACkeY
"" ttachment
IN 95-35 August 28, 1995 LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
Information Date of
Notice No. Subject Issuance Issued to
95-34 Air Actuator and Supply 08/25/95 All holders of OLs or CPs
Air Regulator Problems in for nuclear power reactors.
Copes-Vulcan Pressurizer
Power-Operated Relief Valves
93-83, Potential Loss of Spent 08/24/95 All holders of OLs or CPs
Supp. 1 Fuel Pool Cooling After a for nuclear power reactors.
Loss-of-Coolant Accident
or a Loss of Offsite Power
95-33 Switchgear Fire and 08/23/95 All holders of OLs or CPs
Partial Loss of Offsite for nuclear power reactors.
Power at Waterford
Generating Station, Unit 3
95-10, Potential for Loss of 08/11/95 All holders of OLs or CPs
Supp. 2 Automatic Engineered for nuclear power reactors.
Safety Features Actuation
95-32 Thermo-Lag 330-1 Flame 08/10/95 All holders of OLs or CPs
Spread Test Results for nuclear power reactors.
95-31 Motor-Operated Valve 08/09/95 All holders of OLs or CPs
Failure Caused by Stem for nuclear power reactors.
Protector Pipe Inter- ference
95-30 Susceptibility of Low- 08/03/95 All holders of OLs or CPs
Pressure Coolant Injection for nuclear power reactors.
and Core Spray Injection
Valves to Pressure Locking
94-66, Overspeed of Turbine- 06/16/95 All holders of OLs or CPs
Supp. 1 Driven Pumps Caused by for nuclear power reactors.
Binding in Stems of
Governor Valves
OL - Operating License
CP - Construction Permit
IN 95-35 August 28, 1995 When relying on the ability of the SGs to remove decay heat by natural
circulation of reactor coolant in Mode 5, the following factors are worthy of
consideration: (1) the ability to pressurize and control pressure in the RCS,
(2) secondary side water level in the SGs relied upon for decay heat removal,
(3) availability of a supply of feedwater, and (4) availability of an
auxiliary feedwater pump capable of injecting into the relied-upon SGs.
Consideration should also be given to avoiding the potential for
pressurization of the SG secondary side. It is also important to note that
during the decay heat removal scenario for the natural circulation process, a
mode change (Mode 5 to Mode 4) could occur due to heat up of the RCS.
This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact
the technical contact listed below or the appropriate Office of Nuclear
Reactor Regulation (NRR) project manager.
[Original signed by]
Dennis M. Crutchfield, Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical contact: Brian R. Bonser, RII
(706) 554-9901 Attachment:
List of Recently Issued NRC Information Notices
DOCUMENT NAME: 95-35.IN
To receive a copy of this document, Indicate In the box: AC = Copy without attachment/enclosure 'E'
- Copy with attachmentlenclosura N' = No copy
-
OFFICE OECBIDOP5 E SC/OECB:DOPS NI
NADM:PUB I N SRXB:DSSA IE C/OTSB:DOPS lE
NAME . JTappert* EGoodwin* Tech Ed* RJones* ICGrimes*
DATE 04/20/95 04/27/95 04/28/95 04/25/95 05/16/95
- . _ . . __ _ _. . . ........... _ ._ I
INIIZ'2 D/QD J /&
l1io-7I\_11v
OFFICE _ I_PECB:DRPM
IE IC/PECB:DRPM
4 NAME RKiessel* 1AChaffee* IDCPV((I4fif elI d
DATE 108/10/95 108/17/95 108/1/95 OFFICIAL RECORD COPY
IN 95-35 August 28, 1995 When relying on the ability of the SGs to remove decay heat by natural
circulation of reactor coolant in Mode 5, the following factors are worthy of
consideration: (1)the ability to pressurize and control pressure in the RCS,
(2)secondary side water level in the SGs relied upon for decay heat removal,
(3)availability of a supply of feedwater, and (4) availability of an
auxiliary feedwater pump capable of injecting into the relied-upon SGs.
Consideration should also be given to avoiding the potential for
pressurization of the SG secondary side. It is also important to note that
during the decay heat removal scenario for the natural circulation process, a
mode change (Mode 5 to Mode 4) could occur due to heat up of the RCS.
This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact
the technical contact listed below or the appropriate Office of Nuclear
Reactor Regulation (NRR) project manager.
[Original signed by]
Dennis M. Crutchfield, Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical contact: Brian R. Bonser, RII
(706) 554-9901 Attachment:
List of Recently Issued NRC Information Notices
DOCUMENT NAME: 95-35.IN
T. C ... e fW
ta.b
n .W--f, 6...41.t. o
Inth. hnx* C' = Cnny without attachmentlenclosure E' = COpy with attachmentenclosure 'N" = No copy
OFFICE OECB:DOPS IE N ADM:PUB N SRXB:DSSA IE C/OTSB:DOPS I I
NAME JTappert* EGoodwin* Tech Ed* RJones* CGrimes* I
DATE 04/20/95 04/27/95 04/28/95 04/25/95 05/16/95 OFFICE PECB:DRPM IE IC/PECB:DRPM INID/DAEJA, / SN1_.
lDATE
0T8 10 9 NAME D ARKiessel*
08/10/95 _
_._ AChaffee* _
08/17/95
_ _ _ _ IDC_
.... _
. 10/IV95N nn
_ _
_feld_
_ _ _ _ _ _,._,
,;w 1/
UFFILIAL MUCMV LUPY
IN 95-XX
August xx, 1995 When relying on the ability of the SGs to remove decay heat by natural
circulation of reactor coolant in Mode 5, the following factors are worthy of
consideration: (1) the ability to pressurize and control pressure in the RCS,
(2) secondary side water level in the SGs relied upon for decay heat removal,
(3) availability of a supply of feedwater, and (4) availability of an
auxiliary feedwater pump capable of injecting into the relied-upon SGs.
Consideration should also be given to avoiding the potential for
pressurization of the SG secondary side. It is also important to note that
during the decay heat removal scenario for the natural circulation process, a
mode change (Mode 5 to Mode 4) could occur due to heat up of the RCS.
This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact
the technical contact listed below or the appropriate Office of Nuclear
Reactor Regulation (NRR) project manager.
Dennis M. Crutchfield, Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical contact: Brian R. Bonser, R11
(706) 554-9901 Attachment: List of Recently Issued NRC Information Notices
DOCUMENT NAME: G:\JRT\NATCIRC.BRB
F. cnoov with attchmntl~enclosure 'N = No copy
I- - *- c_ _ . - - E -_
SY I
a _W
OFFICE OECB:DOPS lE SC/OECB:DOPS IN ADM:PUB IN SRX A C/OTSB:DOPS IE
NAME JTappert*_JEGoodwin* Tech Ed* RJones* CGrimes*
DATE 04/20/95 104/27/95 04/28/95 104/25/ 05/16/95 OFFTCEF IPECB:DRPM
H ____ ___ .I . - - - - - . IEIUM/ILJUKFM INIU/UKrm
.......-. w-7 o*,
I R l r %7 NAME RKiessel* 1AC_ ffi ;e fltbj_ DCrutchfield-It ti
DATE 10810L/95 l /95 a.--- -- n n~ff rnnu
UtlIUlAL KLLKU UOUr
IN 95-XX
May xx, 1995 128 0 C
therefore be maintained above the saturation pressure for is
(2620 F). If the RCS pressure at the top of the SG tubes
allowed to fall below the primary fluid saturation temperature, flashing and steam voiding may occur, interrupting or degrading
the natural circulation flow path. Additionally, when system may
pressure is dropped with elevated water temperatures, gases
come out of solution.
When relying on the ability of the SGs to remove decay heat by
natural circulation of reactor coolant in Mode 5, the following
factors are worthy of consideration: (1) the ability to
pressurize and control pressure in the RCS, (2) secondary side (3)
water level in the SGs relied upon for decay heat removal, of an
availability of a supply of feedwater, and (4) availability
auxiliary feedwater pump capable of injecting into the relied- upon SGs. Consideration should also be given to avoiding the
potential for pressurization of the SG secondary side. It is
also important to note that during the decay heat removal (Mode
scenario for the natural circulation process a mode change
5 to Mode 4) could occur due to heat up of the RCS.
This information notice requires no specific action or written in
response. If you have any questions about the information or
this notice, please contact the technical contact listed below
the appropriate Office of Nuclear Reactor Regulation (NRR)
project manager.
Brian K. Grimes, Director
Division of Project Support
Office of Nuclear Reactor
Regulation
Technical contact: Brian R. Bonser, RII
(706) 554-9901 Attachment: List of Recently Issued NRC Information Notices
DOCUMENT NAME: G:\JRT\NATCIRC.BRB
'E' - Copy wIth attachmentlenclosure
To recelv a copy of this document, Indicato In the box: *C - Copy without attachment/enoloue
'N - No copy I
on noon I
OFFICE OECB:DOPS SC/OECB:DOPS ADM:PUB l SRXB:DSSA l C/uiT:uuwr
NAME JTappert * EGoodwin * Tech Ed * RJones * CGr mes*
DATE 4/20/95 4/21/95 4/28/95 4/25/95 5/16/95 OFFICE j BO C/OECB:DOPS D/DOPS
NAME Wessel AChaffee BGrimes
DATE A t J
I/95 1/01 I /95 =============
OFFICIAL RECORD COPY
K)
IN 95-XX
May xx, 1995 the RCS pressure at the top of the SG tubes is allowed to fall
below the primary fluid saturation temperature, flashing and
steam voiding may occur, interrupting or degrading the natural
circulation flow path . Additionally, when system pressure is
dropped with elevated water temperatures, gasses may come out of
solution.
When relying on the ability of the SGs to remove decay heat by
natural circulation of reactor coolant in Mode 5, the folowing
factors are worthy of consideration: (1) the ability to
pressurize and control pressure in the RCS, (2) secondary side
water level, at or above the top of the tubes in the SGs relied
upon for decay heat removal, (3) availability of a supply of
feedwater, and (4) availability of an auxiliary feedwater pump
capable of taking suction from the feedwater source and injecting
into the relied-upon SGs. Consideration should also be given to
avoiding the potential for pressurization of the SG secondary
side. It is also important to note that a mode change (Mode 5 to
Mode 4) could occur during the decay heat removal scenario for
the natural circulation process.
Brian K. Grimes, Director
Division of Project Support
Office of Nuclear Reactor
Regulation
Technical contact: Brian R. Bonser, RII
(706) 554-9901 Attachment: List of Recently Issued NRC Information Notices
DOCUMENT NAME: G:\JRT\NATCIRC.BRB 0i
Cu = Copy without attachment/enclosure 'E' Copy attachment/enclosure
To receive a copy of this document, Indicate In the box:
'Nt = No copy __
OFFICE OECB:DOPS SC/OECB:DOPS ADM:PUB SPB DSSA C/OTSB:DOP
NAME JTappert * EGoodwin * Tech Ed * RJones * CGrimes
DATE 4/20/95 4/27/95 4/28/95 4/25/95 OFFICE OECB:DOPS C/OECB:DOPS L D/DOPSIL
NAME RKiessel AChaffee BGrimes
DATE I/I / /95 / /95 OFFICIAL RECORD COPY
K>
IN 95-XX
May xx, 1995 maintained above the saturation pressure for 2620 F (1280 C].
Flashing and steam voiding may occur interrupting or degrading
the natural circulation flow path if the RCS pressure at the top
of the SG tubes is allowed to fall below the primary fluid
saturation temperature. An added complication of gasses coming
out of solution can occur when system pressure is dropped with
elevated water temperatures.
When relying on the decay heat removal capability of the SGs via
natural circulation of reactor coolant in mode 5 certain criteria
should be considered. These criteria include: 1) the ability to
pressurize and control pressure in the RCS; 2) the SGs relied
upon for decay heat removal have secondary side water level at or
above the top of the tubes; 3) a supply of feedwater is
available; and 4) an auxiliary feedwater pump is available
capable of taking suction from the feedwater source and injecting
into the relied upon SGs. Consideration should also be given to
avoiding the potential for pressurization of the SG secondary
side. It is also important to note that a mode change (mode 5 to
mode 4) will occur during the decay heat removal scenario for the
natural circulation process.
Brian K. Grimes, Director
Division of Project Support
Office of Nuclear Reactor
Regulation
Technical contact: Brian R. Bonser, RII
(706) 554-9901 Attachment: List of Recently Issued NRC Information Notices
DOCUMENT NAME: G:\JRT\NATCIRC.BRB
To receive a copy of this document, Indicate In the box: 'C' - Copy without attachment/enclosure E' - Copy with atachment/enclosure
a No copy
OFFICE IOECB:DOPS SC/OECB:DOPS lAMPUB l [SPLB:DSSA J I
NAME JTappert Voodwin Tech Ed __ones_____
DATE 1q/ Q $ 1 /a) 2I i 1 / f__ I/
OFFICE OECB:DOPS C/OECB:DOPS D/DOPSL
NAME RKiessel IAChaffee BGrimes l
DATE // / /95 / /95 OFFICIAL RECORD COPY
IN 95-XX
May xx, 1995 maintained above the saturation pressure for 128 0 C [2620 F]. If
the RCS pressure at the top of the SG tubes is allowed to fall
below the primary fluid saturation temperature, flashing and
steam voiding may occur interrupting or degrading the natural
circulation flow path . Additionally, when system pressure is
dropped with elevated water temperatures gasses may come out of
solution.
When relying on the decay heat removal capability of the SGs via
natural circulation of reactor coolant in mode 5 the folowing
factors are worthy of consideration. These include: 1) the
ability to pressurize and control pressure in the RCS; 2) the SGs
relied upon for decay heat removal have secondary side water
level at or above the top of the tubes; 3) availability of a
supply of feedwater; and 4) availability of an auxiliary
feedwater pump capable of taking suction from the feedwater
source and injecting into the relied upon SGs. Consideration
should also be given to avoiding the potential for pressurization
of the SG secondary side. It is also important to note that a
mode change (mode 5 to mode 4) could occur during the decay heat
removal scenario for the natural circulation process.
Brian K. Grimes, Director
Division of Project Support
Office of Nuclear Reactor
Regulation
Technical contact: Brian R. Bonser, RII
(706) 554-9901 Attachment: List of Recently Issued NRC Information Notices
DOCUMENT NAME: G:\JRT\NATCIRC.BRB
To raceive a copy of this document, Indicate In Xt box: 'C' - Copy wihout attachmentlenclosure T
E Copy with attachmentlenclosure ¶N'
OFFICE
NAME
DATE
OECB:DOPS
JTappert
/ /
IEGoodwin
1/
SC/OECB:DOPS
/
AMPUB
Tech Ed
/
l
SPLB:DSSA
RJones
//
l
I____
OFFICE OECB:DOPS C/OECB:DOPS _
__D/DOPS
NAME RKiessel AChaffee BGrimes
DATE I / / /95 / /95 OFFICIAL RECORD COPY
|
---|
|
list | - Information Notice 1995-01, DOT Safety Advisory: High Pressure Aluminum Seamless and Aluminum Composite Hoop-Wrapped Cylinders (4 January 1995, Topic: Brachytherapy)
- Information Notice 1995-02, Problems With General Electric CR2940 Contact Blocks In Medium-Voltage Circuit Breakers (17 January 1995)
- Information Notice 1995-02, Problems With General Electric Cr2940 Contact Blocks In Medium-Voltage Circuit Breakers (17 January 1995)
- Information Notice 1995-02, Problems with General Electric CR2940 Contact Blocks in Medium-Voltage Circuit Breakers (17 January 1995)
- Information Notice 1995-03, Loss of Reactor Coolant Inventory and Potential Loss of Emergency Mitigation Functions While in a Shutdown Condition (18 January 1995, Topic: Packing leak, Water hammer)
- Information Notice 1995-04, Excessive Cooldown and Depressurization of the Reactor Coolant System Following Loss of Offsite Power (11 October 1996, Topic: Safe Shutdown, Shutdown Margin, Probabilistic Risk Assessment, Troxler Moisture Density Gauge)
- Information Notice 1995-05, Undervoltage Protection Relay Settings Out of Tolerance Due to Test Equipment Harmonics (20 January 1985)
- Information Notice 1995-06, Potential Blockage of Safety-Related Strainers by Material Brought Inside Containment (25 January 1995, Topic: Foreign Material Exclusion)
- Information Notice 1995-07, Radiopharmaceutical Vial Breakage During Preparation (27 January 1995)
- Information Notice 1995-08, Inaccurate Data Obtained with Clamp-On Ultrasonic Flow Measurement Instruments (30 January 1995)
- Information Notice 1995-08, Inaccurate Data Obtained With Clamp-On Ultrasonic Flow Measurement Instruments (30 January 1995)
- Information Notice 1995-09, Use of Inappropriate Guidelines and Criteria for Nuclear Piping and Pipe Support Evaluation and Design (31 January 1995, Topic: Operability Determination)
- Information Notice 1995-10, Potential for Loss of Automatic Engineered Safety Features Actuation (3 February 1995, Topic: High Energy Line Break)
- Information Notice 1995-11, Failure of Condensate Piping Because of Erosion/Corrosion at Flow-Straightening Device (24 February 1995, Topic: Feedwater Heater)
- Information Notice 1995-12, Potentially Nonconforming Fasteners Supplied by A&G Engineering II, Inc (21 February 1995)
- Information Notice 1995-13, Potential for Data Collection Equipment to Affect Protection System Performance (24 February 1995)
- Information Notice 1995-14, Susceptibility of Containment Sump Recirculation Gate Valves to Pressure Locking (28 February 1995)
- Information Notice 1995-15, Inadequate Logic Testing of Safety-Related Circuits (7 March 1995)
- Information Notice 1995-16, Vibration Caused by Increased Recirculation Flow in a Boiling Water Reactor (9 March 1995)
- Information Notice 1995-17, Reactor Vessel Top Guide and Core Plate Cracking (10 March 1995, Topic: Safe Shutdown, Intergranular Stress Corrosion Cracking, Stress corrosion cracking)
- Information Notice 1995-18, Potential Pressure-Locking of Safety-Related Power-Operated Gate Valves (15 March 1995)
- Information Notice 1995-19, Failure of Reactor Trip Breaker to Open Because of Cutoff Switch Material Lodged in the Trip Latch Mechanism (22 March 1995)
- Information Notice 1995-20, Failures in Rosemount Pressure Transmitters Due to Hydrogen Permeation Into Sensor Cell (22 March 1995)
- Information Notice 1995-21, Unexpected Degradation of Lead Storage Batteries (20 April 1995)
- Information Notice 1995-22, Hardened or Contaminated Lubricant Cause Metal-Clad Circuit Breaker Failures (21 April 1995, Topic: Hardened grease)
- Information Notice 1995-23, Control Room Staffing Below Minimum Regulatory Requirements (24 April 1995)
- Information Notice 1995-24, Summary of Licensed Operator Requalification Inspection Program Findings (25 April 1995, Topic: Job Performance Measure, License Renewal)
- Information Notice 1995-25, Valve Failure During Patient Treatment with Gamma Stereotactic Radiosurgery Unit (11 May 1995, Topic: Overdose)
- Information Notice 1995-26, Defect in Safety-Related Pump Parts Due to Inadequate Treatment (31 May 1995, Topic: Intergranular Stress Corrosion Cracking, Stress corrosion cracking)
- Information Notice 1995-27, NRC Review of Nuclear Energy Institute, Thermo-Lag 330-1 Combustibility Evaluation Methodology Plant Screening Guide. (31 May 1995, Topic: Safe Shutdown, Fire Barrier, Exemption Request, Fire Protection Program)
- Information Notice 1995-28, Emplacement of Support Pads for Spent Fuel Dry Storage Installations at Reactor Sites (5 June 1995, Topic: Safe Shutdown, Tornado Missile, Safe Shutdown Earthquake, Earthquake)
- Information Notice 1995-29, Oversight of Design and Fabrication Activities for Metal Components Used in Spent Fuel Dry Storage Systems (7 June 1995, Topic: Nondestructive Examination)
- Information Notice 1995-30, Susceptibility of Low-Pressure Coolant Injection Valves to Pressure Locking (3 August 1995, Topic: Hydrostatic, Power-Operated Valves, Overspeed trip)
- Information Notice 1995-31, Motor-Operated Valve Failure Caused by Stem Protector Pipe Interference (9 August 1995, Topic: Overspeed)
- Information Notice 1995-32, Thermo-Lag 330-1 Flame Spread Test Results (10 August 1995, Topic: Fire Barrier, Overspeed trip)
- Information Notice 1995-33, Switchgear Fire and Partial Loss of Offsite Power at Waterford Generating Station, Unit 3 (23 August 1995, Topic: Overspeed)
- Information Notice 1995-34, Air Actuator and Supply Air Regulator Problems in Copes-Vulcan Pressurizer Power-Operated Relief Valves (25 August 1995, Topic: Overspeed)
- Information Notice 1995-35, Degraded Ability of Steam Generators to Remove Decay Heat by Natural Circulation (28 August 1995, Topic: Overspeed)
- Information Notice 1995-36, Potential Problems with Post-Fire Emergency Lighting (29 August 1995, Topic: Safe Shutdown, Emergency Lighting, Exemption Request, Overspeed, Manual Operator Action)
- Information Notice 1995-37, Inadequate Offsite Power System Voltages During Design-Basis Events (7 September 1995)
- Information Notice 1995-38, Degradation of Boraflex Neutron Absorber in Spent Fuel Storage Racks (8 September 1995)
- Information Notice 1995-39, Brachytherapy Incidents Involving Treatment Planning Errors (19 September 1995, Topic: Brachytherapy, Underdose)
- Information Notice 1995-40, Supplemental Information to Generic Letter 95-03, Circumferential Cracking of Steam Generator Tubes. (20 September 1995, Topic: Hydrostatic, Nondestructive Examination, Brachytherapy)
- Information Notice 1995-41, Degradation of Ventilation System Charcoal Resulting from Chemical Cleaning of Steam Generators (22 September 1995, Topic: Brachytherapy)
- Information Notice 1995-42, Commission Decision on Resolution of Generic Issue 23, Reactor Coolant Pump Seal Failure. (22 September 1995, Topic: Brachytherapy)
- Information Notice 1995-43, Failure of Bolt-Locking Device on Reactor Coolant Pump Turning Vane (28 September 1995, Topic: Brachytherapy)
- Information Notice 1995-44, Ensuring Compatible Use of Drive Cables Incorporating Industrial Nuclear Company Ball-Type Male Connectors (26 September 1995, Topic: Brachytherapy)
- Information Notice 1995-45, American Power Service Falsification of American Society for Nondestructive Testing Certificates (4 October 1995, Topic: Commercial Grade, Brachytherapy)
- Information Notice 1995-46, Unplanned, Undetected Release of Radioactivity from the Exhaust Ventilation System of a Boiling Water Reactor (6 October 1995, Topic: Brachytherapy)
- Information Notice 1995-47, Unexpected Opening of a Safety/Relief Valve & Complications Involving Suppression Pool Cooling Strainer Blockage (30 November 1995)
... further results |
---|