ML11213A132

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)

Text

Enclosure 3 Generic Communications - Master Table

GENERIC COMMUNICATIONS: MASTER TABLE ITEM TITLE REV ADDITIC)NAL INFORMATION B 71-002 PWR Reactor Trip Circuit Breakers NA Addressed to specific plant(s).

NA Addressed to specific plant(s).

B 71-003 Catastrophic Failure of Main Steam Line Relief Valve Headers NA Addressed to specific plant(s).

B 72-001 Failed Hangers for Emergency Core Cooling System Suction Header B 72-002 Simultaneous Actuation of a NA Addressed to specific plant(s).

Safety Injection Signal on Both Units of a Dual Unit Facility B 72-003 Limitorque Valve Operator Failuress NA Addressed to specific plant(s).

B 73-001 Faulty Overcurrent Trip Delay C TVA: letter dated April 4, 1973 Device in Circuit Breakers for Engineered Safety Systems NRC: IR 390/391 75-5 B 73-002 Malfunction of Containment Purge C TVA: letter dated August 22, 1973 Supply Valve Switch NRC: IR 390/391 75-5 B 73-003 Defective Hydraulic Snubbers and C TVA: letter dated February 7, 1985 Restraints NRC: IR 390/391 85-08 B 73-004 Defective Bergen-Patterson C TVA: memo dated February 7, 1985 Hydraulic Shock Absorbers NRC: IR 390/391 85-08 B 73-005 Manufacturing Defect in BWR NA Boiling Water Reactor Control Rods B 73-006 Inadvertent Criticality in a BWR NA Boiling Water Reactor B 74-001 Valve Deficiencies C TVA: letter dated April 15, 1974 NRC: IR 390/391 75-5 B 74-002 Truck Strike Possibility NA Info Page 1 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 74-003 Failure of Structural or Seismic CI TVA: memo dated January 22, 1985 Support Bolts on Class I Components 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 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 NA Boiling Water Reactor Relief Valves B 74-005 Shipment of an Improperly NA Does not apply to power reactor.

Shielded Source B 74-006 Defective Westinghouse Type W- C TVA: letter dated October 18, 1974 2 Control Switch Component NRC: IR 390/391 75-6 B 74-007 Personnel Exposure - Irradiation NA Does not apply to power reactor.

Facility B 74-008 Deficiency in the ITE Molded Case C TVA: letter dated August 21, 1974 Circuit Breakers, Type HE-3 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 REV ADDITIONAL INFORMATION B 74-014 BWR Relief Valve Discharge to NA Boiling Water Reactor Suppression Pool B 74-015 Misapplication of Cutler-Hammer C TVA: letter dated May 5, 1975 Three Position Maintained Switch Model No. 10250T 06 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 C TVA: letter dated January 2, 1975 Colt Industries (Fairbanks-Morse)

Diesel-Generators NRC: IR 390/391 75-3 B 75-001 Through-Wall Cracks in Core NA Boiling Water Reactor Spray Piping at Dresden-2 B 75-002 Defective Radionics Radiograph NA Does not apply to power reactor.

Exposure Devices and Source Changers B 75-003 Incorrect Lower Disc Spring and CI TVA: letter dated May 16, 1975 Clearance Dimension in Series 8300 and 8302 ASCO Solenoid NRC: IR 390/391 75-6 Valves NRC accepted in IR 50-390/75-6 and 50-391/75-6 (August 21, 1975).

Unit 2 Action:

Modify valves not modified at factory.

B 75-004 Cable Fire at BFNPP CI 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.

ITEM TITLE REV ADDITIONAL INFORMATION B 75-005 Operability of Category I Hydraulic Cl TVA: letter dated June 16, 1975 Shock and Sway Suppressors 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- Cl TVA: letter dated July 31, 1975 2 Control Switches 06 NRC: IR 390/85-25 and 391/85-20 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 NA Does not apply to power reactor.

Shipment B 75-008 PWR Pressure Instrumentation 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 NA Boiling Water Reactor Failure Page 4 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 76-002 Relay Coil Failures - GE Types Cl Unit 2 Action:

HFA, HGA, HKA, HMA Relays Repair or replace relays before preoperational tests.

B 76-003 Relay Malfunctions - GE Type C TVA: letter dated May 17, 1976 STD Relays NRC: IR 390/391 76-6 B 76-004 Cracks in Cold Worked Piping at NA Boiling Water Reactor BWRs B 76-005 Relay Failures - Westinghouse C TVA: letter dated June 7, 1976 BFD Relays NRC: IR 390/391 85-08 B 76-006 Diaphragm Failures in Air C TVA: memo dated January 25, 1985 Operated Auxiliary Actuators for Safety/Relief Valves NRC: IR 390/391 85-08 B 76-007 Crane Hoist Control Circuit C TVA: letter dated October 29, 1976 Modifications NRC: IR 390/391 85-08 B 76-008 Teletherapy Units NA Does not apply to power reactor.

B 77-001 Pneumatic Time Delay Relay C TVA: letter dated July 1, 1977 Setpoint Drift NRC: IR 390/391 85-08 B 77-002 Potential Failure Mechanism in C TVA: letter dated November 11, 1977 Certain Westinghouse AR Relays with Latch Attachments NRC: IR 390/391 85-08 B 77-003 On-Line Testing of the CI Unit 2 Action:

Westinghouse Solid State Protection System Include necessary periodic testing in test procedures.

B 77-004 Calculation Error Affecting The S TVA: letter dated January 23, 1978 Design Performance of a System ---

for Controlling pH of Containment 02 NRC: IR 390/78-11 and 391/78-09 Sump Water Following a LOCA 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 TITLE ADDITIONAL INFORMATION concentration is within a specified range.

B 77-005 Electrical Connector Assemblies C TVA: letter dated January 17, 1978 and B 77-005 A NRC: IR 390/78-11 and 391/78-09 B 77-006 Potential Problems with C Item was applicable only to units with operating license at the time the Containment Electrical Penetration item was issued.

Assemblies NRC: IR 390/391 85-08 B 77-007 Containment Electrical Penetration C TVA: letter dated January 20, 1978 Assemblies at Nuclear Power Plants Under Construction NRC: IR 390/78-11 and 391/78-09 B 77-008 Assurance of Safety and C Item concerns a multi-unit issue that was completed for both units.

Safeguards During an Emergency

- Locking Systems TVA: letter dated March 1, 1978 NRC: IR 390/78-11 and 391/78-09 B 78-001 Flammable Contact - Arm C TVA: letter dated March 20, 1978 Retainers in GE CR120A Relays NRC: IR 390/78-11 and 391/78-09 B 78-002 Terminal Block Qualification C TVA: letter dated March 1, 1978 NRC: IR 390/78-11 and 391/78-09 B 78-003 Potential Explosive Gas Mixture NA Boiling Water Reactor Accumulations Associated with BWR Offgas System Operations B 78-004 Environmental Qualification of CI TVA: letter dated December 19, 1978 Certain Stem Mounted Limit Switches Inside Reactor NRC: IR 390/82-13 and 391/82-10 Closed to EQ Program Containment 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 REV ADDITIONAL INFORMATION B 78-007 Protection Afforded by Air-Line NA Item was applicable only to units with operating license at the time the Respirators and Supplied-Air item was issued.

Hoods B 78-008 Radiation Levels from Fuel NA Item was applicable only to units with operating license at the time the Element Transfer Tubes item was issued.

NRC: IR 390/391 85-08 B 78-009 BWR Drywell Leakage Paths NA Boiling Water Reactor Associated with Inadequate Drywell Closures B 78-010 Bergen-Patterson Hydraulic Shock C TVA: letter dated August 14, 1978 Suppressor Accumulator Spring Coils NRC: IR 390/78-22 and 391/78-19 B 78-011 Examination of Mark I NA Boiling Water Reactor Containment Torus Welds B 78-012 Atypical Weld Material in Reactor C TVA: Westinghouse letter dated October 29, 1979 Pressure Vessel Welds NRC: IR 390/391 81-04 B 78-013 Failures in Source Heads Kay NA Does not apply to power reactor.

Ray, Inc. Gauges Models 7050, 7050B, 7051,7051B, 7060, 7060B, 7061 and 7061B B 78-014 Deterioration of Buna-N NA Boiling Water Reactor Components in ASCO Solenoids B 79-001 Environmental Qualification of C NRC: IR 390/80-06 and 391/80-05 Class lE Equipment B 79-002 Pipe Support Base Plate Designs CI NRC review of HAAUP Program in NUREG-1232, SSER6, and SSER8.

Using Concrete Expansion Anchor Bolts 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.

B 79-003 Longitudinal Weld Defects in C TVA: letter dated July 16, 1981 ASME SA-312 Type 304 SS Pipe Spools Manufactured by NRC: IRs 390/82-21 and 391/82-17; 390/84-35 and 391/84-33 Youngstown Welding &

Engineering Page 7 of 109 * = See last page for status code definition.

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

Complete preservice preventive maintenance on AK-2 Circuit Breakers.

[WAS "NOTE 3."]

REVISION 06 UPDATE:

It has been confirmed that AK-2 Circuit Breakers are not used on Unit 2.

B 79-010 Requalification Training Program NA Item was applicable only to units with operating license at the time the Statistics - - item was issued.

B 79-011 Faulty Overcurrent Trip Device in C TVA: letter dated July 20, 1979 Circuit Breakers for Engineering Safety Systems NRC: IR 390/79-30 and 391/79-25 B 79-012 Short Period Scrams at BWR NA Boiling Water Reactor Facilities B 79-013 Cracking in Feedwater Piping 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 TITLE REV ADDITIONAL INFORMATION B 79-014 Seismic Analysis for As-Built CI NRC review of HAAUP Program in NUREG-1232, SSER6, and SSER8.

Safety-Related Piping Systems 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 TITLE REV ADDITIONAL INFORMATION B 79-021 Temperature Effects on Level C Reviewed in 7.2.5 of both the original 1982 SER and SSER14.

Measurements 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.

NRC Inspection Report 391/2010-605 closed B 79-021.

B 79-022 Possible Leakage of Tubes of NA Does not apply to power reactor.

Tritium Gas Used in Time Pieces for Luminosity NRC: IR 390/80-06 and 391/80-05 B 79-023 Potential Failure of Emergency C TVA: letter dated October 29, 1979 Diesel Generator Field Exciter Transformer NRC: IR 390/80-06 and 391/80-05 B 79-024 Frozen Lines 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.

B 79-025 Failures of Westinghouse BFD C TVA: letter dated January 4, 1980 Relays in Safety-Related Systems NRC: IR 390/80-03 and 391/80-02 B 79-026 Boron Loss from BWR Control NA Boiling Water Reactor Blades TVA responded to the Bulletin on March 1, 1982. Reviewed in 7.5.3 of the B 79-027 Loss of Non-Class 1E I & C Power Cl System Bus During Operation original 1982 SER.

Unit 2 Action: Issue appropriate emergency procedures.

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

ITEM TITLE REV ADDITIONAL INFORMATION B 79-028 Possible Malfunction of NAMCO C TVA: letter dated April 1, 1993 Model EA1 80 Limit Switches at Elevated Temperatures NRC: IR 390/391 93-32 NA Boiling Water Reactor B 80-001 Operability of ADS Valve Pneumatic Supply NA Boiling Water Reactor B 80-002 Inadequate QA for Nuclear Supplied Equipment B 80-003 Loss of Charcoal from Standard C TVA: letter dated March 21, 1980 Type II, 2 Inch, Tray Adsorber Cells NRC: IR 390/80-15 and 391/80-12 B 80-004 Analysis of a PWR Main Steam CI IR 50-390/85-60 and 50-391/85-49 (December 6, 1985) required Line Break with Continued completion of actions that included determination of temperature profiles Feedwater Addition 06 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 TITLE REV ADDITIONAL INFORMATION B 80-010 Contamination of Nonradioactive Cl Unit 2 Actions:

System and Resulting Potential for Unmonitored, Uncontrolled 06 2) Include proper monitoring of non-radioactive systems in procedures.

Release of Radioactivity to Environment 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.

B 80-010 Contamination of Nonradioactive CI Unit 2 Actions: 1) Correct deficiencies involving monitoring of systems.

System and Resulting Potential for Unmonitored, Uncontrolled 06 ....................................................................................................

Release of Radioactivity to ....................................................................................................

Environment 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.

B 80-011 Masonry Wall Design Cl 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.

B 80-012 Decay Heat Removal System CI NRC: IR 390/391 85-08; NUREG/CR 4005 Operability _-

Unit 2 Action: Implement operating instructions and abnormal operating instructions (AOIs) for RHR.

[WAS "NOTE 3."]

Boiling Water Reactor B 80-013 Cracking in Core Spray Spargers NA Boiling Water Reactor B 80-014 Degradation of Scram Discharge NA Volume Capability _

B 80-015 Possible Loss of Emergency C Item concerns a multi-unit issue that was completed for both units.

Notification System with Loss of Offsite Power NRC: IR 390/391 85-08 B 80-016 Potential Misapplication of C TVA: letter dated August 29, 1980 Rosemount, Inc. Models 1151 and 1152 Pressure Transmitters With NRC: IR 390/391 81-17 Either "A"or "D" Output Codes Page 12 of 109

• o= See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 80-017 Failure of 76 of 185 Control Rods NA Boiling Water Reactor to Fully Insert During a Scram at a BWR B 80-018 Maintenance of Adequate CO IR 50-390/85-60 and 50-391/85-49 (Unit 1)

Minimum Flow Thru Centrifugal Charging Pumps Following 06 Unit 2 Action: Implement design and procedure changes.

Secondary Side High Energy Rupture REVISION 06 UPDATE:

NRC Inspection Report 391/2011-604 closed B 80-018.

B 80-019 Mercury-Wetted Matrix Relay in C TVA: letter dated September 4, 1980 Reactor Protective Systems of Operating Nuclear Power Plants NRC: NUREG/CR 4933; IR 390/391 81-17 Designed by CE B 80-020 Failure of Westinghouse Type Cl Unit 2 Action: Modify switches.

W-2 Spring Return to Neutral Control Switches 06 REVISION 06 UPDATE:

The switches were modified.

NRC Inspection Report 391/2011-604 closed B 80-020.

B 80-021 Valve Yokes Supplied by Malcolm C TVA: letter dated May 6, 1981 Foundry Co., Inc.

NRC: 390/391 85-08 B 80-022 Automation Industries, Model NA Does not apply to power reactor.

200-520-008 Sealed-Source Connectors B 80-023 Failures of Solenoid Valves C TVA: letter dated March 31, 1981 Manufactured by Valcor Engineering Corporation NRC: IR 390/391 81-17; NUREG/CR 5292 Page 13 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 80-024 Prevention of Damage Due to Cl Unit 2 Action:

Water Leakage Inside Containment (10/17/80 Indian 06 Confirm that the reactor cavity can not be flooded, resulting in the partial Point 2 Event) 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.

B 80-025 Operating Problems with Target NA Boiling Water Reactor Rock Safety-Relief Valves at BWRs B 81-001 Surveillance of Mechanical NA NRC: IR 390/391 81-17 Snubbers B 81-002 Failure of Gate Type Valves to C TVA: letter dated September 30, 1983 Close Against Differential Pressure NRC: IR 390/391 84-03 B 81-003 Flow Blockage of Cooling Water to C TVA: letters dated July 21, 1981 and March 21, 1983 Safety System Components by Asiatic Clams and Mussels NRC: IR 390/391 81-17 B 82-001 Alteration of Radiographs of C NRC: IR 390/391 85-08 Welds in Piping Subassemblies B 82-002 Degradation of Threaded CI TVA: memo dated February 6, 1985 Fasteners in the Reactor Coolant Pressure Boundary of PWR Plants 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:

The boric acid corrosion program applies to both units.

B 82-003 Stress Corrosion Cracking in Thick- NA Boiling Water Reactor Wall, Large Diameter, Stainless Steel, Recirculation System Piping at BWR Plants Page 14 of 109 * = See last page for status code definition.

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

NRC Inspection Report 391/2011-602 closed B 83-004.

B 83-005 ASME Nuclear Code Pumps and C TVA: letter dated September 7, 1983 Spare Parts Manufactured by the Hayward Tyler Pump Company NRC: IR 390/85-03 and 391/85-04; NUREG/CR 5297 B 83-006 Nonconforming Material Supplied CI TVA: letter dated February 2, 1984 by Tube-Line Facilities 04 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.

ITEM TITLE REV 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 inthe 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 C TVA: letter dated March 22, 1984 Sold by Ray Miller, Inc.

NRC: IR 390/85-03 and 391/85-04 B 83-008 Electrical Circuit Breakers With an C TVA: letter dated March 29, 1984 Undervoltage Trip Feature in Safety-Related Applications Other NRC: IR 390/84-35 and 391/84-33 Than the Reactor Trip System B 84-001 Cracks in BWR Mark 1 NA Boiling Water Reactor Containment Vent Headers B 84-002 Failure of GE Type HFA Relays In C TVA: letter dated July 10, 1984 Use In Class 1 E Safety Systems NRC: IR 390/391 84-42 and IR 390/84-77 and 391/84-54 B 84-003 Refueling Cavity Water Seal 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 Cl TVA: letter dated January 27, 1986 Feedwater Pumps 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 TITLE REV ADDITIONAL INFORMATION B 85-002 Undervoltage Trip Attachment of C Unit 2 Action:

Westinghouse DB-50 Type Reactor Trip Breakers 06 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.

B 85-003 Motor-Operated Valve Common C Superseded by GL 89-10 Mode Failures During Plant Transients Due to Improper Switch Settings B 86-001 Minimum Flow Logic Problems NA Boiling Water Reactor That Could Disable RHR Pumps B 86-002 Static "0" Ring Differential C TVA: letter dated November 20, 1986 Pressure Switches NRC: IR 390/391/90-24 B 86-003 Potential Failure of Multiple ECCS C TVA: letter dated November 14, 1986 Pumps Due to Single Failure of Air-Operated Valve in Minimum Flow NRC: IR 390/391/87-03 Recirculation Line B 86-004 Defective Teletherapy Timer That NA Does not apply to power reactor.

May Not Terminate Treatment Dose B 87-001 Thinning of Pipe Walls in Nuclear C TVA: letter dated September 18, 1987 Power Plants NRC: NUREG/CR 5287 Closed to GL 89-08 Page 17 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 87-002 Fastener Testing to Determine CI TVA: letters dated April 15, 1988, July 6, 1988, Conformance with Applicable September 12, 1988, and January 27, 1989 Material Specifications 03 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.

REVISION 03 UPDATE:

Unit 2 has completed fastener testing as required by this Bulletin.

B 88-001 Defects in Westinghouse Circuit C TVA: letter dated November 15, 1991 Breakers NRC: IR 390/391 93-01 B 88-002 Rapidly Propagating Fatigue CI NRC acceptance letter dated June 7, 1990, for both units.

Cracks in Steam Generator Tubes 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.

B 88-003 Inadequate Latch Engagement in C TVA: letter dated April 13, 1992 HFA Type Latching Relays

• Manufactured by General Electric NRC: IR 390/391 92-13 (GE) Company B 88-004 Potential Safety-Related Pump CI NRC acceptance letter dated May 24, 1990, for both units.

Loss Unit 2 Actions:

• Perform calculations, and

• install check valves to prevent pump to pump interaction.

B 88-005 Nonconforming Materials Supplied .CI NRC reviewed in Appendix EE of SSER16.

by Piping Supplies, Inc. and West Jersey Manufacturing Company Unit 2 Actions:

• Complete review to locate installed WJM material, and

• perform in-situ hardness testing for Unit 2.

B 88-006 Actions to be Taken for the NA Does not apply to power reactor.

Transfer of Model No. SPEC 2-T Radiographic Exposure Device Page 18 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 88-007 Power Oscillations in BWRs NA Boiling Water Reactor B 88-008 Thermal Stresses in Piping Cl NRC acceptance letter dated September 19, 1991, for both units.

Connected to Reactor Cooling Systems Unit 2 Action: Implement program to prevent thermal stratification.

B 88-009 Thimble Tube Thinning in Cl Reviewed in Appendix EE of SSER1 6.

Westinghouse Reactors 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.

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.

B 89-001 Failure of Westinghouse Steam C NRC acceptance letter dated September 26, 1991 for both units.

Generator Tube Mechanical Plugs -

06 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 TITLE REV ADDITIONAL INFORMATION B 89-002 Stress Corrosion Cracking of Cl NRC reviewed in Appendix EE of SSER1 6.

High-Hardness Type 410 Stainless Steel Internal Preloaded Bolting in 06 Unit 2 Actions:

Anchor Darling Model S350W Swing Check Valves or Valves of

• Replace the flapper assembly hold-down bolts fabricated on the 14 Similar Nature (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.

B 89-003 Potential Loss of Required CI TVA: letter dated June 19, 1990 Shutdown Margin During Refueling Operations NRC: IR390/391 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.

B 90-001 Loss of Fill-Oil in Transmitters Co Unit 2 Action:

Manufactured by Rosemount 06 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.

REVISION 06 UPDATE:

NRC Inspection Report 391/2011-603 closed B 90-001.

B 90-002 Loss of Thermal Margin Caused NA Boiling Water Reactor by Channel Box Bow B 91-001 Reporting Loss of Criticality Safety NA Does not apply to power reactor.

Controls Page 20 of 109 * = See last page for status code definition.

ITEM TITLE REV ADIDITIONAL INFORMATION B 92-001 Failure of Thermo-Lag 330 Fire NA Barrier System to Maintain Cabling in Wide Cable Trays and Small 02 Conduits Free From Fire Damage REVISION 02 UPDATE:

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.

B 92-002 Safety Concerns Related to "End of Life" of Aging Theratronics Teletherapy Units NA Does not apply to power reactor.

B 92-003 Release of Patients After Brachytherapy B 93-001 Release of Patients After NA Does not apply to power reactor.

Brachytherapy Treatment with Remote Afterloading Devices B 93-002 Debris Plugging of Emergency C Boiling Water Reactor Core Cooling Suction Strainers 02 -------------------------------------------------------------------------------------------------

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.

B 93-003 Resolution of Issues Related to NA Boiling Water Reactor Reactor Vessel Water Level Instrumentation in BWRs B 94-001 Potential Fuel Pool Draindown NA Addressed to holders of licenses for nuclear power reactors that are Caused by Inadequate permanently shut down with spent fuel in the spent fuel pool Maintenance Practices at Dresden Unit 1 Does not apply to power reactor.

B 94-002 Corrosion Problems in Certain NA Stainless Steel Packagings Used to Transport Uranium Hexafluoride Does not apply to power reactor.

B 95-001 Quality Assurance Program for NA Transportation of Radioactive Material B 95-002 Unexpected Clogging of a NA Boiling Water Reactor Residual Heat Removal Pump Strainer While Operating in Suppression Pool Cooling Mode Page 21 of 109 * = See last page for status code definition.

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

04 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.

B 96-001, Control Rod Insertion Problems CI NRC acceptance letter for Unit 1 dated July 22, 1996 - Initial response last part (PWR) for Unit 2 on September 7, 2007.

06 Unit 2 Action: and provide core map.

REVISION 03 UPDATE:

NRC issued the Safety Evaluation (corrected) for Bulletin 1996-001 on May 3, 2010.

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.

REV ITEM TITLE 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."

B 96-002 Movement of Heavy Loads over Cl NRC closure letter dated May 20, 1998.

Spent Fuel, Over Fuel in the Reactor, or Over Safety-Related 06 Unit 2 Action:

Equipment Unit 2 Heavy Loads Program will be in compliance with NUREG-0612.

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)"

B 96-003 Potential Plugging of ECCS NA Boiling Water Reactor Suction Strainers by Debris in BWRs B 96-004 Chemical, Galvanic, or Other NA Info Reactions in Spent Fuel Storage and Transportation Casks Does not apply to power reactor.

B 97-001 Potential for Erroneous NA Calibration, Dose Rate, or Radiation Exposure Measurements with Certain Victoreen Model 530 and 531SI Electrometer/Dosemeters Does not apply to power reactor.

B 97-002 Puncture Testing of Shipping NA Packages Under 10 CFR Part 71 Page 23 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 01-001 Circumferential Cracking of C NRC acceptance letter dated November 20, 2001 (Unit 1) - Initial Reactor Pressure Vessel (RPV) --- response for Unit 2 on September 7, 2007.

Head Penetration Nozzles 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. ML100539515)"

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. ML100539515)"

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. ML100539515)"

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 TITLE REV ADDITIONAL INFORMATION B 03-002 Leakage from RPV Lower Head Cl NRC acceptance letter dated October 6, 2004 (Unit 1) - Initial response Penetrations and Reactor Coolant - - for Unit 2 on September 7, 2007.

Pressure Boundary Integrity 06 (PWRs) 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)"

B 03-003 Potentially Deficient 1-inch Valves NA Does not apply to power reactor.

for Uranium Hexaflouride Cylinders _ _.

B 03-004 Rebaselining of Data in the C TVA: letter dated December 18, 2003 Nuclear Management and Safeguards System Item concerns a multi-unit issue that was completed for both units.

B 04-001 Inspection of Alloy 82/182/600 Cl Initial response for Unit 2 on September 7, 2007.

Materials Used in the Fabrication of Pressurizer Penetrations and 06 Unit 2 Actions:

Steam Space Piping Connections at PWRs

• Provide details of pressurizer and penetrations, and

• apply Material Stress Improvement Process.

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.

ITEM TITLE REV 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.

REVISION 06 UPDATE:

SSER22 contained the following for NRC Action:

"Closed. NRC Letter dated August 4, 2010 (ADAMS Accession No. ML102080017)"

B 05-001 Material Control and Accounting at C TVA: letters dated March 21, 2005 and May 11, 2005 Reactors and Wet Spent Fuel Storage Facilities Item concerns a multi-unit issue that was completed for both units.

B 05-002 Emergency Preparedness and C TVA: letters dated January 20, 2006 and August 16, 2006.

Response Actions for Security-Based Events Item concerns a multi-unit issue that was completed for both units.

Page 28 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 07-001 Security Officer Attentiveness C Item concerns a multi-unit issue that was completed for both units.

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:

"Closed. NRC Letter dated March 25, 2010 (ADAMS Accession No. ML100770549)"

C 76-001 Crane Hoist Control Circuit C See B 76-007 for additional information.

Modifications C 76-002 Relay Failures - Westinghouse C TVA: letter dated November 22, 1976 informed NRC that these relay BF (AC) and BFD (DC) Relays types'are not used in Class I E circuits.

NRC: IR 50/390/76-11 and 50/391/76-11 C 76-003 Radiation Exposures in Reactor NA Info Cavities C 76-004 Neutron Monitor and Flow Bypass NA Boiling Water Reactor Switch Malfunctions C 76-005 Hydraulic Shock And Sway C TVA: letter dated January 7, 1977 informed NRC that no Grinnell shock Suppressors - Maintenance of suppressors or sway braces have been or will be installed at WBN.

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 NA Item was applicable only to units with operating license at the time the Stagnant, Low Pressure Stainless - - - item was issued.

Piping Containing Boric Acid Solution at PWRs C 76-007 Inadequate Performance by. NA Item was applicable only to units with operating license at the time the Reactor Operating and Support item was issued.

Staff Members C 77-001 Malfunctions of Limitorque Valve NA Info Operators 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 REV ADDITIOINlAL INFORMATION NA Info C 78-002 Proper Lubricating Oil for Terry Turbines NA Info C 78-003 Packaging Greater Than Type A Quantities of Low Specific Activity Radioactive Material for Transport NA Info C 78-004 Installation Errors That Could Prevent Closing of Fire Doors NA Info C 78-005 Inadvertent Safety Injection During Cooldown NA Info C 78-006 Potential Common Mode Flooding of ECCS Equipment Rooms at BWR Facilities NA Info C 78-007 Damaged Components of a Bergen-Paterson Series 25000 Hydraulic Test Stand NA Info C 78-008 Environmental Qualification of Safety-Related Electrical Equipment at Nuclear Power Plants NA Info C 78-009 Arcing of General Electric Company Size 2 Contactors NA Does not apply to power reactor.

C 78-010 Control of Sealed Sources in Radiation Therapy C 78-011 Recirculation MG Set Overspeed NA Boiling Water Reactor Stops C 78-012 HPCI Turbine Control Valve Lift NA Boiling Water Reactor Rod Bending C 78-013 Inoperability of Service Water NA Info Pumps C 78-014 HPCI Turbine Reversing Chamber NA Boiling Water Reactor Hold Down Bolting NA Info C 78-015 Tilting Disc Check Valves Fail to Close with Gravity in Vertical Position NA Info C 78-016 Limitorque Valve Actuators Page 31 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION NA Info- - - - - -

C 78-017 Inadequate Guard Training/Qualification and Falsified Training Records NA Info C 78-018 UL Fire Test C 78-019 Manual Override (Bypass) of NA Info Safety System Actuation Signals C 79-001 Administration of Unauthorized NA Does not apply to power reactor.

Byproduct Material to Humans C 79-002 Failure of 120 Volt Vital AC Power NA Info Supplies NA Info C 79-003 Inadequate Guard Training -

Qualification and Falsified Training Records NA Info C 79-004 Loose Locking Nut on Limitorque Valve Operators NA Info C 79-005 Moisture Leakage in Stranded Wire Conductors C 79-006 Failure to Use Syringe and Bottle NA Does not apply to power reactor.

Shields in Nuclear Medicine C 79-007 Unexpected Speed Increase of NA Boiling Water Reactor Reactor Recirculation MG Set Resulted in Reactor Power Increase C 79-008 Attempted Extortion - Low NA Fuel facilities and operating reactors at the time the item was issued Enriched Uranium NA Info C 79-009 Occurrences of Split or Punctured Regulator Diaphragms in Certain Self Contained Breathing Apparatus NA Info C 79-010 Pipefittings Manufactured from Unacceptable Material NA Info C 79-011 Design/Construction Interface Problem Page 32 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION C 79-012 Potential Diesel Generator NA Info Turbocharger Problem C 79-013 Replacement of Diesel Fire Pump NA Info Starting Contactors C 79-014 Unauthorized Procurement and NA Does not apply to power reactor.

Distribution of XE-133 C 79-015 Bursting of High Pressure Hose NA Item was applicable only to units with operating license at the time the and Malfunction of Relief Valve 0- item was issued.

Ring in Certain Self-Contained Breathing Apparatus C 79-016 Excessive Radiation Exposures to NA Does not apply to power reactor.

Members of the General Public and a Radiographer C 79-017 Contact Problem in SB-12 NA Info Switches on General Electric Company Metalclad Circuit Breakers C 79-018 Proper Installation of Target Rock NA Boiling Water Reactor Safety-Relief Valves C 79-019 Loose Locking Devices on NA Info Ingersoll-Rand Pumps C 79-020 Failure of GTE Sylvania Relay NA Info Type PM Bulletin 7305 Catalog 5U12-1 1-AC with a 120V AC Coil C 79-021 Prevention of Unplanned NA Info Releases of Radioactivity C 79-022 Stroke Times for Power Operated NA Info Relief Valves C 79-023 Motor Starters and Contactors C The Circular did not require a response.

Failed to Operate 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 NA Boiling Water Reactor of Core Spray Pipe Break Detection Equipment on BWRs C 79-025 Shock Arrestor Strut Assembly C The Circular did not require a response.

Interference 01 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.

C 80-001 Service Advice for GE Induction NA Info Disc Relays Info-C 80-002 Nuclear Power Plant Staff Work NA Hours Info C 80-003 Protection from Toxic Gas Hazards NA NA Info C 80-004 Securing of Threaded Locking Devices on Safety-Related Equipment NA Info C 80-005 Emergency Diesel-Generator Lubricating Oil Addition and Onsite Supply C 80-006 Control and Accountability NA Does not apply to power reactor.

Systems for Implant Therapy Sources Boiling Water Reactor C 80-007 Problems with HPCI Turbine Oil NA System Boiling Water Reactor C 80-008 BWR Technical Specification NA Inconsistency - RPS Response Time C 80-009 Problems with Plant Internal NA Info Communications Systems C 80-010 Failure to Maintain Environmental NA Info Qualification of Equipment Page 34 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION C 80-011 Emergency Diesel Generator Lube NA Info Oil Cooler Failures C 80-012 Valve-Shaft-to-Actuator Key May NA Info Fall Out of Place when Mounted Below Horizontal Axis C 80-013 Grid Strap Damage in NA Info Westinghouse Fuel Assemblies C 80-014 Radioactive Contamination of NA Info Plant Demineralized Water System and Resultant Internal Contamination of Personnel C 80-015 Loss of Reactor Coolant Pump NA Info Cooling and Natural Circulation Cooldown C 80-016 Operational Deficiencies in NA Info Rosemount Model 510DU Trip Units and Model 1152 Pressure Transmitters C 80-017 Fuel Pin Damage Due to Water NA Info Jet from Baffle Plate Corner C 80-018 10 CFR 50.59 Safety Evaluations NA Info for Changes to Radioactive Waste Treatment Systems C 80-019 Noncompliance with License NA Does not apply to power reactor.

Requirements for Medical Licensees C 80-020 Changes in Safe-Slab Tank NA Info Dimensions C 80-021 Regulation of Refueling Crews NA Item was applicable only to units with operating license at the time the item was issued.

C 80-022 Confirmation of Employee NA Info Qualifications C 80-023 Potential Defects in Beloit Power NA Info Systems Emergency Generators C 80-024 AECL Teletherapy Unit Malfunction NA Does not apply to power reactor.

Page 35 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION C 80-025 Case Histories of Radiography NA Does not apply to power reactor.

Events C 81-001 Design Problems Involving NA Info Indicating Pushbutton Switches Manufactured by Honeywell Incorporated C 81-002 Performance of NRC-Licensed NA Item was applicable only to units with operating license at the time the Individuals while on Duty - - - item was issued.

NA Info C 81-003 Inoperable Seismic Monitoring Instrumentation NA Info C 81-004 The Role of Shift Technical Advisors and Importance of Reporting Operational Events C 81-005 Self-Aligning Rod End Bushings NA Info for Pipe Supports NA Info NA Info C 81-006 Potential Deficiency Affecting Certain Foxboro 10 to 50 Milliampere Transmitters NA Info C 81-007 Control of Radioactively Contaminated Material NA Info NA Info C 81-008 Foundation Materials C 81-009 Containment Effluent Water that Bypasses Radioactivity Monitor C 81-010 Steam Voiding in the Reactor NA Item was applicable only to units wiith operating license at the time the Coolant System During Decay item was issued.

Heat Removal Cooldown C 81-011 Inadequate Decay Heat Removal NA Boiling Water Reactor During Reactor Shutdown C 81-012 Inadequate Periodic Test NA Info Procedure of PWR Reactor Protection System Page 36 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION C 81-013 Torque Switch Electrical Bypass C The Circular did not require a response.

Circuit for Safeguard Service Valve Motors 01 TVA reported a nonconformance under 10 CFR 50.55e on April 4, 1986 (NCR W367-P), that required closing torque switches were founcd improperly wired. This issue (Torque switch and overload relay bypa ss capability for active safety related valves) is part of the Electrical IssuEks 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 was applicable only to units with operating license at the time the was issued.

to Steam Generators - - - item was applicable only to units with operating license at the time the GL 77-007 Reliability of Standby Diesel NA Item was issued.

Generator Units - - - item GL 77-008 Revised Intrusion Detection NA Info Handbook and Entry Control Systems Handbook was applicable only to units with operating license at the time the GL 78-001 Correction to Letter of 12/15/77 NA Item was issued.

[GL 77-07] - item GL 78-002 Asymmetric Loads Background C NRC Reviewed in SSER15 - Appendix C (June 1995). Resolved by and Revised Request for apprn val of leak-before-break analysis.

Additional Information Page 37 of 109. * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 78-003 Request For Information on Cavity NA Item was applicable only to units with operating license at the time the Annulus Seal Ring --- item was issued.

GL 78-004 GAO Blanket Clearance for Letter NA Item was applicable only to units with operating license at the time the Dated 12/09/77 [GL 77-06] - - -item was issued.

GL 78-005 Internal Distribution of NA Info Correspondence - Asking for Comments on Mass Mailing System GL 78-006 This GL was never issued. NA GL 78-007 This GL was never issued. NA GL 78-008 Enclosing NUREG-0408 Re NA Boiling Water Reactor Mark I Containments, and Granting Exemption from GDC 50 and Enclosing Sample Notice GL 78-009 Multiple-Subsequent Actuations of NA Boiling Water Reactor Safety/Relief Valves Following an Isolation Event 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 Transmittal of Draft NUREG-0219 NA Info for Comment GL 78-015 Request for Information on Control NA See GL 81-007.

of Heavy Loads Near Spent Fuel GL 78-016 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 Duplicate of GL 81-007 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 Boilirng Water Reactor Specifications and Submittal Schedule For BWRs GL 78-025 This GL was never issued. NA GL 78-026 Excessive Control Rod Guide NA Applies only to Babcock and Wilcox designed plants Tube Wear GL 78-027 Forwarding of NUREG-0181 NA Info GL 78-028 Forwarding pages omitted from NA Boiling Water Reactor 07/11/78 letter [GL 78-24]

GL 78-029 Notice of PWR Steam Generator NA Info Conference GL 78-030 Forwarding of NUREG-0219 NA Info Page 39 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 78-031 Notice of Steam Generator NA Info Conference Agenda GL 78-032 Reactor Protection System Power NA Boiling Water Reactor Supplies GL 78-033 Meeting Schedule and Locations NA Info For Upgraded Guard Qualification GL 78-034 Reactor Vessel Atypical Weld C See B 78-12.

Material GL 78-035 Regional Meetings to Discuss NA Info Upgraded Guard Qualifications GL 78-036 Cessation of Plutonium Shipments NA Does not apply to power reactor.

by Air Except In NRC Approved Containers GL 78-037 Revised Meeting Schedule & NA Info Locations For Upgraded Guard Qualifications GL 78-038 Forwarding of 2 Tables of NA Item was applicable only to units with operating license at the time the Appendix I, Draft Radiological item was issued.

Effluent Technical Specifications, PWR, and NUREG-0133 GL 78-039 Forwarding of 2 Tables of NA Boiling Water Reactor Appendix I, Draft Radiological Effluent Technical Specifications, BWR, and NUREG-0133 GL 78-040 Training & Qualification Program NA Info Workshops GL 78-041 Mark II Generic Acceptance NA Boiling Water Reactor Criteria For Lead Plants NA Info GL 78-042 Training and Qualification Program Workshops NA Info GL 79-001 Interservice Procedures for Instructional Systems Development - TRADOC GL 79-002 Transmitting Rev. to Entry Control NA Info Systems Handbook (SAND 77-1033), Intrusion Detection Handbook (SAND 76-0554), and Barrier Penetration Database Page 40 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION NA Info GL 79-003 Offsite Dose Calculation Manual NA Info GL 79-004 Referencing 4/14/78 Letter -

Modifications to NRC Guidance "Review and Acceptance of Spent Fuel Pool Storage and Handling" NA Info GL 79-005 Information Relating to Categorization of Recent Regulatory Guides by the Regulatory Requirements Review Committee NA Info GL 79-006 Contents of the Offsite Dose Calculation Manual NA Info GL 79-007 Seismic (SSE) and LOCA Responses (NUREG-0484)

NA Info GL 79-008 Amendment to 10 CFR 73.55 NA Boiling Water Reactor GL 79-009 Staff Evaluation of Interim Multiple-Consecutive Safety-Relief Valve Actuations GL 79-010 Transmitting Regulatory Guide 2.6 NA Does not apply to power reactor.

for Comment 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.

ITEM TITLE REV 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, NA Info Enhancement of on Site Emergency Diesel Generator Reliability" GL 79-022 Enclosing NUREG-0560, "Staff NA Applies only to Babcock and Wilcox designed plants Report on the Generic Assessment of Feedwater Transients in PWRs Designed by B&W" GL 79-023 NRC Staff Review of Responses NA Boiling Water Reactor to B 79-08 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.

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

Technical Specifications GL 79-031 Submittal of Copies of Response NA Info to 6/29/79 NRC Request [79-25]

GL 79-032 Transmitting NUREG-0578, NA Info "TMI-2 Lessons Learned" GL 79-033 Transmitting NUREG-0576, NA Info "Security Training and Qualification Plans" GL 79-034 New Physical Security Plans NA Does not apply to power reactor.

(FR 43280-285)

GL 79-035 Regional Meetings to Discuss NA Info Impacts on Emergency Planning GL 79-036 Adequacy of Station Electric Cl This GL tracked compliance with BTP PSB-1, "Adequacy of Station Distribution Systems Voltages 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 REV ADDITIONAL INFORMATION GL 79-042 Potentially Unreviewed Safety NA Item was applicable only to units with operating license at the time the Question on Interaction Between item was issued.

Non-Safety Grade Systems and Safety Grade Systems GL 79-043 Reactor Cavity Seal Ring Generic NA Addressed to specific plant(s).

Issue GL 79-044 Referencing 6/29/79 Letter Re NA Item was applicable only to units with operating license at the time the Multiple Equipment Failures item was issued.

GL 79-045 Transmittal of Reports Regarding NA Info Foreign Reactor Operating Experiences GL 79-046 Containment Purge and Venting NA Item was applicable only to units with operating license at the time the During Normal Operation - item was issued.

Guidelines for Valve Operability GL 79-047 Radiation Training NA Info GL 79-048 Confirmatory Requirements NA Boiling Water Reactor Relating to Condensation Oscillation Loads for the Mark I Containment Long Term Program NA Info 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 NA Info GL 79-050 Emergency Plans Submittal Dates GL 79-051 Follow-up Actions Resulting from NA GL 79-51 provided follow-up actions resulting from the Three Mile Island the NRC Staff Reviews Regarding - - Unit 2 accident. GL 79-51 was provided for planning and guidance the TMI-2 Accident 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.

GL 79-052 Radioactive Release at North NA Item was applicable only to units with operating license at the time the Anna Unit 1 and Lessons Learned - - - item was issued.

GL 79-053 ATWS NA Info GL 79-054 Containment Purging and Venting NA Addressed to specific plant(s).

During Normal Operation Page 44 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 79-055 Summary of Meeting Held on NA Info October 12, 1979 to Discuss Responses to Bulletins79-05C and 79-06C and HPI Termination Criteria NA Item was applicable only to units with operating license at the time the GL 79-056 Discussion of Lessons Learned Short Term Requirements --- item was issued.

GL 79-057 Acceptance Criteria for Mark I NA Boiling Water Reactor Long Term Program GL 79-058 ECCS Calculations on Fuel NA Item was applicable only to units with operating license at the time the Cladding item was issued.

GL 79-059 This GL was never issued. NA GL 79-060 Discussion of Lessons Learned NA Info Short Term Requirements GL 79-061 Discussion of Lessons Learned NA Info Short Term Requirements GL 79-062 ECCS Calculations on Fuel NA Item was applicable only to units with operating license at the time the Cladding - - item was issued.

Duplicate of GL 79-058 GL 79-063 Upgraded Emergency Plans 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 GL 79-064 Suspension of All Operating Licenses (PWRs)

NA Info GL 79-065 Radiological Environmental Monitoring Program Requirements - Enclosing Branch Technical Position, Revision 1 NA Info GL 79-066 Additional Information Re 11/09/79 Letter on ECCS Calculations [GL 79-62]

NA Info GL 79-067 Estimates for Evacuation of Various Areas Around Nuclear Power Reactors Page 45 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 79-068 Audit of Small Break LOCA NA Info Guidelines GL 79-069 Cladding Rupture, Swelling, and NA Info Coolant Blockage as a Result of a Reactor Accident NA Info GL 79-070 Environmental Monitoring for Direct Radiation NA Info GL 80-001 NUREG-0630, "Cladding, Swelling and Rupture - Models For LOCA Analysis" GL 80-002 QA Requirements Regarding C TVA: FSAR 9.5.4.2 Diesel Generator Fuel Oil GL 80-003 BWR Control Rod Failures NA Boiling Water Reactor GL 80-004 B 80-01, "Operability of ADS Valve NA Boiling Water Reactor Pneumatic Supply" GL 80-005 B 79-01b, "Environmental NA Info Qualification of Class 1 E Equipment" GL 80-006 Issuance of NUREG-0313, Rev 1, NA Boiling Water Reactor "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping" GL 80-007 This GL was never issued. NA GL 80-008 B 80-02. "Inadequate Quality NA Boiling Water Reactor Assurance for Nuclear Supplied Equipment" GL 80-009 Low Level Radioactive Waste NA Item was applicable only to units with operating license at the time the Disposal - - -item was issued.

GL 80-010 Issuance of NUREG-0588, NA Info "Interim Staff Position On Equipment Qualifications of Safety-Related Electrical Equipment" GL 80-011 B 80-03, "Loss of Charcoal From C GL 80-11 transmitted Bulletin 80-03. TVA responded to B 80-03 on March Standard Type II, 2 Inch, Tray - 21, 1980. See B 80-03 for further information.

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

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

Feedwater Addition" GL 80-013 Qualification of Safety Related NA Item was applicable only to units with operating license at the time the Electrical Equipment item was issued.

GL 80-014 LWR Primary Coolant System S TVA: FSAR 5.2.7.4 Pressure Isolation Valves 02 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 GL 80-015 Request for Additional NA Management and Technical Resources Information Info GL 80-016 B 79-01b, "Environmental NA Qualification of Class 1E Equipment" GL 80-017 Modifications to BWR Control Rod NA Boiling Water Reactor Drive Systems GL 80-018 Crystal River 3 Reactor Trip From NA Applies only to Babcock and Wilcox designed plants Approximately 100% Full Power Info GL 80-019 Resolution of Enhanced Fission NA Gas Release Concern Info 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 TITLE REV ADDITIONAL INFORMATION GL 80-021 B 80-05, "Vacuum Condition CI Closed in IR50-390/84-59 and 50-391/84-45.

Resulting in Damage to Chemical Volume Control System Holdup Unit 2 A.ction:

Tanks" Comple 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 C GL 80-28 transmitted Bulletin 80-08. TVA responded to Containment Liner Penetration - - B 80-08 on July 8, 1980. See B 80-08 for further information.

Welds" 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 es only to Babcock and Wilcox designed plants 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 ng Water Reactor 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 was applicable only to units with operating license at the time the Requirements to Operating item was issued.

Reactors GL 80-038 Summary of Certain Non-Power NA Does not apply to power reactor.

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 Boilirng Water Reactor Spargers" _ -

GL 80-044 Reorganization of Functions and NA Info Assignments Within ONRR/SSPB GL 80-045 Fire Protection Rule NA Item was applicable only to units with operating license at the time the

- - item was issued.

Page 49 of 109 = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 80-046 Generic Technical Activity A-12, C No response was required for this GL, and NUREG-0577 states that the and "Fracture Toughness and lamellar tearing aspect of this issue was resolved by the NUREG. Further, GL 80-047 Additional Guidance on Potential the NUREG states that for plants under review, the fracture toughness for Low Fracture toughness and issue was resolved.

Laminar Tearing on PWR Steam Generator Coolant Pump Supports" GL 80-048 Revision to 5/19/80 Letter On Fire NA Item was applicable only to units with operating license at the time the Protection [GL 80-45] - - - item was issued.

GL 80-049 Nuclear Safeguards Problems NA Info GL 80-050 Generic Activity A-10, "BWR NA Boiling Water Reactor Cracks" GL 80-051 On-Site Storage of Low-Level NA Item was applicable only to units with operating license at the time the Waste item was issued.

GL 80-052 Five Additional TMI-2 Related NA Item was applicable only to units with operating license at the time the Requirements - Erata Sheets to item was issued.

5/7/80 Letter [GL 80-37]

GL 80-053 Decay Heat Removal Capability NA Item was applicable only to units with operating license at the time the item was issued.

GL 80-054 B 80-14, "Degradation of Scram NA Boiling Water Reactor Discharge Volume Capability" GL 80-055 B 80-15, "Possible Loss of Hotline NA Info With Loss of off-Site Power" GL 80-056 Commission Memorandum and NA Info Order on Equipment Qualification GL 80-057 Further Commission Guidance For NA Info Power Reactor Operating Licenses NUREG-0660 and NUREG-0694 GL 80-058 B 80-16, "Potential Misapplication NA Info of Rosemount Inc. Models 1151/1152 Pressure Transmitters With "A" Or "D" Output Codes" Page 50 of 109 * = See last page for status code definition.

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

GL 80-076 Notice of Forthcoming Meeting NA Info With GE to Discussed Proposed BWR Feedwater Nozzle Leakage Detection System GL 80-077 Refueling Water Level - S Unit 2 Action: Address in Technical Specifications, as appropriate.

Technical Specifications Changes 02 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 GL 80-078 Mark I Containment Long-Term NA Program Boiling Water Reactor GL 80-079 B 80-17, Supplement 3, "Failures NA Revealed by Testing Subsequent to Failure of Control Rods to Insert During a Scram At a BWR" GL 80-080 Preliminary Clarification of TMI NA Info Action Plan Requirements GL 80-081 Preliminary Clarification of TMI NA Info Action Plan Requirements -

Addendum to 9/5/80 Letter

[GL 80-80]

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

ITEM TITLE REV ADDITIONAL INFORMATION GL 80-083 Environmental Qualification of NA Info Safety-Related Equipment GL 80-084 BWR Scram System NA Boiling Water Reactor GL 80-085 Implementation of Guidance From NA Info USI A-12, "Potential For LOW Fracture Toughness and Lamellar Tearing On Component Support" NA Info-GL 80-086 Notice of Meeting to Discuss Final Resolution of USI A-12 NA Info 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 NA Item was applicable only to units with operating license at the time the Feedwater Systems item was issued.

GL 80-089 B 79-01b, Supplement 3, NA Info "Environmental Qualification of Class 1E Equipment" GL 80-090 NUREG-0737, TMI (Prior and CI See NUREG items in this list.

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 NA Boiling Water Reactor GL 80-092 B 80-21, "Valve Yokes Supplied by C GL 80-92 transmitted Bulletin 80-21. TVA responded to Malcolm Foundry Company, Inc." ___ 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, NA Boiling Water Reactor NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking" Page 53 of 109 * = See last page for status code definition.

REV ADDITIONAL INFORMATION ITEM TITLE GL 80-096 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 NA Info For Snubber Surveillance GL 80-100 Appendix R to 10 CFR 50 NA Item was applicable only to units with operating license at the time the Regarding Fire Protection - - - item was issued.

Federal Register Notice GL 80-101 Inservice Inspection Programs NA Addressed to specific plant(s).

NA Info GL 80-102 Commission Memorandum and Order of May 23, 1980 (Referencing B 79-01b, Supplement 2 - q.2 & 3 - Sept 30, 1980)

NA Info GL 80-103 Fire Protection - Revised Federal Register Notice NA Info GL 80-104 Orders On Environmental Qualification of Safety Related Electrical Equipment NA Info GL 80-105 Implementation of Guidance For USI A-12, "Potential For Low Fracture toughness and Lamellar Tearing On Component Supports" NA Info GL 80-106 Report On ECCS Cladding Models, NUREG-0630 GL 80-107 BWR Scram Discharge System NA Boiling Water Reactor GL 80-108 Emergency Planning NA Info GIL 80-109 Guidelines For SEP Soil Structure NA Info Interaction Reviews Page 54 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 80-110 Periodic Updating of FSARS NA 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 NA Boiling Water Reactor Control Rods to Insert During a Scram at a BWR" GL 80-112 B 80-25, "Operating Problems NA Info With Target Rock Safety Relief Valves" GL 80-113 Control of Heavy Loads C Superseded by GL 81-007.

GL 81-001 Qualification of Inspection, NA Info Examination, Testing and Audit Personnel GL 81-002 Analysis, Conclusions and NA Info Recommendations Concerning Operator Licensing GL 81-003 Implementation of NUREG-0313, NA Boiling Water Reactor "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping" GL 81-004 Emergency Procedures and C Superseded by Station Blackout Rule.

Training for Station Blackout Events NA Info GL 81-005 Information Regarding The Program For Environmental Qualification of Safety-Related Electrical Equipment NA Info GL 81-006 Periodic Updating of Final Safety Analysis Reports (FSARS)

GL 81-007 Control of Heavy Loads 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.

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

GL 81-012 Fire Protection Rule NA Item was applicable only to units with operating license at the time the item was issued.

GL 81-013 SER For GEXL Correlation For NA Boiling Water Reactor 8X8R Fuel Reload Applications For Appendix D Submittals of The GE topical Report GL 81-014 Seismic Qualification of Auxiliary Cl TVA: FSAR 10.4.9 Feedwater Systems Unit 2 Action: Additional Unit 2 implementing procedures or other activity is required for completion.

[WAS "OL."]

GL 81-015 Environmental Qualification of NA Info 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 Applies only to Babcock and Wilcox designed plants Abnormal Transient Operating -.

Guidelines (ATOG)

GL 81-017 Functional Criteria for Emergency NA Info Response Facilities _ -.

GL 81-018 BWR Scram Discharge System - NA Boiling Water Reactor Clarification of Diverse Instrumentation Requirements GL 81-019 Thermal Shock to Reactor NA Item was applicable only to units with operating license at the time the Pressure Vessels item was issued.

GL 81-020 Safety Concerns Associated With NA Boiling Water Reactor Pipe Breaks in the BWR Scram System Page 56 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 81-021 Natural Circulation Cooldown CI TVA responded December 3, 1981.

Unit 2 Action: Issue operating procedures.

GL 81-022 Engineering Evaluation of the NA Info H. B. Robinson Reactor Coolant System Leak on 1/29/81 GL 81-023 INPO Plant Specific Evaluation NA Info Reports GL 81-024 Multi-Plant Issue B-56, "Control NA Boiling Water Reactor Rods Fail to Fully Insert" GL 81-025 Change in Implementing Schedule NA Info For Submission and Evaluation of Upgraded Emergency Plans GL 81-026 Licensing Requirements for NA Applicants with pending Construction Permits Pending Construction Permit and Manufacturing License Applications Info GL 81-027 Privacy and Proprietary Material in NA Emergency Plans Info GL 81-028 Steam Generator Overfill NA GL 81-029 Simulator Examinations NA Info GL 81-030 Safety Concerns Associated With NA Boiling Water Reactor Pipe Breaks in the BWR Scram System NA GL 81-031 This GL was never issued.

NA Boiling Water Reactor GL 81-032 NUREG-0737, Item II.K.3.44, "Evaluation of Anticipated Transients Combined With Single Failure" NA GL 81-033 This GL was never issued.

NA Boiling Water Reactor GL 81-034 Safety Concerns Associated With Pipe Breaks in the BWR Scram System Page 57 of 109 * = See last page for status code definition.

ITEM REV ADDITIONAL INFORMATION TITLE 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 NA Info NA Item was applicable only to units with operating license at the time the GL 82-005 Post-TMI Requirements

- - - item was issued.

GL 82-006 This GL was never issued. NA GL 82-007 Transmittal of NUREG-0909 NA Boiling Water Reactor Relative to the Ginna Tube Rupture GL 82-008 Transmittal of NUREG-0909 NA Info Relative to the Ginna Tube Rupture GL 82-009 Environmental Qualification of NA Info Safety Related Electrical Equipment Page 58 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 82-010 Post-TMI Requirements 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 NA Item was applicable only to units with operating GL 82-016 NUREG-0737 Technical license at the time the Specifications --- item was issued.

GL 82-017 Inconsistency of Requirements NA Info Between 50.54(T) and 50.15 GL 82-018 Reactor Operator and Senior NA Info Reactor Operator Requalification Examinations NA Info GL 82-019 Submittal of Copies of Documentation to NRC - Copy Requirements for Emergency Plans and Physical Security Plans NA Info GL 82-020 Guidance for Implementing the Standard Review Plan Rule NA Info GL 82-021 Fire Protection Audits GL 82-022 Congressional Request for NA Item was applicable only to units with operating license at the time the Information Concerning Steam item was issued.

Generator Tube Integrity GL 82-023 Inconsistency Between NA Info Requirements of 10CFR 73.40(d) and Standard Technical Specifications For Performing Audits of Safeguards Contingency Plans Page 59 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 82-024 Safety Relief Valve Quencher NA Boiling Water Reactor Loads: BWR MARK IIand III Containments GL 82-025 Integrated IAEA Exercise for NA Item was applicable only to units with operating license at the time the Physical Inventory at LWRS item was issued.

GL 82-026 NUREG-0744, REV. 1, "Pressure NA Item was applicable only to units with operating license at the time the Vessel Material Fracture item was issued.

Toughness" GL 82-027 Transmittal of NUREG-0763, NA Boiling Water Reactor "Guidelines For Confirmatory In-Plant Tests of Safety-Relief Valve Discharge for BWR Plants" GL 82-028 Inadequate Core Cooling CO LICENSE CONDITION - Detectors for Inadequate core cooling (II.F.2)

Instrumentation System 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:

"Closed. Subsumed as part of NRC staff review of Instrumentation and Controls submitted April 8, 2010."

GL 82-029 This GL was never issued. NA GL 82-030 Filings Related to 10 CFR 50 NA Info Production and Utilization Facilities GL 82-031 This GL was never issued. NA GL 82-032 Draft Steam Generator Report NA Item was applicable only to units with operating license at the time the (SAI) - - item was issued.

Page 60 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 82-033 Supplement to NUREG-0737, Cl "Safety Parameter Display System" (SPDS) / "Requirements for "Requirements for Emergency Emergency Response Capability" - NRC reviewed in SSER5, SSER6, Response Capability" 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.

NA GL 82-035 This GL was never issued.

NA GL 82-036 This GL was never issued..

NA GL 82-037 This GL was never issued.

NA GL 82-038 Meeting to Discuss Developments NA Info for Operator Licensing Examinations GL 82-039 Problems With Submittals of NA Info Subsequent Information of CURT 73.21 For Licensing Reviews GL 83-001 Operator Licensing Examination NA Info Site Visit GL 83-002 NUREG-0737 Technical NA Boiling Water Reactor Specifications GL 83-003 This GL was never issued. NA GL 83-004 Regional Workshops Regarding NA Info Supplement 1 to NUREG-0737, "Requirements For Emergency Response Capability" GL 83-005 Safety Evaluation of "Emergency NA Boiling Water Reactor Procedure Guidelines, Revision 2," June 1982 GL 83-006 Certificates and Revised Format NA Info For Reactor Operator and Senior Reactor Operator Licenses GL 83-007 The Nuclear Waste Policy Act of NA Info 1982 Page 61 of 109 * = See last page for status code definition.

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

ITEM TITLE REV ADDITIONAL INFORMATION GL 83-014 Definition of "Key Maintenance NA Info Personnel," (Clarification of Generic Letter 82-12)

GL 83-015 Implementation of Regulatory NA Info Guide 1.150, "Ultrasonic Testing of Reactor Vessel Welds During Preservice & Inservice Examinations, Revision 1" GL 83-016 Transmittal of NUREG-0977 NA Info Relative to the ATWS Events at Salem Generating Station, Unit No.1 GL 83-016a Transmittal of NUREG-0977 NA Info Relative to the ATWS Events at Salem Generating Station, Unit No.1 GL 83-017 Integrity of Requalification NA Info Examinations for Renewal of Reactor Operator and Senior Reactor Operator Licenses GL 83-018 NRC Staff Review of the BWR NA Boiling Water Reactor Owners' Group (BWROG) Control Room Survey Program GL 83-019 New Procedures for Providing NA Item was applicable only to units with operating license at the time the Public Notice Concerning item was issued.

Issuance of Amendments to Operating Licenses GL 83-020 Integrated Scheduling for NA Info Implementation of Plant Modifications GL 83-021 Clarification of Access Control NA Info Procedures for Law Enforcement Visits GL 83-022 Safety Evaluation of "Emergency NA Info Response Guidelines" GL 83-023 Safety Evaluation of "Emergency NA Applies only to Combustion Engineering designed plants Procedure Guidelines" GL 83-024 TMI Task Action Plan Item I.G.1, NA Boiling Water Reactor "Special Low Power Testing and Training," Recommendations for BWRs Page 63 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 83-025 This GL was never issued. NA GL 83-026 Clarification Of Surveillance NA Info Requirements For Diesel Fuel Impurity Level Tests GL 83-027 Surveillance Intervals in Standard NA Info Technical Specifications GL 83-028 "Required Actions Based on C TVA: letters dated November 7, 1983 and Generic Implications of Salem December 4, 1987 ATWS Events:

NRC: IR 50-390, 391/86-04 1.2 - Post Trip Review Data and Information Capability GL 83-028 "Required Actions Based on Cl TVA: letters dated November 7, 1983 and August 24, 1990 Generic Implications of Salem ATWS Events: 06 NRC: letters dated October 20, 1986 and June 18, 1990 2.1 - Equipment Classification and Vendor Interface (Reactor Trip Unit 2 Action:

System Components)

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.

GL 83-028 "Required Actions Based on CI Unit 2 Action:

Generic Implications of Salem ATWS Events: Enter engineering component background data in INPO's Equipment Performance and Information Exchange System (EPIX) for Unit 2.

2.2 - Equipment Classification and Vendor Interface (All SR Components)"

Page 64 of 109 * = See last page for status code definition.

REV ADDITIONAL INFORMATION ITEM TITLE GL 83-028 "Required Actions Based on S TVA: letters dated November 7, 1983, January 17, 1986 and Generic Implications of Salem November 1, 1993 ATWS Events: 02 NRC: letters dated December 10, 1985, October 27, 1986, and July 2, 3.1 - Post-Maintenance Testing 1990; IR 390, 391/86-04 (Reactor Trip System Components)

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 S TVA: letters dated November 7, 1983, January 17, 1986 and Generic Implications of Salem November 1, 1993 ATWS Events: 06 NRC: letters dated December 10, 1985, October 27, 1986, and 3.2 - Post-Maintenance Testing July 2, 1990; IR 390, 391/86-04 (All SR Components)

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 CO TVA: letter dated May 19, 1986 Generic Implications of Salem ATWS Events: 06 4.1 - Reactor Trip System Unit 2 Action:

Reliability (Vendor Related Confirm vendor-recommended DS416 breaker modifications are Modifications) implemented.

REVISION 06 UPDATE:

NRC Inspection Report 391/2011-602 closed GL 83-028, Item 4.1.

GL 83-028 "Required Actions Based on S TVA: letters dated November 7, 1983, February 10, 1986, and Generic Implications of Salem May 19, 1986 ATWS Events: 02 NRC: letters dated July 26, 1985 and June 18, 1992; SSER 16 4.2 - Reactor Trip System Reliability (Preventive Maintenance and Surveillance Program Unit 2 Action: Ensure maintenance instruction procedure and Technical for Reactor Trip Breakers) 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 C TVA: letters dated November 7, 1983, March 22, 1985 Generic Implications of Salem ATWS Events: NRC: IR 50-390/86-04 and 50-391/86-04; letter dated June 18, 1990 4.3 - Reactor Trip System Reliability (Automatic Actuation of Shunt Trip Attachment)

Page 66 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 83-028 "Required Actions Based on S TVA: letters dated November 7, 1983 and July 26, 1985 Generic Implications of Salem ATWS Events: 02 "NRC: letters dated June 28, 1990 and October 9, 1990; SSERs 5 and 16 4.5 - Reactor Trip System Reliability (Automatic Actuation of Shunt Trip Attachment) 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.

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.

GL 83-029 This GL was never issued. NA GL 83-030 Deletion of Standard Technical NA Info Specifications Surveillance Requirement 4.8.1.1.2.d.6 For Diesel Generator Testing GL 83-031 Safety Evaluation of "Abnormal NA Applies only to Babcock and Wilcox designed plants Transient Operating Guidelines" GL 83-032 NRC Staff Recommendations NA Info Regarding Operator Action for Reactor Trip and ATWS GL 83-033 NRC Positions on Certain NA Info Requirements of Appendix R to 10 CFR 50 GL 83-034 This GL was never issued. NA GL 83-035 Clarification of TMI Action Plan NA Info Item I1.K.3.31 GL 83-036 NUREG-0737 Technical NA Boiling Water Reactor Specifications GL 83-037 NUREG-0737 Technical NA Item was applicable only to units with operating license at the time the Specifications - - - item was issued.

GL 83-038 NUREG-0965, "NRC Inventory of NA Info Dams" Page 67 of 109 * = See last page for status code definition.

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

GL 83-040 Operator Licensing Examination NA Info GL 83-041 Fast Cold Starts of Diesel NA Item was applicable only to units with operating license at the time the Generators 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 NA Info "Important To Safety" and "Safety Related" GL 84-002 Notice of Meeting Regarding NA Info Facility Staffing GL 84-003 Availability of NUREG-0933, "A NA Info Prioritization of Generic Safety Issues" GL 84-004 Safety Evaluation of NA Info Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops GL 84-005 Change to NUREG-1021, NA Info "Operator Licensing Examiner Standards" GL 84-006 Operator and Senior Operator NA Does not apply to power reactor.

License Examination Criteria For Passing Grade GL 84-007 Procedural Guidance for Pipe NA Boiling Water Reactor Replacement at BWRs Page 68 of 109 * = See last page for status code definition.

ITEM TITLE REV A DDITIONAL INFORMATION GL 84-008 Interim Procedures for NRC NA Info Management of Plant-Specific Backfitting GL 84-009 Recombiner Capability NA Boiling Water Reactor Requirements of 10 CFR 50.44(c)(3)(ii)

GL 84-010 Administration of Operating Tests NA Info Prior to Initial Criticality GL 84-011 Inspection of BWR Stainless Steel NA Boiling Water Reactor Piping GL 84-012 Compliance With 10 CFR Part 61 NA Info and Implementation of Radiological Effluent Technical Specifications (RETs) and Attendant Process Control Program (PCP)

NA Info GL 84-013 Technical Specification for Snubbers NA Info GL 84-014 Replacement and Requalification Training Program NA Info GL 84-015 Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability NA Info GL 84-016 Adequacy of On-Shift Operating Experience for Near Term Operating License Applicants NA Info GL 84-017 Annual Meeting to Discuss Recent Developments Regarding Operator Training, Qualifications, and Examinations NA Does not apply to power reactor.

GL 84-018 Filing of Applications for Licenses and Amendments NA Info GL 84-019 Availability of Supplement 1 to NUREG-0933, "A Prioritization of Generic Safety Issues" NA Info GL 84-020 Scheduling Guidance for Licensee Submittals of Reloads That Involve Unreviewed Safety Questions Page 69 of 109 * = See last page for status code definition.

nf ITEM TITLE R ADDITIONAL INFORMATION REV GL 84-021 Long Term Low Power Operation NA Info in Pressurized Water Reactors GL 84-022 This GL was never issued. NA GL 84-023 Reactor Vessel Water Level NA Boiling Water Reactor Instrumentation in BWRs GL 84-024 Certification of Compliance to Cl See Special Program for Environmental Qualification.

10 CFR 50.49, Environmental Qualification of Electric Equipment Important To Safety For Nuclear Power Plants GL 85-001 Fire Protection Policy Steering NA Only issued as draft Committee Report GL 85-002 Recommended Actions Stemming Cl TVA responded to the GL on June 17, 1985.

From NRC Integrated Program for the Resolution of Unresolved Unit 2 Action:

Safety Issues Regarding Steam Generator Tube Integrity Perform SG inspection.

GL 85-003 Clarification of Equivalent Control NA Boiling Water Reactor Capacity for Standby Liquid Control Systems GL 85-004 Operating Licensing Examinations NA Info GL 85-005 Inadvertent Boron Dilution Events NA Item was applicable only to units with operating license at the time the item was issued.

Info GL 85-006 Quality Assurance Guidance for NA ATWS Equipment That Is Not Safety-Related Item was applicable only to units with operating license at the time the GL 85-007 Implementation of Integrated NA item was issued.

Schedules for Plant Modifications Info GL 85-008 10 CFR 20.408 Termination NA Reports - Format Info GL 85-009 Technical Specifications For NA Generic Letter 83-28, Item 4.3 GL 85-010 Technical Specification For NA Applies only to Babcock and Wilcox designed plants Generic Letter 83-28, Items 4.3 and 4.4 Page 70 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 85-011 Completion of Phase II of "Control C See GL 81-07.

of Heavy Loads at Nuclear Power Plants," NUREG-0612 GL 85-012 Implementation Of TMI Action Item CI "Implementation of TMI Item II.K.3.5" - Reviewed in 15.5.4 of original 11,K.3.5, "Automatic Trip Of 1982 SER; became License Condition 35. The staff determined that their Reactor Coolant Pumps" 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.

GL 85-013 Transmittal Of NUREG-1 154 NA Info Regarding The Davis-Besse Loss Of Main And Auxiliary Feedwater Event GL 85-014 Commercial Storage At Power NA Item was applicable only to units with operating license at the time the Reactor Sites Of Low Level item was issued.

Radioactive Waste Not Generated By The Utility GL 85-015 Information On Deadlines For NA Item was applicable only to units with operating license at the time the 10 CFR 50.49, "Environmental - - - item was issued.

Qualification Of Electric Equipment Important To Safety At Nuclear Power Plants" NA Info GL 85-016 High Boron Concentrations NA Info GL 85-017 Availability Of Supplements 2 and 3 To NUREG-0933, "A Prioritization Of Generic Safety Issues" Info GL 85-018 Operator Licensing Examinations NA Info GL 85-019 Reporting Requirements On NA Primary Coolant Iodine Spikes GL 85-020 Resolution Of Generic Issue 69: NA Applies only to Babcock and Wilcox designed plants High Pressure Injection/Make-up Nozzle Cracking In Babcock And Wilcox Plants GL 85-021 This GL was never issued. NA GL 85-022 Potential For Loss Of Post-LOCA NA Info Recirculation Capability Due To Insulation Debris Blockage Page 71 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION Boiling Water Reactor GL 86-001 Safety Concerns Associated With NA Pipe Breaks In The BWR Scram System Boiling Water Reactor GL 86-002 Technical Resolution of Generic NA Issue B Thermal Hydraulic Stability GL 86-003 Applications For License NA Info Amendments GL 86-004 Policy Statement On Engineering C TVA responded to GL 86-04 on May 29, 1986. TVA provides engineering Expertise On Shift - expertise on shift in the form of a dedicated Shift Technical Advisor (STA) 01 or an STA qualified Senior Reactor Operator.

GL 86-005 Implementation Of TMI Action Item NA Applies only to Babcock and Wilcox designed plants II.K.3.5, "Automatic Trip Of -

Reactor Coolant Pumps" GL 86-006 Implementation Of TMI Action Item NA Applies only to Combustion Engineering designed plants I1.K.3.5, "Automatic Trip of Reactor Coolant Pumps" 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 U*

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 TITLE REV ADDITIONAL INFORMATION GL 86-010, Fire Endurance Test Acceptance NA Info S1 Criteria for Fire Barrier Systems Used to Separate Redundant Safe Shutdown Trains Within the Same Fire Area NA Does not apply to power reactor.

GL 86-011 Distribution of Products Irradiated in Research NA Does not apply to power reactor.

GL 86-012 Criteria for Unique Purpose Exemption From Conversion From The Use of Heu Fuel GL 86-013 Potential Inconsistency Between NA Applies only to Babcock and Wilcox and Combustion Engineering Plant Safety Analyses and -_- designed plants Technical Specifications NA Info GL 86-014 Operator Licensing Examinations NA Info GL 86-015 Information Relating To Compliance With 10 CFR 50.49, "Environmental Qualification of Electric Equipment Important To Safety For Nuclear Power Plants" NA Info GL 86-016 Westinghouse ECCS Evaluation Models GL 86-017 Availability of NUREG-1 169, NA Boiling Water Reactor "Technical Findings Related to Generic Issue C-8, BWR MSIC Leakage And Treatment Methods" GL 87-001 Public Availability Of The NRC NA Info Operator Licensing Examination Question Bank GL 87-002 Verification of Seismic Adequacy NA Item was applicable only to units with operating license at the time the and of Mechanical and Electrical item was issued.

GL 87-003 Equipment in Operating Reactors, USI A-46 GL 87-004 Temporary Exemption From NA Item was applicable only to units with operating license at the time the Provisions Of The FBI Criminal --- item was issued.

History Rule For Temporary Workers GL 87-005 Request for Additional Information NA Boiling Water Reactor on Assessment of License Measures to Mitigate and/or Identify Potential Degradation of Mark I Drywells Page 73 of 109 * = See last page for status code definition.

TITLE REV ITEM ADDITIONAL INFORMATION GL 87-006 Periodic Verification of Leak Tight NA Item was applicable only to units with operating license at the time the Integrity of Pressure Isolation item was issued.

Valves Info GL 87-007 Information Transmittal of Final NA Rulemaking For Revisions To Operator Licensing - 10 CFR 55 And Confirming Amendments Item was applicable only to units with operating license at the time the GL 87-008 Implementation of 10 CFR 73.55 NA item was issued.

Miscellaneous Amendments and Search Requirements Info GL 87-009 Sections 3.0 And 4.0 of Standard NA Tech Specs on Limiting Conditions For Operation And Surveillance Requirements Item was applicable only to units with operating license at the time the GL 87-010 Implementation of 10 CFR 73.57, NA item was issued.

Requirements For FBI Criminal History Checks Info GL 87-011 Relaxation in Arbitrary NA Intermediate Pipe Rupture Requirements GL 87-012 Loss of Residual Heat Removal C This GL was superseded by GL 88-17.

While The Reactor Coolant System is Partially Filled Does not apply to power reactor.

GL 87-013 Integrity of Requalification NA Examinations At Non-Power Reactors Info GL 87-014 Operator Licensing Examinations NA GL 87-015 Policy Statement On Deferred NA Info Plants _

GL 87-016 Transmittal of NUREG-1262, NA Info "Answers To Questions On Implementation of 10 CFR 55 On Operators' Licenses" GL 88-001 NRC Position on IGSCC in BWR NA Boiling Water Reactor Austenitic Stainless Steel Piping GL 88-002 Integrated Safety Assessment NA Item was applicable only to units with operating license at the time the Program II _- item was issued.

Page 74 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 88-003 Resolution of GSI 93, "Steam CI TVA: letter June 3, 1988. NRC letters dated Binding of Auxiliary Feedwater February 17, 1988 and July 20, 1988 Pumps" NRC: SSER16 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.

GL 88-004 Distribution of Gems Irradiated in NA Does not apply to power reactor.

Research Reactors GL 88-005 Boric Acid Corrosion of Carbon CI NRC acceptance letter dated August 8, 1990 for both units.

Steel Reactor Pressure Boundary Components in PWR plants 06 Unit 2 Action: Implement program.

REVISION 06 UPDATE:

The program has been implemented on Unit 2.

GL 88-006 Removal of Organization Charts NA Info from Technical Specification Administrative Control Requirements GL 88-007 Modified Enforcement Policy C1 See Special Program for Environmental Qualification.

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 NA Info Office of Nuclear Reactor Regulation GL 88-009 Pilot Testing of Fundamentals NA Boiling Water Reactor Examination GL 88-010 Purchase of GSA Approved NA Info Security Containers Page 75 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 88-011 NRC Position on Radiation S NRC acceptance letter dated June 29, 1989, for both units..

Embrittlement of Reactor Vessel Material and its Impact on Plant 02 Unit 2 Action: Submit Pressure Temperature curves.

Operations 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.

GL 88-012 Removal of Fire Protection NA Info Requirements from Technical Specification GL 88-013 Operator Licensing Examinations 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 GL 88-015 Electric Power Systems - NA Inadequate Control Over Design Process Info GL 88-016 Removal of Cycle-Specific NA Parameter Limits from Technical Specifications GL 88-017 Loss of Decay Heat Removal 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.

GL 88-018 Plant Record Storage on Optical NA Info Disks Page 76 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 88-019 Use of Deadly Force by Licensee NA Does not apply to power reactor.

Guards to Prevent Theft of Special Nuclear Material GL 88-020 Individual Plant Examination for S Unit 2 Action: Complete evaluation for Unit 2.

Severe Accident Vulnerabilities 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 NA Info and Procedural Controls for Radiological Effluent Technical Specifications GL 89-002 Actions to Improve the Detection C GL 89-02 did not require a response.

of Counterfeit and Fraudulently Marketed Products 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 NA Info Schedule GL 89-004 Guidelines on Developing OV NRC reviewed in 3.9.6 of SSER14 (Unit 1).

Acceptable Inservice Testing Programs 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 NA Info Examination Page 77 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 89-006 Task Action Plan Item I.D.2 - CI "Safety Parameter Display System" (SPDS) / "Requirements for Safety Parameter Display System Emergency Response Capability" - NRC reviewed in SSER5, SSER6,

- 10 CFR 50.54(f) 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 89-007 Power Reactor Safeguards C TVA: letter dated October 31, 1989 Contingency Planning for Surface Vehicle Bombs NRC: memo dated June 26, 1990 GL 89-008 Erosion/Corrosion-Induced Pipe Cl Unit 1 Flow Accelerated Corrosion Program reviewed in IR 390/94-89 Wall Thinning -- (February 1995).

Unit 2 Actions:

• Prepare procedure, and

. perform baseline inspections.

GL 89-009 ASME Section III Component NA Item was applicable only to units with operating license at the time the Replacements item was issued.

GL 89-010 Safety-Related Motor-Operated Cl NRC accepted approach in September 14, 1990, letter and reviewed in Valve Testing and Surveillance Appendix EE of SSER16.

Unit 2 Action: Implement pressure testing and surveillance program for safety-related MOVs, satisfying the intent of GL 89-10.

GL 89-010 or Involves Main Steam Isolation NA Boiling Water Reactor GL 96-005 Valves GL 89-011 Resolution of Generic Issue 101, NA Boiling Water Reactor "Boiling Water Reactor Water Level Redundancy" GL 89-012 Operator Licensing Examination NA Info GL 89-013 Service Water System Problems CI NRC letters dated July 9, 1990 and June 13, 1997, accepting approach.

Affecting Safety-Related Equipment 06 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.

nf ITEM TITLE ADDITIONAL INFORMATION REV GL 89-014 Line-Item Improvements in NA Info Technical Specifications -

Removal of 3.25 Limit on Extending Surveillance Intervals GL 89-015 Emergency Response Data NA Info System GL 89-016 Installation of a Hardened Wetwell NA Boiling Water Reactor Vent GL 89-017 Planned Administrative Changes NA Info to the NRC Operator Licensing Written Examination Process GL 89-018 Resolution of Unresolved Safety NA Info Issues A-17, "Systems Interactions in Nuclear Power Plants" GL 89-019 Request for Actions Related to CI TVA responded by letter dated March 22, 1990. NRC acceptance letter Resolution of Unresolved Safety - dated October 24, 1990, for both units.

Issue A-47, "Safety Implication of Control Systems in LWR Nuclear Unit 2 Action: Perform evaluation of common mode failures due to fire.

Power Plants" Pursuant to 10 CFR 50.54(f)

GL 89-020 Protected Area Long-Term NA Does not apply to power reactor.

Housekeeping GL 89-021 Request for Information S TVA responded to GL 89-21 with the status of USIs for both units on Concerning Status of - . November 29, 1989. NRC provided an assessment of WBN USI status on Implementation of Unresolved 06 May 1, 1990. The NRC assessment included a list of incomplete USIs for Safety Issue (USI) Requirements 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 C TVA: letter dated December 16, 1981 Loads and Plant Area Flood Runoff Depth At Licensed Nuclear Power Plants Due To Recent Change In Probable Maximum Answer to informal question provided in TVA letter dated Precipitation Criteria Developed by December 16, 1981, and subsequently included in FSAR. GL did not the National Weather Service require a response. No further action required.

GL 89-023 NRC Staff Responses to NA Info Questions Pertaining to Implementation of 10 CFR Part 26 GL 90-001 Request for Voluntary Participation NA Info in NRC Regulatory Impact Survey -

GL 90-002 Alternative Requirements for Fuel NA Info Assemblies in the Design Features Section of Technical Specifications GL 90-003 Relaxation of Staff Position in NA Info Generic Letter 83-28, Item 2.2 Part 2 "Vendor Interface for Safety-Related Components" GL 90-004 Request for Information on the C TVA responded on June 23, 1990 Status of Licensee Implementation of GSIs Resolved with Imposition of Requirements or CAs GL 90-005 Guidance for Performing NA Info Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping Page 80 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 90-006 Resolution of Generic Issues 70, S NRC letter dated January 9, 1991, accepted TVA's response for both "PORV and Block Valve _ _ units.

Reliability," and 94, "Additional 02 LTOP Protection for PWRs" 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.

TS Surveillance Requirement 3.4.11.2 specifies the required testing of each PORV.

GL 90-007 Operator Licensing National NA Info Examination Schedule GL 90-008 Simulation Facility Exemptions NA Info GL 90-009 Alternative Requirements for NA Info Snubber Visual Inspection Intervals and Corrective Actions GL 91 -001 Removal of the Schedule for the NA Info Withdrawal of Reactor Vessel Material Specimens from Technical Specifications Item was applicable only to units with operating license at the time the GL 91-002 Reporting Mishaps Involving LLW NA Forms Prepared for Disposal item was issued.

GL 91-003 Reporting of Safeguards Events NA Info NA Info GL 91-004 Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle NA Info GL 91-005 Licensee Commercial-Grade Procurement and Dedication Programs GL 91-006 Resolution of Generic Issue A-30, NA Item was applicable only to units with operating license at the time the "Adequacy of Safety-Related DC - - -item was issued.

Power Supplies," Pursuant to 10 CFR 50.54(f)

Page 81 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 91-007 GI-23, "Reactor Coolant Pump NA Info Seal Failures" and Its Possible Effect on Station Blackout GL 91-008 Removal of Component Lists from NA Info Technical Specifications GL 91-009 Modification of Surveillance NA Boiling Water Reactor Interval for the Electrical Protective Assemblies in Power Supplies for the Reactor Protection System GL 91-010 Explosives Searches at Protected NA Does not apply to power reactor.

Area Portals GL 91-011 Resolution of Generic Issues NA Item was applicable only to units with operating license at the time the A-48, "LCOs for Class 1E Vital item was issued.

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 NA Info Examination Schedule GL 91-013 Request for Information Related to NA Addressed to specific (non-TVA) plants.

Resolution of Generic Issue 130, "Essential Service Water System Failures @ Multi-Unit Sites" GL 91-014 Emergency Telecommunications NA Info GL 91-015 Operating Experience Feedback NA Info Report, Solenoid-Operated Valve Problems at U.S. Reactors GL 91-016 Licensed Operators' and Other NA Info Nuclear Facility Personnel Fitness for Duty GL 91-017 Generic, Safety Issue 29, "Bolting NA Info Degradation or Failure in Nuclear Power Plants" GL 91-018 Information to Licensees NA GL 91-18 has been superseded by RIS 2005-20.

Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability Page 82 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 91-019 Information to Addressees NA Info Regarding New Telephone Numbers for NRC Offices Located in One White Flint North GL 92-001 Reactor Vessel Structural Integrity 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.

GL 92-002 Resolution of Generic Issue 79, NA Info "Unanalyzed Reactor Vessel (PWR) Thermal Stress During Natural Convection Cooldown" GL 92-003 Compilation of the Current NA Info Licensing Basis: Request for Voluntary Participation in Pilot Program GL 92-004 Resolution of the Issues Related NA Boiling Water Reactor to Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(f)

NA Info GL 92-005 NRC Workshop on the Systematic Assessment of Licensee Performance (SALP) Program NA Info GL 92-006 Operator Licensing National Examination Schedule NA Info GL 92-007 Office of Nuclear Reactor Regulation Reorganization GL 92-008 Thermo-Lag 330-1 Fire Barriers 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.

2) Remove and replace, as required, or prepare an approved deviation.

GL 92-009 Limited Participation by NRC in NA Info the IAEA International Nuclear Event Scale GL 93-001 Emergency Response Data NA Addressed to specific plant(s).

System Test Program Page 83 of 109

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ITEM TITLE REV ADDITIONAL INFORMATION GL 93-002 NRC Public Workshop on NA Info Commercial Grade Procurement and Dedication GL 93-003 Verification of Plant Records NA Info NRC letter dated December 9, 1994, accepted TVA commitments for both GL 93-004 Rod Control System Failure and CO Withdrawal of Rod Control Cluster units.

Assemblies, 10 CFR 50.54(f) 06 Unit 2 Action: Implement modifications and testing.

REVISION 06 UPDATE:

NRC Inspection Report 391/2011-604 closed GL 93-004.

GL 93-005 Line-Item Technical Specifications NA Info Improvements to Reduce Surveillance Requirements for Testing During Power Operation GL 93-006 Research Results on Generic NA Info Safety Issue 106, "Piping and the Use of Highly Combustible Gases in Vital Areas" GL 93-007 Modification of the Technical NA Item was applicable only to units with operating license at the time the Specification Administrative _ .- item was issued.

Control Requirements for Emergency and Security Plans GL 93-008 Relocation of Technical NA Item was applicable only to units with operating license at the time the Specification Tables of Instrument - - - item was issued.

Response Time Limits GL 94-001 Removal of Accelerated Testing NA Item was applicable only to units with operating license at the time the and Special Reporting item was issued.

Requirements for Emergency Diesel Generators NA Boiling Water Reactor GL 94-002 Long-Term Solutions and Upgrade of Interim Operating Recommendations for Thermal-Hydraulic Instabilities in BWRs NA Boiling Water Reactor GL 94-003 IGSCC of Core Shrouds in BWRs GL 94-004 Voluntary Reporting of Additional NA Info Occupational Radiation Exposure Data Page 84 of 109 * = See last page for status code definition.

ITEM REV TITLE ADDITIONAL INFORMATION GL 95-001 NRC Staff Technical Position on NA Does not apply to power reactor.

Fire Protection for Fuel Cycle Facilities GL 95-002 Use of NUMARC/EPRI Report TR- NA Info 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 Cl NRC acceptance letter dated May 16, 1997 (Unit 1) - Initial response for Generator Tubes --- 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 NA Info Evaluation Program Lessons-Learned Issues Page 85 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 95-005 Voltage-Based Repair Criteria for C No specific action or response required by the GL; TVA responded on Westinghouse Steam Generator - - September 7, 2007.

Tubes Affected by Outside 06 Diameter Stress Corrosion ------------- -----.-. -.-.-.-. --.. . ..-.--.-.---.- .----- ... . . . . -.. . ..--.-.-.--.-.---.-.- .

Cracking 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 NA Info Licensing Program GL 95-007 Pressure Locking and Thermal Cl Unit 1 SER for GL 95-07 dated Sept 15, 1999 Binding of Safety-Related Power-Operated Gate Valves 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.

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ITEM TITLE REV 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)"

GL 95-008 10 CFR 50.54(p) Process for NA Info Changes to Security Plans Without Prior NRC Approval Info - - - - - - - -

GL 95-009 Monitoring and Training of NA Shippers and Carriers of Radioactive Materials Info GL 95-010 Relocation of Selected Technical NA Specifications Requirements Related to Instrumentation GL 96-001 Testing of Safety-Related Circuits Cl TVA responded for both units on April 18, 1996.

Unit 2 Action: Implement Recommendations.

GL 96-002 Reconsideration of Nuclear Power NA Info Plant Security Requirements Associated with an Internal Threat GL 96-003 Relocation of the Pressure Cl No response required Temperature Limit Curves and Low Temperature Overpressure 06 Unit 2 Actions:

Protection System Limits

" 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 NA Item was applicable only to units with operating license at the time the Fuel Pool Storage Racks _ item was issued.

GL 96-005 Periodic Verification of Design- CI SE of TVA response to GL 96-05 dated July 21, 1999.

Basis Capability of Safety-Related Motor-Operated Valves 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

• begin testing during the first refueling outage after startup.

GL 96-006 Assurance of Equipment C NRC letter dated April 6, 1999, accepting TVA response for Operability and Containment Unit 1.

Integrity During Design-Basis 06 Accident Conditions 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.

GL 96-007 Interim Guidance on NA Item was applicable only to units with operating license at the time the Transportation of Steam item was issued.

Generators Page 88 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 97-001 Degradation of Control Rod Drive CI NRC acceptance letter dated November 4, 1999 (Unit 1).

Mechanism Nozzle and Other Vessel Closure Head Penetrations 06 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'Cdue 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 June 30, 2010 (ADAMS Accession No. ML100539515)"

GL 97-002 Revised Contents of the Monthly NA Item was applicable only to units with operating license at the time the Operating Report item was issued.

GL 97-003 Annual Financial Update of Surety NA Does not apply to power reactor.

Requirements for Uranium Recovery Licensees NRC acceptance letter dated June 17, 1998 (Unit 1) - Initial response GL 97-004 Assurance of Sufficient Net CI Positive Suction Head for for Unit 2 on September 7, 2007.

Emergency Core Cooling and 06 Containment Heat Removal Pumps 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.

ITEM TITLE REV 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.

REVISION 06 UPDATE:

SSER22 contained the following for NRC Action:

"Closed. NRC Letter dated January 21, 2010 (ADAMS Accession No. ML093631061)"

GL 97-006 Degradation of Steam Generator CI NRC acceptance letter dated October 19, 1999 (Unit 1) - Initial Internals response for Unit 2 on September 7, 2007. TVA responded to a request 06 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 ADDITIONAL INFORMATION ITEM TITLE GL 98-001 Year 2000 Readiness of Computer NA Item was applicable only to units with operating license at the time the Systems at Nuclear Power Plants - - item was issued.

GL 98-002 Loss of Reactor Coolant Inventory Cl Initial response for Unit 2 on September 7, 2007.

and Associated Potential for Loss of Emergency Mitigation Functions 06 Unit 2 Actions:

While in a Shutdown Condition

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 NA Does not apply to power reactor.

Holders' Year 2000 Readiness Programs Page 91 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 98-004 Potential for Degradation of the Cl NRC closure letter dated November 24, 1999 (Unit 1). - Initial response ECCS and the Containment Spray - for Unit 2 on September 7, 2007.

System After a LOCA Because of 06 Construction and Protective Unit 2 Actions:

Coating Deficiencies and Foreign Material in Containment

• 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 NA Boiling Water Reactor 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 NA Info and Safeguards Decision on Bundling Exempt Quantities GL 99-002 Laboratory Testing of Nuclear NA Item was applicable only to units with operating license at the time the Grade Activated Charcoal - - - item was issued.

GL 03-001 Control Room Habitability 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.

ITEM TITLE REV 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)"

GL 04-001 Requirements for Steam Cl NRC acceptance letter dated April 8, 2005 (Unit 1) - Initial response for Generator Tube Inspection Unit 2 on September 7, 2007.

06 Unit 2 Action: Perform baseline inspection.

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 inthe 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 ADDITIONAL INFORMATION Operating License Application Update,' (ADAMS Accession number ML090700378)."

GL 06-002 Grid Reliability and the Impact on CI Initial response for Unit 2 on September 7, 2007.

Plant Risk and the Operability of Offsite Power 06 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.

GL 06-003 Potentially Nonconforming Hemyc CI TVA does not rely on Hemyc or MT materials to protect electrical and and MT Fire Barrier Configurations - instrumentation cables or equipment that provide safe shutdown capability 06 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.

REVISION 02 UPDATE:

NRC issued the Safety Evaluation for Generic Letter 2006-003 on February 25, 2010.

REVISION - --- - -

-UPDATE:

068----

---

---7 04--

SSER22 contained the following for NRC Action:

"Closed. NRC Letter dated February 25, 2010 (ADAMS Accession No. ML100470398)"

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ITEM TITLE REV ADDITIONAL INFORMATION GL 07-001 Inaccessible or Underground CI Initial response for Unit 2 on September 7, 2007.

Power Cable Failures That Disable Accident Mitigation Systems or 06 Unit 2 Action: Complete testing of four additional cables.

Cause Plant Transients 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)"

GL 08-001 Managing Gas Accumulation in 0 Initial response for Unit 2 on October 1, 2008.

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.

ITEM TITLE REV 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 inthe 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- Shift Technical Advisor NA Not applicable to WBN per SSER16.

0737, I.A.1.1 NUREG- Shift Supervisor Responsibilities NA Not applicable to WBN per SSER16.

0737, I.A.1 .2 NUREG- Shift Manning C Closed in SSER16.

0737, I.A.1.3 NUREG- Immediate Upgrade of RO and C Closed in SSER16.

0737, SRO Training and Qualifications I.A.2.1 NUREG- Administration of Training C Closed in SSER16.

0737, Programs I.A.2.3 NUREG- Revise Scope and Criteria for C Closed in SSER16.

0737, Licensing Exams I.A.3.1 Page 98 of 109 * = See last page for status code definition.

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

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.

NUREG- Short Term Accident and Cl NRC reviewed in Appendix EE of SSER16.

0737, Procedure Review I.C.1 Unit 2 Action: Implement upgraded Emergency Operating Procedures, including validation and training.

NUREG- Shift and Relief Turnover C Closed in SSER16.

0737, Procedures I.C.2 NUREG- Shift Supervisor Responsibility C Closed in SSER16.

0737, I.C.3 NUREG- Control Room Access C Closed in SSER16.

0737, I.C.4 NUREG- Feedback of Operating Experience C Closed in SSER16.

0737, I.C.5 Page 99 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION NUREG- Verify Correct Performance of C Closed in SSER16.

0737, Operating Activities I.C.6 NUREG- NSSS Vendor Revision of cI IR 50-390/391 85-08 closed this item for Unit 1, and NRC also reviewed in 0737, Procedures Appendix EE of SSER16.

I.C.7 Unit 2 Action: Revise power ascension and emergency procedures which were reviewed by Westinghouse.

NUREG- Pilot Monitoring of Selected Cl IR 50-390/391 85-08 closed this item for Unit 1, and NRC also reviewed in 0737, Emergency Procedures For Near Appendix EE of SSER16.

I.C.8 Term Operating Licenses Unit 2 Action: Pilot monitor selected emergency procedures for NTOL.

NUREG- Control Room Design Review Cl NRC reviewed in SSER5, SSER6, SSER15, and Appendix EE of 0737, SSER16.

I.D.1 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.

REVISION 06 UPDATE:

SSER22 contained the following for NRC Action:

"Closed in SSER22, Section 18.2" Plant-Safety-Parameter-Display NUREG- Cl NRC reviewed in SSER5, SSER6, and 18.2.2 of SSER15.

0737, Console I.D.2 Unit 2 Action: Install SPDS and have it operational prior to start-up after the first refueling outage.

NUREG- Training During Low-Power Testincg C Closed in SSER16.

0737, I.G.1 NUREG- Reactor Coolant Vent System Cl LICENSE CONDITION - NUREG-0737, ll.B.1, "Reactor Coolant System 0737, Vents" - In the original SER, the NRC found ll.B.1 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.

Page 100 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION NUREG- Plant Shielding CI NRC reviewed in Appendix EE of SSER16.

0737, ll.B.2 Unit 2 Action: Complete Design Review of EQ of equipment for spaces/systems which may be used in post accident operations.

NUREG- Post-Accident Sampling S NRC reviewed in 9.3.2 of SSER16. TVA submitted a TS improvement to 0737, - - eliminate requirements for the Post Accident Sampling System using the ll.B.3 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- Training for Mitigating Core C Closed in SSER16.

0737, Damage ll.B.4 NUREG- Relief and Safety Valve Test Cl NRC reviewed in Technical Evaluation Report attached to Appendix EE of 0737, Requirements SSER15.

ll.D.1 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.

NUREG- Valve Position Indication CI The design was reviewed in the original 1982 SER and found acceptable 0737, _ - pending confirmation of installation of the acoustic monitoring system. In ll.D.3 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.

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ITEM TITLE REV ADDITIONAL INFORMATION NUREG- Auxiliary Feedwater System CI Reviewed in Appendix EE of SSER16.

0737, Evaluation, Modifications II.E.1.1 Unit 2 Action: Perform Auxiliary Feedwater System analysis as it pertains to system failure and flow rate.

NUREG- Auxiliary Feedwater System CI NRC: IR 50-390/84-20 and 50-391/84-16; letters dated 0737, Initiation and Flow __-. March 29, 1985, and October 31, 1995; SSER 16 II.E.1.2 Unit 2 Actions:

" Complete procedures, and

• qualification testing.

NUREG- Emergency Power For Pressurizer Cl NRC: letters dated March 29, 1985, and October 31, 1995; SSER 16 0737, Heaters I1.E.3.1 Reviewed in original 1982 SER.

Unit 2 Action: Implement procedures and testing.

NUREG- Dedicated Hydrogen Penetrations C NRC: IR 50-390/83-27 and 50-391/83-19; SER (NUREG-0847)

0737, I1.E.4.1 NUREG- Containment Isolation S TVA: letters dated October 29, 1981, and February 25, 1985 0737, Dependability I1.E.4.2 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.

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ITEM TITLE REV 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."

NUREG- Accident-Monitoring Cl Reviewed in SSER9.

0737, Instrumentation - Noble Gas II.F.1.2.a. Unit 2 Actions: Install Noble gas, Iodine / particulate sampling, and Containment High Range Monitors.

Unit 2 Action: Install Noble gas monitor for Unit 2.

NUREG- Accident-Monitoring CI Reviewed in SSER9.

0737, Instrumentation - _ -.

II.F.1.2.b. Iodine/Particulate Sampling Unit 2 Actions: Install Noble gas, Iodine / particulate sampling, and Containment High Range Monitors.

Unit 2 Action: Install Iodine / particulate sampling monitor for Unit 2.

NUREG- Accident-Monitoring CI Reviewed in SSER9.

0737, Instrumentation - Containment ___

I1.F.1.2.c. 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.

NUREG- Accident-Monitoring CO Reviewed in SSER9.

0737, Instrumentation - Containment II.F.1.2.d. 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.

NUREG- Accident-Monitoring Cl Reviewed in SSER9.

0737, Instrumentation - Containment ___

II.F.1.2.e. 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- Accident-Monitoring CO Reviewed in SSER9.

0737, Instrumentation - Containment II.F.1.2.f. Hydrogen 06 Unit 2 Action: Verify installation of containment hydrogen accident monitoring instrumentation.

REVISION 06 UPDATE:

NRC Inspection Report 391/2011-604 closed NUREG-0737, II.F.1.2.F.

NUREG- Instrumentation For Detection of 0 LICENSE CONDITION - Detectors for Inadequate core cooling (II.F.2) 0737, 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- Power Supplies For Pressurizer CI Reviewed in original 1982 SER and 8.3.3 of SSER7.

0737, Relief Valves, Block Valves and II.G.1 Level Indicators 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- Review ESF Valves C NRC: letter dated March 29, 1985; SSER 16

0737, I1.K.1.5 NUREG- Operability Status Cl Unit 2 Action: Confirm multi-unit operation will have no impact on 0737, - -_ administrative procedures with respect to operability II.K.1.10 status.

NUREG- Trip Per Low-Level B/S C NRC: letter dated March 29, 1985; SSER 16

0737, II.K.1.17 Page 104 of 109

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ITEM TITLE REV ADDITIONAL INFORMATION NUREG- Effect of High Pressure Injection C LICENSE CONDITION - Effect of high pressure injection for small break 0737, for Small Break LOCA With No - -. LOCA with no auxiliary feedwater II.K.2.13 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.

NUREG- Voiding in the Reactor Coolant C LICENSE CONDITION - Voiding in the reactor coolant system 0737, System _ -. (NUREG-0737, I1.K.2.17)

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.

NUREG- Auto PORV Isolation C Reviewed in SSER5 and resolved based on NRC conclusion that there is 0737, _ -_ no need for an automatic PORV isolation system (NRC letter dated June II.K.3.1 29, 1990).

NUREG- Report on PORV Failures C Reviewed in SSER5 and resolved based on NRC conclusion that there is 0737, _ -_ no need for an automatic PORV isolation system (NRC letter dated June I1.K.3.2 29, 1990).

NUREG- Reporting SV/RV C (Action from GL 82-16) - NRC reviewed in Appendix EE of SSER16.

0737, Failures/Challenges I1.K.3.3 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."

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ITEM TITLE REV ADDITIONAL INFORMATION NUREG- Auto Trip of RCPS CI Reviewed in 15.5.4 of original 1982 SER; became License Condition 35.

0737, The staff determined that their review of Item II.K.3.5 did not have to be II.K.3.5 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- PID Controller Cl Reviewed in original 1982 SER.

0737, 11.K.3.9 06 Unit 2 Action: Set the derivative time constant to zero.

REVISION 06 UPDATE:

The derivative time constant was set to zero.

NUREG- Anticipatory Trip at High Power S NRC: letter dated October 31, 1995; SSER 16

0737, I1.K.3.10 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- Confirm Existence of Anticipatory C Closed in SSER16.

0737, Reactor Trip Upon Turbine Trip II.K.3.12 NUREG- Report On Outage of Emergency C LICENSE CONDITION - Report on outage of emergency core cooling 0737, Core Cooling System system (NUREG-0737, I1.K.3.17)

II.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.

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ITEM TITLE REV ADDITIONAL INFORMATION NUREG- Power On Pump Seals C NRC reviewed and closed in IR 390/84-35 based on Diesel Generator 0737, -- (DG) power to pump sealing cooling system.

II.K.3.25 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- Small Break LOCA Methods C TVA: letter dated October 29, 1981

0737, I1.K.3.30 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.

NUREG- Plant Specific Analysis C The staff determined in SSER4 that their review of Items II.K.3.30 and 0737, _l__ II.K.3.31 did not have to be completed to support the full-power license I1.K.3.31 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.

Unit 2 Action: Complete analysis for Unit 2.

REVISION 06 UPDATE:

Page 107 of 109 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION The analysis has been completed for Unit 2.

NRC Inspection Report 391/2011-603 closed NUREG-0737, II.K.3.31.

NUREG- Emergency Preparedness, Short C LICENSE CONDITION - Emergency Preparedness (NUREG-0737, 0737, Term III.A.1, III.A.2)

III.A.1.1 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.

NUREG- Upgrade Emergency Support C LICENSE CONDITION - Emergency Preparedness (NUREG-0737, 0737, Facilities III.A.1, III.A.2)

III.A.1.2 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.

NUREG- Emergency Preparedness C LICENSE CONDITION - Emergency Preparedness (NUREG-0737, 0737, III.A.1, III.A.2)

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.

NUREG- Primary Coolant Outside S Resolved for Unit I only in SSER10; reviewed in Appendix EE of 0737, Containment SSER16.

II1.D.1.1 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.

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ITEM TITLE REV ADDITIONAL INFORMATION NUREG- In-Plant Iodine Radiation Cl NRC reviewed in Appendix EE of SSER16.

0737, Monitoring _ -

II1.D.3.3 Unit 2 Action: Complete modifications for Unit 2.

NUREG- Control-Room Habitability Cl TVA: letter dated October 29, 1981

0737, II1.D.3.4 06 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.

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