Text

Enclosure 3 - Generic Communications - Master Table
	ML11213A132
Person / Time
Site:	Watts Bar
Issue date:	07/27/2011
From:	; Tennessee Valley Authority
To:	; Office of Nuclear Reactor Regulation
References
	Download: ML11213A132 (110)
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Enclosure 3 Generic Communications

-Master Table GENERIC COMMUNICATIONS:

MASTER TABLE ITEM B 71-002 TITLE PWR Reactor Trip Circuit Breakers REV ADDITIC NA Addressed to specific plant(s).)NAL INFORMATION B 71-003 Catastrophic Failure of Main Steam Line Relief Valve Headers B 72-001 Failed Hangers for Emergency Core Cooling System Suction Header B 72-002 Simultaneous Actuation of a Safety Injection Signal on Both Units of a Dual Unit Facility B 72-003 Limitorque Valve Operator Failures B 73-001 Faulty Overcurrent Trip Delay Device in Circuit Breakers for Engineered Safety Systems B 73-002 Malfunction of Containment Purge Supply Valve Switch NA Addressed to specific plant(s).NA Addressed to specific plant(s).NA Addressed to specific plant(s).s NA Addressed to specific plant(s).C TVA: letter dated April 4, 1973 NRC: IR 390/391 75-5 C TVA: letter dated August 22, 1973 NRC: IR 390/391 75-5 C TVA: letter dated February 7, 1985 NRC: IR 390/391 85-08 C TVA: memo dated February 7, 1985 NRC: IR 390/391 85-08 NA Boiling Water Reactor B 73-003 Defective Hydraulic Snubbers and Restraints B 73-004 Defective Bergen-Patterson Hydraulic Shock Absorbers B 73-005 Manufacturing Defect in BWR Control Rods B 73-006 Inadvertent Criticality in a BWR B 74-001 Valve Deficiencies B 74-002 Truck Strike Possibility NA Boiling Water Reactor C TVA: letter dated April 15, 1974 NRC: IR 390/391 75-5 NA Info Page 1 of 109* = See last page for status code definition.

ITEM B 74-003 TITLE Failure of Structural or Seismic Support Bolts on Class I Components REV CI 06 ADDITIONAL INFORMATION TVA: memo dated January 22, 1985 NRC: IR 390/391 85-08 Approach accepted in IR 50-390/85-08 and 50-391/85-08 (March 29, 1985).Unit 2 Action: Implement per NUREG-0577 as was done for Unit 1.REVISION 06 UPDATE: Corrective action for this item consisted of a bolting reheat treatment program for both units; it has been completed.

B 74-004 Malfunction of Target Rock Safety Relief Valves B 74-005 Shipment of an Improperly Shielded Source B 74-006 Defective Westinghouse Type W-2 Control Switch Component B 74-007 Personnel Exposure -Irradiation Facility B 74-008 Deficiency in the ITE Molded Case Circuit Breakers, Type HE-3 NA NA C Boiling Water Reactor Does not apply to power reactor.TVA: letter dated October 18, 1974 NRC: IR 390/391 75-6 NA Does not apply to power reactor.C TVA: letter dated August 21, 1974 NRC: IR 390/391 75-5 B 74-009 Deficiency in GE Model 4KV C TVA: letter dated September 20, 1974 Magne-Blast Circuit Breakers NRC: IR 390/391 76-6 B 74-010 Failures in 4-Inch Bypass Pipe at NA Boiling Water Reactor Dresden 2 B 74-011 Improper Wiring of Safety Injection C NRC: IR 390/391 75-6 Logic at Zion 1 & 2 B 74-012 Incorrect Coils in Westinghouse C NRC: IR 390/391 75-5 Type SG Relays at Trojan B 74-013 Improper Factory Wiring on GE C TVA: letter dated December 24, 1974 Motor Control Centers at Fort Calhoun NRC: IR 390/391 75-5 Page 2 of 109* = See last page for status code definition.

ITEM TITLE B 74-014 BWR Relief Valve Discharge to Suppression Pool B 74-015 Misapplication of Cutler-Hammer Three Position Maintained Switch Model No. 10250T REV NA C 06 ADDITIONAL INFORMATION Boiling Water Reactor TVA: letter dated May 5, 1975 NRC: IR 390/391 75-5 Unit 2 Action: Install modified A3 Cutler-Hammer 10250T switches.REVISION 06 UPDATE: It has been confirmed that WBN Unit 2 never had the faulty switches.NRC Inspection Report 391/2010-605 .closed B 74-015.B 74-016 Improper Machining of Pistons in Colt Industries (Fairbanks-Morse)

Diesel-Generators B 75-001 Through-Wall Cracks in Core Spray Piping at Dresden-2 B 75-002 Defective Radionics Radiograph Exposure Devices and Source Changers B 75-003 Incorrect Lower Disc Spring and Clearance Dimension in Series 8300 and 8302 ASCO Solenoid Valves C TVA: letter dated January 2, 1975 NRC: IR 390/391 75-3 NA Boiling Water Reactor NA CI Does not apply to power reactor.TVA: letter dated May 16, 1975 NRC: IR 390/391 75-6 NRC accepted in IR 50-390/75-6 and 50-391/75-6 (August 21, 1975).Unit 2 Action: B 75-004 Cable Fire at BFNPP CI Modify valves not modified at factory.NRC: IR 390/391 85-08 Closed to Fire Protection CAP Part of Fire Protection CAP Page 3 of 109* = See last page for status code definition.

REV ITEM B 75-005 TITLE Operability of Category I Hydraulic Shock and Sway Suppressors ADDITIONAL INFORMATION Cl TVA: letter dated June 16, 1975 NRC: IR 390/391 75-6 NRC accepted in IR 50-390/75-6 and 50-391/75-6 (August 21, 1975).Unit 2 Action: Install proper suppressors.

B 75-006 Defective Westinghouse Type OT-2 Control Switches Cl TVA: letter dated July 31, 1975 NRC: IR 390/85-25 and 391/85-20 06 Unit 2 Action: Inspect Westinghouse Type OT-2 control switches.[WAS "NOTE 3."]REVISION 06 UPDATE: All Unit 2 Type OT-2 switches procured or refurbished are inspected and tested.B 75-007 Exothermic Reaction in Radwaste Shipment B 75-008 PWR Pressure Instrumentation NA Does not apply to power reactor.S NRC: IR 390/391 85-08 02 Unit 2 Action: Ensure that Technical Specifications and Site Operating Instructions address importance of maintaining temperature and pressure within prescribed limits.REVISION 02 UPDATE: Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.Adherence to Pressure and Temperature limits is required by the following portions of the Unit 2 TS: 1.1 [definition of "PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)"];

3.4.3 ["RCS Pressure and Temperature (P/T) Limits"];

3.4.12 ["Cold Overpressure Mitigation System (COMS)"];

and 5.9.6 ["Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)"].B 76-001 BWR Isolation Condenser Tube Failure NA Boiling Water Reactor Page 4 of 109* = See last page for status code definition.

ITEM B 76-002 TITLE Relay Coil Failures -GE Types HFA, HGA, HKA, HMA Relays REV Cl Unit 2 Action: ADDITIONAL INFORMATION B 76-003 Relay Malfunctions

-GE Type C STD Relays B 76-004 Cracks in Cold Worked Piping at NA BWRs B 76-005 Relay Failures -Westinghouse C BFD Relays B 76-006 Diaphragm Failures in Air C Operated Auxiliary Actuators for Safety/Relief Valves B 76-007 Crane Hoist Control Circuit C Modifications B 76-008 Teletherapy Units NA B 77-001 Pneumatic Time Delay Relay C Setpoint Drift B 77-002 Potential Failure Mechanism in C Certain Westinghouse AR Relays with Latch Attachments B 77-003 On-Line Testing of the CI Westinghouse Solid State Protection System B 77-004 Calculation Error Affecting The S Design Performance of a System ---for Controlling pH of Containment 02 Sump Water Following a LOCA Repair or replace relays before preoperational tests.TVA: letter dated May 17, 1976 NRC: IR 390/391 76-6 Boiling Water Reactor TVA: letter dated June 7, 1976 NRC: IR 390/391 85-08 TVA: memo dated January 25, 1985 NRC: IR 390/391 85-08 TVA: letter dated October 29, 1976 NRC: IR 390/391 85-08 Does not apply to power reactor.TVA: letter dated July 1, 1977 NRC: IR 390/391 85-08 TVA: letter dated November 11, 1977 NRC: IR 390/391 85-08 Unit 2 Action: Include necessary periodic testing in test procedures.

TVA: letter dated January 23, 1978 NRC: IR 390/78-11 and 391/78-09 Unit 2 Action: Ensure Technical Specifications includes limit on Boron concentration.

REVISION 02 UPDATE: Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.TS Surveillance Requirement 3.6.11.5 requires verification that the boron Page 5 of 109* = See last page for status code definition.

R REV ITEM B 77-005 and B 77-005 A B 77-006 TITLE Electrical Connector Assemblies C Potential Problems with Containment Electrical Penetration Assemblies C ADDITIONAL INFORMATION concentration is within a specified range.TVA: letter dated January 17, 1978 NRC: IR 390/78-11 and 391/78-09 Item was applicable only to units with operating license at the time the item was issued.NRC: IR 390/391 85-08 TVA: letter dated January 20, 1978 NRC: IR 390/78-11 and 391/78-09 Item concerns a multi-unit issue that was completed for both units.B 77-007 Containment Electrical Penetration Assemblies at Nuclear Power Plants Under Construction B 77-008 Assurance of Safety and Safeguards During an Emergency-Locking Systems C C TVA: letter dated March 1, 1978 B 78-001 Flammable Contact -Arm Retainers in GE CR120A Relays B 78-002 Terminal Block Qualification B 78-003 Potential Explosive Gas Mixture Accumulations Associated with BWR Offgas System Operations B 78-004 Environmental Qualification of Certain Stem Mounted Limit Switches Inside Reactor Containment NRC: IR 390/78-11 and 391/78-09 C TVA: letter dated March 20, 1978 NRC: IR 390/78-11 and 391/78-09 C TVA: letter dated March 1, 1978 NA CI NRC: IR 390/78-11 and 391/78-09 Boiling Water Reactor TVA: letter dated December 19, 1978 NRC: IR 390/82-13 and 391/82-10 Closed to EQ Program IR 50-390/82-13 and 50-391/82-10 (April 22, 1982) accepted approach.Unit 2 Action: Ensure NAMCO switches have been replaced.B 78-005 Malfunctioning of Circuit Breaker C TVA: letter dated June 12, 1978 Auxiliary Contact Mechanism

-GE Model CR105X NRC: IR 390/78-17 and 391/78-15 B 78-006 Defective Cutler-Hammer Type M C NRC: IR 390/78-22 and 391/78-19 Relays With DC Coils Page 6 of 109* = See last page for status code definition.

ITEM TITLE B 78-007 Protection Afforded by Air-Line Respirators and Supplied-Air Hoods B 78-008 Radiation Levels from Fuel Element Transfer Tubes B 78-009 BWR Drywell Leakage Paths Associated with Inadequate Drywell Closures B 78-010 Bergen-Patterson Hydraulic Shock Suppressor Accumulator Spring Coils B 78-011 Examination of Mark I Containment Torus Welds B 78-012 Atypical Weld Material in Reactor Pressure Vessel Welds B 78-013 Failures in Source Heads Kay Ray, Inc. Gauges Models 7050, 7050B, 7051,7051B, 7060, 7060B, 7061 and 7061B B 78-014 Deterioration of Buna-N Components in ASCO Solenoids B 79-001 Environmental Qualification of Class lE Equipment B 79-002 Pipe Support Base Plate Designs Using Concrete Expansion Anchor Bolts B 79-003 Longitudinal Weld Defects in ASME SA-312 Type 304 SS Pipe Spools Manufactured by Youngstown Welding &Engineering REV NA NA ADDITIONAL INFORMATION Item was applicable only to units with operating license at the time the item was issued.Item was applicable only to units with operating license at the time the item was issued.NRC: IR 390/391 85-08 NA Boiling Water Reactor C TVA: letter dated August 14, 1978 NRC: IR 390/78-22 and 391/78-19 NA Boiling Water Reactor C NA NA C TVA: Westinghouse letter dated October 29, 1979 NRC: IR 390/391 81-04 Does not apply to power reactor.Boiling Water Reactor NRC: IR 390/80-06 and 391/80-05 CI NRC review of HAAUP Program in NUREG-1232, SSER6, and SSER8.Unit 2 Actions: Addressed in CAP/SP.Conduct a complete review of affected support calculations, and perform the necessary revisions to design documents and field modifications to achieve compliance.

C TVA: letter dated July 16, 1981 NRC: IRs 390/82-21 and 391/82-17; 390/84-35 and 391/84-33 Page 7 of 109* = See last page for status code definition.

ITEM TITLE REV C ADDITIONAL INFORMATION B 79-004 Incorrect Weights for Swing Check Valves Manufactured by Velan Engineering Corporation B 79-005 Nuclear Incident at TMI B 79-006 Review of Operational Errors and System Misalignments Identified During the Three Mile Island Incident B 79-007 Seismic Stress Analysis of Safety-Related Piping B 79-008 Events Relevant to BWRs Identified During TMI Incident B 79-009 Failure of GE Type AK-2 Circuit Breaker in Safety Related Systems TVA: letter dated October 20, 1980 NRC: IR 390/83-15 and 391/83-11 NA Applies only to Babcock and Wilcox designed plants C NRC: IR 390/80-06 and 391/80-05 C TVA- letter dated May 31, 1979 NRC: IR 390/79-30 and 391/79-25 NA Boiling Water Reactor Cl TVA: letter dated June 20, 1979 06 Unit 2 Action: Complete preservice preventive maintenance on AK-2 Circuit Breakers.[WAS "NOTE 3."]----------------------------------------------------------------------------------------------------

REVISION 06 UPDATE: B 79-010 Requalification Training Program Statistics B 79-011 Faulty Overcurrent Trip Device in Circuit Breakers for Engineering Safety Systems B 79-012 Short Period Scrams at BWR Facilities B 79-013 Cracking in Feedwater Piping It has been confirmed that AK-2 Circuit Breakers are not used on Unit 2.NA Item was applicable only to units with operating license at the time the--item was issued.C TVA: letter dated July 20, 1979 NRC: IR 390/79-30 and 391/79-25 NA Boiling Water Reactor C Item was applicable only to units with operating license at the time the--item was issued.TVA: letter dated December 1, 1983 NRC: IR 390/391 85-08 Page 8 of 109* = See last page for status code definition.

ITEM B 79-014 TITLE Seismic Analysis for As-Built Safety-Related Piping Systems REV CI ADDITIONAL INFORMATION NRC review of HAAUP Program in NUREG-1232, SSER6, and SSER8.Unit 2 Actions: " Addressed in CAP/SP.* Initiate a Unit 2 hanger walkdown and hanger analysis program similar to the program for Unit 1.* Complete re-analysis of piping and associated supports as necessary.

Perform modifications as required by re-analysis.

B 79-015 Deep Draft Pump Deficiencies C TVA: letter dated January 24, 1992 NRC: IR 390/391 95-70 B 79-016 Vital Area Access Controls NA Item was applicable only to units with operating license at the time the_ --item was issued.NRC: IR 390/80-06 and 391/80-05 B 79-017 Pipe Cracks in Stagnant Borated NA Item was applicable only to units with operating license at the time the Water Systems at PWR Plants _ item was issued.NRC: IR 390/80-06 and 391/80-05; NUREG/ CR 5286 B 79-018 Audibility Problems Encountered NA Item was applicable only to units with operating license at the time the on Evacuation of Personnel from _ -. item was issued.High-Noise Areas NRC: IR 390/80-06 and 391/80-05 B 79-019 Packaging of Low-Level NA Item was applicable only to units with operating license at the time the Radioactive Waste for Transport

-item was issued.and Burial NRC: IR 390/80-06 and 391/80-05 B 79-020 Packaging, Transport and Burial of NA Item was applicable only to units with operating license at the time the Low-Level Radioactive Waste item was issued.NRC: IR 390/80-06 and 391/80-05 Page 9 of 109* P= See last page for status code definition.

ITEM B 79-021 TITLE Temperature Effects on Level Measurements REV C ADDITIONAL INFORMATION Reviewed in 7.2.5 of both the original 1982 SER and SSER14.06 Unit 2 Action: Update accident calculation.

CONFIRMATORY ISSUE -address IEB 79-21 to alleviate temperature dependence problem associated with measuring SG water level In SSER14, NRC concurred with TVA's assessment to not insulate the steam generator water level instrument reference leg.Unit 2 Action: Update accident calculation.

REVISION 06 UPDATE: The calculations were updated.B 79-022 Possible Leakage of Tubes of Tritium Gas Used in Time Pieces for Luminosity B 79-023 Potential Failure of Emergency Diesel Generator Field Exciter Transformer B 79-024 Frozen Lines NRC Inspection Report 391/2010-605 closed B 79-021.NA Does not apply to power reactor.NRC: IR 390/80-06 and 391/80-05 C TVA: letter dated October 29, 1979 NRC: IR 390/80-06 and 391/80-05 Cl Unit 2 Actions:* Insulate the section of piping in the containment spray full-flow test line that is exposed to outside air.* Confirm installation of heat tracing on the sensing lines off the feedwater flow elements.TVA: letter dated January 4, 1980 B 79-025 Failures of Westinghouse BFD Relays in Safety-Related Systems B 79-026 Boron Loss from BWR Control Blades B 79-027 Loss of Non-Class 1 E I & C Power System Bus During Operation C NRC: IR 390/80-03 and 391/80-02 NA Boiling Water Reactor Cl TVA responded to the Bulletin on March 1, 1982. Reviewed in 7.5.3 of the original 1982 SER.Unit 2 Action: Issue appropriate emergency procedures.

Page 10 of 109* = See last page for status code definition.

ITEM TITLE REV C ADDITIONAL INFORMATION TVA: letter dated April 1, 1993 B 79-028 Possible Malfunction of NAMCO Model EA1 80 Limit Switches at Elevated Temperatures B 80-001 Operability of ADS Valve Pneumatic Supply B 80-002 Inadequate QA for Nuclear Supplied Equipment B 80-003 Loss of Charcoal from Standard Type II, 2 Inch, Tray Adsorber Cells B 80-004 Analysis of a PWR Main Steam Line Break with Continued Feedwater Addition NRC: IR 390/391 93-32 NA Boiling Water Reactor NA Boiling Water Reactor C TVA: letter dated March 21, 1980 CI 06 NRC: IR 390/80-15 and 391/80-12 IR 50-390/85-60 and 50-391/85-49 (December 6, 1985) required completion of actions that included determination of temperature profiles inside and outside of containment following a MSLB for Unit 1.Unit 2 Action: Complete analysis for Unit 2.----------------------------------------------------------------------------------------------------

REVISION 06 UPDATE: The analysis for Unit 2 was completed.

B 80-005 Vacuum Condition Resulting in CI Closed in IR 50-390/84-59 and 50-391/84-45.

Damage to Chemical Volume Control System Holdup Tanks Unit 2 Action: Complete surveillance procedures for Unit 2.B 80-006 Engineered Safety Feature Reset CI TVA response dated March 11, 1982. Reviewed in 7.3.5 of the original Control 1982 SER.Unit 2 Action: Perform verification during the preoperational testing.B 80-007 BWR Jet Pump Assembly Failure NA Boiling Water Reactor B 80-008 Examination of Containment Liner C TVA: letter dated July 8, 1980 Penetration Welds NRC: IR 390/391 81-19 B 80-009 Hydramotor Actuator Deficiencies C TVA: letter dated January 15, 1981 NRC: NUREG/ CR 5291; IR 390/391 85-08; IR 390/85-60 and 391/85-49 Page 11 of 109* = See last page for status code definition.

ITEM B 80-010 TITLE REV ADDITIONAL INFORMATION Unit 2 Actions: Contamination of Nonradioactive System and Resulting Potential for Unmonitored, Uncontrolled Release of Radioactivity to Environment Cl 06 2) Include proper monitoring of non-radioactive systems in procedures.

REVISION 06 UPDATE: B 80-010 Contamination of Nonradioactive CI System and Resulting Potential for Unmonitored, Uncontrolled 06 Release of Radioactivity to Environment B 80-011 Masonry Wall Design Cl B 80-012 Decay Heat Removal System CI Operability

_ -Chemistry procedure CM-3.01 (System Chemistry Specification) includes a radiation monitoring system for non-radioactive systems and provides appropriate surveillance limits. Additionally, it provides required actions if the surveillance limits are not met.Unit 2 Actions: 1) Correct deficiencies involving monitoring of systems.....................................................................................................

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REVISION 06 UPDATE: Chemistry procedure CM-3.01 (System Chemistry Specification) includes a radiation monitoring system for non-radioactive systems and provides appropriate surveillance limits. Additionally, it provides required actions if the surveillance limits are not met.NRC accepted all but completion of corrective actions in IR 50-390/93-01 and 50-391/93-01(February 25, 1993) and closed for Unit I in IR 50-390/95-46 (August 1, 1995).Unit 2 Action: Complete implementation for Unit 2.NRC: IR 390/391 85-08; NUREG/CR 4005 Unit 2 Action: Implement operating instructions and abnormal operating instructions (AOIs) for RHR.[WAS "NOTE 3."]Boiling Water Reactor Boiling Water Reactor Item concerns a multi-unit issue that was completed for both units.NRC: IR 390/391 85-08 TVA: letter dated August 29, 1980 NRC: IR 390/391 81-17 B 80-013 Cracking in Core Spray Spargers NA B 80-014 Degradation of Scram Discharge NA Volume Capability

_B 80-015 Possible Loss of Emergency C Notification System with Loss of Offsite Power B 80-016 Potential Misapplication of C Rosemount, Inc. Models 1151 and 1152 Pressure Transmitters With Either "A" or "D" Output Codes Page 12 of 109* o= See last page for status code definition.

ITEM B 80-017 TITLE Failure of 76 of 185 Control Rods to Fully Insert During a Scram at a BWR REV NA ADDITIONAL INFORMATION Boiling Water Reactor IR 50-390/85-60 and 50-391/85-49 (Unit 1)Unit 2 Action: Implement design and procedure changes.B 80-018 Maintenance of Adequate Minimum Flow Thru Centrifugal Charging Pumps Following Secondary Side High Energy Rupture CO 06 REVISION 06 UPDATE: B 80-019 Mercury-Wetted Matrix Relay in Reactor Protective Systems of Operating Nuclear Power Plants Designed by CE B 80-020 Failure of Westinghouse Type W-2 Spring Return to Neutral Control Switches C Cl 06 NRC Inspection Report 391/2011-604 closed B 80-018.TVA: letter dated September 4, 1980 NRC: NUREG/CR 4933; IR 390/391 81-17 Unit 2 Action: Modify switches.REVISION 06 UPDATE: The switches were modified.B 80-021 Valve Yokes Supplied by Malcolm Foundry Co., Inc.NRC Inspection Report 391/2011-604 closed B 80-020.C TVA: letter dated May 6, 1981 B 80-022 Automation Industries, Model 200-520-008 Sealed-Source Connectors B 80-023 Failures of Solenoid Valves Manufactured by Valcor Engineering Corporation NA NRC: 390/391 85-08 Does not apply to power reactor.C TVA: letter dated March 31, 1981 NRC: IR 390/391 81-17; NUREG/CR 5292 Page 13 of 109* = See last page for status code definition.

ITEM B 80-024 TITLE Prevention of Damage Due to Water Leakage Inside Containment (10/17/80 Indian Point 2 Event)REV Cl Unit 2 Action: ADDITIONAL INFORMATION 06 Confirm that the reactor cavity can not be flooded, resulting in the partial or total submergence of the reactor vessel unnoticed by the reactor operators.

REVISION 06 UPDATE: It was confirmed that the reactor cavity can not be flooded, resulting in the partial or total submergence of the reactor vessel unnoticed by the reactor operators.

NA Boiling Water Reactor B 80-025 Operating Problems with Target Rock Safety-Relief Valves at BWRs B 81-001 Surveillance of Mechanical Snubbers B 81-002 Failure of Gate Type Valves to Close Against Differential Pressure B 81-003 Flow Blockage of Cooling Water to Safety System Components by Asiatic Clams and Mussels B 82-001 Alteration of Radiographs of Welds in Piping Subassemblies B 82-002 Degradation of Threaded Fasteners in the Reactor Coolant Pressure Boundary of PWR Plants NA C NRC: IR 390/391 81-17 TVA: letter dated September 30, 1983 NRC: IR 390/391 84-03 C TVA: letters dated July 21, 1981 and March 21, 1983 NRC: IR 390/391 81-17 C NRC: IR 390/391 85-08 CI TVA: memo dated February 6, 1985 06 NRC: IR 390/391 85-08 Approach accepted in IR 50-390/85-08 and 50-391/85-08 (March 29, 1985).Unit 2 Action: Implement same approach as Unit 1.REVISION 06 UPDATE: B 82-003 Stress Corrosion Cracking in Thick-Wall, Large Diameter, Stainless Steel, Recirculation System Piping at BWR Plants NA The boric acid corrosion program applies to both units.Boiling Water Reactor Page 14 of 109* = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 82-004 Deficiencies in Primary Containment Electrical Penetration Assemblies B 83-001 Failure of Trip Breakers (Westinghouse DB-50) to Open on Automatic Trip Signal B 83-002 Stress Corrosion Cracking in Large-Diameter Stainless Steel Recirculation System Piping at BWR Plants B 83-003 Check Valve Failures in Raw Water Cooling Systems of Diesel Generators B 83-004 Failure of the Undervoltage Trip Function of Reactor Trip Breakers C TVA: letter dated January 24, 1983 NRC: IR 390/83-10 and 391/83-08 C NRC: IRs 390/391 85-08 and 390/391 92-13 NA Boiling Water Reactor NA C 06 Addressed by Inservice Testing for Construction Permit holders NRC: IR 390/391 85-08 Unit 2 Action: Install new undervoltage attachment with wider grooves on the reactor trip breakers.....................................................................................................

REVISION 06 UPDATE: New breakers have been installed on Unit 2.B 83-005 ASME Nuclear Code Pumps and C Spare Parts Manufactured by the Hayward Tyler Pump Company B 83-006 Nonconforming Material Supplied CI by Tube-Line Facilities 04 NRC Inspection Report 391/2011-602 closed B 83-004.TVA: letter dated September 7, 1983 NRC: IR 390/85-03 and 391/85-04; NUREG/CR 5297 TVA: letter dated February 2, 1984 NRC: IR 390/391 84-03; NUREG/CR 4934 NRC SER for both units dated September 23, 1991, provided an alternate acceptance for fittings supplied by Tube-Line.

Unit 2 Action: Implement as necessary.

REVISION 04 UPDATE: NRC Inspection Report Nos. 50-390/90-02 and 50-391/90-02 found the Page 15 of 109* = See last page for status code definition.

REV ITEM TITLE ADDITIONAL INFORMATION proposed alternative to ASME code paragraph NA-3451 (a) to be acceptable.

It noted that TVA must revise the FSAR to document this deviation from ASME Section III requirements.

TVA letter to NRC dated October 11, 2007, stated the Unit 1 exemption is applicable to Unit 2 and was submitted to the NRC as being required for Unit 2 construction.

Final action was to incorporate the exemption in the Unit 2 FSAR. This exemption is documented in Unit 2 FSAR Section 3.2 in paragraph 3.2.3.2 and Table 3.2-2a as explained in Note 4. of the table.B 83-007 Apparently Fraudulent Products Sold by Ray Miller, Inc.B 83-008 Electrical Circuit Breakers With an Undervoltage Trip Feature in Safety-Related Applications Other Than the Reactor Trip System B 84-001 Cracks in BWR Mark 1 Containment Vent Headers B 84-002 Failure of GE Type HFA Relays In Use In Class 1 E Safety Systems B 84-003 Refueling Cavity Water Seal C TVA: letter dated March 22, 1984 NRC: IR 390/85-03 and 391/85-04 C TVA: letter dated March 29, 1984 NRC: IR 390/84-35 and 391/84-33 NA Boiling Water Reactor C TVA: letter dated July 10, 1984 NRC: IR 390/391 84-42 and IR 390/84-77 and 391/84-54 Cl Reviewed in IR 390/93-11.

Unit 2 Action: Ensure appropriate abnormal operating instructions (AOIs) are used for Unit 2.B 85-001 Steam Binding of Auxiliary Feedwater Pumps Cl TVA: letter dated January 27, 1986 NRC: IR 390/391 90-20 NRC accepted approach in letter dated July 20, 1988, and reviewed response in Appendix EE of SSER16.Unit 2 Action: Procedures and hardware will be in place to ensure recognition of indications of steam binding and maintenance of system operability until check valves are repaired and back leakage stopped.Page 16 of 109* = See last page for status code definition.

ITEM B 85-002 TITLE Undervoltage Trip Attachment of Westinghouse DB-50 Type Reactor Trip Breakers REV C 06 ADDITIONAL INFORMATION Unit 2 Action: Install automatic shunt trip on the Westinghouse DS-416 reactor trip breakers on Unit 2.REVISION 06 UPDATE: New breakers (including an automatic shunt trip) have been installed on Unit 2.NRC Inspection Report 391/2011-602 closed B 85-002.C Superseded by GL 89-10 B 85-003 Motor-Operated Valve Common Mode Failures During Plant Transients Due to Improper Switch Settings B 86-001 Minimum Flow Logic Problems That Could Disable RHR Pumps B 86-002 Static "0" Ring Differential Pressure Switches B 86-003 Potential Failure of Multiple ECCS Pumps Due to Single Failure of Air-Operated Valve in Minimum Flow Recirculation Line B 86-004 Defective Teletherapy Timer That May Not Terminate Treatment Dose B 87-001 Thinning of Pipe Walls in Nuclear Power Plants NA C Boiling Water Reactor TVA: letter dated November 20, 1986 NRC: IR 390/391/90-24 C TVA: letter dated November 14, 1986 NRC: IR 390/391/87-03 NA Does not apply to power reactor.C TVA: letter dated September 18, 1987 NRC: NUREG/CR 5287 Closed to GL 89-08 Page 17 of 109* = See last page for status code definition.

ITEM B 87-002 TITLE Fastener Testing to Determine Conformance with Applicable Material Specifications REV CI 03 ADDITIONAL INFORMATION TVA: letters dated April 15, 1988, July 6, 1988, September 12, 1988, and January 27, 1989 NRC: letter dated August 18, 1989 NRC closed in letter dated August 18, 1989.Unit 2 Action: Complete for Unit 2, using information used for Unit 1, as applicable.

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REVISION 03 UPDATE: B 88-001 Defects in Westinghouse Circuit Breakers B 88-002 Rapidly Propagating Fatigue Cracks in Steam Generator Tubes C CI Unit 2 has completed fastener testing as required by this Bulletin.TVA: letter dated November 15, 1991 NRC: IR 390/391 93-01 NRC acceptance letter dated June 7, 1990, for both units.Unit 2 Actions:* Evaluate E/C data to determine anti-vibration bar penetration depth;* perform T/H analysis to identify susceptible tubes;* modify, if necessary.

TVA: letter dated April 13, 1992 NRC: IR 390/391 92-13 B 88-003 Inadequate Latch Engagement in HFA Type Latching Relays*Manufactured by General Electric (GE) Company B 88-004 Potential Safety-Related Pump Loss C CI NRC acceptance letter dated May 24, 1990, for both units.Unit 2 Actions:* Perform calculations, and B 88-005 Nonconforming Materials Supplied by Piping Supplies, Inc. and West Jersey Manufacturing Company.CI* install check valves to prevent pump to pump interaction.

NRC reviewed in Appendix EE of SSER16.Unit 2 Actions:* Complete review to locate installed WJM material, and B 88-006 Actions to be Taken for the Transfer of Model No. SPEC 2-T Radiographic Exposure Device NA* perform in-situ hardness testing for Unit 2.Does not apply to power reactor.Page 18 of 109* = See last page for status code definition.

ITEM B 88-007 TITLE*REV NA ADDITIONAL INFORMATION Power Oscillations in BWRs Boiling Water Reactor B 88-008 Thermal Stresses in Piping Connected to Reactor Cooling Systems B 88-009 Thimble Tube Thinning in Westinghouse Reactors Cl NRC acceptance letter dated September 19, 1991, for both units.Unit 2 Action: Implement program to prevent thermal stratification.

Cl Reviewed in Appendix EE of SSER1 6.06 Unit 2 Action: TVA letter dated March 11, 1994, for both units committed to establish a program and inspect the thimble tubes during the first refueling outage.REVISION 06 UPDATE: Unit 2 is installing the Westinghouse In-core, Information, Surveillance, and Engineering (WINCISE) system. Westinghouse has analyzed WINCISE to exhibit essentially no wear due to vibrations, and should there be a breach of the thimble tube there would not be a loss of into the seal table room, Therefore, the thimble tubes for WINCISE do not need eddy current testing.B 88-010 Nonconforming Molded-Case Cl Unit 2 Action: Replace those circuits not traceable to a circuit breaker Circuit Breakers manufacturer.

B 88-011 Pressurizer Surge Line Thermal Cl NRC SER on "Leak-Before-Break" (April 28, 1993) and reviewed in Stratification Appendix EE of SSER16.B 89-001 Failure of Westinghouse Steam C Generator Tube Mechanical Plugs -06 Unit 2 Actions:* Complete modifications to accommodate Surge Line thermal movements, and* incorporate a temperature limitation during heatup and cooldown operations into Unit 2 procedures.

NRC acceptance letter dated September 26, 1991 for both units.Unit 2 Action: Remove SG tube plugs.REVISION 06 UPDATE: The SG tube plugs were removed.NRC Inspection Report 391/2011-602 closed B 89-001.Page 19 of 109* = See last page for status code definition.

ITEM B 89-002 TITLE Stress Corrosion Cracking of High-Hardness Type 410 Stainless Steel Internal Preloaded Bolting in Anchor Darling Model S350W Swing Check Valves or Valves of Similar Nature REV Cl 06 ADDITIONAL INFORMATION NRC reviewed in Appendix EE of SSER1 6.Unit 2 Actions:* Replace the flapper assembly hold-down bolts fabricated on the 14 (12 valves are installed)

Atwood and Morrell Mark No. 47W450-53 check valves.* Replacement bolts are to be fabricated from ASTM F593 Alloy 630.* A review of the remaining Unit 2 safety related swing check valves will be performed.

REVISION 06 UPDATE:* Bolts fabricated from ASTM F593 Alloy 630 have been procured.* The review of the remaining Unit 2 safety related swing check valves was completed.

Needed corrective actions were initiated.

TVA: letter dated June 19, 1990 NRC: IR 390/391 94-04 and letter dated June 22, 1990 NRC acceptance letter dated June 22, 1990.Unit 2 Action: Ensure that requirements for fuel assembly configuration, fuel loading and training are included in Unit 2.Unit 2 Action: Implement applicable recommendations from this Bulletin including identification of potentially defective transmitters and an enhanced surveillance program which monitors transmitters for loss of fill oil.B 89-003 Potential Loss of Required CI Shutdown Margin During Refueling Operations B 90-001 Loss of Fill-Oil in Transmitters Co Manufactured by Rosemount 06 B 90-002 Loss of Thermal Margin Caused NA by Channel Box Bow B 91-001 Reporting Loss of Criticality Safety NA Controls REVISION 06 UPDATE: NRC Inspection Report 391/2011-603 closed B 90-001.Boiling Water Reactor Does not apply to power reactor.Page 20 of 109* = See last page for status code definition.

ITEM B 92-001 TITLE Failure of Thermo-Lag 330 Fire Barrier System to Maintain Cabling in Wide Cable Trays and Small Conduits Free From Fire Damage REV ADI NA 02 REVISION 02 UPDATE: DITIONAL INFORMATION B 92-002 Safety Concerns Related to "End of Life" of Aging Theratronics Teletherapy Units B 92-003 Release of Patients After Brachytherapy B 93-001 Release of Patients After Brachytherapy Treatment with Remote Afterloading Devices B 93-002 Debris Plugging of Emergency Core Cooling Suction Strainers This bulletin was provided for information only to plants with construction permits. See Generic Letter 92-08 for Thermo-lag related actions.NA Does not apply to power reactor.NA Does not apply to power reactor.NA C 02 Does not apply to power reactor.Boiling Water Reactor-------------------------------------------------------------------------------------------------

B 93-003 Resolution of Issues Related to Reactor Vessel Water Level Instrumentation in BWRs B 94-001 Potential Fuel Pool Draindown Caused by Inadequate Maintenance Practices at Dresden Unit 1 B 94-002 Corrosion Problems in Certain Stainless Steel Packagings Used to Transport Uranium Hexafluoride B 95-001 Quality Assurance Program for Transportation of Radioactive Material B 95-002 Unexpected Clogging of a Residual Heat Removal Pump Strainer While Operating in Suppression Pool Cooling Mode NA NA NA NA NA REVISION 02 UPDATE: In Rev. 01, this was characterized as "NA -BWR only". This Bulletin was provided for Information to holders of construction permits. No WBN response was found.B-93-02 was closed in IR 50-390/94-04 and 50-391/94-04.

Boiling Water Reactor Addressed to holders of licenses for nuclear power reactors that are permanently shut down with spent fuel in the spent fuel pool Does not apply to power reactor.Does not apply to power reactor.Boiling Water Reactor Page 21 of 109* = See last page for status code definition.

ITEM B 96-001, first part TITLE Control Rod Insertion Problems (PWR)REV Cl 04 ADDITIONAL INFORMATION NRC acceptance letter for Unit 1 dated July 22, 1996 -Initial response for Unit 2 on September 7, 2007.Unit 2 Action: Issue Emergency Operating Procedure.

REVISION 02 UPDATE: Unit 2 will load all new RFA-2 fuel for the initial fuel load.REVISION 03 UPDATE: NRC issued the Safety Evaluation (corrected) for Bulletin 1996-001 on May 3, 2010.....................................................................................................

....................................................................................................

REVISION 04 UPDATE: Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.NRC acceptance letter for Unit 1 dated July 22, 1996 -Initial response for Unit 2 on September 7, 2007.Unit 2 Action: and provide core map.....................................................................................................

REVISION 03 UPDATE: NRC issued the Safety Evaluation (corrected) for Bulletin 1996-001 on May 3, 2010.B 96-001, Control Rod Insertion Problems CI last part (PWR)06 REVISION 04 UPDATE: Corrected status from "OVW to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.REVISION 06 UPDATE: SSER22 contained the following for NRC Action: "Closed. NRC letter dated May 3,2010 (ADAMS Accession No.ML101200035) required Confirmatory Action (See Appendix HH)" The applicable item from SER22, Appendix HH for this item is Open Page 22 of 109* o= See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION Item 5, "Verify timely submittal of pre-startup core map and perform technical review. (TVA letter dated September 7, 2007, ADAMS Accession No. ML072570676)." TVA to NRC letter dated April 6, 2011 provided the following response to Open Item 5: "Attachment 1 provides the requested core map." NRC closure letter dated May 20, 1998.Unit 2 Action: Unit 2 Heavy Loads Program will be in compliance with NUREG-0612.

B 96-002 Movement of Heavy Loads over Spent Fuel, Over Fuel in the Reactor, or Over Safety-Related Equipment Cl 06 REVISION 02 UPDATE: NRC issued the Safety Evaluation for Bulletin 1996-002 on March 4, 2010.REVISION 06 UPDATE: SSER22 contained the following for NRC Action: "Closed. NRC letter dated March 4, 2010 (ADAMS Accession No.ML100480062)" Boiling Water Reactor B 96-003 Potential Plugging of ECCS Suction Strainers by Debris in BWRs B 96-004 Chemical, Galvanic, or Other Reactions in Spent Fuel Storage and Transportation Casks B 97-001 Potential for Erroneous Calibration, Dose Rate, or Radiation Exposure Measurements with Certain Victoreen Model 530 and 531SI Electrometer/Dosemeters B 97-002 Puncture Testing of Shipping Packages Under 10 CFR Part 71 NA NA Info NA NA Does not apply to power reactor.Does not apply to power reactor.Page 23 of 109* = See last page for status code definition.

ITEM B 01-001 TITLE Circumferential Cracking of Reactor Pressure Vessel (RPV)Head Penetration Nozzles REV ADDITIONAL INFORMATION C NRC acceptance letter dated November 20, 2001 (Unit 1) -Initial--- response for Unit 2 on September 7, 2007.06 Unit 2 Action: Perform baseline inspection.

REVISION 02 UPDATE: Unit 2 Actions:* Perform baseline inspection.

Evaluate or repair as necessary.

REVISION 03 UPDATE: NRC issued the Safety Evaluation for Bulletin 2001-001 on June 30, 2010.REVISION 04 UPDATE: Corrected status from "OV" to "Cl" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.----------------------------------------------------------------------------------------------------

REVISION 06 UPDATE: The baseline inspection was performed with evaluations and repairs as necessary.

SSER22 contained the following for NRC Action: "Closed. See NRC Letter dated June 30, 2010 (ADAMS Accession No.ML 100539515)" NRC Inspection Report 391/2011-602 closed B 01-001.Page 24 of 109* = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 02-001 RPV Head Degradation and C NRC review of Unit 1's 15 day response in letter dated May 20, 2002 -Reactor Coolant Pressure -Initial response for Unit 2 on September 7, 2007.Boundary Integrity 06 Unit 2 Action: Perform baseline inspection.

REVISION 02 UPDATE: Unit 2 Actions:* Perform baseline inspection.

Evaluate or repair as necessary.

REVISION 03 UPDATE: NRC issued the Safety Evaluation for Bulletin 2002-001 on June 30, 2010.REVISION 04 UPDATE: Corrected status from "OV" to "Cl" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.REVISION 06 UPDATE: The baseline inspection was performed with evaluations and repairs as necessary.

SSSER22 contained the following for NRC Action: "Closed. See NRC Letter dated June 30, 2010 (ADAMS Accession No.ML 100539515)" NRC Inspection Report 391/2011-602 closed B 02-001.Page 25 of 109* = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 02-002 RPV Head and Vessel Head C NRC acceptance letter dated December 20, 2002 (Unit 1) -Initial Penetration Nozzle Inspection

-response for Unit 2 on September 7, 2007.Programs 06 Unit 2 Action: Perform baseline inspection.

....................................................................................................

REVISION 02 UPDATE: Unit 2 Actions:* Perform baseline inspection.

Evaluate or repair as necessary.

....................................................................................................

REVISION 03 UPDATE: NRC issued the Safety Evaluation for Bulletin 2002-002 on June 30, 2010.REVISION 04 UPDATE: Corrected status from "OV" to "Cl" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.REVISION 06 UPDATE: The baseline inspection was performed with evaluations and repairs as necessary.

SSSER22 contained the following for NRC Action: "Closed. See NRC Letter dated June 30, 2010 (ADAMS Accession No.ML 100539515)" NRC Inspection Report 391/2011-602 closed B 02-002.B 03-001 Potential Impact of Debris NA TVA: letter dated September 7, 2007 Blockage on Emergency Sump Recirculation at PWRs Page 26 of 109* = See last page for status code definition.

ITEM B 03-002 TITLE Leakage from RPV Lower Head Penetrations and Reactor Coolant Pressure Boundary Integrity (PWRs)REV ADDITIONAL INFORMATION Cl NRC acceptance letter dated October 6, 2004 (Unit 1) -Initial response--for Unit 2 on September 7, 2007.06 Unit 2 Action: Perform baseline inspection.

REVISION 02 UPDATE: NRC issued the Safety Evaluation for Bulletin 2003-002 on January 21, 2010.Unit 2 Actions:* Perform baseline inspection.

Evaluate or repair as necessary.

........................................................................................

REVISION 06 UPDATE: SSER22 contained the following for NRC Action: "Closed. NRC Letter dated January 21, 2010 (ADAMS Accession No.ML093631061)" Does not apply to power reactor.TVA: letter dated December 18, 2003 Item concerns a multi-unit issue that was completed for both units.Initial response for Unit 2 on September 7, 2007.Unit 2 Actions:* Provide details of pressurizer and penetrations, and* apply Material Stress Improvement Process.B 03-003 B 03-004 B 04-001 Potentially Deficient 1-inch Valves NA for Uranium Hexaflouride Cylinders

_ _.Rebaselining of Data in the C Nuclear Management and Safeguards System Inspection of Alloy 82/182/600 Materials Used in the Fabrication of Pressurizer Penetrations and Steam Space Piping Connections at PWRs Cl 06 REVISION 02 UPDATE: TVA provided details of the pressurizer and penetrations on September 29, 2008. This letter committed to: Prior to placing the pressurizer in service, TVA will apply the Material Stress Improvement Process (MSIP) to the Pressurizer Power Operated Relief Valve connections, the safety relief valve connections, the spray line nozzle and surge line nozzle connections.

Page 27 of 109* = See last page for status code definition.

REV ITEM TITLE ADDITIONAL INFORMATION TVA will perform a bare metal visual (BMV) inspection of the upper pressurizer Alloy 600 locations at the first refueling outage.REVISION 03 UPDATE: April 1, 2010, letter committed to: TVA will perform NDE prior to and after performance of the MSIP. If circumferential cracking is observed in either pressure boundary or non-pressure boundary portions of any locations covered under the scope of the bulletin, TVA will develop plans to perform an adequate extent-of-condition evaluation, and TVA will discuss those plans with cognizant NRC technical staff prior to starting Unit 2.After performing the BMV inspection during the first refueling outage, if any evidence of apparent reactor coolant pressure boundary leakage is discovered, then NDE capable of determining crack orientation will be performed in order to accurately characterize the flaw, the orientation, and extent. TVA will develop plans to perform an adequate extent of condition evaluation, and plans to possibly expand the scope of NDE to other components in the pressurizer will be discussed with NRC technical staff prior to restarting of Unit 2.REVISION 04 UPDATE: NRC issued the Safety Evaluation for Bulletin 2004-001 on August 4, 2010.--------------------------------------------------------------------------------------------------

B 05-001 Material Control and Accounting at Reactors and Wet Spent Fuel Storage Facilities REVISION 06 UPDATE: SSER22 contained the following for NRC Action: "Closed. NRC Letter dated August 4, 2010 (ADAMS Accession No.ML102080017)" C TVA: letters dated March 21, 2005 and May 11, 2005 Item concerns a multi-unit issue that was completed for both units.C TVA: letters dated January 20, 2006 and August 16, 2006.Item concerns a multi-unit issue that was completed for both units.B 05-002 Emergency Preparedness and Response Actions for Security-Based Events Page 28 of 109* = See last page for status code definition.

ITEM B 07-001 TITLE REV C ADDITIONAL INFORMATION Item concerns a multi-unit issue that was completed for both units.Security Officer Attentiveness 06 REVISION 05 UPDATE: The NRC closed this bulletin via letter dated March 25, 2010 (ADAMS Accession No. ML100770549).

REVISION 06 UPDATE: SSER22 contained the following for NRC Action: C 76-001 Crane Hoist Control Circuit Modifications C 76-002 Relay Failures -Westinghouse BF (AC) and BFD (DC) Relays C 76-003 Radiation Exposures in Reactor Cavities C 76-004 Neutron Monitor and Flow Bypass Switch Malfunctions C 76-005 Hydraulic Shock And Sway Suppressors

-Maintenance of Bleed and Lock-Up Velocities on ITT Grinnell's Model Nos. -Fig. 200 And Fig. 201, Catalog Ph-74-R C 76-006 Stress Corrosion Cracks in Stagnant, Low Pressure Stainless Piping Containing Boric Acid Solution at PWRs C 76-007 Inadequate Performance by.Reactor Operating and Support Staff Members C 77-001 Malfunctions of Limitorque Valve Operators"Closed. NRC Letter dated March 25, 2010 (ADAMS Accession No.ML 100770549)" C See B 76-007 for additional information.

C TVA: letter dated November 22, 1976 informed NRC that these relay types'are not used in Class I E circuits.NRC: IR 50/390/76-11 and 50/391/76-11 NA Info NA Boiling Water Reactor C TVA: letter dated January 7, 1977 informed NRC that no Grinnell shock suppressors or sway braces have been or will be installed at WBN.NA Item was applicable only to units with operating license at the time the---item was issued.NA Item was applicable only to units with operating license at the time the item was issued.NA Info Page 29 of 109* = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION C 77-002a Potential Heavy Spring Flooding NA Item was applicable only to units with operating license at the time the (CP) _ item was issued.C 77-003 Fire Inside a Motor Control Center NA Info C 77-004 Inadequate Lock Assemblies NA Info C 77-005 Fluid Entrapment in Valve Bonnets NA Info C 77-006 Effects of Hydraulic Fluid on NA Info Electrical Cables C 77-007 Short Period During Reactor NA Boiling Water Reactor Startup --C 77-008 Failure of Feedwater Sample NA Item was applicable only to units with operating license at the time the Probe --item was issued.C 77-009 Improper Fuse Coordination in NA Boiling Water Reactor BWR Standby Liquid Control System Control Circuits C 77-010 Vacuum Conditions Resulting in NA Item was applicable only to units with operating license at the time the Damage to Liquid Process Tanks _ -. item was issued.C 77-011 Leakage of Containment Isolation NA Info Valves with Resilient Seats C 77-012 Dropped Fuel Assemblies at BWR NA Boiling Water Reactor Facilities C 77-013 Reactor Safety Signals Negated NA Info During Testing _ -.C 77-014 Separation of Contaminated Water NA Info Systems from Noncontaminated Plant Systems C 77-015 Degradation of Fuel Oil Flow to the NA Info Emergency Diesel Generator C 77-016 Emergency Diesel Generator NA Info Electrical Trip Lock-Out Features _ -.C 78-001 Loss of Well Logging Source NA Does not apply to power reactor.Page 30 of 109* = See last page for status code definition.

ITEM TITLE C 78-002 Proper Lubricating Oil for Terry Turbines C 78-003 Packaging Greater Than Type A Quantities of Low Specific Activity Radioactive Material for Transport C 78-004 Installation Errors That Could Prevent Closing of Fire Doors C 78-005 Inadvertent Safety Injection During Cooldown C 78-006 Potential Common Mode Flooding of ECCS Equipment Rooms at BWR Facilities C 78-007 Damaged Components of a Bergen-Paterson Series 25000 Hydraulic Test Stand C 78-008 Environmental Qualification of Safety-Related Electrical Equipment at Nuclear Power Plants C 78-009 Arcing of General Electric Company Size 2 Contactors C 78-010 Control of Sealed Sources in Radiation Therapy C 78-011 Recirculation MG Set Overspeed Stops C 78-012 HPCI Turbine Control Valve Lift Rod Bending C 78-013 Inoperability of Service Water Pumps C 78-014 HPCI Turbine Reversing Chamber Hold Down Bolting C 78-015 Tilting Disc Check Valves Fail to Close with Gravity in Vertical Position C 78-016 Limitorque Valve Actuators REV ADDITIOIN NA Info NA Info NA Info NA Info NA Info NA Info NA Info NA Info NA Does not apply to power reactor.lAL INFORMATION NA NA Boiling Water Reactor Boiling Water Reactor NA Info NA Boiling Water Reactor NA Info NA Info Page 31 of 109* = See last page for status code definition.

REV ITEM TITLE ADDITIONAL INFORMATION C 78-017 Inadequate Guard Training/Qualification and Falsified Training Records C 78-018 UL Fire Test C 78-019 Manual Override (Bypass) of Safety System Actuation Signals C 79-001 Administration of Unauthorized Byproduct Material to Humans C 79-002 Failure of 120 Volt Vital AC Power Supplies C 79-003 Inadequate Guard Training -Qualification and Falsified Training Records C 79-004 Loose Locking Nut on Limitorque Valve Operators C 79-005 Moisture Leakage in Stranded Wire Conductors C 79-006 Failure to Use Syringe and Bottle Shields in Nuclear Medicine C 79-007 Unexpected Speed Increase of Reactor Recirculation MG Set Resulted in Reactor Power Increase C 79-008 Attempted Extortion

-Low Enriched Uranium C 79-009 Occurrences of Split or Punctured Regulator Diaphragms in Certain Self Contained Breathing Apparatus C 79-010 Pipefittings Manufactured from Unacceptable Material C 79-011 Design/Construction Interface Problem NA Info- -----NA Info NA Info NA NA Does not apply to power reactor.Info NA Info NA Info NA Info NA NA Does not apply to power reactor.Boiling Water Reactor NA Fuel facilities and operating reactors at the time the item was issued NA Info NA Info NA Info Page 32 of 109* = See last page for status code definition.

REV ITEM TITLE ADDITIONAL INFORMATION C 79-012 Potential Diesel Generator Turbocharger Problem C 79-013 Replacement of Diesel Fire Pump Starting Contactors C 79-014 Unauthorized Procurement and Distribution of XE-133 C 79-015 Bursting of High Pressure Hose and Malfunction of Relief Valve 0-Ring in Certain Self-Contained Breathing Apparatus C 79-016 Excessive Radiation Exposures to Members of the General Public and a Radiographer C 79-017 Contact Problem in SB-12 Switches on General Electric Company Metalclad Circuit Breakers C 79-018 Proper Installation of Target Rock Safety-Relief Valves C 79-019 Loose Locking Devices on Ingersoll-Rand Pumps C 79-020 Failure of GTE Sylvania Relay Type PM Bulletin 7305 Catalog 5U12-1 1-AC with a 120V AC Coil C 79-021 Prevention of Unplanned Releases of Radioactivity C 79-022 Stroke Times for Power Operated Relief Valves NA Info NA NA NA NA Info Does not apply to power reactor.Item was applicable only to units with operating license at the time the item was issued.Does not apply to power reactor.NA Info NA Boiling Water Reactor NA Info NA Info NA Info NA Info C 79-023 Motor Starters and Contactors Failed to Operate C The Circular did not require a response.01 TVA reported a nonconformance under 10 CFR 50.55e on January 17, 1980, that four motor starters of this type had been located in the 480V control and auxiliary vent boards at WBN. Gould factory representatives supervised the replacement of the carrier assemblies in accordance with the Gould instructions.

The starters with replaced carriers were acceptable.

NRC IR 50-390/80-03 and 50-391/80-02 reviewed and closed the associated nonconformance reports.Page 33 of 109* = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION C 79-024 Proper Installation and Calibration of Core Spray Pipe Break Detection Equipment on BWRs C 79-025 Shock Arrestor Strut Assembly Interference NA C 01 Boiling Water Reactor The Circular did not require a response.TVA reported a nonconformance under 10 CFR 50.55e on March 6, 1980, that a review had determined that nine installed supports had brackets with the potential of hindering full function of the support.Additional supports that were not installed had the same potential problem. TVA initially determined that the supports would be modified in accordance with a vendor approved drawing. TVA subsequently determined that no actual problem existed and no field work was required.NRC IR 50-390/83-15 and 50-391/83-11 reviewed and closed the associated nonconformance reports.NA Info C 80-001 Service Advice for GE Induction Disc Relays C 80-002 Nuclear Power Plant Staff Work Hours C 80-003 Protection from Toxic Gas Hazards C 80-004 Securing of Threaded Locking Devices on Safety-Related Equipment C 80-005 Emergency Diesel-Generator Lubricating Oil Addition and Onsite Supply C 80-006 Control and Accountability Systems for Implant Therapy Sources C 80-007 Problems with HPCI Turbine Oil System C 80-008 BWR Technical Specification Inconsistency

-RPS Response Time C 80-009 Problems with Plant Internal Communications Systems C 80-010 Failure to Maintain Environmental Qualification of Equipment NA NA-Info- ----Info NA Info NA Info NA NA NA NA NA Does not apply to power reactor.Boiling Water Reactor Boiling Water Reactor Info Info Page 34 of 109* = See last page for status code definition.

ITEM TITLE C 80-011 Emergency Diesel Generator Lube Oil Cooler Failures C 80-012 Valve-Shaft-to-Actuator Key May Fall Out of Place when Mounted Below Horizontal Axis C 80-013 Grid Strap Damage in Westinghouse Fuel Assemblies C 80-014 Radioactive Contamination of Plant Demineralized Water System and Resultant Internal Contamination of Personnel C 80-015 Loss of Reactor Coolant Pump Cooling and Natural Circulation Cooldown C 80-016 Operational Deficiencies in Rosemount Model 510DU Trip Units and Model 1152 Pressure Transmitters C 80-017 Fuel Pin Damage Due to Water Jet from Baffle Plate Corner C 80-018 10 CFR 50.59 Safety Evaluations for Changes to Radioactive Waste Treatment Systems C 80-019 Noncompliance with License Requirements for Medical Licensees C 80-020 Changes in Safe-Slab Tank Dimensions C 80-021 Regulation of Refueling Crews REV NA Info NA Info NA Info NA Info NA Info NA Info NA Info NA Info ADDITIONAL INFORMATION NA NA NA NA Does not apply to power reactor.Info Item was applicable only to units with operating license at the time the item was issued.Info C 80-022 Confirmation of Employee Qualifications C 80-023 Potential Defects in Beloit Power Systems Emergency Generators C 80-024 AECL Teletherapy Unit Malfunction NA Info NA Does not apply to power reactor.Page 35 of 109* = See last page for status code definition.

ITEM C 80-025 TITLE REV NA ADDITIONAL INFORMATION Case Histories of Radiography Events Does not apply to power reactor.C 81-001 Design Problems Involving Indicating Pushbutton Switches Manufactured by Honeywell Incorporated C 81-002 Performance of NRC-Licensed Individuals while on Duty C 81-003 Inoperable Seismic Monitoring Instrumentation C 81-004 The Role of Shift Technical Advisors and Importance of Reporting Operational Events C 81-005 Self-Aligning Rod End Bushings for Pipe Supports C 81-006 Potential Deficiency Affecting Certain Foxboro 10 to 50 Milliampere Transmitters C 81-007 Control of Radioactively Contaminated Material C 81-008 Foundation Materials C 81-009 Containment Effluent Water that Bypasses Radioactivity Monitor C 81-010 Steam Voiding in the Reactor Coolant System During Decay Heat Removal Cooldown C 81-011 Inadequate Decay Heat Removal During Reactor Shutdown C 81-012 Inadequate Periodic Test Procedure of PWR Reactor Protection System NA Info NA Item was applicable only to units with operating license at the time the---item was issued.NA Info NA Info NA Info NA Info NA Info NA Info NA Info NA Info NA Item was applicable only to units wi item was issued.ith operating license at the time the NA Boiling Water Reactor NA Info Page 36 of 109* = See last page for status code definition.

ITEM C 81-013 TITLE Torque Switch Electrical Bypass Circuit for Safeguard Service Valve Motors REV C ADDITIONAL INFORMATION The Circular did not require a response.01 TVA April founc bypa IssuE reported a nonconformance under 10 CFR 50.55e on 4, 1986 (NCR W367-P), that required closing torque switches were d improperly wired. This issue (Torque switch and overload relay ss capability for active safety related valves) is part of the Electrical ks Corrective Action Program for WBN Unit 2.C 81-014 Main Steam Isolation Valve NA Info Failures to Close C 81-015 Unnecessary Radiation Exposures NA Info to the Public and Workers During Events Involving Thickness and Level Measuring Devices GL 77-001 Intrusion Detection Systems NA Info Handbook GL 77-002 Fire Protection Functional NA Info Responsibilities GL 77-003 Transmittal of NUREG-0321, "A NA Info Study of the Nuclear Regulatory Commission Quality Assurance Program" GL 77-004 Shipments of Contaminated NA Info Components From NRC Licensed Power Facilities to Vendors &Service Companies GL 77-005 Nonconformity of Addressees of NA Info Items Directed to the Office of Nuclear Reactor Regulation GL 77-006 Enclosing Questionnaire Related NA Item to Steam Generators

---item GL 77-007 Reliability of Standby Diesel NA Item Generator Units ---item GL 77-008 Revised Intrusion Detection NA Info Handbook and Entry Control Systems Handbook GL 78-001 Correction to Letter of 12/15/77 NA Item[GL 77-07] -item GL 78-002 Asymmetric Loads Background C NRC and Revised Request for apprn Additional Information was applicable only to units with operating license at the time the was issued.was applicable only to units with operating license at the time the was issued.was applicable only to units with operating license at the time the was issued.Reviewed in SSER15 -Appendix C (June 1995). Resolved by val of leak-before-break analysis.Page 37 of 109.* = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 78-003 Request For Information on Cavity Annulus Seal Ring GL 78-004 GAO Blanket Clearance for Letter Dated 12/09/77 [GL 77-06]GL 78-005 Internal Distribution of Correspondence

-Asking for Comments on Mass Mailing System GL 78-006 This GL was never issued.GL 78-007 This GL was never issued.GL 78-008 Enclosing NUREG-0408 Re Mark I Containments, and Granting Exemption from GDC 50 and Enclosing Sample Notice GL 78-009 Multiple-Subsequent Actuations of Safety/Relief Valves Following an Isolation Event NA Item was applicable only to units with operating license at the time the--- item was issued.NA Item was applicable only to units with operating license at the time the---item was issued.NA Info NA NA NA Boiling Water Reactor NA Boiling Water Reactor GL 78-010 Guidance on Radiological NA Info Environmental Monitoring

_ -GL 78-011 Guidance on Spent Fuel Pool NA Info Modifications GL 78-012 Notice of Meeting Regarding NA Info"Implementation of 10 CFR 73.55 Requirements and Status of Research GL 78-013 Forwarding of NUREG-0219 NA Info GL 78-014 GL 78-015 GL 78-016 Transmittal of Draft NUREG-0219 NA Info for Comment Request for Information on Control NA See GL 81-007.of Heavy Loads Near Spent Fuel Request for Information on Control NA Info of Heavy Loads Near Spent Fuel Pools Page 38 of 109* = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 78-017 Corrected Letter on Heavy Loads NA Info Over Spent Fuel GL 78-018 Corrected Letter on Heavy Loads NA Dupli Over Spent Fuel GL 78-019 Enclosing Sandia Report SAND NA Info 77-0777, "Barrier Technology Handbook" GL 78-020 Enclosing

-"A Systematic NA Info Approach to the Conceptual Design of Physical Protection Systems for Nuclear Facilities GL 78-021 Transmitting NUREG/CR-0181, NA Info"Concerning Barrier and Penetration Data Needed for Physical Security System Assessment" GL 78-022 Revision to Intrusion Detection NA Info Systems and Entry Control Systems Handbooks and Nuclear Safeguards Technology Handbook GL 78-023 Manpower Requirements for NA Info Operating Reactors GL 78-024 Model Appendix I Technical NA Boilir Specifications and Submittal Schedule For BWRs GL 78-025 This GL was never issued. NA cate of GL 81-007 ng Water Reactor GL 78-026 Excessive Control Rod Guide Tube Wear GL 78-027 Forwarding of NUREG-0181 GL 78-028 Forwarding pages omitted from 07/11/78 letter [GL 78-24]GL 78-029 Notice of PWR Steam Generator Conference GL 78-030 Forwarding of NUREG-0219 NA Applies only to Babcock and Wilcox designed plants NA Info NA Boiling Water Reactor NA Info NA Info Page 39 of 109* = See last page for status code definition.

ITEM TITLE REV NA Info ADDITIONAL INFORMATION GL 78-031 Notice of Steam Generator Conference Agenda GL 78-032 Reactor Protection System Power Supplies GL 78-033 Meeting Schedule and Locations For Upgraded Guard Qualification GL 78-034 Reactor Vessel Atypical Weld Material GL 78-035 Regional Meetings to Discuss Upgraded Guard Qualifications GL 78-036 Cessation of Plutonium Shipments by Air Except In NRC Approved Containers GL 78-037 Revised Meeting Schedule &Locations For Upgraded Guard Qualifications GL 78-038 Forwarding of 2 Tables of Appendix I, Draft Radiological Effluent Technical Specifications, PWR, and NUREG-0133 GL 78-039 Forwarding of 2 Tables of Appendix I, Draft Radiological Effluent Technical Specifications, BWR, and NUREG-0133 GL 78-040 Training & Qualification Program Workshops GL 78-041 Mark II Generic Acceptance Criteria For Lead Plants GL 78-042 Training and Qualification Program Workshops GL 79-001 Interservice Procedures for Instructional Systems Development

-TRADOC GL 79-002 Transmitting Rev. to Entry Control Systems Handbook (SAND 77-1033), Intrusion Detection Handbook (SAND 76-0554), and Barrier Penetration Database NA NA Boiling Water Reactor Info C See B 78-12.NA NA Info Does not apply to power reactor.NA Info NA NA Item was applicable only to units with operating license at the time the item was issued.Boiling Water Reactor NA Info NA Boiling Water Reactor NA Info NA Info NA Info Page 40 of 109* = See last page for status code definition.

ITEM TITLE GL 79-003 Offsite Dose Calculation Manual GL 79-004 Referencing 4/14/78 Letter -Modifications to NRC Guidance"Review and Acceptance of Spent Fuel Pool Storage and Handling" GL 79-005 Information Relating to Categorization of Recent Regulatory Guides by the Regulatory Requirements Review Committee GL 79-006 Contents of the Offsite Dose Calculation Manual GL 79-007 Seismic (SSE) and LOCA Responses (NUREG-0484)

GL 79-008 Amendment to 10 CFR 73.55 GL 79-009 Staff Evaluation of Interim Multiple-Consecutive Safety-Relief Valve Actuations GL 79-010 Transmitting Regulatory Guide 2.6 for Comment REV ADDITIONAL INFORMATION NA Info NA Info NA Info NA Info NA Info NA Info NA Boiling Water Reactor NA Does not apply to power reactor.GL 79-011 Transmitting "Summary of NA Info Operating Experience with Recalculating Steam Generators, January 1979," NUREG-0523 GL 79-012 ATWS -Enclosing Letter to GE, NA Info with NUREG-0460, Vol. 3 GL 79-013 Schedule for Implementation and NA Info Resolution of Mark I Containment

_ -_Long Term Program GL 79-014 Pipe Crack Study Group -NA Info Enclosing NUREG-0531 and _ _Notice GL 79-015 Steam Generators

-Enclosing NA Info Summary of Operating Experience

_ -with Recirculating Steam Generators, NUREG-0523 Page 41 of 109* = See last page for status code definition.

REV ITEM TITLE ADDITIONAL INFORMATION GL 79-016 Meeting Re Implementation of NA Info Physical Security Requirements GL 79-017 Reliability of Onsite Diesel NA Info Generators at Light Water Reactors GL 79-018 Westinghouse Two-Loop NSSS NA Addressed to specific plant(s).GL 79-019 NRC Staff Review of Responses NA Addressed to specific plant(s).to Bs 79-06 and 79-06a GL 79-020 Cracking in Feedwater Lines C See B 79-13.GL 79-021 Enclosing NUREG/CR-0660, Enhancement of on Site Emergency Diesel Generator Reliability" GL 79-022 Enclosing NUREG-0560, "Staff Report on the Generic Assessment of Feedwater Transients in PWRs Designed by B&W" GL 79-023 NRC Staff Review of Responses to B 79-08 NA Info NA Applies only to Babcock and Wilcox designed plants NA Boiling Water Reactor GL 79-024 Multiple Equipment Failures in NA GL 79-24 provided a discussion of an inadvertent reactor scram and Safety-Related Systems _ -. safety injection during monthly surveillance tests of the safeguards system 01 at a PWR facility.

The GL requested a review to determine if similar errors had or could have occurred at other PWRs. The GL further requested a review of management policies and procedures to assure that multiple equipment failures in safety-related systems will be vigorously pursued and analyzed to identify significant reduction in the ability of safety systems to function as required.

A response was requested within 30 days of receipt of the GL with the results of these reviews. TVA does not have a record of receiving or responding to this GL. Thus, TVA concluded that this item was applicable only to PWRs with an operating license at the time the GL was issued.GL 79-025 Information Required to Review NA Info Corporate Capabilities

_ .GL 79-026 Upgraded Standard Technical NA Info Specification Bases Program GL 79-027 Operability Testing of Relief and NA Boiling Water Reactor Safety Relief Valves Page 42 of 109* = See last page for status code definition.

ITEM TITLE R REV ADDITIONAL INFORMATION GL 79-028 Evaluation of Semi-Scale Small NA Break Experiment GL 79-029 Transmitting NUREG-0473, NA Revision 2, Draft Radiological Effluent Technical Specifications GL 79-030 Transmitting NUREG-0472, NA Revision 2, Draft Radiological

-Technical Specifications GL 79-031 Submittal of Copies of Response NA to 6/29/79 NRC Request [79-25]GL 79-032 Transmitting NUREG-0578, NA"TMI-2 Lessons Learned" GL 79-033 Transmitting NUREG-0576, NA"Security Training and Qualification Plans" GL 79-034 New Physical Security Plans NA (FR 43280-285)

GL 79-035 Regional Meetings to Discuss NA Impacts on Emergency Planning GL 79-036 Adequacy of Station Electric Cl Distribution Systems Voltages Info Info Info Info Info Info Does not apply to power reactor.Info This GL tracked compliance with BTP PSB-1, "Adequacy of Station Electric Distribution System Voltages." Unit 2 Action: Perform verification during the preoperational testing.GL 79-037 Amendment to 10 CFR 73.55 NA Info Deferral from 8/1/79 to 11/1/79 GL 79-038 BWR Off-Gas Systems -NA Boiling Water Reactor Enclosing NUREG/CR-0727 GL 79-039 Transmitting Division 5 Draft NA Does not apply to power reactor.Regulatory Guide and Value Impact Statement GL 79-040 Follow-up Actions Resulting from NA Item was applicable only to units with operating license at the time the the NRC Staff Reviews Regarding item was issued.the TMI-2 Accident GL 79-041 Compliance with 40 CFR 190, NA Info EPA Uranium Fuel Cycle Standard Page 43 of 109* = See last page for status code definition.

ITEM TITLE GL 79-042 Potentially Unreviewed Safety Question on Interaction Between Non-Safety Grade Systems and Safety Grade Systems GL 79-043 Reactor Cavity Seal Ring Generic Issue GL 79-044 Referencing 6/29/79 Letter Re Multiple Equipment Failures GL 79-045 Transmittal of Reports Regarding Foreign Reactor Operating Experiences GL 79-046 Containment Purge and Venting During Normal Operation

-Guidelines for Valve Operability GL 79-047 Radiation Training REV NA NA NA NA NA ADDITIONAL INFORMATION Item was applicable only to units with operating license at the time the item was issued.Addressed to specific plant(s).Item was applicable only to units with operating license at the time the item was issued.Info Item was applicable only to units with operating license at the time the item was issued.NA Info GL 79-048 Confirmatory Requirements Relating to Condensation Oscillation Loads for the Mark I Containment Long Term Program GL 79-049 Summary of Meetings Held on 9/18-20/79 to Discuss Potential Unreviewed Safety Question on Systems Interaction for B&W PI GL 79-050 Emergency Plans Submittal Dates GL 79-051 Follow-up Actions Resulting from the NRC Staff Reviews Regarding the TMI-2 Accident GL 79-052 Radioactive Release at North Anna Unit 1 and Lessons Learned GL 79-053 ATWS GL 79-054 Containment Purging and Venting During Normal Operation NA Boiling Water Reactor NA Info NA Info NA GL 79-51 provided follow-up actions resulting from the Three Mile Island--Unit 2 accident.

GL 79-51 was provided for planning and guidance 01 purposes.

Its principal element was a report titled 'TMMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations" (NUREG-0573).

This GL and the NUREG were superseded by GL 80-90 and NUREG-0737.

See GL 80-90 for further information.

NA Item was applicable only to units with operating license at the time the---item was issued.NA Info NA Addressed to specific plant(s).Page 44 of 109* = See last page for status code definition.

ITEM TITLE REV NA Info ADDITIONAL INFORMATION GL 79-055 Summary of Meeting Held on October 12, 1979 to Discuss Responses to Bulletins79-05C and 79-06C and HPI Termination Criteria GL 79-056 Discussion of Lessons Learned Short Term Requirements GL 79-057 Acceptance Criteria for Mark I Long Term Program GL 79-058 ECCS Calculations on Fuel Cladding GL 79-059 This GL was never issued.GL 79-060 Discussion of Lessons Learned Short Term Requirements GL 79-061 Discussion of Lessons Learned Short Term Requirements GL 79-062 ECCS Calculations on Fuel Cladding NA Item was applicable only to units with operating license at the time the--- item was issued.NA Boiling Water Reactor NA NA NA NA Item was applicable only to units with operating license at the time the item was issued.Info Info NA Item was applicable only to units with operating license at the time the--item was issued.GL 79-063 Upgraded Emergency Plans GL 79-064 Suspension of All Operating Licenses (PWRs)GL 79-065 Radiological Environmental Monitoring Program Requirements

-Enclosing Branch Technical Position, Revision 1 GL 79-066 Additional Information Re 11/09/79 Letter on ECCS Calculations

[GL 79-62]GL 79-067 Estimates for Evacuation of Various Areas Around Nuclear Power Reactors Duplicate of GL 79-058 C GL 79-63 advised applicants for licenses of proposed rulemaking that--NRC concurrence in State and local emergency plans would be a 01 condition for issuing an operating license. TVA responded to GL 79-63 on January 3, 1980, and confirmed the intent to revise the Emergency Plan to address the NRC requirements.

NA Info NA Info NA Info NA Info Page 45 of 109* = See last page for status code definition.

ITEM TITLE REV NA Info ADDITIONAL INFORMATION GL 79-068 Audit of Small Break LOCA Guidelines GL 79-069 Cladding Rupture, Swelling, and Coolant Blockage as a Result of a Reactor Accident GL 79-070 Environmental Monitoring for Direct Radiation GL 80-001 NUREG-0630, "Cladding, Swelling and Rupture -Models For LOCA Analysis" GL 80-002 QA Requirements Regarding Diesel Generator Fuel Oil GL 80-003 BWR Control Rod Failures NA Info NA Info NA Info C TVA: FSAR 9.5.4.2 NA GL 80-004 B 80-01, "Operability of ADS Valve NA Pneumatic Supply" GL 80-005 B 79-01b, "Environmental NA Qualification of Class 1 E Equipment" GL 80-006 Issuance of NUREG-0313, Rev 1, NA"Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping" GL 80-007 This GL was never issued. NA Boiling Water Reactor Boiling Water Reactor Info Boiling Water Reactor GL 80-008 B 80-02. "Inadequate Quality Assurance for Nuclear Supplied Equipment" GL 80-009 Low Level Radioactive Waste Disposal GL 80-010 Issuance of NUREG-0588,"Interim Staff Position On Equipment Qualifications of Safety-Related Electrical Equipment" GL 80-011 B 80-03, "Loss of Charcoal From Standard Type II, 2 Inch, Tray Absorber Cells" NA Boiling Water Reactor NA Item was applicable only to units with operating license at the time the---item was issued.NA Info C GL 80-11 transmitted Bulletin 80-03. TVA responded to B 80-03 on March-21, 1980. See B 80-03 for further information.

01 Page 46 of 109* = See last page for status code definition.

REV ITEM TITLE GL 80-012 B 80-04, "Analysis of a PWR Main NA Steam Line Break With Continued

_Feedwater Addition" GL 80-013 Qualification of Safety Related NA Electrical Equipment GL 80-014 LWR Primary Coolant System S Pressure Isolation Valves 02 ADDITIONAL INFORMATION Info Item was applicable only to units with operating license at the time the item was issued.TVA: FSAR 5.2.7.4 NRC: 1.14.2 of SSER 6 NRC reviewed in 1.14.2 of SSER6.Unit 2 Action: Incorporate guidance into Technical Specifications.

....................................................................................................

REVISION 02 UPDATE: Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.TS Surveillance Requirement 3.4.13.1 verifies RCS operational leakage by performance of an RCS water inventory balance.Info Info Boiling Water Reactor Applies only to Babcock and Wilcox designed plants Info Info GL 80-015 Request for Additional NA Management and Technical Resources Information GL 80-016 B 79-01b, "Environmental NA Qualification of Class 1 E Equipment" GL 80-017 Modifications to BWR Control Rod NA Drive Systems GL 80-018 Crystal River 3 Reactor Trip From NA Approximately 100% Full Power GL 80-019 Resolution of Enhanced Fission NA Gas Release Concern GL 80-020 Actions Required From OL NA Applicants of NSSS Designs by W and CE Resulting From NRC B&O Task Force Review of TMI2 Accident Page 47 of 109* = See last page for status code definition.

ITEM GL 80-021 TITLE B 80-05, "Vacuum Condition Resulting in Damage to Chemical Volume Control System Holdup Tanks" REV CI ADDITIONAL INFORMATION Closed in IR 50-390/84-59 and 50-391/84-45.

Unit 2 A Comple.ction: te surveillance procedures for Unit 2.GL 80-022 Transmittal of NUREG-0654, NA Info"Criteria For Preparation and Evaluation of Radiological Emergency Response Plan" GL 80-023 Change of Submittal Date For NA Info Evaluation Time Estimates GL 80-024 Transmittal of Information on NRC NA Info"Nuclear Data Link Specifications" --GL 80-025 B 80-06, "Engineering Safety NA Info Feature (ESF) Reset Controls" _GL 80-026 Qualifications of Reactor Operators NA Info GL 80-027 B 80-07, "BWR Jet Pump NA Boiling Water Reactor Assembly Failure" GL 80-028 B 80-08, "Examination of Containment Liner Penetration Welds" C GL 80-28 transmitted Bulletin 80-08. TVA responded to--B 80-08 on July 8, 1980. See B 80-08 for further information.

01 GL 80-029 Modifications to Boiling Water NA Boiling Water Reactor Reactor Control Rod Drive Systems GL 80-030 Clarification of The Term NA Item was applicable only to units with operating license at the time the"Operable" As It Applies to Single _ item was issued.Failure Criterion For Safety Systems Required by TS GL 80-031 B 80-09, "Hydramotor Actuator NA Info Deficiencies" GL 80-032 Information Request on C GL 80-32 transmitted NRC questions on masonry walls.Category I Masonry Walls -. TVA provided the information requested by letters dated February 12, Employed by Plants Under 01 1981, for reinforced walls and August 20, 1981, for nonreinforced walls.CP and OL Review TVA provided a final response on January 22, 1982. See B 80-11 for further information.

Page 48 of 109* = See last page for status code definition.

ITEM TITLE REV ITEMTITL REVADDITIONAL INFORMATION GL 80-033 Actions Required From OL NA Appli Applicants of B&W Designed NSSS Resulting From NRC B&O Task Force Review of TMI2 Accident GL 80-034 Clarification of NRC Requirements NA Info for Emergency Response Facilities at Each Site GL 80-035 Effect of a DC Power Supply NA Boilir Failure on ECCS Performances GL 80-036 B 80-10, "Contamination of NA Info Non-Radioactive System and Resulting Potential For Unmonitored, Uncontrolled Release to Environment" GL 80-037 Five Additional TMI-2 Related NA Item Requirements to Operating item Reactors GL 80-038 Summary of Certain Non-Power NA Does Reactor Physical Protection Requirements GL 80-039 B 80-11, "Masonry Wall Design" NA Info GL 80-040 Transmittal of NUREG-0654, NA Info"Report of the B&O Task Force" and Appropriate NUREG-0626,"Generic Evaluation of FW Transient and Small Break LOCA" GL 80-041 Summary of Meetings Held on NA Info April 22 &23, 1980 With Representatives of the Mark I Owners Group GL 80-042 B 80-12, "Decay Heat Removal NA Info System Operability" -_GL 80-043 B 80-13, "Cracking In Core Spray NA Boilir Spargers" _ -GL 80-044 Reorganization of Functions and NA Info Assignments Within ONRR/SSPB GL 80-045 Fire Protection Rule NA Item--item es only to Babcock and Wilcox designed plants ng Water Reactor was applicable only to units with operating license at the time the was issued.not apply to power reactor.ng Water Reactor was applicable only to units with operating license at the time the was issued.Page 49 of 109= See last page for status code definition.

ITEM GL 80-046 and GL 80-047 TITLE REV C Generic Technical Activity A-12,"Fracture Toughness and Additional Guidance on Potential for Low Fracture toughness and Laminar Tearing on PWR Steam Generator Coolant Pump Supports" ADDITIONAL INFORMATION No response was required for this GL, and NUREG-0577 states that the lamellar tearing aspect of this issue was resolved by the NUREG. Further, the NUREG states that for plants under review, the fracture toughness issue was resolved.GL 80-048 Revision to 5/19/80 Letter On Fire Protection

[GL 80-45]GL 80-049 Nuclear Safeguards Problems GL 80-050 Generic Activity A-10, "BWR Cracks" GL 80-051 On-Site Storage of Low-Level Waste GL 80-052 Five Additional TMI-2 Related Requirements

-Erata Sheets to 5/7/80 Letter [GL 80-37]GL 80-053 Decay Heat Removal Capability GL 80-054 B 80-14, "Degradation of Scram Discharge Volume Capability" GL 80-055 B 80-15, "Possible Loss of Hotline With Loss of off-Site Power" GL 80-056 Commission Memorandum and Order on Equipment Qualification GL 80-057 Further Commission Guidance For Power Reactor Operating Licenses NUREG-0660 and NUREG-0694 GL 80-058 B 80-16, "Potential Misapplication of Rosemount Inc. Models 1151/1152 Pressure Transmitters With "A" Or "D" Output Codes" NA Item was applicable only to units with operating license at the time the---item was issued.NA Info NA Boiling Water Reactor NA NA Item was applicable only to units with operating license at the time the item was issued.Item was applicable only to units with operating license at the time the item was issued.NA NA NA Item was applicable only to units with operating license at the time the item was issued.Boiling Water Reactor Info NA Info NA Info NA Info Page 50 of 109* = See last page for status code definition.

ITEM TITLE*REV NA Info ADDITIONAL INFORMATION GL 80-059 Transmittal of Federal Register Notice RE Regional Meetings to Discuss Environmental Qualification of Electrical Equipment GL 80-060 Request for Information Regarding Evacuation Times GL 80-061 TMI-2 Lessons Learned NA NA Info Info GL 80-062 TMI-2 Lessons Learned NA Boiling Water Reactor GL 80-063 B 80-17, "Failure of Control Rods to Insert During a Scram at a BWR" GL 80-064 Scram Discharge Volume Designs GL 80-065 Request for Estimated Construction Completion and Fuel Load Schedules GL 80-066 B 80-17, Supplement 1, "Failure of Control Rods to Insert During a Scram at a BWR" GL 80-067 Scram Discharge Volume NA Boiling Water Reactor NA NA Boiling Water Reactor Info NA Boiling Water Reactor NA Boiling Water Reactor GL 80-068 B 80-17, Supplement 2, "Failures NA Boiling Water Reactor Revealed by Testing Subsequent

_to Failure of Control Rods to Insert During a Scram at a BWR" GL 80-069 B 80-18, "Maintenance of NA Info Adequate Minimum Flow Through _ .Centrifugal Charging Pumps Following Secondary Side HELB" GL 80-070 B 80-19, "Failures of Mercury- NA Info Wetted Matrix Relays in RPS of Operating Nuclear Power Plants Designed by GE" GL 80-071 B 80-20, "Failures of NA Info Westinghouse Type W-2 Spring _ -Return to Neutral Control Switches" Page 51 of 109* = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 80-072 Interim Criteria For Shift Staffing NA Info GL 80-073 "Functional Criteria For NA Info Emergency Response Facilities," _ -_NUREG-0696 GL 80-074 Notice of Forthcoming Meeting NA Info With Representatives of EPRI to Discuss Program For Resolution of USI A-12, "Fracture Toughness Issue" GL 80-075 Lessons Learned Tech. Specs. NA Item was applicable only to units with operating license at the time the---item was issued.NA Info GL 80-076 Notice of Forthcoming Meeting With GE to Discussed Proposed BWR Feedwater Nozzle Leakage Detection System GL 80-077 Refueling Water Level -Technical Specifications Changes S 02 Unit 2 Action: Address in Technical Specifications, as appropriate.

REVISION 02 UPDATE:, Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.TS LCO 3.9.7 requires the refueling cavity water level to be maintained greater than or equal to 23 feet above the top of the reactor vessel flange during movement of irradiated fuel assemblies within containment.

Boiling Water Reactor Boiling Water Reactor GL 80-078 Mark I Containment Long-Term Program GL 80-079 B 80-17, Supplement 3, "Failures Revealed by Testing Subsequent to Failure of Control Rods to Insert During a Scram At a BWR" GL 80-080 Preliminary Clarification of TMI Action Plan Requirements GL 80-081 Preliminary Clarification of TMI Action Plan Requirements

-Addendum to 9/5/80 Letter[GL 80-80]GL 80-082 B 79-01b, Supplement 2,"Environmental Qualification of Class 1E Equipment" NA NA NA Info NA Info NA Info Page 52 of 109* = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 80-083 Environmental Qualification of Safety-Related Equipment GL 80-084 BWR Scram System GL 80-085 Implementation of Guidance From USI A-12, "Potential For LOW Fracture Toughness and Lamellar Tearing On Component Support" GL 80-086 Notice of Meeting to Discuss Final Resolution of USI A-12 GL 80-087 Notice of Meeting to Discuss Status of EPRI-Proposed Resolution of the USI A-12 Fracture Toughness Issue GL 80-088 Seismic Qualification of Auxiliary Feedwater Systems GL 80-089 B 79-01b, Supplement 3,"Environmental Qualification of Class 1 E Equipment" GL 80-090 NUREG-0737, TMI (Prior and future GLs, with the exception of certain discrete scopes, have been screened into NUREG list for those applicable to Watts Bar 2)GL 80-091 ODYN Code Calculation GL 80-092 B 80-21, "Valve Yokes Supplied by Malcolm Foundry Company, Inc." NA Info NA Boiling Water Reactor NA Info NA -----Info- ----NA Info NA NA CI Item was applicable only to units with operating license at the time the item was issued.Info See NUREG items in this list.NA Boiling Water Reactor C GL 80-92 transmitted Bulletin 80-21. TVA responded to___ B 80-21 on May 6, 1981. See B 80-21 for further information.

01 GL 80-093 Emergency Preparedness NA Does not apply to power reactor.GL 80-094 Emergency Plan NA Info GL 80-095 Generic Technical Activity A-10, NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking" NA Boiling Water Reactor Page 53 of 109* = See last page for status code definition.

ITEM GL 80-096 TITLE REV ADDITIONAL INFORMATION Fire Protection NA Addressed to specific plant(s).GL 80-097 B 80-23, "Failures of Solenoid NA Info Valves Manufactured by Valcor Engineering Corporation" GL 80-098 B 80-24, "Prevention of Damage NA Info Due to Water Leakage Inside Containment" GL 80-099 Technical Specifications Revisions For Snubber Surveillance GL 80-100 Appendix R to 10 CFR 50 Regarding Fire Protection

-Federal Register Notice GL 80-101 Inservice Inspection Programs GL 80-102 Commission Memorandum and Order of May 23, 1980 (Referencing B 79-01b, Supplement 2 -q.2 & 3 -Sept 30, 1980)GL 80-103 Fire Protection

-Revised Federal Register Notice GL 80-104 Orders On Environmental Qualification of Safety Related Electrical Equipment GL 80-105 Implementation of Guidance For USI A-12, "Potential For Low Fracture toughness and Lamellar Tearing On Component Supports" GL 80-106 Report On ECCS Cladding Models, NUREG-0630 GL 80-107 BWR Scram Discharge System NA Info NA Item was applicable only to units with operating license at the time the--item was issued.NA Addressed to specific plant(s).NA Info NA Info NA Info NA Info NA Info NA Boiling Water Reactor GL 80-108 Emergency Planning NA Info GIL 80-109 Guidelines For SEP Soil Structure Interaction Reviews NA Info Page 54 of 109* = See last page for status code definition.

ITEM GL 80-110 TITLE Periodic Updating of FSARS REV NA NA ADDITIONAL INFORMATION Item was applicable only to units with operating license at the time the item was issued.GL 80-111 B 80-17, Supplement 4, "Failure of Control Rods to Insert During a Scram at a BWR" GL 80-112 B 80-25, "Operating Problems With Target Rock Safety Relief Valves" GL 80-113 Control of Heavy Loads GL 81-001 Qualification of Inspection, Examination, Testing and Audit Personnel GL 81-002 Analysis, Conclusions and Recommendations Concerning Operator Licensing GL 81-003 Implementation of NUREG-0313,"Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping" GL 81-004 Emergency Procedures and Training for Station Blackout Events GL 81-005 Information Regarding The Program For Environmental Qualification of Safety-Related Electrical Equipment GL 81-006 Periodic Updating of Final Safety Analysis Reports (FSARS)GL 81-007 Control of Heavy Loads Boiling Water Reactor NA Info C Superseded by GL 81-007.NA Info NA NA Info Boiling Water Reactor C Superseded by Station Blackout Rule.NA Info NA Info Cl "Movement of Heavy Loads Over Spent Fuel, Over Fuel in the Reactor, or Over Safety-Related Equipment" -NRC closure letter dated May 20, 1998.LICENSE CONDITION

-Control of heavy loads (NUREG-0612)

The staff concluded in SSER1 3 that the license condition was no longer necessary based on their review of TVA's response to NUREG-0612 guidelines for Phase I in TVA letter dated July 28, 1993.Unit 2 Action: Unit 2 Heavy Loads Program will be in compliance with NUREG-0612.

Page 55 of 109* = See last page for status code definition.

R RE BilnVatrReco ITEM GL 81-008 TITLE ODYN Code ADDITIONAL INFORMATION NA Boiling Water Reactor GL 81-009 BWR Scram Discharge System GL 81-010 Post-TMI Requirements For The Emergency Operations Facility GL 81-011 BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking (NUREG-0619)

GL 81-012 Fire Protection Rule GL 81-013 SER For GEXL Correlation For 8X8R Fuel Reload Applications For Appendix D Submittals of The GE topical Report GL 81-014 Seismic Qualification of Auxiliary Feedwater Systems NA Boiling Water Reactor NA Info NA NA NA Cl Boiling Water Reactor Item was applicable only to units with operating license at the time the item was issued.Boiling Water Reactor TVA: FSAR 10.4.9 Unit 2 Action: Additional Unit 2 implementing procedures or other activity is required for completion.

[WAS "OL."]Info Applies only to Babcock and Wilcox designed plants Info Boiling Water Reactor GL 81-015 Environmental Qualification of NA Class 1 E Electrical Equipment

--__Clarification of Staff's Handling of Proprietary Information GL 81-016 NUREG-0737, Item I.C.1 SER on NA Abnormal Transient Operating

-.Guidelines (ATOG)GL 81-017 Functional Criteria for Emergency NA Response Facilities

_ -.GL 81-018 BWR Scram Discharge System -NA Clarification of Diverse Instrumentation Requirements GL 81-019 Thermal Shock to Reactor NA Pressure Vessels GL 81-020 Safety Concerns Associated With NA Pipe Breaks in the BWR Scram System Item was applicable only to units with operating license at the time the item was issued.Boiling Water Reactor Page 56 of 109* = See last page for status code definition.

ITEM TITLE GL 81-021 Natural Circulation Cooldown REV CI ADDITIONAL INFORMATION GL 81-022 Engineering Evaluation of the NA H. B. Robinson Reactor Coolant System Leak on 1/29/81 GL 81-023 INPO Plant Specific Evaluation NA Reports GL 81-024 Multi-Plant Issue B-56, "Control NA Rods Fail to Fully Insert" GL 81-025 Change in Implementing Schedule NA For Submission and Evaluation of Upgraded Emergency Plans GL 81-026 Licensing Requirements for NA Pending Construction Permit and Manufacturing License Applications GL 81-027 Privacy and Proprietary Material in NA Emergency Plans TVA responded December 3, 1981.Unit 2 Action: Issue operating procedures.

Info Info Boiling Water Reactor Info Applicants with pending Construction Permits Info Info GL 81-028 Steam Generator Overfill NA GL 81-029 Simulator Examinations NA Info GL 81-030 Safety Concerns Associated With Pipe Breaks in the BWR Scram System GL 81-031 This GL was never issued.GL 81-032 NUREG-0737, Item II.K.3.44,"Evaluation of Anticipated Transients Combined With Single Failure" GL 81-033 This GL was never issued.GL 81-034 Safety Concerns Associated With Pipe Breaks in the BWR Scram System NA Boiling Water Reactor NA NA Boiling Water Reactor NA NA Boiling Water Reactor Page 57 of 109* = See last page for status code definition.

REV ITEM TITLE ADDITIONAL INFORMATION GL 81-035 Safety Concerns Associated With NA Boiling Water Reactor Pipe Breaks in the BWR Scram System GL 81-036 Revised Schedule for Completion NA Info of TMI Action Plan Item Il.D.1,"Relief and Safety Valve Testing" GL 81-037 ODYN Code Reanalysis NA Boiling Water Reactor Requirements

_ _GL 81-038 Storage of Low Level Radioactive NA Info Wastes at Power Reactor Sites _ _GL 81-039 NRC Volume Reduction Policy NA Info GL 81-040 Qualifications of Reactor Operators NA Info GL 82-001 New Applications Survey NA Info GL 82-002 Commission Policy on Overtime NA Info GL 82-003 High Burnup MAPLHGR Limits NA Boiling Water Reactor GL 82-004 Use of INPO See-in Program GL 82-005 Post-TMI Requirements GL 82-006 This GL was never issued.GL 82-007 Transmittal of NUREG-0909 Relative to the Ginna Tube Rupture GL 82-008 Transmittal of NUREG-0909 Relative to the Ginna Tube Rupture GL 82-009 Environmental Qualification of Safety Related Electrical Equipment NA Info NA Item was applicable only to units with operating license at the time the---item was issued.NA NA Boiling Water Reactor NA NA Info Info Page 58 of 109* = See last page for status code definition.

REV ITEM GL 82-010 TITLE Post-TMI Requirements ADDITIONAL INFORMATION NA Item was applicable only to units with operating license at the time the---item was issued.GL 82-011 Transmittal of NUREG-0916 NA Info Relative to the Restart of R. E.Ginna Nuclear Power Plant GL 82-012 Nuclear Power Plant Staff Working NA Info Hours GL 82-013 Reactor Operator and Senior NA Info Reactor Operator Examinations GL 82-014 Submittal of Documents to the NA Info NRC GL 82-015 This GL was never issued. NA GL 82-016 NUREG-0737 Technical Specifications GL 82-017 Inconsistency of Requirements Between 50.54(T) and 50.15 GL 82-018 Reactor Operator and Senior Reactor Operator Requalification Examinations GL 82-019 Submittal of Copies of Documentation to NRC -Copy Requirements for Emergency Plans and Physical Security Plans GL 82-020 Guidance for Implementing the Standard Review Plan Rule GL 82-021 Fire Protection Audits GL 82-022 Congressional Request for Information Concerning Steam Generator Tube Integrity GL 82-023 Inconsistency Between Requirements of 10CFR 73.40(d)and Standard Technical Specifications For Performing Audits of Safeguards Contingency Plans NA Item was applicable only to units with operating--- item was issued.NA Info license at the time the NA Info NA Info NA Info NA Info NA Item was applicable only to units with operating license at the time the item was issued.NA Info Page 59 of 109* = See last page for status code definition.

ITEM TITLE GL 82-024 Safety Relief Valve Quencher Loads: BWR MARK II and III Containments GL 82-025 Integrated IAEA Exercise for Physical Inventory at LWRS GL 82-026 NUREG-0744, REV. 1, "Pressure Vessel Material Fracture Toughness" GL 82-027 Transmittal of NUREG-0763,"Guidelines For Confirmatory In-Plant Tests of Safety-Relief Valve Discharge for BWR Plants" GL 82-028 Inadequate Core Cooling Instrumentation System REV NA NA NA ADDITIONAL INFORMATION Boiling Water Reactor Item was applicable only to units with operating license at the time the item was issued.Item was applicable only to units with operating license at the time the item was issued.NA Boiling Water Reactor CO LICENSE CONDITION

-Detectors for Inadequate core cooling (II.F.2)06 In the original SER, the review of the ICC instrumentation was incomplete.

The January 24, 1992, letter superseded the previous responses on this issue. TVA letter for Units 1 and 2 dated January 24, 1992, committed to install Westinghouse ICCM-86 and associated hardware.

NRC completed the review for Units 1 and 2 in SSER1 0. For Unit 2 due to obsolescence of the ICCM-86 system, TVA intends to install the Westinghouse Common Q Post-Accident Monitoring System.Unit 2 Action: Install Westinghouse Common 0 PAM system.'Closed. Subsumed as part of NRC staff review of Instrumentation and----------------

REVISION 06 UPDATE: SSER22 contained the following for NRC Action: GL 82-029 This GL was never issued. NA"Closed. Subsumed as part of NRC staff review of Instrumentation and Controls submitted April 8, 2010." Info GL 82-030 Filings Related to 10 CFR 50 Production and Utilization Facilities GL 82-031 This GL was never issued.GL 82-032 Draft Steam Generator Report (SAI)NA NA NA Item was applicable only to units with operating license at the time the--item was issued.Page 60 of 109* = See last page for status code definition.

ITEM GL 82-033 TITLE Supplement to NUREG-0737,"Requirements for Emergency Response Capability" REV Cl ADDITIONAL INFORMATION"Safety Parameter Display System" (SPDS) / "Requirements for Emergency Response Capability" -NRC reviewed in SSER5, SSER6, and 18.2.2 of SSER15.Unit 2 Action: Install SPDS and have it operational prior to start-up after the first refueling outage.GL 82-034 This GL was never issued.GL 82-035 This GL was never issued.GL 82-036 This GL was never issued..GL 82-037 This GL was never issued.NA NA NA NA GL 82-038 Meeting to Discuss Developments for Operator Licensing Examinations GL 82-039 Problems With Submittals of Subsequent Information of CURT 73.21 For Licensing Reviews GL 83-001 Operator Licensing Examination Site Visit GL 83-002 NUREG-0737 Technical Specifications GL 83-003 This GL was never issued.NA Info NA Info NA Info NA Boiling Water Reactor NA GL 83-004 Regional Workshops Regarding Supplement 1 to NUREG-0737,"Requirements For Emergency Response Capability" GL 83-005 Safety Evaluation of "Emergency Procedure Guidelines, Revision 2," June 1982 GL 83-006 Certificates and Revised Format For Reactor Operator and Senior Reactor Operator Licenses GL 83-007 The Nuclear Waste Policy Act of 1982 NA Info NA Boiling Water Reactor NA NA Info Info Page 61 of 109* = See last page for status code definition.

ITEM GL 83-008 GL 83-009 GL 83-010a GL 83-010b GL 83-010c TITLE REV Modification of Vacuum Breakers NA on Mark I Containments Review of Combustion NA Engineering Owners' Group Emergency Procedures Guideline Program Resolution of TMI Action Item NA IIK.3.5., "Automatic Trip of Reactor Coolant Pumps" Resolution of TMI Action Item NA 11,K.3.5., "Automatic Trip of Reactor Coolant Pumps" Resolution of TMI Action Item Cl ILK.3.5., "Automatic Trip of Reactor Coolant Pumps" ADDITIONAL INFORMATION Boiling Water Reactor Applies only to Combustion Engineering designed plants Applies only to Combustion Engineering designed plants Applies only to Combustion Engineering designed plants TVA: letters dated January 5, 1984 and June 25, 1984 NRC: letter dated June 8, 1990.Unit 2 Action: Incorporate emergency response guidelines into applicable procedures.

[WAS "NOTE 3."]GL 83-01 Od Resolution of TMI Action Item NA Item was applicable only to units with operating license at the time the 11.K.3.5., "Automatic Trip of --. item was issued.Reactor Coolant Pumps" GL 83-010e Resolution of TMI Action Item NA Applies only to Babcock and Wilcox designed plants IIK.3.5., "Automatic Trip of Reactor Coolant Pumps" GL 83-01Of Resolution of TMI Action Item NA Applies only to Babcock and Wilcox designed plants 11,K.3.5., "Automatic Trip of Reactor Coolant Pumps" GL 83-011 Licensee Qualification for NA Item was applicable only to units with operating license at the time the Performing Safety Analyses in _ -. item was issued.Support of Licensing Actions GL 83-012 Issuance of NRC FORM 398 -NA Info Personal Qualifications Statement

-Licensee GL 83-013 Clarification of Surveillance NA Info Requirements for HEPA Filters and Charcoal Absorber Units In Standard Technical Specifications on ESF Cleanup Systems Page 62 of 109* = See last page for status code definition.

REV ITEM TITLE ADDITIONAL INFORMATION GL 83-014 Definition of "Key Maintenance Personnel," (Clarification of Generic Letter 82-12)GL 83-015 Implementation of Regulatory Guide 1.150, "Ultrasonic Testing of Reactor Vessel Welds During Preservice

& Inservice Examinations, Revision 1" GL 83-016 Transmittal of NUREG-0977 Relative to the ATWS Events at Salem Generating Station, Unit No.1 GL 83-016a Transmittal of NUREG-0977 Relative to the ATWS Events at Salem Generating Station, Unit No.1 GL 83-017 Integrity of Requalification Examinations for Renewal of Reactor Operator and Senior Reactor Operator Licenses GL 83-018 NRC Staff Review of the BWR Owners' Group (BWROG) Control Room Survey Program GL 83-019 New Procedures for Providing Public Notice Concerning Issuance of Amendments to Operating Licenses GL 83-020 Integrated Scheduling for Implementation of Plant Modifications GL 83-021 Clarification of Access Control Procedures for Law Enforcement Visits GL 83-022 Safety Evaluation of "Emergency Response Guidelines" GL 83-023 Safety Evaluation of "Emergency Procedure Guidelines" GL 83-024 TMI Task Action Plan Item I.G.1,"Special Low Power Testing and Training," Recommendations for BWRs NA Info NA Info NA Info NA Info NA Info NA NA Boiling Water Reactor Item was applicable only to units with operating license at the time the item was issued.NA Info NA Info NA Info NA Applies only to Combustion Engineering designed plants NA Boiling Water Reactor Page 63 of 109* = See last page for status code definition.

REV ITEM TITLE ADDITIONAL INFORMATION GL 83-025 This GL was never issued.GL 83-026 Clarification Of Surveillance Requirements For Diesel Fuel Impurity Level Tests GL 83-027 Surveillance Intervals in Standard Technical Specifications GL 83-028 "Required Actions Based on Generic Implications of Salem ATWS Events: 1.2 -Post Trip Review Data and Information Capability GL 83-028 "Required Actions Based on Generic Implications of Salem ATWS Events: 2.1 -Equipment Classification and Vendor Interface (Reactor Trip System Components)

NA NA Info NA Info C TVA: letters dated November 7, 1983 and December 4, 1987 NRC: IR 50-390, 391/86-04 Cl TVA: letters dated November 7, 1983 and August 24, 1990 06 NRC: letters dated October 20, 1986 and June 18, 1990 Unit 2 Action: Ensure that required information on Critical Structures and Components is properly incorporated into procedures.

[WAS "NOTE 3."]....................................................................................................

REVISION 06 UPDATE: Confirmed that required information on Critical Structures and Components is properly incorporated into procedures.

CI Unit 2 Action: Enter engineering component background data in INPO's Equipment Performance and Information Exchange System (EPIX) for Unit 2.GL 83-028 "Required Actions Based on Generic Implications of Salem ATWS Events: 2.2 -Equipment Classification and Vendor Interface (All SR Components)" Page 64 of 109* = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 83-028 "Required Actions Based on Generic Implications of Salem ATWS Events: S TVA: letters dated November 7, 1983, January 17, 1986 and November 1, 1993 02 3.1 -Post-Maintenance Testing (Reactor Trip System Components)

NRC: letters dated December 10, 1985, October 27, 1986, and July 2, 1990; IR 390, 391/86-04 Unit 2 Action: Test and maintenance procedures and Technical Specifications will include post-maintenance operability testing of safety-related components of the reactor trip system.REVISION 02 UPDATE: Developmental Revision A of the Unit 2 TS (including the TS Bases) was submitted on March 4, 2009.The Bases for TS Surveillance Requirement 3.0.1 states, in part, "Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE.

This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2." GL 83-028 "Required Actions Based on Generic Implications of Salem ATWS Events: 3.2 -Post-Maintenance Testing (All SR Components)

S TVA: letters dated November 7, 1983, January 17, 1986 and November 1, 1993 06 NRC: letters dated December 10, 1985, October 27, 1986, and July 2, 1990; IR 390, 391/86-04 Unit 2 Action: Test and maintenance procedures and Technical Specifications will include post-maintenance operability testing of other (than reactor trip system) safety-related components.

REVISION 02 UPDATE: Developmental Revision A of the Unit 2 TS (including the TS Bases) was submitted on March 4, 2009.The Bases for TS Surveillance Requirement 3.0.1 states, in part, "Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE.

This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2." REVISION 06 UPDATE: Watts Bar's Preventative Maintenance Program is not unit specific; no further action is required for Unit 2.Page 65 of 109* = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 83-028 "Required Actions Based on Generic Implications of Salem ATWS Events: 4.1 -Reactor Trip System Reliability (Vendor Related Modifications)

CO TVA: letter dated May 19, 1986 06 Unit 2 Action: Confirm vendor-recommended DS416 breaker modifications are implemented.

--. -. -.- .----------------------------

.-.-.-- .------ .-.-.-.---

.-.- ....-.-.----.-.-.-

GL 83-028 "Required Actions Based on Generic Implications of Salem ATWS Events: S 02 4.2 -Reactor Trip System Reliability (Preventive Maintenance and Surveillance Program for Reactor Trip Breakers)REVISION 06 UPDATE: NRC Inspection Report 391/2011-602 closed GL 83-028, Item 4.1.TVA: letters dated November 7, 1983, February 10, 1986, and May 19, 1986 NRC: letters dated July 26, 1985 and June 18, 1992; SSER 16 Unit 2 Action: Ensure maintenance instruction procedure and Technical Specifications support reliable reactor trip breaker operation.

REVISION 02 UPDATE: Developmental Revision B of the Unit 2 TS was submitted on February 2, 2010.Item 17. (Reactor Trip Breakers) of TS Table 3.3.1-1 states the requirement for the reactor trip breakers.GL 83-028 "Required Actions Based on Generic Implications of Salem ATWS Events: 4.3 -Reactor Trip System Reliability (Automatic Actuation of Shunt Trip Attachment)

C TVA: letters dated November 7, 1983, March 22, 1985 NRC: IR 50-390/86-04 and 50-391/86-04; letter dated June 18, 1990 Page 66 of 109* = See last page for status code definition.

ITEM GL 83-028 TITLE"Required Actions Based on Generic Implications of Salem ATWS Events: 4.5 -Reactor Trip System Reliability (Automatic Actuation of Shunt Trip Attachment)

REV S ADDITIONAL INFORMATION TVA: letters dated November 7, 1983 and July 26, 1985 02 "NRC: letters dated June 28, 1990 and October 9, 1990;SSERs 5 and 16 Unit 2 Action: Address in Technical Specifications, as appropriate.

GL 83-029 This GL was never issued.GL 83-030 Deletion of Standard Technical Specifications Surveillance Requirement 4.8.1.1.2.d.6 For Diesel Generator Testing GL 83-031 Safety Evaluation of "Abnormal Transient Operating Guidelines" GL 83-032 NRC Staff Recommendations Regarding Operator Action for Reactor Trip and ATWS GL 83-033 NRC Positions on Certain Requirements of Appendix R to 10 CFR 50 GL 83-034 This GL was never issued.REVISION 02 UPDATE: Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.Item 18. (Reactor Trip Breaker Undervoltage and Shunt Trip Mechanisms) of TS Table 3.3.1-1 states the requirement for the shunt trip attachment.

NA NA Info NA Applies only to Babcock and Wilcox designed plants NA Info NA Info NA GL 83-035 Clarification of TMI Action Plan Item I1.K.3.31 GL 83-036 NUREG-0737 Technical Specifications GL 83-037 NUREG-0737 Technical Specifications GL 83-038 NUREG-0965, "NRC Inventory of Dams" NA NA Info Boiling Water Reactor NA Item was applicable only to units with operating license at the time the---item was issued.NA Info Page 67 of 109* = See last page for status code definition.

ITEM GL 83-039 GL 83-040 TITLE Voluntary Survey of Licensed NA Info REV ADDITIONAL INFORMATION Operators---

Operator Licensing Examination NA Info GL 83-041 Fast Cold Starts of Diesel NA Generators Item was applicable only to units with operating license at the time the item was issued.GL 83-042 Clarification to GL 81-07 NA Info Regarding Response to NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants" GL 83-043 Reporting Requirements of NA Info 10 CFR 50, Sections 50.72 and _ -50.73, and Standard Technical Specifications GL 83-044 Availability of NUREG-1021, NA Info"Operator Licensing Examiner _ -_Standards" GL 84-001 NRC Use Of The Terms"Important To Safety" and "Safety Related" GL 84-002 Notice of Meeting Regarding Facility Staffing GL 84-003 Availability of NUREG-0933, "A Prioritization of Generic Safety Issues" GL 84-004 Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops GL 84-005 Change to NUREG-1021,"Operator Licensing Examiner Standards" GL 84-006 Operator and Senior Operator License Examination Criteria For Passing Grade GL 84-007 Procedural Guidance for Pipe Replacement at BWRs NA Info NA Info NA Info NA Info NA Info NA Does not apply to power reactor.NA Boiling Water Reactor Page 68 of 109* = See last page for status code definition.

ITEM TITLE GL 84-008 Interim Procedures for NRC Management of Plant-Specific Backfitting GL 84-009 Recombiner Capability Requirements of 10 CFR 50.44(c)(3)(ii)

GL 84-010 Administration of Operating Tests Prior to Initial Criticality GL 84-011 Inspection of BWR Stainless Steel Piping*REV A NA Info NA Boiling Water Reactor NA Info NA Boiling Water Reactor NA Info DDITIONAL INFORMATION GL 84-012 Compliance With 10 CFR Part 61 and Implementation of Radiological Effluent Technical Specifications (RETs) and Attendant Process Control Program (PCP)GL 84-013 Technical Specification for Snubbers GL 84-014 Replacement and Requalification Training Program GL 84-015 Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability GL 84-016 Adequacy of On-Shift Operating Experience for Near Term Operating License Applicants GL 84-017 Annual Meeting to Discuss Recent Developments Regarding Operator Training, Qualifications, and Examinations GL 84-018 Filing of Applications for Licenses and Amendments GL 84-019 Availability of Supplement 1 to NUREG-0933, "A Prioritization of Generic Safety Issues" GL 84-020 Scheduling Guidance for Licensee Submittals of Reloads That Involve Unreviewed Safety Questions NA Info NA Info NA Info NA Info NA Info NA Does not apply to power reactor.NA Info NA Info Page 69 of 109* = See last page for status code definition.

R REV nf ITEM TITLE ADDITIONAL INFORMATION GL 84-021 Long Term Low Power Operation in Pressurized Water Reactors GL 84-022 This GL was never issued.GL 84-023 Reactor Vessel Water Level Instrumentation in BWRs GL 84-024 Certification of Compliance to 10 CFR 50.49, Environmental Qualification of Electric Equipment Important To Safety For Nuclear Power Plants GL 85-001 Fire Protection Policy Steering Committee Report GL 85-002 Recommended Actions Stemming From NRC Integrated Program for the Resolution of Unresolved Safety Issues Regarding Steam Generator Tube Integrity GL 85-003 Clarification of Equivalent Control Capacity for Standby Liquid Control Systems GL 85-004 Operating Licensing Examinations GL 85-005 Inadvertent Boron Dilution Events GL 85-006 Quality Assurance Guidance for ATWS Equipment That Is Not Safety-Related GL 85-007 Implementation of Integrated Schedules for Plant Modifications GL 85-008 10 CFR 20.408 Termination Reports -Format GL 85-009 Technical Specifications For Generic Letter 83-28, Item 4.3 GL 85-010 Technical Specification For Generic Letter 83-28, Items 4.3 and 4.4 NA Info NA NA Boiling Water Reactor Cl NA Cl NA NA NA NA NA NA NA See Special Program for Environmental Qualification.

Only issued as draft TVA responded to the GL on June 17, 1985.Unit 2 Action: Perform SG inspection.

Boiling Water Reactor Info Item was applicable only to units with operating license at the time the item was issued.Info Item was applicable only to units with operating license at the time the item was issued.Info Info NA Applies only to Babcock and Wilcox designed plants Page 70 of 109* = See last page for status code definition.

ITEM TITLE REV C ADDITIONAL INFORMATION GL 85-011 Completion of Phase II of "Control of Heavy Loads at Nuclear Power Plants," NUREG-0612 GL 85-012 Implementation Of TMI Action Item 11,K.3.5, "Automatic Trip Of Reactor Coolant Pumps" GL 85-013 Transmittal Of NUREG-1 154 Regarding The Davis-Besse Loss Of Main And Auxiliary Feedwater Event GL 85-014 Commercial Storage At Power Reactor Sites Of Low Level Radioactive Waste Not Generated By The Utility GL 85-015 Information On Deadlines For 10 CFR 50.49, "Environmental Qualification Of Electric Equipment Important To Safety At Nuclear Power Plants" GL 85-016 High Boron Concentrations GL 85-017 Availability Of Supplements 2 and 3 To NUREG-0933, "A Prioritization Of Generic Safety Issues" GL 85-018 Operator Licensing Examinations See GL 81-07.CI "Implementation of TMI Item II.K.3.5" -Reviewed in 15.5.4 of original 1982 SER; became License Condition

35. The staff determined that their review of Item II.K.3.5 did not have to be completed to support the full power license and considered this license condition resolved in SSER4.The item was further reviewed in Appendix EE of SSER1 6.Unit 2 Action: Implement modifications as required.NA Info NA Item was applicable only to units with operating license at the time the item was issued.NA Item was applicable only to units with operating license at the time the---item was issued.NA Info NA Info NA GL 85-019 Reporting Requirements On NA Primary Coolant Iodine Spikes GL 85-020 Resolution Of Generic Issue 69: NA High Pressure Injection/Make-up Nozzle Cracking In Babcock And Wilcox Plants GL 85-021 This GL was never issued. NA Info Info Applies only to Babcock and Wilcox designed plants GL 85-022 Potential For Loss Of Post-LOCA Recirculation Capability Due To Insulation Debris Blockage NA Info Page 71 of 109* = See last page for status code definition.

REV ITEM TITLE ADDITIONAL INFORMATION GL 86-001 Safety Concerns Associated With NA Pipe Breaks In The BWR Scram System GL 86-002 Technical Resolution of Generic NA Issue B-19 -Thermal Hydraulic Stability GL 86-003 Applications For License NA Amendments GL 86-004 Policy Statement On Engineering C Expertise On Shift -01 GL 86-005 Implementation Of TMI Action Item NA II.K.3.5, "Automatic Trip Of -Reactor Coolant Pumps" GL 86-006 Implementation Of TMI Action Item NA I1.K.3.5, "Automatic Trip of Reactor Coolant Pumps" Boiling Water Reactor Boiling Water Reactor Info TVA responded to GL 86-04 on May 29, 1986. TVA provides engineering expertise on shift in the form of a dedicated Shift Technical Advisor (STA)or an STA qualified Senior Reactor Operator.Applies only to Babcock and Wilcox designed plants Applies only to Combustion Engineering designed plants GL 86-007 Transmittal of NUREG-1190 NA Info Regarding The San Onofre Unit 1 --.Loss of Power and Water Hammer Event GL 86-008 Availability of Supplement 4 to NA Info NUREG-0933, "A Prioritization of _Generic Safety Issues" GL 86-009 Technical Resolution of Generic S N-1 Loop operation was addressed in original 1982 SER (4.4.7).Issue B-59, (N-i) Loop Operation

--.in BWRs and PWRs nj Unit 2 Action: Confirm Technical Specifications prohibit(N-i) Loop Operation.

REVISION 02 UPDATE: Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.TS LCO 3.4.4 requires that four Reactor Coolant System loops be operable and in operation during Modes 1 and 2.GL 86-010 Implementation of Fire Protection NA Info Requirements

_ -.Page 72 of 109* = See last page for status code definition.

ITEM GL 86-010, S1 TITLE REV NA Info ADDITIONAL INFORMATION Fire Endurance Test Acceptance Criteria for Fire Barrier Systems Used to Separate Redundant Safe Shutdown Trains Within the Same Fire Area GL 86-011 Distribution of Products Irradiated in Research GL 86-012 Criteria for Unique Purpose Exemption From Conversion From The Use of Heu Fuel GL 86-013 Potential Inconsistency Between Plant Safety Analyses and Technical Specifications GL 86-014 Operator Licensing Examinations GL 86-015 Information Relating To Compliance With 10 CFR 50.49,"Environmental Qualification of Electric Equipment Important To Safety For Nuclear Power Plants" GL 86-016 Westinghouse ECCS Evaluation Models GL 86-017 Availability of NUREG-1 169,"Technical Findings Related to Generic Issue C-8, BWR MSIC Leakage And Treatment Methods" GL 87-001 Public Availability Of The NRC Operator Licensing Examination Question Bank GL 87-002 Verification of Seismic Adequacy and of Mechanical and Electrical GL 87-003 Equipment in Operating Reactors, USI A-46 GL 87-004 Temporary Exemption From Provisions Of The FBI Criminal History Rule For Temporary Workers GL 87-005 Request for Additional Information on Assessment of License Measures to Mitigate and/or Identify Potential Degradation of Mark I Drywells NA Does not apply to power reactor.NA Does not apply to power reactor.NA Applies only to Babcock and Wilcox and Combustion Engineering

-_- designed plants NA Info NA Info NA Info NA NA NA Boiling Water Reactor Info Item was applicable only to units with operating license at the time the item was issued.NA Item was applicable only to units with operating license at the time the--- item was issued.NA Boiling Water Reactor Page 73 of 109* = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 87-006 Periodic Verification of Leak Tight NA Integrity of Pressure Isolation Valves GL 87-007 Information Transmittal of Final NA Rulemaking For Revisions To Operator Licensing

-10 CFR 55 And Confirming Amendments GL 87-008 Implementation of 10 CFR 73.55 NA Miscellaneous Amendments and Search Requirements GL 87-009 Sections 3.0 And 4.0 of Standard NA Tech Specs on Limiting Conditions For Operation And Surveillance Requirements GL 87-010 Implementation of 10 CFR 73.57, NA Requirements For FBI Criminal History Checks GL 87-011 Relaxation in Arbitrary NA Intermediate Pipe Rupture Requirements GL 87-012 Loss of Residual Heat Removal C While The Reactor Coolant System is Partially Filled GL 87-013 Integrity of Requalification NA Examinations At Non-Power Reactors GL 87-014 Operator Licensing Examinations NA Item was applicable only to units with operating license at the time the item was issued.Info Item was applicable only to units with operating license at the time the item was issued.Info Item was applicable only to units with operating license at the time the item was issued.Info This GL was superseded by GL 88-17.Does not apply to power reactor.Info GL 87-015 Policy Statement On Deferred NA Info Plants _GL 87-016 Transmittal of NUREG-1262,"Answers To Questions On Implementation of 10 CFR 55 On Operators' Licenses" GL 88-001 NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping GL 88-002 Integrated Safety Assessment Program II NA Info NA Boiling Water Reactor NA Item was applicable only to units with operating license at the time the_ -item was issued.Page 74 of 109* = See last page for status code definition.

ITEM GL 88-003 TITLE Resolution of GSI 93, "Steam Binding of Auxiliary Feedwater Pumps" REV CI ADDITIONAL INFORMATION TVA: letter June 3, 1988. NRC letters dated February 17, 1988 and July 20, 1988 NRC: SSER16 GL 88-004 Distribution of Gems Irradiated in Research Reactors GL 88-005 Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR plants NRC accepted approach in letter dated July 20, 1988, and reviewed response in Appendix EE of SSER16.Unit 2 Action: Procedures and hardware will be in place to ensure recognition of indications of steam binding and maintenance of system operability until check valves are repaired and back leakage stopped.NA Does not apply to power reactor.CI NRC acceptance letter dated August 8, 1990 for both units.06 Unit 2 Action: Implement program.REVISION 06 UPDATE: GL 88-006 Removal of Organization Charts from Technical Specification Administrative Control Requirements GL 88-007 Modified Enforcement Policy Relating to 10 CFR 50.49,"Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants" GL 88-008 Mail Sent or Delivered to the Office of Nuclear Reactor Regulation GL 88-009 Pilot Testing of Fundamentals Examination GL 88-010 Purchase of GSA Approved Security Containers The program has been implemented on Unit 2.NA Info C1 See Special Program for Environmental Qualification.

NA Info NA Boiling Water Reactor NA Info Page 75 of 109* = See last page for status code definition.

ITEM GL 88-011 TITLE NRC Position on Radiation Embrittlement of Reactor Vessel Material and its Impact on Plant Operations REV S 02 ADDITIONAL INFORMATION NRC acceptance letter dated June 29, 1989, for both units..Unit 2 Action: Submit Pressure Temperature curves.REVISION 02 UPDATE: Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.WCAP-17035-NP "Watts Bar Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation" was submitted with the TS.Info GL 88-012 Removal of Fire Protection Requirements from Technical Specification GL 88-013 Operator Licensing Examinations NA NA Info GL 88-014 Instrument Air Supply System Cl NRC letter dated July 26, 1990, closing the issue.Problems Affecting Safety-Related

-Equipment 04 Unit 2 Action: Complete Unit 2 implementation.

REVISION 04 UPDATE: The compressed air system is a common system at Watts Bar; therefore, the requirements for this GL have been satisfied for Unit 2.Watts Bar revised the response in a letter dated July 14, 1995.NRC letter dated July 27, 1995, stated that their conclusion as stated on July 26,1990, had not changed and that their effort was complete.Info Info GL 88-015 Electric Power Systems -Inadequate Control Over Design Process GL 88-016 Removal of Cycle-Specific Parameter Limits from Technical Specifications GL 88-017 Loss of Decay Heat Removal GL 88-018 Plant Record Storage on Optical Disks NA NA Cl NRC acceptance letter dated March 8, 1995 (Unit 1).Unit 2 Action: Implement modifications to provide RCS temperature, RV level and RHR system performance.

NA Info Page 76 of 109* = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 88-019 GL 88-020 Use of Deadly Force by Licensee Guards to Prevent Theft of Special Nuclear Material Individual Plant Examination for Severe Accident Vulnerabilities NA Does not apply to power reactor.S Unit 2 Action: Complete evaluation for Unit 2.06 REVISION 02 UPDATE: The Probabilistic Risk Assessment Individual Plant Examination Summary Report was submitted on February 9, 2010.REVISION 04 UPDATE: The Individual Plant Examination of External Events Design Report was submitted on April 30, 2010.REVISION 06 UPDATE: The NRC issued Requests for Additional Information (RAIs) on November 12, 2010.TVA responded to the RAls on December 17, 2010, and April 1, 2011.GL 89-001 Implementation of Programmatic and Procedural Controls for Radiological Effluent Technical Specifications GL 89-002 Actions to Improve the Detection of Counterfeit and Fraudulently Marketed Products NA Info C GL 89-02 did not require a response.01 WBN Unit 2 program for procurement and dedication of materials is based in part on and complies with the guidance of GL 89-02. The program is implemented through project procedures.

GL 89-003 Operator Licensing Examination Schedule GL 89-004 Guidelines on Developing Acceptable Inservice Testing Programs NA Info OV NRC reviewed in 3.9.6 of SSER14 (Unit 1).Unit 2 Action: Submit an ASME Section Xl Inservice Test Program for the first ten year interval six months before receiving an Operating License.GL 89-005 Pilot Testing of the Fundamentals Examination NA Info Page 77 of 109* = See last page for status code definition.

ITEM GL 89-006 TITLE Task Action Plan Item I.D.2 -Safety Parameter Display System-10 CFR 50.54(f)REV CI ADDITIONAL INFORMATION"Safety Parameter Display System" (SPDS) / "Requirements for Emergency Response Capability" -NRC reviewed in SSER5, SSER6, and 18.2.2 of SSER15.GL 89-007 Power Reactor Safeguards Contingency Planning for Surface Vehicle Bombs GL 89-008 Erosion/Corrosion-Induced Pipe Wall Thinning Unit 2 Action: Install SPDS and have it operational prior to start-up after the first refueling outage.C TVA: letter dated October 31, 1989 NRC: memo dated June 26, 1990 Cl Unit 1 Flow Accelerated Corrosion Program reviewed in IR 390/94-89--(February 1995).Unit 2 Actions:* Prepare procedure, and GL 89-009 ASME Section III Component NA Replacements GL 89-010 Safety-Related Motor-Operated Cl Valve Testing and Surveillance

.perform baseline inspections.

Item was applicable only to units with operating license at the time the item was issued.NRC accepted approach in September 14, 1990, letter and reviewed in Appendix EE of SSER16.Unit 2 Action: Implement pressure testing and surveillance program for safety-related MOVs, satisfying the intent of GL 89-10.NA Boiling Water Reactor GL 89-010 or Involves Main Steam Isolation GL 96-005 Valves GL 89-011 Resolution of Generic Issue 101,"Boiling Water Reactor Water Level Redundancy" GL 89-012 Operator Licensing Examination GL 89-013 Service Water System Problems Affecting Safety-Related Equipment NA Boiling Water Reactor NA Info CI 06 NRC letters dated July 9, 1990 and June 13, 1997, accepting approach.Unit 2 Actions: 1) Implement initial performance testing of the heat exchangers; and 2) Establish eddy current baseline data for the Containment Spray heat exchangers.

REVISION 06 UPDATE: NRC Inspection Report 391/2011-602 closed GL 89-013.Page 78 of 109* = See last page for status code definition.

ITEM TITLE GL 89-014 Line-Item Improvements in Technical Specifications

-Removal of 3.25 Limit on Extending Surveillance Intervals GL 89-015 Emergency Response Data System GL 89-016 Installation of a Hardened Wetwell Vent GL 89-017 Planned Administrative Changes to the NRC Operator Licensing Written Examination Process GL 89-018 Resolution of Unresolved Safety Issues A-17, "Systems Interactions in Nuclear Power Plants" GL 89-019 Request for Actions Related to Resolution of Unresolved Safety Issue A-47, "Safety Implication of Control Systems in LWR Nuclear Power Plants" Pursuant to 10 CFR 50.54(f)GL 89-020 Protected Area Long-Term Housekeeping GL 89-021 Request for Information Concerning Status of Implementation of Unresolved Safety Issue (USI) Requirements REV nf ADDITIONAL INFORMATION NA Info NA Info NA Boiling Water Reactor NA Info NA Info CI TVA responded by letter dated March 22, 1990. NRC acceptance letter-dated October 24, 1990, for both units.Unit 2 Action: Perform evaluation of common mode failures due to fire.NA Does not apply to power reactor.S TVA responded to GL 89-21 with the status of USIs for both units on-.November 29, 1989. NRC provided an assessment of WBN USI status on 06 May 1, 1990. The NRC assessment included a list of incomplete USIs for WBN. USIs were initially reviewed for WBN in the SER Appendix C. USIs were subsequently reviewed in SSER 15 Appendix C (June 1995) and SSER 16 (September 1995).Unit 2 actions:* Provide a status of WBN Unit 2 USIs.* Complete implementation of USIs.REVISION 02 UPDATE: Status of USIs was provided by Enclosure 2 of TVA letter dated September 26, 2008.The applicable USIs are either closed, deleted, or captured in either the SER Framework or the Generic Communications Framework, or they are part of the CAPs and SPs.Page 79 of 109* = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION

REVISION 06 UPDATE: Updated status of USIs was provided on January 25, 2011.GL 89-022 Potential For Increased Roof Loads and Plant Area Flood Runoff Depth At Licensed Nuclear Power Plants Due To Recent Change In Probable Maximum Precipitation Criteria Developed by the National Weather Service C TVA: letter dated December 16, 1981 Answer to informal question provided in TVA letter dated December 16, 1981, and subsequently included in FSAR. GL did not require a response.

No further action required.GL 89-023 NRC Staff Responses to Questions Pertaining to Implementation of 10 CFR Part 26 NA Info GL 90-001 Request for Voluntary Participation NA Info in NRC Regulatory Impact Survey -GL 90-002 Alternative Requirements for Fuel Assemblies in the Design Features Section of Technical Specifications GL 90-003 Relaxation of Staff Position in Generic Letter 83-28, Item 2.2 Part 2 "Vendor Interface for Safety-Related Components" GL 90-004 Request for Information on the Status of Licensee Implementation of GSIs Resolved with Imposition of Requirements or CAs GL 90-005 Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping NA Info NA Info C TVA responded on June 23, 1990 NA Info Page 80 of 109* = See last page for status code definition.

ITEM GL 90-006 TITLE Resolution of Generic Issues 70,"PORV and Block Valve Reliability," and 94, "Additional LTOP Protection for PWRs" REV ADDITIONAL INFORMATION S NRC letter dated January 9, 1991, accepted TVA's response for both_ _ units.02 Unit 2 Actions: 1) Revise operating instruction and surveillance procedure; and 2) Incorporate testing requirements in the Technical Specifications.

REVISION 02 UPDATE: Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.GL 90-007 GL 90-008 TS Surveillance Requirement 3.4.11.2 specifies the required testing of each PORV.Operator Licensing National NA Info Examination Schedule Simulation Facility Exemptions NA Info GL 90-009 Alternative Requirements for Snubber Visual Inspection Intervals and Corrective Actions GL 91 -001 Removal of the Schedule for the Withdrawal of Reactor Vessel Material Specimens from Technical Specifications GL 91-002 Reporting Mishaps Involving LLW Forms Prepared for Disposal GL 91-003 Reporting of Safeguards Events GL 91-004 Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle GL 91-005 Licensee Commercial-Grade Procurement and Dedication Programs GL 91-006 Resolution of Generic Issue A-30,"Adequacy of Safety-Related DC Power Supplies," Pursuant to 10 CFR 50.54(f)NA Info NA Info NA NA Item was applicable only to units with operating license at the time the item was issued.Info NA Info NA Info NA Item was applicable only to units with operating license at the time the---item was issued.Page 81 of 109* = See last page for status code definition.

REV ITEM TITLE ADDITIONAL INFORMATION GL 91-007 GI-23, "Reactor Coolant Pump Seal Failures" and Its Possible Effect on Station Blackout GL 91-008 Removal of Component Lists from Technical Specifications GL 91-009 Modification of Surveillance Interval for the Electrical Protective Assemblies in Power Supplies for the Reactor Protection System GL 91-010 Explosives Searches at Protected Area Portals GL 91-011 Resolution of Generic Issues A-48, "LCOs for Class 1 E Vital Instrument Buses", and 49,"Interlocks and LCOs for Class 1 E Tie Breakers," Pursuant to 10 CFR 50.54 GL 91-012 Operator Licensing National Examination Schedule GL 91-013 Request for Information Related to Resolution of Generic Issue 130,"Essential Service Water System Failures @ Multi-Unit Sites" GL 91-014 Emergency Telecommunications GL 91-015 Operating Experience Feedback Report, Solenoid-Operated Valve Problems at U.S. Reactors GL 91-016 Licensed Operators' and Other Nuclear Facility Personnel Fitness for Duty GL 91-017 Generic, Safety Issue 29, "Bolting Degradation or Failure in Nuclear Power Plants" GL 91-018 Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability NA Info NA Info NA Boiling Water Reactor NA Does not apply to power reactor.NA Item was applicable only to units with operating license at the time the item was issued.NA NA Info Addressed to specific (non-TVA) plants.NA Info NA Info NA Info NA Info NA GL 91-18 has been superseded by RIS 2005-20.Page 82 of 109* = See last page for status code definition.

ITEM TITLE REV NA Info ADDITIONAL INFORMATION GL 91-019 Information to Addressees Regarding New Telephone Numbers for NRC Offices Located in One White Flint North GL 92-001 Reactor Vessel Structural Integrity GL 92-002 Resolution of Generic Issue 79,"Unanalyzed Reactor Vessel (PWR) Thermal Stress During Natural Convection Cooldown" GL 92-003 Compilation of the Current Licensing Basis: Request for Voluntary Participation in Pilot Program GL 92-004 Resolution of the Issues Related to Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(f)GL 92-005 NRC Workshop on the Systematic Assessment of Licensee Performance (SALP) Program GL 92-006 Operator Licensing National Examination Schedule GL 92-007 Office of Nuclear Reactor Regulation Reorganization GL 92-008 Thermo-Lag 330-1 Fire Barriers C By letter dated May 11, 1994, for both units NRC confirmed TVA had_ --provided the information requested in GL 92-01. NRC issued GL 92-01 revision 1, supplement 1 on May 19, 1995. By letter dated July 26, 1996, NRC closed GL 92-01, Revision 1, Supplement 1 for both Watts Bar units.NA Info NA NA Info Boiling Water Reactor NA Info NA Info NA Info OV TVA configurations for Thermo-Lag 330-1 were reviewed in SSER18 and_ -. accepted in NRC letter dated January 6, 1998 (includes a supplemental SE).Unit 2 Actions: 1) Review Watts Bar design and installation requirements for Thermolag 330-1 fire barrier system and evaluate the Thermolag currently installed in Unit 2.GL 92-009 Limited Participation by NRC in the IAEA International Nuclear Event Scale GL 93-001 Emergency Response Data System Test Program NA 2) Remove and replace, as required, or prepare an approved deviation.

Info NA Addressed to specific plant(s).Page 83 of 109* o= See last page for status code definition.

ITEM TITLE REV NA Info ADDITIONAL INFORMATION GL 93-002 NRC Public Workshop on Commercial Grade Procurement and Dedication GL 93-003 Verification of Plant Records GL 93-004 Rod Control System Failure and Withdrawal of Rod Control Cluster Assemblies, 10 CFR 50.54(f)NA Info CO 06 NRC letter dated December 9, 1994, accepted TVA commitments for both units.Unit 2 Action: Implement modifications and testing.REVISION 06 UPDATE: GL 93-005 Line-Item Technical Specifications Improvements to Reduce Surveillance Requirements for Testing During Power Operation GL 93-006 Research Results on Generic Safety Issue 106, "Piping and the Use of Highly Combustible Gases in Vital Areas" GL 93-007 Modification of the Technical Specification Administrative Control Requirements for Emergency and Security Plans GL 93-008 Relocation of Technical Specification Tables of Instrument Response Time Limits GL 94-001 Removal of Accelerated Testing and Special Reporting Requirements for Emergency Diesel Generators GL 94-002 Long-Term Solutions and Upgrade of Interim Operating Recommendations for Thermal-Hydraulic Instabilities in BWRs GL 94-003 IGSCC of Core Shrouds in BWRs GL 94-004 Voluntary Reporting of Additional Occupational Radiation Exposure Data NA NRC Inspection Report 391/2011-604 closed GL 93-004.Info NA Info NA Item was applicable only to units with operating license at the time the_ .-item was issued.NA Item was applicable only to units with operating license at the time the---item was issued.NA Item was applicable only to units with operating license at the time the item was issued.NA Boiling Water Reactor NA Boiling Water Reactor NA Info Page 84 of 109* = See last page for status code definition.

ITEM REV TITLE ADDITIONAL INFORMATION GL 95-001 NRC Staff Technical Position on Fire Protection for Fuel Cycle Facilities GL 95-002 Use of NUMARC/EPRI Report TR-102348, "Guideline on Licensing Digital Upgrades," in Determining the Acceptability of Performing Analog-to-Digital Replacements under 10 CFR 50.59 GL 95-003 Circumferential Cracking of Steam Generator Tubes NA Does not apply to power reactor.NA Info Cl NRC acceptance letter dated May 16, 1997 (Unit 1) -Initial response for--- Unit 2 on September 7, 2007. TVA responded to a request for additional 06 information on December 17, 2007.Unit 2 Action: Perform baseline inspection.

REVISION 02 UPDATE: Unit 2 Action:* Perform baseline inspection.

Evaluate or repair as necessary.

On January 21, 2010, NRC issued the Safety Evaluation for the following Generic Letters: 1995-03, 1995-05, 1997-05, 1997-06, 2004-01, and 2006-01.REVISION 06 UPDATE: SSER22 contained the following for NRC Action: "Closed. NRC Letter dated January 21, 2010 (ADAMS Accession No.ML093631061)." 100% of the steam generator tubes have been inspected.

GL 95-004 Final Disposition of the Systematic Evaluation Program Lessons-Learned Issues NA Info Page 85 of 109* = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 95-005 Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking C No specific action or response required by the GL; TVA responded on--September 7, 2007.06-------------

.-. -. -.-.-. --. ....- .--.-.---

.- .----- ... ....-. ....--.-.-.-

-.-. ---. -. -.REVISION 02 UPDATE: On January 21, 2010, NRC issued the Safety Evaluation for the following Generic Letters: 1995-03, 1995-05, 1997-05, 1997-06, 2004-01, and 2006-01..............

.........

... ...... ...... ........ ..... ............ ...........

.REVISION 06 UPDATE: SSER22 contained the following for NRC Action: "Closed. NRC Letter dated January 21, 2010 (ADAMS Accession No.ML093631061)." GL 95-006 Changes in the Operator Licensing Program GL 95-007 Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves NA Info Cl Unit 1 SER for GL 95-07 dated Sept 15, 1999 06 Unit 2 Actions:* Perform evaluation for pressure locking and thermal binding of safety related power-operated gate valves, and* take corrective actions for those valves identified as being susceptible.

REVISION 03 UPDATE: April 1, 2010, letter committed to evaluate missing GL 89-10 motor-operated valves for susceptibility to pressure locking and thermal binding.REVISION 04 UPDATE: NRC letter dated July 29, 2010, provided RAIs on the GL.TVA letter dated July 30, 2010, answered the RAIs and provided the following commitments:

EDCRs 53292 and 53287 shall be implemented to eliminate the potential for pressure locking prior to startup.* Valves 2-FCV-63-25 and -26 will be evaluated for impact due to new parameters from the JOG Topical Report MPR 2524A prior to startup.Page 86 of 109* o= See last page for status code definition.

REV ITEM TITLE ADDITIONAL INFORMATION NRC issued the Safety Evaluation for GL 1995-007 on August 12, 2010.REVISION 06 UPDATE: TVA letter to NRC dated July 30, 2010, documented that none of the missing Watts Bar Unit 2 GL 89-10 valves are GL 95-07 valves.SSER22 contained the following for NRC Action: "Closed. NRC Letter dated August 12, 2010 (ADAMS Accession No.ML100190443)" NA Info GL 95-008 10 CFR 50.54(p) Process for Changes to Security Plans Without Prior NRC Approval GL 95-009 Monitoring and Training of Shippers and Carriers of Radioactive Materials GL 95-010 Relocation of Selected Technical Specifications Requirements Related to Instrumentation GL 96-001 Testing of Safety-Related Circuits GL 96-002 Reconsideration of Nuclear Power Plant Security Requirements Associated with an Internal Threat GL 96-003 Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits NA NA Info --------Info Cl TVA responded for both units on April 18, 1996.Unit 2 Action: Implement Recommendations.

NA Info Cl No response required 06 Unit 2 Actions: " Submit Pressure Temperature limits, and* similar to Unit 1, upon approval, incorporate into licensee-controlled document.REVISION 06 UPDATE: The Pressure and Temperature Limits Report (PTLR) was submitted via TVA to NRC letter dated February 2, 2010.The PTLR was incorporated in the system description for the Reactor Coolant System (WBN2-68-4001).

Page 87 of 109* = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 96-004 Boraflex Degradation in Spent Fuel Pool Storage Racks GL 96-005 Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves NA Item was applicable only to units with operating license at the time the_ item was issued.CI SE of TVA response to GL 96-05 dated July 21, 1999.Unit 2 Actions:* Implement the Joint Owner's Group recommended GL 96-05 MOV PV program, as described in Topical Report No. OG-97-018, and GL 96-006 Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions

begin testing during the first refueling outage after startup.C NRC letter dated April 6, 1999, accepting TVA response for Unit 1.06 Unit 2 Action: Implement modification to provide containment penetration relief.REVISION 02 UPDATE: NRC issued the Safety Evaluation for Generic Letter 1996-006 on January 21, 2010.REVISION 06 UPDATE: SSER22 contained the following for NRC Action: "Closed. NRC Letter dated January 21, 2010 (ADAMS Accession No.ML100130227)." Modification to provide containment penetration relief was implemented.

NRC Inspection Report 391/2011-603 closed GL 96-006.Item was applicable only to units with operating license at the time the item was issued.GL 96-007 Interim Guidance on NA Transportation of Steam Generators Page 88 of 109* = See last page for status code definition.

ITEM GL 97-001 TITLE Degradation of Control Rod Drive Mechanism Nozzle and Other Vessel Closure Head Penetrations

REV CI 06 ADDITIONAL INFORMATION NRC acceptance letter dated November 4, 1999 (Unit 1).Unit 2 Action: Provide a report to address the inspection program.REVISION 03 UPDATE: NRC issued the Safety Evaluation for Generic Letter 97-001 on June 30, 2010.....................................................................................................

REVISION 04 UPDATE: Corrected status from "OV" to "c'C due to NRC issuance of Safety Evaluation as noted in Revision 03 update.GL 97-002 Revised Contents of the Monthly Operating Report GL 97-003 Annual Financial Update of Surety Requirements for Uranium Recovery Licensees GL 97-004 Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps NA NA CI 06 REVISION 06 UPDATE: SSER22 contained the following for NRC Action: "Closed. NRC Letter dated June 30, 2010 (ADAMS Accession No.ML100539515)" Item was applicable only to units with operating license at the time the item was issued.Does not apply to power reactor.NRC acceptance letter dated June 17, 1998 (Unit 1) -Initial response for Unit 2 on September 7, 2007.Unit 2 Actions: " Install new sump strainers, and* perform other modification-related activities identical to Unit 1.REVISION 02 UPDATE: NRC issued the Safety Evaluation for Generic Letter 1997-004 on February 18, 2010.REVISION 06 UPDATE: See the REVISION 06 UPDATE for GL 04-002 for new commitments.

Page 89 of 109* = See last page for status code definition.

REV ITEM TITLE ADDITIONAL INFORMATION SSER22 contained the following for NRC Action: "Closed. NRC Letter dated February 18, 2010 (ADAMS Accession No.ML100200375)" GL 97-005 Steam Generator Tube Inspection CI NRC acceptance letter dated September 22, 1998 (Unit 1) -Initial Techniques

-response for Unit 2 on September 7, 2007.06 Unit 2 Action: Employ the same approach used on the original Unit 1 SGs. TVA responded to a request for additional information on December 17, 2007.REVISION 02 UPDATE: On January 21, 2010, NRC issued the Safety Evaluation for the following Generic Letters: 1995-03, 1995-05, 1997-05, 1997-06, 2004-01, and 2006-01.GL 97-006 Degradation of Steam Generator CI Internals 06 REVISION 06 UPDATE: SSER22 contained the following for NRC Action: "Closed. NRC Letter dated January 21, 2010 (ADAMS Accession No.ML093631061)" NRC acceptance letter dated October 19, 1999 (Unit 1) -Initial response for Unit 2 on September 7, 2007. TVA responded to a request for additional information on December 17, 2007.Unit 2 Action: Perform SG inspections during each refueling outage.REVISION 02 UPDATE: On January 21, 2010, NRC issued the Safety Evaluation for the following Generic Letters: 1995-03, 1995-05, 1997-05, 1997-06, 2004-01, and 2006-01.REVISION 06 UPDATE: SSER22 contained the following for NRC Action: "Closed. NRC Letter dated January 21, 2010 (ADAMS Accession No.ML093631061)" Page 90 of 109= See last page for status code definition.

REV ITEM TITLE ADDITIONAL INFORMATION GL 98-001 Year 2000 Readiness of Computer Systems at Nuclear Power Plants GL 98-002 Loss of Reactor Coolant Inventory and Associated Potential for Loss of Emergency Mitigation Functions While in a Shutdown Condition NA Item was applicable only to units with operating license at the time the--item was issued.Cl Initial response for Unit 2 on September 7, 2007.06 Unit 2 Actions: 1) Review the ECCS designs to ensure they do not contain design features which can render them susceptible to common-cause failures; and 2) document the results.REVISION 02 UPDATE: NRC issued the Safety Evaluation for Generic Letter 1998-002 on March 3, 2010.REVISION 03 UPDATE: NRC issued the Safety Evaluation for Generic Letter 98-002 on May 11, 2010. This letter noted that it superseded the SE issued by NRC on March 3, 2010.April 1, 2010, letter committed to ensure that the guidance added to the Unit 1 procedure as a result of the review of NRC GL 98-02 is incorporated into the Unit 2 procedures.

Specifically, when decreasing power, valve HCV-74-34, Refueling Water Return (normally locked closed valve) has a hold order placed with specific release criteria before entry into Mode 4 and to remove the hold order before entry into Mode 3 when returning to power.REVISION 06 UPDATE: SSER22 contained the following for NRC Action: "Closed. NRC Letter dated May 11, 2010 (ADAMS Accession No.ML101200155)" GL 98-003 NMSS Licensees' and Certificate Holders' Year 2000 Readiness Programs NA Does not apply to power reactor.Page 91 of 109* = See last page for status code definition.

REV ITEM TITLE ADDITIONAL INFORMATION GL 98-004 Potential for Degradation of the ECCS and the Containment Spray System After a LOCA Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment Cl NRC closure letter dated November 24, 1999 (Unit 1). -Initial response-for Unit 2 on September 7, 2007.06 Unit 2 Actions:* Install new sump strainers, and* perform other modification-related activities identical to Unit 1.REVISION 02 UPDATE: NRC issued the Safety Evaluation for Generic Letter 1998-004 on February 1, 2010.REVISION 06 UPDATE: See the REVISION 06 UPDATE for GL 04-002 for new commitments.

SSER22 contained the following for NRC Action: "Closed. NRC Letter dated February 1,2010 (ADAMS Accession No.ML100260594)" GL 98-005 Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds GL 99-001 Recent Nuclear Material Safety and Safeguards Decision on Bundling Exempt Quantities GL 99-002 Laboratory Testing of Nuclear Grade Activated Charcoal GL 03-001 Control Room Habitability NA Boiling Water Reactor NA Info NA Item was applicable only to units with operating license at the time the---item was issued.S Initial response for Unit 2 on September 7, 2007 06 Unit 2 Action: Incorporate TSTF-448 into Technical Specifications.

REVISION 02 UPDATE: NRC issued the Safety Evaluation for Generic Letter 2003-01 on February 1, 2010.Page 92 of 109* = See last page for status code definition.

REV ITEM TITLE ADDITIONAL INFORMATION Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.TS Surveillance Requirement 3.7.10.4 requires performance of a Control Room Envelope (CRE) unfiltered air inleakage test in accordance with the CRE Habitability Program.TS 5.7.2.20 provides for the CRE Habitability Program.These portions of the Unit 2 TS were based on the Unit 1 TS which incorporated TSTF-448 per Amendment 70 (NRC approved A70 on 10/08/2008).

REVISION 06 UPDATE: SSER22 contained the following for NRC Action: "Closed. NRC Letter dated February 1, 2010 (ADAMS Accession No.ML100270076)" NRC acceptance letter dated April 8, 2005 (Unit 1) -Initial response for Unit 2 on September 7, 2007.Unit 2 Action: Perform baseline inspection.

GL 04-001 Requirements for Steam Cl Generator Tube Inspection 06 REVISION 02 UPDATE: On January 21, 2010, NRC issued the Safety Evaluation for the following Generic Letters: 1995-03, 1995-05, 1997-05, 1997-06, 2004-01, and 2006-01.REVISION 06 UPDATE: SSER22 contained the following for NRC Action: "Closed. NRC Letter dated January 21, 2010 (ADAMS Accession No.ML093631061)" 100% of the steam generator tubes have been inspected.

Page 93 of 109* = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 04-002 Potential Impact of Debris OV NRC Audit Report dated February 7, 2007 (Unit 1) -Initial response for Blockage on Emergency

-Unit 2 on September 7, 2007.Recirculation During Design Basis 06 Accidents at PWRs Unit 2 Actions:* Install new sump strainers, and* perform other modification-related activities identical to Unit 1.REVISION 06 UPDATE: Additional TVA letters concerning GL 2004-02 were sent to the NRC on the following dates:-January 29, 2008,-May 19, 2008,-September 10, 2010,-March 4, 2011, and-April 29, 2011.The March 4, 2011, letter provided a response that superseded previous responses and commitments.

It provided the following new commitments:

-Unit 2 will install sump modifications per the requirements of Generic Letter (GL) 2004-02 prior to Unit 2 fuel load.-A confirmatory walkdown for loose debris will be performed on Unit 2 after containment work is completed and the containment has been cleaned. This walkdown will be completed prior to startup.-New throttle valves will be installed in the CVCS and SI injection lines to the RCS. The new valves will be opened sufficiently to preclude downstream blockage.-The current Unit 1 TVA protective coating program contains requirements for conducting periodic visual examinations of Coating Service Level I and Level II protective coatings.

The Unit 2 program will be the same.-Procedural controls will be put in place at WBN Unit 2 to ensure that potential quantities of post-accident debris are maintained within the bounds of the analyses and design bases that support ECCS and CSS recirculation functions.

-TVA will complete the WBN in-vessel downstream effects evaluation discussed in the supplemental response to Generic Letter 2004-02 following issuance of the final NRC Safety Evaluation Report (SER) for Topical Report No. WCAP-16793-NP, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous, and Chemical Debris in the Recirculating Fluid."-The design basis of the modified emergency sump strainer has been incorporated into the plant's current licensing basis. The WBN Unit 2 FSAR will be amended to include this information.

-Unit 1 and Unit 2 share a common protective coatings program.-Amendment 103 to the Unit 2 FSAR was submitted to the NRC on Page 94 of 109* = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION March 15, 2010. This amendment included the design basis of the modified emergency sump strainer.GL 06-001 Steam Generator Tube Integrity S Initial response for Unit 2 on September 7, 2007.and Associated Technical Specifications 06 Unit 2 Action: Incorporate TSTF-449 into Technical Specifications.

REVISION 02 UPDATE: On January 21, 2010, NRC issued the Safety Evaluation for the following Generic Letters: 1995-03, 1995-05, 1997-05, 1997-06, 2004-01, and 2006-01.Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.TS 5.7.2.12 is the Steam Generator (SG) Program. This program is implemented to ensure that SG tube integrity is maintained.

Unit 2 TS 5.7.2.12 was based on Unit 1 TS 5.7.2.12.

Unit 1 TS 5.7.2.1.12 was based on TSTF-449 (NRC approved Unit 1 TS A65 on 1/03/2006).

REVISION 06 UPDATE: SSER22 contained the following for NRC Action: "Closed. NRC Letter dated January 21, 2010 (ADAMS Accession No.ML093631061) (See Appendix HH)" The applicable item from SER22, Appendix HH for this item is Open item 6, "Verify implementation of TSTF-449. (TVA letter dated September 7, 2007, ADAMS Accession No. ML072570676)." TVA to NRC letter dated April 6, 2011 provided the following response to Open Item 6: "Amendment 65 to the Unit 1 TS revised the existing steam generator tube surveillance program and was modeled after TSTF-449, Rev. 4. The NRC approved Amendment 65 via letter dated November 3, 2006, 'Watts Bar Nuclear Plant, Unit 1 -Issuance of Amendment Regarding Steam Generator Tube Integrity (TS-05-10) (TAC No. MC9271).'

Revision 82 made the associated changes to the Unit 1 TS Bases.Developmental Revision A to the Unit 2 TS and TS Bases made the equivalent changes to the Unit 2 TS / TS Bases. Affected TS sections include the following:

LEAKAGE definition in 1.1, LCO 3.4.13 (RCS Operational LEAKAGE), LCO 3.4.17 (SG Tube Integrity), 5.7.2.12 (Steam Generator (SG) Program), and 5.9.9 (Steam Generator Tube Inspection Report).Developmental Revision A of the Unit 2 TS was submitted to the NRC via letter dated March 4, 2009, 'Watts Bar Nuclear Plant (WBN) Unit 2 -Page 95 of 109* = See last page for status code definition.

ITEM TITLE REV CI 06 GL 06-002 Grid Reliability and the Impact on Plant Risk and the Operability of Offsite Power ADDITIONAL INFORMATION Operating License Application Update,' (ADAMS Accession number ML090700378)." Initial response for Unit 2 on September 7, 2007.Unit 2 Action: Complete the two unit baseline electrical calculations and implementing procedures.

REVISION 02 UPDATE: NRC issued the Safety Evaluation for Generic Letter 2006-002 on January 20, 2010.REVISION 06 UPDATE: SSER22 contained the following for NRC Action: "Closed. NRC Letter dated January 21, 2010 (ADAMS Accession No.ML093631061) (See Appendix HH)" Note that the correct date and ADAMS Accession No. are January 20, 2010, and ML100080768, respectively.

TVA does not rely on Hemyc or MT materials to protect electrical and instrumentation cables or equipment that provide safe shutdown capability during a postulated fire.Unit 2 Action: Addressed in CAP/SP.The Fire Protection Corrective Action Program will ensure Unit 2 conforms with NRC requirements and applicable guidelines.

GL 06-003 Potentially Nonconforming Hemyc CI and MT Fire Barrier Configurations

-06 REVISION 02 UPDATE: NRC issued the Safety Evaluation for Generic Letter 2006-003 on February 25, 2010.---- -- ----- ---- ------04-- ----7 0- -- ---- ----- ----8---- ----- ---REVISION 06 UPDATE: SSER22 contained the following for NRC Action: "Closed. NRC Letter dated February 25, 2010 (ADAMS Accession No.ML1 00470398)" Page 96 of 109* o= See last page for status code definition.

ITEM GL 07-001 TITLE Inaccessible or Underground Power Cable Failures That Disable Accident Mitigation Systems or Cause Plant Transients REV CI 06 ADDITIONAL INFORMATION Initial response for Unit 2 on September 7, 2007.Unit 2 Action: Complete testing of four additional cables.REVISION 02 UPDATE: NRC issued the Safety Evaluation for Generic Letter 2007-001 on January 26, 2010.REVISION 04 UPDATE: NRC Inspection Report 391/2010-603 closed GL 2007-001.----------------------------------------------------------------------------------------------------

REVISION 06 UPDATE: The four additional cables passed the testing.SSER22 contained the following for NRC Action: "Closed. NRC Letter dated January 26, 2010 (ADAMS Accession No.ML100120052)" Initial response for Unit 2 on October 1, 2008.GL 08-001 Managing Gas Accumulation in 0 Emergency Core Cooling, Decay ---Heat Removal, and Containment 06 Spray Systems REVISION 02 UPDATE: Unit 2 Actions:-TVA will provide a submittal within 45 days of completion of the engineering for the ECCS, RHR, and CSS systems.-WBN Unit 2 will complete the required modifications and provide a submittal consistent with the information requested in the GL 90 days prior to fuel load.REVISION 06 UPDATE: The submittal was provided in TVA to NRC letter dated March 11, 2011.This submittal satisfied the above Unit 2 actions and generated the following new commitments:

-TVA will evaluate adopting the revised ISTS SR 3.5.2.3 (NUREG 1431)at WBN within 6 months of NRC approval of the Traveler.-Complete evaluation of CS pump 2A-A pipe chase horizontal suction Page 97 of 109* = See last page for status code definition.

REV ITEM TITLE ADDITIONAL INFORMATION piping for venting. Add a vent valve to this location or conduct periodic UT examinations if necessary.

(90 days prior to fuel load.)-Add vent valves to selected locations in the ECCS and RHRS piping to enhance filling and venting. (90 days prior to fuel load.)-Complete walk down survey of ECCS and RHRS piping and evaluate the piping for latent voids that could exceed 5% of the pipe cross sectional area. (90 days prior to fuel load.)-Operating procedures are being revised to improve instructions for filling and venting portions of the ECCS discharge pipe. (90 days prior to fuel load.)-Complete Preoperational tests on ECCS and RHRS systems to confirm Unit 1 operating experience showing no gas intrusion/accumulation issues. (90 days prior to fuel load.)-Periodic venting procedures used to meet SR 3.5.2.3 are being revised to require that, for an extended gas release, a report is entered into the Corrective Action Program. (90 days prior to fuel load.)NUREG-0737, I.A.1.1 NUREG-0737, I.A.1.2 NUREG-0737, I.A.1.3 NUREG-0737, I.A.2.1 NUREG-0737, I.A.2.3 NUREG-0737, I.A.3.1 Shift Technical Advisor NA Not applicable to WBN per SSER16.Shift Supervisor Responsibilities NA Not applicable to WBN per SSER16.Shift Manning C Closed in SSER16.Immediate Upgrade of RO and SRO Training and Qualifications Administration of Training Programs Revise Scope and Criteria for Licensing Exams C Closed in SSER16.C Closed in SSER16.C Closed in SSER16.Page 98 of 109* = See last page for status code definition.

ITEM NUREG-0737, I.B.1.2 TITLE Independent Safety Engineering Group REV ADDITIONAL INFORMATION OV LICENSE CONDITION

-Independent Safety Engineering Group (ISEG)(NUREG-0737, I.B.1.2)06 Resolved for Unit 1 only in SSER8.Unit 2 action: Implement the alternate ISEG that was approved for the rest of the TVA units including WBN Unit 1 by NRC on August 26, 1999. The function will be performed by the site engineering organizations.

REVISION 06 UPDATE: By letter of March 2, 1999, TVA proposed to eliminate the ISEG function from the fleet-wide nuclear organization.

NRC safety evaluation of August 26,1999 shows that the NRC accepted the elimination of the ISEG with alternate organizational responsibilities.

provided in TVA-NQA-PLN89A and TVA-NPOD89-A.

By letter of August 26, 1999, TVA revised Topical Report TVA-NPOD89-A, Rev 8 to describe the alternate organizations responsible for the management and operation of TVA's nuclear projects that replaced the ISEG function.The developmental Unit 2 TS were modeled after the Unit 1 TS. There is no reference to the ISEG.The current revision of TVA-NQA-PLN89-A (24A1) was written to include Unit 2.The current revision of TVA-NPOD89-A (18) was written to include Unit 2.Cl NRC reviewed in Appendix EE of SSER16.Unit 2 Action: Implement upgraded Emergency Operating Procedures, including validation and training.NUREG-0737, I.C.1 NUREG-0737, I.C.2 NUREG-0737, I.C.3 NUREG-0737, I.C.4 NUREG-0737, I.C.5 Short Term Accident and Procedure Review Shift and Relief Turnover Procedures Shift Supervisor Responsibility C Closed in SSER16.C Closed in SSER16.Control Room Access C Closed in SSER16.Feedback of Operating Experience C Closed in SSER16.Page 99 of 109* = See last page for status code definition.

ITEM NUREG-0737, I.C.6 TITLE REV C ADDITIONAL INFORMATION Verify Correct Performance of Operating Activities Closed in SSER16.NUREG-0737, I.C.7 NUREG-0737, I.C.8 NUREG-0737, I.D.1 NSSS Vendor Revision of Procedures Pilot Monitoring of Selected Emergency Procedures For Near Term Operating Licenses Control Room Design Review cI IR 50-390/391 85-08 closed this item for Unit 1, and NRC also reviewed in Appendix EE of SSER16.Unit 2 Action: Revise power ascension and emergency procedures which were reviewed by Westinghouse.

Cl IR 50-390/391 85-08 closed this item for Unit 1, and NRC also reviewed in Appendix EE of SSER16.Unit 2 Action: Pilot monitor selected emergency procedures for NTOL.Cl NRC reviewed in SSER5, SSER6, SSER15, and Appendix EE of SSER16.06 Unit 2 Actions:* Complete the CRDR process.* Perform rewiring in accordance with ECN 5982.* Take advantage of the completed Human Engineering reviews to ensure appropriate configuration for Unit 2 control panels.See CRDR Special Program.NUREG-0737, I.D.2 NUREG-0737, I.G.1 NUREG-0737, ll.B.1 Plant-Safety-Parameter-Display Console Training During Low-Power Testinc REVISION 06 UPDATE: SSER22 contained the following for NRC Action: "Closed in SSER22, Section 18.2" Cl NRC reviewed in SSER5, SSER6, and 18.2.2 of SSER15.Unit 2 Action: Install SPDS and have it operational prior to start-up after the first refueling outage.g C Closed in SSER16.Cl LICENSE CONDITION

-NUREG-0737, ll.B.1, "Reactor Coolant System Vents" -In the original SER, the NRC found TVA's commitment to install reactor coolant vents acceptable pending verification.

This was completed for Unit 1 only in SSER5 (IR 390/84-37).

Unit 2 Action: Verify installation of reactor coolant vents.Reactor Coolant Vent System Page 100 of 109* = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION NUREG-0737, ll.B.2 Plant Shielding CI NRC reviewed in Appendix EE of SSER16.NUREG-0737, ll.B.3 Post-Accident Sampling Unit 2 Action: Complete Design Review of EQ of equipment for spaces/systems which may be used in post accident operations.

S NRC reviewed in 9.3.2 of SSER16. TVA submitted a TS improvement to--eliminate requirements for the Post Accident Sampling System using the 02 Consolidated Line Item Improvement Process in a letter dated October 31, 2001.Unit 2 Actions: Unit 2 Technical Specifications will eliminate requirements for the Post-Accident Sampling System.REVISION 02 UPDATE: Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.Rev. 0 of the Unit 1 TS contained 5.7.2.6, "Post Accident Sampling." Amendment 34 to the Unit 1 TS (approved by the NRC on January 14, 2002) deleted 5.7.2.6, "Post Accident Sampling." The markup for Unit 2 Developmental Revision A noted that Unit 2 had deleted 5.7.2.6, "Post Accident Sampling" also.NUREG-0737, ll.B.4 NUREG-0737, ll.D.1 Training for Mitigating Core Damage Relief and Safety Valve Test Requirements C Closed in SSER16.Cl NRC reviewed in Technical Evaluation Report attached to Appendix EE of SSER15.Unit 2 Actions: 1) Testing of relief and safety valves;2) Reanalysis of fluid transient loads for pressurizer relief and safety valve supports and any required modifications;

3) Modifications to pressurizer safety valves, PORVs, PORV block valves and associated piping; and 4) Change motor operated block valves.CI The design was reviewed in the original 1982 SER and found acceptable

_ -pending confirmation of installation of the acoustic monitoring system. In SSER5 (IR 390/84-35), the staff closed the LICENSE CONDITION for Unit 1 only.Unit 2 Action: Verify installation of the acoustic monitoring system to PORV to indicate position.NUREG-0737, ll.D.3 Valve Position Indication Page 101 of 109* a= See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION NUREG-0737, II.E.1.1 NUREG-0737, II.E.1.2 Auxiliary Feedwater System Evaluation, Modifications Auxiliary Feedwater System Initiation and Flow CI Reviewed in Appendix EE of SSER16.Unit 2 Action: Perform Auxiliary Feedwater System analysis as it pertains to system failure and flow rate.CI NRC: IR 50-390/84-20 and 50-391/84-16; letters dated__-. March 29, 1985, and October 31, 1995; SSER 16 Unit 2 Actions: " Complete procedures, and* qualification testing.Cl NRC: letters dated March 29, 1985, and October 31, 1995; SSER 16 NUREG-0737, I1.E.3.1 Emergency Power For Pressurizer Heaters Reviewed in original 1982 SER.Unit 2 Action: Implement procedures and testing.NUREG-0737, I1.E.4.1 NUREG-0737, I1.E.4.2 Dedicated Hydrogen Penetrations C NRC: IR 50-390/83-27 and 50-391/83-19; SER (NUREG-0847)

Containment Isolation Dependability S TVA: letters dated October 29, 1981, and February 25, 1985 02 NRC: letters dated March 29, 1985, July 12, 1990 and October 31, 1995;SSER 16.OUTSTANDING ISSUE for NRC to complete review of information provided by TVA to address Containment Purging During Normal Plant Operation LICENSE CONDITION

-Containment isolation dependability In the original 1982 SER, NRC concluded that WBN met all the requirements of NUREG-0737, item II.E.4.2 except subsection (6)concerning containment purging during normal operation.

In SSER3, the outstanding issue was closed and the LICENSE CONDITION was left open.NRC completed the review and issued a Technical Evaluation Report for both units on July 12, 1990. NRC concluded that the isolation valves can close against the buildup of pressure in the event of a design basis accident if the lower containment isolation valves are physically blocked to an opening angle of 50 degrees or less. (SSER5)Unit 2 Action: Reflect valve opening restriction in the Technical Specifications.

....................................................................................................

Page 102 of 109* = See last page for status code definition.

ITEM TITLE REV NUREG- Accident-Monitoring Cl 0737, Instrumentation

-Noble Gas II.F.1.2.a.

ADDITIONAL INFORMATION REVISION 02 UPDATE: Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.TS Surveillance Requirement 3.6.3.7 requires verification that the valves are "blocked to restrict the valve from opening > 50 degrees." Reviewed in SSER9.Unit 2 Actions: Install Noble gas, Iodine / particulate sampling, and Containment High Range Monitors.Unit 2 Action: Install Noble gas monitor for Unit 2.Reviewed in SSER9.Unit 2 Actions: Install Noble gas, Iodine / particulate sampling, and Containment High Range Monitors.NUREG-0737, II.F.1.2.b.

NUREG-0737, I1.F.1.2.c.

NUREG-0737, II.F.1.2.d.

NUREG-0737, II.F.1.2.e.

Accident-Monitoring CI Instrumentation

-_ -.Iodine/Particulate Sampling Unit 2 Action: Install Iodine / particulate sampling monitor for Unit 2.Accident-Monitoring CI Reviewed in SSER9.Instrumentation

-Containment

___High Range Monitoring Unit 2 Actions: Install Noble gas, Iodine / particulate sampling, and Containment High Range Monitors.Unit 2 Action: Install high range in-containment monitor for Unit 2.Accident-Monitoring CO Reviewed in SSER9.Instrumentation

-Containment Pressure 06 Unit 2 Action: Verify installation of containment pressure indication.

REVISION 06 UPDATE: NRC Inspection Report 391/2011-604 closed NUREG-0737, II.F.1.2.d.

Accident-Monitoring Cl Reviewed in SSER9.Instrumentation

-Containment

___Water Level Unit 2 Action: Verify installation of containment water level monitors.Page 103 of 109* = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION NUREG-0737, II.F.1.2.f.

Accident-Monitoring Instrumentation

-Containment Hydrogen CO Reviewed in SSER9.06 Unit 2 Action: Verify installation of containment hydrogen accident monitoring instrumentation.

NUREG-0737, II.F.2 Instrumentation For Detection of Inadequate Core-Cooling REVISION 06 UPDATE: NRC Inspection Report 391/2011-604 closed NUREG-0737, II.F.1.2.F.

0 LICENSE CONDITION

-Detectors for Inadequate core cooling (II.F.2)In the original SER, the review of the ICC instrumentation was incomplete.

The January 24, 1992, letter superseded the previous responses on this issue. TVA letter for Units 1 and 2 dated January 24, 1992, committed to install Westinghouse ICCM-86 and associated hardware.

NRC completed the review for Units 1 and 2 in SSER10. For Unit 2 due to obsolescence of the ICCM-86 system, TVA intends to install the Westinghouse Common Q Post-Accident Monitoring System.Unit 2 Action: Install Westinghouse Common Q PAM system.NUREG-0737, II.G.1 Power Supplies For Pressurizer Relief Valves, Block Valves and Level Indicators CI Reviewed in original 1982 SER and 8.3.3 of SSER7.06 Unit 2 Action: Implement modifications such that PORVS and associated Block Valves are powered from same train but different buses.REVISION 06 UPDATE: Modifications were implemented such that PORVS and associated Block Valves are powered from same train but different buses.NUREG-0737, I1.K.1.5 NUREG-0737, II.K.1.10 NUREG-0737, II.K.1.17 Review ESF Valves C NRC: letter dated March 29, 1985; SSER 16 Operability Status Cl Unit 2 Action: Confirm multi-unit operation will have no impact on--_ administrative procedures with respect to operability status.Trip Per Low-Level B/S C NRC: letter dated March 29, 1985; SSER 16 Page 104 of 109* P= See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION NUREG-0737, II.K.2.13 Effect of High Pressure Injection for Small Break LOCA With No Auxiliary Feedwater NUREG-0737, I1.K.2.17 Voiding in the Reactor Coolant System C LICENSE CONDITION

-Effect of high pressure injection for small break--. LOCA with no auxiliary feedwater (NUREG-0737, II.K.2.13)

In SSER4, the staff concluded that there was reasonable assurance that vessel integrity would be maintained for small breaks with an extended loss of all feedwater and that the USI A-49, "Pressurized Thermal Shock," review did not have to be completed to support the full-power license.They considered this condition resolved.C LICENSE CONDITION

-Voiding in the reactor coolant system_ -. (NUREG-0737, I1.K.2.17)

The staff reviewed the generic resolution of this license condition in SSER4 and approved the study in question, thereby resolving this license condition.

C Reviewed in SSER5 and resolved based on NRC conclusion that there is_ -_ no need for an automatic PORV isolation system (NRC letter dated June 29, 1990).C Reviewed in SSER5 and resolved based on NRC conclusion that there is_ -_ no need for an automatic PORV isolation system (NRC letter dated June 29, 1990).C (Action from GL 82-16) -NRC reviewed in Appendix EE of SSER16.NUREG-0737, II.K.3.1 NUREG-0737, I1.K.3.2 NUREG-0737, I1.K.3.3 Auto PORV Isolation Report on PORV Failures Reporting SV/RV Failures/Challenges 06 Unit 2 Action: Include, as necessary, in Technical Specifications submittal.

REVISION 02 UPDATE: Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.Rev. 0 of the Unit 1 TS contained 5.9.4 (Monthly Operating Reports) which implemented the above commitment for Unit 1.Amendment 57 to the Unit 1 TS (approved by the NRC on March 21, 2005) deleted this section of the TS.The markup for Unit 2 Developmental Revision A noted that Unit 2 will apply this change, and the Unit 2 TS will contain no requirement for Monthly Operating Reports.REVISION 06 UPDATE: SSER22 contained the following for NRC Action: "Closed in SSER22, Section 13.5.3." Page 105 of 109* = See last page for status code definition.

REV ITEM NUREG-0737, II.K.3.5 TITLE Auto Trip of RCPS ADDITIONAL INFORMATION CI Reviewed in 15.5.4 of original 1982 SER; became License Condition 35.The staff determined that their review of Item II.K.3.5 did not have to be completed to support the full power license and considered this license condition resolved in SSER4. The item was further reviewed in Appendix EE of SSER16.Unit 2 Action: Implement modifications as required.NUREG-0737, 11.K.3.9 PID Controller Cl Reviewed in original 1982 SER.06 Unit 2 Action: Set the derivative time constant to zero.REVISION 06 UPDATE: The derivative time constant was set to zero.S NRC: letter dated October 31, 1995; SSER 16 NUREG-0737, I1.K.3.10 Anticipatory Trip at High Power 02 Unit 2 Action: Unit 2 Technical Specifications and surveillance procedures will address this issue.REVISION 02 UPDATE: Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.Items 14.a. (Turbine Trip -Low Fluid Oil Pressure) and 14.b. (Turbine Trip -Turbine Stop Valve Closure) of TS Table 3.3.1-1 are the trips of interest.

The table and the Bases for these items state that below the P-9 setpoint, these trips do not actuate a reactor trip.Per item 16.d. (Power Range Neutron Flux, P-9) of TS Table 3.3.1-1, the Nominal Trip Setpoint for P-9 is"50% RTP" and the Allowable Value is"< 52.4% RTP." NUREG-0737, II.K.3.12 NUREG-0737, II.K.3.17 Confirm Existence of Anticipatory Reactor Trip Upon Turbine Trip Report On Outage of Emergency Core Cooling System C Closed in SSER16.C LICENSE CONDITION

-Report on outage of emergency core cooling system (NUREG-0737, I1.K.3.17)

In the original 1982 SER, the NRC accepted TVA's commitment to develop and implement a plan to collect emergency core cooling system outage information.

In SSER3, the staff accepted a revised commitment from an October 28, 1983, letter to participate in the nuclear power reliability data system and comply with the requirements of 10 CFR 50.73.Page 106 of 109* = See last page for status code definition.

REV ITEM TITLE ADDITIONAL INFORMATION NUREG-0737, II.K.3.25 Power On Pump Seals C NRC reviewed and closed in IR 390/84-35 based on Diesel Generator-- (DG) power to pump sealing cooling system.06 Unit 2 Action: Ensure DG power is provided to pump sealing cooling system.REVISION 06 UPDATE: It was confirmed that DG power is provided to pump sealing cooling system.NRC Inspection Report 391/2010-605 closed NUREG-0737, II.K.3.25.

NUREG-0737, I1.K.3.30 Small Break LOCA Methods C TVA: letter dated October 29, 1981 06 NRC: letters dated March 29, 1985, and July 24, 1986; SSER 16 The staff determined in SSER4 that their review of Items I1.K.3.30 and I1.K.3.31 did not have to be completed to support the full-power license and considered this LICENSE CONDITION resolved in SSER4. In SSER5, the staff further reviewed responses to these items, and concluded that the Units 1 and 2 FSAR methods and analysis met the requirements of II.K.3.30 and I1.K.3.31.

This item was further reviewed in Appendix EE of SSER16.Unit 2 Action: Complete analysis for Unit 2.REVISION 06 UPDATE: The analysis has been completed for Unit 2.NRC Inspection Report 391/2011-603 closed NUREG-0737, II.K.3.30.

C The staff determined in SSER4 that their review of Items II.K.3.30 and_l__ II.K.3.31 did not have to be completed to support the full-power license 06 and considered this LICENSE CONDITION resolved in SSER4. In SSER5, the staff further reviewed responses to these items, and concluded that the Units 1 and 2 FSAR methods and analysis met the requirements of I1.K.3.30 and II.K.3.31.

This item was further reviewed in Appendix EE of SSER16.NUREG-0737, I1. K.3.31 Plant Specific Analysis Unit 2 Action: Complete analysis for Unit 2.----------------------------------------------------------------------------------------------------

REVISION 06 UPDATE: Page 107 of 109* = See last page for status code definition.

REV ITEM TITLE ADDITIONAL INFORMATION The analysis has been completed for Unit 2.NRC Inspection Report 391/2011-603 closed NUREG-0737, II.K.3.31.

NUREG-0737, III.A.1.1 Emergency Preparedness, Short Term NUREG-0737, III.A.1.2 Upgrade Emergency Support Facilities NUREG-0737, III.A.2 Emergency Preparedness C LICENSE CONDITION

-Emergency Preparedness (NUREG-0737, III.A.1, III.A.2)The NRC review of Emergency Preparedness in SSER1 3 superseded the review in the original 1982 SER. In SSER1 3, the staff concluded that the WBN Radiological Emergency Plan (REP) provided an adequate planning basis for an acceptable state of onsite emergency preparedness, and the LICENSE CONDITION was deleted. The NRC completed the review of the REP in SSER20.C LICENSE CONDITION

-Emergency Preparedness (NUREG-0737, III.A.1, III.A.2)The NRC review of Emergency Preparedness in SSER13 superseded the review in the original 1982 SER. In SSER13, the staff concluded that the WBN Radiological Emergency Plan (REP) provided an adequate planning basis for an acceptable state of onsite emergency preparedness, and the LICENSE CONDITION was deleted. The NRC completed the review of the REP in SSER20.C LICENSE CONDITION

-Emergency Preparedness (NUREG-0737, III.A.1, III.A.2)The NRC review of Emergency Preparedness in SSER13 superseded the review in the original 1982 SER. In SSER13, the staff concluded that the WBN Radiological Emergency Plan (REP) provided an adequate planning basis for an acceptable state of onsite emergency preparedness, and the LICENSE CONDITION was deleted. The NRC completed the review of the REP in SSER20.S Resolved for Unit I only in SSER10; reviewed in Appendix EE of SSER16.02 Unit 2 Actions: Include the waste gas disposal system in the leakage reduction program and incorporate in Unit 2 Technical Specifications.

REVISION 02 UPDATE: Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.TS 5.7.2.4 is the Primary Coolant Sources Outside Containment program.This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable.

This program includes the "Waste Gas" system.NUREG-0737, II1.D.1.1 Primary Coolant Outside Containment Page 108 of 109* = See last page for status code definition.

ITEM TITLE REV NUREG- In-Plant Iodine Radiation Cl 0737, Monitoring

_ -II1.D.3.3 NUREG- Control-Room Habitability Cl 0737, II1.D.3.4 06 ADDITIONAL INFORMATION NRC reviewed in Appendix EE of SSER16.Unit 2 Action: Complete modifications for Unit 2.TVA: letter dated October 29, 1981 NRC: SSER16 NRC reviewed in SER and in Appendix EE of SSER16.Unit 2 Action: Complete with CRDR completion.

REVISION 06 UPDATE: SSER22 contained the following for NRC Action: "Closed in SSER22, Section 6.4"...........................................................................................................................................

STATUS CODE DEFINITIONS C: CLOSED: Previous staff review of NUREG-0847 and/or supplements has closed the item either for both units at WBN or explicitly for WBN Unit 2.CI: CLOSED/IMPLEMENTATION:

Staff has approved either for both units at WBN or explicitly for WBN Unit 2; there is no change to the approved design; and implementation is recommended through Regional Inspection.

CO: CLOSED -OPEN: Staff has approved closure of the item; however, TVA actions remain to be completed.

CT: CLOSED/TECHNICAL SPECIFICATIONS:

Item has been approved either for both units at WBN or explicitly for WBN Unit 2; however, a change to the original approval requires submittal of the Technical Specifications and staff review.NA: NOT APPLICABLE:

Justification as to why a section / subsection is not applicable is provided in the ADDITIONAL INFORMATION column.0: OPEN: No action or documentation is provided that shows the staff has reviewed the item for WBN Unit 2.OT: OPEN/TECHNICAL SPECIFICATIONS:

No action or documentation is provided that shows the staff has reviewed the item for WBN Unit 2, and the resolution is through submittal of a Technical Specification.

OV: OPENNALIDATION:

The proposed approach has been approved for Watts Bar Unit 1; the same approach is proposed for use on WBN Unit 2 without change.S: SUBMITTED:

Information has been submitted, and is under review by NRC staff.Page 109 of 109* = See last page for status code definition.

ML11213A132

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