|
|
Line 14: |
Line 14: |
| | page count = 5 | | | page count = 5 |
| }} | | }} |
| {{#Wiki_filter:v)UNITED STATES NUCLEAR REGULATORY | | {{#Wiki_filter:v) |
| | UNITED STATES |
|
| |
|
| COMMISSION | | NUCLEAR REGULATORY COMMISSION |
|
| |
|
| OFFICE OF NUCLEAR REACTOR REGULATION_ | | OFFICE OF NUCLEAR REACTOR REGULATION_ |
| WASHINGTON, D.C. 20555 6 September | | WASHINGTON, D.C. 20555 6 September 3, 1992 - |
| | 7 ' |
| | NRC INFORMATION NOTICE 92-65: SAFETY SYSTEM PROBLEM flCAUSED BY MODIFICATIONS |
|
| |
|
| 3, 1992 -7 'NRC INFORMATION
| | THAT WERE NOT ADEQUATE-V REVIEWED AND TESTED |
| | |
| NOTICE 92-65: SAFETY SYSTEM PROBLEM flCAUSED BY MODIFICATIONS
| |
| | |
| THAT WERE NOT ADEQUATE-V | |
| | |
| REVIEWED AND TESTED | |
|
| |
|
| ==Addressees== | | ==Addressees== |
| All holders of operating | | All holders of operating licenses or construction permits for nuclear power |
| | |
| licenses or construction | |
|
| |
|
| permits for nuclear power reactors.
| | reactors. |
|
| |
|
| ==Purpose== | | ==Purpose== |
| The U.S. Nuclear Regulatory | | The U.S. Nuclear Regulatory Commission (NRC) is issuing this information |
| | |
| Commission (NRC) is issuing this information | |
| | |
| notice to alert addressees
| |
| | |
| to problems caused by inadequate
| |
| | |
| review and testing of safety system modifications.
| |
| | |
| 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.Description
| |
| | |
| of Circumstances
| |
| | |
| The following
| |
| | |
| describes
| |
| | |
| two examples of safety system design errors that went undetected
| |
| | |
| since construction, because design changes were not thoroughly
| |
| | |
| reviewed and tested.On October 10, 1991, during post overhaul testing, personnel
| |
| | |
| at Arkansas Nuclear One, Unit 1, observed that one of the high-pressure
| |
| | |
| safety injection (HPSI) pumps was losing its lubricating
| |
|
| |
|
| oil at a rate of more than 15 gallons per hour as a result of oil spraying from the bearings.
| | notice to alert addressees to problems caused by inadequate review and testing |
|
| |
|
| The licensee found that the oil would always leak at this 'rate during emergency
| | of safety system modifications. It is expected that recipients will review |
|
| |
|
| operation
| | the information for applicability to their facilities'and consider actions, as |
|
| |
|
| because of excessive
| | appropriate, to avoid similar problems. However, suggestions contained in |
|
| |
|
| oil pressure caused by the simultaneous
| | this information notice are not NRC requirements; therefore, no specific |
|
| |
|
| operation
| | action or written response is required. |
|
| |
|
| of two oil -pumps that-served | | ==Description of Circumstances== |
| | The following describes two examples of safety system design errors that went |
|
| |
|
| the HPSI pump. This condition
| | undetected since construction, because design changes were not thoroughly |
|
| |
|
| had-existed
| | reviewed and tested. |
|
| |
|
| since the plant began operation.
| | On October 10, 1991, during post overhaul testing, personnel at Arkansas |
|
| |
|
| The bearings for each of the HPSI pumps are supplied with lubricating
| | Nuclear One, Unit 1, observed that one of the high-pressure safety injection |
|
| |
|
| oil by two oil pumps, one attached directly to the HPSI pump itself and the other a separate electric backup pump. Originally | | (HPSI) pumps was losing its lubricating oil at a rate of more than 15 gallons |
|
| |
|
| the electric oil pumps were intended to be used during start up of a HPSI pump or to replace a malfunctioning | | per hour as a result of oil spraying from the bearings. The licensee found |
|
| |
|
| attached oil pump. The electric oil pumps could be started manually and would start automatically
| | that the oil would always leak at this 'rate during emergency operation because |
|
| |
|
| when the oil pressure decreased
| | of excessive oil pressure caused by the simultaneous operation of two oil - |
| | pumps that-served the HPSI pump. This condition had-existed since the plant |
|
| |
|
| below a certain point. The licensee continues
| | began operation. |
|
| |
|
| to use this method of control when the HPSI pumps are used for normal reactor water makeup. However, during construction, the licensee decided that the HPSI pumps would be more reliable if the electric lubricating
| | The bearings for each of the HPSI pumps are supplied with lubricating oil by |
|
| |
|
| oil pumps ran continuously | | two oil pumps, one attached directly to the HPSI pump itself and the other a |
|
| |
|
| during emergency operation.
| | separate electric backup pump. Originally the electric oil pumps were |
|
| |
|
| Consequently, the licensee modified the emergency
| | intended to be used during start up of a HPSI pump or to replace a |
|
| |
|
| controls to keep 9208280105
| | malfunctioning attached oil pump. The electric oil pumps could be started |
| ?DR J4E fltc6-S-O c a0 7 0-3 W I
| |
|
| |
|
| IN 92-65 September
| | manually and would start automatically when the oil pressure decreased below a |
|
| |
|
| 3, 1992 the electric oil pumps-operating
| | certain point. The licensee continues to use this method of control when the |
|
| |
|
| whenever an emergency
| | HPSI pumps are used for normal reactor water makeup. However, during |
|
| |
|
| safety features actuation
| | construction, the licensee decided that the HPSI pumps would be more reliable |
|
| |
|
| system (ESFAS) signal was present. Anticipating
| | if the electric lubricating oil pumps ran continuously during emergency |
|
| |
|
| that the simultaneous
| | operation. Consequently, the licensee modified the emergency controls to keep |
|
| |
|
| operation
| | 9208280105 ?DR J4E fltc6-S-O |
|
| |
|
| of both oil pumps could cause excessive
| | c a0 7 0-3 WI |
|
| |
|
| oil pressure, the licensee added an oil- pressure relief valve to the oil system. However, the relief valve settings were not appropriately
| | IN 92-65 September 3, 1992 the electric oil pumps-operating whenever an emergency safety features |
|
| |
|
| selected to prevent oil spraying from the bearings.In September
| | actuation system (ESFAS) signal was present. Anticipating that the |
|
| |
|
| 1991, the Gulf States Utilities
| | simultaneous operation of both oil pumps could cause excessive oil pressure, the licensee added an oil- pressure relief valve to the oil system. However, the relief valve settings were not appropriately selected to prevent oil |
|
| |
|
| Company, licensee for the River Bend Station, discovered
| | spraying from the bearings. |
|
| |
|
| that the outlet valves for the hydrogen mixing system would immediately
| | In September 1991, the Gulf States Utilities Company, licensee for the River |
|
| |
|
| close if an operator attempted
| | Bend Station, discovered that the outlet valves for the hydrogen mixing system |
|
| |
|
| to start up the system by opening these valves when a loss-of-coolant | | would immediately close if an operator attempted to start up the system by |
|
| |
|
| accident (LOCA) signal was present. An interlock | | opening these valves when a loss-of-coolant accident (LOCA) signal was |
|
| |
|
| prevented | | present. An interlock prevented the mixing system fans from operating with |
|
| |
|
| the mixing system fans from operating | | the outlet valves closed. Consequently, the hydrogen mixing system would have |
|
| |
|
| with the outlet valves closed. Consequently, the hydrogen mixing system would have been inoperable
| | been inoperable if a LOCA signal were present. This condition had existed |
|
| |
|
| if a LOCA signal were present. This condition
| | since the plant was constructed. |
| | |
| had existed since the plant was constructed.
| |
|
| |
|
| The River Bend Station is a boiling water reactor with a Mark III containment | | The River Bend Station is a boiling water reactor with a Mark III containment |
|
| |
|
| structure. | | structure. This containment structure consists of two chambers, a large outer |
|
| |
|
| This containment
| | primary containment and a drywell which is inside the primary containment and |
|
| |
|
| structure
| | surrounds the reactor vessel. This system suppresses the steam pressure |
|
| |
|
| consists of two chambers, a large outer primary containment
| | released during a LOCA by directing the steam through the suppression pool |
|
| |
|
| and a drywell which is inside the primary containment
| | water into the primary containment. After the initial pressure suppression is |
|
| |
|
| and surrounds
| | complete following a LOCA, hydrogen created by the zirconium-water reaction |
|
| |
|
| the reactor vessel. This system suppresses | | would be mainly concentrated in the drywell. The hydrogen mixing system is |
|
| |
|
| the steam pressure released during a LOCA by directing | | provided to reduce the concentration of the hydrogen in the drywell by moving |
|
| |
|
| the steam through the suppression | | it into the primary containment where it is diluted and reduced in |
|
| |
|
| pool water into the primary containment.
| | concentration by the hydrogen recombiners. |
|
| |
|
| After the initial pressure suppression
| | The redundant hydrogen mixing systems each have two lines penetrating the |
|
| |
|
| is complete following
| | drywell; an outlet line having a recirculating fan to draw suction from the |
|
| |
|
| a LOCA, hydrogen created by the zirconium-water
| | drywell and an inlet line that allows diluted air to reenter the drywell. |
|
| |
|
| reaction would be mainly concentrated
| | Each of these lines has two isolation valves which are normally closed during |
|
| |
|
| in the drywell. The hydrogen mixing system is provided to reduce the concentration
| | plant operation. In 1983, during construction, the licensee added a LOCA |
|
| |
|
| of the hydrogen in the drywell by moving it into the primary containment
| | interlock to the hydrogen mixing system that would automatically close all |
|
| |
|
| where it is diluted and reduced in concentration
| | eight of the mixing system valves upon receiving a LOCA signal. In 1984, the |
|
| |
|
| by the hydrogen recombiners.
| | licensee revised the control logic for the mixing system valves to |
|
| |
|
| The redundant
| | automatically override a LOCA signal when the operator opened the drywell |
|
| |
|
| hydrogen mixing systems each have two lines penetrating
| | inlet valves. However, the licensee did not provide this LOCA override |
|
| |
|
| the drywell; an outlet line having a recirculating | | capability for the outlet line valves. |
|
| |
|
| fan to draw suction from the drywell and an inlet line that allows diluted air to reenter the drywell.Each of these lines has two isolation
| | Discussion |
|
| |
|
| valves which are normally closed during plant operation.
| | In both of these cases, the licensee changed the design with the intention of |
|
| |
|
| In 1983, during construction, the licensee added a LOCA interlock
| | increasing the reliability of safety systems. However, because the licensees |
|
| |
|
| to the hydrogen mixing system that would automatically
| | did not adequately review and test the designs, these changes introduced |
|
| |
|
| close all eight of the mixing system valves upon receiving
| | errors that could have prevented the systems from performing their safety |
|
| |
|
| a LOCA signal. In 1984, the licensee revised the control logic for the mixing system valves to automatically
| | functions as intended. |
|
| |
|
| override a LOCA signal when the operator opened the drywell inlet valves. However, the licensee did not provide this LOCA override capability
| | At Arkansas Nuclear One, the licensee intended to increase the reliability of |
|
| |
|
| for the outlet line valves.Discussion
| | the HPSI system by causing both HPSI oil pumps to operate simultaneously when |
|
| |
|
| In both of these cases, the licensee changed the design with the intention
| | an ESFAS signal was present. However, the oil pumps had apparently never been |
|
| |
|
| of increasing
| | run simultaneously for any extended period untiltthe recent overhaul test. |
|
| |
|
| the reliability | | IN 92-65 September 3, 1992 The licensee routinely conducted the' required periodic pump surveillance tests |
|
| |
|
| of safety systems. However, because the licensees did not adequately
| | with the HPSI operating in the normal-reactor makeup'mode with only one oil |
|
| |
|
| review and test the designs, these changes introduced
| | pump running at a time. The licensee tested the effectiveness of the |
|
| |
|
| errors that could have prevented
| | ESFAS signal during each refueling outage. However, the test only required |
|
| |
|
| the systems from performing | | verification that the test signal would actuate the HPSI system and did not |
|
| |
|
| their safety functions
| | result in the simultaneous operation'of the two oil pumps for an extended |
|
| |
|
| as intended.At Arkansas Nuclear One, the licensee intended to increase the reliability
| | time. As a result, neither of these tests revealed the oil leakage problem. |
|
| |
|
| of the HPSI system by causing both HPSI oil pumps to operate simultaneously
| | The licensee estimated'that a HPSI 'pump would have performed satisfactorily |
|
| |
|
| when an ESFAS signal was present. However, the oil pumps had apparently
| | for only 80 minutes without-operator action to replenish the oil or to stop |
|
| |
|
| never been run simultaneously
| | the electric oil pumps. With an ESFAS signal present, the electric oil pumps |
|
| |
|
| for any extended period untiltthe
| | cannot be stopped from the control room,'but must be'stopped by opening local |
|
| |
|
| recent overhaul test.
| | power supply breakers. |
|
| |
|
| IN 92-65 September
| | The licensee has modified the oil pressure relief'valve settings to minimize |
|
| |
|
| 3, 1992 The licensee routinely
| | the oil leakage. Procedures were established that instruct the operators to |
|
| |
|
| conducted
| | stop the electric oil pumps 15 minutes after an ESFAS actuation of the pumps. |
|
| |
|
| the' required periodic pump surveillance | | At River Bend, the control logic to automatically close all of the mixing |
|
| |
|
| tests with the HPSI operating
| | system valves was provided to ensure that the drywell integrity would be |
|
| |
|
| in the normal-reactor
| | restored if a LOCA occurred during a mixing system test with the valves open. |
|
| |
|
| makeup'mode
| | Apparently, the LOCA override for the inlet valves was provided later to |
|
| |
|
| with only one oil pump running at a time. The licensee tested the effectiveness
| | permit the drywell to be depressurized to clear a false LOCA signal that might |
|
| |
|
| of the ESFAS signal during each refueling | | be caused by a loss of offsite power. The false LOCA signal could be |
|
| |
|
| outage. However, the test only required verification
| | generated by the drywell pressure rise that would accompany a loss of drywell |
|
| |
|
| that the test signal would actuate the HPSI system and did not result in the simultaneous
| | cooling. Since the drywell could be depressurized without opening the outlet |
|
| |
|
| operation'of
| | valves, the LOCA override was not provided for these valves. The need to open |
|
| |
|
| the two oil pumps for an extended time. As a result, neither of these tests revealed the oil leakage problem.The licensee estimated'that | | the outlet to operate the hydrogen mixing was apparently not considered for |
|
| |
|
| a HPSI 'pump would have performed
| | this change. Normal surveillance testing did not reveal this design error |
|
| |
|
| satisfactorily
| | because it was never conducted with a LOCA signal present. |
|
| |
|
| for only 80 minutes without-operator
| | When the licensee discovered this design error, it declared both hydrogen |
|
| |
|
| action to replenish
| | mixing trains inoperable and commenced shutting down the reactor. The |
|
| |
|
| the oil or to stop the electric oil pumps. With an ESFAS signal present, the electric oil pumps cannot be stopped from the control room,'but | | licensee then developed a LOCA bypass procedure for the hydrogen mixing |
|
| |
|
| must be'stopped
| | system. |
|
| |
|
| by opening local power supply breakers.The licensee has modified the oil pressure relief'valve
| | These events highlight the importance of thoroughly reviewing any safety- related design change, including considering the effect of the change on all |
|
| |
|
| settings to minimize the oil leakage. Procedures
| | related systems. The events also show the need for completely testing the |
|
| |
|
| were established
| | systems affected by the design change under conditions that simulate as nearly |
|
| |
|
| that instruct the operators | | as possible those conditions that are expected to exist when the systems are |
|
| |
|
| to stop the electric oil pumps 15 minutes after an ESFAS actuation
| | needed. |
|
| |
|
| of the pumps.At River Bend, the control logic to automatically
| | IN 92-65 September 3, 1992 This information notice requires no specific action or written response. If |
|
| |
|
| close all of the mixing system valves was provided to ensure that the drywell integrity
| | you have any questions about the information in this notice, please contact |
|
| |
|
| would be restored if a LOCA occurred during a mixing system test with the valves open.Apparently, the LOCA override for the inlet valves was provided later to permit the drywell to be depressurized
| | -the technical contact listed below or the appropriate Office of Nuclear |
|
| |
|
| to clear a false LOCA signal that might be caused by a loss of offsite power. The false LOCA signal could be generated
| | Reactor Regulation (NRR) project manager. |
|
| |
|
| by the drywell pressure rise that would accompany
| | harles E. Rossi, Director' |
| | Division of-Operational Events Assessment |
|
| |
|
| a loss of drywell cooling. Since the drywell could be depressurized
| | Office of Nuclear Reactor Regulation |
|
| |
|
| without opening the outlet valves, the LOCA override was not provided for these valves. The need to open the outlet to operate the hydrogen mixing was apparently
| | Technical contact: Thomas F. Westerman, RIV |
|
| |
|
| not considered
| | (817) 860-8145 Attachment: List of Recently Issued NRC Information Notices |
|
| |
|
| for this change. Normal surveillance
| | V)Attachment |
| | |
| testing did not reveal this design error because it was never conducted
| |
| | |
| with a LOCA signal present.When the licensee discovered
| |
| | |
| this design error, it declared both hydrogen mixing trains inoperable
| |
| | |
| and commenced
| |
| | |
| shutting down the reactor. The licensee then developed
| |
| | |
| a LOCA bypass procedure
| |
| | |
| for the hydrogen mixing system.These events highlight
| |
| | |
| the importance
| |
| | |
| of thoroughly
| |
| | |
| reviewing
| |
| | |
| any safety-related design change, including
| |
| | |
| considering
| |
| | |
| the effect of the change on all related systems. The events also show the need for completely
| |
| | |
| testing the systems affected by the design change under conditions
| |
| | |
| that simulate as nearly as possible those conditions
| |
| | |
| that are expected to exist when the systems are needed.
| |
| | |
| IN 92-65 September
| |
| | |
| 3, 1992 This information
| |
|
| |
|
| notice requires no specific action or written response.
| | IN 92-65 September 3, 1992 LIST OF RECENTLY ISSUED |
|
| |
|
| If you have any questions
| | NRC INFORMATION NOTICES |
|
| |
|
| about the information
| | Information Date of |
|
| |
|
| in this notice, please contact-the technical
| | Notice No. Subject Issuance Issued to |
|
| |
|
| contact listed below or the appropriate
| | 92-64 Nozzle Ring Settings 08/28/92 All holders of OLs or CPs |
|
| |
|
| Office of Nuclear Reactor Regulation (NRR) project manager.harles E. Rossi, Director'Division of-Operational
| | on Low Pressure Water- for nuclear power reactors. |
|
| |
|
| ===Events Assessment===
| | Relief Valves |
| Office of Nuclear Reactor Regulation
| |
|
| |
|
| Technical
| | 92-63 Cracked Insulators in 08/26/92 All holders of OLs or CPs |
|
| |
|
| contact: Thomas F. Westerman, RIV (817) 860-8145 Attachment:
| | ASL Dry Type Transformers for nuclear power reactors. |
| List of Recently Issued NRC Information
| |
|
| |
|
| Notices
| | Manufactured by Westing- house Electric Corporation |
|
| |
|
| V)Attachment
| | 92-62 Emergency Response 08/24/92 All U.S. Nuclear Regulatory |
|
| |
|
| IN 92-65 September
| | Information Require- Commission licensees. |
|
| |
|
| 3, 1992 LIST OF RECENTLY ISSUED NRC INFORMATION
| | ments for Radioactive |
|
| |
|
| NOTICES Information
| | Material Shipments |
|
| |
|
| Date of Notice No. Subject Issuance Issued to 92-64 Nozzle on Low Relief Ring Settings Pressure Water-Valves 92-63 92-62 92-61 60 92-59 92-58 Cracked Insulators
| | 92-61 Loss of High Head 08/20/92 All holders of OLs or CPs |
|
| |
|
| in ASL Dry Type Transformers
| | Safety Injection for nuclear power reactors. |
|
| |
|
| Manufactured
| | 60 Valve Stem Failure 08/20/92 All holders of OLs or CPs |
|
| |
|
| by Westing-house Electric Corporation | | Caused by Embrittlement for pressurized water |
|
| |
|
| Emergency
| | reactors (PWRs). |
|
| |
|
| Response Information
| | 92-59 Horizontally-Installed 08/18/92 All holders of OLs or CPs |
|
| |
|
| Require-ments for Radioactive
| | Motor-Operated Gate for nuclear power reactors. |
|
| |
|
| Material Shipments Loss of High Head Safety Injection Valve Stem Failure Caused by Embrittlement
| | Valves |
|
| |
|
| Horizontally-Installed
| | 92-58 Uranium Hexafluoride 08/12/92 All Fuel Cycle Licensees. |
|
| |
|
| Motor-Operated
| | Cylinders - Deviations |
|
| |
|
| ===Gate Valves Uranium Hexafluoride===
| | in Coupling Welds |
| Cylinders
| |
|
| |
|
| -Deviations | | 92-57 Radial Cracking of- 08/11/92 All holders of OLs or CPs |
|
| |
|
| in Coupling Welds 08/28/92 08/26/92 08/24/92 08/20/92 08/20/92 08/18/92 08/12/92 All holders of OLs or CPs for nuclear power reactors.All holders of OLs or CPs for nuclear power reactors.All U.S. Nuclear Regulatory
| | Shroud Support Access for boiling water reactors |
|
| |
|
| Commission
| | Hole Cover Welds (BWRs). |
|
| |
|
| licensees.
| | 92-56 Counterfeit Valves in 08/06/92 All holders of OLs or CPs |
|
| |
|
| All holders of OLs or CPs for nuclear power reactors.All holders of OLs or CPs for pressurized
| | the Commercial Grade for nuclear power reactors. |
|
| |
|
| water reactors (PWRs).All holders of OLs or CPs for nuclear power reactors.All Fuel Cycle Licensees.
| | Supply System |
|
| |
|
| 92-57 Radial Cracking of-Shroud Support Access Hole Cover Welds 08/11/92 All holders for boiling (BWRs).of OLs or CPs water reactors 92-56 92-55 Counterfeit | | 92-55 Current Fire Endurance 07/27/92 All holders of OLs or CPs |
|
| |
|
| Valves in the Commercial
| | Test Results for for nuclear power reactors. |
|
| |
|
| Grade Supply System Current Fire Endurance Test Results for Thermo-Lag
| | Thermo-Lag Fire Barrier |
|
| |
|
| Fire Barrier Material 08/06/92 07/27/92 All holders of OLs or CPs for nuclear power reactors.All holders of OLs or CPs for nuclear power reactors.OL = Operating
| | Material |
|
| |
|
| License CP = Construction | | OL = Operating License |
|
| |
|
| Permit}} | | CP = Construction Permit}} |
|
| |
|
| {{Information notice-Nav}} | | {{Information notice-Nav}} |
Safety System Problems Caused by Modifications That Were Not Adequately Reviewed and TestedML031200373 |
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: |
09/03/1992 |
---|
From: |
Rossi C Office of Nuclear Reactor Regulation |
---|
To: |
|
---|
References |
---|
IN-92-065, NUDOCS 9208280105 |
Download: ML031200373 (5) |
|
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>. |
v)
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION_
WASHINGTON, D.C. 20555 6 September 3, 1992 -
7 '
NRC INFORMATION NOTICE 92-65: SAFETY SYSTEM PROBLEM flCAUSED BY MODIFICATIONS
THAT WERE NOT ADEQUATE-V REVIEWED AND TESTED
Addressees
All holders of operating licenses or construction permits for nuclear power
reactors.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to problems caused by inadequate review and testing
of safety system modifications. 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.
Description of Circumstances
The following describes two examples of safety system design errors that went
undetected since construction, because design changes were not thoroughly
reviewed and tested.
On October 10, 1991, during post overhaul testing, personnel at Arkansas
Nuclear One, Unit 1, observed that one of the high-pressure safety injection
(HPSI) pumps was losing its lubricating oil at a rate of more than 15 gallons
per hour as a result of oil spraying from the bearings. The licensee found
that the oil would always leak at this 'rate during emergency operation because
of excessive oil pressure caused by the simultaneous operation of two oil -
pumps that-served the HPSI pump. This condition had-existed since the plant
began operation.
The bearings for each of the HPSI pumps are supplied with lubricating oil by
two oil pumps, one attached directly to the HPSI pump itself and the other a
separate electric backup pump. Originally the electric oil pumps were
intended to be used during start up of a HPSI pump or to replace a
malfunctioning attached oil pump. The electric oil pumps could be started
manually and would start automatically when the oil pressure decreased below a
certain point. The licensee continues to use this method of control when the
HPSI pumps are used for normal reactor water makeup. However, during
construction, the licensee decided that the HPSI pumps would be more reliable
if the electric lubricating oil pumps ran continuously during emergency
operation. Consequently, the licensee modified the emergency controls to keep
9208280105 ?DR J4E fltc6-S-O
c a0 7 0-3 WI
IN 92-65 September 3, 1992 the electric oil pumps-operating whenever an emergency safety features
actuation system (ESFAS) signal was present. Anticipating that the
simultaneous operation of both oil pumps could cause excessive oil pressure, the licensee added an oil- pressure relief valve to the oil system. However, the relief valve settings were not appropriately selected to prevent oil
spraying from the bearings.
In September 1991, the Gulf States Utilities Company, licensee for the River
Bend Station, discovered that the outlet valves for the hydrogen mixing system
would immediately close if an operator attempted to start up the system by
opening these valves when a loss-of-coolant accident (LOCA) signal was
present. An interlock prevented the mixing system fans from operating with
the outlet valves closed. Consequently, the hydrogen mixing system would have
been inoperable if a LOCA signal were present. This condition had existed
since the plant was constructed.
The River Bend Station is a boiling water reactor with a Mark III containment
structure. This containment structure consists of two chambers, a large outer
primary containment and a drywell which is inside the primary containment and
surrounds the reactor vessel. This system suppresses the steam pressure
released during a LOCA by directing the steam through the suppression pool
water into the primary containment. After the initial pressure suppression is
complete following a LOCA, hydrogen created by the zirconium-water reaction
would be mainly concentrated in the drywell. The hydrogen mixing system is
provided to reduce the concentration of the hydrogen in the drywell by moving
it into the primary containment where it is diluted and reduced in
concentration by the hydrogen recombiners.
The redundant hydrogen mixing systems each have two lines penetrating the
drywell; an outlet line having a recirculating fan to draw suction from the
drywell and an inlet line that allows diluted air to reenter the drywell.
Each of these lines has two isolation valves which are normally closed during
plant operation. In 1983, during construction, the licensee added a LOCA
interlock to the hydrogen mixing system that would automatically close all
eight of the mixing system valves upon receiving a LOCA signal. In 1984, the
licensee revised the control logic for the mixing system valves to
automatically override a LOCA signal when the operator opened the drywell
inlet valves. However, the licensee did not provide this LOCA override
capability for the outlet line valves.
Discussion
In both of these cases, the licensee changed the design with the intention of
increasing the reliability of safety systems. However, because the licensees
did not adequately review and test the designs, these changes introduced
errors that could have prevented the systems from performing their safety
functions as intended.
At Arkansas Nuclear One, the licensee intended to increase the reliability of
the HPSI system by causing both HPSI oil pumps to operate simultaneously when
an ESFAS signal was present. However, the oil pumps had apparently never been
run simultaneously for any extended period untiltthe recent overhaul test.
IN 92-65 September 3, 1992 The licensee routinely conducted the' required periodic pump surveillance tests
with the HPSI operating in the normal-reactor makeup'mode with only one oil
pump running at a time. The licensee tested the effectiveness of the
ESFAS signal during each refueling outage. However, the test only required
verification that the test signal would actuate the HPSI system and did not
result in the simultaneous operation'of the two oil pumps for an extended
time. As a result, neither of these tests revealed the oil leakage problem.
The licensee estimated'that a HPSI 'pump would have performed satisfactorily
for only 80 minutes without-operator action to replenish the oil or to stop
the electric oil pumps. With an ESFAS signal present, the electric oil pumps
cannot be stopped from the control room,'but must be'stopped by opening local
power supply breakers.
The licensee has modified the oil pressure relief'valve settings to minimize
the oil leakage. Procedures were established that instruct the operators to
stop the electric oil pumps 15 minutes after an ESFAS actuation of the pumps.
At River Bend, the control logic to automatically close all of the mixing
system valves was provided to ensure that the drywell integrity would be
restored if a LOCA occurred during a mixing system test with the valves open.
Apparently, the LOCA override for the inlet valves was provided later to
permit the drywell to be depressurized to clear a false LOCA signal that might
be caused by a loss of offsite power. The false LOCA signal could be
generated by the drywell pressure rise that would accompany a loss of drywell
cooling. Since the drywell could be depressurized without opening the outlet
valves, the LOCA override was not provided for these valves. The need to open
the outlet to operate the hydrogen mixing was apparently not considered for
this change. Normal surveillance testing did not reveal this design error
because it was never conducted with a LOCA signal present.
When the licensee discovered this design error, it declared both hydrogen
mixing trains inoperable and commenced shutting down the reactor. The
licensee then developed a LOCA bypass procedure for the hydrogen mixing
system.
These events highlight the importance of thoroughly reviewing any safety- related design change, including considering the effect of the change on all
related systems. The events also show the need for completely testing the
systems affected by the design change under conditions that simulate as nearly
as possible those conditions that are expected to exist when the systems are
needed.
IN 92-65 September 3, 1992 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.
harles E. Rossi, Director'
Division of-Operational Events Assessment
Office of Nuclear Reactor Regulation
Technical contact: Thomas F. Westerman, RIV
(817) 860-8145 Attachment: List of Recently Issued NRC Information Notices
V)Attachment
IN 92-65 September 3, 1992 LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
Information Date of
Notice No. Subject Issuance Issued to
92-64 Nozzle Ring Settings 08/28/92 All holders of OLs or CPs
on Low Pressure Water- for nuclear power reactors.
Relief Valves
92-63 Cracked Insulators in 08/26/92 All holders of OLs or CPs
ASL Dry Type Transformers for nuclear power reactors.
Manufactured by Westing- house Electric Corporation
92-62 Emergency Response 08/24/92 All U.S. Nuclear Regulatory
Information Require- Commission licensees.
ments for Radioactive
Material Shipments
92-61 Loss of High Head 08/20/92 All holders of OLs or CPs
Safety Injection for nuclear power reactors.
60 Valve Stem Failure 08/20/92 All holders of OLs or CPs
Caused by Embrittlement for pressurized water
reactors (PWRs).
92-59 Horizontally-Installed 08/18/92 All holders of OLs or CPs
Motor-Operated Gate for nuclear power reactors.
Valves
92-58 Uranium Hexafluoride 08/12/92 All Fuel Cycle Licensees.
Cylinders - Deviations
in Coupling Welds
92-57 Radial Cracking of- 08/11/92 All holders of OLs or CPs
Shroud Support Access for boiling water reactors
Hole Cover Welds (BWRs).
92-56 Counterfeit Valves in 08/06/92 All holders of OLs or CPs
the Commercial Grade for nuclear power reactors.
Supply System
92-55 Current Fire Endurance 07/27/92 All holders of OLs or CPs
Test Results for for nuclear power reactors.
Thermo-Lag Fire Barrier
Material
OL = Operating License
CP = Construction Permit
|
---|
|
list | - Information Notice 1992-01, Cable Damage Caused by Inadequate Cable Installation Procedures and Controls (3 January 1992)
- Information Notice 1992-02, Relap5/MOD3 Computer Code Error Associated with the Conservation of Energy Equation (3 January 1992)
- Information Notice 1992-02, Relap5/Mod3 Computer Code Error Associated with the Conservation of Energy Equation (3 January 1992)
- Information Notice 1992-03, Remote Trip Function Failures in General Electric F-Frame Molded-Case Circuit Breakers (6 January 1992)
- Information Notice 1992-04, Potter and Brumfield Model Mdr Rotary Relay Failures (6 January 1992, Topic: Probabilistic Risk Assessment)
- Information Notice 1992-04, Potter and Brumfield Model MDR Rotary Relay Failures (6 January 1992, Topic: Probabilistic Risk Assessment)
- Information Notice 1992-05, Potential Coil Insulations Breakdown in Abs RXMH2 Relays (8 January 1992)
- Information Notice 1992-05, Potential Coil Insulations Breakdown in Abs Rxmh2 Relays (8 January 1992)
- Information Notice 1992-05, Potential Coil Insulations Breakdown in ABS RXMH2 Relays (8 January 1992)
- Information Notice 1992-06, Reliability of ATWS Mitigation System and Other NRC Required Equipment Not Controlled by Plant Technical Specifications (15 January 1992)
- Information Notice 1992-06, Reliability of ATWS Mitigation System and Other NRC Required Equipment not Controlled by Plant Technical Specifications (15 January 1992)
- Information Notice 1992-07, Rapid Flow-induced Erosion/Corrosion of Feedwater Piping (9 January 1992)
- Information Notice 1992-08, Revised Protective Action Guidance for Nuclear Incidents (23 January 1992)
- Information Notice 1992-09, Overloading and Subsequent Lock Out of Electrical Buses During Accident Conditions (30 January 1992)
- Information Notice 1992-10, Brachytherapy Incidents Involving Iridium-192 Wire Used in Endobronchial Treatments (31 January 1992, Topic: Brachytherapy)
- Information Notice 1992-10, Brachytherapy Incidents Involving Iridium-192 Wire used in Endobronchial Treatments (31 January 1992, Topic: Brachytherapy)
- Information Notice 1992-11, Soil and Water Contamination at Fuel Cycle Facilities (5 February 1992, Topic: Brachytherapy)
- Information Notice 1992-12, Effects of Cable Leakage Currents on Instrument Settings and Indications (10 February 1992, Topic: Brachytherapy)
- Information Notice 1992-13, Inadequate Control Over Vehicular Traffic at Nuclear Power Plant Sites (18 February 1992, Topic: Brachytherapy)
- Information Notice 1992-14, Uranium Oxide Fires at Fuel Cycle Facilities (21 February 1992, Topic: Brachytherapy)
- Information Notice 1992-15, Failure of Primary Systems Compression Fitting (24 February 1992)
- Information Notice 1992-16, Loss of Flow from the Residual Heat Removal Pump During Refueling Cavity Draindown (25 February 1992, Topic: Reactor Vessel Water Level, Temporary Modification, Brachytherapy)
- Information Notice 1992-17, NRC Inspections of Programs Being Developed at Nuclear Power Plants in Response to Generic Letter 89-10 (26 February 1992, Topic: Stroke time)
- Information Notice 1992-18, Potential for Loss of Remote Shutdown Capability During a Control Room Fire (28 February 1992, Topic: Hot Short, Safe Shutdown)
- Information Notice 1992-19, Misapplication of Potter and Brumfield Mdr Rotary Relays (2 March 1992)
- Information Notice 1992-19, Misapplication of Potter and Brumfield MDR Rotary Relays (2 March 1992)
- Information Notice 1992-20, Inadequate Local Leak Rate Testing (3 March 1992)
- Information Notice 1992-21, Spent Fuel Pool Reactivity Calculations (24 March 1992)
- Information Notice 1992-23, Results of Validation Testing of Motor-Operated Valve Diagnostic Equipment (27 March 1992)
- Information Notice 1992-24, Distributor Modification to Certain Commercial-Grade Agastat Electrical Relays (30 March 1992)
- Information Notice 1992-25, Pressure Locking of Motor-Operated Flexible Wedge Gate Valves (2 April 1992, Topic: Stroke time, Hydrostatic)
- Information Notice 1992-27, Thermally Induced Accelerated Aging and Failure of ITE/Gould A.C. Relays used in Safety-Related Applications (3 April 1992)
- Information Notice 1992-27, Thermally Induced Accelerated Aging and Failure of Ite/Gould A.C. Relays Used in Safety-Related Applications (3 April 1992)
- Information Notice 1992-28, Inadequate Fire Suppression System Testing (8 April 1992, Topic: Safe Shutdown)
- Information Notice 1992-29, Potential Breaker Miscoordination Caused by Instantaneous Trip Circuitry (17 April 1992)
- Information Notice 1992-30, Falsification of Plant Records (23 April 1992)
- Information Notice 1992-31, Electrical Connection Problem in Johnson Yokogawa Corporation YS-80 Programmable Indicating Controllers (27 April 1992)
- Information Notice 1992-32, Problems Identified with Emergency Ventilation Systems for Near-Site (Within 10 Miles) Emergency Operations Facilities and Technical Support Centers (29 April 1992)
- Information Notice 1992-32, Problems Identified with Emergency Ventilation Systems for Near-Site (within 10 Miles) Emergency Operations Facilities and Technical Support Centers (29 April 1992)
- Information Notice 1992-33, Increased Instrument Response Time When Pressure Dampening Devices Are Installed (30 April 1992)
- Information Notice 1992-33, Increased Instrument Response Time When Pressure Dampening Devices are Installed (30 April 1992)
- Information Notice 1992-34, New Exposure Limits for Airborne Uranium and Thorium (6 May 1992)
- Information Notice 1992-35, Higher than Predicted Erosion/Corrosion in Unisolable Reactor Coolant Pressure Boundary Piping Inside Containment at a Boiling Water Reactor (6 May 1992)
- Information Notice 1992-35, Higher than Predicted Erosion/Corrosion in Unisolable Reactor Coolant Pressure Boundary Piping inside Containment at a Boiling Water Reactor (6 May 1992)
- Information Notice 1992-36, Intersystem LOCA Outside Containment (7 May 1992)
- Information Notice 1992-37, Implementation of the Deliberate Misconduct Rule (8 May 1992)
- Information Notice 1992-38, Implementation Date for the Revision to the EPA Manual of Protective Action Guides and Protective Actions for Nuclear Incidents (26 May 1992, Topic: Brachytherapy)
- Information Notice 1992-39, Unplanned Return to Criticality During Reactor Shutdown (13 May 1992, Topic: Fuel cladding)
- Information Notice 1992-40, Inadequate Testing of Emergency Bus Undervoltage Logic Circuitry (27 May 1992)
- Information Notice 1992-41, Consideration of Stem Rejection Load In Calculation of Required Valve Thrust (29 May 1992, Topic: Anchor Darling)
... further results |
---|