ML080850253

Unit 2 - Generic Communications Status for Unit 2 - Restructured Tables

ML080850253
Person / Time
Site:	Watts Bar
Issue date:	03/20/2008
From:	Bajestani M Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML080850253 (132)

Text

Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 March 20, 2008

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U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop: OWFN P1-35 Washington, D.C. 20555-0001 Gentlemen:

In the Matter of Docket No. 50-391 Tennessee Valley Authority )

WATTS BAR NUCLEAR PLANT (WBN) - UNIT 2 - GENERIC COMMUNICATIONS STATUS FOR UNIT 2 - RESTRUCTURED TABLES The purpose of this letter is to provide a comprehensive list of Generic Communications for Watts Bar Nuclear Plant (WBN) Unit 2, reviewed and categorized by status consistent with previously submitted tables. In Reference 1, TVA provided a summary of the status of Generic Letters and Bulletins issued prior to 1995 for WBN Unit 2. In Reference 2, TVA provided the initial responses to Generic Letters and Bulletins for WBN Unit 2 issued subsequent to the licensing of WBN Unit 1. TVA further reviewed Generic Communications as part of the regulatory framework for the completion of construction and licensing activities for WBN Unit 2 that was submitted in Reference 3. Based on subsequent discussions with the staff, TVA agreed to supplement the Regulatory Framework tables for the Safety Evaluation Report and Supplements (NUREG-0847) to allow ease of review and provide additional definitions for section status. Reference 4 provided the restructured tables. In the tables provided in Reference 4, Bulletins, Generic Letters, and NUREG-0737 items were addressed only if they were explicitly included in the Safety Evaluation Report and its Supplements related to the operation of WBN Units 1 and 2 (NUREG-0847). In Reference 4, TVA committed to provide additional information on Generic Communications. This letter provides tables for Generic Communications that are consistent with the tables provided in Reference 4 for the Regulatory Framework for the completion of construction and licensing activities for WBN Unit 2.

KA(2-R

U.S. Nuclear Regulatory Commission Page 2 March 20, 2008 The following provides the seven definitions used in the attached tables:

1. Closed (C): 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.

2. Closed/Implementation (Cl): 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.

3. Closed/Technical Specification (CT): 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.

4. Not Applicable (NA): Justification as to why a section/subsection is not applicable is provided in the Additional Information section of the tables.

5. Open (0): No action or documentation is provided that shows that the staff has reviewed the item for WBN Unit 2.

6. Open/Technical Specifications (OT): No action or documentation is provided that shows that the staff has reviewed the item for WBN Unit 2, and the resolution is through submittal of a Technical Specification.

7. Open/Validation (OV): The proposed approach has been approved for WBN Unit 1; the same approach used is proposed for WBN Unit 2 without change.

To present the results of TVA's review of the WBN Unit 2 Generic Communications, a series of tables (Tables 1 - 8) were developed. These tables list Bulletins, Circulars, and Generic Letters and provide the current status for WBN Unit 2. In cases where there were revisions to Generic Communications, the latest revision is addressed. A brief description of these tables is provided below:

• Table 1 - Generic Communications Matrix Master Table This table provides the comprehensive review of the WBN Unit 2 Generic Communications. The seven results described above were applied in this master table.

" Table 2 - Generic Communications Status = Closed

" Table 3 - Generic Communications Status = Closed/Implementation

" Table 4 - Generic Communications Status = Closed/Technical Specifications

" Table 5 - Generic Communications Status = Not Applicable

• Table 6 - Generic Communications Status = Open

U.S. Nuclear Regulatory Commission Page 3 March 20, 2008

" Table 7 - Generic Communications Status = Open/Technical Specifications

" Table 8 - Generic Communications Status = OpenNalidation Should TVA determine, based on further review or other emerging issues, that a different approach or additional action is appropriate, TVA will submit such changes to the NRC for review and concurrence. TVA will also provide periodic updates to the generic communication tables as actions are completed.

If you have any questions, please contact me at (423) 365-2351.

Sincerely,

-*'- Masoud Bajestani Watts Bar Unit 2 Vice President Enclosure cc: See page 4

References:

1. TVA letter dated September 7, 2007, 'Watts Bar Nuclear Plant (WBN) - Unit 2 -

Generic Communications Issued Prior to 1995"

2. TVA letter dated September 7, 2007, 'Watts Bar Nuclear Plant (WBN) - Unit 2 -

Initial Responses to Bulletins and Generic Letters"

3. TVA letter dated January 29, 2008, 'Watts Bar Nuclear Plant (WBN) Unit 2 -

Regulatory Framework for the Completion of Construction and Licensing Activities for Unit 2"

4. TVA letter dated March 13, 2008, "Watts Bar Nuclear Plant (WBN) Unit 2 -

Regulatory Framework for the Completion of Construction and Licensing Activities for Unit 2 - Restructured Tables"

U.S. Nuclear Regulatory Commission Page 4 March 20, 2008 cc (Enclosure):

Lakshminarasimh Raghavan U.S. Nuclear Regulatory Commission MS 08H4A One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 Joseph Williams, Senior Project Manager (WBN Unit 2)

U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 Loren R. Plisco, Deputy Regional Administrator for Construction U. S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center, Suite 23T85 61 Forsyth Street, SW, Atlanta, Georgia 30303-8931 U. S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-8931 NRC Resident Inspector Unit 2 Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381

GENERIC COMMUNICATIONS MASTER TABLE

GENERIC COMMUNICATIONS MASTER TABLE ITEM TITLE ADDITIONAL INFORMATION B 71-001 Involves Main Steam Isolation Valves NA Boiling Water Reactor B 71-002 PWR Reactor Trip Circuit Breakers NA Addressed to specific plant(s).

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

Relief Valve Headers B 72-001 Failed Hangers for Emergency Core NA Addressed to specific plant(s).

Cooling System Suction Header B 72-002 Simultaneous Actuation of a Safety NA Addressed to specific plant(s).

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

B 73-001 Faulty Overcurrent Trip Delay Device in C TVA: letter dated April 4, 1973 Circuit Breakers for Engineered Safety Systems NRC: IR 390/391 75-5 B 73-002 Malfunction of Containment Purge Supply C TVA: letter dated August 22, 1973 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 Hydraulic C TVA: memo dated February 7, 1985 Shock Absorbers NRC: IR 390/391 85-08 B 73-005 Manufacturing Defect in BWR Control Rods NA Boiling Water Reactor 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

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B 74-002 Truck Strike Possibility NA Info B 74-003 Failure of Structural or Seismic Support Cl TVA: memo dated January 22, 1985 Bolts on Class I Components 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.

B 74-004 Malfunction of Target Rock Safety Relief NA Boiling Water Reactor Valves B 74-005 Shipment of an Improperly Shielded Source NA Does not apply to power reactor.

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

ITEM TITLE

• ADDITIONAL INFORMATION B 74-006 Defective Westinghouse Type W-2 Control C TVA: letter dated October 18, 1974 Switch Component NRC: I1 390/391 75-6 B 74-007 Personnel Exposure - Irradiation Facility NA Does not apply to power reactor.

B 74-008 Deficiency in the ITE Molded Case Circuit C TVA: letter dated August 21, 1974 Breakers, Type HE-3 NRC: IR 390/391 75-5 B 74-009 Deficiency in GE Model 4KV Magne-Blast C TVA: letter dated September 20, 1974 Circuit Breakers NRC: IR 390/391 76-6 B 74-010 Failures in 4-Inch Bypass Pipe at Dresden 2 NA Boiling Water Reactor B 74-011 Improper Wiring of Safety Injection Logic at C NRC: IR 390/391 75-6 Zion 1 & 2 B 74-012 Incorrect Coils in Westinghouse Type SG C NRC: IR 390/391 75-5 Relays at Trojan B 74-013 Improper Factory Wiring on GE Motor C TVA: letter dated December 24, 1974 Control Centers at Fort Calhoun NRC: IR 390/391 75-5 B 74-014 BWR Relief Valve Discharge to NA Boiling Water Reactor Suppression Pool B 74-015 Misapplication of Cutler-Hammer Three CI TVA: letter dated May 5, 1975 Position Maintained Switch Model No.

10250T NRC: IR 390/391 75-5 Unit 2 Action: Install modified A3 Cutler-Hammer 10250T switches.

B 74-016 Improper Machining of Pistons in Colt C TVA: letter dated January 2, 1975 Industries (Fairbanks-Morse)

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

Devices and Source Changers B 75-003 Incorrect Lower Disc Spring and Clearance CI TVA: letter dated May 16, 1975 Dimension in Series 8300 and 8302 ASCO Solenoid Valves 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: 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 2 of 60 * = See last page for status code definition.........................................

Page 2 of 60 * = See last page for status code definition.

ITEM TITLE

• ADDITIONAL INFORMATION B 75-005 Operability of Category I Hydraulic Shock CI TVA: letter dated June 16, 1975 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-2 Control Cl TVA: letter dated July 31, 1975 Switches NRC: IR 390/85-25 and 391/85-20 Unit 2 Action: Inspect Westinghouse Type OT-2 control switches.

...... ...... ...... ... ...... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... o ... ... ... ... ... ... ... ... ... .. .. .. .. .. .

B 75-007 Exothermic Reaction in Radwaste Shipment NA Does not apply to power reactor.

B 75-008 PWR Pressure Instrumentation CT NRC: IR 390/391 85-08 Unit 2 Action: Ensure that Technical Specifications and Site Operating Instructions address importance of maintaining temperature and pressure within prescribed limits.

B 76-001 BWR Isolation Condenser Tube Failure NA Boiling Water Reactor B 76-002 Relay Coil Failures - GE Types HFA, Cl Unit 2 Action: Repair or replace relays before HGA, HKA, HMA Relays preoperational tests.

B 76-003 Relay Malfunctions - GE Type STD Relays C TVA: letter dated May 17, 1976 NRC: IR 390/391 76-6 B 76-004 Cracks in Cold Worked Piping at BWRs NA Boiling Water Reactor B 76-005 Relay Failures - Westinghouse BFD C TVA: letter dated June 7, 1976 Relays NRC: IR 390/391 85-08 B 76-006 Diaphragm Failures in Air Operated C TVA: memo dated January 25, 1985 Auxiliary Actuators for Safety/Relief Valves NRC: IR 390/391 85-08 B 76-007 Crane Hoist Control Circuit Modifications C TVA: letter dated October 29, 1976 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 Setpoint Drift C TVA: letter dated July 1, 1977 NRC: IR 390/391 85-08 B 77-002 Potential Failure Mechanism in Certain C TVA: letter dated November 11, 1977 Westinghouse AR Relays with Latch Attachments NRC: IR 390/391 85-08

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ITEM TITLE ADDITIONAL INFORMATION B 77-003 On-Line Testing of the Westinghouse Solid Cl Unit 2 Action: Include necessary periodic testing in test State Protection System procedures.

B 77-004 Calculation Error Affecting The Design CT TVA: letter dated January 23, 1978 Performance of a System for Controlling pH of Containment Sump Water Following a NRC: IR 390/78-11 and 391/78-09 LOCA Unit 2 Action: Ensure Technical Specifications includes limit on Boron concentration.

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

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 Safeguards C Item concerns a multi-unit issue that was completed for During an Emergency - Locking Systems both units.

TVA: letter dated March 1, 1978 NRC: IR 390/78-11 and 391/78-09 B 78-001 Flammable Contact - Arm Retainers in C TVA: letter dated March 20, 1978 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 Certain Stem Cl TVA: letter dated December 19, 1978 Mounted Limit Switches Inside Reactor Containment NRC: IR 390/82-13 and 391/82-10 Closed to EQ Program IR 50-390/82-13 and 50-391/82-10 (April 22, 1982) accepted approach.

Unit 2 Action: Ensure NAMCO switches have been replaced.

B 78-005 Malfunctioning of Circuit Breaker Auxiliary C TVA: letter dated June 12, 1978 Contact Mechanism - GE Model CR105X NRC: IR 390/78-17 and 391/78-15 Page 4 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION B 78-006 Defective Cutler-Hammer Type M Relays C NRC: IR 390/78-22 and 391/78-19 With DC Coils

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B 78-007 Protection Afforded by Air-Line Respirators NA Item was applicable only to units with operating license at and Supplied-Air Hoods the time the item was issued.

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

NRC: IR 390/391 85-08 B 78-009 BWR Drywell Leakage Paths Associated NA Boiling Water Reactor 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 Containment Torus NA Boiling Water Reactor Welds B 78-012 Atypical Weld Material in Reactor Pressure C TVA: Westinghouse letter dated October 29, 1979 Vessel Welds NRC: IR 390/391 81-04 B 78-013 Failures in Source Heads Kay Ray, Inc. NA Does not apply to power reactor.

Gauges Models 7050, 7050B, 7051, 7051B, 7060, 7060B, 7061 and 7061B B 78-014 Deterioration of Buna-N Components in NA Boiling Water Reactor ASCO Solenoids B 79-001 Environmental Qualification of Class 1E C NRC: IR 390/80-06 and 391/80-05 Equipment B 79-002 Pipe Support Base Plate Designs Using Cl NRC review of HAAUP Program in NUREG-1 232, SSER6, Concrete Expansion Anchor Bolts and SSER8.

Unit 2 Actions: Addressed in CAP/SP. Conduct a complete review of affected support calculations, and perform the necessary revisions to design documents and field modifications to achieve compliance.

B 79-003 Longitudinal Weld Defects in ASME SA-312 C TVA: letter dated July 16, 1981 Type 304 SS Pipe Spools Manufactured by Youngstown Welding & Engineering NRC: IRs 390/82-21 and 391/82-17; 390/84-35 and 391/84-33 B 79-004 Incorrect Weights for Swing Check Valves C TVA: letter dated October 20, 1980 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 System C NRC: IR 390/80-06 and 391/80-05 Misalignments Identified During the Three Mile Island Incident B 79-007 Seismic Stress Analysis of Safety-Related C TVA: letter dated May 31, 1979 Piping NRC: IR 390/79-30 and 391/79-25 Page 5 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION B 79-008 Events Relevant to BWRs Identified During NA Boiling Water Reactor TMI Incident B 79-009 Failure of GE Type AK-2 Circuit Breaker in Cl TVA: letter dated June 20, 1979 Safety Related Systems Unit 2 Action: Complete preservice preventive maintenance on AK-2 Circuit Breakers.

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

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

Unit 2 Actions: Addressed in CAP/SP. Initiate a Unit 2 hanger walkdown and hanger analysis program similar to the program for Unit 1. Complete re-analysis of piping and associated supports as necessary. Perform modifications as required by re-analysis.

B 79-015 Deep Draft Pump Deficiencies C TVA: letter dated January 24, 1992 NRC: IR 390/391 95-70 B 79-016 Vital Area Access Controls NA Item was applicable only to units with operating license at the time the item was issued.

NRC: IR 390/80-06 and 391/80-05 B 79-017 Pipe Cracks in Stagnant Borated Water NA Item was applicable only to units with operating license at Systems at PWR Plants the time the item was issued.

NRC: IR 390/80-06 and 391/80-05; NUREG/ CR 5286 B 79-018 Audibility Problems Encountered on NA Item was applicable only to units with operating license at Evacuation of Personnel from High-Noise the time the item was issued.

Areas NRC: IR 390/80-06 and 391/80-05 Page 6 of 60 * = See last page for status code definition.

ITEM TITLE

• ADDITIONAL INFORMATION B 79-019 Packaging of Low-Level Radioactive Waste NA Item was applicable only to units with operating license at for Transport and Burial the time the item was issued.

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 Low-Level Radioactive Waste the time the item was issued.

NRC: IR 390/80-06 and 391/80-05 B 79-021 Temperature Effects on Level CI Reviewed in 7.2.5 of both the original 1982 SER and Measurements SSER14.

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.

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

Used in Time Pieces for Luminosity NRC: IR 390/80-06 and 391/80-05 B 79-023 Potential Failure of Emergency Diesel C TVA: letter dated October 29, 1979 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 Relays in C TVA: letter dated January 4, 1980 Safety-Related Systems NRC: IR 390/80-03 and 391/80-02 B 79-026 Boron Loss from BWR Control Blades NA Boiling Water Reactor B 79-027 Loss of Non-Class 1E I & C Power System Cl TVA responded to the Bulletin on March 1, 1982. Reviewed Bus During Operation in 7.5.3 of the original 1982 SER.

Unit 2 Action: Issue appropriate emergency procedures.

B 79-028 Possible Malfunction of NAMCO Model C TVA: letter dated April 1, 1993 EA1 80 Limit Switches at Elevated Temperatures NRC: IR 390/391 93-32 B 80-001 Operability of ADS Valve Pneumatic Supply NA Boiling Water Reactor B 80-002 Inadequate QA for Nuclear Supplied NA Boiling Water Reactor Equipment Page 7 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION B 80-003 Loss of Charcoal from Standard Type II, C TVA: letter dated March 21, 1980 2 Inch, Tray Adsorber Cells NRC: IR 390/80-15 and 391/80-12 B 80-004 Analysis of a PWR Main Steam Line Break CI IR 50-390/85-60 and 50-391/85-49 (December 6, 1985) with Continued Feedwater Addition required completion of actions that included determination of temperature profiles inside and outside of containment following a MSLB for Unit 1.

Unit 2 Action: Complete analysis for Unit 2.

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

Chemical Volume Control System Holdup Tanks Unit 2 Action: Complete surveillance procedures for Unit 2.

B 80-006 Engineered Safety Feature Reset Control Cl TVA response dated March 11, 1982. Reviewed in 7.3.5 of the original 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 B 80-010 Contamination of Nonradioactive System Cl Unit 2 Actions: 1) Correct deficiencies involving monitoring and Resulting Potential for Unmonitored, of systems; and 2) include proper monitoring of Uncontrolled Release of Radioactivity to non-radioactive systems in procedures.

Environment 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 1 in IR 50-390/95-46 (August 1, 1995).

Unit 2 Action: Complete implementation for Unit 2.

B 80-012 Decay Heat Removal System Operability Cl NRC: IR 390/391 85-08; NUREG/CR 4005 Unit 2 Action: Implement operating instructions and abnormal operating instructions (AOIs) for RHR.

B 80-013 Cracking in Core Spray Spargers NA Boiling Water Reactor B 80-014 Degradation of Scram Discharge Volume NA Boiling Water Reactor Capability B 80-015 Possible Loss of Emergency Notification C Item concerns a multi-unit issue that was completed for System with Loss of Offsite Power both units.

NRC: IR 390/391 85-08.

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ITEM TITLE ADDITIONAL INFORMATION B 80-016 Potential Misapplication of Rosemount, Inc. C TVA: letter dated August 29, 1980 Models 1151 and 1152 Pressure Transmitters With Either "A"or "D"Output NRC: IR 390/391 81-17 Codes B 80-017 Failure of 76 of 185 Control Rods to Fully NA Boiling Water Reactor Insert During a Scram at a BWR B 80-018 Maintenance of Adequate Minimum Flow Cl IR 50-390/85-60 and 50-391/85-49 (Unit 1)

Thru Centrifugal Charging Pumps Following Secondary Side High Energy Rupture Unit 2 Action: Implement design and procedure changes.

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

Return to Neutral Control Switches

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

NRC: 390/391 85-08 B 80-022 Automation Industries, Model 200-520-008 NA Does not apply to power reactor.

Sealed-Source Connectors B 80-023 Failures of Solenoid Valves Manufactured C TVA: letter dated March 31, 1981 by Valcor Engineering Corporation NRC: IR 390/391 81-17; NUREG/CR 5292

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

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

B 82-003 Stress Corrosion Cracking in Thick-Wall, NA Boiling Water Reactor Large Diameter, Stainless Steel, Recirculation System Piping at BWR Plants Page 9 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION B 82-004 Deficiencies in Primary Containment C TVA: letter dated January 24, 1983 Electrical Penetration Assemblies NRC: IR 390/83-10 and 391/83-08 B 83-001 Failure of Trip Breakers (Westinghouse C NRC: IRs 390/391 85-08 and 390/391 92-13 DB-50) to Open on Automatic Trip Signal

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B 83-002 Stress Corrosion Cracking in Large- NA Boiling Water Reactor Diameter Stainless Steel Recirculation System Piping at BWR Plants B 83-003 Check Valve Failures in Raw Water Cooling NA Addressed by Inservice Testing for Construction Permit Systems of Diesel Generators holders B 83-004 Failure of the Undervoltage Trip Function of Cl NRC: IR 390/391 85-08 Reactor Trip Breakers Unit 2 Action: Install new undervoltage attachment with wider grooves on the reactor trip breakers.

B 83-005 ASME Nuclear Code Pumps and Spare C TVA: letter dated September 7, 1983 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 by Cl TVA: letter dated February 2, 1984 Tube-Line Facilities 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.

B 83-007 Apparently Fraudulent Products Sold by C TVA: letter dated March 22, 1984 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 Than the NRC: IR 390/84-35 and 391/84-33 Reactor Trip System B 84-001 Cracks in BWR Mark 1 Containment Vent NA Boiling Water Reactor Headers B 84-002 Failure of GE Type HFA Relays In Use In C TVA: letter dated July 10, 1984 Class 1E Safety Systems NRC: IR 390/391 84-42 and IR 390/84-'7 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 (AOls) are used for Unit 2.

Page 10 of 60 * = See last page for status code definition.

ITEM TITLE

• ADDITIONAL INFORMATION B 85-001 Steam Binding of Auxiliary Feedwater Cl TVA: letter dated January 27, 1986 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.

B 85-002 Undervoltage Trip Attachment of Cl Unit 2 Action: Install automatic shunt trip on the Westinghouse DB-50 Type Reactor Trip Westinghouse DS-416 reactor trip breakers on Unit 2.

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

Terminate Treatment Dose B 87-001 Thinning of Pipe Walls in Nuclear Power C TVA: letter dated September 18, 1987 Plants NRC: NUREG/CR 5287 Closed to GL 89-08 B 87-002 Fastener Testing to Determine Cl TVA: letters dated April 15, 1988, July 6, 1988, Conformance with Applicable Material September 12, 1988, and January 27, 1989 Specifications 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.

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

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.

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

ITEM TITLE ADDITIONAL INFORMATION B 88-003 Inadequate Latch Engagement in HFA C TVA: letter dated April 13, 1992 Type Latching Relays Manufactured by General Electric (GE) Company NRC: IR 390/391 92-13 Potential Safety-Related Pump Loss B 88-004 Cl NRC acceptance letter dated May 24, 1990, for both units.

Unit 2 Action: Perform calculations and install check valves to prevent pump to pump interaction.

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

Piping Supplies, Inc. and West Jersey Manufacturing Company Unit 2 Action: 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 Transfer of NA Does not apply to power reactor.

Model No. SPEC 2-T Radiographic Exposure Device B 88-007 Power Oscillations in BWRs NA Boiling Water Reactor B 88-008 Thermal Stresses in Piping Connected to Cl NRC acceptance letter dated September 19, 1991, for both Reactor Cooling Systems units.

Unit 2 Action: Implement program to prevent thermal stratification.

B 88-009 Thimble Tube Thinning in Westinghouse CI Reviewed in Appendix EE of SSER16.

Reactors 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.

B 88-010 Nonconforming Molded-Case Circuit Cl Unit 2 Action: Replace those circuits not traceable to a Breakers circuit breaker manufacturer.

B 88-011 Pressurizer Surge Line Thermal Cl NRC SER on "Leak-Before-Break" (April 28, 1993) and Stratification reviewed in Appendix EE of SSER16.

Unit 2 Action: 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 Generator Cl NRC acceptance letter dated September 26, 1991 for both Tube Mechanical Plugs units.

Unit 2 Action: Remove SG tube plugs.

B 89-002 Stress Corrosion Cracking of Cl NRC reviewed in Appendix EE of SSER16.

High-Hardness Type 410 Stainless Steel Internal Preloaded Bolting in Anchor Unit 2 Actions: Replace the flapper assembly hold-down Darling Model S350W Swing Check Valves bolts fabricated on the 14 (12 valves are installed) Atwood or Valves of Similar Nature 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.

Page 12 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION B 89-003 Potential Loss of Required Shutdown CI TVA: letter dated June 19, 1990 Margin During Refueling Operations NRC: IR 390/391 94-04 and letter dated June 22, 1990 NRC acceptance letter dated June 22, 1990.

Unit 2 Action: Ensure that requirements for fuel assembly configuration, fuel loading and training are included in Unit 2.

B 90-001 Loss of Fill-Oil in Transmitters CI Unit 2 Action: Implement applicable recommendations from Manufactured by Rosemount this Bulletin including identification of potentially defective transmitters and an enhanced surveillance program which monitors transmitters for loss of fill oil.

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

B 92-001 Failure of Thermo-Lag 330 Fire Barrier OV TVA configurations for Thermo-Lag 330-1 were reviewed in System to Maintain Cabling in Wide Cable SSER18 and accepted in NRC letter dated January 6, 1998 Trays and Small Conduits Free From Fire (includes a supplemental SE).

Damage 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.

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

Aging Theratronics Teletherapy Units B 92-003 Release of Patients After Brachytherapy NA Does not apply to power reactor.

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

Treatment with Remote Afterloading Devices B 93-002 Debris Plugging of Emergency Core NA Boiling Water Reactor Cooling Suction Strainers B 93-003 Resolution of Issues Related to Reactor NA Boiling Water Reactor Vessel Water Level Instrumentation in BWRs B 94-001 Potential Fuel Pool Draindown Caused by NA Addressed to holders of licenses for nuclear power reactors Inadequate Maintenance Practices at that are permanently shut down with spent fuel in the spent Dresden Unit 1 fuel pool B 94-002 Corrosion Problems in Certain Stainless NA Does not apply to power reactor.

Steel Packagings Used to Transport Uranium Hexafluoride B 95-001 Quality Assurance Program for NA Does not apply to power reactor.

Transportation of Radioactive Material B 95-002 Unexpected Clogging of a Residual Heat NA Boiling Water Reactor Removal Pump Strainer While Operating in Suppression Pool Cooling Mode Page 13 of 60 * = See last page for status code definition.

ITEM TITLE

• ADDITIONAL INFORMATION B 96-001 Control Rod Insertion Problems (PWR) OV NRC acceptance letter for Unit 1 dated July 22, 1996 -

Initial response for Unit 2 on September 7, 2007.

Unit 2 Action: Issue Emergency Operating Procedure and provide core map.

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

Fuel, Over Fuel in the Reactor, or Over Safety-Related Equipment Unit 2 Action: Unit 2 Heavy Loads Program will be in compliance with NUREG-0612.

B 96-003 Potential Plugging of ECCS Suction NA Boiling Water Reactor Strainers by Debris in BWRs B 96-004 Chemical, Galvanic, or Other Reactions in NA Info Spent Fuel Storage and Transportation Casks B 97-001 Potential for Erroneous Calibration, Dose NA Does not apply to power reactor.

Rate, or Radiation Exposure Measurements with Certain Victoreen Model 530 and 531SI Electrometer/Dosemeters B 97-002 Puncture Testing of Shipping Packages NA Does not apply to power reactor.

Under 10 CFR Part 71 B 01-001 Circumferential Cracking of Reactor OV NRC acceptance letter dated November 20, 2001 (Unit 1)

Pressure Vessel (RPV) Head Penetration - Initial response for Unit 2 on September 7, 2007.

Nozzles Unit 2 Action: Perform baseline inspection.

B 02-001 RPV Head Degradation and Reactor OV NRC review of Unit 1's 15 day response in letter dated Coolant Pressure Boundary Integrity May 20, 2002 - Initial response for Unit 2 on September 7, 2007.

Unit 2 Action: Perform baseline inspection.

B 02-002 RPV Head and Vessel Head Penetration OV NRC acceptance letter dated December 20, 2002 (Unit 1)

Nozzle Inspection Programs - Initial response for Unit 2 on September 7, 2007.

Unit 2 Action: Perform baseline inspection.

B 03-001 Potential Impact of Debris Blockage on NA *ITVA: letter dated September 7, 2007 Emergency Sump Recirculation at PWRs B 03-002 Leakage from RPV Lower Head OV NRC acceptance letter dated October 6, 2004 (Unit 1) -

Penetrations and Reactor Coolant Pressure Initial response for Unit 2 on September 7, 2007.

Boundary Integrity (PWRs)

Unit 2 Action: Perform baseline inspection.

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

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

Page 14 of 60 * = See last page for status code definition.

ITEM TITLE

• ADDITIONAL INFORMATION B 04-001 Inspection of Alloy 82/182/600 Materials OV Initial response for Unit 2 on September 7, 2007.

Used in the Fabrication of Pressurizer Penetrations and Steam Space Piping Unit 2 Actions: Provide details of pressurizer and Connections at PWRs penetrations and apply Material Stress Improvement Process.

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 Response C TVA: letters dated January 20, 2006 and August 16, 2006.

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

B 07-001 Security Officer Attentiveness OV Item concerns a multi-unit issue that was completed for both units.

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

C 76-002 Relay Failures - Westinghouse BF (AC) C TVA: letter dated November 22, 1976 informed NRC that and BFD (DC) Relays these relay types are not used in Class 1E circuits.

NRC: IR50/390/76-11 and 50/391/76-11 C 76-003 Radiation Exposures in Reactor Cavities NA Info C 76-004 Neutron Monitor and Flow Bypass Switch NA Boiling Water Reactor Malfunctions C 76-005 Hydraulic Shock And Sway Suppressors - C TVA: letter dated January 7, 1977 informed NRC that no Maintenance of Bleed and Lock-Up Grinnell shock suppressors or sway braces have been or Velocities on ITT Grinnell's Model Nos. - will be installed at WBN.

Fig. 200 And Fig. 201, Catalog Ph-74-R C 76-006 Stress Corrosion Cracks in Stagnant, Low NA Item was applicable only to units with operating license at Pressure Stainless Piping Containing Boric the time the item was issued.

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

C 77-001 Malfunctions of Limitorque Valve Operators NA Info C 77-002a Potential Heavy Spring Flooding (CP) NA Item was applicable only to units with operating license at the time the 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 inValve Bonnets NA Info C 77-006 Effects of Hydraulic Fluid on Electrical NA Info Cables C 77-007 Short Period During Reactor Startup NA Boiling Water Reactor Page 15 of 60 * = See last page for status code definition.

ITEM TITLE

• ADDITIONAL INFORMATION C 77-008 Failure of Feedwater Sample Probe NA Item was applicable only to units with operating license at the time the item was issued.

C 77-009 Improper Fuse Coordination in BWR NA Boiling Water Reactor Standby Liquid Control System Control Circuits C 77-010 Vacuum Conditions Resulting in Damage to NA Item was applicable only to units with operating license at Liquid Process Tanks the time the item was issued.

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C 77-011 Leakage of Containment Isolation Valves NA Info with Resilient Seats C 77-012 Dropped Fuel Assemblies at BWR Facilities NA Boiling Water Reactor C 77-013 Reactor Safety Signals Negated During NA Info 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 Electrical Trip NA Info Lock-Out Features C 78-001 Loss of Well Logging Source NA Does not apply to power reactor.

C 78-002 Proper Lubricating Oil for Terry Turbines NA Info C 78-003 Packaging Greater Than Type A Quantities NA Info of Low Specific Activity Radioactive Material for Transport C 78-004 Installation Errors That Could Prevent NA Info Closing of Fire Doors C 78-005 Inadvertent Safety Injection During NA Info Cooldown C 78-006 Potential Common Mode Flooding of ECCS NA Info Equipment Rooms at BWR Facilities

................................................................................................................................. i........

C 78-007 Damaged Components of a NA Info Bergen-Paterson Series 25000 Hydraulic Test Stand C 78-008 Environmental Qualification of NA Info Safety-Related Electrical Equipment at Nuclear Power Plants C 78-009 Arcing of General Electric Company NA Info Size 2 Contactors C 78-010 Control of Sealed Sources in Radiation NA Does not apply to power reactor.

Therapy C 78-011 Recirculation MG Set Overspeed Stops NA Boiling Water Reactor C 78-012 HPCI Turbine Control Valve Lift Rod NA Boiling Water Reactor Bending Page 16 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION C 78-013 Inoperability of Service Water Pumps NA Info C 78-014 HPCI Turbine Reversing Chamber Hold NA Boiling Water Reactor Down Bolting C 78-015 Tilting Disc Check Valves Fail to Close with NA Info Gravity in Vertical Position C 78-016 Limitorque Valve Actuators NA Info C 78-017 Inadequate Guard Training/Qualification NA Info and Falsified Training Records C 78-018 UL Fire Test NA Info

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C 78-019 Manual Override (Bypass) of Safety System NA Info Actuation Signals C 79-001 Administration of Unauthorized Byproduct NA Does not apply to power reactor.

Material to Humans C 79-002 Failure of 120 Volt Vital AC Power Supplies NA Info C 79-003 Inadequate Guard Training - Qualification NA Info and Falsified Training Records C 79-004 Loose Locking Nut on Limitorque Valve NA Info Operators C 79-005 Moisture Leakage in Stranded Wire NA Info Conductors C 79-006 Failure to Use Syringe and Bottle Shields in NA Does not apply to power reactor.

Nuclear Medicine C 79-007 Unexpected Speed Increase of Reactor NA Boiling Water Reactor Recirculation MG Set Resulted in Reactor Power Increase C 79-008 Attempted Extortion - Low Enriched NA Fuel facilities and operating reactors at the time the item Uranium was issued C 79-009 Occurrences of Split or Punctured NA Info Regulator Diaphragms in Certain Self Contained Breathing Apparatus C 79-010 Pipefittings Manufactured from NA Info Unacceptable Material

............. ................................................................................. L...........................................

C 79-011 Design/Construction Interface Problem NA Info C 79-012 Potential Diesel Generator Turbocharger NA Info Problem C 79-013 Replacement of Diesel Fire Pump Starting NA Info Contactors C 79-014 Unauthorized Procurement and Distribution NA Does not apply to power reactor.

of XE-133 Page 17 of 60 *=See last page for status code definition.

ITEM TITLE

• ADDITIONAL INFORMATION C 79-015 Bursting of High Pressure Hose and NA Item was applicable only to units with operating license at Malfunction of Relief Valve O-Ring in the time the item was issued.

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 Switches on NA Info 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 Ingersoll-Rand NA Info Pumps C 79-020 Failure of GTE Sylvania Relay Type PM NA Info Bulletin 7305 Catalog 5U12-11-AC with a 120V AC Coil C 79-021 Prevention of Unplanned Releases of NA Info Radioactivity C 79-022 Stroke Times for Power Operated Relief NA Info Valves C 79-023 Motor Starters and Contactors Failed to NA Info Operate C 79-024 Proper Installation and Calibration of Core NA Boiling Water Reactor Spray Pipe Break Detection Equipment on BWRs C 79-025 Shock Arrestor Strut Assembly Interference NA Info

_ _....m ..................................................................................................................................

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

Implant Therapy Sources C 80-007 Problems with HPCI Turbine Oil System NA Boiling Water Reactor C 80-008 BWR Technical Specification NA Boiling Water Reactor 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 18 of 60 * = See last page for status code definition.

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

for Medical Licensees C 80-020 Changes in Safe-Slab Tank Dimensions NA Info 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 Qualifications NA Info C 80-023 Potential Defects in Beloit Power Systems NA Info Emergency Generators C 80-024 AECL Teletherapy Unit Malfunction NA Does not apply to power reactor.

C 80-025 Case Histories of Radiography Events NA Does not apply to power reactor.

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

C 81-003 Inoperable Seismic Monitoring NA Info Instrumentation C 81-004 The Role of Shift Technical Advisors and NA Info Importance of Reporting Operational Events C 81-005 Self-Aligning Rod End Bushings for Pipe NA Info Supports C 81-006 Potential Deficiency Affecting Certain NA Info Foxboro 10 to 50 Milliampere Transmitters Page 19 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION C 81-007 Control of Radioactively Contaminated NA Info Material C 81-008 Foundation Materials NA Info C 81-009 Containment Effluent Water that Bypasses NA Info Radioactivity Monitor C 81-010 Steam Voiding in the Reactor Coolant NA Item was applicable only to units with operating license at System During Decay Heat Removal the time the item was issued.

Cooldown

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C 81-011 Inadequate Decay Heat Removal During NA Boiling Water Reactor Reactor Shutdown C 81-012 Inadequate Periodic Test Procedure of NA Info PWR Reactor Protection System C 81-013 Torque Switch Electrical Bypass Circuit for NA Info Safeguard Service Valve Motors C 81-014 Main Steam Isolation Valve Failures to NA Info Close C 81-015 Unnecessary Radiation Exposures to the NA Info Public and Workers During Events Involving Thickness and Level Measuring Devices GL 77-001 Intrusion Detection Systems Handbook NA Info GL 77-002 Fire Protection Functional Responsibilities NA Info GL 77-003 Transmittal of NUREG-0321, "A Study of NA Info the Nuclear Regulatory Commission Quality Assurance Program" GL 77-004 Shipments of Contaminated Components NA Info From NRC Licensed Power Facilities to Vendors & Service Companies GL 77-005 Nonconformity of Addressees of Items NA Info Directed to the Office of Nuclear Reactor Regulation GL 77-006 Enclosing Questionnaire Related to Steam NA Item was applicable only to units with operating license at Generators the time the item was issued.

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

GL 77-008 Revised Intrusion Detection Handbook and NA Info Entry Control Systems Handbook

................................................................................................................................ =.........

GL 78-001 Correction to Letter of 12/15/77 [GL 77-07] NA Item was applicable only to units with operating license at the time the item was issued.

GL 78-002 Asymmetric Loads Background and C NRC: Reviewed in SSER15 - Appendix C (June 1995).

Revised Request for Additional Information Resolved by approval of leak-before-break analysis.

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

Page 20 of 60 * = See last page for status code definition.

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

GL 78-005 Internal Distribution of Correspondence - NA Info 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 Mark I NA Boiling Water Reactor 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 Environmental NA Info Monitoring GL 78-011 Guidance on Spent Fuel Pool Modifications NA Info 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 for NA Info Comment GL 78-015 Request for Information on Control of NA See GL 81-007.

Heavy Loads Near Spent Fuel GL 78-016 Request for Information on Control of NA Info Heavy Loads Near Spent Fuel Pools GL 78-017 Corrected Letter on Heavy Loads Over NA Info Spent Fuel GL 78-018 Corrected Letter on Heavy Loads Over NA Duplicate of GL 81-007 Spent Fuel GL 78-019 Enclosing Sandia Report SAND 77-0777, NA Info "Barrier Technology Handbook" GL 78-020 Enclosing - "A Systematic Approach to the NA Info 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 Systems NA Info and Entry Control Systems Handbooks and Nuclear Safeguards Technology Handbook GL 78-023 Manpower Requirements for Operating NA Info Reactors Page 21 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 78-024 Model Appendix I Technical Specifications NA Boiling Water Reactor and Submittal Schedule For BWRs GL 78-025 This GL was never issued. NA GL 78-026 Excessive Control Rod Guide Tube Wear NA Applies only to Babcock and Wilcox designed plants GL 78-027 Forwarding of NUREG-0181 NA Info GL 78-028 Forwarding pages omitted from 07/11/78 NA Boiling Water Reactor letter [GL 78-24]

GL 78-029 Notice of PWR Steam Generator NA Info Conference GL 78-030 Forwarding of NUREG-0219 NA Info GL 78-031 Notice of Steam Generator Conference NA Info Agenda GL 78-032 Reactor Protection System Power Supplies NA Boiling Water Reactor GL 78-033 Meeting Schedule and Locations For NA Info Upgraded Guard Qualification GL 78-034 Reactor Vessel Atypical Weld Material C See B 78-12.

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

Except In NRC Approved Containers GL 78-037 Revised Meeting Schedule & Locations For NA Info Upgraded Guard Qualifications

.......................................................... °.,.............................................................................

GL 78-038 Forwarding of 2 Tables of Appendix I, Draft NA Item was applicable only to units with operating license at Radiological Effluent Technical the time the item was issued.

Specifications, PWR, and NUREG-0133 GL 78-039 Forwarding of 2 Tables of Appendix I, Draft NA Boiling Water Reactor Radiological Effluent Technical Specifications, BWR, and NUREG-0133 GL 78-040 Training & Qualification Program NA Info Workshops GL 78-041 Mark II Generic Acceptance Criteria For NA Boiling Water Reactor Lead Plants GL 78-042 Training and Qualification Program NA Info Workshops GL 79-001 Interservice Procedures for Instructional NA Info Systems Development - TRADOC GL 79-002 Transmitting Rev. to Entry Control Systems NA Info Handbook (SAND 77-1033), Intrusion Detection Handbook (SAND 76-0554), and Barrier Penetration Database GL 79-003 Offsite Dose Calculation Manual NA Info Page 22 of 60 * = See last page for status code definition.

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

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

Comment GL 79-011 Transmitting "Summary of Operating NA Info Experience with Recalculating Steam Generators, January 1979," NUREG-0523 GL 79-012 ATWS - Enclosing Letter to GE, with NA Info 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 - Enclosing NA Info NUREG-0531 and Notice GL 79-015 Steam Generators - Enclosing Summary NA Info of Operating Experience with Recirculating Steam Generators, NUREG-0523 GL 79-016 Meeting Re Implementation of Physical NA Info Security Requirements GL 79-017 Reliability of Onsite Diesel Generators at NA Info Light Water Reactors GL 79-018 Westinghouse Two-Loop NSSS NA Addressed to specific plant(s).

GL 79-019 NRC Staff Review of Responses to NA Addressed to specific plant(s).

Bs 79-06 and 79-06a GL 79-020 Cracking in Feedwater Lines C See B 79-13.

GL 79-021 Enclosing NUREG/CR-0660, Enhancement NA Info of on Site Emergency Diesel Generator Reliability" GL 79-022 Enclosing NUREG-0560, "Staff Report on NA Applies only to Babcock and Wilcox designed plants the Generic Assessment of Feedwater Transients in PWRs Designed by B&W" Page 23 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 79-023 NRC Staff Review of Responses to B 79-08 NA Boiling Water Reactor GL 79-024 Multiple Equipment Failures in NA Item was applicable only to units with operating license at Safety-Related Systems the time the item was issued.

GL 79-025 Information Required to Review Corporate NA Info Capabilities GL 79-026 Upgraded Standard Technical Specification NA Info Bases Program GL 79-027 Operability Testing of Relief and Safety NA Boiling Water Reactor Relief Valves GL 79-028 Evaluation of Semi-Scale Small Break NA Info Experiment GL 79-029 Transmitting NUREG-0473, Revision 2, NA Info Draft Radiological Effluent Technical Specifications GL 79-030 Transmitting NUREG-0472, Revision 2, NA Info Draft Radiological Technical Specifications GL 79-031 Submittal of Copies of Response to 6/29/79 NA Info NRC Request [79-25]

GL 79-032 Transmitting NUREG-0578, "TMI-2 Lessons NA Info Learned" GL 79-033 Transmitting NUREG-0576, "Security NA Info 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 Impacts on NA Info Emergency Planning GL 79-036 Adequacy of Station Electric Distribution Cl This GL tracked compliance with BTP PSB-1, "Adequacy of Systems Voltages Station Electric Distribution System Voltages."

Unit 2 Action: Perform verification during the preoperational testing.

GL 79-037 Amendment to 10 CFR 73.55 Deferral from NA Info 8/1/79 to 11/1/79 GL 79-038 BWR Off-Gas Systems - Enclosing NA Boiling Water Reactor NUREG/CR-0727

........................................... e, ..........................................................................................

GL 79-039 Transmitting Division 5 Draft Regulatory NA Does not apply to power reactor.

Guide and Value Impact Statement GL 79-040 Follow-up Actions Resulting from the NRC NA Item was applicable only to units with operating license at Staff Reviews Regarding the TMI-2 Accident the time the item was issued.

GL 79-041 Compliance with 40 CFR 190, EPA NA Info Uranium Fuel Cycle Standard GL 79-042 Potentially Unreviewed Safety Question on NA Item was applicable only to units with operating license at Interaction Between Non-Safety Grade the time the item was issued.

Systems and Safety Grade Systems Page 24 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 79-043 Reactor Cavity Seal Ring Generic Issue NA Addressed to specific plant(s).

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

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

Operability GL 79-047 Radiation Training NA Info GL 79-048 Confirmatory Requirements Relating to NA Boiling Water Reactor Condensation Oscillation Loads for the Mark I Containment Long Term Program GL 79-049 Summary of Meetings Held on 9/18-20/79 NA Info to Discuss Potential Unreviewed Safety Question on Systems Interaction for B&W PI GL 79-050 Emergency Plans Submittal Dates NA Info GL 79-051 Follow-up Actions Resulting from the NRC , NA Info Staff Reviews Regarding the TMI-2 Accident GL 79-052 Radioactive Release at North Anna Unit 1 NA Item was applicable only to units with operating license at and Lessons Learned the time the item was issued.

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

Normal Operation 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 GL 79-056 Discussion of Lessons Learned Short Term NA Item was applicable only to units with operating license at Requirements the time the item was issued.

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

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

Duplicate of GL 79-058 GL 79-063 Upgraded Emergency Plans NA Info Page 25 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 79-064 Suspension of All Operating Licenses NA Info (PWRs)

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

GL 79-067 Estimates for Evacuation of VariousAreas NA Info Around Nuclear Power Reactors GL 79-068 Audit of Small Break LOCA Guidelines NA Info GL 79-069 Cladding Rupture, Swelling, and Coolant NA Info Blockage as a Result of a Reactor Accident GL 79-070 Environmental Monitoring for Direct NA Info Radiation GL 80-001 NUREG-0630, "Cladding, Swelling and NA Info Rupture - Models For LOCA Analysis" GL 80-002 QA Requirements Regarding Diesel C TVA: FSAR 9.5.4.2 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 Qualification of NA Info Class 1E 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 Assurance for NA Boiling Water Reactor Nuclear Supplied Equipment" GL 80-009 Low Level Radioactive Waste Disposal NA Item was applicable only to units with operating license at the time the item was issued.

GL 80-010 Issuance of NUREG-0588, "Interim Staff NA Info Position On Equipment Qualifications of Safety-Related Electrical Equipment" GL 80-011 B 80-03, "Loss of Charcoal From Standard NA Info Type II,2 Inch, Tray Absorber Cells" GL 80-012 B 80-04, "Analysis of a PWR Main Steam NA Info Line Break With Continued Feedwater Addition" GL 80-013 Qualification of Safety Related Electrical NA Item was applicable only to units with operating license at Equipment the time the item was issued.

Page 26 of 60 * = See last page for status code definition.

ITEM TITLE

• ADDITIONAL INFORMATION GL 80-014 LWR Primary Coolant System Pressure CT TVA: FSAR 5.2.7.4 Isolation Valves NRC: 1.14.2 of SSER 6 NRC reviewed in 1.14.2 of SSER6.

Unit 2 Action: Incorporate guidance into Technical Specifications.

GL 80-015 Request for Additional Management and NA Info Technical Resources Information GL 80-016 B 79-01b, "Environmental Qualification of NA Info Class 1E Equipment" GL 80-017 Modifications to BWR Control Rod Drive NA Boiling Water Reactor Systems GL 80-018 Crystal River 3 Reactor Trip From NA Applies only-to Babcock and Wilcox designed plants Approximately 100% Full Power GL 80-019 Resolution of Enhanced Fission Gas NA Info Release Concern GL 80-020. Actions Required From OL Applicants of NA Info NSSS Designs by W and CE Resulting From NRC B&O Task Force Review of TMI2 Accident GL 80-021 B 80-05, "Vacuum Condition Resulting in Cl 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.

.......................................... m.........................................................................*............ 0.........

GL 80-022 Transmittal of NUREG-0654, "Criteria For NA Info Preparation and Evaluation of Radiological Emergency Response Plan" GL 80-023 Change of Submittal Date For Evaluation NA Info Time Estimates GL 80-024 Transmittal of Information on NRC "Nuclear NA. Info Data Link Specifications" GL 80-025 B 80-06, "Engineering Safety Feature (ESF) NA Info Reset Controls" GL 80-026 Qualifications of Reactor Operators NA Info GL 80-027 B 80-07, "BWR Jet Pump Assembly Failure" NA Boiling Water Reactor GL 80-028 B 80-08, "Examination of Containment NA Info Liner Penetration Welds" GL 80-029 Modifications to Boiling Water Reactor NA Boiling Water Reactor Control Rod Drive Systems GL 80-030 Clarification of The Term "Operable" As It NA Item was applicable only to units with operating license at Applies to Single Failure Criterion For the time the item was issued.

Safety Systems Required by TS 8

GL 80-031 B 80-09, "Hydramotor Actuator Deficiencies" NA Info Page 27 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 80-032 Information Request on Category I Masonry NA Addressed by B 80-11.

Walls Employed by Plants Under CP and OL Review GL 80-033 Actions Required From OL Applicants of NA Applies only to Babcock and Wilcox designed plants B&W Designed NSSS Resulting From NRC B&O Task Force Review of TMI2 Accident GL 80-034 Clarification of NRC Requirements for NA Info Emergency Response Facilities at Each Site GL 80-035 Effect of a DC Power Supply Failure on NA Boiling Water Reactor 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 Requirements to Operating Reactors the time the item was issued.

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

Physical Protection Requirements GL 80-039 B 80-11, "Masonry Wall Design" NA Info GL 80-040 Transmittal of NUREG-0654, "Report of the NA Info 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 System NA Info Operability" GL 80-043 B 80-13, "Cracking In Core Spray Spargers" NA Boiling Water Reactor 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.

GL 80-046/47 Generic Technical Activity A-12, "Fracture C No response was required for this GL, and NUREG-0577 Toughness and Additional Guidance on states that the lamellar tearing aspect of this issue was Potential for Low Fracture toughness and resolved by the NUREG. Further, the NUREG states that Laminar Tearing on PWR Steam Generator for plants under review, the fracture toughness issue was Coolant Pump Supports" resolved.

GL 80-048 Revision to 5/19/80 Letter On Fire NA Item was applicable only to units with operating license at Protection [GL 80-45] the time the item was issued.

GL 80-049 Nuclear Safeguards Problems NA Info GL 80-050 Generic Activity A-1 0, "BWR Cracks" NA Boiling Water Reactor GL 80-051 On-Site Storage of Low-Level Waste NA Item was applicable only to units with operating license at the time the item was issued.

........................................................................................................................... i..............

Page 28 of 60 P 2=6See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 80-052 Five Additional TMI-2 Related NA Item was applicable only to units with operating license at Requirements - Erata Sheets to 5/7/80 the time the item was issued.

Letter [GL 80-37]

.............................................................. -------------............. .......................... ...---...............- o---

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

GL 80-054 B 80-14, "Degradation of Scram Discharge NA Boiling Water Reactor Volume Capability" GL 80-055 B 80-15, "Possible Loss of Hotline With NA Info Loss of off-Site Power" GL 80-056 Commission Memorandum and Order on NA Info Equipment Qualification GL 80-057 Further Commission Guidance For Power NA Info Reactor Operating Licenses NUREG-0660 and NUREG-0694 GL 80-058 B 80-16, "Potential Misapplication of NA Info Rosemount Inc. Models 1151/1152 Pressure Transmitters With "A" Or "D" Output Codes" GL 80-059 Transmittal of Federal Register Notice RE NA Info 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 to Insert NA Boiling Water Reactor During a Scram at a BWR" GL 80-064 Scram Discharge Volume Designs NA Boiling Water Reactor GL 80-065 Request for Estimated Construction NA Info Completion and Fuel Load Schedules GL 80-066 B 80-17, Supplement 1, "Failure of Control NA Boiling Water Reactor Rods to Insert During a Scram at a BWR" GL 80-067 Scram Discharge Volume NA Boiling Water Reactor GL 80-068 B 80-17, Supplement 2, "Failures Revealed NA Boiling Water Reactor by Testing Subsequent to Failure of Control Rods to Insert During a Scram at a BWR" GL 80-069 B 80-18, "Maintenance of Adequate NA Info Minimum Flow Through Centrifugal Charging Pumps Following Secondary Side HELB" GL 80-070 B 80-19, "Failures of Mercury-Wetted Matrix NA Info Relays in RPS of Operating Nuclear Power Plants Designed by GE" Page 29 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 80-071 B 80-20, "Failures of Westinghouse Type NA Info W-2 Spring Return to Neutral Control Switches" GL 80-072 Interim Criteria For Shift Staffing NA Info GL 80-073 "Functional Criteria For Emergency NA Info Response Facilities," NUREG-0696 GL 80-074 Notice of Forthcoming Meeting With NA Info 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 With GE to NA Info Discussed Proposed BWR Feedwater Nozzle Leakage Detection System

............... J..........................................................................................................................

GL 80-077 Refueling Water Level - Technical CT Unit 2 Action: Address in Technical Specifications, as Specifications Changes appropriate.

GL 80-078 Mark I Containment Long-Term Program NA Boiling Water Reactor

................................................................................................... m......................................

GL 80-079 B 80-17, Supplement 3, "Failures Revealed NA Boiling Water Reactor by Testing Subsequent to Failure of Control Rods to Insert During a Scram At a BWR" GL 80-080 Preliminary Clarification of TMI Action Plan NA Info Requirements GL 80-081 Preliminary Clarification of TMI Action Plan NA Info Requirements - Addendum to 9/5/80 Letter [GL 80-80]

GL 80-082 B 79-01b, Supplement 2, "Environmental NA Info Qualification of Class 1 E Equipment" 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" GL 80-086 Notice of Meeting to Discuss Final NA Info Resolution of USI A-12

................................................... m......................................................................................

GL 80-087 Notice of Meeting to Discuss Status of NA Info EPRI-Proposed Resolution of the USI A-12 Fracture Toughness Issue GL 80-088 Seismic Qualification of Auxiliary Feedwater NA Item was applicable only to units with operating license at Systems the time the item was issued.

GL 80-089 B 79-01b, Supplement 3, "Environmental NA Info Qualification of Class 1 E Equipment" Page 30 of 60 * = See last page for status code definition.

ITEM TITLE

• ADDITIONAL INFORMATION GL 80-090 NUREG-0737, TMI (Prior and future GLs, Cl See NUREG items in this list.

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 Malcolm NA Info Foundry Company, Inc."

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-1 0, NA Boiling Water Reactor NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking"

............................................................. w............................................................................

GL 80-096 Fire Protection NA Addressed to specific plant(s).

GL 80-097 B 80-23, "Failures of Solenoid Valves NA Info Manufactured by Valcor Engineering Corporation" GL 80-098 B 80-24, "Prevention of Damage Due to NA Info Water Leakage Inside Containment" GL 80-099 Technical Specifications Revisions For NA Info Snubber Surveillance GL 80-100 Appendix R to 10 CFR 50 Regarding Fire NA Item was applicable only to units with operating license at Protection - Federal Register Notice the time the item was issued.

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

.................................................................. m.......................................................................

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

GL 80-103 Fire Protection - Revised Federal Register NA Info Notice GL 80-104 Orders On Environmental Qualification of NA Info Safety Related Electrical Equipment GL 80-105 Implementation of Guidance For USI A-12, NA Info

""Potential For Low Fracture toughness and Lamellar Tearing On Component Supports" GL 80-106 Report On ECCS Cladding Models, NA Info NUREG-0630 GL 80-107 BWR Scram Discharge System NA Boiling Water Reactor GL 80-108 Emergency Planning NA Info GL 80-109 Guidelines For SEP Soil Structure NA Info Interaction Reviews GL 80-110 Periodic Updating of FSARS NA Item was applicable only to units with operating license at the time the item was issued.

Page 31 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 80-111 B 80-17, Supplement 4, "Failure of Control NA Boiling Water Reactor Rods to Insert During a Scram at a BWR" GL 80-112 B 80-25, "Operating Problems With Target NA Info Rock Safety Relief Valves" GL 80-113 Control of Heavy Loads C Superseded by GL 81-007.

GL 81-001 Qualification of Inspection, Examination, NA Info 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 Training for C Superseded by Station Blackout Rule.

Station Blackout Events GL 81-005 Information Regarding The Program For NA Info Environmental Qualification of Safety-Related Electrical Equipment GL 81-006 Periodic Updating of Final Safety Analysis NA Info 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.

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 Control Rod NA Boiling Water Reactor 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 8X8R Fuel NA Boiling Water Reactor Reload Applications For Appendix D Submittals of The GE topical Report Page 32 of 60 * = See last page for status code definition.

ITEM TITLE

• ADDITIONAL INFORMATION GL 81-014 Seismic Qualification of Auxiliary Feedwater Cl TVA: FSAR 10.4.9 Systems Unit 2 Action: Additional Unit 2 implementing procedures or other activity is required for completion.

GL 81-015 Environmental Qualification of Class 1E NA Info Electrical Equipment - Clarification of Staffs Handling of Proprietary Information GL 81-016 NUREG-0737, Item I.C.1 SER on Abnormal NA Applies only to Babcock and Wilcox designed plants 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 Pressure Vessels NA Item was applicable only to units with operating license at the time the item was issued.

GL 81-020 Safety Concerns Associated With Pipe NA Boiling Water Reactor Breaks in the BWR Scram System 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 Reports NA Info GL 81-024 Multi-Plant Issue B-56, "Control Rods Fail NA Boiling Water Reactor to Fully Insert"

...................... *............................................................................................ 7,----------------------.

GL 81-025 Change in Implementing Schedule For NA Info Submission and Evaluation of Upgraded.

Emergency Plans GL 81-026 Licensing Requirements for Pending NA Applicants with pending Construction Permits Construction Permit and Manufacturing License Applications GL 81-027 Privacy and Proprietary Material in NA Info Emergency Plans GL 81-028 Steam Generator Overfill NA Info GL 81-029 Simulator Examinations NA Info GL 81-030 Safety Concerns Associated With Pipe NA Boiling Water Reactor Breaks in the BWR Scram System GL 81-031 This GL was never issued. NA GL 81-032 NUREG-0737, Item I1.K.3.44, "Evaluation of NA Boiling Water Reactor Anticipated Transients Combined With Single Failure" Page 33 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 81-033 This GL was never issued. NA GL 81-034 Safety Concerns Associated With Pipe NA Boiling Water Reactor Breaks in the BWR Scram System GL 81-035 Safety Concerns Associated With Pipe NA Boiling Water Reactor Breaks in the BWR Scram System GL 81-036 Revised Schedule for Completion of TMI NA Info Action Plan Item Il.D.1, "Relief and Safety Valve Testing" GL 81-037 ODYN Code Reanalysis Requirements NA Boiling Water Reactor GL 81-038 Storage of Low Level Radioactive Wastes NA Info 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 GL 82-005 Post-TMI Requirements NA Item was applicable only to units with operating license at the time the item was issued.

GL 82-006 This GL was never issued. NA GL 82-007 Transmittal of NUREG-0909 Relative to the NA Boiling Water Reactor Ginna Tube Rupture GL 82-008 Transmittal of NUREG-0909 Relative to the NA Info Ginna Tube Rupture GL 82-009 Environmental Qualification of Safety NA Info Related Electrical Equipment 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 Relative to the NA Info Restart of R. E. Ginna Nuclear Power Plant GL 82-012 Nuclear Power Plant Staff Working Hours NA Info GL 82-013 Reactor Operator and Senior Reactor NA Info Operator Examinations GL 82-014 Submittal of Documents to the NRC NA Info GL 82-015 This GL was never issued. NA GL 82-016 NUREG-0737 Technical Specifications NA Item was applicable only to units with operating license at the time the item was issued.

Page 34 of 60 * = See last page for status code definition.

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

GL 82-023 Inconsistency Between Requirements of NA Info 10CFR 73.40(d) and Standard Technical Specifications For Performing Audits of Safeguards Contingency Plans GL 82-024 Safety Relief Valve Quencher Loads: BWR NA Boiling Water Reactor MARK II and III Containments GL 82-025 Integrated IAEA Exercise for Physical NA Item was applicable only to units with operating license at Inventory at LWRS the time the item was issued.

......................................................................................... =..............................................

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

GL 82-027 Transmittal of NUREG-0763, "Guidelines NA Boiling Water Reactor For Confirmatory In-Plant Tests of Safety-Relief Valve Discharge for BWR Plants" GL 82-028 Inadequate Core Cooling Instrumentation 0 LICENSE CONDITION - Detectors for Inadequate core System 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 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 Q PAM system.

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

Page 35 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 82-033 Supplement to NUREG-0737, CI "Safety Parameter Display System" (SPDS) /

"Requirements for Emergency Response "Requirements for Emergency Response Capability" -

Capability" NRC reviewed in SSER5, SSER6, and 18.2.2 of SSER15.

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

GL 82-034 This GL was never issued. 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 for NA Info Operator Licensing Examinations GL 82-039 Problems With Submittals of Subsequent NA Info Information of CURT 73.21 For Licensing Reviews GL 83-001 Operator Licensing Examination Site Visit NA Info GL 83-002 NUREG-0737 Technical Specifications NA Boiling Water Reactor 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 For NA Info Reactor Operator and Senior Reactor Operator Licenses GL 83-007 The Nuclear Waste Policy Act of 1982 NA Info

........................................................................................... *.......................u.......................

GL 83-008 Modification of Vacuum Breakers on Mark I NA Boiling Water Reactor Containments GL 83-009 Review of Combustion Engineering NA Applies only to Combustion Engineering designed plants Owners' Group Emergency Procedures Guideline Program GL 83-01Oa Resolution of TMI Action Item I1.K.3.5., NA Applies only to Combustion Engineering designed plants "Automatic Trip of Reactor Coolant Pumps"

......................................................... 7 .. *- . --. .- .- ..- . . . . .. . . . . -- .. . . . . .. . . . --. .. - .- .- .- . .. . .- . . .

GL 83-010b Resolution of TMI Action Item I1.K.3.5., NA Applies only to Combustion Engineering designed plants "Automatic Trip of Reactor Coolant Pumps" Page 36 of 60 * = See last page for status code definition.

ITEM TITLE

• ADDITIONAL INFORMATION GL 83-010c Resolution of TMI Action Item Il.K.3.5., Cl TVA: letters dated January 5, 1984 and June 25, 1984 "Automatic Trip of Reactor Coolant Pumps" NRC: letter dated June 8, 1990.

Unit 2 Action: Incorporate emergency response guidelines into applicable procedures.

GL 83-010d Resolution of TMI Action Item Il.K.3.5., NA Item was applicable only to units with operating license at "Automatic Trip of Reactor Coolant Pumps" the time the item was issued.

GL 83-010e Resolution of TMI Action Item I1.K.3.5., NA Applies only to Babcock and Wilcox designed plants "Automatic Trip of Reactor Coolant Pumps" GL 83-01Of Resolution of TMI Action Item II.K.3.5., NA Applies only to Babcock and Wilcox designed plants "Automatic Trip of Reactor Coolant Pumps" GL 83-011 Licensee Qualification for Performing NA Item was applicable only to units with operating license at Safety Analyses in Support of Licensing the time the item was issued.

Actions GL 83-012 Issuance of NRC FORM 398 - Personal NA Info Qualifications Statement - Licensee GL 83-013 Clarification of Surveillance Requirements NA Info for HEPA Filters and Charcoal Absorber Units In Standard Technical Specifications on ESF Cleanup Systems GL 83-014 Definition of "Key Maintenance Personnel," NA Info (Clarification of Generic Letter 82-12)

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

Operating Licenses GL 83-020 IntegratedScheduling for Implementation of NA Info Plant Modifications GL 83-021 Clarification of Access Control Procedures NA Info for Law Enforcement Visits Page 37 of 60 * = See last page for status code definition.

ITEM TITLE

• ADDITIONAL INFORMATION GL 83-022 Safety Evaluation of "Emergency Response NA Info 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 1.G.1, NA Boiling Water Reactor "Special Low Power Testing and Training,"

Recommendations for BWRs

..................... *......................................................................................................n..............

GL 83-025 This GL was never issued. NA GL 83-026 Clarification Of Surveillance Requirements NA Info For Diesel Fuel Impurity Level Tests GL 83-027 Surveillance Intervals in Standard NA Info Technical Specifications GL 83-028 "Required Actions Based on Generic C TVA: letters dated November 7, 1983 and Implications of Salem ATWS Events: December 4, 1987 1.2 - Post Trip Review Data and NRC: IR 50-390, 391/86-04 Information Capability GL 83-028 "Required Actions Based on Generic Cl TVA: letters dated November 7, 1983 and August 24, 1990 Implications of Salem ATWS Events:

NRC: letters dated October 20, 1986 and June 18, 1990 2.1 - Equipment Classification and Vendor Interface (Reactor Trip --

System Components)

Unit 2 Action: Ensure that required information on Critical Structures and Components is properly incorporated into procedures.

GL 83-028 "Required Actions Based on Generic Cl Unit 2 Action: Enter engineering component background Implications of Salem ATWS Events: data in INPO's Equipment Performance and Information Exchange System (EPIX) for Unit 2.

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

GL 83-028 "Required Actions Based on Generic CT TVA: letters dated November 7, 1983, January 17, 1986 Implications of Salem ATWS Events: and November 1, 1993 3.1 - Post-Maintenance Testing NRC: letters dated December 10, 1985, October 27, 1986, (Reactor Trip System Components) and July 2, 1990; IR 390, 391/86-04 Unit 2 Action: Test and maintenance procedures and Technical Specifications will include post-maintenance operability testing of safety-related components of the reactor trip system.

Page 38 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 83-028 "Required Actions Based on Generic CT TVA: letters dated November 7, 1983, January 17, 1986 Implications of Salem ATWS Events: and November 1, 1993 3.2- Post-Maintenance Testing (All SR NRC: letters dated December 10, 1985, October 27, 1986, Components) and July 2, 1990; IR 390, 391/86-04 Unit 2 Action: Test and maintenance procedures and Technical Specifications will include post-maintenance operability testing of other (than reactor trip system) safety-related components.

GL 83-028 "Required Actions Based on Generic CI TVA: letter dated May 19, 1986 Implications of Salem ATWS Events:

4.1 - Reactor Trip System Reliability (Vendor Related Modifications) Unit 2 Action: Confirm vendor-recommended DS416 breaker modifications are implemented.

GL 83-028 "Required Actions Based on Generic CT TVA: letters dated November 7, 1983, February 10, 1986, Implications of Salem ATWS Events: and May 19, 1986 4.2 - Reactor Trip System Reliability NRC: letters dated July 26, 1985 and June 18, 1992; (Preventive Maintenance and SSER 16 Surveillance Program for Reactor Trip Breakers)

Unit 2Action: Ensure maintenance instruction procedure and Technical Specifications support reliable reactor trip breaker operation.

GL 83-028 "Required Actions Based on Generic C TVA: letters dated November 7, 1983, March 22, 1985 Implications of Salem ATWS Events:

NRC: IR 50-390/86-04 and 50-391/86-04; letter dated 4.3 - Reactor Trip System Reliability June 18, 1990 (Automatic Actuation of Shunt Trip Attachment)

GL 83-028 "Required Actions Based on Generic CT TVA: letters dated November 7, 1983 and July 26, 1985 Implications of Salem ATWS Events:

NRC: letters dated June 28, 1990 and October 9, 1990; 4.5 - Reactor Trip System Reliability SSERs 5 and 16 (Automatic Actuation of Shunt Trip Attachment)

Unit 2 Action: Address in Technical Specifications, as appropriate.

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 Transient NA Applies only to Babcock and Wilcox designed plants Operating Guidelines"

................................................................................................................................... e_._

GL 83-032 NRC Staff Recommendations Regarding NA Info Operator Action for Reactor Trip and ATWS Page 39 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 83-033 NRC Positions on Certain Requirements of NA Info Appendix R to 10 CFR 50 GL 83-034 This GL was never issued. NA GL 83-035 Clarification of TMI Action Plan Item NA Info II.K.3.31 GL 83-036 NUREG-0737 Technical Specifications NA Boiling Water Reactor

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GL 83-037 NUREG-0737 Technical Specifications NA Item was applicable only to units with operating license at the time the item was issued.

............................... a .........................................................................................................

GL 83-038 NUREG-0965, "NRC Inventory of Dams" NA Info GL 83-039 Voluntary Survey of Licensed Operators NA Info GL 83-040 Operator Licensing Examination NA Info GL 83-041 Fast Cold Starts of Diesel Generators NA Item was applicable only to units with operating license at the time the item was issued.

............................................................. ................................................ .°. ..........................

GL 83-042 Clarification to GL 81-07 Regarding NA Info Response to NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants" GL 83-043 Reporting Requirements of 10 CFR 50, NA Info Sections 50.72 and 50.73, and Standard Technical Specifications GL 83-044 Availability of NUREG-1021, "Operator NA Info Licensing Examiner Standards" GL 84-001 NRC Use Of The Terms "Important To NA Info Safety" and "Safety Related" GL 84-002 Notice of Meeting Regarding Facility NA Info Staffing GL 84-003 Availability of NUREG-0933, "A NA Info Prioritization of Generic Safety Issues" GL 84-004 Safety Evaluation of Westinghouse Topical NA Info Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops GL 84-005 Change to NUREG-1021, "Operator NA Info Licensing Examiner Standards" GL 84-006 Operator and Senior Operator License NA Does not apply to power reactor.

Examination Criteria For Passing Grade GL 84-007 Procedural Guidance for Pipe Replacement NA Boiling Water Reactor at BWRs GL 84-008 Interim Procedures for NRC Management NA Info of Plant-Specific Backfitting GL 84-009 Recombiner Capability Requirements of 10 NA Boiling Water Reactor CFR 50.44(c)(3)(ii)

Page 40 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 84-010 Administration of Operating Tests Prior to NA Info Initial Criticality

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

GL 84-013 Technical Specification for Snubbers NA Info GL 84-014 Replacement and Requalification Training NA Info Program GL 84-015 Proposed Staff Actions to Improve and NA Info Maintain Diesel Generator Reliability GL 84-016 Adequacy of On-Shift Operating Experience NA Info for Near Term Operating License Applicants GL 84-017 Annual Meeting to Discuss Recent NA Info Developments Regarding Operator Training, Qualifications, and Examinations GL 84-018 Filing of Applications for Licenses and NA Does not apply to power reactor.

Amendments GL 84-019 Availability of Supplement 1 to NA Info NUREG-0933, "A Prioritization of Generic Safety Issues" GL 84-020 Scheduling Guidance for Licensee NA Info Submittals of Reloads That Involve Unreviewed Safety Questions GL 84-021 Long Term Low Power Operation in NA Info 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 Committee NA Only issued as draft Report GL 85-002 Recommended Actions Stemming From Cl TVA responded to the GL on June 17, 1985.

NRC Integrated Program for the Resolution of Unresolved Safety Issues Regarding Unit 2 Action: Perform SG inspection.

Steam Generator Tube Integrity GL 85-003 Clarification of Equivalent Control Capacity NA Boiling Water Reactor for Standby Liquid Control Systems GL 85-004 Operating Licensing Examinations NA Info Page 41 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 85-005 Inadvertent Boron Dilution Events NA Item was applicable only to units with operating license at the time the item was issued.

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

GL 85-008 10 CFR 20.408 Termination Reports - NA Info Format GL 85-009 Technical Specifications For Generic Letter NA Info 83-28, Item 4.3 GL 85-010 Technical Specification For Generic Letter NA Applies only to Babcock and Wilcox designed plants 83-28, Items 4.3 and 4.4 GL 85-011 Completion of Phase IIof "Control of Heavy C See GL 81-07.

Loads at Nuclear Power Plants,"

NUREG-0612 GL 85-012 Implementation Of TMI Action Item I1.K.3.5, Cl "Implementation of TMI Item I1.K.3.5" - Reviewed in 15.5.4 "Automatic Trip Of Reactor Coolant Pumps" of original 1982 SER; became License Condition 35. The staff determined that their review of Item II.K.3.5 did not have to be completed to support the full power license and considered this license condition resolved in SSER4. The item was further reviewed in Appendix EE of SSER16.

Unit 2 Action: Implement modifications as required.

GL 85-013 Transmittal Of NUREG-1154 Regarding NA Info The Davis-Besse Loss Of Main And Auxiliary Feedwater Event GL 85-014 Commercial Storage At Power Reactor NA Item was applicable only to units with operating license at Sites Of Low Level Radioactive Waste Not the time the item was issued.

Generated By The Utility GL 85-015 Information On Deadlines For NA Item was applicable only to units with operating license at 10 CFR 50.49, "Environmental Qualification the time the item was issued.

Of Electric Equipment Important To Safety At Nuclear Power Plants" GL 85-016 High Boron Concentrations NA Info GL 85-017 Availability Of Supplements 2 and 3 To NA Info NUREG-0933, "A Prioritization Of Generic Safety Issues" GL 85-018 Operator Licensing Examinations NA Info GL 85-019 Reporting Requirements On Primary NA Info Coolant Iodine Spikes GL 85-020 Resolution Of Generic Issue 69: High NA Applies only to Babcock and Wilcox designed plants Pressure Injection/Make-up Nozzle Cracking In Babcock And Wilcox Plants GL 85-021 This GL was never issued. NA Page 42 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 85-022 Potential For Loss Of Post-LOCA NA Info Recirculation Capability Due To Insulation Debris Blockage GL 86-001 Safety Concerns Associated With Pipe NA Boiling Water Reactor Breaks In The BWR Scram System GL 86-002 Technical Resolution of Generic Issue NA Boiling Water Reactor B Thermal Hydraulic Stability GL 86-003 Applications For License Amendments NA Info GL 86-004 Policy Statement On Engineering Expertise NA Info On Shift GL 86-005 Implementation Of TMI Action Item Il.K.3.5, NA Applies only to Babcock and Wilcox designed plants "Automatic Trip Of Reactor Coolant Pumps" GL 86-006 Implementation Of TMI Action Item I1.K.3.5, NA Applies only to Combustion Engineering designed plants "Automatic Trip of Reactor Coolant Pumps" GL 86-007 Transmittal of NUREG-1 190 Regarding The NA Info 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 Issue CT N-1 Loop operation was addressed in original 1982 SER B-59, (N-i) Loop Operation in BWRs and (4.4.7).

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

GL 86-010 Implementation of Fire Protection NA Info Requirements GL 86-010, Fire Endurance Test Acceptance Criteria NA Info S1 for Fire Barrier Systems Used to Separate Redundant Safe Shutdown Trains Within the Same Fire Area GL 86-011 Distribution of Products Irradiated in NA Does not apply to power reactor.

Research GL 86-012 Criteria for Unique Purpose Exemption NA Does not apply to power reactor.

From Conversion From The Use of Heu Fuel GL 86-013 Potential Inconsistency Between Plant NA Applies only to Babcock and Wilcox and Combustion Safety Analyses and Technical Engineering designed plants Specifications GL 86-014 Operator Licensing Examinations NA Info GL 86-015 Information Relating To Compliance With NA Info 10 CFR 50.49, "Environmental Qualification of Electric Equipment Important To Safety For Nuclear Power Plants" GL 86-016 Westinghouse ECCS Evaluation Models NA Info Page 43 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 86-017 Availability of NUREG-1169, "Technical NA Boiling Water Reactor Findings Related to Generic Issue C-8, BWR MSIC Leakage And Treatment Methods" GL 87-001 Public Availability Of The NRC Operator NA Info Licensing Examination Question Bank GL 87-002 & Verification of Seismic Adequacy of NA Item was applicable only to units with operating license at 003 Mechanical and Electrical Equipment in the time the item was issued.

Operating Reactors, USI A-46

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GL 87-004 Temporary Exemption From Provisions Of NA Item was applicable only to units with operating license at The FBI Criminal History Rule For the time the item was issued.

Temporary Workers

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GL 87-005 Request for Additional Information on NA, Boiling Water Reactor Assessment of License Measures to Mitigate and/or Identify Potential Degradation of Mark I Drywells GL 87-006 Periodic Verification of Leak Tight Integrity NA Item was applicable only to units with operating license at of Pressure Isolation Valves the time the item was issued.

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

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

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

Reactor Coolant System is Partially Filled GL 87-013 Integrity of Requalification Examinations At NA Does not apply to power reactor.

Non-Power Reactors GL 87-014 Operator Licensing Examinations NA Info GL 87-015 Policy Statement On Deferred Plants NA Info GL 87-016 Transmittal of NUREG-1262, "Answers To NA Info Questions On Implementation of 10 CFR 55 On Operators' Licenses" GL 88-001 NRC Position on IGSCC in BWR Austenitic NA Boiling Water Reactor Stainless Steel Piping GL 88-002 Integrated Safety Assessment Program II NA Item was applicable only to units with operating license at the time the item was issued.

Page 44 of 60 * = See last page for status code definition.

ITEM TITLE

• ADDITIONAL INFORMATION GL 88-003 Resolution of GSI 93, "Steam Binding of Cl TVA: letter June 3, 1988. NRC letters dated Auxiliary Feedwater Pumps" February 17, 1988 and July 20, 1988 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 Research NA Does not apply to power reactor.

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

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

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

10 CFR 50.49, "Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants" GL 88-008 Mail Sent or Delivered to the Office of NA Info Nuclear Reactor Regulation GL 88-009 Pilot Testing of Fundamentals Examination NA Boiling Water Reactor GL 88-010 Purchase of GSA Approved Security NA Info Containers GL 88-011 NRC Position on Radiatibn Embrittlement Cl NRC acceptance letter dated June 29, 1989, for both units.

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

-- -- -- -- -- -- -- -- 7 -- -- ----- -- "..........--- -- -- --

GL 88-012 Removal of Fire Protection Requirements NA Info from Technical Specification GL 88-013 Operator Licensing Examinations NA Info GL 88-014 Instrument Air Supply System Problems Cl NRC letter dated July 26, 1990, closing the issue.

Affecting Safety-Related Equipment Unit 2 Action: Complete Unit 2 implementation.

GL 88-015 Electric Power Systems - Inadequate NA Info Control Over Design Process GL 88-016 Removal of Cycle-Specific Parameter NA Info 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.

Page 45 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 88-018 Plant Record Storage on Optical Disks NA Info GL 88-019 Use of Deadly Force by Licensee Guards to NA Does not apply to power reactor.

Prevent Theft of Special Nuclear Material GL 88-020 Individual Plant Examination for Severe 0 Unit 2 Action: Complete evaluation for Unit 2.

Accident Vulnerabilities GL 89-001 Implementation of Programmatic and NA Info Procedural Controls for Radiological Effluent Technical Specifications

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GL 89-002 Actions to Improve the Detection of NA Info Counterfeit and Fraudulently Marketed Products GL 89-003 Operator Licensing Examination Schedule NA Info GL 89-004 Guidelines on Developing Acceptable OV NRC reviewed in 3.9.6 of SSER14 (Unit 1).

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 GL 89-006 Task Action Plan Item I.D.2 - Safety Cl "Safety Parameter Display System" (SPDS) /

Parameter Display System - 10 CIFR "Requirements for Emergency Response Capability" -

50.54(f) NRC reviewed in SSER5, SSER6, and 18.2.2 of SSER15.

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

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

Unit 2 Actions: Prepare procedure and perform baseline inspections.

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

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GL 89-010 Safety-Related Motor-Operated Valve C1 NRC accepted approach in September 14, 1990, letter and Testing and Surveillance reviewed in Appendix EE of SSER16.

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

GL 89-011 Resolution of Generic Issue 101, "Boiling NA Boiling Water Reactor Water Reactor Water Level Redundancy" GL 89-012 Operator Licensing Examination NA Info Page 46 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 89-013 Service Water System Problems Affecting Cl NRC letters dated July 9, 1990 and June 13, 1997, Safety-Related Equipment accepting approach.

Unit 2 Actions: 1) Implement initial performance testing of the heat exchangers; and 2) Establish eddy current baseline data for the Containment Spray heat exchangers.

GL 89-014 Line-Item Improvements in Technical NA Info Specifications - Removal of 3.25 Limit on Extending Surveillance Intervals GL 89-015 Emergency Response Data System NA Info GL 89-016 Installation of a Hardened Wetwell Vent NA Boiling Water Reactor GL 89-017 Planned Administrative Changes to the NA Info NRC Operator Licensing Written Examination Process GL 89-018 Resolution of Unresolved Safety Issues NA Info A-17, "Systems Interactions in Nuclear Power Plants" GL 89-019 Request for Actions Related to Resolution Cl TVA responded by letter dated March 22, 1990. NRC of Unresolved Safety Issue A-47, "Safety acceptance letter dated October 24, 1990, for both units.

Implication of Control Systems in LWR Nuclear Power Plants" Pursuant to Unit 2 Action: Perform evaluation of common mode failures 10 CFR 50.54(f) due to fire.

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

GL 89-021 Request for Information Concerning Status NA Info of Implementation of Unresolved Safety Issue (USI) Requirements GL 89-022 Potential For Increased Roof Loads and C TVA: letter dated December 16, 1981 Plant Area Flood Runoff Depth At Licensed Nuclear Power Plants Due To Recent Change In Probable Maximum Precipitation Criteria Developed by the National Weather Answer to informal question provided in TVA letter dated Service December 16, 1981, and subsequently included in FSAR.

GL did not require a response. No further action required.

GL 89-023 NRC Staff Responses to Questions NA Info Pertaining to Implementation of 10 CFR Part 26 GL 90-001 Request for Voluntary Participation in NRC NA Info 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 Generic NA Info Letter 83-28, Item 2.2 Part 2 "Vendor Interface for Safety-Related Components" GL 90-004 Request for Information on the Status of C TVA responded on June 23, 1990 Licensee Implementation of GSIs Resolved with Imposition of Requirements or CAs Page 47 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 90-005 Guidance for Performing Temporary NA Info Non-Code Repair of ASME Code Class 1,2, and 3 Piping

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GL 90-006 Resolution of Generic Issues 70, "PORV CT NRC letter dated January 9, 1991, accepted TVA's and Block Valve Reliability," and 94, response for both units.

"Additional LTOP Protection for PWRs" Unit 2 Actions: 1) Revise operating instruction and surveillance procedure; and 2) Incorporate testing requirements inthe Technical Specifications.

GL 90-007 Operator Licensing National Examination NA Info Schedule GL90-008 Simulation Facility Exemptions NA Info GL 90-009 Alternative Requirements for Snubber NA Info Visual Inspection Intervals and Corrective Actions G3L 91-001 Removal of the Schedule for the NA Info Withdrawal of Reactor Vessel Material Specimens from Technical Specifications GL 91-002 Reporting Mishaps Involving LLW Forms NA Item was applicable only to units with operating license at Prepared for Disposal . the time the item was issued.

GL 91-003 Reporting of Safeguards Events NA Info GL 91-004 Changes in Technical Specification NA Info Surveillance Intervals to Accommodate a 24-Month Fuel Cycle

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GL 91-005 Licensee Commercial-Grade Procurement NA Info and Dedication Programs GL 91-006 Resolution of Generic Issue A-30, NA Item was applicable only to units with operating license at "Adequacy of Safety-Related DC Power the time the item was issued.

Supplies," Pursuant to 10 CFR 50.54(f)

GL 91-007 GI-23, "Reactor Coolant Pump Seal NA Info 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 Interval for the NA Boiling Water Reactor Electrical Protective Assemblies in Power Supplies for the Reactor Protection System GL 91-010 Explosives Searches at Protected Area NA Does not apply to power reactor.

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

49, "Interlocks and LCOs for Class 1E Tie Breakers," Pursuant to 10 CFR 50.54 GL 91-012 Operator Licensing National Examination NA Info Schedule Page 48 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION 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 Report, NA Info Solenoid-Operated Valve Problems at U.S.

Reactors GL 91-016 Licensed Operators' and Other Nuclear NA Info Facility Personnel Fitness for Duty GL 91-017 Generic Safety Issue 29, "Bolting NA Info Degradation or Failure in Nuclear Power Plants"

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GL 91-018 Information to Licensees Regarding Two NA GL 91-18 has been superseded by RIS 2005-20.

NRC Inspection Manual Sections on

• Resolution of Degraded and Nonconforming Conditions and on Operability GL 91-019 Information to Addressees Regarding New NA Info 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 Licensing NA Info Basis: Request for Voluntary Participation in Pilot Program GL 92-004 Resolution of the Issues Related to Reactor NA Boiling Water Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(f)

GL 92-005 NRC Workshop on the Systematic NA Info Assessment of Licensee Performance (SALP) Program GL 92-006 Operator Licensing National Examination NA Info Schedule GL 92-007 Office of Nuclear Reactor Regulation NA Info Reorganization Page 49 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION 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 the IAEA NA Info International Nuclear Event Scale GL 93-001 Emergency Response Data System Test NA Addressed to specific plant(s).

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

Assemblies, 10 CFR 50.54(f)

Unit 2 Action: Implement modifications and testing.

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 Safety Issue NA Info 106, "Piping and the Use of Highly Combustible Gases in Vital Areas" GL 93-007 Modification of the Technical Specification NA Item was applicable only to units with operating license at Administrative Control Requirements for the time the item was issued.

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

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

Emergency Diesel Generators GL 94-002 Long-Term Solutions and Upgrade of NA Boiling Water Reactor Interim Operating Recommendations for Thermal-Hydraulic Instabilities in BWRs GL 947003 IGSCC of Core Shrouds in BWRs NA Boiling Water Reactor GL 94-004 Voluntary Reporting of Additional NA Info Occupational Radiation Exposure Data GL 95-001 NRC Staff Technical Position on Fire NA Does not apply to power reactor.

Protection for Fuel Cycle Facilities GL 95-002 Use of NUMARC/EPRI Report TR-102348, NA Info "Guideline on Licensing Digital Upgrades,"

in Determining the Acceptability of Performing Analog-to-Digital Replacements under 10 CFR 50.59 Page 50 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 95-003 Circumferential Cracking of Steam OV NRC acceptance letter dated May 16, 1997 (Unit 1) -

Generator Tubes Initial response for Unit 2 on September 7, 2007. TVA responded to a request for additional information on December 17, 2007.

Unit 2 Action: Perform baseline inspection.

GL 95-004 Final Disposition of the Systematic NA Info Evaluation Program Lessons-Learned Issues GL 95-005 Voltage-Based Repair Criteria for C No specific action or response required by the GL; TVA Westinghouse Steam Generator Tubes responded on September 7, 2007.

Affected by Outside Diameter Stress Corrosion Cracking GL 95-006 Changes in the Operator Licensing Program NA Info GL 95-007 Pressure Locking and Thermal Binding of Cl Unit 1 SER for GL 95-07 dated Sept 15, 1999 Safety-Related Power-Operated Gate Valves Unit 2 Action: 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.

GL 95-008 10 CFR 50.54(p) Process for Changes to NA Info Security Plans Without Prior NRC Approval NA Info GL 95-009 Monitoring and Training of Shippers and Carriers of Radioactive Materials NA Info GL 95-010 Relocation of Selected Technical Specifications Requirements Related to Instrumentation GL 96-00 1 Testing of Safety-Related Circuits CI TVA responded for both units on April 18, 1996.

Unit 2 Action: Implement Recommendations.

GL 96-002 Reconsideration of Nuclear Power Plant NA Info Security Requirements Associated with an Internal Threat GL 96-003 Relocation of the Pressure Temperature Cl No response required Limit Curves and Low Temperature Overpressure Protection System Limits Unit 2 Action: Submit Pressure Temperature limits and similar to Unit 1, upon approval, incorporate into licensee-controlled document.

GL 96-004 Boraflex Degradation in Spent Fuel Pool NA Item was applicable only to units with operating license at Storage Racks the time the item was issued.

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

Capability of Safety-Related Motor-Operated Valves Unit 2 Action: 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 Operability and OV NRC letter dated April 6, 1999, accepting TVA response for Containment Integrity During Unit 1.

Design-Basis Accident Conditions Unit 2 Action: Implement modification to provide containment penetration relief.

Page 51 of 60 * = See last page for status code definition.

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

GL 97-001 Degradation of Control Rod Drive OV NRC acceptance letter dated November 4, 1999 (Unit 1).

Mechanism Nozzle and Other Vessel Closure Head Penetrations Unit 2 Action: Provide a report to address the inspection program.

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

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

Requirements for Uranium Recovery Licensees GL 97-004 Assurance of Sufficient Net Positive OV NRC acceptance letter dated June 17, 1998 (Unit 1) -

Suction Head for Emergency Core Cooling Initial response for Unit 2 on September 7, 2007.

and Containment Heat Removal Pumps Unit 2 Actions: Install new sump strainers, and perform other modification-related activities identical to Unit 1.

GL 97-005 Steam Generator Tube Inspection OV NRC acceptance letter dated September 22, 1998 Techniques (Unit 1) - Initial response for Unit 2 on September 7, 2007.

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.

GL 97-006 Degradation of Steam Generator Internals OV NRC acceptance letter dated October 19, 1999 (Unit 1) -

Initial response for Unit 2 on September 7, 2007. TVA, responded to a request for additional information on December 17, 2007.

Unit 2 Action: Perform SG inspections during each refueling outage.

~~~~

... .... ~ ~ ~ . ~.. ~.... ~.. ~ .. ~ ..~ .to. ~ ~.. N Item.

.... was. .. applicable only. units with.

.. .operating.. .... .. . license... at.

GL 98-001 Year 2000 Readiness of Computer NA Item was applicable only to units with operating license at Systems at Nuclear Power Plants the time the item was issued.

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

Associated Potential for Loss of Emergency Mitigation Functions While in a Shutdown Unit 2 Actions: 1) Review the ECCS designs to ensure they Condition do not contain design features which can render them susceptible to common-cause failures; and 2) document the results.

GL 98-003 NMSS Licensees' and Certificate Holders' NA Does not apply to power reactor.

Year 2000 Readiness Programs GL 98-004 Potential for Degradation of the ECCS and OV NRC closure letter dated November 24, 1999 (Unit 1). -

the Containment Spray System After a Initial response for Unit 2 on September 7, 2007.

LOCA Because of Construction and Protective Coating Deficiencies and Unit 2 Actions: Install new sump strainers, and perform Foreign Material in Containment other modification-related activities identical to Unit 1.

GL 98-005 Boiling Water Reactor Licensees Use of the NA Boiling Water Reactor BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds GIL 99-001 Recent Nuclear Material Safety and NA Info Safeguards Decision on Bundling Exempt Quantities Page 52 of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 99-002 Laboratory Testing of Nuclear Grade NA Item was applicable only to units with operating license at Activated Charcoal the time the item was issued.

GL 03-001 Control Room Habitability OT Initial response for Unit 2 on September 7, 2007 Unit 2 Action: Incorporate TSTF-448 into Technical Specifications.

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

Unit 2 Action: Perform baseline inspection.

GL 04-002 Potential Impact of Debris Blockage on OV NRC Audit Report dated February 7, 2007 (Unit 1) - Initial Emergency Recirculation During Design response for Unit 2 on September 7, 2007.

Basis Accidents at PWRs Unit 2 Actions: Install new sump strainers, and perform other modification-related activities identical to Unit 1.

GL 06-001 Steam Generator Tube Integrity and OT Initial response for Unit 2 on September 7, 2007.

Associated Technical Specifications Unit 2 Action: Incorporate TSTF-449 into Technical Specifications.

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

Risk and the Operability of Offsite Power Unit 2 Action: Complete the two unit baseline electrical calculations and implementing procedures.

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

Unit 2 Action: Addressed in CAP/SP. The Fire Protection Corrective Action Program will ensure Unit 2 conforms with NRC requirements and applicable guidelines.

GL 07-001 Inaccessible or Underground Power Cable OV Initial response for Unit 2 on September 7, 2007.

Failures That Disable Accident Mitigation Systems or Cause Plant Transients Unit 2 Action: Complete testing of four additional cables.

GL 08-001 Managing Gas Accumulation in Emergency 0 Core Cooling, Decay Heat Removal, and Containment Spray Systems NUREG- Shift Technical Advisor NA Not applicable to WBN per SSER16.

0737, L.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 SRO C Closed in SSER16.

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

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

0737, I.A.3.1 Exams Page 53 of 60 * = See last page for status code definition.

ITEM TITLE

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

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.

NUREG- Short Term Accident and Procedure Review CI NRC reviewed in Appendix EE of SSER16.

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

NUREG- Shift and Relief Turnover Procedures C Closed in SSER16.

0737, 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 NUREG- Verify Correct Performance of Operating C Closed in SSER16.

0737, I.C.6 Activities NUREG- NSSS Vendor Revision of Procedures CI IR 50-390/391 85-08 closed this item for Unit 1, and NRC 0737, I.C.7 also reviewed in Appendix EE of SSER16.

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

NUREG- Pilot Monitoring of Selected Emergency Cl IR 50-390/391 85-08 closed this item for Unit 1, and NRC 0737, I.C.8 Procedures For Near Term Operating also reviewed in Appendix EE of SSER16.

Licenses Unit 2 Action: Pilot monitor selected emergency procedures for NTOL.

........................................................................... m..............................................................

NUREG- Control Room Design Review OV NRC reviewed in SSER5, SSER6, SSER15, and Appendix 0737, I.D.1 EE of SSER16.

Unit 2 Actions: Complete the CRDR process. Perform rewiring in accordance with ECN 5982. Take advantage of the completed Human Engineering reviews to ensure appropriate configuration for Unit 2 control panels. See CRDR Special Program.

NUREG- Plant-Safety-Parameter-Display Console Cl NRC reviewed in SSER5, SSER6, and 18.2.2 of SSER15.

0737, 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 Testing C Closed in SSER16.

0737, I.G.1 Page 54of 60 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION NUREG- Reactor Coolant Vent System Cl LICENSE CONDITION - NUREG-0737, ll.B.1, "Reactor 0737, ll.B.1 Coolant System Vents" - In the original SER, the NRC found TVA's commitment to install reactor coolant vents acceptable pending verification. This was completed for Unit 1 only in SSER5 (IR 390/84-37).

Unit 2 Action: Verify installation of reactor coolant vents.

NUREG- Plant Shielding Cl NRC reviewed in Appendix EE of SSER16.

0737, 1l.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 CT NRC reviewed in 9.3.2 of SSER16. TVA submitted a 0737, ll.B.3 TS improvement to eliminate requirements for the Post Accident Sampling System using the 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.

NUREG- Training for Mitigating Core Damage C Closed in SSER16.

0737, ll.B.4 NUREG- Relief and Safety Valve Test Requirements Cl NRC reviewed in Technical Evaluation Report attached to 0737, ll.D.1 Appendix EE of SSER15.

Unit 2 Actions: 1) Testing of relief and safety valves;

2) Reanalysis of fluid transient loads for pressurizer relief and safety valve supports and any required modifications;

3) Modifications to pressurizer safety valves, PORVs, PORV block valves and associated piping; and 4) Change motor operated block valves.

NUREG- Valve Position Indication Cl The design was reviewed in the original 1982 SER and 0737, ll.D.3 found acceptable pending confirmation of installation of the acoustic monitoring system. In SSER5 (IR 390/84-35), the staff closed the LICENSE CONDITION for Unit 1 only.

Unit 2 Action: Verify installation of the acoustic monitoring system to PORV to indicate position.

NUREG- Auxiliary Feedwater System Evaluation, Cl Reviewed in Appendix EE of SSER16.

0737, I1.E.1.1 Modifications Unit 2 Action: Perform Auxiliary Feedwater System analysis as it pertains to system failure and flow rate.

NUREG- Auxiliary Feedwater System Initiation and Cl NRC: IR 50-390/84-20 and 50-391/84-16; letters dated 0737, I1.E.1.2 Flow March 29, 1985, and October 31, 1995; SSER 16 Unit 2 Action: Complete procedures and qualification testing.

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

Unit 2 Action: Implement procedures and testing.

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ITEM TITLE ADDITIONAL INFORMATION NUREG- Dedicated Hydrogen Penetrations C NRC: IR 50-390/83-27 and 50-391/83-19; SER 0737, II.E.4.1 (NUREG-0847)

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

NUREG- Accident-Monitoring Instrumentation - Cl Reviewed in SSER9.

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

NUREG- Accident-Monitoring Instrumentation - Cl Reviewed in SSER9.

0737, Iodine/Particulate Sampling II.F.1.2.B. Unit 2 Actions: Install Noble gas, Iodine / particulate sampling, and Containment High Range Monitors.

NUREG- Accident-Monitoring Instrumentation - CI Reviewed in SSER9.

0737, Containment High Range Monitoring I1.F.1.2.C. 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 Instrumentation - CI Reviewed in SSER9.

0737, Containment Pressure II.F.1.2.D. Unit 2 Action: Verify installation of containment pressure indication.

NUREG- Accident-Monitoring Instrumentation - CI Reviewed in SSER9.

0737, Containment Water Level II.F.1.2.E. Unit 2 Action: Verify installation of containment water level monitors.

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ITEM TITLE ADDITIONAL INFORMATION NUREG- Accident-Monitoring Instrumentation - Cl Reviewed in SSER9.

0737. Containment Hvdroaen II.F.1I.2.F. Unit 2 Action: Verify installation of containment hydrogen accident monitoring instrumentation.

NUREG- Instrumentation For Detection of " 0 LICENSE CONDITION - Detectors for Inadequate core 0737, II.F.2 Inadequate Core-Cooling 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 Relief CI Reviewed in original 1982 SER and 8.3.3 of SSER7.

0737, II.G.1 Valves, Block Valves and Level Indicators Unit 2 Action: Implement modifications 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 CI Unit 2 Action: Confirm multi-unit operation will have no 0737, impact on administrative procedures with respect to I1.K.1.10 operability status.

NUREG- Trip Per Low-Level B/S C NRC: letter dated March 29, 1985; SSER 16 0737, II.K.1.17 NUREG- Effect of High Pressure Injection for Small C LICENSE CONDITION - Effect of high pressure injection 0737, Break LOCA With No Auxiliary Feedwater for small break LOCA with no auxiliary feedwater II.K.2.13 (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 System C LICENSE CONDITION - Voiding in the reactor coolant 0737, system (NUREG-0737, II.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 0737, II.K.3.1 conclusion that there is no need for an automatic PORV isolation system (NRC letter dated June 29, 1990).

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

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ITEM TITLE ADDITIONAL INFORMATION NUREG- Reporting SV/RV Failures/Challenges CT (Action from GL 82-16) - NRC reviewed in Appendix EE of 0737, II.K.3.3 SSER16.

Unit 2 Action: Include, as necessary, in Technical Specifications submittal.

NUREG- Auto Trip of RCPS Cl Reviewed in 15.5.4 of original 1982 SER; became License 0737, II.K.3.5 Condition 35. The staff determined that their review of Item II.K.3.5 did not have to be completed to support the full power license and considered this license condition resolved in SSER4. The item was further reviewed in Appendix EE of SSER16.

Unit 2 Action: Implement modifications as required.

NUREG- PID Controller Cl Reviewed in original 1982 SER.

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

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

0737, II.K.3.10 Unit 2 Action: Unit 2 Technical Specifications and surveillance procedures will address this issue.

NUREG- Confirm Existence of Anticipatory Reactor C Closed in SSER16.

0737, Trip Upon Turbine Trip II.K.3.12 NUREG- Report On Outage of Emergency Core C LICENSE CONDITION - Report on outage of emergency 0737, Cooling System core cooling system (NUREG-0737, II.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 NUREG- Power On Pump Seals Cl NRC reviewed and closed in IR 390/84-35 based on Diesel 0737, Generator (DG) power to pump sealing cooling system.

II.K.3.25 Unit 2 Action: Ensure DG power is provided to pump sealing cooling system.

NUREG& Small Break LOCA Methods Cl TVA: letter dated October 29, 1981 0737, I1.K.3.30 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 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.

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ITEM TITLE ADDITIONAL INFORMATION NUREG- Plant Specific Analysis Cl The staff determined in SSER4 that their review of Items 0737, I1.K.3.30 and II.K.3.31 did not have to be completed to Il.K.3.31 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 I1.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.

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

II1.A.1.1 The NRC review of Emergency Preparedness in SSER1 3 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- Upgrade Emergency Support Facilities C LICENSE CONDITION - Emergency Preparedness 0737, (NUREG-0737, III.A.1, III.A.2)

III.A.1.2 The NRC review of Emergency Preparedness in SSER1 3 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 0737, III.A.2 (NUREG-0737, III.A.1, III.A.2)

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

The NRC completed the review of the REP in SSER20.

NUREG- Primary Coolant Outside Containment OT Resolved for Unit 1 only in SSER10; reviewed in Appendix 0737, EE of SSER16.

I11.D.1.1 Unit 2 Actions: Include the waste gas disposal system in the leakage reduction program and incorporate in Unit 2 Technical Specifications.

NUREG- In-Plant Iodine Radiation Monitoring Cl NRC reviewed in Appendix EE of SSER16.

0737, III.D.3.3 Unit 2 Action: Complete modifications for Unit 2.

NUREG- Control-Room Habitability OV TVA: letter dated October 29, 1981 0737, II1.D.3.4 NRC: SSER16 NRC reviewed in SER and in Appendix EE of SSER16.

Unit 2 Action: Complete with CRDR completion.

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ITEM TITLE

• ADDITIONAL INFORMATION 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.

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.

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GENERIC COMMUNICATIONS STATUS = CLOSED

GENERIC COMMUNICATIONS STATUS = CLOSED ITEM TITLE

• ADDITIONAL INFORMATION B 73-001 Faulty Overcurrent Trip Delay Device in C TVA: letter dated April 4, 1973 Circuit Breakers for Engineered Safety Systems NRC: IR 390/391 75-5 B 73-002 Malfunction of Containment Purge Supply C TVA: letter dated August 22, 1973 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 Hydraulic C TVA: memo dated February 7, 1985 Shock Absorbers NRC: IR 390/391 85-08 B 74-001 Valve Deficiencies C TVA: letter dated April 15, 1974 NRC: IR 390/391 75-5 B 74-006 Defective Westinghouse Type W-2 Control C TVA: letter dated October 18, 1974 Switch Component NRC: IR 390/391 75-6 B 74-008 Deficiency in the ITE Molded Case Circuit C TVA: letter dated August 21, 1974 Breakers, Type HE-3 NRC: IR 390/391 75-5 B 74-009 Deficiency in GE Model 4KV Magne-Blast C TVA: letter dated September 20, 1974 Circuit Breakers NRC: IR 390/391 76-6 B 74-011 Improper Wiring of Safety Injection Logic at C NRC: IR 390/391 75-6 Zion 1 & 2 B 74-012 Incorrect Coils in Westinghouse Type SG C NRC: IR 390/391 75-5 Relays at Trojan B 74-013 Improper Factory Wiring on GE Motor C TVA: letter dated December 24, 1974 Control Centers at Fort Calhoun NRC: IR 390/391 75-5

......................................................................................................... = . ... ... ... ... ..

B 74-016 Improper Machining of Pistons in Colt C TVA: letter dated January 2, 1975 Industries (Fairbanks-Morse)

Diesel-Generators NRC: IR 390/391 75-3 B 76-003 Relay Malfunctions - GE Type STD Relays C TVA: letter dated May 17, 1976 NRC: IR 390/391 76-6 B 76-005 Relay Failures - Westinghouse BFD C TVA: letter dated June 7, 1976 Relays NRC: IR 390/391 85-08 B 76-006 Diaphragm Failures in Air Operated C TVA: memo dated January 25, 1985 Auxiliary Actuators for Safety/Relief Valves NRC: IR 390/391 85-08 Page I of 8 * = See last page for status code definition.

ITEM TITLE

• ADDITIONAL INFORMATION B 76-007 Crane Hoist Control Circuit Modifications C TVA: letter dated October 29, 1976 NRC: IR 390/391 85-08 B 77-001 Pneumatic Time Delay Relay Setpoint Drift C TVA: letter dated July 1, 1977 NRC: IR 390/391 85-08 B 77-002 Potential Failure Mechanism in Certain C TVA: letter dated November 11, 1977 Westinghouse AR Relays with Latch Attachments NRC: IR 390/391 85-08 B 77-005 & Electrical Connector Assemblies C TVA: letter dated January 17, 1978 77-005 A NRC: IR 390/78-11 and 391/78-09 B 77-006 Potential Problems with Containment C Item was applicable only to units with operating license at Electrical Penetration Assemblies the time the item was issued.

NRC: IR 390/391 85-08

.................................................................... ,.............m........................................................

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

................................................................................. n........................................................

B 77-008 Assurance of Safety and Safeguards C Item concerns a multi-unit issue that was completed for During an Emergency - Locking Systems both units.

TVA: letter dated March 1, 1978 NRC: IR 390/78-11 and 391/78-09 B 78-001 Flammable Contact - Arm Retainers in C TVA: letter dated March 20, 1978 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 78-005 Malfunctioning of Circuit Breaker Auxiliary C TVA: letter dated June 12, 1978 Contact Mechanism - GE Model CR105X NRC: IR 390/78-17 and 391/78-15 B 78-006 Defective Cutler-Hammer Type M Relays C NRC: IR 390/78-22 and 391/78-19 With DC Coils 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 78-012 Atypical Weld Material in Reactor Pressure. C TVA: Westinghouse letter dated October 29, 1979 Vessel Welds NRC: IR 390/391 81-04 B 79-001 Environmental Qualification of Class 1E C NRC: IR 390/80-06 and 391/80-05 Equipment Page 2 of 8 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION B 79-003 Longitudinal Weld Defects in ASME SA-312 C TVA: letter dated July 16, 1981 Type 304 SS Pipe Spools Manufactured by Youngstown Welding & Engineering NRC: IRs 390/82-21 and 391/82-17; 390/84-35 and 391/84-33 B 79-004 Incorrect Weights for Swing Check Valves C TVA: letter dated October 20, 1980 Manufactured by Velan Engineering Corporation NRC: IR 390/83-15 and 391/83-11 B 79-006 Review of Operational Errors and System C NRC: IR 390/80-06 and 391/80-05 Misalignments Identified During the Three Mile Island Incident B 79-007 Seismic Stress Analysis of Safety-Related C TVA: letter dated May 31, 1979 Piping NRC: IR 390/79-30 and 391/79-25 B 79-011 Faulty Overcurrent Trip Device in Circuit C TVA: letter dated July 20, 1979 Breakers for Engineering Safety Systems NRC: IR 390/79-30 and 391/79-25 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 B 79-015 Deep Draft Pump Deficiencies C TVA: letter dated January 24, 1992 NRC: IR 390/391 95-70 B 79-023 Potential Failure of Emergency Diesel C TVA: letter dated October 29, 1979 Generator Field Exciter Transformer NRC: IR 390/80-06 and 391/80-05 B 79-025 Failures of Westinghouse BFD Relays in C TVA: letter dated January 4, 1980 Safety-Related Systems NRC: IR 390/80-03 and 391/80-02 B 79-028 Possible Malfunction of NAMCO Model C TVA: letter dated April 1, 1993 EA1 80 Limit Switches at Elevated Temperatures NRC: IR 390/391 93-32 B 80-003 Loss of Charcoal from Standard Type II, C TVA: letter dated March 21, 1980 2 Inch, Tray Adsorber Cells NRC: IR 390/80-15 and 391/80-12 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 3 of 8 * = See last page for status code definition.

ITEM TITLE

• ADDITIONAL INFORMATION B 80-015 Possible Loss of Emergency Notification C Item concerns a multi-unit issue that was completed for System with Loss of Offsite Power both units.

NRC: IR 390/391 85-08 B 80-016 Potential Misapplication of Rosemount, Inc. C TVA: letter dated August 29, 1980 Models 1151 and 1152 Pressure Transmitters With Either "A"or "D" Output NRC: IR 390/391 81-17 Codes B 80-019 Mercury-Wetted Matrix Relay in Reactor C. TVA: letter dated September 4, 1980 Protective Systems of Operating Nuclear Power Plants Designed by CE NRC: NUREG/CR 4933; IR 390/391 81-17 B 80-021 Valve Yokes Supplied by Malcolm Foundry C TVA: letter dated May 6, 1981 Co., Inc.

NRC: 390/391 85-08 B 80-023 Failures of Solenoid Valves Manufactured C TVA: letter dated March 31, 1981 by Valcor Engineering Corporation NRC: IR 390/391 81-17; NUREG/CR 5292 B 81-002 Failure of Gate Type Valves to Close C TVA: letter dated September 30, 1983 Against Differential Pressure NRC: IR 390/391 84-03 B 81-003 Flow Blockage of Cooling Water to Safety C TVA: letters dated July 21, 1981 and March 21, 1983 System Components by Asiatic Clams and Mussels NRC: IR 390/391 81-17 B 82-001 Alteration of Radiographs of Welds in C NRC: IR 390/391 85-08 Piping Subassemblies B 82-004 Deficiencies in Primary Containment C TVA: letter dated January 24, 1983 Electrical Penetration Assemblies NRC: IR 390/83-10 and 391/83-08 B 83-001 Failure of Trip Breakers (Westinghouse C NRC: IRs 390/391 85-08 and 390/391 92-13 DB-50) to Open on Automatic Trip Signal B 83-005 ASME Nuclear Code Pumps and Spare C TVA: letter dated September 7, 1983 Parts Manufactured by the Hayward Tyler Pump Company NRC: IR 390/85-03 and 391/85-04; NUREG/CR 5297 B 83-007 Apparently Fraudulent Products Sold by C TVA: letter dated March 22, 1984 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 Than the NRC: IR 390/84-35 and 391/84-33 Reactor Trip System B 84-002 Failure of GE Type HFA Relays In Use In C TVA: letter dated July 10, 1984 Class 1 E Safety Systems NRC: IR 390/391 84-42 and IR 390/84-77 and 391/84-54 B 85-003 Motor-Operated Valve Common Mode C Superseded by GL 89-10 Failures During Plant Transients Due to Improper Switch Settings Page 4 of 8 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION B 86-002 Static "0"Ring Differential Pressure C TVA: letter dated November 20, 1986 Switches NRC: IR 390/391/90-24 B 86-003 Potential Failure of Multiple ECCS Pumps C TVA: letter dated November 14, 1986 Due to Single Failure of Air-Operated Valve in Minimum Flow Recirculation Line NRC: IR 390/391/87-03 B 87-001 Thinning of Pipe Walls in Nuclear Power C TVA: letter dated September 18, 1987 Plants NRC: NUREG/CR 5287 Closed to GL 89-08

..........................................................................................................................- o...............

B 88-001 Defects in Westinghouse Circuit Breakers C TVA: letter dated November 15, 1991 NRC: IR 390/391 93-01 B 88-003 Inadequate LatchEngagement in HFA C TVA: letter dated April 13, 1992 Type Latching Relays Manufactured by General Electric (GE) Company NRC: IR 390/391 92-13 B 03-004 Rebaselining of Data in the Nuclear C TVA: letter dated December 18, 2003 Management and Safeguards System Item concerns a multi-unit issue that was completed for both units.

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 Response C TVA: letters dated January 20, 2006 and August 16, 2006.

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

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

C 76-002 Relay Failures - Westinghouse BF (AC) C TVA: letter dated November 22, 1976 informed NRC that and BFD (DC) Relays these relay types are not used in Class 1E circuits.

NRC: IR 50/390/76-11 and 50/391/76-11 C 76-005 Hydraulic Shock And Sway Suppressors - C TVA: letter dated January 7, 1977 informed NRC that no Maintenance of Bleed and Lock-Up Grinnell shock suppressors or sway braces have been or Velocities on ITT Grinnell's Model Nos. - will be installed at WBN.

Fig. 200 And Fig. 201, Catalog Ph-74-R GL 78-002 Asymmetric Loads Background and C NRC: Reviewed in SSER15 - Appendix C (June 1995).

Revised Request for Additional Information Resolved by approval of leak-before-break analysis.

GL 78-034 Reactor Vessel Atypical Weld Material C See B 78-12.

GL 79-020 Cracking in Feedwater Lines C See B 79-13.

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ITEM TITLE ADDITIONAL INFORMATION GL 80-002 QA Requirements Regarding Diesel C TVA: FSAR 9.5.4.2 Generator Fuel Oil GL 80-046/47 Generic Technical Activity A-12, "Fracture C No response was required for this GL, and NUREG-0577 Toughness and Additional Guidance on states that the lamellar tearing aspect of this issue was Potential for Low Fracture toughness and resolved by the NUREG. Further, the NUREG states that Laminar Tearing on PWR Steam Generator for plants under review, the fracture toughness issue was Coolant Pump Supports" resolved.

GL 80-113 Control of Heavy Loads C Superseded by GL 81-007.

GL 81-004 Emergency Procedures and Training for C Superseded by Station Blackout Rule.

Station Blackout Events

...~~~~~~~~~ ~ ~~ ~ . . . . .......... . . TVA. leter date . Nov

. .......... . mbe 7,1983 and.........

GL 83-028 "Required Actions Based on Generic C TVA: letters dated November 7, 1983 and Implications of Salem ATWS Events: December 4, 1987 1.2 - Post Trip Review Data and NRC: IR 50-390, 391/86-04 Information Capability GL 83-028 "Required Actions Based on Generic C TVA: letters dated November 7, 1983, March 22, 1985 Implications of Salem ATWS Events:

NRC: IR 50-390/86-04 and 50-391/86-04; letter dated 4.3 - Reactor Trip System Reliability June 18, 1990 (Automatic Actuation of Shunt Trip Attachment)

GL 85-011 Completion of Phase II of "Control of Heavy C See GL 81-07.

Loads at Nuclear Power Plants,"

NUREG-0612 GL 87-012 Loss of Residual Heat Removal While The C This GL was superseded by GL 88-17.

Reactor Coolant System is Partially Filled GL 89-007 Power Reactor Safeguards Contingency C TVA: letter dated October 31, 1989 Planning for Surface Vehicle Bombs NRC: memo dated June 26, 1990 GL 89-022 Potential For Increased Roof Loads and C TVA: letter dated December 16, 1981 Plant Area Flood Runoff Depth At Licensed Nuclear Power Plants Due To Recent Change In Probable Maximum Precipitation Criteria Developed by the National Weather Answer to informal question provided in TVA letter dated Service December 16, 1981, and subsequently included in FSAR.

GL did not require a response. No further action required.

GL 90-004 Request for Information on the Status of C TVA responded on June 23, 1990 Licensee Implementation of GSIs Resolved with Imposition of Requirements or CAs 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 95-005 Voltage-Based Repair Criteria for C No specific action or response required by the GL; TVA Westinghouse Steam Generator Tubes responded on September 7, 2007.

Affected by Outside Diameter Stress Corrosion Cracking Page 6 of 8 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION NUREG- Shift Manning C Closed in SSER16.

0737, I.A.1.3

........................ m.................................................................................................................

NUREG- Immediate Upgrade of RO and SRO C Closed in SSER16.

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

0737, I.A.2.3

........................ m...........................................................................*......................................

NUREG- Revise Scope and Criteria for Licensing C Closed in SSER16.

0737, I.A.3.1 Exams NUREG- Shift and Relief Turnover Procedures C Closed in SSER16.

0737, I.C.2

................................................................................................ .o.. .....................................

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 NUREG- Verify Correct Performance of Operating C Closed in SSER16.

0737, I.C.6 Activities

........................ *..................................................................n...............................................

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

0737, I.G.1 NUREG- Training for Mitigating. Core Damage C Closed in SSER16.

0737, ll.B.4 NUREG- Dedicated Hydrogen Penetrations C NRC: IR 50-390/83-27 and 50-391/83-19; SER 0737, I1.E.4.1 (NUREG-0847)

NUREG- Review ESF Valves C NRC: letter dated March 29, 1985; SSER 16 0737, I1.K.1.5 NUREG- Trip Per Low-Level B/S C NRC: letter dated March 29, 1985; SSER 16 0737, II.K.1.17 NUREG- Effect of High Pressure Injection for Small C LICENSE CONDITION - Effect of high pressure injection 0737, Break LOCA With No Auxiliary Feedwater for small break LOCA with no auxiliary feedwater II.K.2.13 (NUREG-0737, I1.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 System C LICENSE CONDITION - Voiding in the reactor coolant 0737, system (NUREG-0737, II.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 0737, II.K.3.1 conclusion that there is no need for an automatic PORV isolation system (NRC letter dated June 29, 1990).

7.............................................................................

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

NUREG- Confirm Existence of Anticipatory Reactor C Closed in SSER16.

0737, Trip Upon Turbine Trip I1.K.3.12 NUREG- Report On Outage of Emergency Core C LICENSE CONDITION - Report on outage of emergency 0737, Cooling System core cooling 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 NUREG- Emergency Preparedness, Short Term C LICENSE CONDITION - Emergency Preparedness 0737, (NUREG-0737, 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 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- Upgrade Emergency Support Facilities C LICENSE CONDITION - Emergency Preparedness

.0737, (NUREG-0737, III.A.1, III.A.2)

II1.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 0737, III.A.2 (NUREG-0737, III.A.1, III.A.2)

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

The NRC completed the review of the REP in SSER20.

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.

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GENERIC COMMUNICATIONS STATUS = CLOSED/IMPLEMENTATION

GENERIC COMMUNICATIONS STATUS = CLOSED/IMPLEMENTATION ITEM TITLE

• ADDITIONAL INFORMATION B 74-003 Failure of Structural or Seismic Support Cl TVA: memo dated January 22, 1985 Bolts on Class I Components 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.

B 74-015 Misapplication of Cutler-Hammer Three Cl TVA: letter dated May 5, 1975 Position Maintained Switch Model No.

10250T NRC: IR 390/391 75-5 Unit 2 Action: Install modified A3 Cutler-Hammer 10250T switches.

B 75-003 Incorrect Lower Disc Spring and Clearance Cl TVA: letter dated May 16, 1975 Dimension in Series 8300 and 8302 ASCO Solenoid Valves 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: Modify valves not modified at factory.

B 75-004 Cable Fire at BFNPP Cl NRC: IR 390/391 85-08 Closed to Fire Protection CAP Part of Fire Protection CAP B 75-005 Operability of Category I Hydraulic Shock CI TVA: letter dated June 16, 1975 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-2 Control Cl TVA: letter dated July 31, 1975 Switches NRC: IR 390/85-25 and 391/85-20 Unit 2 Action: Inspect Westinghouse Type OT-2 control switches.

B 76-002 Relay Coil Failures - GE Types HFA, CI Unit 2 Action: Repair or replace relays before HGA, HKA, HMA Relays preoperational tests.

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ITEM TITLE

• ADDITIONAL INFORMATION B 77-003 On-Line Testing of the Westinghouse Solid CI Unit 2 Action: Include necessary periodic testing in test State Protection System procedures.

B 78-004 Environmental Qualification of Certain Stem Cl TVA: letter dated December 19, 1978 Mounted Limit Switches Inside Reactor Containment NRC: IR 390/82-13 and 391/82-10 Closed to EQ Program IR 50-390/82-13 and 50-391/82-10 (April 22, 1982) accepted approach.

Unit 2 Action: Ensure NAMCO switches have been replaced.

B 79-002 Pipe Support Base Plate Designs Using Cl NRC review of HAAUP Program in NUREG-1232, SSER6, Concrete Expansion Anchor Bolts and SSER8.

Unit 2 Actions: Addressed in CAP/SP. Conduct a complete review of affected support calculations, and perform the necessary revisions to design documents and field modifications to achieve compliance.

B 79-009 Failure of GE Type AK-2 Circuit Breaker in Cl TVA: letter dated June 20, 1979 Safety Related Systems Unit 2 Action: Complete preservice preventive maintenance on AK-2 Circuit Breakers.

B 79-014 Seismic Analysis for As-Built CI NRC review of HAAUP Program in NUREG-1232, SSER6, Safety-Related Piping Systems and SSER8.

Unit 2 Actions: Addressed in CAP/SP. Initiate a Unit 2 hanger walkdown and hanger analysis program similar to the program for Unit 1. Complete re-analysis of piping and associated supports as necessary. Perform modifications as required by re-analysis.

B 79-021 Temperature Effects on Level Cl Reviewed in 7.2.5 of both the original 1982 SER and Measurements SSER14.

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.

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 79-027 Loss of Non-Class 1E I & C Power System Cl TVA responded to the Bulletin on March 1, 1982. Reviewed Bus During Operation in 7.5.3 of the original 1982 SER.

Unit 2 Action: Issue appropriate emergency procedures.

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ITEM TITLE ADDITIONAL INFORMATION B 80-004 Analysis of a PWR Main Steam Line Break CI IR 50-390/85-60 and 50-391/85-49 (December 6, 1985) with Continued Feedwater Addition required completion of actions that included determination of temperature profiles inside and outside of containment following a MSLB for Unit 1.

Unit 2 Action: Complete analysis for Unit 2.

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

Chemical Volume Control System Holdup Tanks Unit 2 Action: Complete surveillance procedures for Unit 2.

B 80-006 Engineered Safety Feature Reset Control Cl TVA response dated March 11, 1982. Reviewed in 7.3.5 of the original 1982 SER.

Unit 2 Action: Perform verification during the preoperational testing.

B 80-010 Contamination of Nonradioactive System Cl Unit 2 Actions: 1) Correct deficiencies involving monitoring and Resulting Potential for Unmonitored, of systems; and 2) include proper monitoring of Uncontrolled Release of Radioactivity to non-radioactive systems in procedures.

Environment 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 1 in IR 50-390/95-46 (August 1, 1995).

Unit 2 Action: Complete implementation for Unit 2.

B 80-012 Decay Heat Removal System Operability Cl NRC: IR 390/391 85-08; NUREG/CR 4005 Unit 2 Action: Implement operating instructions and abnormal operating instructions (AOIs) for RHR.

B 80-018 Maintenance of Adequate Minimum Flow Cl IR 50-390/85-60 and 50-391/85-49 (Unit 1)

Thru Centrifugal Charging Pumps Following Secondary Side High Energy Rupture Unit 2 Action: Implement design and procedure changes.

B 80-020 Failure of Westinghouse Type W-2 Spring Cl Unit 2 Action: Modify switches.

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

B 82-002 Degradation of Threaded Fasteners in the Cl TVA: memo dated February 6, 1985 Reactor Coolant Pressure Boundary of PWR Plants 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.

B 83-004 Failure of the Undervoltage Trip Function of Cl NRC: IR 390/391 85-08 Reactor Trip Breakers Unit,2 Action: Install new undervoltage attachment with wider grooves on the reactor trip breakers.

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ITEM TITLE ADDITIONAL INFORMATION B 83-006 Nonconforming Material Supplied by CI TVA: letter dated February 2, 1984 Tube-Line Facilities 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.

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 Feedwater Cl TVA: letter dated January 27, 1986 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.

B 85-002 Undervoltage Trip Attachment of Cl Unit 2 Action: Install automatic shunt trip on the Westinghouse DB-50 Type Reactor Trip Westinghouse DS-416 reactor trip breakers on Unit 2.

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

B 88-002 Rapidly Propagating Fatigue Cracks in Cl NRC acceptance letter dated June 7, 1990, for both units.

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 88-004 Potential Safety-Related Pump Loss Cl NRC acceptance letter dated May 24, 1990, for both units.

Unit 2 Action: Perform calculations and install check valves to prevent pump to pump interaction.

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

Piping Supplies, Inc. and West Jersey Manufacturing Company Unit 2 Action: Complete review to locate installed WJM material and perform in-situ hardness testing for Unit 2.

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ITEM TITLE ADDITIONAL INFORMATION B 88-008 Thermal Stresses in Piping Connected to Cl NRC acceptance letter dated September 19, 1991, for both Reactor Cooling Systems units.

Unit 2 Action: Implement program to prevent thermal stratification.

B 88-009 Thimble Tube Thinning in Westinghouse CI Reviewed in Appendix EE of SSER16.

Reactors 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.

B 88-010 Nonconforming Molded-Case Circuit CI Unit 2 Action: Replace those circuits not traceable to a Breakers circuit breaker manufacturer.

B 88-011 Pressurizer Surge Line Thermal CI NRC SER on "Leak-Before-Break" (April 28, 1993) and Stratification reviewed in Appendix EE of SSER16.

Unit 2 Action: 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 Generator CI NRC acceptance letter dated September 26, 1991 for both Tube Mechanical Plugs units.

Unit 2 Action: Remove SG tube plugs.

B 89-002 Stress Corrosion Cracking of CI NRC reviewed in Appendix EE of SSER16.

High-Hardness Type 410 Stainless Steel Internal Preloaded Bolting in Anchor Unit 2 Actions: Replace the flapper assembly hold-down Darling Model S350W Swing Check Valves bolts fabricated on the 14 (12 valves are installed) Atwood or Valves of Similar Nature and Morrell Mark No. 47W450-53 check valves.

Replacement bolts are to be fabricated from ASTM F593 Alloy 630. A review of the remaining Unit2 safety related swing check valves will be performed.

B 89-003 Potential Loss of Required Shutdown Cl TVA: letter dated June 19, 1990 Margin During Refueling Operations NRC: IR 390/391 94-04 and letter dated June 22, 1990 NRC acceptance letter dated June 22, 1990.

Unit 2 Action: Ensure that requirements for fuel assembly configuration, fuel loading and training are included in Unit 2.

B 90-001 Loss of Fill-Oil in Transmitters Cl Unit 2 Action: Implement applicable recommendations from Manufactured by Rosemount this Bulletin including identification of potentially defective transmitters and an enhanced surveillance program which monitors transmitters for loss of fill oil.

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

Fuel, Over Fuel in the Reactor, or Over Safety-Related Equipment Unit 2 Action: Unit 2 Heavy Loads Program will be in compliance with NUREG-0612.

GL 79-036 Adequacy of Station Electric Distribution CI This GL tracked compliance with BTP PSB-1, "Adequacy of Systems Voltages Station Electric Distribution System Voltages."

Unit 2 Action: Perform verification during the preoperational testing.

............... *................................................................................................m..........................

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ITEM TITLE ADDITIONAL INFORMATION GL 80-021 B 80-05, "Vacuum Condition Resulting in Cl 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.

GL 80-090 NUREG-0737, TMI (Prior and future GLs, Cl See NUREG items in this list.

with the exception of certain discrete scopes, have been screened into NUREG list for those applicable to Watts Bar 2)

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 SSER13 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.

GL 81-014 Seismic Qualification of Auxiliary Feedwater Cl TVA: FSAR 10.4.9 Systems Unit 2 Action: Additional Unit 2 implementing procedures or other activity is required for completion.

GL 81-021 Natural Circulation Cooldown Cl TVA responded December 3, 1981.

Unit 2 Action: Issue operating procedures.

GL 82-033 Supplement to NUREG-0737, Cl "Safety Parameter Display System" (SPDS) /

"Requirements for Emergency Response "Requirements for Emergency Response Capability" -

Capability" NRC reviewed in SSER5, SSER6, and 18.2.2 of SSER15.

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

GL 83-01Oc Resolution of TMI Action Item II.K.3.5., CI TVA: letters dated January 5, 1984 and June 25, 1984 "Automatic Trip of Reactor Coolant Pumps" NRC: letter dated June 8, 1990.

Unit 2 Action: Incorporate emergency response guidelines into applicable procedures.

GL 83-028 "Required Actions Based on Generic Cl TVA: letters dated November 7, 1983 and August 24, 1990 Implications of Salem ATWS Events:

NRC: letters dated October 20, 1986 and June 18, 1990 2.1 - Equipment Classification and Vendor Interface (Reactor Trip System Components)

Unit 2 Action: Ensure that required information on Critical Structures and Components is properly incorporated into procedures.

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ITEM TITLE

• ADDITIONAL INFORMATION GL 83-028 "Required Actions Based on Generic Cl Unit 2 Action: Enter engineering component background Implications of Salem ATWS Events: data in INPO's Equipment Performance and Information Exchange System (EPIX) for Unit 2.

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

GL 83-028 "Required Actions Based on Generic Cl TVA: letter dated May 19, 1986 Implications of Salem ATWS Events:

4.1 - Reactor Trip System Reliability (Vendor Related Modifications) Unit 2 Action: Confirm vendor-recommended DS416 breaker modifications are implemented.

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 85-002 Recommended Actions Stemming From Cl TVA responded to the GL on June 17, 1985.

NRC Integrated Program for the Resolution of Unresolved Safety Issues Regarding Unit 2 Action: Perform SG inspection.

Steam Generator Tube Integrity GL 85-012 Implementation Of TMI Action Item I1.K.3.5, CI "Implementation of TMI Item II.K.3.5" - Reviewed in 15.5.4 "Automatic Trip Of Reactor Coolant Pumps" of original 1982 SER; became License Condition 35. The staff determined that their review of Item I1.K.3.5 did not have to be completed to support the full power license and considered this license condition resolved in SSER4. The item was further reviewed in Appendix EE of SSER16.

Unit 2 Action: Implement modifications as required.

GL 88-003 Resolution of GSI 93, "Steam Binding of CI TVA: letter June 3, 1988. NRC letters dated Auxiliary Feedwater Pumps" February 17, 1988 and July 20, 1988 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 88-005 Boric Acid Corrosion of Carbon Steel Cl NRC acceptance letter dated August 8, 1990 for both units.

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

GL 88-007 Modified Enforcement Policy Relating to CI See Special Program for Environmental Qualification.

10 CFR 50.49, "Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants" GL 88-011 NRC Position on Radiation Embrittlement CI NRC acceptance letter dated June 29, 1989, for both units.

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

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ITEM TITLE ADDITIONAL INFORMATION GL 88-014 Instrument Air Supply System Problems Cl NRC letter dated July 26, 1990, closing the issue.

Affecting Safety-Related Equipment Unit 2 Action: Complete Unit 2 implementation.

GL 88-017 Loss of Decay Heat Removal CI 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 89-006 Task Action Plan Item i.D.2 - Safety Cl "Safety Parameter Display System" (SPDS) /

Parameter Display System - 10 CFR "Requirements for Emergency Response Capability" -

50.54(f) NRC reviewed in SSER5, SSER6, and 18.2.2 of SSER15.

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

.............................................................. =...........................................................................

GL 89-008 Erosion/Corrosion-Induced Pipe Wall Cl Unit 1 Flow Accelerated Corrosion Program reviewed in Thinning IR 390/94-89 (February 1995).

Unit 2 Actions: Prepare procedure and perform baseline inspections.

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

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

GL 89-013 Service Water System Problems Affecting CI NRC letters dated July 9, 1990 and June 13, 1997, Safety-Related Equipment accepting approach.

Unit 2 Actions: 1) Implement initial performance testing of the heat exchangers; and 2) Establish eddy current baseline data for the Containment Spray heat exchangers.

GL 89-019 Request for Actions Related to Resolution Cl TVA responded by letter dated March 22, 1990. NRC of Unresolved Safety Issue A-47, "Safety acceptance letter dated October 24, 1990, for both units.

Implication of Control Systems in LWR Nuclear Power Plants" Pursuant to Unit 2 Action: Perform evaluation of common mode failures 10 CFR 50.54(f) due to fire.

GL 93-004 Rod Control System Failure and Cl NRC letter dated December 9, 1994, accepted TVA Withdrawal of Rod Control Cluster commitments for both units.

Assemblies, 10 CFR 50.54(f)

Unit 2 Action: Implement modifications and testing.

GL 95-007 Pressure Locking and Thermal Binding of Cl Unit 1 SER for GL 95-07 dated Sept 15, 1999 Safety-Related Power-Operated Gate Valves Unit 2 Action: 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.

GL 96-001 Testing of Safety-Related Circuits Cl . TVA responded for both units on April 18, 1996.

Unit 2 Action: Implement Recommendations.

GL 96-003 Relocation of the Pressure Temperature Cl No response required Limit Curves and Low Temperature Overpressure Protection System Limits Unit 2 Action: Submit Pressure Temperature limits and similar to Unit 1, upon approval, incorporate into licensee-controlled document.

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ITEM TITLE ADDITIONAL INFORMATION GL 96-005 Periodic Verification of Design-Basis Cl SE of TVA response to GL 96-05 dated July 21, 1999.

Capability of Safety-Related Motor-Operated Valves Unit 2 Action: 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.

NUREG- Short Term Accident and Procedure Review CI NRC reviewed in Appendix EE of SSER16.

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

NUREG- NSSS Vendor Revision of Procedures Cl IR 50-390/391 85-08 closed this item for Unit 1, and NRC 0737, I.C.7 also reviewed in Appendix EE of SSER16.

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

NUREG- Pilot Monitoring of Selected Emergency Cl IR 50-390/391 85-08 closed this item for Unit 1, and NRC 0737, I.C.8 Procedures For Near Term Operating also reviewed in Appendix EE of SSER16.

Licenses Unit 2 Action: Pilot monitor selected emergency procedures for NTOL.

NUREG- Plant-Safety-Parameter-Display Console CI NRC reviewed in SSER5, SSER6, and 18.2.2 of SSER15.

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

NUREG- Reactor Coolant Vent System Cl LICENSE CONDITION - NUREG-0737, Il.B.1, "Reactor 0737, lI.B.1 Coolant System Vents" - In the original SER, the NRC found TVA's commitment to install reactor coolant vents acceptable pending verification. This was completed for Unit 1 only in SSER5 (IR 390/84-37).

Unit 2 Action: Verify installation of reactor coolant vents.

NUREG- Plant Shielding Cl NRC reviewed in Appendix EE of SSER16.

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

NUREG- Relief and Safety Valve Test Requirements Cl NRC reviewed in Technical Evaluation Report attached to 0737, lI.D.1 Appendix EE of SSER15.

Unit 2 Actions: 1) Testing of relief and safety valves;

2) Reanalysis of fluid transient loads for pressurizer relief and safety valve supports and any required modifications;

3) Modifications to pressurizer safety valves, PORVs, PORV block valves and associated piping; and 4) Change motor operated block valves.

NUREG- Valve Position Indication CI The design was reviewed in the original 1982 SER and 0737, lI.D.3 found acceptable pending confirmation of installation of the acoustic monitoring system. In SSER5 (IR 390/84-35), the staff closed the LICENSE CONDITION for Unit 1 only.

Unit 2 Action: Verify installation of the acoustic monitoring system to PORV to indicate position.

NUREG- Auxiliary Feedwater System Evaluation, Cl Reviewed in Appendix EE of SSER16.

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

Page 9 of I11 * = See last page for status code definition.

ITEM TITLE

• ADDITIONAL INFORMATION NUREG- Auxiliary Feedwater System Initiation and CI NRC: IR 50-390/84-20 and 50-391/84-16; letters dated 0737, II.E.1.2 Flow March 29, 1985, and October 31, 1995; SSER 16 Unit 2 Action: Complete procedures and qualification testing.

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

Unit 2 Action: Implement procedures and testing.

NUREG- Accident-Monitoring Instrumentation - Cl Reviewed in SSER9.

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

NUREG- Accident-Monitoring Instrumentation - CI Reviewed in SSER9.

0737, Iodine/Particulate Sampling II.F.1.2.B. Unit 2 Actions: Install Noble gas, Iodine / particulate sampling, and Containment High Range Monitors.

NUREG- Accident-Monitoring Instrumentation - CI Reviewed in SSER9.

0737, Containment High Range Monitoring I1.F.1.2.C. 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 Instrumentation - Cl Reviewed in SSER9.

0737, Containment Pressure II.F.1.2.D. Unit 2 Action: Verify installation of containment pressure indication.

NUREG- Accident-Monitoring Instrumentation - Cl Reviewed in SSER9.

0737, Containment Water Level II.F.1.2.E. Unit 2 Action: Verify installation of containment water level monitors.

NUREG- Accident-Monitoring Instrumentation - CI Reviewed in SSER9.

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

NUREG- Power Supplies For Pressurizer Relief Cl Reviewed in original 1982 SER and 8.3.3 of SSER7.

0737, II.G.1 Valves, Block Valves and Level Indicators Unit 2 Action: Implement modifications such that PORVS and associated Block Valves are powered from same train but different buses.

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

Page 10 of 11 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION NUREG- Auto Trip of RCPS Cl Reviewed in 15.5.4 of original 1982 SER; became License 0737, I1.K.3.5 Condition 35. The staff determined that their review of Item II.K.3.5 did not have to be completed to support the full power license and considered this license condition resolved in SSER4. The item was further reviewed in Appendix EE of SSER16.

Unit 2 Action: Implement modifications as required.

NUREG- PID Controller Cl Reviewed in original 1982 SER.

0737, I1.K.3.9 Unit 2 Action: Set the derivative time constant to zero.

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

I1.K.3.25 Unit 2 Action: Ensure DG power is provided to pump sealing cooling system.

NUREG- Small Break LOCA Methods Cl TVA: letter dated October 29, 1981

0737, II.K.3.30 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 II.K.3.31. This item was further reviewed in Appendix EE of SSER16.

Unit 2 Action: Complete analysis for Unit 2.

NUREG- Plant Specific Analysis Cl The staff determined in SSER4 that their review of Items 0737, II.K.3.30 and II.K.3.31 did not have to be completed to 11.K.3.31 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 SSER1 6.

Unit 2 Action: Complete analysis for Unit 2.

NUREG- In-Plant Iodine Radiation Monitoring Cl NRC reviewed in Appendix EE of SSER16.

0737, II1.D.3.3 Unit 2 Action: Complete modifications for Unit 2.

STATUS CODE DEFINITIONS 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.

Page 11 of 11 * = See last page for status code definition.

GENERIC COMMUNICATIONS STATUS = CLOSED/TECHNICAL SPECIFICATIONS

GENERIC COMMUNICATOONS STATUS = CLOSED/TECHNICAL SPECIFICATIONS ITEM TITLE

• ADDITIONAL INFORMATION B 75-008 PWR Pressure Instrumentation CT NRC: IR 390/391 85-08 Unit 2 Action: Ensure that Technical Specifications and Site Operating Instructions address importance of maintaining temperature and pressure within prescribed limits.

B 77-004 Calculation Error Affecting The Design CT TVA: letter dated January 23, 1978 Performance of a System for Controlling pH of Containment Sump Water Following a NRC: IR 390/78-11 and 391/78-09 LOCA Unit 2 Action: Ensure Technical Specifications includes limit on Boron concentration.

GL 80-014 LWR Primary Coolant System Pressure CT TVA: FSAR 5.2.7.4 Isolation Valves NRC: 1.14.2 of SSER6 NRC reviewed in 1.14.2 of SSER6.

Unit 2 Action: Incorporate guidance into Technical Specifications.

GL 80-077 Refueling Water Level - Technical CT Unit 2 Action: Address in Technical Specifications, as Specifications Changes appropriate.

GL 83-028 "Required Actions Based on Generic CT TVA: letters dated November 7, 1983, January 17, 1986 Implications of Salem ATWS Events: and November 1, 1993 3.1 - Post-Maintenance Testing NRC: letters dated December 10, 1985, October 27, 1986, (Reactor Trip System Components) and July 2, 1990; IR 390, 391/86-04 Unit 2 Action: Test and maintenance procedures and Technical Specifications will include post-maintenance operability testing of safety-related components of the reactor trip system.

GL 83-028 "Required Actions Based on Generic CT TVA: letters dated November 7, 1983, January 17, 1986 Implications of Salem ATWS Events: and November 1, 1993 3.2 - Post-Maintenance Testing (All SR NRC: letters dated December 10, 1985, October 27, 1986, Components) and July 2, 1990; IR 390, 391/86-04 Unit 2 Action: Test and maintenance procedures and Technical Specifications will include post-maintenance operability testing of other (than reactor trip system) safety-related components.

Page 1 of 3 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 83-028 "Required Actions Based on Generic CT TVA: letters dated November 7, 1983, February 10, 1986, Implications of Salem ATWS Events: and May 19, 1986 4.2 - Reactor Trip System Reliability NRC: letters dated July 26, 1985 and June 18, 1992; (Preventive Maintenance and SSER 16 Surveillance Program for Reactor Trip Breakers)

• Unit 2 Action: Ensure maintenance instruction procedure and Technical Specifications support reliable reactor trip breaker operation.

GL 83-028 "Required Actions Based on Generic CT TVA: letters dated November 7, 1983 and July 26, 1985 Implications of Salem ATWS Events:

NRC: letters dated June 28, 1990 and October 9, 1990; 4.5 - Reactor Trip System Reliability SSERs 5 and 16 (Automatic Actuation of Shunt Trip Attachment)

Unit 2 Action: Address in Technical Specifications, as appropriate.

GL 86-009 Technical Resolution of Generic Issue CT N-1 Loop operation was addressed in original 1982 SER

• B-59, (N-i) Loop Operation in BWRs and (4.4.7).

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

GL 90-006 Resolution of Generic Issues 70, "PORV CT NRC letter dated January 9, 1991, accepted TVA's and Block Valve Reliability," and 94, response for both units.

"Additional LTOP Protection for PWRs" Unit 2 Actions: 1) Revise operating instruction and surveillance procedure; and 2) Incorporate testing requirements in the Technical Specifications.

NUREG- Post-Accident Sampling CT NRC reviewed in 9.3.2 of SSER16. TVA submitted a 0737, lI.B.3 TS improvement to eliminate requirements for the Post Accident Sampling System using the 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.

Page 2 of 3 * = See last page for status code definition.

ITEM TITLE

• ADDITIONAL INFORMATION NUREG- Containment Isolation Dependability CT TVA: letters dated October 29, 1981, and 0737, I1.E.4.2 February 25, 1985 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.

NUREG- Reporting SV/RkV Failures/Challenges CT (Action from GL 82-16) - NRC reviewed in Appendix EE of 0737, lI.K.3.3 SSER16.

Unit 2 Action: Include, as necessary, in Technical Specifications submittal.

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

0737, II.K.3.10 Unit 2 Action: Unit 2 Technical Specifications and surveillance procedures will address this issue.

STATUS CODE DEFINITIONS 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.

Page 3 of 3 * = See last page for status code definition.

GENERIC COMMUNICATIONS STATUS = NOT APPLICABLE

GENERIC COMMUNICATIONS STATUS = NOT APPLICABLE ITEM TITLE ADDITIONAL INFORMATION B 71-001 Involves Main Steam Isolation Valves NA Boiling Water Reactor B 71-002 PWR Reactor Trip Circuit Breakers NA Addressed to specific plant(s).

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

Relief Valve Headers

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B 72-001 Failed Hangers for Emergency Core NA Addressed to specific plant(s).

Cooling System Suction Header B 72-002 Simultaneous Actuation of a Safety NA Addressed to specific plant(s).

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

B 73-005 Manufacturing Defect in BWR Control Rods NA Boiling Water Reactor B 73-006 Inadvertent Criticality in a BWR NA Boiling Water Reactor B 74-002 Truck-Strike Possibility NA Info B 74-004 Malfunction of Target Rock Safety Relief NA Boiling Water Reactor Valves

.......... .o..............................................................................................................................

B 74-005 Shipment of an Improperly Shielded Source NA Does not apply to power reactor.

B 74-007 Personnel Exposure - Irradiation Facility NA Does not apply to power reactor.

B 74-010 Failures in 4-Inch Bypass Pipe at Dresden 2 NA Boiling Water Reactor B 74-014 BWR Relief Valve Discharge to NA Boiling Water Reactor Suppression Pool B 75-001 Through-Wall Cracks in Core Spray Piping NA Boiling Water Reactor at Dresden-2 B 75-002 Defective Radionics Radiograph Exposure NA Does not apply to power reactor.

Devices and Source Changers B 75-007 Exothermic Reaction in Radwaste Shipment NA Does not apply to power reactor.

B 76-001 BWR Isolation Condenser Tube Failure NA Boiling Water Reactor B 76-004 Cracks in Cold Worked Piping at BWRs NA Boiling Water Reactor B 76-008 Teletherapy Units NA Does not apply to power reactor.

B 78-003 Potential Explosive Gas Mixture NA Boiling Water Reactor Accumulations Associated with BWR Offgas System Operations B 78-007 Protection Afforded by Air-Line Respirators NA Item was applicable only to units with operating license at and Supplied-Air Hoods the time the item was issued.

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

NRC: IR 390/391 85-08 Page 1 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION B 78-009 BWR Drywell Leakage Paths Associated NA Boiling Water Reactor with Inadequate Drywell Closures B 78-011 Examination of Mark I Containment Torus NA Boiling Water Reactor Welds B 78-013 Failures in Source Heads Kay Ray, Inc. NA Does not apply to power reactor.

Gauges Models 7050, 7050B, 7051, 7051 B, 7060, 7060B, 7061 and 7061B

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B 78-014 Deterioration of Buna-N Components in NA Boiling Water Reactor ASCO Solenoids B 79-005 Nuclear Incident at TMI NA Applies only to Babcock and Wilcox designed plants B 79-008 Events Relevant to BWRs Identified During NA Boiling Water Reactor TMI Incident B 79-010 Requalification Training Program Statistics NA Item was applicable only to units with operating license at the time the item was issued.

B 79-012 Short Period Scrams at BWR Facilities NA Boiling Water Reactor

... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... ... n...... ... ... ... ... ... .... .. . ... ... .

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 Water NA Item was applicable only to units with operating license at Systems at PWR Plants the time the item was issued.

NRC: IR 390/80-06 and 391/80-05; NUREG/ CR 5286 B 79-018 Audibility Problems Encountered on NA Item was applicable only to units with operating license at Evacuation of Personnel from High-Noise the time the item was issued.

Areas NRC: IR 390/80-06 and 391/80-05 B 79-019 Packaging of Low-Level Radioactive Waste NA Item was applicable only to units with operating license at for Transport and Burial the time the item was issued.

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 Low-Level Radioactive Waste the time the item was issued.

NRC: IR 390/80-06 and 391/80-05 B 79-022 Possible Leakage of Tubes of Tritium Gas NA Does not apply to power reactor.

Used in Time Pieces for Luminosity NRC: IR 390/80-06 and 391/80-05 B 79-026 Boron Loss from BWR Control Blades NA - Boiling Water Reactor B 80-001 Operability of ADS Valve Pneumatic Supply NA Boiling Water Reactor Page 2 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION B 80-002 Inadequate QA for Nuclear Supplied NA Boiling Water Reactor Equipment B 80-007 BWR Jet Pump Assembly Failure NA Boiling Water Reactor B 80-013 Cracking in Core Spray Spargers NA Boiling Water Reactor B 80-014 Degradation of Scram Discharge Volume NA Boiling Water Reactor Capability B 80-017 Failure of 76 of 185 Control Rods to Fully NA Boiling Water Reactor Insert During a Scram at a BWR B 80-022 Automation Industries, Model 200-520-008 NA Does not apply to power reactor.

Sealed-Source Connectors B 80-025 Operating Problems with Target Rock NA Boiling Water Reactor Safety-Relief Valves at BWRs B 81-001 Surveillance of Mechanical Snubbers NA NRC: IR 390/391 81-17 B 82-003 Stress Corrosion Cracking in Thick-Wall, NA Boiling Water Reactor Large Diameter, Stainless Steel, Recirculation System Piping at BWR Plants B 83-002 Stress Corrosion Cracking in Large- NA Boiling Water Reactor Diameter Stainless Steel Recirculation System Piping at BWR Plants

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B 83-003 Check Valve Failures in Raw Water Cooling NA Addressed by Inservice Testing for Construction Permit Systems of Diesel Generators holders B 84-001 Cracks in BWR Mark 1 Containment Vent NA Boiling Water Reactor Headers B 86-001 Minimum Flow Logic Problems That Could NA Boiling Water Reactor Disable RHR Pumps B 86-004 Defective Teletherapy Timer That May Not NA Does not apply to power reactor.

Terminate Treatment Dose B 88-006 Actions to be Taken for the Transfer of NA Does not apply to power reactor.

Model No. SPEC 2-T Radiographic Exposure Device B 88-007 Power Oscillations in BWRs NA Boiling Water Reactor B 90-002 Loss of Thermal Margin Caused by NA Boiling Water Reactor Channel Box Bow B 91-001 Reporting Loss of Criticality Safety Controls NA Does not apply to power reactor.

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

Aging Theratronics Teletherapy Units B

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

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

Treatment with Remote Afterloading Devices B 93-002 Debris Plugging of Emergency Core NA Boiling Water Reactor Cooling Suction Strainers Page 3 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION B 93-003 Resolution of Issues Related to Reactor NA Boiling Water Reactor Vessel Water Level Instrumentation in BWRs B 94-001 Potential Fuel Pool Draindown Caused by NA Addressed to holders of licenses for nuclear power reactors Inadequate Maintenance Practices at that are permanently shut down with spent fuel in the spent Dresden Unit 1 fuel pool

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B 94-002 Corrosion Problems in Certain Stainless NA Does not apply to power reactor.

Steel Packagings Used to Transport Uranium Hexafluoride B 95-001 Quality Assurance Program for NA Does not apply to power reactor.

Transportation of Radioactive Material B 95-002 Unexpected Clogging of a Residual Heat NA Boiling Water Reactor Removal Pump Strainer While Operating in Suppression Pool Cooling Mode B 96-003 Potential Plugging of ECCS Suction NA Boiling Water Reactor Strainers by Debris in BWRs B 96-004 Chemical, Galvanic, or Other Reactions in NA Info Spent Fuel Storage and Transportation Casks B 97-001 Potential for Erroneous Calibration, Dose NA Does not apply to power reactor.

Rate, or Radiation Exposure Measurements with Certain Victoreen Model 530 and 531SI Electrometer/Dosemeters B 97-002 Puncture Testing of Shipping Packages NA Does not apply to power reactor.

Under 10 CFR Part 71 B 03-001 Potential Impact of Debris Blockage on NA TVA: letter dated September 7, 2007 Emergency Sump Recirculation at PWRs B 03-003 Potentially Deficient 1-inch Valves for NA Does not apply to power reactor.

Uranium Hexaflouride Cylinders C 76-003 Radiation Exposures in Reactor Cavities NA Info C 76-004 Neutron Monitor and Flow Bypass Switch NA Boiling Water Reactor Malfunctions C 76-006 Stress Corrosion Cracks in Stagnant, Low NA Item was applicable only to units with operating license at Pressure Stainless Piping Containing Boric the time the item was issued.

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

C 77-001 Malfunctions of Limitorque Valve Operators NA Info C 77-002a Potential Heavy Spring Flooding (CP) NA Item was applicable only to units with operating license at the time the 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 Electrical NA Info Cables Page 4 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION C 77-007 Short Period During Reactor Startup NA Boiling Water Reactor *.....................................

C 77-008 Failure of Feedwater Sample Probe NA Item was applicable only to units with operating license at the time the item was issued.

C 77-009 Improper Fuse Coordination in BWR NA Boiling Water Reactor Standby Liquid Control System Control Circuits C 77-010 Vacuum Conditions Resulting in Damage to NA Item was applicable only to units with operating license at Liquid Process Tanks the time the item was issued.

C 77-011 Leakage of Containment Isolation Valves NA Info with Resilient Seats C 77-012 Dropped Fuel Assemblies at BWR Facilities NA Boiling Water Reactor C 77-013 Reactor Safety Signals Negated-During NA Info Testing

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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 EmergencyDiesel Generator C 77-016 Emergency Diesel Generator Electrical Trip NA Info Lock-Out Features C 78-001 Loss of Well Logging Source NA Does not apply to power reactor.

C 78-002 Proper Lubricating Oil for Terry Turbines NA Info ......................................................

C 78-003 Packaging Greater Than Type A Quantities NA Info of Low Specific Activity Radioactive Material for Transport C 78-004 Installation Errors That Could Prevent NA Info Closing of Fire Doors C 78-005 Inadvertent Safety Injection During NA Info Cooldown C 78-006 Potential Common Mode Flooding of ECCS NA Info Equipment Rooms at BWR Facilities C 78-007 Damaged Components of a NA Info Bergen-Paterson Series 25000 Hydraulic Test Stand C 78-008 Environmental Qualification of NA Info Safety-Related Electrical Equipment at Nuclear Power Plants C 78-009 Arcing of General Electric Company NA Info Size 2 Contactors C 78-010 Control of Sealed Sources in Radiation NA Does not apply to power reactor.

Therapy C 78-011 Recirculation MG Set Overspeed Stops NA Boiling Water Reactor C 78-012 HPCI Turbine Control Valve Lift Rod NA Boiling Water Reactor Bending Page 5 of 33 * = See last page for status code definition.

ITEM TITLE

• ADDITIONAL INFORMATION C 78-013 Inoperability of Service Water Pumps NA Info

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C 78-014 HPCI Turbine Reversing Chamber Hold NA Boiling Water Reactor Down Bolting C 78-015 Tilting Disc Check Valves Fail to Close with NA Info Gravity in Vertical Position C 78-016 Limitorque Valve Actuators NA Info C 78-017 Inadequate Guard Training/Qualification NA Info and Falsified Training Records C 78-018 UL Fire Test NA Info C 78-019 Manual Override (Bypass) of Safety System NA Info Actuation Signals C 79-001 Administration of Unauthorized Byproduct NA Does not apply to power reactor.

Material to Humans C 79-002 Failure of 120 Volt Vital AC Power Supplies NA Info C 79-003 Inadequate Guard Training - Qualification NA Info and Falsified Training Records C 79-004 Loose Locking Nut on Limitorque Valve NA Info Operators C 79-005 Moisture Leakage in Stranded Wire NA Info Conductors C 79-006 Failure to Use Syringe and Bottle Shields in NA Does not apply to power reactor.

Nuclear Medicine C 79-007 Unexpected Speed Increase of Reactor NA Boiling Water Reactor Recirculation MG Set Resulted in Reactor Power Increase C 79-008 Attempted Extortion - Low Enriched NA Fuel facilities and operating reactors at the time the item Uranium was issued C 79-009 Occurrences of Split or Punctured NA Info Regulator Diaphragms in Certain Self Contained Breathing Apparatus C 79-010 Pipefittings Manufactured from NA Info Unacceptable Material C 79-011 Design/Construction Interface Problem NA Info C 79-012 Potential Diesel Generator Turbocharger NA Info Problem C 79-013 Replacement of Diesel Fire Pump Starting NA Info Contactors C 79-014 Unauthorized Procurement and Distribution NA Does not apply to power reactor.

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

Certain Self-Contained Breathing Apparatus Page 6 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION 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 Switches on NA Info 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 Ingersoll-Rand NA Info Pumps C 79-020 Failure of GTE Sylvania Relay Type PM NA Info Bulletin 7305 Catalog 5U12-1 1-AC with a 120V AC Coil C 79-021 Prevention of Unplanned Releases of NA Info Radioactivity C 79-022 Stroke Times for Power Operated Relief NA Info Valves m..................................................................................................

C 79-023 Motor Starters and Contactors Failed to NA Info Operate C 79-024 Proper Installation and Calibration of Core NA Boiling Water Reactor Spray Pipe Break Detection Equipment on BWRs C 79-025 Shock Arrestor Strut Assembly Interference NA Info *.............................

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

Implant Therapy Sources C 80-007 Problems with HPCI Turbine Oil System NA Boiling Water Reactor C 80-008 BWR Technical Specification NA Boiling Water Reactor 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

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C 80-011 Emergency Diesel Generator Lube Oil NA Info Cooler Failures C 80-012 Valve-Shaft-to-Actuator Key May Fall Out of NA Info Place when Mounted Below Horizontal Axis Page 7 of 33 *=See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION C 80-013 Grid Strap Damage in Westinghouse Fuel NA Info Assemblies C 80-014 Radioactive Contamination of Plant NA Info Demineralized Water System and Resultant Internal Contamination of Personnel C 80-015 Loss of Reactor Coolant Pump Cooling and NA Info Natural Circulation Cooldown C 80-016 Operational Deficiencies in Rosemount NA Info Model 510DU Trip Units and Model 1152 Pressure Transmitters C 80-017 Fuel Pin Damage Due to Water Jet from NA Info Baffle Plate Corner

. .. . .. . .. ... . .. . .. . ... .. . .. . .. . .. . .. . .. . .. . .. .7 .. ... . .. . .. . .. ... . .. . .. . .. . .. . .. . .. . .. . .. . .. ... . .. ... . .. . .. . .. . .. . .. . .. . .. . .. . .. . .. . .. . .. . .

C 80-018 10 CFR 50.59 Safety Evaluations for NA Info Changes to Radioactive Waste Treatment Systems C 80-019 Noncompliance with License Requirements NA Does not apply to power reactor.

for Medical Licensees C 80-020 Changes in Safe-Slab Tank Dimensions NA Info

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

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C 80-025 Case Histories of Radiography Events NA Does not apply to power reactor.

C 81-001 Design Problems Involving Indicating NA Info Pushbutton Switches Manufactured by Honeywell Incorporated

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C 81-002 Performance of NRC-Licensed Individuals NA Item was applicable only to units with operating license at while on Duty the time the item was issued.

C 81-003 Inoperable Seismic Monitoring NA Info Instrumentation C 81-004 The Role of Shift Technical Advisors and NA Info Importance of Reporting Operational Events C 81-005 Self-Aligning Rod End Bushings for Pipe NA Info Supports C 81-006 Potential Deficiency Affecting Certain NA Info Foxboro 10 to 50 Milliampere Transmitters C 81-007 Control of Radioactively Contaminated NA Info Material C 81-008 Foundation Materials NA Info C 81-009 Containment Effluent Water that Bypasses NA Info Radioactivity Monitor Page 8 of 33 *=See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION C 81-010 Steam Voiding in the Reactor Coolant NA Item was applicable only to units with operating license at System During Decay Heat Removal the time the item was issued.

Cooldown C 81-011 Inadequate Decay Heat Removal During NA Boiling Water Reactor Reactor Shutdown C 81-012 Inadequate Periodic Test Procedure of NA Info PWR Reactor Protection System C 81-013 Torque Switch Electrical Bypass Circuit for NA Info Safeguard Service Valve Motors C 81-014 Main Steam Isolation Valve Failures to NA Info Close C 81-015 Unnecessary Radiation Exposures to the NA Info Public and Workers During Events Involving Thickness and Level Measuring Devices -

GL 77-001 Intrusion Detection Systems Handbook NA Info GL 77-002 Fire Protection Functional Responsibilities NA Info GL 77-003 Transmittal of NUREG-0321, "A Study of NA Info the Nuclear Regulatory Commission Quality Assurance Program" GL 77-004 Shipments of Contaminated Components NA Info From NRC Licensed Power Facilities to Vendors & Service Companies GL 77-005 Nonconformity of Addressees of Items NA Info Directed to the Office of Nuclear Reactor Regulation GL 77-006 Enclosing Questionnaire Related to Steam NA Item was applicable only to units with operating license at Generators the time the item was issued.

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

GL 77-008 Revised Intrusion Detection Handbook and NA Info Entry Control Systems Handbook GL 78-001 Correction to Letter of 12/15/77 [GL 77-07] NA Item was applicable only to units with operating license at the time the item was issued.

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

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

GL 78-005 Internal Distribution of Correspondence - NA Info Asking for Comments on Mass Mailing System GL 78-006 This GL was never issued. NA

..................................................................... m....................................................................

GL 78-007 This GL was never issued. NA GL 78-008 Enclosing NUREG-0408 Re Mark I NA Boiling Water Reactor Containments, and Granting Exemption from GDC 50 and Enclosing Sample Notice Page 9 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 78-009 Multiple-Subsequent Actuations of NA Boiling Water Reactor Safety/Relief Valves Following an Isolation Event GL 78-010 Guidance on Radiological Environmental NA Info Monitoring GL 78-011 Guidance on Spent Fuel Pool Modifications NA Info 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 for NA Info Comment GL 78-015 Request for Information on Control of NA See GL 81-007.

Heavy Loads Near Spent Fuel GL 78-016 Request for Information on Control of NA Info Heavy Loads Near Spent Fuel Pools GL 78-017 Corrected Letter on Heavy Loads Over NA Info Spent Fuel GL 78-018 Corrected Letter on Heavy Loads Over NA Duplicate of GL 81-007 Spent Fuel GL 78-019 Enclosing Sandia Report SAND 77-0777, NA Info "Barrier Technology Handbook" GL 78-020 Enclosing - "A Systematic Approach to the NA Info 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 Systems NA Info and Entry Control Systems Handbooks and Nuclear Safeguards Technology Handbook GL 78-023 Manpower Requirements for Operating NA Info Reactors GL 78-024 Model Appendix I Technical Specifications NA Boiling Water Reactor and Submittal Schedule For BWRs GL 78-025 This GL was never issued. NA GL 78-026 Excessive Control Rod Guide Tube Wear NA Applies only to Babcock and Wilcox designed plants GL 78-027 Forwarding of NUREG-0181 NA Info GL 78-028 Forwarding pages omitted from 07/11/78 NA Boiling Water Reactor 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 10 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 78-031 Notice of Steam Generator Conference NA Info Agenda GL 78-032 Reactor Protection System Power Supplies NA Boiling Water Reactor GL 78-033 Meeting Schedule and Locations For NA Info Upgraded Guard Qualification

..................... m....................................................................................................................

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

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

Specifications, PWR, and NUREG-0133 GL 78-039 Forwarding of 2 Tables of Appendix I, Draft NA Boiling Water Reactor Radiological Effluent Technical Specifications, BWR, and NUREG-0133 m.. .......................................................................................................................................

GL 78-040 Training & Qualification Program NA Info Workshops GL 78-041 Mark II Generic Acceptance Criteria For NA Boiling Water Reactor Lead Plants GL 78-042 Training and Qualification Program NA Info Workshops GL 79-001 Interservice Procedures for Instructional NA Info Systems Development - TRADOC GL 79-002 Transmitting Rev. to Entry Control Systems NA Info Handbook (SAND 77-1033), Intrusion Detection Handbook (SAND 76-0554), and Barrier Penetration Database GL 79-003 Offsite Dose Calculation Manual NA Info GL 79-004 Referencing 4/14/78 Letter - Modifications NA Info to NRC Guidance "Review and Acceptance of Spent Fuel Pool Storage and Handling" GL 79-005 Information Relating to Categorization of NA Info Recent Regulatory Guides by the Regulatory Requirements Review Committee GL 79-006 Contents of the Offsite Dose Calculation NA Info Manual GL 79-007 Seismic (SSE) and LOCA Responses NA Info (NUREG-0484)

GL 79-008 Amendment to 10 CFR 73.55 NA Info GL 79-009 Staff Evaluation of Interim NA Boiling Water Reactor Multiple-Consecutive Safety-Relief Valve Actuations Page 11 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 79-010 Transmitting Regulatory Guide 2.6 for NA Does not apply to power reactor.

Comment GL 79-011 Transmitting "Summary of Operating NA Info Experience with Recalculating Steam Generators, January 1979," NUREG-0523 GL 79-012 ATWS - Enclosing Letter to GE, with NA Info 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 - Enclosing NA Info NUREG-0531 and Notice GL 79-015 Steam Generators - Enclosing Summary NA Info of Operating Experience with Recirculating Steam Generators, NUREG-0523 GL 79-016 Meeting Re Implementation of Physical NA Info Security Requirements GL 79-017 Reliability of Onsite Diesel Generators at NA Info Light Water Reactors GL 79-018 Westinghouse Two-Loop NSSS NA Addressed to specific plant(s).

GL 79-019 NRC Staff Review of Responses to NA Addressed to specific plant(s).

Bs 79-06 and 79-06a GL 79-021 Enclosing NUREG/CR-0660, Enhancement NA Info of on Site Emergency Diesel Generator Reliability" GL 79-022 Enclosing NUREG-0560, "Staff Report on NA Applies only to Babcock and Wilcox designed plants the Generic Assessment of Feedwater Transients in PWRs Designed by B&W" GL 79-023 NRC Staff Review of Responses to B 79-08 NA Boiling Water Reactor GL 79-024 Multiple Equipment Failures in NA Item was applicable only to units with operating license at Safety-Related Systems the time the item was issued.

GL 79-025 Information Required to Review Corporate NA Info Capabilities GL 79-026 Upgraded Standard Technical Specification NA Info Bases Program GL 79-027 Operability Testing of Relief and Safety NA Boiling Water Reactor Relief Valves ...............................

GL 79-028 Evaluation of Semi-Scale Small Break NA Info Experiment GL 79-029 Transmitting NUREG-0473, Revision 2, NA Info Draft Radiological Effluent Technical Specifications GL 79-030 Transmitting NUREG-0472, Revision 2, NA Info Draft Radiological Technical Specifications Page 12 of 33 * = See last page for status code definition.

ITEM TITLE

• ADDITIONAL INFORMATION GL 79-031 Submittal of Copies of Response to 6/29/79 NA Info NRC Request [79-25]

GL 79-032 Transmitting NUREG-0578, "TMI-2 Lessons NA Info Learned" GL 79-033 Transmitting NUREG-0576, "Security NA Info 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 Impacts on NA Info Emergency Planning GL 79-037 Amendment to 10 CFR 73.55 Deferral from NA Info 8/1/79 to 11/1/79 GL 79-038 BWR Off-Gas Systems - Enclosing NA Boiling Water Reactor NUREG/CR-0727 GL 79-039 Transmitting Division 5 Draft Regulatory NA Does not apply to power reactor.

Guide and Value Impact Statement

.................................... -.......................... m..........................................................................

GL 79-040 Follow-up Actions Resulting from the NRC NA Item was applicable only to units with operating license at Staff Reviews Regarding the TMI-2 Accident the time the item was issued.

GL 79-041 Compliance with 40 CFR 190, EPA NA Info Uranium Fuel Cycle Standard GL 79-042 Potentially Unreviewed Safety Question on NA Item was applicable only to units with operating license at Interaction Between Non-Safety Grade the time the item was issued.

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

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

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

Operability GL 79-047 Radiation Training NA Info GL 79-048 Confirmatory Requirements Relating to NA Boiling Water Reactor Condensation Oscillation Loads for the Mark I Containment Long Term Program GL 79-049 Summary of Meetings Held on 9/18-20/79 NA Info to Discuss Potential Unreviewed Safety Question on Systems Interaction for B&W PI ... ... ... ... ... ... ... ... ... ... ... ... ... .. .. .. .. .. .. ...

GL 79-050 Emergency Plans Submittal Dates NA Info GL 79-051 Follow-up Actions Resulting from the NRC NA Info Staff Reviews Regarding the TMI-2 Accident GL 79-052 Radioactive Release at North Anna Unit 1 NA Item was applicable only to units with operating license at and Lessons Learned the time the item was issued.

GL 79-053 ATWS NA Info Page 13 of 33 P = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 79-054 Containment Purging and Venting During NA Addressed to specific plant(s).

Normal Operation 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 GL 79-056 Discussion of Lessons Learned Short Term NA Item was applicable only to units with operating license at Requirements the time the item was issued.

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

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

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

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

GL 79-067 Estimates for Evacuation of Various Areas NA Info Around Nuclear Power Reactors GL 79-068 Audit of Small Break LOCA Guidelines NA Info GL 79-069 Cladding Rupture, Swelling, and Coolant NA Info Blockage as a Result of a Reactor Accident GL 79-070 Environmental Monitoring for Direct NA Info Radiation GL 80-001 NUREG-0630, "Cladding, Swelling and NA Info Rupture - Models For LOCA Analysis" 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 Qualification of NA Info Class 1E Equipment" Page 14 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION 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 Assurance for NA Boiling Water Reactor Nuclear Supplied Equipment" GL 80-009 Low Level Radioactive Waste Disposal NA Item was applicable only to units with operating license at the time the 'item was issued.

GL 80-010 Issuance of NUREG-0588, "Interim Staff NA Info Position On Equipment Qualifications of Safety-Related Electrical Equipment" GL 80-011 B 80-03, "Loss of Charcoal From Standard NA Info Type II, 2 Inch, Tray Absorber Cells" GL 80-012 B 80-04, "Analysis of a PWR Main Steam NA Info Line Break With Continued Feedwater Addition" GL 80-013 Qualification of Safety Related Electrical NA Item was applicable only to units with operating license at Equipment the time the item was issued.

GL 80-015 Request for Additional Management and NA Info Technical Resources Information GL 80-016 B 79-01b, "Environmental Qualification of NA Info Class 1E Equipment" GL 80-017 Modifications to BWR Control Rod Drive NA Boiling Water Reactor Systems GL 80-018 Crystal River 3 Reactor Trip From NA Applies only to Babcock and Wilcox designed plants Approximately 100% Full Power GL 80-019 Resolution of Enhanced Fission Gas NA Info Release Concern GL 80-020 Actions Required From OL Applicants of NA Info NSSS Designs by W and CE Resulting From NRC B&O Task Force Review of TMI2 Accident GL 80-022 Transmittal of NUREG-0654, "Criteria For NA Info Preparation and Evaluation of Radiological Emergency Response Plan" GL 80-023 Change of Submittal Date For Evaluation NA Info Time Estimates GL 80-024 Transmittal of Information on NRC "Nuclear NA Info Data Link Specifications" GL 80-025 B 80-06, "Engineering Safety Feature (ESF) NA Info Reset Controls" GL 80-026 Qualifications of Reactor Operators NA Info GL 80-027 B 80-07, "BWR Jet Pump Assembly Failure" NA Boiling Water Reactor GL 80-028 B 80-08, "Examination of Containment NA Info Liner Penetration Welds" Page 15 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 80-029 Modifications to Boiling Water Reactor NA Boiling Water Reactor Control Rod Drive Systems GL 80-030 Clarification of The Term "Operable" As It NA Item was applicable only to units with operating license at Applies to Single Failure Criterion For the time the item was issued.

Safety Systems Required by TS GL 80-031 B 80-09, "Hydramotor Actuator Deficiencies" NA Info GL 80-032 Information Request on Category I Masonry NA Addressed by B 80-11.

Walls Employed by Plants Under CP and OL Review GL 80-033 Actions Required From OL Applicants of NA Applies only to Babcock and Wilcox designed plants B&W Designed NSSS Resulting From NRC B&O Task Force Review of TMI2 Accident GL 80-034 Clarification of NRC Requirements for NA Info Emergency Response Facilities at Each Site GL 80-035 Effect of a DC Power Supply Failure on NA Boiling Water Reactor 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 Requirements to Operating Reactors the time the item was issued.

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

Physical Protection Requirements GL 80-039 B 80-11, "Masonry Wall Design" NA Info GL 80-040 Transmittal of NUREG-0654, "Report of the NA Info 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 System NA Info Operability" GL 80-043 B 80-13, "Cracking In Core Spray Spargers" NA Boiling Water Reactor 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.

GL 80-048 Revision to 5/19/80 Letter On Fire NA Item was applicable only to units with operating license at Protection [GL 80-45] the time the item was issued.

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

Page 16 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 80-052 Five Additional TMI-2 Related NA Item was applicable only to units with operating license at Requirements - Erata Sheets to 5/7/80 the time the item was issued.

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. ,.o GL 80-054 B 80-14, "Degradation of Scram Discharge NA Boiling Water Reactor Volume Capability" GL 80-055 B 80-15, "Possible Loss of Hotline With NA Info Loss of off-Site Power" GL 80-056 Commission Memorandum and Order on NA Info Equipment Qualification GL 80-057 Further Commission Guidance For Power NA Info Reactor Operating Licenses NUREG-0660 and NUREG-0694 GL 80-058 B 80-16, "Potential Misapplication of NA Info Rosemount Inc. Models 1151/1152 Pressure Transmitters With "A" Or "D" Output Codes" GL 80-059 Transmittal of Federal Register Notice RE NA Info 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 to Insert NA Boiling Water Reactor During a Scram at a BWR" GL 80-064 Scram Discharge Volume Designs NA Boiling Water Reactor GL 80-065 Request for Estimated Construction NA Info Completion and Fuel Load Schedules GL 80-066 B 80-17, Supplement 1, "Failure of Control NA Boiling Water Reactor Rods to Insert During a Scram at a BWR" GL 80-067 Scram Discharge Volume NA Boiling Water Reactor GL 80-068 B 80-17, Supplement 2, "Failures Revealed NA Boiling Water Reactor by Testing Subsequent to Failure of Control Rods to Insert During a Scram at a BWR" GL 80-069 B 80-18, "Maintenance of Adequate NA Info Minimum Flow Through Centrifugal Charging Pumps Following Secondary Side HELB" GL 80-070 B 80-19, "Failures of Mercury-Wetted Matrix NA Info Relays in RPS of Operating Nuclear Power Plants Designed by GE" GL 80-071 B 80-20, "Failures of Westinghouse Type NA Info W-2 Spring Return to Neutral Control Switches" Page 17 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 80-072 Interim Criteria For Shift Staffing NA Info GL 80-073 "Functional Criteria For Emergency NA Info Response Facilities," NUREG-0696 GL 80-074 Notice of Forthcoming Meeting With NA Info 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 With GE to NA Info Discussed Proposed BWR Feedwater Nozzle Leakage Detection System GL 80-078 Mark I Containment Long-Term Program NA Boiling Water Reactor GL 80-079 B 80-17, Supplement 3, "Failures Revealed NA Boiling Water Reactor by Testing Subsequent to Failure of Control Rods to Insert During a Scram At a BWR" GL 80-080 Preliminary Clarification of TMI Action Plan NA Info Requirements GL 80-081 Preliminary Clarification of TMI Action Plan NA Info Requirements - Addendum to 9/5/80 Letter [GL 80-80]

GL 80-082 B 79-01b, Supplement 2, "Environmental NA Info Qualification of Class IE Equipment" GL 80-083 Environmental Qualification of NA Info Safety-Related Equipment m...............................................

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" GL 80-086 Notice of Meeting to Discuss Final NA Info Resolution of USI A-12 GL 80-087 Notice of Meeting to Discuss Status of NA Info EPRI-Proposed Resolution of the USI A-12 Fracture Toughness Issue GL 80-088 Seismic Qualification of Auxiliary Feedwater NA Item was applicable only to units with operating license at Systems the time the item was issued.

GL 80-089 B.79-01 b, Supplement 3, "Environmental NA Info Qualification of Class 1E Equipment" GL 80-091 ODYN Code Calculation NA Boiling Water Reactor GL 80-092 B 80-21, "Valve Yokes Supplied by Malcolm NA Info Foundry Company, Inc."

GL 80-093 Emergency Preparedness NA Does not apply to power reactor.

GL 80-094 Emergency Plan NA Info Page 18 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION 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"

...................................................... U...................................................................................

GL 80-096 Fire Protection NA Addressed to specific plant(s).

GL 80-097 B 80-23, "Failures of Solenoid Valves NA Info Manufactured by Valcor Engineering Corporation" GL 80-098 B 80-24, "Prevention of Damage Due to NA Info Water Leakage Inside Containment" GL 80-099 Technical Specifications Revisions For NA Info Snubber Surveillance GL 80-100 Appendix R to 10 CFR 50 Regarding Fire NA Item was applicable only to units with operating license at Protection - Federal Register Notice the time the item was issued.

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

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

GL 80-103 Fire Protection - Revised Federal Register NA Info Notice GL 80-104 Orders On Environmental Qualification of NA Info Safety Related Electrical Equipment GL 80-105 Implementation of Guidance For USI A-12, NA Info "Potential.For Low Fracture toughness and Lamellar Tearing On Component Supports" GL 80-106 Report On ECCS Cladding Models, NA Info NUREG-0630 GL 80-107 BWR Scram Discharge System NA Boiling Water Reactor S............................................................................................................. . ... ...... ..

GL 80-108 Emergency Planning NA Info GL 80-109 Guidelines For SEP Soil Structure NA Info Interaction Reviews 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 Control NA Boiling Water Reactor Rods to Insert During a Scram at a BWR" GL 80-112 B 80-25, "Operating Problems With Target NA Info Rock Safety Relief Valves" GL 81-001 Qualification of Inspection, Examination, NA Info 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" Page 19 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 81-005 Information Regarding The Program For NA Info Environmental Qualification of Safety-Related Electrical Equipment GL 81-006 Periodic Updating of Final Safety Analysis NA Info Reports (FSARS)

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 Control Rod NA Boiling Water Reactor 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 8X8R Fuel NA Boiling Water Reactor Reload Applications For Appendix D Submittals of The GE topical Report GL 81-015 Environmental Qualification of Class 1E NA Info Electrical Equipment - Clarification of Staffs Handling of Proprietary Information GL 81-016 NUREG-0737, Item I.C.1 SER on Abnormal NA - Applies only to Babcock and Wilcox designed plants 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 Pressure Vessels NA Item was applicable only to units with operating license at the time the item was issued.

GL 81-020 Safety Concerns Associated With Pipe NA Boiling Water Reactor Breaks in the BWR Scram System 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 Reports NA Info GL 81-024 Multi-Plant Issue B-56, "Control Rods Fail NA Boiling Water Reactor to Fully Insert" GL 81-025 Change in Implementing Schedule For NA Info Submission and Evaluation of Upgraded Emergency Plans GL 81-026 Licensing Requirements for Pending NA Applicants with pending Construction Permits Construction Permit and Manufacturing License Applications GL 81-027 Privacy and Proprietary Material in NA Info Emergency Plans GL 81-028 Steam Generator Overfill NA Info Page 20 of 33 * = See last page for status code definition.,

ITEM TITLE ADDITIONAL INFORMATION GL 81-029 Simulator Examinations NA Info GL 81-030 Safety Concerns Associated With Pipe NA Boiling Water Reactor Breaks in the BWR Scram System GL 81-031 This GL was never issued. NA GL 81-032 NUREG-0737, Item II.K.3.44, "Evaluation of NA Boiling Water Reactor Anticipated Transients Combined With Single Failure" GL 81-033 This GL was never issued. NA GL 81-034 Safety Concerns Associated With Pipe NA Boiling Water Reactor Breaks in the BWR Scram System GL 81-035 Safety Concerns Associated With Pipe NA Boiling Water Reactor Breaks in the BWR Scram System GL 81-036 Revised Schedule for Completion of TMI NA Info Action Plan Item Il.D.1, "Relief and Safety Valve Testing" GL 81-037 ODYN Code Reanalysis Requirements NA Boiling Water Reactor GL 81-038 Storage of Low Level Radioactive Wastes NA Info 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 GL 82-005 Post-TMI Requirements NA Item was applicable only to units with operating license at the time the item was issued.

GL 82-006 This GL was never issued. NA GL 82-007 Transmittal of NUREG-0909 Relative to the NA Boiling Water Reactor Ginna Tube Rupture GL 82-008 Transmittal of NUREG-0909 Relative to the NA Info Ginna Tube Rupture GL 82-009 Environmental Qualification of Safety NA Info Related Electrical Equipment 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 Relative to the NA Info Restart of R. E. Ginna Nuclear Power Plant GL 82-012 Nuclear Power Plant Staff Working Hours NA Info GL 82-013 Reactor Operator and Senior Reactor NA Info Operator Examinations GL 82-014 Submittal of Documents to the NRC NA Info Page 21 of 33 * = See last page for status code definition.

ITEM TITLE

• ADDITIONAL INFORMATION GL 82-015 This GL was never issued. NA GL 82-016 NUREG-0737 Technical Specifications NA Item was applicable only to units with operating license at the time the item was issued.

GL 82-017 Inconsistency of Requirements Between NA Info 50.54(T) and 50.15 GL 82-018 Reactor Operator and Senior Reactor NA Info Operator Requalification Examinations GL 82-019 Submittal of Copies of Documentation to NA Info NRC - Copy Requirements for Emergency Plans and Physical Security Plans GL 82-020 Guidance for Implementing the Standard NA Info Review Plan Rule

.......................................... m...............................................................................................

GL 82-021 Fire Protection Audits NA Info GL 82-022 Congressional Request for Information NA Item was applicable only to units with operating license at Concerning Steam Generator Tube Integrity the time the item was issued.

GL 82-023 Inconsistency Between Requirements of NA Info 10CFR 73.40(d) and Standard Technical Specifications For Performing Audits of Safeguards Contingency Plans GL 82-024 Safety Relief Valve Quencher Loads: BWR NA Boiling Water Reactor MARK II and III Containments GL 82-025 Integrated IAEA Exercise for Physical NA Item was applicable only to units with operating license at Inventory at LWRS the time the item was issued.

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

GL 82-027 Transmittal of NUREG-0763, "Guidelines NA Boiling Water Reactor For Confirmatory In-Plant Tests of Safety-Relief Valve Discharge for BWR Plants" GL 82-029 This GL was never issued. NA GL 82-030 Filings Related to 10 CFR 50 Production NA Info and Utilization Facilities GL 82-031 This GL was never issued. NA GL 82-032 Draft SteamGenerator Report (SAI) NA Item was applicable only to units with operating license at the time the item was issued.

GL 82-034 This GL was never issued. NA GL 82-036 This GL was never issued. NA GL 82-037 This GL was never issued. NA GIL 82-036 This GL was never issued. NA GL 82-038 Meeting to Discuss Developments for NA Info Operator Licensing Examinations GL 82-039 Problems With Submittals of Subsequent NA Info Information of CURT 73.21 For Licensing Reviews Page 22 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 83-001 Operator Licensing Examination Site Visit NA Info GL 83-002 NUREG-0737 Technical Specifications NA Boiling Water Reactor 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

.... *.....................................................n................................................................................

GL 83-006 Certificates and Revised Format For NA Info Reactor Operator and Senior Reactor Operator Licenses GL 83-007 The Nuclear Waste Policy Act of 1982 NA Info GL 83-008 Modification of Vacuum Breakers on Mark I NA Boiling Water Reactor Containments GL 83-009 Review of Combustion Engineering NA Applies only to Combustion Engineering designed plants Owners' Group Emergency Procedures Guideline Program GL 83-010a Resolution of TMI Action Item I1.K.3.5., NA Applies only to Combustion Engineering designed plants "Automatic Trip of Reactor Coolant Pumps" GL 83-010b Resolution of TMI Action Item I1.K.3.5., NA Applies only to Combustion Engineering designed plants "Automatic Trip of Reactor Coolant Pumps" GL 83-01Od Resolution of TMI Action Item I1.K.3.5., NA Item was applicable only to units with operating license at "Automatic Trip of Reactor Coolant Pumps" the time the item was issued.

GL 83-01Oe Resolution of TMI Action Item II.K.3.5., NA Applies only to Babcock and Wilcox designed plants "Automatic Trip of Reactor Coolant Pumps" GL 83-01Of Resolution of TMI Action Item II.K.3.5., NA Applies only to Babcock and Wilcox designed plants "Automatic Trip of Reactor Coolant Pumps" GL 83-011 Licensee Qualification for Performing NA Item was applicable only to units with operating license at Safety Analyses in Support of Licensing 'the time the item was issued.

Actions GL 83-012 Issuance of NRC FORM 398 - Personal NA Info Qualifications Statement - Licensee GL 83-013 Clarification of Surveillance Requirements NA Info for HEPA Filters and Charcoal Absorber Units In Standard Technical Specifications on ESF Cleanup Systems GL 83-014 Definition of "Key Maintenance Personnel," NA Info (Clarification of Generic Letter 82-12)

GL 83-015 Implementation of Regulatory Guide 1.150, NA Info "Ultrasonic Testing of Reactor Vessel Welds During Preservice & Inservice Examinations, Revision 1" GL 83-016 Transmittal of NUREG-0977 Relative to the NA Info ATWS Events at Salem Generating Station, Unit No.1 Page 23 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 83-016a Transmittal of NUREG-0977 Relative to the NA Info ATWS Events at Salem Generating Station, Unit No.1 GL 83-017 Integrity of Requalification Examinations for NA Info Renewal of Reactor Operator and Senior Reactor Operator Licenses GL 83-018 NRC Staff Review of the BWR Owners' NA Boiling Water Reactor Group (BWROG) Control Room Survey Program GL 83-019 New Procedures for Providing Public Notice NA Item was applicable only to units with operating license at Concerning Issuance of Amendments to the time the item was issued.

Operating Licenses GL 83-020 Integrated Scheduling for Implementation of NA Info Plant Modifications GL 83-021 Clarification of Access Control Procedures NA Info for Law Enforcement Visits GL 83-022 Safety Evaluation of "Emergency Response NA Info 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 GL 83-025 This GL was never issued. NA GL 83-026 Clarification Of Surveillance Requirements NA Info For Diesel Fuel Impurity Level Tests GL 83-027 Surveillance Intervals in Standard NA Info Technical Specifications 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 Transient NA Applies only to Babcock and Wilcox designed plants Operating Guidelines" GL 83-032 NRC Staff Recommendations Regarding NA Info Operator Action for Reactor Trip and ATWS

................................................................................................. s........................................

GL 83-033 NRC Positions on Certain Requirements of NA Info Appendix R to 10 CFR 50 GL 83-034 This GL was never issued. NA GL 83-035 Clarification of TMI Action Plan Item NA Info I1.K.3.31 ... ... ... ... ... ... ... ... ... ...

GL 83-036 NUREG-0737 Technical Specifications NA Boiling Water Reactor GL 83-037 NUREG-0737 Technical Specifications NA Item was applicable only to units with operating license at the time the item was issued.

.... *o..................................................... ,o...............................................................................

GL 83-038 NUREG-0965, "NRC Inventory of Dams" NA Info Page 24 of 33 * = See last page for status code definition.

ITEM TITLE

• ADDITIONAL INFORMATION GL 83-039 Voluntary Survey of Licensed Operators NA Info GL 83-040 Operator Licensing Examination NA Info GL 83-041 Fast Cold Starts of Diesel Generators NA Item was applicable only to units with operating license at the time the item was issued.

GL 83-042 Clarification to GL 81-07 Regarding NA Info Response to NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants" GL 83-043 Reporting Requirements of 10 CFR 50, NA Info Sections 50.72 and 50.73, and Standard Technical Specifications GL 83-044 Availability of NUREG-1021, "Operator NA Info Licensing Examiner Standards" GL 84-001 NRC Use Of The Terms "Important To NA Info Safety" and "Safety Related" GL 84-002 Notice of Meeting Regarding Facility NA Info Staffing GL 84-003 Availability of NUREG-0933, "A NA Info Prioritization of Generic Safety Issues" GL 84-004 Safety Evaluation of Westinghouse Topical NA Info Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops GL 84-005 Change to NUREG-1021, "Operator NA Info Licensing Examiner Standards" GL 84-006 Operator and Senior Operator License NA Does not apply to power reactor.

Examination Criteria For Passing Grade GL 84-007 Procedural Guidance for Pipe Replacement NA Boiling Water Reactor at BWRs GL 84-008 Interim Procedures for NRC Management NA Info of Plant-Specific Backfitting GL 84-009 Recombiner Capability Requirements of 10 NA Boiling Water Reactor CFR 50.44(c)(3)(ii)

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

GL 84-013 Technical Specification for Snubbers NA Info GL 84-014 Replacement and Requalification Training NA Info Program GL 84-015 Proposed Staff Actions to Improve and NA Info Maintain Diesel Generator Reliability Page 25 of 33 = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 84-016 Adequacy of On-Shift Operating Experience NA Info for Near Term Operating License Applicants GL 84-017 Annual Meeting to Discuss Recent NA Info Developments Regarding Operator Training, Qualifications, and Examinations GL 84-018 Filing of Applications for Licenses and NA Does not apply to power reactor.

Amendments GL 84-019 Availability of Supplement 1 to NA Info NUREG-0933, "A Prioritization of Generic Safety Issues" GL 84-020 Scheduling Guidance for Licensee NA Info Submittals of Reloads That Involve Unreviewed Safety Questions GL 84-021 Long Term Low Power Operation in NA Info 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 85-001 Fire Protection Policy Steering Committee NA Only issued as draft Report GL 85-003 Clarification of Equivalent Control Capacity NA Boiling Water Reactor 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.

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

GL 85-008 10 CFR 20.408 Termination Reports - NA Info Format GL 85-009 Technical Specifications For Generic Letter NA Info 83-28, Item 4.3 GL 85-010 Technical Specification For Generic Letter NA Applies only to Babcock and Wilcox designed plants 83-28, Items 4.3 and 4.4 GL 85-013 Transmittal Of NUREG-1154 Regarding NA Info The Davis-Besse Loss Of Main And Auxiliary Feedwater Event GL 85-014 Commercial Storage At Power Reactor NA Item was applicable only to units with operating license at Sites Of Low Level Radioactive Waste Not the time the item was issued.

Generated By The Utility GL 85-015 Information On Deadlines For NA Item was applicable only to units with operating license at 10 CFR 50.49, "Environmental Qualification the time the item was issued.

Of Electric Equipment Important To Safety At Nuclear Power Plants" GL 85-016 High Boron Concentrations NA Info Page 26 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 85-017 Availability Of Supplements 2 and 3 To NA Info NUREG-0933, "A Prioritization Of Generic Safety Issues" GL 85-018 Operator Licensing Examinations NA Info GL 85-019 Reporting Requirements On Primary NA Info Coolant Iodine Spikes GL 85-020 Resolution Of Generic Issue 69: High NA Applies only to Babcock and Wilcox designed plants 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 GL 86-001 Safety Concerns Associated With Pipe NA Boiling Water Reactor Breaks In The BWR Scram System GL 86-002 Technical Resolution of Generic Issue NA Boiling Water Reactor B Thermal Hydraulic Stability GL 86-003 Applications For License Amendments NA Info GL 86-004 Policy Statement On Engineering Expertise NA Info On Shift GL 86-005 Implementation Of TMI Action Item II.K.3.5, NA Applies only to Babcock and Wilcox designed plants "Automatic Trip Of Reactor Coolant Pumps" GL 86-006 Implementation Of TMI Action Item II.K.3.5, NA Applies only to Combustion Engineering designed plants "Automatic Trip of Reactor Coolant Pumps" GL 86-007 Transmittal of NUREG-1 190 Regarding The NA Info 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 86-010 Implementation of Fire Protection NA Info Requirements GL 86-010, Fire Endurance Test Acceptance Criteria NA Info

$1 for Fire Barrier Systems Used to Separate Redundant Safe Shutdown Trains Within the Same Fire Area GL 86-011 Distribution of Products Irradiated in NA Does not apply to power reactor.

Research GL 86-012 Criteria for Unique Purpose Exemption NA Does not apply to power reactor.

From Conversion From The Use of Heu Fuel GL 86-013 Potential Inconsistency Between Plant NA Applies only to Babcock and Wilcox and Combustion Safety Analyses and Technical Engineering designed plants Specifications

.......................................... m..............................................................................................

GIL 86-014 Operator Licensing Examinations NA Info Page 27 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 86-015 Information Relating To Compliance With NA Info 10 CFR 50.49, "Environmental Qualification of Electric Equipment Important To Safety For Nuclear Power Plants" m.............................................................................................................................

GL 86-016 Westinghouse ECCS Evaluation Models NA Info GL 86-017 Availability of NUREG-1169, "Technical NA Boiling Water Reactor Findings Related to Generic Issue C-8, BWR MSIC Leakage And Treatment Methods" GL 87-001 Public Availability Of The NRC Operator NA Info Licensing Examination Question Bank GL 87-002 & Verification of Seismic Adequacy of NA Item was applicable only to units with operating license at 003 Mechanical and Electrical Equipment in the time the item was issued.

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

Temporary Workers GL 87-005 Request for Additional Information on NA Boiling Water Reactor Assessment of License Measures to Mitigate and/or Identify Potential Degradation of Mark I Drywells GL 87-006 Periodic Verification of Leak Tight Integrity NA Item was applicable only to units with operating license at of Pressure Isolation Valves the time the item was issued.

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

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

Checks GL 87-011 Relaxation in Arbitrary Intermediate Pipe NA Info Rupture Requirements GL 87-013 Integrity of Requalification Examinations At NA Does not apply to power reactor.

Non-Power Reactors GL 87-014 Operator Licensing Examinations NA Info GL 87-015 Policy Statement On Deferred Plants NA Info GL 87-016 Transmittal of NUREG-1262, "Answers To NA Info Questions On Implementation of 10 CFR 55 On Operators' Licenses" GL 88-001 NRC Position on IGSCC in BWR Austenitic NA Boiling Water Reactor Stainless Steel Piping GL 88-002 Integrated Safety Assessment Program II NA Item was applicable only to units with operating license at the time the item was issued.

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Page 28 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 88-004 Distribution of Gems Irradiated in Research NA Does not apply to power reactor.

Reactors GL 88-006 Removal of Organization Charts from NA Info Technical Specification Administrative Control Requirements GL 88-008 Mail Sent or Delivered to the Office of NA Info Nuclear Reactor Regulation GL 88-009 Pilot Testing of Fundamentals Examination - NA Boiling Water Reactor GL 88-010 Purchase of GSA Approved Security NA Info Containers GL 88-012 Removal of Fire Protection Requirements NA Info from Technical Specification GL 88-013 Operator Licensing Examinations NA Info GL 88-015 Electric Power Systems - Inadequate NA Info Control Over Design Process GL 88-016 Removal of Cycle-Specific, Parameter NA Info Limits from ... ...

Technical... ...

Specifications GL 88-018 Plant Record Storage on Optical Disks NA Info GL 88-019 Use of Deadly Force by Licensee Guards to NA Does not apply to power reactor.

Prevent Theft of Special Nuclear Material GL 89-001 Implementation of Programmatic and NA Info Procedural Controls for Radiological Effluent Technical Specifications GL 89-002 Actions to Improve the Detection of NA Info Counterfeit and Fraudulently Marketed Products GL 89-003 Operator Licensing Examination.Schedule NA Info GL 89-005 Pilot Testing of the Fundamentals NA Info Examination GL 89-009 ASME Section III Component Replacements NA Item was applicable only to units with operating license at the time the item was issued.

GL 89-011 Resolution of Generic Issue 101, "Boiling NA Boiling Water Reactor Water Reactor Water Level Redundancy" GL 89-012. Operator Licensing Examination NA Info GL 89-014 Line-Item Improvements in Technical NA Info-Specifications - Removal of 3.25 Limit on Extending Surveillance Intervals GL 89-015 Emergency Response Data System NA Info GL 89-016 Installation of a Hardened Wetwell Vent NA Boiling Water Reactor GL 89-017 Planned Administrative Changes to the NA Info NRC Operator Licensing Written Examination Process Page 29 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 89-018 Resolution of Unresolved Safety Issues NA Info A-17, "Systems Interactions in Nuclear Power Plants" GL 89-020 Protected Area Long-Term Housekeeping NA Does not apply to power reactor.

GL 89-021 Request for Information Concerning Status NA Info of Implementation of Unresolved Safety Issue (USI) Requirements GL 89-023 NRC Staff Responses to Questions NA Info Pertaining to Implementation of 10 CFR Part 26 _:........................................................................................

GL 90-001 Request for Voluntary Participation in NRC NA Info 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 Generic NA Info Letter 83-28, Item 2.2 Part 2 "Vendor Interface for Safety-Related Components"

............................................................. n;............................................ m................................

GL 90-005 Guidance for Performing Temporary NA Info Non-Code Repair of ASME Code Class 1,2, and 3 Piping GL 90-007 Operator Licensing National Examination NA Info Schedule GL 90-008 Simulation Facility Exemptions NA Info GL 90-009 Alternative Requirements for Snubber NA Info 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 GL 91-002 Reporting Mishaps Involving LLW Forms NA Item was applicable only to units with operating license at Prepared for Disposal the time the item was issued.

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

Supplies," Pursuant to 10 CFR 50.54(f)

GL 91-007 GI-23, "Reactor Coolant Pump Seal NA Info Failures" and Its Possible Effect on Station Blackout GL 91-008 Removal of Component Lists from NA Info Technical Specifications Page 30 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 91-009 Modification of Surveillance Interval for the NA Boiling Water Reactor Electrical Protective Assemblies in Power Supplies for the Reactor Protection System GL 91-010 Explosives Searches at Protected Area NA Does not apply to poWer reactor.

Portals o..........................................................................................

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

49, "Interlocks and LCOs for Class 1 E Tie Breakers," Pursuant to 10 CFR 50.54 GL 91-012 Operator Licensing National Examination NA Info 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 Report, NA Info Solenoid-Operated Valve Problems at U.S.

Reactors GL 91-016 Licensed Operators' and Other Nuclear NA Info 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 Regarding Two NA GL 91-18 has been superseded by RIS 2005-20.

NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability GL 91-019 Information to Addressees Regarding New NA Info Telephone Numbers for NRC Offices Located in One White Flint North 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 Licensing NA Info Basis: Request for Voluntary Participation in Pilot Program GL 92-004 Resolution of the Issues Related to Reactor NA Boiling Water Reactor Vessel Water Level Instrumentation in BWRs ...

Pursuant to 10 CFR 50.54(f)

........................................................................................................... 7*--- --- -- --- -- -

GL 92-005 NRC Workshop on the Systematic NA Info Assessment of Licensee Performance (SALP) Program GL 92-006 Operator Licensing National Examination NA Info Schedule GL 92-007 Office of Nuclear Reactor Regulation NA Info Reorganization Page 31 of 33 * = See last page for status code definition.

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

Program GL 93-002 NRC Public Workshop on Commercial NA Info Grade Procurement and Dedication GL 93-003 Verification of Plant Records NA Info 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 Safety Issue NA Info 106, "Piping and the Use of Highly Combustible Gases in Vital Areas" GL 93-007 Modification of the Technical Specification NA Item was applicable only to units with operating license at Administrative Control Requirements for the time the item was issued.

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

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

Emergency Diesel Generators GL 94-002 Long-Term Solutions and Upgrade of NA Boiling Water Reactor Interim Operating Recommendations for Thermal-Hydraulic Instabilities in BWRs GL 94-003 IGSCC of Core Shrouds in BWRs NA Boiling Water Reactor GL 94-004 Voluntary Reporting of Additional NA Info Occupational Radiation Exposure Data GL 95-001 NRC Staff Technical Position on Fire NA Does not apply to power reactor.

Protection for Fuel Cycle Facilities GL 95-002 Use of NUMARC/EPRI Report TR-102348, NA Info "Guideline on Licensing Digital Upgrades,"

in Determining the Acceptability of Performing Analog-to-Digital Replacements under 10 CFR 50.59 GL 95-004 Final Disposition of the Systematic NA Info Evaluation Program Lessons-Learned Issues GL 95-006 Changes in the Operator Licensing Program NA Info GL 95-008 10 CFR 50.54(p) Process for Changes to NA Info Security Plans Without Prior NRC Approval GL 95-009 Monitoring and Training of Shippers and NA Info Carriers of Radioactive Materials GL 95-010 Relocation of Selected Technical NA Info Specifications Requirements Related to Instrumentation Page 32 of 33 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION GL 96-002 Reconsideration of Nuclear Power Plant NA Info Security Requirements Associated with an Internal Threat GL 96-004 Boraflex Degradation in Spent Fuel Pool NA Item was applicable only to units with operating license at Storage Racks the time the item was issued.

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

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

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

Requirements for Uranium Recovery Licensees GL 98-001 Year 2000 Readiness of Computer NA Item was applicable only to units with operating license at Systems at Nuclear Power Plants the time the item was issued.

GL 98-003 NMSS Licensees' and Certificate Holders' NA Does not apply to power reactor.

Year 2000 Readiness Programs GL 98-005 Boiling Water Reactor Licensees Use of the NA Boiling Water Reactor BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds GL 99-001 Recent Nuclear Material Safety and NA Info Safeguards Decision on Bundling Exempt Quantities GL 99-002 Laboratory Testing of Nuclear Grade NA Item was applicable only to units with operating license at Activated Charcoal the time the item was issued.

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 STATUS CODE DEFINITIONS NA: NOT APPLICABLE: Justification as to why a section / subsection is not applicable is provided in the ADDITIONAL INFORMATION column.

Page 33 of 33 * = See last page for status code definition.

GENERIC COMMUNICATIONS STATUS = OPEN

GENERIC COMMUNICATIONS STATUS = OPEN ITEM TITLE ADDITIONAL INFORMATION GL 82-028 Inadequate Core Cooling Instrumentation 0 LICENSE CONDITION - Detectors for Inadequate core System 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 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 Q PAM system.

GL 88-020 Individual Plant Examination for Severe 0 Unit 2 Action: Complete evaluation for Unit 2.

Accident Vulnerabilities GL 08-001 Managing Gas Accumulation in Emergency 0 Core Cooling, Decay Heat Removal, and Containment Spray Systems

..................................................................... a ...................................................................

NUREG- Instrumentation For Detection of 0 LICENSE CONDITION - Detectors for Inadequate core 0737, II.F.2 Inadequate Core-Cooling 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.

STATUS CODE DEFINITIONS 0: OPEN: No action or documentation is provided that shows the staff has reviewed the item for WBN Unit 2.

Page 1 of I * = See last page for status code definition.

GENERIC COMMUNICATIONS STATUS = OPEN/TECHNICAL SPECIFICATIONS

GENERIC COMMUNICATIONS STATUS = OPEN/TECHNICAL SPECIFICATIONS ITEM TITLE ADDITIONAL INFORMATION GL 03-001 Control Room Habitability OT Initial response for Unit 2 on September 7, 2007 Unit 2 Action: Incorporate TSTF-448 into Technical Specifications.

............................................................................................................................ .o............

GL 06-001 Steam Generator Tube Integrity and OT Initial response for Unit 2 on September 7, 2007.

Associated Technical Specifications Unit 2 Action: Incorporate TSTF-449 into Technical Specifications.

NUREG- Primary Coolant Outside Containment OT Resolved for Unit 1 only in SSER10; reviewed in Appendix 0737, EE of SSER16.

II1.D.1.1 Unit 2 Actions: Include the waste gas disposal system in the leakage reduction program and incorporate in Unit 2 Technical Specifications.

STATUS CODE DEFINITIONS 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.

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

GENERIC COMMUNICATIONS STATUS = OPENNALIDATION

GENERIC COMMUNICATIONS STATUS = OPEN/VALIDATION ITEM TITLE COMUNICTION GENERC ADDITIONAL INFORMATION B 92-001 Failure of Thermo-Lag 330 Fire Barrier OV TVA configurations for Thermo-Lag 330-1 were reviewed in System to Maintain Cabling inWide Cable SSER18 and accepted in NRC letter dated January 6, 1998 Trays and Small Conduits Free From Fire (includes a supplemental SE).

Damage 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.

B 96-001 Control Rod Insertion Problems (PWR) OV NRC acceptance letter for Unit 1 dated July 22, 1996 -

Initial response for Unit 2 on September 7, 2007.

Unit 2 Action: Issue Emergency Operating Procedure and provide core map.

OV NRC acceptance letter dated November 20, 2001 (Unit 1)

B 01-001 Circumferential Cracking of Reactor Pressure Vessel (RPV) Head Penetration - Initial response for Unit 2 on September 7, 2007.

Nozzles Unit 2 Action: Perform baseline inspection.

B 02-001 RPV Head Degradation and Reactor OV NRC review of Unit l's 15 day response in letter dated Coolant Pressure Boundary Integrity May 20, 2002 - Initial response for Unit 2 on September 7, 2007.

Unit 2 Action: Perform baseline inspection.

B 02-002 RPV Head and Vessel Head Penetration OV NRC acceptance letter dated December 20, 2002 (Unit 1)

Nozzle Inspection Programs - Initial response for Unit 2 on September 7, 2007.

Unit 2 Action: Perform baseline inspection.

B 03-002 Leakage from RPV Lower Head OV NRC acceptance letter dated October 6, 2004 (Unit 1) -

Penetrations and Reactor Coolant Pressure Initial response for Unit 2 on September 7, 2007.

Boundary Integrity (PWRs)

Unit 2 Action: Perform baseline inspection.

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

Used in the Fabrication of Pressurizer Penetrations and Steam Space Piping Unit 2 Actions: Provide details of pressurizer and Connections at PWRs penetrations and apply Material Stress Improvement Process.

B 07-001 Security Officer Attentiveness OV Item concerns a multi-unit issue that was completed for both units.

GL 89-004 Guidelines on Developing Acceptable OV NRC reviewed in 3.9.6 of SSER14 (Unit 1).

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.

Page I of 3 * = See last page for status code definition.

ITEM TITLE ADDITIONAL INFORMATION 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 95-003 Circumferential Cracking of Steam OV NRC acceptance letter dated May 16, 1997 (Unit 1) -

Generator Tubes Initial response for Unit 2 on September 7, 2007. TVA responded to a request for additional information on December 17, 2007.

Unit 2 Action: Perform baseline inspection.

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

Design-Basis Accident Conditions Unit 2 Action: Implement modification to provide containment penetration relief.

GL 97-001 Degradation of Control Rod Drive OV NRC acceptance letter dated November 4, 1999 (Unit 1).

Mechanism Nozzle and Other Vessel Closure Head Penetrations Unit 2 Action: Provide a report to address the inspection program.

GL 97-004 Assurance of Sufficient Net Positive OV NRC acceptance letter dated June 17, 1998 (Unit 1) -

Suction Head for Emergency Core Cooling Initial response for Unit 2 on September 7, 2007.

and Containment Heat Removal Pumps Unit 2 Actions: Install new sump strainers, and perform other modification-related activities identical to Unit 1.

GL 97-005 Steam Generator Tube Inspection OV NRC acceptance letter dated September 22, 1998 Techniques (Unit 1) - Initial response for Unit 2 on September 7, 2007.

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.

GL 97-006 Degradation of Steam Generator Internals OV NRC acceptance letter dated October 19, 1999 (Unit 1) -

Initial response for Unit 2 on September 7, 2007. TVA responded to a request for additional information on December 17, 2007.

Unit 2 Action: Perform SG inspections during each refueling outage.

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

Associated Potential for Loss of Emergency Mitigation Functions While in a Shutdown Unit 2 Actions: 1) Review the ECCS designs to ensure they Condition do not contain design features which can render them susceptible to common-cause failures; and 2) document the results.

GL 98-004 Potential for Degradation of the ECCS and OV NRC closure letter dated November 24, 1999 (Unit 1). -

the Containment Spray System After a Initial response for Unit 2 on September 7, 2007.

LOCA Because of Construction and Protective Coating Deficiencies and Unit 2 Actions: Install new sump strainers, and perform Foreign Material in Containment other modification-related activities identical to Unit 1.

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

Unit 2 Action: Perform baseline inspection.

GL 04-002 Potential Impact of Debris Blockage on OV NRC Audit Report dated February 7, 2007 (Unit 1) - Initial Emergency Recirculation During Design response for Unit 2 on September 7, 2007.

Basis Accidents at PWRs Unit 2 Actions: Install new sump strainers, and perform other modification-related activities identical to Unit 1.

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

Risk and the Operability of Offsite Power Unit 2 Action: Complete the two unit baseline electrical calculations and implementing procedures.

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

Unit 2 Action: Addressed in CAP/SP. The Fire Protection Corrective Action Program will ensure Unit 2 conforms with NRC requirements and applicable guidelines.

Inaccessible or Underground Power Cable OV Initial response for Unit 2 on September 7, 2007.

GL 07-001 Failures That Disable Accident Mitigation Systems or Cause Plant Transients Unit 2 Action: Complete testing of four additional cables.

NUREG- Independent Safety Engineering Group OV LICENSE CONDITION - Independent Safety Engineering 0737, I.B.1.2 Group (ISEG) (NUREG-0737, I.B.1.2)

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.

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NUREG- Control Room Design Review OV NRC reviewed in SSER5, SSER6, SSER15, and Appendix 0737, I.D.1 EE of SSER16.

Unit 2 Actions: Complete the CRDR process. Perform rewiring in accordance with ECN 5982. Take advantage of the completed Human Engineering reviews to ensure appropriate configuration for Unit 2 control panels. See CRDR Special Program.

NUREG- Control-Room Habitability OV TVA: letter dated October 29, 1981 0737, II1.D.3.4 NRC: SSER16 NRC reviewed in SER and in Appendix EE of SSER16.

Unit 2 Action: Complete with CRDR completion.

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STATUS CODE DEFINITIONS 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.

Page 3 of 3 * = See last page for status code definition.