Text

Watts Bar Nuclear Plant, Unit 2 - Status of Regulatory Framework for the Completion of Construction and Licensing for Unit 2 - Revision 5 and Status of Generic Communications for Unit 2 - Revision 5
	ML110210486
Person / Time
Site:	Watts Bar
Issue date:	01/21/2011
From:	Bajestani M; Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	TAC MD8314, TAC ME6311
	Download: ML110210486 (210)
	v • d • e

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

January 21, 2011

U.S. Nuclear Regulatory Commission

ATTN: Document Control Desk Washington, D.C. 20555-0001

Watts Bar Nuclear Plant, Unit 2 NRC Docket No. 50-391

Subject:

Watts Bar Nuclear Plant (WBN) Unit 2 - Status of Regulatory Framework for the Completion of Construction and Licensing for Unit 2 - Revision 5 (TAC No. MD6311), and Status of Generic Communications for Unit 2 -

Revision 5 (TAC No. MD8314)

Reference:

1. Letter from TVA to NRC dated October 28, 2010, "Watts Bar Nuclear Plant (WBN) Unit 2 - Status of Regulatory Framework for the Completion of Construction and Licensing for Unit 2 - Revision 4 (TAC No. MD6311), and Status of Generic Communications for Unit 2 - Revision 4 (TAC No. MD8314)"

This letter provides an updated status of the Regulatory Framework for the completion of construction and licensing activities for WBN Unit 2 as well as an updated status of Generic Communications for WBN Unit 2. TVA's last revision to these two status updates, Revision 4, was submitted on October 28, 2010 (Reference 1).

For the Regulatory Framework, Enclosure 1 provides the revised Regulatory Framework Master, and Enclosure 2 provides a version of the table showing only those items revised in this Revision 5.

For the Generic Communications, Enclosure 3 provides the revised Generic Communications Master, and Enclosure 4 provides a version of the table showing only those items revised in this Revision 5.

U.S.Nuclear Regulatory Commission Page 2 January 21, 2011 The following is the status of the items which are applicable to WBN Unit 2.The status codes are defined on the last page of each enclosure.

SERf GENERIC STATUS SSER COMM.TOTAL C (CLOSED)204 119 323 CI (CLOSED/IMPLEMENTATION) 18 116 134 CT (CLOSED/TECHNICAL 0 0 0 SPECIFICATIONS) 0 (OPEN)50 3 53 OT (OPEN/TECHNICAL 1 0 1 SPECI FICATIONS)

OV (OPEN/VALIDATION) 7 7 14 S (SUBMITTED) 65 20 85 TOTAL 345 265 610 There are no new regulatory commitments associated with this submittal.

If you have any questions, please contact William Crouch at (423)365-2004.Respectfully, f 1'bIi;;!t,,,---..-'

---Masoud BaJestanl Watts BfY Unit 2 Vice President

Enclosures:

1.SER and Supplements Review Matrix-Master Table 2.SER and Supplements Review Matrix-Revision 5 Changes 3.Generic Communications

-Master Table 4.Generic Communications

-Revision 5 Changes U.S. Nuclear Regulatory Commission

Page 3 January 21, 2011 cc (Enclosures):

U. S. Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, Georgia 30303-1257

NRC Resident Inspector Unit 2 Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381

Enclosure 1

SER and Supplement s Review Matrix - Master Table SAFETY EVALUATION REPORT AND SUPPLEMENTS(NUREG-0847) REVIEW MATRIX:

MASTER TABLESER SECTIONSSER #ADDITIONAL INFORMATION

REV. Overview only NA100..Overview only NA110..Overview only NA111..Overview only NA112..Overview only NA113..Overview only NA114..Overview only NA120..Overview only NA130..Overview only NA131..Overview only NA132..Overview only NA140..Overview only NA150..Overview only NA160..Overview only NA170..Overview only NA180..Overview only NA190..Overview only NA1100..* = See last page for status code definition.Page 1 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 0Approved for both units in SER.

C200..0Approved for both units in SER.

C210..0Approved for both units in SER.

C211..0Approved for both units in SER.

C212..21SRP requirement. Unit 2 Action:Update FSAR for present and projected population over the lifetime of the plant.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Status in SSER21 is Open (NRR).

Amendment 94 to the Unit 2 FSAR was submitted on August 27, 2009.Part of this amendment revised population information in Section 2.1.3.

S213..0221"CONCLUSIONS" left open until all items in subsection are closed.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:Status in SSER21 is Open (NRR).

Amendment 94 to the Unit 2 FSAR was submitted on August 27, 2009.

Part of this amendment revised population information in Unit 2 FSAR Section 2.1.3.

S214..020Approved for both units in SER.

C220..* = See last page for status code definition.Page 2 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21SRP requirement. Unit 2 Action: Update FSAR for potential external hazards and hazardous materials.

REVISION 02 UPDATE:

Status in SSER21 is Open (NRR).

Amendment 94 to the Unit 2 FSAR was submitted on August 27, 2009.

Part of this amendment revised the description of hazardous material shipped past the plant in Section 2.2.2.2.

S221..0221SRP requirement. Unit 2 Action:

Update FSAR for projected annual number of aircraft flights.

REVISION 02 UPDATE:

Status in SSER21 is Open (NRR).

Amendment 94 to the Unit 2 FSAR was submitted on August 27, 2009.Part of this amendment revised information concerning airports and numbers of aircraft flights in 2.2.2.5.

S222..0221"CONCLUSIONS" left open until all items in subsection are closed.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Status in SSER21 is Open (NRR).

Amendment 94 to the Unit 2 FSAR was submitted on August 27, 2009.Part of this amendment revised the description of hazardous material shipped past the plant in Section 2.2.2.2.

S223..020Approved for both units in SER.

C230..0Approved for both units in SER.

C231..0Approved for both units in SER.

C232..0See 13.3.3 (Emergency Preparedness Evaluation Conclusions).

C233..* = See last page for status code definition.Page 3 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 14TVA updated information on portions of the metrology program in FSAR amendment 83. This was reviewed and found acceptable in SSER14.

C234..0114TVA updated information on portions of the metrology program in FSAR amendment 83. This was reviewed and found acceptable in SSER14.

C235..010Approved for both units in SER.

C240..0Approved for both units in SER.

C241..0Approved for both units in SER.

C242..0REVISION 02 UPDATE:Approved for both units in SER.

O243..020Approved for both units in SER.

C244..0GL 89-22, "Potential For Increased Roof Load Due to Changes in Maximum Precipitation" - Answer to informal question provided in TVA letter dated December 16, 1981, and subsequently included in FSAR. GL did not require a response. No further action required.

C245..0Approved for both units in SER.

C246..0Approved for both units in SER.

C247..21CONFIRMATORY ISSUE for design basis groundwater level for ERCW pipelineAmendment 50 to the FSAR (May 1, 1984) provided a description of the analysis used to determine the 25-year groundwater level for the ERCW pipeline. Staff closed issue in SSER3.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Status in SSER21 is Open (NRR).

O248..02* = See last page for status code definition.Page 4 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21SRP requirement. Unit 2 Action: Update FSAR for present and projected use of local and regional groundwater.

REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O249..0221Staff found flood emergency plan and draft Technical Specifications acceptable in original 1982 SER.Unit 2 Action:

Address in Technical Specifications as appropriate.

REVISION 02 UPDATE:

Status in SSER21 is Open (Inspection).

Amendment B of the Technical Requirements Manual (TRM) was submitted on February 2, 2010.TRM TLCO 3.7.2 provides the Flood Protection Plan.

S2410..02Addressed in 2.4.6.

NA2411..Addressed in 2.4.7.

NA2412..Addressed in 2.4.9.

NA2413..Addressed in 2.4.10.

NA2414..0Approved for both units in SER.

C250..0Approved for both units in SER.

C251..0Approved for both units in SER.

C252..0Approved for both units in SER.

C253..* = See last page for status code definition.Page 5 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 11CONFIRMATORY ISSUE for design differential settlement of piping and electrical componentsAnalysis was presented to staff in September 1983. Staff found analysis and results acceptable. Staff closed issue in SSER3.--------------------

CONFIRMATORY ISSUE for analysis of sheetpile walls Staff performed audit in September 1982, and determined TVA had used reasonable assumptions. Staff closed issue in SSER3.---------------------

CONFIRMATORY ISSUE for material and geometric damping in soil-structure interaction (SSI) analysisStaff performed audit in September 1982, and determined TVA had used reasonable assumptions. Staff closed issue in SSER3.---------------------

OUTSTANDING ISSUE (1) on liquefaction beneath ERCW pipelines and Class 1E electrical conduit.Amendment 50 to the FSAR (May 1, 1984) provided a description of the underground barriers along the ERCW pipelines. Staff agreed the barriers provide sufficient confinement to any liquefied soil. Staff closed issue in SSER3.--------------------FSAR amendment 54-63 was reviewed in SSER9. NRC determined that the conclusions previously issued in the SER and SSER3 remained unchanged.---------------------The Special Program (SP) for Soil Liquefaction was reviewed in SSER11. NRC IR 50-390/92-45 and 50-391/92-45 concluded that TVA had correctly implemented the SP and that it was closed. SSER11 accepted the implementation for WBN Unit 1. Per TVA letter dated August 3, 2007, implementation of the Soil Liquefaction SP is complete for both units.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 03 UPDATE:

NRC IR 50-391/2009-605 noted that the Soil Liquefaction SP was closed for Unit 2.C254..030Approved for both units in SER.

C255..0Approved for both units in SER.

C256..0Approved for both units in SER.

C260..0Approved for both units in SER.

C300..* = See last page for status code definition.Page 6 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 0Approved for both units in SER.

C310..0Approved for both units in SER.

C311..0Approved for both units in SER.

C312..14In SSER14, the staff reviewed revisions to Table 3.2-2, "Summary of Criteria - Mechanical System Components", and found the table acceptable.

C320..018CONFIRMATORY ISSUE for seismic classification of structures, systems, and components important to safetyThe staff reviewed Amendment 49 to FSAR and actions implemented by TVA to address ERCW seismic classification in SSER3 and found them acceptable, pending verification of actions. Staff closed issue on ERCW seismic category upgrade and seismic classification in SSER5.--------------------

CONFIRMATORY ISSUE for ERCW upgrade to seismic category 1Staff verified that required portion of ERCW had been upgraded or replaced satisfactorily in SSER5 and closed this issue.--------------------

In SSER6, the staff addressed and resolved an issue on Category I boundary.

OUTSTANDING ISSUE involving seismic classification of cable trays and conduitsIn SSER6, staff identified an issue on seismic classification of cable trays and conduits being categorized as I(L). In its May 8, 1991, letter, TVA proposed to analyze conduits as Seismic Category I subsystems. Additionally, in a September 18, 1991 letter, TVA agreed to perform cable tray qualification using conventional linear elastic analysis methods, considering nonlinear response behavior on a case-by-case basis and to submit these cases to the staff for approval. The staff resolved this issue in SSER8.

C321..01* = See last page for status code definition.Page 7 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21Section 3.2.2 of SSER3 discusses confirmatory issues for seismic classification and upgrade of ERCW that are already included in 3.2.1.--------------------

Staff accepted implementation of Heat Code Traceability CAP for Unit 1 in SSER7.Unit 2 Action:

Complete CAP using Unit 1 approach.

Staff reviewed updated information in Amendment 68 on use of codes and standards in SSER9 and stated that prior conclusions were unchanged.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).TVA's September 26, 2008, letter proposed the use of the Unit 1 approach to resolve the Heat Code Traceability CAP.In SSER21, the Heat Code Traceability CAP was resolved. Completion of Heat Code Traceability CAP is tracked under 23.2.9.

CI322..020Approved for both units in SER.

C330..0Approved for both units in SER.

C331..0Approved for both units in SER.

C332..0Approved for both units in SER.

C340..0Approved for both units in SER.

C341..Addressed in 3.4.1.

NA342..0Approved for both units in SER.

C350..14In SSER9, the staff determined that a new spectrum used for the design of a new DG building and other Category I structures built after 1979 was acceptable. In SSER14, clarification in Amendment 79 on internal missile sources was reviewed and did not change prior conclusions. Staff also reviewed revised information on turbine missiles and concluded that impact of potential missiles was insignificant.

C351..01* = See last page for status code definition.Page 8 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 2CONFIRMATORY ISSUE for modifications to protect Diesel GeneratorsTVA submitted a proposed design modification for installation of a reinforced concrete curb around the diesel exhaust stacks to protect them from damage in a letter dated November 24, 1982. The staff found this acceptable and closed this issue in SSER2.

C352..0Approved for both units in SER.

C353..0121In SSER6, the staff accepted TVA approa ches involving arbitrary intermediate breaks, determination of intermediate break locations and analysis of jet impingement loads.In SSER11, the staff reviewed results of the MELB Special Program and determined that the conclusion in the SER finding plant design for protection against piping failures outside containment was still valid.Unit 2 Action: Complete Special Program using the Unit 1 approach.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (NRR).TVA's September 26, 2008, letter proposed the use of the Unit 1 approach to resolve the MELB SP.In SSER21, the MELB Special Program was resolved. Completion of MELB SP is tracked under 23.3.8.

CI360..02* = See last page for status code definition.Page 9 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21OUTSTANDING ISSUE involving main steam line break (MSLB) outside containmentIn a letter dated November 30, 1992, TVA submitted a new evaluation for both Units 1 and 2 accounting for increased environmental temperatures in the MSVV rooms due to release of superheated steam and later submitted, by letter dated March 28, 1994, additional information related to the assumptions made in this analysis for both units. The staff reviewed this information together with their detailed evaluation and acceptance of the same methodology applied at Sequoyah and concluded that the MSLB analysis for the WBN MSVV rooms, including the effects of superheated steam, was acceptable and identified this issue as resolved in SSER14.--------------------

In SSER14, the staff reviewed the construction of response spectra for the steel containment vessel resulting from the compartment pressure transients caused by pipe break and TVA modeling of the SCV for both units (see TVA letter dated December 30, 1993) and concluded that the methodology for obtaining shell dynamic displacements and construction of spectra were acceptable.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Status in SSER21 is Open (NRR).

O361..0214The 3.6.2 discussion in SSER14 on response spectra for the SCV refers to the evaluation provided in 3.6.1.

C362..0112New section in SRP 1987. Approved for both units in Appendix J of SSER5. The staff concluded in SSER12 that TVA may eliminate pressurizer surge line rupture from the design basis for Units 1 and 2.

C363..01* = See last page for status code definition.

Page 10 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21The staff concluded in SSER6 that FSAR section 3.7 which was added to describe Set A, Set B and Set C seismic analysis was consistent with the Seismic Analysis CAP.Unit 2 Action:

Complete CAP using the Unit 1 approach.

REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

TVA's September 26, 2008, letter proposed the use of the Unit 1 approach to resolve the Seismic Analysis CAP .In SSER21, the Seismic Analysis CAP was resolved. Completion of the Seismic Analysis CAP is tracked under 23.2.16.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 03 UPDATE:NRC IR 50-391/2010-602 noted that the Seismic Analysis CAP was closed for Unit 2.C370..0321OUTSTANDING ISSUE involving update of FSAR for seismic design issuesThe staff reviewed FSAR Amendment 68 and found that required changes had been incorporated into the FSAR, as committed to in TVA letter dated December 18, 1990, for Units 1 and 2, and issue was deemed resolved in SSER6. SSER9 stated the Seismic Analysis CAP was acceptably implemented for Unit 1. SSER16 discusses use of a vertical PGA of .15g rather than .18g for Set B spectra and determined that it was acceptable.Unit 2 Action: Complete CAP using Unit 1 approach.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (NRR).

TVA's September 26, 2008, letter proposed the use of the Unit 1 approach to resolve the Seismic Analysis CAP .In SSER21, the Seismic Analysis CAP was resolved. Completion of the Seismic Analysis CAP is tracked under 23.2.16.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 03 UPDATE:

NRC IR 50-391/2010-602 noted that the Seismic Analysis CAP was closed for Unit 2.C371..03* = See last page for status code definition.

Page 11 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 213.7.2.1.2: OUTSTANDING ISSUE involving mass eccentricityIn a letter dated May 8, 1991, for Units 1 and 2, TVA provided clarification that actual mass eccentricities from such items as equipment hatch and lock used in evaluating the steel containment vessel for an earthquake load were replaced by a 5% accidental eccentricity. This was demonstrated to be conservative. TVA also proposed a revision to the FSAR to document this change. The staff found this acceptable and resolved this issue in SSER8.--------------------

3.7.2.1.2: OUTSTANDING ISSUE involving comparison of Set A vs. Set B responseThe staff considered this item (opened in SSER6) resolved in SSER11 based on audits and inspections since SSER6.Unit 2 Action: Complete Seismic Analysis CAP using the Unit 1 approach.--------------------

In SSER16, the staff discussed the review and acceptability of the NSSS-ICS modeling for seismic analysis.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (NRR).TVA's September 26, 2008, letter proposed the use of the Unit 1 approach to resolve the Seismic Analysis CAP .In SSER21, the Seismic Analysis CAP was resolved. Completion of the Seismic Analysis CAP is tracked under 23.2.16.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 03 UPDATE:

NRC IR 50-391/2010-602 noted that the Seismic Analysis CAP was closed for Unit 2.C372..03* = See last page for status code definition.

Page 12 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21OUTSTANDING ISSUE involving number of peak cycles to be used for OBEIn SSER6, the staff identified an issue involving the number of peak cycles to be used for OBE. In a letter dated May 8, 1991, for both units, TVA proposed to revise the FSAR for ASME Section III Class I piping analysis to include the assumption of 5 OBEs and 1 SSE and a minimum of 10 peak stress cycles per event. The staff accepted this in SSER8.--------------------

OUTSTANDING ISSUE involving use of code cases, damping factors for conduit and use of worst case, critical case and bounding caseIn SSER6, the staff identified outstanding issues involving code case use, damping factors for conduit and use of worst case, critical case and bounding case. Deficiencies identified in the use of worst case, critical case and bounding calculations were resolved in IR 50-390/93-201, and this issue was considered resolved for Unit 1 in SSER12. Unit 2 Action: Addressed in CAP/SP. The Unit 1 approach will be used for Unit 2.

OUTSTANDING ISSUE involving 1.2 multi mode factor In SSER6, the staff identified an issue involving a 1.2 multi-mode factor. In SSER8, the staff continued to review the use of a multi-mode factor of 1.2. The staff reviewed verification studies performed by TVA to justify the use of a 1.2 multi-mode factor in seismic evaluation of certain sub systems in SSER8 and SSER9 and, after TVA provided further confirmation of supporting calculations, the use of Complete Quadratic Combinations and validity of two degree of freedom predictions in a letter dated October 10, 1991, for both units, the staff considered this issue resolved in SSER9.--------------------Conduit Supports Corrective Action Program. Process was reviewed and determined to be acceptable for Unit 1 in SER dated September 1, 1989.Unit 2 Action: Addressed in CAP/SP. The Unit 1 approach will be used for Unit 2.

In SSER6, the staff reviewed several other seismic analysis considerations including combination of components of earthquake motion, use of load factors in simplified analysis of equipment, consideration of torsional effects of eccentric masses in piping analysis; damping values for cable trays, HVAC and equipment and components; analysis of mounting for equipment and components; and loads and load combinations used in design of HVAC ducts and supports and found them acceptable.In SSER7, the staff reviewed the seismic design of the Refueling Water Storage Tank, the only safety related above ground vertical steel tank in the plant, and found it acceptable.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

C373..03* = See last page for status code definition.

Page 13 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. TVA's September 26, 2008, letter proposed the use of the Unit 1 approach to resolve the Seismic Analysis CAP and the Conduit Supports CAP.In SSER21, the Seismic Analysis CAP was resolved. Completion of the Seismic Analysis CAP is tracked under 23.2.16. In SSER21, the Conduit Supports CAP was resolved. Completion of the Electrical Conduit and Conduit Supports CAP is tracked under 23.2.16.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 03 UPDATE:

NRC IR 50-391/2010-602 noted that the Seismic Analysis CAP was closed for Unit 2.0Approved for both units in SER.

C374..21OUTSTANDING ISSUE involving load combinations and stress allowablesIn response to staff concerns regarding use of ductility ratio when considering thermally induced stresses, TVA stated in a letter dated April 6, 1992, for both units, that they would use a methodology consistent with SRP 3.8.4 for the design of steel members and use the linear elastic provision of DG-C 1.6.12, Rev. 1, "Evaluation of Steel Structures with Thermal Restraint," except for the energy balance provision of Section C.2.3.1. The staff found this acceptable. TVA also agreed, in its May 8, 1991, letter for both units, that any further sampling of structural welds after the issuance of NCIG-2, Rev. 2 would be to that revision. This issue was resolved in SSER9.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O380..023CONFIRMATORY ISSUE - verify buckling methodologyIn response to staff concern, TVA submitted a letter dated May 16, 1984, for both units, stating that TVA calculations already accounted for new information from NRC-sponsored research programs, particularly information concerning reinforcement around shell (vessel) opening. Based on their review of the response, the staff closed this issue in SSER3.

C381..017The staff accepted implementation of the Concrete Quality Special Program for Unit 1 in SSER7. This program is considered closed for Unit 2 based on the work performed for Unit 1. The was identified in a TVA letter dated August 3, 2007, WBN - Unit 2 - Reactivation of Construction Activities C382..01* = See last page for status code definition.

Page 14 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21The staff reviewed materials, allowable stresses and load cases for the watertight equipment hatch cover in an FSAR Table in Amendment and found them acceptable for both units in SSER14.The staff reviewed allowable stresses for Category I structural steel and found them acceptable for both units in SSER16.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O383..020Approved for both units in SER.

C384..0Approved for both units in SER.

C390..13OUTSTANDING ISSUE involving assumption in piping analysis for water-hammer due to check valve slamIn SSER6, the NRC expressed concern regarding TVA's piping analysis that postulated failure of certain supports, TVA submitted an August 4, 1992, letter stating that, where possible, supports were upgraded in the analysis to maintain structural integrity during the postulated loading scenario. The issue was resolved in SSER13.Unit 2 Action: Modify supports as needed.

OV391..0114The staff reviewed "Pre-operational Vibration and Dynamic Effects Testing on Piping", and found this area acceptable in SSER14.

C392..01* = See last page for status code definition.

Page 15 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 153.9.3.1: OUTSTANDING ISSUE involving use of experience data to qualify category I(L) pipingThe staff identified a concern regarding the use of experience data as a method of seismic qualification of Category I(L) piping in SSER6. TVA stated in a letter dated December 18, 1990 for both units, that it was performing a verification program to validate the original seismic design basis for Category I(L) piping, including a screening criteria based on earthquake experience data to identify items requiring further evaluation and bounding case analysis to demonstrate the conservatism of the screening criteria. In a September 20, 1991, for both units, letter, TVA provided revised criteria for the bounding case analysis. Based on the staff's evaluation, the issue was considered resolved in SSER8.---------------------3.9.3.3: LICENSE CONDITION - Relief and safety valve testing (II.D.1)

Staff found TVA approach in response to this issue, using information from EPRI valve test program and performing modifications to safety and relief discharge piping and supports, was acceptable. Issue was considered resolved in SSER3.--------------------

3.9.3.3: OUTSTANDING ISSUE involving operating characteristics of main steam safety valvesThe staff identified a concern with operating characteristics of main steam safety valves in SSER6. In a letter dated June 21, 1991, TVA responded to NRC concerns regarding the design and installation of MSSVs stated that all valves and piping components were analyzed for all MSSV discharge loads acting simultaneously, combined with other required loads and this was accepted by the staff. In the same letter, TVA also provided the method used to establish the MSSV adjustment ring settings for plant valves and this was acceptable to the staff. This resolved the issue in SSER7.Unit 2 Action:

Provide basis of applicability of Unit 1 MSSV analysis to Unit 2.--------------------3.9.3.4: CONFIRMATORY ISSUE involving baseplate flexibility and its effect on anchor bolt loadsThe staff continued to review baseplate flexibility and its effect on anchor bolt loads. The issue remained open in SSER6. The TVA response to this issue, in a letter dated July 26, 1991, for both units, described an update to the previous response for B 79-02 and its civil design standard for concrete anchorage, which incorporated an increase in anchor stiffness and consideration of prying forces for thin baseplates analyzed by hand. The staff determined that this adequately resolved the issue in SSER8.--------------------

3.9.3.4: OUTSTANDING ISSUE involving stiffness and deflection limits for seismic Category I pipe supportsThe staff questioned new support stiffness and deflection limits for seismic Category I pipe supports in SSER6. The TVA program to demonstrate that change in design criteria which uses stiffness and deflection limits for Category I pipe supports did not compromise the adequacy of pipe supports, was submitted in a TVA letter dated September 30, 1991, for both units, and was found to be acceptable by the staff and the issue was resolved in SSER8.

S393..02* = See last page for status code definition.

Page 16 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. --------------------3.9.3.4: OUTSTANDING ISSUE, staff was awaiting TVA concurrence on their position with respect to margin for critical buckling of pipe supportsIn a letter dated May 14, 1984, TVA provided results of a sampling program and determined that compressive stresses for pipe supports did not exceed acceptance criteria established by NRC and staff considered this issue resolved in SSER4.--------------------The staff reviewed proposed new criteria for service load combinations and associated stress limits for ASME Code Class 1, 2, and 3 pipe supports in SSER6 and found them acceptable.In SSER15, the staff found the response to NUREG-0737, Item II.D.1, "Performance Testing of Relief and Safety Valves," acceptable.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:TVA determined that the Unit 1 MSSV analysis was applicable to Unit 2.

Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.

Section 10.1 was amended to reference the Westinghouse safety evaluation that evaluated the effect of the MSSV blowdown on the LOCA related FSAR analysis results.0Approved for both units in SER.

C394..0Approved for both units in SER.

C395..* = See last page for status code definition.

Page 17 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 20LICENSE CONDITION on inservice testing of pumps and valvesThe staff stated that they were reviewing TVA's response to GL 89-04, addressing acceptable IST programs and the license condition on inservice testing of pumps and valves remained open in SSER5. TVA committed to submit a revised ASME Section XI Inservice Pump and Valve Test Program six months before the projected date of operating license issuance in an August 21, 1989, letter. On this basis, the staff considered that the proposed license condition was no longer required in SSER12.--------------------

OUTSTANDING ISSUE required that Technical Specifications include limiting condition for operation that requires plant shutdown or system isolation when leak limits are not met. Staff had not reviewed Technical Specifications.The safety evaluation in SSER14 states that the staff did not find any IST issues that would prevent issuance of an operating license for Unit 1. The item was resolved in SSER14. Unit 2 Action: Submit Technical Specifications.

In SSER18, the staff approved a proposed alternative for set pressure testing of the three pressurizer safety relief valves that provide overpressure protection for the reactor coolant system.In SSER20, the staff discussed 13 issues that remained to be resolved for the pump and valve inservice testing program and stated that they had been addressed in a manner that complies with the staff's position and they granted relief for an additional relief request.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.TS LCO 3.4.13 provides the requirements for RCS Operational Leakage. Included in this is a requirement to shutdown the unit if leakage can not be reduced to within limits within the specified time frame.TS LCO 3.4.14 provides the requirements for RCS Pressure Isolation Valve Leakage. Included in this is a requirement to shutdown the unit if leakage can not be reduced to within limits within the specified time frame.TS 5.7.2.11 provides the Inservice Testing Program.

S396..02Area not addressed in 1981 Standard Review Plan.

NA397..Area not addressed in 1981 Standard Review Plan.

NA398..* = See last page for status code definition.

Page 18 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21In SSER1 the staff discussed their evaluation of the TVA program for qualification of electrical and mechanical equipment for seismic and other loads, and opened the OUTSTANDING ISSUE involving adequacy of frequency test, peak broadening of response spectra, reconciling actual field mounting by welding vs. testing configuration mounted by bolting and need for surveillance and maintenance programs to address aging.The staff provided a status of these issues in SSER3 and closed peak broadening of response spectra, use of damping values, consideration of nozzle loads, and status of seismic qualification. Other specific issues were closed in this supplement as well.In SSER5, the staff stated that this issue remained open.In a letter dated December 1, 1982, TVA provided justification for single-frequency tests to seismically qualify the Reactor Protection System cabinet. This showed that test response spectra (TRS) were substantially higher than broadened required response spectra (RRS) throughout the required frequency range. The staff evaluated test results and building seismic behavior and considered this aspect of the testing issue closed in SSER6.Staff concerns on the impact of aging on seismic performance were resolved in SSER6 based on discussions with TVA technical personnel and review of maintenance and surveillance instruction manuals.There was a specific issue on installing spacers for the 125V DC vital batteries as was done during qualification testing and required by the manufacturer. The issue was closed in SSER6 when it was determined that spacers had been installed.With regard to the overall issue on adequacy of testing, the staff performed an audit as part of Appendix S of SSER9. This included a review of the TVA approach, criteria and action plan to address effect of directional coupling and verification that acceleration at each device location is less than .95g because relay chatter at higher acceleration levels is expected. TRS enveloped RRS for all directions. The staff found the above to be in accordance with SRP 3.10 and IEEE 344-1975 and closed the issue.For reconciling the impact for equipment actually mounted using welding but tested with mounting by bolting, in-situ test results were provided to NRC (in letters dated April 30, 1985, and January 30, 1986) along with Westinghouse report on seismic qualification by analysis and testing for the main control board. The staff reviewed these results and on the basis of the consistency of all results provided, concluded that the issue was resolved in SSER6.Unit 2 Action:

Complete Equipment Seismic Qualification CAP using the Unit 1 approach.

In SSER4, the staff reviewed an issue on the vibration of deep draft pumps and found it acceptable.In SSER8, the staff accepted a proposed revision to FSAR Section 3.7.3.16 to indicate that the alternative seismic qualification method is to follow the requirements of IEEE Standard 344-1971 and address the guidelines of SRP Section 3.10.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

CI3100..02* = See last page for status code definition.

Page 19 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. TVA's September 26, 2008, letter proposed the use of the Unit 1 approach to resolve the Equipment Seismic Qualification CAP .In SSER21, the Equipment Seismic Qualification CAP was resolved. Completion of the Equipment Seismic Qualification CAP is tracked under 23.2.6.21OUTSTANDING ISSUE - TVA program not submitted at time of SERThe EQ program was submitted after issuance of the SER. It was reviewed and found acceptable in SSER15.Unit 2 Action: Complete EQ Special Program.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (NRR).

TVA's September 26, 2008, letter proposed the use of the Unit 1 approach to resolve the EQ SP.In SSER21, the Environmental Qualification Special Program was resolved. The EQ program is tracked under 23.3.4.

CI3110..02Addressed in 3.9.1 through 3.9.3.

NA3120..Addressed in 3.9.1 through 3.9.3.

NA3121..Addressed in 3.9.1 through 3.9.3.

NA3122..Addressed in 3.9.1 through 3.9.3.

NA3123..Addressed in 3.9.1 through 3.9.3.

NA3124..Addressed in 3.9.1 through 3.9.3.

NA3125..Addressed in 3.9.1 through 3.9.3.

NA3126..Area not addressed in 1981 Standard Review Plan.

NA3130..0Approved for both units in SER.

C400..0Approved for both units in SER.

C410..0Approved for both units in SER.

C420..* = See last page for status code definition.

Page 20 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 13In SSER13, NRC determined that internal fuel rod pressure was not key design information that needed to be included in the WBN Unit 1 Technical Specifications.Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.FSAR Chapter 4 was updated to address the application of the second generation Robust Fuel Assembly design (RFA-2)

S421..022CONFIRMATORY ISSUE on cladding collapse calculationsThe staff reviewed the calculation for the predicted cladding collapse for the most limiting Watts Bar fuel and found it acceptable. Staff closed issue in SSER2. Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.

FSAR Chapter 4 was updated to address the application of RFA-2 fuel.

S422..0213CONFIRMATORY ISSUE - identify margins and to offset reduction in DNBR due to fuel rod bowing and incorporating residual bow penalty into the Technical Specifications.In SSER2, the staff concluded TVA had an acceptable means of analyzing the effects of fuel rod bowing and determining any residual rod bowing penalties on the departure from nucleate boiling ratio and total peaking power. Staff closed the issue in SSER2.In SSER10, NRC reviewed design loading conditions for the reactor vessel internals and raised an issue on the seismic analysis of the control rod drive mechanisms (CRDMs). TVA's letter dated June 15, 1993, for both units discussed CRDM seismic operability. In SSER13, the NRC documented that concerns related to CRDM seismic qualification had been resolved.Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.

FSAR Chapter 4 was updated to address the application of RFA-2 fuel.

S423..02* = See last page for status code definition.

Page 21 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 0Approved for both units in SER.

C424..0"FUEL DESIGN CONCLUSIONS" left open until all items in subsection are closed.Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.

FSAR Chapter 4 was updated to address the application of RFA-2 fuel.

S425..020Approved for both units in SER.

C430..13In SSER13, NRC reviewed the V5H fuel design and found use of V5H fuel acceptable.Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.FSAR Chapter 4 was updated to address the application of RFA-2 fuel.

S431..0215In SSER13, NRC reviewed the V5H fuel design and found use of V5H fuel acceptable.Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------In SSER15, NRC reviewed TVA's proposed changes to the FSAR from a reanalysis of Pressurized Thermal Shock. The analysis was subsequently incorporated into the FSAR.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.

FSAR Chapter 4 was updated to address the application of RFA-2 fuel.

S432..02* = See last page for status code definition.

Page 22 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 13In SSER13, NRC reviewed the V5H fuel design and found use of V5H fuel acceptable.Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.

FSAR Chapter 4 was updated to address the application of RFA-2 fuel.

S433..0213In SSER13, NRC reviewed the V5H fuel design and found use of V5H fuel acceptable.Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.

FSAR Chapter 4 was updated to address the application of RFA-2 fuel.

S434..020Approved for both units in SER.

C440..0Approved for both units in SER.

C441..12In SSER12, NRC evaluated a change in reactor coolant flow (upflow) for both units. NRC concluded in a July 28, 1993 letter for both units that the proposed upflow modification was acceptable.----------------------Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.

FSAR Chapter 4 was updated to address the application of RFA-2 fuel.

S442..02* = See last page for status code definition.

Page 23 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 16OUTSTANDING ISSUE concerning removal of RTD bypass systemThis outstanding issue was opened in SSER6. Staff issued an SER dated June 13, 1989, for Unit 1 only that approved replacement of the RTD bypass system with an Eagle-21 microprocessor system for monitoring reactor coolant temperature. NRC provided their initial assessment of the RTD bypass removal for WBN Unit 1 in SSER8. This SER was reproduced in SSER8, Appendix R. In SSER16, NRC reviewed the flow measurement uncertainty value for the reactor coolant system. TVA letter dated December 5, 2007, informs NRC of intent to use Eagle-21 for Unit 2. NRC requested additional information December 27, 2007. TVA provided the requested information by letter dated February 28, 2008. By letter dated May 7, 2008, NRC provided a list of specific issues to be addressed in a future amendment application for Eagle-21 for WBN Unit 2. Unit 2 Action: Provide the additional information for NRC review.

In SSER12, NRC evaluated a change in reactor coolant flow (upflow) for both units. NRC concluded that the proposed upflow modification was acceptable.--------------------

In SSER13, NRC reviewed thermal hydraulic methodologies and concluded that the V5H thermal-hydraulic design was acceptable for Watts Bar.Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.

Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.FSAR Chapter 4 was updated to address the application of RFA-2 fuel.

S443..0213In SSER13, NRC reviewed TVA's responses to a request for additional information concerning fuel rod bowing and crud buildup for WBN Unit 1.Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.FSAR Chapter 4 was updated to address the application of RFA-2 fuel.

S444..02* = See last page for status code definition.

Page 24 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 16CONFIRMATORY ISSUE / LICENSE CONDITION on review of Loose Parts Monitoring System (LPMS) startup report and inclusion of limiting conditions for LPMS in Technical SpecificationsTVA letters dated February 25, 1982, and November 10, 1982, provided a description of operator training and an evaluation of conformance to RG 1.133. In SSER3, the staff closed the confirmatory issue and opened a license condition to track submittal of the startup test results and the alert level setting. In SSER5, the staff closed the LICENSE CONDITION to a TVA commitment to provide the startup test results and the alert level settings made in a letter dated September 19, 1990, for both units. In SSER16, NRC reviewed additional information and revised commitments associated with the LPMS. For Unit 2 due to obsolescence, TVA will replace the LPMS. Unit 2 Action: Provide the startup test results and the alert level settings.

O445..010Approved for both units in SER.

C446..0"Technical Resolution of Generic Issue B-59-(N-1) Loop Operation in BWRs and PWRs - N-1 Loop operation was addressed in original 1982 SER (4.4.7). Unit 2 Action: Confirm Technical Specifications prohibit (N-1) Loop Operation.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.TS LCO 3.4.4 requires that four Reactor Coolant System loops be operable and in operation during Modes 1 and 2.

S447..0210LICENSE CONDITION - Detectors for Inadequate core cooling (II.F.2)GL 82-28 / NUREG-0737, II.F.2, "Inadequate Core Cooling Instrumentation System" - 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.

O448..010"CONCLUSION" left open until all items in subsection are closed.

O449..010Approved for both units in SER.

C450..0Approved for both units in SER.

C451..* = See last page for status code definition.

Page 25 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 0Approved for both units in SER.

C452..0Approved for both units in SER.

C460..0Approved for both units in SER.

C500..6The staff stated that the Eagle 21 microprocessor system was an acceptable replacement of the resistance temperature detector (RTD) bypass system for monitoring reactor cooling temperature in SSER5. In SSER6, the staff noted that TVA had incorporated the information for this new design into the FSAR and said they would track results of the review of this design change as an outstanding issue - Removal of RTD Bypass System (See 4.4.3).Unit 2 Action:

Provide additional information for NRC review per 7.2.1.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.

S510..020Approved for both units in SER.

C520..0Approved for both units in SER.

C521..15OUTSTANDING ISSUE on staff review of sensitivity study of required safety valve flow rate versus trip parameterTVA letter dated April 18, 1983, provided the safety valve sizing information and information on differences with the reference plant. Staff closed issue in SSER2.--------------------In SSER15, the staff stated that subject to resolution of NUREG-737 Items II.D.1 (performance testing of relief and safety valves) and II.D.3 (indication of relief and safety valve position), overpressure protection at hot operating conditions will comply with the guidelines of SRP 5.2.2 and requirements of GDC 15. They noted that these items were found to be acceptable.

C522..010Approved for both units in SER.

C523..* = See last page for status code definition.

Page 26 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 16LICENSE CONDITION - Inservice inspection (ISI) programThe ISI program is required to be submitted within 6 months of the date of issuance of the operating license. The applicable ASME Code edition and addenda are determined by reference to 50.55a(b) 12 months preceding the date of issuance of the OL. The staff reiterated this in SSER10. In SSER12, the LICENSE CONDITION was resolved by a TVA commitment to submit the program within six months after receiving the operating license. Unit 2 Action:

Submit Unit 2 ISI program.

OUTSTANDING ISSUE - Unit 2 PSI program submitted April 30, 1990, with a partial listing of relief requests. This item tracked the staff review.In the SER, the preservice inspection program was still under review. NRC reviewed the Unit 1 PSI program in SSERs 10, 12, and 16. Unit 2 Action:

Submit Unit 2 PSI program.

REVISION 03 UPDATE:

Preservice Inspection Plan, Program No. WBN-2 PSI, Revision 3 was submitted to the NRC on June 17, 2010 (ADAMS Accession No. ML101680561).--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 05 UPDATE:

Corrected status from "O" to "S." S524..0521In SSER9, the staff stated that since the UHI system has been eliminated from the WB design , the previous discussion of this system in the SER no longer applies, but the conclusions reached in the SER were still valid. In SSER11, the staff reviewed valve stem leakage and stated that the staff's prior conclusions about valve stem leakage were not affected. In SSER12, the staff retracted the requirement identified in the SER that if leakage is alarmed and confirmed in a flow path with no indicators, then the Technical Specifications require a water inventory material balance be initiated within one hour. The staff also provided a clarification of SER wording related to detection of intersystem leakage through check valves and stated that this did not change prior staff conclusions and the reactor coolant pressure boundary system remains acceptable.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:In SSER21 the status is Open (NRR).

O525..02* = See last page for status code definition.

Page 27 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 16In SSER16, the staff reviewed the analysis of the RPV and internal components and found the use of the WECAN computer code acceptable.

C526..010Approved for both units in SER.

C530..21The staff reviewed TVA's submittal on reactor vessel irradiation in SSER11 and stated that the WB reactor vessels acceptably satisfy the requirements of 10 CFR 50.61. In SSER14, the staff determined that TVA complied with all the requirements in the current Appendix G, 10 CFR Part 50 without exemptions and the previously approved exemptions were no longer needed.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O531..0216OUTSTANDING ISSUE - P-T limits for Unit 2 not provided. Staff will review as part of Unit 2 Technical Specifications.In the original 1982 SER, NRC indicated that the review of the Unit 2 P-T limits would be completed as part of the review of the Unit 2 Technical Specifications. In SSER16, the staff found the pressure temperature limits methodology and the pressure temperature limits report for Unit 1 acceptable.Unit 2 action: Submit P-T limits.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.WCAP-17035-NP "Watts Bar Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation" was submitted with the TS.S532..02* = See last page for status code definition.

Page 28 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 0OUTSTANDING ISSUE for staff to complete evaluation of Unit 2 after receipt of P-T limitsIn the original 1982 SER, NRC indicated that the review of the Unit 2 P-T limits would be completed as part of the review of the Unit 2 Technical Specifications. Unit 2 action: Submit P-T limits.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.WCAP-17035-NP "Watts Bar Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation" was submitted with the TS.S533..020Approved for both units in SER.

C540..0Approved for both units in SER.

C541..45.4.2.2: OUTSTANDING ISSUE for staff to evaluate TVA's proposed resolution to concerns about flow induced vibrations in Model D-3 SGs pre-heat regionIn the original 1982 SER, the staff concluded that because of the generic problem of tube degradation caused by flow induced vibration in Westinghouse model D steam generators, operation would be limited to 50%. In SSER1, the staff continued to monitor activities associated with proposed modifications to the pre-heater region of the SGs to reduce impingement of water on tubes in this area and eliminate the vibration responsible for wear of the SG tubes. TVA's May 27, 1983, letter committed to implement the NUREG-0966 modifications to address this. In SSER4, the staff concluded the modification was acceptable to operate at 100%. In a letter dated December 17, 2008, TVA confirmed that these modifications were performed for WBN Unit 2.

C542..01* = See last page for status code definition.

Page 29 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21CONFIRMATORY ISSUE to verify installation of an RHR flow alarm and proper function of dump valves when actuated manuallyIn the SER, staff accepted TVA's commitment to provide, before startup, an RHR flow alarm to alert the operator to initiate alternate cooling modes in the event of loss of RHR pump suction. SSER2 resolved testing of dump valves. The staff verified that the alarm had been installed in SSER5, resolving the confirmatory issue.Unit 2 action: Verify alarm installation.

CONFIRMATORY ISSUE involving natural circulation test to demonstrate ability to cool down and depressurize the plant, and that boron mixing is sufficient under such circumstances; or, if necessary, other applicable tests before startup after first refuelingBranch Technical Position requires a natural circulation test with supporting analysis to demonstrate the ability to cool down and depressurize the plant and that boron mixing is sufficient. Comparison with performance of previously tested plants of similar design is acceptable, if justified. July 11, 1991, TVA letter, for both units, provided an assessment of the acceptability of the Diablo Canyon natural circulation tests to WBN. In SSER10, the NRC found the methods and conclusions acceptable. The staff corrected the wording in SSER10 in SSER11 and stated that this did not alter the conclusion reached.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (NRR).

O543..020Approved for both units in SER.

C544..21LICENSE CONDITION - NUREG-0737, II.B.1, "Reactor Coolant System Vents" - In the original SER, the NRC found TVA's commitment to install reactor coolant vents acceptable pending verification. In SSER2, the staff found venting guidelines acceptable. Installation was completed for Unit 1 only in SSER5 (IR 390/84-37) and the staff stated that the LC was no longer necessary. In SSER12, the staff included the safety evaluation for the RCSV system. The staff concluded that the high point vent system was acceptable subject to satisfactory completion of seven items that were described as on-going or planned activities associated with completion of the WB licensing process. They stated that none required additional review with respect to the SER nor would they change the SER, provided they were satisfactorily completed. TVA was asked to submit a letter prior to receipt of an OL stating how and when these items were completed. The staff stated that when these items were satisfactorily implemented, the RCSV system would be acceptable. Unit 2 Action: Verify installation of reactor coolant vents.

REVISION 02 UPDATE:

The status in SSER21 is Open (Inspection).

CI545..02* = See last page for status code definition.

Page 30 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 0Approved for both units in SER.

C600..0Approved for both units in SER.

C610..0Approved for both units in SER.

C611..0Approved for both units in SER.

C612..0Approved for both units in SER.

C613..0Approved for both units in SER.

C620..* = See last page for status code definition.

Page 31 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 216.2.1.1: CONFIRMATORY ISSUE involves reviewing analysis that ensures that containment external pressure will not exceed design value of 2.0 psiIn the original 1982 SER, NRC indicated it would confirm the contention that containment external pressure transients could not exceed the design value of 2.0 psig. TVA submitted the information June 4, 1982. In SSER3, NRC concluded that the design provided adequate protection against damage from external pressure transients.--------------------

In SSER5, the staff reviewed a revised long term containment analysis for the design basis LOCA in support of a proposed reduction in the limit for minimum allowable weight of ice in the condenser and found it acceptable.

Additionally, the staff verified that containment pressure and water level monitors were installed in Unit 1. Thus, License Conditions 6d and 6e were resolved (these are discussed with the other NUREG-0737 issues).In SSER7, the staff resolved their concerns regarding local temperatures near MSLBs inside containment and their impact on equipment qualification.In SSER12, the staff reviewed TVA's basis for deleting requirements for a 20,000 ppm boron concentration in the boron injection tank and determined that this would not significantly affect the environmental response of the containment or the safe shutdown equipment therein.In SSER14, the staff reviewed revisions to a number of containment design parameters and concluded that none affect conclusions reached in the SER or supplements.In SSER15, the staff reviewed the containment barrier seals and associated surveillance requirements and concluded that a revised divider barrier seal surveillance program was appropriate for Unit 1.Unit 2 Action: Review Unit 2 Technical Specifications with respect to divider barrier seal surveillance program.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.TS 3.6.13 provides the Limiting Condition for Operation for Divider Barrier Integrity.

S621..02* = See last page for status code definition.

Page 32 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21In SSER7, the staff determined that hot standby was an acceptable mode following a main steamline break and the containment cooling system modifications were acceptable.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (NRR).

TVA's September 26, 2008, letter proposed the use of the Unit 1 approach to resolve the Containment Cooling Special Program .In SSER21, the Containment Cooling SP was resolved. Completion of the Containment Cooling SP is tracked under 23.3.2.

CI622..0216In SSER16, the staff reviewed Amendment 89 to the FSAR and deletion of the high-radiation signal from the auxiliary building exhaust vent monitors and found it acceptable.

C623..01* = See last page for status code definition.

Page 33 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21CONFIRMATORY ISSUE to install safety grade isolation valves on 1" chemical feed lines joining feedwater lines to main steam line.LICENSE CONDITION - Modification of chemical feedlines In the original 1982 SER, the containment isolation provisions for the main and auxiliary feedwater lines, feedwater bypass lines and the chemical feedlines to the steam generators did not meet GDC 57. This was resolved by FSAR Amendment 55. In SSER5, the NRC concluded that the containment isolation provisions for the main and auxiliary feedwater lines, feedwater bypass lines and the chemical feedlines were acceptable.--------------------OUTSTANDING ISSUE for NRC to complete review of information provided by TVA to address Containment Purging During Normal Plant OperationLICENSE CONDITION - Containment isolation dependabilityIn 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 TER 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.--------------------

OUTSTANDING ISSUE involving containment isolation using closed systems This outstanding issue was opened in SSER7. In SSER12, the NRC concluded that the systems in question were "closed loops outside containment" and reaffirmed the previous conclusion of acceptability.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (Inspection).Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.TS Surveillance Requirement 3.6.3.7 requires verification that the valves are "blocked to restrict the valve from opening > 50 degrees." S624..02* = See last page for status code definition.

Page 34 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21OUTSTANDING ISSUE for review of TVA provided additional information relative to discussion added to FSAR to address analysis of the production and accumulation of hydrogen within containment following onset of a LOCAIn the original 1982 SER, NRC indicated that additional information was required concerning the analysis of the production and accumulation of hydrogen within the containment during a design basis LOCA. This information was provided in FSAR amendments and evaluated by NRC in SSER4. In SSER4, the NRC concluded that the design of the combustible gas control system was acceptable and the outstanding issue closed.Unit 2 Action:

The hydrogen recombiners will be removed from the Unit 2 design and licensing basis based on 10 CFR 50.44 (final rule September 16, 2003) and abandoned in place.This portion has a status of Open.--------------------LICENSE CONDITION - (6f) Accident monitoring instrumentationII.F.1 - containment hydrogenIn SSER5, NRC closed the LICENSE CONDITION for Unit 1 only (IR 390/84-85). Unit 2 Action:

Verify installation of containment hydrogen accident monitoring instrumentation.This portion has a status of Closed/Implementation only per NRC May 28, 2008, letter.--------------------

LICENSE CONDITION - (9) Hydrogen control measures In the original 1982 SER, an LC was raised to track resolution of Unresolved Safety Issue A-48, "Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment." In SSER8, the NRC reviewed the hydrogen mitigation system (igniters) and concluded it met the requirements of the final rule {10 CFR 50.44(c)(3)}.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (NRR).

Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009. This amendment deleted the hydrogen recombiners from the Unit 2 FSAR.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 04 UPDATE:EDCR 52329 was initiated to abandon in place Unit 2 hydrogen recombiners.

Technical Specifications (TS) / TS BASES 3.6.7 (Hydrogen Recombiners) were deleted in Developmental Revision B which was submitted on February 2, 2010.

S625..04* = See last page for status code definition.

Page 35 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21In SSER4, the staff approved exemption from certain requirements of Appendix J to 10 CFR 50 for both units. In SSER19, the staff found a revised schedule for the exemption approved in SSER4 acceptable.In SSER5, the staff found there was no radiological consequence to an increase in the bypass leakage rate for the emergency gas treatment system and found the increase acceptable.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O626..024CONFIRMATORY ISSUE for TVA to confirm that the lowest temperatures which will be experienced by the limiting materials of the reactor containment pressure boundary under the conditions cited by GDC 51 will be in compliance with the temperatures identified in the staff's analysis of fracture toughness requirements for load bearing component of the containment systemIn SSER4, NRC reviewed the confirmatory information submitted and concluded for both units that the reactor containment pressure boundary materials will behave in a non-brittle manner and the requirements of GDC 51 were satisfied. NRC provided the technical basis in Appendix H of SSER4.

C627..0Approved for both units in SER.

C630..0111OUTSTANDING ISSUE - involving removal of upper head injection systemThe Upper Head Injection (UHI) system design was approved in the original 1982 SER. TVA letter dated September 19, 1985, informed NRC that UHI would not be installed on Unit 2. The staff stated in SSER6 that they were continuing to review TVA's submittal. In SSER7, NRC concluded it was acceptable to delete UHI from both units. In SSER11, the staff stated that the revision of the design code for ECCS piping from B31.1 to ASME Section III did not change the conclusions made in the SER and previous SSERs.--------------------

Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.

This amendment revised the FSAR to address the application of RFA-2 fuel.

S631..02* = See last page for status code definition.

Page 36 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 5In SSER5, the staff reviewed TVA's approach to maintaining ECCS effectiveness by ensuring that no single failure would be able to energize the coils of the valve operators and found it acceptable. The staff also reviewed TVA's response to Issue 4 of NUREG-0138, Resequencing of ECCS loads following SI signal reset followed by a loss of offsite power.--------------------

Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.

This amendment revised the FSAR to address the application of RFA-2 fuel.

S632..029OUTSTANDING ISSUE - involving containment sump screen designIn the original 1982 SER, the staff approved the proposed sump design in the FSAR. A deviation between the installed and proposed design was discovered during an NRC inspection. In SSER9, the staff concluded that the as-installed sump screen was acceptable.--------------------CONFIRMATORY ISSUE - provide a detailed survey of insulation material that could be debris post-LOCAIn the original 1982 SER, NRC found the design of the containment sump against debris acceptable subject to the acceptability of a detailed survey of insulation materials. In SSER2, the NRC review of the survey confirmed the staff's initial conclusion that the design to provide protection against sump debris was acceptable.--------------------

Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.

This amendment revised the FSAR to address the application of RFA-2 fuel.

S633..020Approved for both units in SER.

C634..0Closure based on 6.3.1 to 6.3.3.

O635..01* = See last page for status code definition.

Page 37 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21In SSER5, the staff concluded that removal of the main control room air intake chlorine detector was acceptable. In SSER11, they stated that FSAR Amendment 69 on control room isolation did not change previous conclusions. In SSER16, the staff concluded that the control room design satisfied the requirements of GDC 19 and the guidelines of NUREG-0737, Item III.D.3.4.In SSER18, the staff reviewed updated control room air flow rate data and dose analysis, as provided in Amendment 90, and determined that the changes did not affect conclusions reached in the SER or its supplements.See 18.1.0 also.

REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O640..020Approved for both units in SER.

C650..21In SSER5, the staff found the Reactor Building Purge Ventilation System acceptable.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (NRR).

O651..020Approved for both units in SER.

C652..0Approved for both units in SER.

C653..0Approved for both units in SER.

C654..* = See last page for status code definition.

Page 38 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 15OUTSTANDING ISSUE on additional information required on preservice inspection program and identification of plant specific areas where ASME Code Section XI requirements cannot be met and supporting technical justificationNRC reviewed the preservice inspection program (PSI) for Unit 1 only in SSER10 and on the basis of a TVA commitment to submit an inservice inspection program within 6 months after receiving an operating license, considered a proposed LC for an ISI no longer required. In SSER15, the staff reviewed Revisions 24 and 25 to the preservice inspection program and concluded that the changes included therein were acceptable. Unit 2 Action: Submit Unit 2 PSI program.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 03 UPDATE:Preservice Inspection Plan, Program No. WBN-2 PSI, Revision 3 was submitted to the NRC on June 17, 2010 (ADAMS Accession No. ML101680561).--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 05 UPDATE:

Corrected status from "O" to "S." S660..050Approved for both units in SER.

C700..0Approved for both units in SER.

C710..16In SSER13, NRC reviewed the Eagle-21 upgrade for WBN Unit 1 only. TVA letter dated December 5, 2007, informs NRC of intent to use Eagle-21 for Unit 2. NRC requested additional information December 27, 2007. TVA provided the requested information by letter dated February 28, 2008. By letter dated May 7, 2008, NRC provided a list of specific issues to be addressed in a future amendment application for Eagle-21 for WBN Unit 2. Unit 2 Action: Provide the additional information for NRC review.

By letter dated August 21, 1995 for both units, TVA provided additional justification for a deviation from Position C.6(a) of RG 1.118 "Periodic Testing of Electrical Power and Protection Systems" Revision 2. In SSER16, the NRC found the deviation acceptable.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.

S711..02* = See last page for status code definition.

Page 39 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 0Approved for both units in SER.

C712..15In the SER, NRC indicated that a review of the setpoint methodology would be performed with a review of the Technical Specifications. In SSER4, NRC reviewed the methodology used to determine setpoints for Watts Bar Units 1 and 2 and determined that it was acceptable. By letter dated July 29, 1994, for both units, TVA submitted a topical report titled "Westinghouse Setpoint Methodology for Protection Systems, Watts Bar Units 1 and 2, Eagle 21 Version" (WCAP-12096, Revision 6). In SSER15, the NRC concluded the setpoint methodology was acceptable based on (1) previous acceptance of Westinghouse setpoint methodology at other plants, (2) the similarity between the Watts Bar and previously approved designs such as Sequoyah, and (3) the Watts Bar setpoint methodology is in compliance with RG 1.105 and ISA S6704.Staff requested discussion of methodology for determining, setting, and evaluating as-found setpoints for drift susceptible instruments. Unit 2 action: Resolve this issue using the BFN TS-453 precedent (see NRC ML061680008).--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Developmental Revision B of the Unit 2 Technical Specifications (TS) and TS Bases was submitted on February 2, 2010.As part of the submittal, TVA incorporated TSTF-493, Revision 4, "Clarify Application of Setpoint Methodology for LSSS Functions," into Section 3.3 of the TS and TS Bases.TVA submitted WCAP-17044, "Westinghouse Setpoint Methodology for Protection Systems" on February 5, 2010.

S713..020Approved for both units in SER.

C720..15In SSER13, NRC reviewed the Eagle-21 upgrade for WBN Unit 1 only. In SSER15, the NRC reviewed the WBN Unit 1 EMI/RFI report and concluded that the EMI/RFI issue was resolved for WBN Unit 1. TVA letter dated December 5, 2007, informs NRC of intent to use Eagle-21 for Unit 2. NRC requested additional information December 27, 2007. TVA provided the requested information by letter dated February 28, 2008. By letter dated May 7, 2008, NRC provided a list of specific issues to be addressed in a future amendment application for Eagle-21 for WBN Unit 2. Unit 2 Action: Provide the additional information for NRC review.

REVISION 02 UPDATE:

TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.

S721..02* = See last page for status code definition.

Page 40 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 0Approved for both units in SER.

C722..0Approved for both units in SER.

C723..0Approved for both units in SER.

C724..21CONFIRMATORY ISSUE - address IEB 79-21 to alleviate temperature dependence problem associated with measuring SG water levelIn SSER2, NRC accepted TVA's commitment to insulate the steam generator water level reference legs to alleviate the temperature dependence problem. By letter dated July 27, 1994, TVA submitted an evaluation for both units and determined that it was not necessary to insulate the SG reference legs at WBN. In SSER14, NRC concurred with TVA's assessment to not insulate the steam generator water level instrument reference leg.Unit 2 Action:

Update accident calculation.

REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O725..0213In SSER13, NRC reviewed the Eagle-21 upgrade for WBN Unit 1 only. TVA letter dated December 5, 2007, informs NRC of intent to use Eagle-21 for Unit 2. NRC requested additional information December 27, 2007. TVA provided the requested information by letter dated February 28, 2008. By letter dated May 7, 2008, NRC provided a list of specific issues to be addressed in a future amendment application for Eagle-21 for WBN Unit 2. Unit 2 Action: Provide the additional information for NRC review.

"CONCLUSIONS" left open until all actions in subsection are closed.

REVISION 02 UPDATE:

TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.

S726..02* = See last page for status code definition.

Page 41 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 13In SSER13, NRC reviewed the Eagle-21 upgrade for WBN Unit 1 only. TVA letter dated December 5, 2007, informs NRC of intent to use Eagle-21 for Unit 2. NRC requested additional information December 27, 2007. TVA provided the requested information by letter dated February 28, 2008. By letter dated May 7, 2008, NRC provided a list of specific issues to be addressed in a future amendment application for Eagle-21 for WBN Unit 2. Unit 2 Action: Provide the additional information for NRC review.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.

S730..0214In SSER13, NRC reviewed the Eagle-21 upgrade for WBN Unit 1 only. TVA letter dated December 5, 2007, informs NRC of intent to use Eagle-21 for Unit 2. NRC requested additional information December 27, 2007. TVA provided the requested information by letter dated February 28, 2008. By letter dated May 7, 2008, NRC provided a list of specific issues to be addressed in a future amendment application for Eagle-21 for WBN Unit 2. Unit 2 Action: Provide the additional information for NRC review.

In SSER14, NRC reviewed TVA's FSAR amendment 81 sect ion 7.3.2.2.6, with respect to a deviation from IEEE Standard 279-1971. Manual initiation of both steamline isolation and switchover from injection to recirculation following a loss-of-primary-coolant accident are performed at the component level only. In SSER14, NRC agreed with TVA's justification.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.

S731..022CONFIRMATORY ISSUE is commitment to make a design change to provide protection that prevents debris from entering containment sump level sensorsIn the original SER, staff identified a concern that debris in the containment sump could block the inlets to the differential pressure transmitters and result in a loss of the permissive signal to the initiation logic for the automatic switchover from the injection to the recirculation mode of the emergency core cooling system. In a September 15, 1983, letter TVA notified NRC that the level sensors had been moved from inside the sump wall to outside the sump wall with the sense line opening protected by a cap with small holes. Staff closed the issue in SSER2.

C732..0Approved for both units in SER.

C733..0Approved for both units in SER.

C734..* = See last page for status code definition.

Page 42 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21CONFIRMATORY ISSUE - perform confirmatory tests to satisfy IEB 80-06 (to ensure that no device will change position solely due to reset action) and staff review of electrical schematics for modifications that ensure that valves remain in emergency mode after ESF resetIn the original SER, staff concluded that the design modifications for Bulletin 80-06 were acceptable subject to review of the electrical schematics that were not available at the time. In SSER3, the staff found the modifications acceptable and closed the confirmatory issue.Unit 2 Action:

Perform verification during preoperational testing.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (Inspection).

CI735..0213In SSER13, NRC reviewed the Eagle-21 upgrade for WBN Unit 1 only. TVA letter dated December 5, 2007, informs NRC of intent to use Eagle-21 for Unit 2. NRC requested additional information December 27, 2007. TVA provided the requested information by letter dated February 28, 2008. By letter dated May 7, 2008, NRC provided a list of specific issues to be addressed in a future amendment application for Eagle-21 for WBN Unit 2. Unit 2 Action: Provide the additional information for NRC review.

"CONCLUSIONS" left open until all actions in subsection are closed.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.

S736..020Approved for both units in SER.

C740..0Approved for both units in SER.

C741..21By letter dated September 26, 1985, TVA requested a deviation from 10 CFR Part 50, Appendix R, Section III.L.2.d for use of the SG saturation temperatures to approximate reactor coolant system cold leg temperatures. This was approved for both units by SE dated May 17, 1991. The SE was discussed in SSER7. The staff concluded that this was an acceptable deviation.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (NRR).

O742..02* = See last page for status code definition.

Page 43 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 0Approved for both units in SER.

C743..0Approved for both units in SER.

C750..0Approved for both units in SER.

C751..21OUTSTANDING ISSUE involving RG 1.97 instruments following course of an accidentIn the original 1982 SER, the staff stated that WBN did not use RG 1.97, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plants and Environs Conditions During and Following an Accident," for the design because the design predated the RG. In SSER7, an outstanding issue was opened. TVA provided NRC information on exceptions to RG 1.97. A detailed review was performed for both units (Appendix V of SSER9). The staff concluded that WBN conforms to or has adequately justified deviations from the guidance of RG 1.97, Revision 2. TVA submitted additional deviations for both units in letters dated May 9, 1994, and April 21, 1995. In SSER14 and SSER15, the additional deviations to RG 1.97 were reviewed and accepted by NRC.NUREG-0737, II.F.1.2, ""Accident Monitoring Instrumentation" - Reviewed in SSER9.Unit 2 Actions: Install Noble gas, Iodine / particulate sampling, and Containment High Range Monitors. CI in NRC May 28, 2008, letter.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (NRR).

O752..0221B 79-27, "Loss of Non-class 1E I&C Power System Bus During Operation" - TVA responded to the Bulletin on March 1, 1982. Reviewed in 7.5.3 of the original 1982 SER. Unit 2 Action: Issue appropriate emergency procedures.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (Inspection).

CI753..02* = See last page for status code definition.

Page 44 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21"CONCLUSIONS" left CI until all items in subsection are closed.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (Inspection).

CI754..020Approved for both units in SER.

C760..0Approved for both units in SER.

C761..0Approved for both units in SER.

C762..0Approved for both units in SER.

C763..0Approved for both units in SER.

C764..4CONFIRMATORY ISSUE - install switches on the main control board for the operator to manually arm this system (overpressure protection provided by pressurizer PORVs)In the original 1982 SER, the staff found the design of the overpressure protection during low temperature features acceptable pending review of the drawings and FSAR description. In SSER4, the staff documented completion of the review and closed the confirmatory issue.

C765..0Approved for both units in SER.

C766..0Approved for both units in SER.

C767..0Approved for both units in SER.

C768..4Approved for both units SER subject to completion of Confirmatory Issue in 7.6.5.C769..0Approved for both units in SER.

C770..0Approved for both units in SER.

C771..* = See last page for status code definition.

Page 45 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 13LICENSE CONDITION - Status monitoring system, Bypassed and Inoperable Status Indication (BISI) In the original 1982 SER, the staff requested TVA address RG 1.47, "Bypassed and Inoperable Status Indications for Nuclear Power Plant Safety Systems." TVA addressed RG 1.47 by letters for both units dated January 29, 1987, and October 22, 1990. In SSER7, the staff documented completion of the review and closed the issue. By letter dated February 18, 1994, for both units, TVA submitted a re-evaluation of BISI that excluded components that would not be rendered inoperable more than once a year in accordance with RG 1.47 position C.3(b). In SSER13, NRC reviewed the revision and concluded that it was acceptable.

C772..010Approved for both units in SER.

C773..0Approved for both units in SER.

C774..0Approved for both units in SER.

C775..0Approved for both units in SER.

C776..0Approved for both units in SER.

C777..21ATWS Mitigation design was reviewed and approved for both units by a Safety Evaluation Report issued December 28, 1989. This SER is also in Appendix W of SSER9. Outstanding Issue was Technical Specifications requirements. In SSER14, NRC reviewed the revision of FSAR Figure 7.3-3 for the AMSAC automatic initiation signal to start the turbine driven and motor driven auxiliary feedwater pumps and considered the issue resolved. Unit 2 Action:

Address in Technical Specifications as appropriate.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.ATWS is not addressed in either the Unit 1 TS or the Unit 2 TS; nor is it addressed in the Standard TS (NUREG-1431).

S778..020Approved for both units in SER.

C780..* = See last page for status code definition.

Page 46 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21NUREG-0737, II.D.3, "Valve Position Indication" - The design was reviewed in the original 1982 SER and found acceptable pending confirmation of installation of the acoustic monitoring system. In SSER5 (IR 390/84-35), the staff closed the LICENSE CONDITION for Unit 1 only. By letter dated November 7, 1994, for both units, TVA provided a revised response for NUREG-0737 Item II.D.3. TVA revised the design by relocating the accelerometers for valve position indication to downstream of the relief valves. This change was reviewed in SSER14. The revision did not change the function of the position indication hardware and did not alter the previous review.Unit 2 Action: Verify installation of the acoustic monitoring system to PORV to indicate position. CI in NRC May 28, 2008, letter.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O781..0221NUREG-0737, II.E.1.2, "Auxiliary Feedwater System Initiation and Flow Indication"Unit 2 Action: Complete procedures and qualification testing.

REVISION 02 UPDATE:

The status in SSER21 is Open (Inspection).

CI782..0221NUREG-0737, II.K.3.9, "Proportional Integral Derivative Controller Modification" - Reviewed in original 1982 SER. Unit 2 Action: Set the derivative time constant to zero.

REVISION 02 UPDATE:

The status in SSER21 is Open (Inspection).

O783..02* = See last page for status code definition.

Page 47 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21NUREG-0737, II.K.3.10, "Anticipatory Trip At High Power"In SSER4, NRC concluded that TVA had adequately addressed the requirements of NUREG-0737 Item II.K.3.10 for removal of the anticipatory reactor trip on turbine trip at or below 50% power. Unit 2 Action: Unit 2 Technical Specifications and surveillance procedures will address this issue.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (Inspection).Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.Items 14.a. (Turbine Trip - Low Fluid Oil Pressure) and 14.b. (Turbine Trip - Turbine Stop Valve Closure) of TS Table 3.3.1-1 are the trips of interest. The table and the Bases for these items state that below the P-9 setpoint, these trips do not actuate a reactor trip.Per item 16.d. (Power Range Neutron Flux, P-9) of TS Table 3.3.1-1, the Nominal Trip Setpoint for P-9 is"50% RTP" and the Allowable Value is "< 52.4% RTP."

S784..020NUREG-0737, II.K.3.12, "Confirm Existence of Anticipatory Reactor Trip Upon Turbine Trip"Approved for both units in the SER C785..01Area not addressed in 1981 Standard Review Plan.

NA790..0Approved for both units in SER.

C800..0Approved for both units in SER.

C810..0Approved for both units in SER.

C820..13Approved for both units in SER. In SSER13, NRC reviewed TVA's analysis of grid stability on loss of both units. The NRC conclusions in the SER remained valid.C821..01* = See last page for status code definition.

Page 48 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 218.2.2.1 CONFIRMATORY ISSUE - document additional information in FSAR on control power supplies and distribution system for the Watts Bar Hydro Plant SwitchyardIn the original 1982 SER, NRC concluded that the offsite power system circuits at the Watts Bar Hydro Plant Switchyard met GDC 17 pending documentation in the FSAR. The information was added to the FSAR. In SSER2, NRC closed the issue. In SSER13, the staff reviewed revised information incorporated into FSAR amendment 71 for both units and concluded that it supported the original conclusion in SSER2.---------------------

8.2.2.2 OUTSTANDING ISSUE involving compliance of design changes to the offsite power system with GDC 17 and 18.In SSER2 and 3, NRC continued the review of the offsite electrical power system. By letter dated June 20, 1991, for both units, NRC requested additional information on Section 8 of the FSAR. TVA responded for both units by letter dated September 13, 1991. In SSER13, the NRC reviewed the design changes to minimize the probability of losing all AC power, compliance with GDC 17 and minimizing the probability of a two unit trip following a one unit trip. These issues were resolved in SSER13. Additional review was done in SSER14, but the conclusions remained valid.--------------------

8.2.2.3 Compliance with GDC 17 for the Duration of the Offsite System ContingenciesBy letter dated June 20, 1991, for both units, NRC requested additional information on Section 8 of the FSAR. TVA responded for both units by letter dated September 13, 1991. In SSER13, NRC reviewed the load shed scheme described in FSAR amendment 71 that reduces loads from common station service transformers A and B including contingency for both units trip and a 161-kV supply contingency. In SSER15, NRC determined that entering the LCO for one offsite circuit inoperable was appropriate. No open items were identified.----------------8.2.2.4 Minimizing the Probability of a Two-Unit Trip Following a One-Unit TripBy letter dated June 20, 1991, for both units, NRC requested additional information on Section 8 of the FSAR. TVA responded for both units by letter dated September 13, 1991. In FSAR amendment 71, TVA described the transfer of power sources on trip of a unit's main generator. In SSER13, NRC evaluated the design and determined that the concern was resolved.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O822..020Approved for both units in SER.

C823..0Approved for both units in SER.

C824..* = See last page for status code definition.

Page 49 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 0Approved for both units in SER.

C830..* = See last page for status code definition.

Page 50 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 208.3 Fifth Diesel GeneratorIn SSER10, NRC reviewed the design of the fifth diesel generator. In SSER19, NRC accepted TVA's commitment to perform modifications and surveillances including preoperational testing before declaring the fifth diesel generator operable as a replacement for one of the four diesel generators. TVA stated in a submittal dated July 28, 1993, that they did not plan to place the additional diesel generator in service.---------------------

8.3.1.1: CONFIRMATORY ISSUE - incorporate new design that provides dedicated transformer for each preferred offsite circuit in FSARIn the original 1982 SER, NRC concluded that the offsite power system with a dedicated transformer for each preferred offsite circuit met GDC 17 pending documentation in the FSAR. The information was added to the FSAR. In SSER2, NRC closed the issue. In SSER13, NRC reviewed additional changes though FSAR amendment 75 and concluded that the design was acceptable.---------------------

8.3.1 DG Starting and Control Circuit Logic In SSER10, NRC reviewed the DG starting and control circuit logic. No open items were identified.---------------------

8.3.1.2 Low and Degraded Grid Voltage Condition In the SER, NRC stated they would verify the adequacy of TVA's analysis regarding Branch Technical Position PSB-1 once preoperational testing was completed. In SSER13, the NRC reviewed information on the load shed and diesel start relays. In SSER14 NRC clarified the requirements. In SSER20, NRC reviewed the preoperational test for Unit 1. Unit 2 Action: Include the setpoint in the Technical Specifications for the load shed relays and similar minimum limits for the diesel start relays.---------------------8.3.1.6: CONFIRMATORY ISSUE - provide diesel generator reliability qualification test reportIn SSER2, NRC indicated that it would verify DG qualification testing. TVA provided a copy of the DG qualification test report. In SSER7, the NRC concluded that the DGs had been satisfactorily tested in accordance with IEEE 387-1977.---------------------8.3.1.6: LICENSE CONDITION (12) - Diesel generator reliability qualification testing at normal operating temperatureIn the original 1982 SER, NRC required that the capability of the DGs to start at normal temperature be demonstrated. TVA's August 31, 1983, letter confirmed tests had been performed on a DG identical to those at WBN. In SSER2, NRC closed the issue.---------------------8.3.1.7 Possible Interconnection Between Redundant Divisions Through Normal and Alternate Power to the Battery Charger S831..02* = See last page for status code definition.

Page 51 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. By letter dated June 20, 1991, for both units, NRC requested additional information on Section 8 of the FSAR. TVA responded for both units by letter dated September 13, 1991. In SSER13, the NRC reviewed the use of alternate feeders to the battery chargers and inverters and concluded a Technical Specification surveillance for monitoring the position of these supply breakers resolved the item. Unit 2 Action:

Include the surveillance requirement in the Technical Specifications.

8.3.1.10 No-load Operation of the Diesel Generator By letter dated June 20, 1991, for both units, NRC requested additional information on Section 8 of the FSAR. TVA responded for both units by letter dated September 13, 1991. In SSER13, the NRC reviewed the information provided and concluded the issue was resolved. In SSER14, NRC added additional clarification but did not change the conclusions.---------------------

8.3.1.11 Test and Inspection of the Vital Power System By letter dated June 20, 1991, for both units, NRC requested additional information on Section 8 of the FSAR. TVA responded for both units by letter dated September 13, 1991. In SSER13, the NRC reviewed TVA's plan for test and inspection of the vital ac system and concluded the issue was resolved.---------------------

8.3.1.12 The Capability and Independence of Offsite and Onsite Sources When Paralleling During TestingBy letter dated June 20, 1991, for both units, NRC requested additional information on Section 8 of the FSAR. TVA responded for both units by letter dated September 13, 1991. In SSER13, the NRC reviewed the Emergency Diesel Generators response to a loss-of-offsite-power (LOOP). TVA submitted additional information for both units by letters dated February 7, 1994 and June 29, 1994. In SSER14, NRC concluded that the issue was resolved.---------------------8.3.1.13 Use of an Idle Start Switch for Diesel Generators By letter dated June 20, 1991, for both units, NRC requested additional information on Section 8 of the FSAR. TVA responded for both units by letter dated September 13, 1991. In SSER13, the NRC reviewed the information presented on the local idle start switch and concluded the issue was resolved.---------------------

8.3.1.14 Master Fuse List Program In SSER9, NRC provided a safety evaluation of the Master Fuse List Special Program (SP) for Unit 1 (Appendix U). In SSER13, NRC referenced the evaluation.Unit 2 Action: Resolve the SP for WBN Unit 2 with the Unit 1 approach.

* = See last page for status code definition.

Page 52 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. ----------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

Revised "SSER18" to "SSER19" item 8.3 above to fix typographical error in Regulatory Framework.

Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.8.3.1.2: TS Table 3.3.5-1 provides Diesel Generator start and load shed relay trip setpoints and allowable values.8.3.1.7: TS surveillance requirements SR 3.8.4.3 and SR 3.8.7.1 provide surveillances to check the alignment of battery charger alternate feeder breakers and inverters.

8.3.1.14: TVA's September 26, 2008, letter proposed the use of the Unit 1 approach to resolve the Master Fuse List Special Program.In SSER21 the Containment Cooling SP was resolved. Completion of the Master Fuse List SP is tracked under 23.3.5.* = See last page for status code definition.

Page 53 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 148.3.2.2: LICENSE CONDITION - DC monitoring and annunciation systemIn SSER3, the staff determined that some items were omitted from the design of the DG DC monitoring and annunciation system. By letter dated June 20, 1991, for both units, NRC requested additional information on Section 8 of the FSAR. TVA responded for both units by letter dated September 13, 1991. In SSER13, NRC closed the issue.--------------------

8.3.2.4: CONFIRMATORY ISSUE - include diesel generator design analysis in FSARIn the original 1982 SER, staff indicated the design analysis for demonstrating compliance of the DGs with regulatory requirements and guidelines was acceptable pending incorporation of the analysis in the FSAR. The analysis was incorporated in the FSAR, and the issue closed in SSER2. By letter dated June 20, 1991, for both units, NRC requested additional information on Section 8 of the FSAR. TVA responded for both units by letter dated September 13, 1991. In SSER13, NRC indicated that the issue was resolved.---------------------

8.3.2.5 Non-safety Loads Powered from the DC Distribution System and Vital InvertersBy letter dated June 20, 1991, for both units, NRC requested additional information on Section 8 of the FSAR. TVA responded for both units by letter dated September 13, 1991. In SSER13, NRC indicated that the issue was resolved.---------------------8.3.2.5.1 Transfer of Loads Between Power Supplies Associated with the Same Load Group but Different UnitsBy letter dated June 20, 1991, for both units, NRC requested additional information on Section 8 of the FSAR. TVA responded for both units by letter dated September 13, 1991. In SSER13, NRC reviewed the information provided. Additional information was requested for both units by letter dated March 28, 1994. TVA responded for both units by letter dated June 29, 1994.

In SSER14, NRC indicated that the issue was resolved.---------------------

8.3.2.7 The Fifth Vital Battery System By letter dated June 20, 1991, for both units, NRC requested additional information on Section 8 of the FSAR. TVA responded for both units by letter dated September 13, 1991. In SSER13, NRC indicated that the issue was resolved.---------------------

8.3.2.8 Reenergizing the Battery Charger from the Onsite Power Sources Versus Automatically Immediately Following a Loss of Offsite PowerBy letter dated June 20, 1991, for both units, NRC requested additional information on Section 8 of the FSAR. TVA responded for both units by letter dated September 13, 1991. In SSER13, NRC indicated that the issue was resolved.C832..01* = See last page for status code definition.

Page 54 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 218.3.3.1.1: CONFIRMATORY ISSUE involving submergence of electrical equipment as result of a LOCAIn the original 1982 SER and SSER3, staff stated that the design for the automatic deenergizing of loads as a result of a LOCA would be verified as part of the site visit. During the August 1991, visit and in a letter for both units dated September 13, 1991, TVA committed to revise the FSAR. The information was added to the FSAR in amendment 71. In SSER13, NRC closed the issue.---------------------

8.3.3.1.3 Failure Analysis of Circuits Associated with Cables and Cable Splices Unqualified for SubmergenceBy letter dated June 20, 1991, for both units, NRC requested additional information on Section 8 of the FSAR. TVA responded for both units by letter dated September 13, 1991. In SSER13, NRC reviewed the submergence calculation and closed the issue.Unit 2 Action:

Revise calculation for WBN Unit 2.---------------------8.3.3.1.2: CONFIRMATORY ISSUE - verify design for bypass of thermal overload protective deviceIn the original 1982 SER, NRC indicated that the design for bypass of thermal overload protective devices on safety-related motor operated valves would be verified during the electrical drawing review. The staff subsequently reviewed the drawings and closed the issue in SSER2.---------------------8.3.3.1.4 Use of Waterproof Splices in Potentially Submersible Sections of Underground Duct RunsBy letter dated June 20, 1991, for both units, NRC requested additional information on Section 8 of the FSAR. TVA responded for both units by letter dated September 13, 1991. In SSER13 and 14, NRC raised a concern on splice usage in raceways. TVA submitted additional information for both units by letters dated November 18, 1994, and January 5, 1995. In SSER15, NRC found that TVA had adequately justified the acceptability of the installed splices at Watts Bar.---------------------

8.3.3.1.5 Dow Corning RTV-3140 Used to Repair Damaged Kapton Insulated ConductorsIn SSER15, NRC reviewed the use of RTV-3140. TVA submitted the technical basis for use in a December 6, 1994, letter for both units. TVA completed additional testing and told the NRC of the limited use of this repair method for both units by letter dated February 10, 1995. In SSER15, NRC found the use of RTV-3140 acceptable for the limited use described.---------------------8.3.3.1.6 Cable Damage Near Splices and Terminations In SSER16, NRC reviewed TVA's corrective action plan for Construction Deficiency Report 390/95-02 and found the limited inspections for damaged Class 1E cables to 10 CFR 50.49 installations acceptable. This was a WBN Unit 1 only CDR.

S833..02* = See last page for status code definition.

Page 55 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. ---------------------8.3.3.2: CONFIRMATORY ISSUE - revise FSAR to reflect requirements of shared safety systemsIn the original 1982 SER, the staff stated that the description and analysis of shared onsite AC and DC systems was under review but was acceptable pending revision of the FSAR. In SSER3, the confirmatory issue was left open to track additional information to be incorporated in the FSAR. In a letter dated September 13, 1991, TVA provided the additional information. In SSER13, NRC closed the issue. In SSER14, NRC added additional clarification.---------------------

8.3.3.2.2 Sharing of AC Distribution Systems and Standby Power Supplies Between Units 1 and 2In the SER and SSER3, NRC reviewed the design to the guidelines of RG 1.81 and determined it was acceptable pending revision to the FSAR. NRC noted discrepancies in the FSAR. By letter dated June 20, 1991, for both units, NRC requested additional information on Section 8 of the FSAR. TVA responded for both units by letter dated September 13, 1991. In SSER13, NRC closed the issue. ---------------------

8.3.3.2.3: CONFIRMATORY ISSUE for design of sharing raceway systems between unitsIn the original SER, NRC indicated that the design for sharing of raceway systems between units would be verified during the electrical drawing review. The staff confirmed that cable routing was in accordance with accepted separation criteria and closed the issue in SSER2.---------------------

8.3.3.2.4: LICENSE CONDITION - Possible sharing of DC control power to AC switchgearIn the original 1982 SER, staff required that all possible interconnections between redundant divisions through normal and alternate power sources to various loads be identified in the FSAR. TVA letter dated January 17, 1984, provided the information. NRC closed the issue in SSER3. ---------------------

8.3.3.3: LICENSE CONDITION - Testing of associated circuits In the original 1982 SER, staff required that protective devices used to isolate non-Class 1E from Class 1E circuits be of high quality commensurate with their importance to safety and be periodically tested. TVA letter dated January 17, 1984, provided the information. NRC closed the issue in SSER3. ---------------------8.3.3.3: LICENSE CONDITION - Testing of non-class 1E cables In the original 1982 SER, staff required that protective devices used to isolate non-Class 1E from Class 1E circuits be of high quality commensurate with their importance to safety and be periodically tested. TVA letter dated January 17, 1984, provided additional information. NRC closed the issue in SSER3.* = See last page for status code definition.

Page 56 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. ---------------------8.3.3.3 Physical Independence (Compliance with GDC 17)

By letter dated June 20, 1991, for both units, NRC requested additional information on Section 8 of the FSAR. TVA responded for both units by letter dated September 13, 1991. The information was incorporated into the FSAR by amendment 71. Surveillance requirements for the testing of protective devices used to protect Class 1E circuits from failure of non-Class 1E circuits were incorporated into the Technical Requirements Manual (TRM). This issue was closed based on review of the TRM in SSER13.Unit 2 Action:

Incorporate testing requirements into the Unit 2 TRM.

8.3.3.3 Physical Independence (Compliance with GDC 17)

In SSER13, NRC cited differences between RG 1.75 and the WBN design criteria (WB-DC-30-4). In SSER14, NRC continued the review. NRC requested additional information for both WBN units by letter dated March 28, 1994. TVA responded for both WBN units by letters dated July 29, 1994, January 11, 1995, and June 5, 1995. In SSER16, NRC found separation between open cable trays (including cables in free air) adequate.---------------------8.3.3.5.1 Compliance with Regulatory Guides 1.108 and 1.118 In SSERs 13, 14 and 15, NRC reviewed WBN compliance with RGs 1.108 and 1.118. In SSER13, NRC reviewed WBN's use of temporary jumper wires when portable test equipment is used during testing. The justification was documented in the FSAR. In SSER14 and 15, NRC reviewed Class 1E standby power system testing, testing DG full load rejection capability and non-class 1E circuitry for transmitting signals needed for starting DGs. NRC concluded that the features were appropriately tested.---------------------8.3.3.5.2: CONFIRMATORY ISSUE - incorporate commitment to test only one of four diesel generators at one timeIn the original 1982 SER, the NRC found the commitment to test DGs one at a time acceptable pending its incorporation into the FSAR. In SSER2, NRC reviewed the documentation and closed the issue.---------------------

8.3.3.5.3 Time Constraints for Stability of EDG During No-Load Startup TestingIn SSER16, NRC reviewed and approved changes to the no load emergency diesel generator testing surveillance requirements.Unit 2 Action: Incorporate into WBN Unit 2 TS surveillances.

8.3.3.6: CONFIRMATORY ISSUE involving evaluation of penetrations' ability to withstand failure of overcurrent protection deviceIn the original 1982 SER, staff required a reevaluation of the penetrations' * = See last page for status code definition.

Page 57 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. capability to withstand, without seal failure, the total range of available time-current characteristics assuming a single failure of any overcurrent protective device. In SSER3, staff found the results of the evaluation acceptable pending the information being incorporated in the FSAR. The staff reviewed the FSAR and closed the issue for both units in SSER7.---------------------

8.3.3.6: LICENSE CONDITION - Testing of reactor coolant pump breakersIn the original 1982 SER, staff required that the redundant fault current protective devices for the reactor coolant pump circuits meet RG 1.63. In SSER2, staff reviewed the design and concluded it met RG 1.63.---------------------

8.3.3.6 Compliance with GDC 50 By letter dated June 20, 1991, for both units, NRC requested additional information on Section 8 of the FSAR. TVA responded for both units by letter dated September 13, 1991. The information was incorporated into the FSAR in amendment 70. In SSER13, NRC indicated that the issue was resolved.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

Developmental Revision B of the Unit 2 Technical Specifications (TS) and Technical Requirements Manual (TRM) was submitted on February 2, 2010.8.3.3.3: TRM TR 3.8.1 specifies testing of circuit breakers that are used as isolation devices protecting 1E busses from non-qualified loads.8.3.3.5.3: TS Sections 3.8.1.7, 3.8.1.12, 3.8.1.15 and 3.8.1.21 require that voltage and frequency remain within specified limits following a fast start.Station Blackout (SBO) - SE for both units - March 18, 1993; SSE for both units - September 9, 1993. Unit 2 Action:

Implement SBO requirements.

CI840..Area not addressed in 1981 Standard Review Plan.

NA850..Area not addressed in 1981 Standard Review Plan.

NA851..10In SSER10, the staff completed its review of the additional DG building and that review is documented in Sections 9.2.1, 9.4.5, 9.5, 9.5.1, 9.5.4, 9.5.6, 9.5.7 and 9.5.8 of SSER10.

C900..01* = See last page for status code definition.

Page 58 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 5In response to TVA letters requesting relief from the requirement of 10 CFR 70.24 to have a criticality monitor installed in the fuel storage area until irradiated fuel is placed in the area, the staff granted an exemption from the requirement in SSER5.

C910..010Approved for both units in SER.

C911..21In SSER5, the staff acknowledged notification by TVA of a contract with DOE for DOE to accept spent fuel from WB and stated that they had no more concerns about this issue.In SSER15, the staff reviewed TVA's proposed resolution of the Boraflex degradation issue and found it acceptable.In SSER16, the staff reviewed changes in design basis with respect to placement of fuel assembly, and structural aspects of rack fabrication deficiencies, considering that TVA planned to replace the racks by the first scheduled refueling outage. The staff noted that the replacement racks have approximately the same capacity as the original WB racks. The staff concluded that the proposed changes were acceptable provided that no single rack load exceeded 80% of its original capacity.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O912..0221In SSER11, the staff reviewed TVA's revised commitment regarding testing of spent fuel pool cooling pumps and found it acceptable.As a result of a submittal filed as a petition pursuant to 10 CFR 2.206 regarding spent fuel storage safety issues, the staff reevaluated the spent fuel cooling capability at WB considering the identified issues and concluded that the spent fuel cooling system satisfied the requirements of GDC 44 with regard to transferring heat from the spent fuel to an ultimate heat sink under normal operating and accident conditions in SSER15.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O913..0213LICENSE CONDITION - Control of heavy loads (NUREG-0612)The staff noted in SSER3 that they were reviewing TVA's submittals regarding NUREG-0612 and 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: Implement NEI guidance on heavy loads.

OV914..01Addressed in 9.1.4.

NA915..* = See last page for status code definition.

Page 59 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 0Approved for both units in SER.

C920..18In SSER9, the staff noted that Amendment 65 indicated that ERCW provided cooling to the instrument room chillers, instead of room coolers and stated that conclusions in the SER and supplements were still valid. In SSER10, the staff reviewed discrepancies between FSAR figures pertaining to the raw cooling water system and its valving and TVA's clarification of these discrepancies, and considered them resolved.In SSER18, the staff concluded that ERCW does not conform to GDC 5 for two-unit operation.Unit 2 Action: Appropriate measures will be taken to ensure that the ERCW system is fully capable of meeting design requirements for two unit operation.

O921..015CONFIRMATORY ISSUE - relocate component cooling thermal barrier booster pumps above probable maximum flood (PMF) level before receipt of an OLTVA committed to relocate the pumps above PMF level and the staff found this acceptable. Implementation for this issue was resolved for Unit 1 in SSER5 when the staff verified in IR 390/84-20 that the pumps had been relocated. Additionally, IR 390/83-06 and 391/83-05 verified that the 4 booster pumps had been relocated and the construction deficiency reports identifying this issue for both units were closed. Unit 2 Action: Verify relocation of pumps for Unit 2.

CI922..010Approved for both units in SER.

C923..9In SSER9, the staff noted that potable water requirements were incorrectly stated in the SER, but this change did not affect the conclusions reached in the SER.C924..010Approved for both units in SER.

C925..12In SSER12, the staff noted that FSAR Amendment 72 revised the reserved amount of condensate for each units auxiliary feedwater system from 2000,000 gallons to 210,000 gallons and that this did not change the conclusions reached in the SER or supplements.

C926..010Approved for both units in SER.

C930..0Approved for both units in SER.

C931..* = See last page for status code definition.

Page 60 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21LICENSE CONDITION - Post-Accident Sampling SystemIn SSER3, the staff identified the criteria from Item II.B.3 in NUREG-0737 that were unresolved in the SER and reviewed TVA responses for these items. The staff stated that the post-accident sampling system met all of the criteria and was acceptable. They also stated that the proposed procedure for estimating the degree of reactor core damage was acceptable on an interim basis and that TVA would be required to provide a final procedure for estimating the degree of core damage before start-up following the first refueling outage. In SSER5, the staff stated that due to the 5 year delay in WB licensing, TVA should commit to submitting the procedure at an earlier date.TVA submitted a final procedure for estimating degree of core damage by letter dated June 10, 1994, and the license condition was deleted in SSER14.In SSER16, the staff reviewed TVA's revised emergency plan implementing procedure governing the use of the methodology provided in the June 10, 1994, submittal, and other plant data, for addressing degree of reactor core damage and found the methodology and implementing procedure acceptable.Unit 2 Action:

Eliminate requirement for Post-Accident Sampling System in Technical Specifications (Identified as CT in NRC letter dated May 28, 2008).--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (NRR).

Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.Rev. 0 of the Unit 1 TS contained 5.7.2.6, "Post Accident Sampling."

Amendment 34 to the Unit 1 TS (approved by the NRC on January 14, 2002) deleted 5.7.2.6, "Post Accident Sampling." The markup for Unit 2 Developmental Revision A noted that Unit 2 had deleted 5.7.2.6, "Post Accident Sampling" also.

S932..020Approved for both units in SER.

C933..0Approved for both units in SER.

C934..0Approved for both units in SER.

C940..9In SSER9, the staff clarified control room isolation after activation of SI signal from either unit, or upon detection of high radiation or smoke concentration in outside air supply stream and stated that conclusions reached in SER and supplements were still valid.

C941..010Approved for both units in SER.

C942..* = See last page for status code definition.

Page 61 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 0Approved for both units in SER.

C943..0Approved for both units in SER.

C944..21In SSER9, the staff reviewed the design of the additional DG building ventilation system (FSAR Amendment 66 submittal dated May 20, 1991, for both units) and determined that conclusion reached in SER was still valid and design was acceptable.In SSER10, the staff had concerns regarding periodic testing of the ventilation system for the additional DG building; muffler room exhaust fan failure or exhaust blockage; missile protection for the muffler fan exhaust structure; and potential for blockage and turbine missile damage of air intake structures. These were all resolved in SSER10, with the exception of the potential for external blockage of the air intake structure by missile impact. In SSER11 the staff found TVA's response and procedural change to address potential blockage of the air intake structure by missile impact acceptable. TVA stated in a submittal dated July 28, 1993, that they did not plan to place the additional diesel generator in service.In SSER14, the staff clarified statements made in the SER by stating that none of the ventilation systems for the ERCW pumping station was safety related, but the failure of both mechanical equipment room ventilation fans would not prevent operation of any safety related equipment. Thus, the conclusions reached in the SER were still valid, and the systems were still acceptable.In SSER16, the staff reviewed design changes to the DG building ventilation system, since the original design was reviewed, and concluded that the judgments made in the SER and supplements did not change and the system was still acceptable.In SSER19, the staff clarified their statements about the diesel engine room exhaust fans, stating that since the fans automatically start when the DG starts, DG testing results in operation of the diesel engine room exhaust fans.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (NRR).

O945..0210In SSER10, the staff reviewed 55 questions previously asked concerning the 4 original DGs for applicability to the additional DG and additional responses from TVA and had no concerns.

C950..01* = See last page for status code definition.

Page 62 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 199.5.1.2: OUTSTANDING ISSUE for Fire Protection Program9.5.1.3: CONFIRMATORY ISSUE - Electrical penetrations documentation9.5.1.3: LICENSE CONDITION - Fire protection program In SSER10, the staff noted that the fire hazard analysis for the additional DG building would be included in the WB Fire Protection report. The staff reviewed the building design for compliance with BTP 9.5-1, Appendix A and found it in conformance with the BTP. They also asked TVA to verify that the fire fighting systems installed in the DG building meet GDC 3 and stated that TVA's response satisfied their concerns.In SSER18, the staff concluded that the Fire Protection program for Watts Bar conformed to the requirements of 10 CFR 50.48 and was acceptable except for the fire barrier seal program and emergency lighting inside the Reactor Building. Additionally, the staff considered the confirmatory issue involving electrical penetration documentation resolved in SSER18 on the basis of the safety evaluation of the revised Fire Protection program included in Appendix FF of SSER18. In Appendix FF of SSER19, a safety evaluation of the Fire Protection program contains a detailed evaluation of fire barrier penetration seals. The staff concluded that TVA's penetration seal program adequately demonstrates the fire resistive rating of the penetrations, and that they conform to the guidelines of Positions D.1.j and D.3.d of Appendix A to BTP 9.5.1 and were acceptable. The safety evaluation also includes TVA's revised position on emergency lighting, which was found to be acceptable.

C951..0121LICENSE CONDITION - Performance testing of communications systemThe staff resolved this license condition in SSER5 based on TVA's letter of March 18, 1985 for both units, which described its testing of communications systems. Unit 2 Action: Perform testing of communication systems on Unit 2.

REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O952..020Approved for both units in SER.

C953..* = See last page for status code definition.

Page 63 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 129.5.4.1: CONFIRMATORY ISSUE - include required language in operating instruction to ensure no-load and low-load operation is minimized and revise operating procedures to address increased diesel generator load after it has run for an extended period of time at low or no loadIn SSER5, the staff verified that plant operating procedures had been revised to incorporate requirements that ensure that operational no-load and low-load conditions will not harm the diesel generators.---------------------

9.5.4.1: LICENSE CONDITION - Diesel Generator reliability The staff verified that the modifications necessary to comply withNUREG/CR-0660 had been completed and, as stated above, requirements had been incorporated into operating procedures. Thus, this license condition was resolved in SSER5.---------------------

9.5.4.1: OUTSTANDING ISSUE for staff to complete review to determine if diesel generator auxiliary support systems can perform their design safety functions under all conditions, after receipt of all requested information.In SSER5, the staff resolved the issue of the completeness of its review of the emergency diesel engine lubrication oil system.---------------------

9.5.4.1: OUTSTANDING ISSUE to design skid-mounted piping and components from the day tank to the diesel engine as seismic Category I and to ASME Section III, Class 3The staff reviewed standards to which emergency diesel engine skid mounted auxiliary system piping and associated components were designed, as well as the testing and inspections to be performed on these systems, as provided in TVA letters dated February 15, 1985, March 18, 1985, and August 30, 1985, and concluded that they were acceptable in SSER5. The staff considered this issue resolved. They stated that this resolution applied to the fuel oil, cooling water, air starting, lubrication, and combustion air intake and exhaust systems (9.5.4.2, 9.5.5, 9.5.

6, 9.5.7 and 9.5.8).---------------------

9.5.4.2: CONFIRMATORY ISSUE - provide missile protection for fuel oil storage tank vent linesThe staff found TVA's commitment to provide missile protection for the fuel oil storage tank vent lines acceptable and verified that the protection had been installed and considered this issue resolved in SSER5.---------------------In SSER9, the staff stated that the conclusions reached in the SER, SSER3 and SSER5 regarding the EDG auxiliary supports systems applied to the additional EDG. This conclusion applied to sections 9.5.5, 9.5.6, 9.5.7 and 9.5.8, as well.In SSER10, the staff questioned tornado missile protection and seismic requirements for the additional DG fuel oil storage tank fill lines and found TVA's response acceptable. The staff questioned the difference between the design of the fuel oil transfer pump for the additional DG and the design of the DG building storage pumps, and found TVA's explanation and proposed clarification to the FSAR acceptable. TVA stated in a submittal dated July 28, 1993, that they did not plan to place the additional diesel generator in service.

C954..01* = See last page for status code definition.

Page 64 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. In SSER11, the staff noted the revised capacity of the 7-day fuel oil storage tank identified in FSAR Amendment 69 and stated that it still exceeded the amount needed for a 7-day supply and, therefore, did not affect the staff's conclusions reached in the SER or supplements.In SSER12, the staff determined that the fire watch required when routing a hose from a fuel oil delivery vehicle to the DG tank manway openings in the DG building was no longer required based on TVA actions in response to other fire protection requirements.------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------The status in SSER21 is Open (NRR).11OUTSTANDING ISSUE to design engine cooling water system piping and components for all engines up to the engine interface, including auxiliary skid mounted piping, to ASME Section III, Class 3The staff reviewed standards to which emergency diesel engine skid mounted auxiliary system piping and associated components were designed, as well as the testing and inspections to be performed on these systems, and concluded that they were acceptable in SSER5. The staff considered this issue resolved. This resolution applies to the fuel oil, cooling water, air starting, lubrication, and combustion air intake and exhaust systems.--------------------In SSER5, the staff also resolved concerns regarding ambient DG room temperature and its impact on pre-heating DG units, the time period the DG is capable of operating fully loaded without secondary cooling, and the possibility of the cooling water system becoming air bound due to the expansion tank location.In SSER11, the staff noted that FSAR Amendment 70 stated that coolant temperature would be maintained between 125 and 155 degrees F, not the 115 and 125 stated in the SER. They stated that this clarification did not alter the staff's conclusions previously reached in the SER or its supplements.

C955..01* = See last page for status code definition.

Page 65 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21OUTSTANDING ISSUE to design engine air-starting system piping components for all engines up to the engine interface, including auxiliary skid mounted piping, to ASME Section III, Class 3The staff reviewed standards to which emergency diesel engine skid mounted auxiliary system piping and associated components were designed, as well as the testing and inspections to be performed on these systems, and concluded that they were acceptable in SSER5. The staff considered this issue resolved. This resolution applies to the fuel oil, cooling water, air starting, lubrication, and combustion air intake and exhaust systems.--------------------

In SSER10, the staff questioned protection of the additional DG electrical starting system components from water spray, and whether diesel engine control functions supplied by the air starting system could interfere with the engines' ability to perform its safety function once it has started. TVA stated in a submittal dated July 28, 1993, that they did not plan to place the additional diesel generator in service.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O956..02* = See last page for status code definition.

Page 66 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21OUTSTANDING ISSUE to perform additional modification, or provide justification for acceptability of proposed modification, to ensure lubrication of all wearing parts of the diesel engine either on an interim or continuous basis and to provide a more detailed description of the lubricating oil system and a description of the diesel engine crankcase explosion protection featuresIn response to a staff concern regarding dry diesel engine starting, TVA proposed using the manufacturers' modification and provided justification for its ability to ensure lubrication of all parts of the diesel engine. The staff found this acceptable in SSER3.TVA submittal of March 18, 1985, responded to a staff request to describe the features that protect the diesel engine crankcase from exploding. In SSER5, on the basis of this submittal, the staff concluded that the emergency diesel engine lubrication oil system can perform its safety function and is acceptable. This issue was resolved.--------------------OUTSTANDING ISSUE to design standby diesel engine lube oil system piping and components up to the engine interface, including skid mounted piping, to ASME Section III, Class 3The staff reviewed standards to which emergency diesel engine skid mounted auxiliary system piping and associated components were designed, as well as the testing and inspections to be performed on these systems, and concluded that they were acceptable in SSER5. The staff considered this issue resolved. This resolution applies to the fuel oil, cooling water, air starting, lubrication, and combustion air intake and exhaust systems.--------------------

In SSER10, the staff questioned the ability to replenish the additional DG lube oil system without interrupting operation of the DG and found TVA's provision to replenish lube oil acceptable. TVA stated in a submittal dated July 28, 1993, that they did not plan to place the additional diesel generator in service.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O957..02* = See last page for status code definition.

Page 67 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21OUTSTANDING ISSUE to design standby diesel engine combustion air intake and exhaust system piping and components up to the engine interface to ASME Section III, Class 3 and recommendations of RG 1.26The staff reviewed standards to which emergency diesel engine skid mounted auxiliary system piping and associated components were designed, as well as the testing and inspections to be performed on these systems, and concluded that they were acceptable in SSER5. The staff considered this issue resolved. This resolution applies to the fuel oil, cooling water, air starting, lubrication, and combustion air intake and exhaust systems.--------------------

In SSER10, the staff expressed a concern regarding products of combustion from a fire in the air intake/muffler room, or from the DG exhaust gases, impacting the additional DG or the other DGs. TVA's response addressed the concern. The staff also questioned inspection, surveillance and testing of the DG exhaust system and found the system design adequate to address their concern. In addition, the staff questioned pressure losses through the DG air intake and exhaust systems and determined that their designs were acceptable. TVA stated in a submittal dated July 28, 1993, that they did not plan to place the additional diesel generator in service.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O958..020Approved for both units in SER.

C1000..0Approved for both units in SER.

C1010..21In SSER5, the staff agreed that the interval between periodic turbine valve testing could be increased for WB from weekly to monthly.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O1020..0212In SSER12, the staff reviewed the revised description of the 3 independent overspeed turbine trip systems, consistent with FSAR Amendment 77, and stated that this review did not alter the conclusions reached in the SER and the system remained acceptable.

C1021..010Approved for both units in SER.

C1022..0Approved for both units in SER.

C1030..* = See last page for status code definition.

Page 68 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21In SSER12, the staff described changes to the MSIV closing signals as a result of changes to the Eagle-21 process protection system. They stated that the conclusions reached in the SER were still valid and the main steam system remained acceptable.In SSER19, the staff evaluated a revision in FSAR Amendment 91 to the closure time of the MSIVs from 5 seconds after receiving a closure signal to 6 seconds and concluded it was acceptable.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (NRR).

O1031..020Approved for both units in SER.

C1032..0Approved for both units in SER.

C1033..5LICENSE CONDITION - Secondary water chemistry monitoring and control programThe staff determined that the secondary water chemistry monitoring and control program was being included in the administrative section of the Technical Specifications and resolved this for Unit 1 in SSER5. Unit 2 Action: Take same action for Unit 2.

REVISION 02 UPDATE:

Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.Section 5.7.2.13 provides information about the Secondary Water Chemistry Program.S1034..020Approved for both units in SER.

C1040..9In SSER9, the staff clarified the description of the main condenser and stated that this clarification did not affect the conclusion reached in the SER.

C1041..010Approved for both units in SER.

C1042..0Approved for both units in SER.

C1043..* = See last page for status code definition.

Page 69 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21In SSER5, the staff concluded that periodic stroking of the turbine bypass system valves may be performed according to plant operating procedures and no Technical Specification was necessary to ensure this testing.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (NRR).

O1044..020Approved for both units in SER.

C1045..0Approved for both units in SER.

C1046..14In SSER14, the staff evaluated changes that TVA made in Amendment 82 to the FSAR adding a new feedwater isolation signal and clarifying the isolation signal generated by a reactor trip, and stated that the revisions did not affect the conclusions reached in the SER. The staff also corrected an unrelated error they made in the SER regarding the time for the main feedwater regulation valves to close after receipt of a feedwater isolation signal and stated that the conclusions reached in the SER remained valid.

C1047..010Approved for both units in SER.

C1048..14In SSER14, the staff discussed reductions in auxiliary feedwater pump design-basis flow rates and new minimum flow requirements. They reviewed TVA's reanalysis of design-basis events and concluded that the revised flow rates were acceptable and the conclusions reached in the SER remained valid.

C1049..010Approved for both units in SER.

C1100..16This item remains open pending closure of 11.4.0 and 11.5.0 OV1110..0116In SSER4, the staff evaluated the revised description contained in FSAR Revision 49 and 54 and determined that the conclusions reached in the original SER were not affected by the revisions.In SSER16, the staff superseded its previous review of the liquid waste management system. The staff concluded that TVA had submitted sufficient design information for both Units 1 and 2 liquid waste management system in accordance with 10 CFR 50.34a requirements and that the LWMS for Watts Bar Units 1 and 2 met the acceptance criteria of SRP Section 11.2 and was, therefore, acceptable.

C1120..01* = See last page for status code definition.

Page 70 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 16In the SER, the staff identified that the hydrogen and oxygen monitoring system did not meet the acceptance criteria because redundant monitors had not been provided and because the system was not designed to automatically initiate action to mitigate the potential for explosion in the event of high oxygen content. This issue was addressed by Technical Specifications discussed in the original SER and in SSER8 but was later resolved in SSER16. Based upon NRC review of TVA's February 17, 1995, letter (submitted on both dockets), the staff accepted the WBN's system approach of preclusive of gas buildup, as allowed by SRP Section 11.3 guidelines, if TVA submitted an administrative program to satisfy administrative controls for TS 5.7.2.15, "Explosive Gas and Storage Tank Radioactivity Monitoring Program." As stated in TVA's letter dated July 21, 1995, the program would provide for monitoring and control of potential explosive mixtures, limit the concentration of oxygen, and surveillance to ensure that the limits are not exceeded. As a result of an SSER16 review, the staff concluded that the GWMS for Watts Bar Units 1 and 2 met the acceptance criteria of SRP Section 11.3 and was acceptable.

C1130..0116On the basis of its review in SSER16, the staff found the process control program for Watts Bar acceptable and concluded that the solid waste management system for Watts Bar Unit 1 conformed to the acceptance criteria of SRP Section 11.4 and was, therefore, acceptable.Unit 2 Action:

Provide system description and information on QA provisions for Unit 2 Solid Waste Management System and information on the Process Control Program.

OV1140..0120In SSER16, the staff updated its review to Amendment 89, and TVA's submittal dated February 17, 1995. The staff concluded that the process and effluent radiological monitoring and sampling system for Watts Bar Unit 1 complied with 10 CFR 20.1302 and GDCs 60, 63, and 64. The staff also concluded that the system design conformed to the guidelines of NUREG-0737, RGs 1.21 and 4.15, and applicable guidelines of RG 1.97 (Rev. 2).

Thus, the system met the acceptance criteria of SRP Section 11.5 and was, therefore, acceptable.In SSER20, the staff agreed that TVA did not commit to RG-4.15, Revision 1 as reflected in TVA's July 21, 1995 letter. In that letter, TVA had stated that the radiation monitoring system generally agrees with and satisfies the intent of the RG 4.15 except for specific calibration techniques and frequencies. The staff then reiterated its earlier finding stated in SSER16, Section 11.5.1, that the radiation monitoring system for Watts Bar Unit 1 meets the intent and purpose of RG 4.15, with respect to quality assurance provisions for the system. The staff modified one sentence from SSER16 and then concluded by stating that the other conclusions given in SSER16 continued to be valid.Unit 2 Action:

Provide system description and information on QA provisions for the Unit 2 Radiation Monitoring System OV1150..01* = See last page for status code definition.

Page 71 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21In SSER8, the staff reviewed the preoperational REMP program provided by letter dated June 14, 1991 (submitted for both dockets) The staff concluded in SSER Section 1.6.1, "Offsite Radiological Monitoring Program," that the Watts Bar preoperational REMP as proposed was adequate to provide baseline data which will assist in verifying radioactivity concentrations and related public exposures during plant operation, and was therefore acceptable. The staff provided a safety evaluation for both units via a September 10, 1991 letter. In SSER16, the staff superseded previous evaluations provided in this section by Sections 11.1 through 11.5 of this supplement, except for the material in Section 11.6.1 of SSER8, which was unaffected by supplement 16.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O1160..020This item will remain open pending resolution of Item 11.7.2.

OT1170..01* = See last page for status code definition.

Page 72 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21LICENSE CONDITION (6a) - Accident monitoring instrumentation II.F.1 - Noble Gas monitorIn SSER5, TVA submitted letter dated April 26, 1985, on the Unit 1 docket which stated that the Unit 2 shield building vent monitor could not be installed by the time Unit 1 fuel load was scheduled in 1985 because of procurement problems. Since the 1985 fuel load was delayed, TVA subsequently committed in letter dated October 11, 1990, that this monitor and its sampler would be operational before fuel was loaded in Unit 1. This commitment eliminated the staff's concern and resolved the proposed License Condition 6a.Also, in SSER5, TVA letter dated November 8, 1983 (submitted on both Unit 1 and Unit 2 dockets) requested an exception to the requirement to monitor pressurized-water reactor steam safety valve discharge and atmospheric steam dump valve discharge to be monitored by high-range noble gas effluent monitors by stating that adequate instrumentation was provided to detect a steam generator tube rupture. The staff disagreed with this approach which resulted in TVA subsequently committing in a letter dated October 11, 1990 (submitted on both dockets) that the required high range noble gas effluent monitor would be operational before fuel load. This commitment resolved the staff's concern and eliminated the need for License Condition 6a.---------------------

LICENSE CONDITION (6b) - Accident monitoring instrumentationII.F.1 - Iodine particulate samplingSee 7.5.2.

In addition, in SSER5, by letter dated April 26, 1985, submitted on the Unit 1 docket, TVA committed to have the capability for continuous collection in place (i.e., procedures and any minor system modifications necessary) before exceeding 5-percent power. The staff evaluated this commitment and found it acceptable. Since 1985 licensing of Watts Bar was delayed, TVA subsequently committed via letter dated January 3, 1991, as discussed in SSER6 that the procedural revision and upgrade of the radiation monitors would be done by Unit 1 fuel load. Thus License Condition 6b was resolved in SSER6. In SSER6, TVA via letter dated January 3, 1991, committed to have the procedural revision and upgrade of the radiation monitors by fuel load. This commitment ensured the plant would have the capability for continuous collection of post accident gaseous effluents by fuel load.---------------------

In SSER5, the staff noted that the WBN design did not include a high-range noble gas effluent monitor as described in NUREG-0737, Item II.F.1, Attachment 1, for the auxiliary building vent because the release is diverted to the shield building vent for design-basis accidents. A low-range to high-range radiation monitor is provided in the shield building ventilation stack. By letter dated November 22, 1983, TVA requested an exception to NUREG-0737, Item II.F.1, concerning the installation ofhigh-range noble gas monitors on the auxiliary building vent at Watts Bar. TVA provided the staff additional information at a meeting on December 20, 1983, and subsequently in a submittal dated January 24, 1984. The staff concluded that the auxiliary building vent was not considered to be a potential accident release pathway and, therefore, the Watts Bar Nuclear Plant design, as described above, does not need to be changed to provide for the addition of a high-range noble gas effluent monitor, as described in NUREG-0737, Item II.F.1, Attachment 1, for the auxiliary building vent.The above items were identified as CI by NRC in May 28, 2008, letter.----------------------------------------------------------------------------------------------------

CI1171..02* = See last page for status code definition.

Page 73 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. ----------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (Inspection).16NUREG-0737, III.D.1.1, "Primary Coolant Outside Containment" - Resolved for Unit 1 only in SSER10; reviewed in Appendix EE of SSER16. Unit 2 Actions: Include the waste gas disposal system in the leakage reduction program and incorporate in Unit 2 Technical Specifications.---------------------

In SSER5, TVA by letter dated October 4, 1984, submitted a justification for excluding the waste gas system from the leak reduction program under NUREG-0737, Item III.D.1.1. The staff has evaluated the TVA's submittal and found that sufficient information had not been submitted to provide assurance that significant quantities of radioactive materials would not enter the waste gas system in the event of an accident.On this basis, the staff concluded that the leakage reduction program was acceptable if the following systems were to be included leakage reduction program: (1) residual heat removal, (2) containment spray, (3) safety injection, (4) chemical and volume control, (5) sampling, and (6) waste gas. The staff proposed License Condition 24 and would be resolved if TVA accepted the change as stated above. In SSER6, the staff reviewed TVA's letter dated March 27, 1986, and agreed that TVA had justified excluding the WGDS from the program. In SSER10, the staff resolved Condition 24, when upon review of TVA letter dated August 27, 1992, they noted that WGDS specification was included in the draft TS Section 5.7.2.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

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

S1172..0214Approved for both units in SER.

C1200..* = See last page for status code definition.

Page 74 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21In SSER10, the staff updated its evaluation based upon review of FSAR Amendments 65 through 71 and TVA letter dated January 3, 1991 submitted on U1 docket only. The staff acknowledged that TVA would soon revise FSAR again due to reflect recent changes to 10 CFR Part 20.In SSER14, the staff reviewed the revised FSAR to reflect the 10 CFR Part 20 changes. Details of the staff's review are delineated in the sections that follow.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (NRR).

O1210..0221In SSER14, the staff reviewed the revised FSAR discussion of ALARA design and operational considerations in this section that were made to clarify that the total effective dose equivalent for each individual would be maintained ALARA. As revised, FSAR Section 12.1 was consistent with the requirements in 10 CFR 20.1101 and 20.1702 and was, therefore, acceptable to the staff.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O1220..0221In SSER14, the staff reviewed the revised FSAR descriptions of the radioactive sources expected to result from normal plant operations, anticipated operational occurrences, and accident conditions. The staff concluded that the descriptions of plant radioactive sources, as revised, conformed to the acceptance criteria in SRP Section 12.2 and were, therefore, acceptable to the staff.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O1230..02* = See last page for status code definition.

Page 75 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21In SSER10, the staff reviewed revised operational test frequency of area radiation monitors from monthly to quarterly and found that TVA's program met the provisions of 10 CFR 20.1601(c) and the acceptance criteria in SRP Section 12.3 and was, therefore, acceptable.In SSER14, the staff reviewed FSAR Amendment 84 in light of the revised requirements of 10 CFR Part 20. The staff found these sections, as amended, complied with the acceptance criteria in the SRP and was acceptable to the staff. In addition, the staff reviewed revised FSAR Section which specified the radiation dose rate design criteria for the placement and configuration of plant system valves This section as amended was consistent with the staff's conclusion that Watts Bar can be operated within the dose limits and that radiation doses can be maintained ALARA. Therefore, these changes were acceptable to the staff.In SSER18, the staff reviewed FSAR Amendments 89 and 90 in which TVA had revised the discussions of the installed area radiation monitoring and the fixed airborne radiation monitoring systems. In addition, Amendment 90 revised the estimated maximum radiation dose rates depicted on the radiation zone maps for several areas in the plant. The staff also reviewed FSAR text changes that clarified the distinctions between a monitor calibration, a monitor channel operational test, and a check source functional test and deleted discussions of fixed airborne radiation monitors in the Unit 2 hot sample room and the Unit 1 control room and were replaced with portable continuous air monitors (CAMs). The staff found this acceptable since it did not change the staff's conclusion documented in SSER14.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O1240..0221In SSER14, the staff reviewed FSAR Amendment 88 which revised the discussion of the estimate of personnel internal exposures to address the new 10 CFR Part 20 requirements. The staff concluded that this section as amended provided reasonable assurance that the requirements of 10 CFR 20.1502 and 20.1703 would be met. In addition, the staff reviewed FSAR Amendment 84 which updated the predicted maximum annual doses resulting from plan operation and determined that this section as amended provides reasonable assurance that the radiation doses resulting from plant operations would not exceed the limits in 10 CFR 20.1301.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (NRR).

O1250..02* = See last page for status code definition.

Page 76 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21OUTSTANDING ISSUE involving Health Physics ProgramThe staff reviewed TVA's RADCON program (formerly the HP program) and found that the WBN organizational structure can provide adequate support for the RADCON program and that organizational changes described in the FSAR amendments met the staff's acceptance criteria. They considered this issue resolved in SSER10. In SSER14, the staff reviewed the revised FSAR sections (through Amendment 88), and found them acceptable.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (NRR).

O1260..020Approved for both units in SER.

C1270..21NUREG-0737, II.B.2, "Plant Shielding" - NRC reviewed in Appendix EE of SSER16. In SSER14, the staff reviewed FSAR Amendment 88 which revised the discussion of shielding for accident conditions. The staff stated that this change did not affect the staff's previous conclusion that Watts Bar conformed to the positions in NUREG-0737 Item II.B.2, and was therefore, acceptable to the staff. Identified as CI in NRC letter dated May 28, 2008.Unit 2 Action: Complete Design Review of EQ of equipment for spaces/systems which may be used in post accident operations. CI in NRC May 28, 2008, letter.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O1271..0221NUREG-0737, II.F.1.2.C., "Accident Monitoring Instrumentation" - In SSER5, the staff resolved this license condition for Unit 1 (IR 390/84-09 & IR 390/84-

28) due to verification that TVA's commitments regarding the high range in-containment monitor were satisfactory and that it was installed. Identified as CI in NRC letter dated May 28, 2008.Unit 2 Action: Install high range in-containment monitor for Unit 2. CI in NRC May 28, 2008, letter.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (NRR).

O1272..02* = See last page for status code definition.

Page 77 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21NUREG-0737, III.D.3.3, "In-plant Monitoring of I2 radiation monitoring" - NRC reviewed in Appendix EE of SSER16. Identified as CI in NRC letter dated May 28, 2008.Unit 2 Action: Complete modifications for Unit 2. CI in NRC May 28, 2008, letter.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O1273..020Approved for both units in SER.

C1300..16In SSER16, NRC reviewed the organizational information presented in TVA Topical Report TVA-NPOD89. NRC approval of the topical report and its revisions superseded the staff review in the SER.

C1310..010Approved for both units in SER.

C1311..0Approved for both units in SER.

C1312..8LICENSE CONDITION - Use of experienced personnel during startupIn the original 1982 SER, NRC provided a LICENSE CONDITION to ensure TVA augmented the shift staff with individuals that had prior experience with large pressurized water reactor operations. In SSER8, NRC reviewed TVA's commitment in the FSAR and the Nuclear Quality Assurance Plan to comply with RG 1.8, "Personnel Selection and Training,". NRC staff considered that this provided adequate assurance, and eliminated the LICENSE CONDITION.Unit 2 Action: Submit staffing and NQAP for two unit operation.

O1313..010Approved for both units in SER.

C1320..10In SSER9, NRC reviewed TVA's certification for licensed operator training programs and FSAR Chapter 13 revision to reflect the training program . NRC determined that these were acceptable. In SSER10, NRC reviewed changes to the initial test program for TMI Item I.G.1, "Training During Low Power Testing." NRC found the training requirement satisfied.

C1321..010Approved for both units in SER.

C1322..13In SSER13, NRC reviewed the Watts Bar Nuclear Plant Radiological Emergency Plan submitted February 12, 1993. This review superseded the review in the SER.Unit 2 Action:

Submit WBN REP for two unit operation.

O1330..01* = See last page for status code definition.

Page 78 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 20In SSER13, NRC reviewed the Watts Bar Nuclear Plant Radiological Emergency Plan submitted February 12, 1993. This review superseded the review in the SER. In SSER20, NRC completed the review including the findings of the Federal Emergency Management Agency.Unit 2 Action: Submit WBN REP for two unit operation.

O1331..0120In SSER13, NRC reviewed the Watts Bar Nuclear Plant Radiological Emergency Plan submitted February 12, 1993. This review superseded the review in the 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. In SSER20, NRC completed the review and found that the REP complied with NRC requirements and was acceptable for the full-power license of WBN Unit 1.Unit 2 Action:

Submit WBN REP for two unit operation.

O1332..0120LICENSE CONDITION - Emergency Preparedness (NUREG-0737, III.A.1, III.A.2, III.A.2)The NRC review of Emergency Preparedness in SSER13 superseded the review in the original 1982 SER. In SSER13, the staff concluded that the WBN Radiological Emergency Plan (REP) provided an adequate planning basis for an acceptable state of onsite emergency preparedness, and the LICENSE CONDITION was deleted. In SSER20, NRC completed the review and found that the REP complied with NRC requirements and was acceptable for the full-power license of WBN Unit 1.Unit 2 Action:

Submit WBN REP for two unit operation.

O1333..018LICENSE CONDITION - Independent Safety Engineering Group (ISEG) (NUREG-0737, I.B.1.2)In SSER8, NRC indicated that the ISEG would be established as part of the Technical Specifications. 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.

OV1340..010Approved for both units in SER.

C1350..21Approved for both units in SER.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O1351..02* = See last page for status code definition.

Page 79 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21OUTSTANDING ISSUE involving operating, maintenance and emergency proceduresIn the original 1982 SER, this issue was used to track the staff's review of the emergency operating procedures generation package. In SSER9, the staff concluded that the outstanding issue was no longer needed as the staff no longer performed such reviews. The emergency operating procedure development program review is performed under IP 42000, "Emergency Operating Procedures." This inspection will be performed before issuance of an operating license. In SSER10, NRC reviewed TVA's plan for vendor review of the power ascension test procedures and the Emergency Operating Instructions (EOIs). Based on the Watts Bar plant specific simulator, NRC determined that a License Condition to ensure consistency with the Sequoyah EOIs was no longer necessary.Unit 2 Action:

Issue operating, maintenance and emergency procedures.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:The status in SSER21 is Open (Inspection).

CI1352..02* = See last page for status code definition.

Page 80 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21LICENSE CONDITION - Report on outage of emergency core cooling system (NUREG-0737, 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.-------------------Reporting of Safety Valve and Relief Valve Failures and Challenges (II.K.3.3)

In SSER16, NRC reviewed TVA revised commitment to report failures and challenges to PORVs and safety valves in accordance with the Technical Specifications. Unit 2 Action: Include, as necessary, in the Technical Specifications.CT in NRC May 28, 2008, letter.

REVISION 02 UPDATE:

The status in SSER21 is Open (Inspection).

Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.Rev. 0 of the Unit 1 TS contained 5.9.4 (Monthly Operating Reports) which implemented the above commitment for Unit 1.Amendment 57 to the Unit 1 TS (approved by the NRC on March 21, 2005) deleted this section of the TS. The markup for Unit 2 Developmental Revision A noted that Unit 2 will apply this change, and the Unit 2 TS will contain no requirement for Monthly Operating Reports.

S1353..02* = See last page for status code definition.

Page 81 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21OUTSTANDING ISSUE to file appropriate revision to the Physical Security PlanIn the original 1982 SER, the staff identified certain outstanding issues with TVA's Physical Security Plan. In SSER1 NRC evaluated revisions to the plan submitted July 29, 1982. In SSER15, NRC provided a safety evaluation that concluded that WBN conforms to the requirements of 10 CFR 50.73.---------------------

LICENSE CONDITION - Physical security of fuel in containment In SSER1, part of the Physical Security Plan (PSP) was not in accordance with the regulation. TVA submitted a new PSP on June 17, 1992. In SSER10, the staff concluded that the provisions for protection of the containment during major refueling and maintenance met the intent of the regulation.------------------

LICENSE CONDITION - Land Vehicle Bomb Control Program In SSER20, NRC added a license condition for WBN Unit 1 to fully implement the Surface Vehicle Bomb Rule by February 17, 1996. TVA letter to NRC dated February 15, 1996, (submitted for both units) notified NRC that Watts Bar had fully implemented the program.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

O1360..02* = See last page for status code definition.

Page 82 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21LICENSE CONDITION - Report changes to Initial Test ProgramIn the original 1982 SER, this LICENSE CONDITION was intended to require TVA report to NRC within 30 days of modifying an approved initial test. In SSER7, the NRC accepted a commitment in TVA's July 1, 1991, letter to notify NRC within 30 days of any changes to the Startup Test Program made under 10 CFR 50.59. Unit 2 Action:

Notify NRC within 30 days of any changes to the Startup Test Program made under 10 CFR 50.59.---------------------

In SSER3, the staff reviewed additional information and FSAR amendments through 46 addressing concerns identified by the staff in the FSAR. They concluded in SSER3 that the Initial Test Program (ITP), with the exception of open items as a result of modifications made to the program in subsequent amendments (through 53) for which the staff requested additional information, would meet the acceptance criteria of SRP section 14.2 and successful completion of the program would demonstrate functional adequacy of structures, systems and components.In SSER5, the staff reviewed TVA submittals to address the open items from SSER3 and FSAR amendments through 55, and concluded that the program met the acceptance criteria of the SRP and was acceptable.In SSER9, the staff stated that TVA commitments to reinstate the loss-of-offsite-power test for Unit 2 and revise the acceptance criteria for the reactor building purge system air flow rate (TVA letter dated July 10, 1991, for both units) were found acceptable to address two issues identified by the staff during their review of the FSAR through Amendment 67.In SSER10, the staff agreed with TVA that there was no need to perform any natural recirculation test for Units 1 and 2 (See subsection 5.4.3.)In SSER12, the staff evaluated the ITP based on Amendment 74 to the FSAR, which addressed most of the staff's concerns raised during review of Amendment 69, in which the ITP was completely revised. The staff found that Chapter 14, as revised by Amendment 74, was generally adequate and in accordance with review criteria with the exception of 7 items, which would be evaluated in later supplements.In SSER14, the staff evaluated changes made by TVA in Amendments 84 and 86, as well as 5 TVA letters submitted during 1994 to resolve the issues identified by the staff in SSER12, and changes made in FSAR Amendment 88 to address concerns still open prior to that amendment. The staff found that, with the exception of open items that remained open pending receipt and review of TVA's responses, the WB Units 1 and 2 ITP description contained in FSAR Chapter 14, updated through Amendment 88, was generally comprehensive and encompassed the major phases of the program requirements.In SSER16, SSER18 and SSER19, the staff evaluated the ITP through amendments 89, 90 and 91 respectively and stated each time that it found the program to be comprehensive and encompassing the major phases of the testing program guidance presented in the SRP. ---------------

A Unit 2 issue to verify capability of each common station service transformer to carry load required to supply ESF loads of 1 unit under LOCA condition in addition to power required for shutdown of non-accident unit was raised in SSER14, and the NRC stated that before an OL can be issued for Unit 2, TVA would have to demonstrate the capability of each CSST to carry the loads of S1400..05* = See last page for status code definition.

Page 83 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. one unit under LOCA conditions in addition to power required for shutting down the non-accident unit. TVA agreed with the NRC position in a January 5, 1995, letter and the issue was resolved in SSER16. Unit 2 Action:

Amend FSAR Chapter 14 to reflect the capability of each CSST to carry the loads of one unit under LOCA conditions in addition to power required for shutting down the non-accident unit.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (Inspection).

Amendment 97 to the Unit 2 FSAR was submitted on January 11, 2010 (ADAMS Accession No. ML100191421) .Table 14.2-1 was revised to clarify the testing requirement.-------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 05 UPDATE:

As a result of the response to NRC RAI 14 - 1, item 6. of Table 14.2-1 was revised again as part of Amendment 100 to the Unit 2 FSAR. Amendment 100 was submitted on September 1, 2010 (ADAMS Accession No. ML102500171).0Approved for both units in SER.

C1500..Area not addressed in 1981 Standard Review Plan.

NA1501..Area not addressed in 1981 Standard Review Plan.

NA1502..0Approved for both units in SER.

C1510..Addressed in 15.2.1 NA1511..Addressed in 15.2.1 NA1512..Addressed in 15.2.1 NA1513..Addressed in 15.2.1 NA1514..Addressed in 15.2.1 and 15.4.2.

NA1515..* = See last page for status code definition.

Page 84 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 0Approved for both units in SER.Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.

Chapter 15 was updated to address the application of RFA-2 fuel.

S1520..0214In SSER13, NRC reviewed TVA's use of the FACTRAN computer code for LOCA temperature distribution. NRC concluded that the transient analysis was acceptable. In SSER14, NRC approved the trip time delay functional upgrade as part of the Eagle 21 process protection system for low-low steam generator reactor trip. TVA letter dated December 5, 2007, informs NRC of intent to use Eagle-21 for Unit 2. NRC requested additional information December 27, 2007. TVA provided the requested information by letter dated February 28, 2008. By letter dated May 7, 2008, NRC provided a list of specific issues to be addressed in a future amendment application for Eagle-21 for WBN Unit 2. Unit 2 Action: Provide the additional information for NRC review.

REVISION 02 UPDATE:

TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.

S1521..020Approved for both units in SER.

C1522..18In SSER18, NRC reviewed FSAR amendment 90. In FSAR amendment 90, TVA revised for the transient event of inadvertent ECCS actuation for both Units. TVA provided additional information for both units by letter dated October 12, 1995. In SSER18, NRC found the reanalysis acceptable.Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.

Chapter 15 was updated to address the application of RFA-2 fuel.

S1523..02* = See last page for status code definition.

Page 85 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 1415.2.4.1 Uncontrolled Rod Cluster Assembly Bank Withdrawal from Zero-Power ConditionIn SSER7, NRC reviewed additional analysis submitted for both units for a two pump, zero power, rod withdrawal. The NRC concluded the revision was acceptable. In SSER13, NRC accepted a change to a limiting condition for operation and bases changes to include a requirement that two reactor coolant pumps should be running whenever rods are capable of withdrawal in Mode 4.Unit 2 Action:

Submit Technical Specifications. --------------------15.2.4.4: OUTSTANDING ISSUE for evaluation of Boron dilution and single failure criteriaIn a letter dated November 2, 1984, TVA stated that the boron dilution alarm system receives signals from two independent channels which are independently powered. Additionally, testing of these circuits was described. The staff concluded in SSER4 that the system is adequately protected from single failure and closed this item. In SSER14, NRC reviewed a reanalysis of the accident associated with uncontrolled boron dilution and accepted the analysis.--------------------

15.2.4.6 Rod Cluster Control Assembly EjectionIn SSER14, NRC accepted a change to the maximum cladding temperature for the rod ejection accident made in FSAR amendment 80.--------------------Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.TS Limiting Condition for Operation 3.4.6 requires two RCS loops with both loops in operation when the rod control system is capable of rod withdrawal.

Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.

Chapter 15 was updated to address the application of RFA-2 fuel.

S1524..024Approved for both units in SER subject to completion of Outstanding Issue in 15.2.4.4.C1525..Addressed in 15.2.1.

NA1526..* = See last page for status code definition.

Page 86 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. Addressed in 15.2.1.

NA1527..0Approved for both units in SER.

C1530..15In SSER12, NRC reviewed the reanalysis of small break loss of coolant analysis (SBLOCA) for Units 1 and 2. NRC found the analysis acceptable. In SSER15, NRC reviewed additional changes to the SBLOCA for Units 1 and 2.Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.

Chapter 15 was updated to address the application of RFA-2 fuel.

S1531..0214In SSER3, NRC reviewed proposed changes to the boron concentration requirement in the Boron Injection Tank and found them acceptable. In SSER14, NRC reviewed TVA application of the new steamline protection feature associated with the Eagle 21 upgrade for WBN Unit 1. The model resulted in the reanalysis of two ruptures: the main feedline and a steamline break outside of containment.Unit 2 Action:

Perform analysis.

Unit 2 Action:

Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

WCAP-13462, "Summary Report Process Protection System Eagle 21 Upgrade, NSLB, MSS and TTD Implementation Watts Bar Units 1 and 2" Revision 2 is applicable to WBN Unit 2. The main feedline and steam line break outside of containment are analyzed in WCAP-13462. NRC has previously reviewed and accepted this analysis for Unit 1 in SSER14.

Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.

Chapter 15 was updated to address the application of RFA-2 fuel.

S1532..02* = See last page for status code definition.

Page 87 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 14In SSER14, NRC reviewed TVA application of the new steamline protection feature associated with the Eagle 21 upgrade for WBN Unit 1. The model resulted in the reanalysis of two ruptures: the main feedline and a steamline break outside of containment.Unit 2 Action: Perform analysis.--------------------

Unit 2 action:

Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:WCAP-13462, "Summary Report Process Protection System Eagle 21 Upgrade, NSLB, MSS and TTD Implementation Watts Bar Units 1 and 2" Revision 2 is applicable to WBN Unit 2. The main feedline and steam line break outside of containment are analyzed in WCAP-13462. NRC has previously reviewed and accepted this analysis for Unit 1 in SSER14.

Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.Chapter 15 was updated to address the application of RFA-2 fuel.

S1533..0214In SSER14, NRC reviewed this section based on VANTAGE 5H fuel and found it acceptable.Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.Chapter 15 was updated to address the application of RFA-2 fuel.

S1534..0214In SSER14, NRC reviewed this section based on VANTAGE 5H fuel and found it acceptable.Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.Chapter 15 was updated to address the application of RFA-2 fuel.

S1535..02* = See last page for status code definition.

Page 88 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21LICENSE CONDITION - Anticipated Transients Without Scram (Generic Letter 83-28 Item 4.3)In SSER3, NRC performed an initial review of Generic Letter 83-28 for the Salem anticipated transients without scram events. A new license condition was established for GL 83-28 Item 4.3. In SSER5, the staff found TVA's response to a number of items in GL 83-28 acceptable, including Item 4.3, and thus eliminated this license condition. In a letter dated June 18, 1990, for both units, NRC confirmed that all issues under Item 4.3 were fully resolved. In SSER6, NRC continued the review. In SSER10, NRC completed the review of TVA's submittals for GL 83-28 and found them acceptable. In SSER11, a reference to Item 4.3 that was omitted in SSER10 was added. In SSER12, NRC provided additional information on Items 3.1.3 and 3.2.3. NRC noted that TVA reported that there would be no post maintenance test requirements in the Technical Specifications for either the reactor trip system or other safety related components which could degrade safety. The NRC had no further concerns.CI in May 28, 2008, NRC letter.

REVISION 02 UPDATE:

The status in SSER21 is Open (Inspection).

CI1536..020Approved for both units in SER.

C1537..0Approved for both units in SER.

C1540..18In SSER5, NRC reviewed a change to the estimated fractions in leakage pathways for the release of radioactive material following a LOCA. In SSER9, NRC corrected the filter efficiency for organic iodine. The conclusions reached in the SER and supplements remained unchanged. In SSER15, NRC reviewed revised short term atmospheric relative concentration factors. The conclusions reached in the SER and supplements remained unchanged.

In FSAR amendment 90, TVA increased the amount of leakage that enters the auxiliary building following a LOCA. In SSER18, NRC confirmed this was within the guidelines of 10 CFR Part 100.Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.Chapter 15 was updated to address the application of RFA-2 fuel.

S1541..02* = See last page for status code definition.

Page 89 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 15In SSER15, NRC reviewed revised short term atmospheric relative concentration factors. The conclusions reached in the SER and supplements remained unchanged.Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.Chapter 15 was updated to address the application of RFA-2 fuel.

S1542..0215LICENSE CONDITION - Steam Generator tube ruptureIn SSER2, NRC performed an initial evaluation of an actual Steam Generator Tube Rupture (SGTR) that occurred at Ginna. As part of the Westinghouse Owners Group (WOG), WBN committed to implement all corrective actions recommended by the WOG. In SSER5, NRC reviewed the WOG SGTR analysis and determined that plant specific information was required. In SSER12, the staff identified 5 items that required resolution involving1) operator action times; 2) radiation offsite consequence analysis;

3) systems, 4) associated components credited for accident mitigation in SG tube rupture emergency operating procedures; and 5) system compatibility with bounding analysis. Items 2-5 were resolved in SSER12. In SSER14, the staff stated that a revised SG tube rupture analysis was more conservative and did not alter the conclusions of their Original safety evaluation. With regard to operator response times, TVA letters dated April 21, 1994, and August 15, 1994, and NRC letter dated June 28, 1994, dealt with simulator runs to address response times and operator performance during simulated SG tube ruptures. The staff concluded, after review of the TVA letters, that the times assumed in the tube rupture analysis were satisfactorily verified and deleted this condition. In SSER15, NRC reviewed revised short term atmospheric relative concentration factors. The conclusions reached in the SER and supplements remained unchanged.Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.Chapter 15 was updated to address the application of RFA-2 fuel.

S1543..02* = See last page for status code definition.

Page 90 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 15In SSER15, NRC reviewed revised short term atmospheric relative concentration factors. The conclusions reached in the SER and supplements remained unchanged.Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.Chapter 15 was updated to address the application of RFA-2 fuel.

S1544..0215In SSER4, NRC reevaluated the consequences of a fuel handling accident inside primary containment. NRC concluded WBN met the relevant requirements of GDC 61. In SSER15, NRC reviewed revised short term atmospheric relative concentration factors. The conclusions reached in the SER and supplements remained unchanged.Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.Chapter 15 was updated to address the application of RFA-2 fuel.

S1545..020Approved for both units in SER.Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.Chapter 15 was updated to address the application of RFA-2 fuel.

S1546..020Approved for both units in SER.Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.

Chapter 15 was updated to address the application of RFA-2 fuel.

S1547..02* = See last page for status code definition.

Page 91 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 0Approved for both units in SER.

C1550..4LICENSE CONDITION - Effect of high pressure injection for small beak LOCA with no auxiliary feedwater (NUREG-0737, II.K.2.13)In SSER4, the staff concluded that there was reasonable assurance that vessel integrity would be maintained for small breaks with an extended loss of all feedwater and that the USI A-49, "Pressurized Thermal Shock," review did not have to be completed to support the full-power license. NRC considered this condition resolved. C in NRC May 28, 2008 letter.

C1551..4LICENSE CONDITION - Voiding in the reactor coolant system (NUREG-0737, II.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.

C1552..5LICENSE CONDITION - PORV isolation system (NUREG-0737, II.K.3.1, II.K.3.2)NUREG-0737, II.K.3.1, II.K.3.2, "Auto PORV isolation/Report on PORV Failures" - Reviewed in SSER5 and resolved based on NRC conclusion that there is no need for an automatic PORV isolation system (NRC letter dated June 29, 1990). C in NRC May 28, 2008 letter.

C1553..21"Implementation of TMI Item II.K.3.5 (Automatic Trip of Reactor Coolant Pumps" - Reviewed in 15.5.4 of original 1982 SER; became License Condition 35. The staff determined that their review of Item II.K.3.5 did not have to be completed to support the full power license and considered this license condition resolved in SSER4. The item was further reviewed in Appendix EE of SSER16. CI in NRC May 28, 2008, letter.Unit 2 Action: Implement modifications as required.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Status in SSER21 is Open (Inspection).

CI1554..02* = See last page for status code definition.

Page 92 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21NUREG-0737, II.K.3.30, "Small Break LOCA Methods" and NUREG-0737, II.K.3.31, "Plant Specific Analysis" - The staff determined that their review of Items II.K.3.30 and II.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. Both of these items were CI in NRC May 28, 2008, letter.Unit 2 Action: Complete analysis for Unit 2.

REVISION 02 UPDATE:

Status in SSER21 is Open (Inspection).

Unit 2 FSAR Amendment 97 was submitted on January 11, 2010.

It documents SBLOCA analysis being performed using the NOTRUMP computer code. Use of the NOTRUMP evaluation model meets the requirements of II.K.3.31.

S1555..020Approved for both units in SER.

C1560..0Approved for both units in SER.

C1561..Unit 2 Action: Submit Technical Specifications.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

Developmental Revision A of the Unit 2 Technical Specifications was submitted on March 4, 2009.Developmental Revision B of the Unit 2 Technical Specifications was submitted on February 2, 2010.

S1600..02Area not addressed in 1981 Standard Review Plan.

NA1610..0Approved for both units in SER.

C1700..0Approved for both units in SER. See 17.3.

C1710..010Approved for both units in SER. See 17.3.

C1720..01* = See last page for status code definition.

Page 93 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 15OUTSTANDING ISSUE - QA programThe staff reviewed the description of the QA program in SSER2 and stated that they had resolved the list of open items for which the QA program for the operations phase applies with TVA and concluded that the description was in compliance with NRC regulations. The staff reviewed the organization for the QA program and the NQA Plan, and presented their conclusions in SSER5.

They concluded that the program was acceptable for the operations phase of Watts Bar. It was noted, however, that Amendment 63 stated that identification of safety related features would be addressed later and the staff left the outstanding issue unresolved. In SSER10, the staff reviewed additional revisions to the QA program and stated that they did not change the staff's conclusions reached in SSER5. In SSER13, the staff concluded that TVA had established appropriate programmatic controls for identification of safety related features and considered this issue resolved. In SSER15, the staff listed additional revisions to the QA program without comment.

C1730..010Approved for both units in SER. See 17.3.

C1740..01Area not addressed in 1981 Standard Review Plan.

NA1750..10 CFR 50.65- Maintenance RuleUnit 2 Action: Implement Maintenance Rule for Unit 2 systems 1 month prior to fuel load

REVISION 05 UPDATE:

TVA letter to NRC dated November 17, 2010 (ADAMS Accession No. ML103210644) revised this commitment to read "Implement Maintenance Rule for Unit 2 systems by October 21, 2011." OV1760..050See 18.1.

NA1800..* = See last page for status code definition.

Page 94 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21NUREG-0737, I.D.1, "Control Room Design Review" - NRC reviewed in SSER5, SSER6, SSER15, and Appendix EE of SSER16. In SSER6, the staff concluded that the DCRCR program implemented for Unit 1 satisfied the programmatic requirements of Supplement 1, NUREG-0737. In SSER15, the staff conducted a final onsite audit of the Unit 1 DCRDR and concluded that the product implemented conformed to the DCRDR requirements of Supplement 1, NUREG-0737 and that the DCRDR special program had been effectively implemented. In SSER16, the staff reviewed a TVA reclassification of a human engineering deficiency and concluded that it was satisfactory.Unit 2 Actions: Complete the CRDR process. Perform rewiring in accordance with ECN 5982. Take advantage of the completed Human Engineering reviews to ensure appropriate configuration for Unit 2 control panels. See CRDR Special Program.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

TVA's September 26, 2008, letter proposed the use of the Unit 1 approach to resolve the CRDR SP.

In SSER21, the Detailed Control Room Design Review (CRDR) Special Program was resolved. Completion of CRDR is tracked under 23.3.3.

CI1810..0221"CONCLUSIONS" left open until all items in subsection are closed.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

TVA's September 26, 2008, letter proposed the use of the Unit 1 approach to resolve the CRDR SP.

In SSER21, the Detailed Control Room Design Review (CRDR) Special Program was resolved.

CI1820..02* = See last page for status code definition.

Page 95 of 96 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 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 applicalbe is provided in the ADDITIONAL INFORMATION column.

O:OPEN: No action or documentation is provided that shows the staff has reviewed the item for WBN Unit 2.OT: OV: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.OPEN/VALIDATION: The proposed approach has been approved for Watts Bar Unit 1; the same approach is proposed for use on WBN Unit 2 without change.

S:SUBMITTED: Information has been submitted, and is under review by NRC staff.* = See last page for status code definition.

Page 96 of 96 Enclosure 2

SER and Supplements Review Matrix - Revision 5 Changes SAFETY EVALUATION REPORT AND SUPPLEMENTS(NUREG-0847) REVIEW MATRIX: REVISION 5 CHANGESSER SECTIONSSER #ADDITIONAL INFORMATION

REV. 16LICENSE CONDITION - Inservice inspection (ISI) programThe ISI program is required to be submitted within 6 months of the date of issuance of the operating license. The applicable ASME Code edition and addenda are determined by reference to 50.55a(b) 12 months preceding the date of issuance of the OL. The staff reiterated this in SSER10. In SSER12, the LICENSE CONDITION was resolved by a TVA commitment to submit the program within six months after receiving the operating license. Unit 2 Action: Submit Unit 2 ISI program.--------------------

OUTSTANDING ISSUE - Unit 2 PSI program submitted April 30, 1990, with a partial listing of relief requests. This item tracked the staff review.In the SER, the preservice inspection program was still under review. NRC reviewed the Unit 1 PSI program in SSERs 10, 12, and 16. Unit 2 Action: Submit Unit 2 PSI program.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 03 UPDATE:Preservice Inspection Plan, Program No. WBN-2 PSI, Revision 3 was submitted to the NRC on June 17, 2010 (ADAMS Accession No. ML101680561).--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 05 UPDATE:

Corrected status from "O" to "S." S524..05* = See last page for status code definition.Page 1 of 5 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 15OUTSTANDING ISSUE on additional information required on preservice inspection program and identification of plant specific areas where ASME Code Section XI requirements cannot be met and supporting technical justificationNRC reviewed the preservice inspection program (PSI) for Unit 1 only in SSER10 and on the basis of a TVA commitment to submit an inservice inspection program within 6 months after receiving an operating license, considered a proposed LC for an ISI no longer required. In SSER15, the staff reviewed Revisions 24 and 25 to the preservice inspection program and concluded that the changes included therein were acceptable. Unit 2 Action: Submit Unit 2 PSI program.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 03 UPDATE:Preservice Inspection Plan, Program No. WBN-2 PSI, Revision 3 was submitted to the NRC on June 17, 2010 (ADAMS Accession No. ML101680561).--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 05 UPDATE:

Corrected status from "O" to "S." S660..05* = See last page for status code definition.Page 2 of 5 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 21LICENSE CONDITION - Report changes to Initial Test ProgramIn the original 1982 SER, this LICENSE CONDITION was intended to require TVA report to NRC within 30 days of modifying an approved initial test. In SSER7, the NRC accepted a commitment in TVA's July 1, 1991, letter to notify NRC within 30 days of any changes to the Startup Test Program made under 10 CFR 50.59. Unit 2 Action:

Notify NRC within 30 days of any changes to the Startup Test Program made under 10 CFR 50.59.---------------------

In SSER3, the staff reviewed additional information and FSAR amendments through 46 addressing concerns identified by the staff in the FSAR. They concluded in SSER3 that the Initial Test Program (ITP), with the exception of open items as a result of modifications made to the program in subsequent amendments (through 53) for which the staff requested additional information, would meet the acceptance criteria of SRP section 14.2 and successful completion of the program would demonstrate functional adequacy of structures, systems and components.In SSER5, the staff reviewed TVA submittals to address the open items from SSER3 and FSAR amendments through 55, and concluded that the program met the acceptance criteria of the SRP and was acceptable.In SSER9, the staff stated that TVA commitments to reinstate the loss-of-offsite-power test for Unit 2 and revise the acceptance criteria for the reactor building purge system air flow rate (TVA letter dated July 10, 1991, for both units) were found acceptable to address two issues identified by the staff during their review of the FSAR through Amendment 67.In SSER10, the staff agreed with TVA that there was no need to perform any natural recirculation test for Units 1 and 2 (See subsection 5.4.3.)In SSER12, the staff evaluated the ITP based on Amendment 74 to the FSAR, which addressed most of the staff's concerns raised during review of Amendment 69, in which the ITP was completely revised. The staff found that Chapter 14, as revised by Amendment 74, was generally adequate and in accordance with review criteria with the exception of 7 items, which would be evaluated in later supplements.In SSER14, the staff evaluated changes made by TVA in Amendments 84 and 86, as well as 5 TVA letters submitted during 1994 to resolve the issues identified by the staff in SSER12, and changes made in FSAR Amendment 88 to address concerns still open prior to that amendment. The staff found that, with the exception of open items that remained open pending receipt and review of TVA's responses, the WB Units 1 and 2 ITP description contained in FSAR Chapter 14, updated through Amendment 88, was generally comprehensive and encompassed the major phases of the program requirements.In SSER16, SSER18 and SSER19, the staff evaluated the ITP through amendments 89, 90 and 91 respectively and stated each time that it found the program to be comprehensive and encompassing the major phases of the testing program guidance presented in the SRP. ---------------

A Unit 2 issue to verify capability of each common station service transformer to carry load required to supply ESF loads of 1 unit under LOCA condition in addition to power required for shutdown of non-accident unit was raised in SSER14, and the NRC stated that before an OL can be issued for Unit 2, TVA would have to demonstrate the capability of each CSST to carry the loads of S1400..05* = See last page for status code definition.Page 3 of 5 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. one unit under LOCA conditions in addition to power required for shutting down the non-accident unit. TVA agreed with the NRC position in a January 5, 1995, letter and the issue was resolved in SSER16. Unit 2 Action:

Amend FSAR Chapter 14 to reflect the capability of each CSST to carry the loads of one unit under LOCA conditions in addition to power required for shutting down the non-accident unit.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 02 UPDATE:

The status in SSER21 is Open (Inspection).

Amendment 97 to the Unit 2 FSAR was submitted on January 11, 2010 (ADAMS Accession No. ML100191421) .Table 14.2-1 was revised to clarify the testing requirement.-------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 05 UPDATE:

As a result of the response to NRC RAI 14 - 1, item 6. of Table 14.2-1 was revised again as part of Amendment 100 to the Unit 2 FSAR. Amendment 100 was submitted on September 1, 2010 (ADAMS Accession No. ML102500171).10 CFR 50.65- Maintenance RuleUnit 2 Action: Implement Maintenance Rule for Unit 2 systems 1 month prior to fuel load--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------REVISION 05 UPDATE:TVA letter to NRC dated November 17, 2010 (ADAMS Accession No. ML103210644) revised this commitment to read "Implement Maintenance Rule for Unit 2 systems by October 21, 2011." OV1760..05* = See last page for status code definition.Page 4 of 5 SER SECTIONSSER #ADDITIONAL INFORMATION

REV. 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.

O:OPEN: No action or documentation is provided that shows the staff has reviewed the item for WBN Unit 2.

OT: OV: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.OPEN/VALIDATION: The proposed approach has been approved for Watts Bar Unit 1; the same approach is proposed for use on WBN Unit 2 without change.

S:SUBMITTED: Information has been submitted, and is under review by NRC staff.* = See last page for status code definition.Page 5 of 5 Enclosure 3

Generic Communications - Master Table GENERIC COMMUNICATIONS: MASTER TABLEITEMTITLE*ADDITIONAL INFORMATION REVB 71-002PWR Reactor Trip Circuit Breakers NAAddressed to specific plant(s).B 71-003Catastrophic Failure of Main Steam Line Relief Valve Headers NAAddressed to specific plant(s).B 72-001Failed Hangers for Emergency Core Cooling System Suction Header NAAddressed to specific plant(s).B 72-002Simultaneous Actuation of a Safety Injection Signal on Both Units of a Dual Unit Facility NAAddressed to specific plant(s).B 72-003Limitorque Valve Operator Failures NAAddressed to specific plant(s).B 73-001Faulty Overcurrent Trip Delay Device in Circuit Breakers for Engineered Safety Systems CTVA: letter dated April 4, 1973NRC: IR 390/391 75-5B 73-002Malfunction of Containment Purge Supply Valve Switch CTVA: letter dated August 22, 1973NRC: IR 390/391 75-5B 73-003Defective Hydraulic Snubbers and Restraints CTVA: letter dated February 7, 1985NRC: IR 390/391 85-08B 73-004Defective Bergen-Patterson Hydraulic Shock Absorbers CTVA: memo dated February 7, 1985NRC: IR 390/391 85-08B 73-005Manufacturing Defect in BWR Control Rods NABoiling Water ReactorB 73-006Inadvertent Criticality in a BWR NABoiling Water ReactorB 74-001Valve Deficiencies CTVA: letter dated April 15, 1974NRC: IR 390/391 75-5B 74-002Truck Strike Possibility NAInfoPage 1 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 74-003Failure of Structural or Seismic Support Bolts on Class I Components CITVA: memo dated January 22, 1985NRC: 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 74-004Malfunction of Target Rock Safety Relief Valves NABoiling Water ReactorB 74-005Shipment of an Improperly Shielded Source NADoes not apply to power reactor.B 74-006Defective Westinghouse Type W-2 Control Switch Component CTVA: letter dated October 18, 1974NRC: IR 390/391 75-6B 74-007Personnel Exposure - Irradiation Facility NADoes not apply to power reactor.B 74-008Deficiency in the ITE Molded Case Circuit Breakers, Type HE-3 CTVA: letter dated August 21, 1974NRC: IR 390/391 75-5B 74-009Deficiency in GE Model 4KV Magne-Blast Circuit Breakers CTVA: letter dated September 20, 1974NRC: IR 390/391 76-6B 74-010Failures in 4-Inch Bypass Pipe at Dresden 2 NABoiling Water ReactorB 74-011Improper Wiring of Safety Injection Logic at Zion 1 & 2 CNRC: IR 390/391 75-6B 74-012Incorrect Coils in Westinghouse Type SG Relays at Trojan CNRC: IR 390/391 75-5B 74-013Improper Factory Wiring on GE Motor Control Centers at Fort Calhoun CTVA: letter dated December 24, 1974NRC: IR 390/391 75-5B 74-014BWR Relief Valve Discharge to Suppression Pool NABoiling Water ReactorPage 2 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 74-015Misapplication of Cutler-Hammer Three Position Maintained Switch Model No.

10250T CITVA: letter dated May 5, 1975NRC: IR 390/391 75-5

Unit 2 Action: Install modified A3 Cutler-Hammer 10250T switches.B 74-016Improper Machining of Pistons in Colt Industries (Fairbanks-Morse)

Diesel-Generators CTVA: letter dated January 2, 1975NRC: IR 390/391 75-3B 75-001Through-Wall Cracks in Core Spray Piping at Dresden-2 NABoiling Water ReactorB 75-002Defective Radionics Radiograph Exposure Devices and Source Changers NADoes not apply to power reactor.B 75-003Incorrect Lower Disc Spring and Clearance Dimension in Series 8300 and 8302 ASCO Solenoid Valves CITVA: letter dated May 16, 1975NRC: 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 75-004Cable Fire at BFNPP CINRC: IR 390/391 85-08 Closed to Fire Protection CAP

Part of Fire Protection CAPB 75-005Operability of Category I Hydraulic Shock and Sway Suppressors CITVA: letter dated June 16, 1975NRC: 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.Page 3 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 75-006Defective Westinghouse Type OT-2 Control Switches CITVA: letter dated July 31, 1975NRC: IR 390/85-25 and 391/85-20

Unit 2 Action: Inspect Westinghouse Type OT-2 control switches.

[WAS "NOTE 3."]B 75-007Exothermic Reaction in Radwaste Shipment NADoes not apply to power reactor.B 75-008PWR Pressure Instrumentation SNRC: 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.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:

Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.Adherence to Pressure and Temperature limits is required by the following portions of the Unit 2 TS: 1.1 [definition of "PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)"]; 3.4.3 ["RCS Pressure and Temperature (P/T) Limits"]; 3.4.12 ["Cold Overpressure Mitigation System (COMS)"]; and 5.9.6 ["Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)"].

02B 76-001BWR Isolation Condenser Tube Failure NABoiling Water ReactorB 76-002Relay Coil Failures - GE Types HFA, HGA, HKA, HMA Relays CIUnit 2 Action: Repair or replace relays before preoperational tests.B 76-003Relay Malfunctions - GE Type STD Relays CTVA: letter dated May 17, 1976NRC: IR 390/391 76-6B 76-004Cracks in Cold Worked Piping at BWRs NABoiling Water ReactorB 76-005Relay Failures - Westinghouse BFD Relays CTVA: letter dated June 7, 1976NRC: IR 390/391 85-08Page 4 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 76-006Diaphragm Failures in Air Operated Auxiliary Actuators for Safety/Relief Valves CTVA: memo dated January 25, 1985NRC: IR 390/391 85-08B 76-007Crane Hoist Control Circuit Modifications CTVA: letter dated October 29, 1976NRC: IR 390/391 85-08B 76-008Teletherapy Units NADoes not apply to power reactor.B 77-001Pneumatic Time Delay Relay Setpoint Drift CTVA: letter dated July 1, 1977NRC: IR 390/391 85-08B 77-002Potential Failure Mechanism in Certain Westinghouse AR Relays with Latch Attachments CTVA: letter dated November 11, 1977NRC: IR 390/391 85-08B 77-003On-Line Testing of the Westinghouse Solid State Protection System CIUnit 2 Action: Include necessary periodic testing in test procedures.B 77-004Calculation Error Affecting The Design Performance of a System for Controlling pH of Containment Sump Water Following a LOCA STVA: letter dated January 23, 1978NRC: IR 390/78-11 and 391/78-09

Unit 2 Action: Ensure Technical Specifications includes limit on Boron concentration.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:

Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.TS Surveillance Requirement 3.6.11.5 requires verification that the boron concentration is within a specified range.

02B 77-005 andB 77-005 AElectrical Connector Assemblies CTVA: letter dated January 17, 1978NRC: IR 390/78-11 and 391/78-09B 77-006Potential Problems with Containment Electrical Penetration Assemblies CItem was applicable only to units with operating license at the time the item was issued.

NRC: IR 390/391 85-08Page 5 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 77-007Containment Electrical Penetration Assemblies at Nuclear Power Plants Under Construction CTVA: letter dated January 20, 1978NRC: IR 390/78-11 and 391/78-09B 77-008Assurance of Safety and Safeguards During an Emergency - Locking Systems CItem concerns a multi-unit issue that was completed for both units.--------------------TVA: letter dated March 1, 1978 NRC: IR 390/78-11 and 391/78-09B 78-001Flammable Contact - Arm Retainers in GE CR120A Relays CTVA: letter dated March 20, 1978NRC: IR 390/78-11 and 391/78-09B 78-002Terminal Block Qualification CTVA: letter dated March 1, 1978NRC: IR 390/78-11 and 391/78-09B 78-003Potential Explosive Gas Mixture Accumulations Associated with BWR Offgas System Operations NABoiling Water ReactorB 78-004Environmental Qualification of Certain Stem Mounted Limit Switches Inside Reactor Containment CITVA: letter dated December 19, 1978NRC: 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 78-005Malfunctioning of Circuit Breaker Auxiliary Contact Mechanism - GE Model CR105X CTVA: letter dated June 12, 1978NRC: IR 390/78-17 and 391/78-15B 78-006Defective Cutler-Hammer Type M Relays With DC Coils CNRC: IR 390/78-22 and 391/78-19B 78-007Protection Afforded by Air-Line Respirators and Supplied-Air Hoods NAItem was applicable only to units with operating license at the time the item was issued.B 78-008Radiation Levels from Fuel Element Transfer Tubes NAItem was applicable only to units with operating license at the time the item was issued.

NRC: IR 390/391 85-08Page 6 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 78-009BWR Drywell Leakage Paths Associated with Inadequate Drywell Closures NABoiling Water ReactorB 78-010Bergen-Patterson Hydraulic Shock Suppressor Accumulator Spring Coils CTVA: letter dated August 14, 1978NRC: IR 390/78-22 and 391/78-19B 78-011Examination of Mark I Containment Torus Welds NABoiling Water ReactorB 78-012Atypical Weld Material in Reactor Pressure Vessel Welds CTVA: Westinghouse letter dated October 29, 1979NRC: IR 390/391 81-04B 78-013Failures in Source Heads Kay Ray, Inc. Gauges Models 7050, 7050B, 7051, 7051B, 7060, 7060B, 7061 and 7061B NADoes not apply to power reactor.B 78-014Deterioration of Buna-N Components in ASCO Solenoids NABoiling Water ReactorB 79-001Environmental Qualification of Class 1E Equipment CNRC: IR 390/80-06 and 391/80-05B 79-002Pipe Support Base Plate Designs Using Concrete Expansion Anchor Bolts CINRC review of HAAUP Program in NUREG-1232, SSER6, and SSER8. Unit 2 Actions:Addressed in CAP/SP.Conduct a complete review of affected support calculations, and perform the necessary revisions to design documents and field modifications to achieve compliance.B 79-003Longitudinal Weld Defects in ASME SA-312 Type 304 SS Pipe Spools Manufactured by Youngstown Welding &

Engineering CTVA: letter dated July 16, 1981NRC: IRs 390/82-21 and 391/82-17; 390/84-35 and 391/84-33B 79-004Incorrect Weights for Swing Check Valves Manufactured by Velan Engineering Corporation CTVA: letter dated October 20, 1980NRC: IR 390/83-15 and 391/83-11B 79-005Nuclear Incident at TMI NAApplies only to Babcock and Wilcox designed plantsB 79-006Review of Operational Errors and System Misalignments Identified During the Three Mile Island Incident CNRC: IR 390/80-06 and 391/80-05Page 7 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 79-007Seismic Stress Analysis of Safety-Related Piping CTVA: letter dated May 31, 1979NRC: IR 390/79-30 and 391/79-25B 79-008Events Relevant to BWRs Identified During TMI Incident NABoiling Water ReactorB 79-009Failure of GE Type AK-2 Circuit Breaker in Safety Related Systems CITVA: letter dated June 20, 1979

Unit 2 Action:

Complete preservice preventive maintenance on AK-2 Circuit Breakers.[WAS "NOTE 3."]B 79-010Requalification Training Program Statistics NAItem was applicable only to units with operating license at the time the item was issued.B 79-011Faulty Overcurrent Trip Device in Circuit Breakers for Engineering Safety Systems CTVA: letter dated July 20, 1979NRC: IR 390/79-30 and 391/79-25B 79-012Short Period Scrams at BWR Facilities NABoiling Water ReactorB 79-013Cracking in Feedwater Piping CItem 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 85-08B 79-014Seismic Analysis for As-Built Safety-Related Piping Systems CINRC review of HAAUP Program in NUREG-1232, SSER6, and SSER8.Unit 2 Actions:

Addressed in CAP/SP.Initiate a Unit 2 hanger walkdown and hanger analysis program similar to the program for Unit 1. Complete re-analysis of piping and associated supports as necessary. Perform modifications as required by re-analysis.B 79-015Deep Draft Pump Deficiencies CTVA: letter dated January 24, 1992NRC: IR 390/391 95-70Page 8 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 79-016Vital Area Access Controls NAItem was applicable only to units with operating license at the time the item was issued.

NRC: IR 390/80-06 and 391/80-05B 79-017Pipe Cracks in Stagnant Borated Water Systems at PWR Plants NAItem was applicable only to units with operating license at the time the item was issued.NRC: IR 390/80-06 and 391/80-05; NUREG/ CR 5286B 79-018Audibility Problems Encountered on Evacuation of Personnel from High-Noise Areas NAItem was applicable only to units with operating license at the time the item was issued.

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

NRC: IR 390/80-06 and 391/80-05B 79-020Packaging, Transport and Burial of Low-Level Radioactive Waste NAItem was applicable only to units with operating license at the time the item was issued.

NRC: IR 390/80-06 and 391/80-05B 79-021Temperature Effects on Level Measurements CIReviewed in 7.2.5 of both the original 1982 SER and SSER14. Unit 2 Action:

Update accident calculation.

CONFIRMATORY ISSUE - address IEB 79-21 to alleviate temperature dependence problem associated with measuring SG water levelIn 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 79-022Possible Leakage of Tubes of Tritium Gas Used in Time Pieces for Luminosity NADoes not apply to power reactor.

NRC: IR 390/80-06 and 391/80-05Page 9 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 79-023Potential Failure of Emergency Diesel Generator Field Exciter Transformer CTVA: letter dated October 29, 1979NRC: IR 390/80-06 and 391/80-05B 79-024Frozen Lines CIUnit 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 79-025Failures of Westinghouse BFD Relays in Safety-Related Systems CTVA: letter dated January 4, 1980NRC: IR 390/80-03 and 391/80-02B 79-026Boron Loss from BWR Control Blades NABoiling Water ReactorB 79-027Loss of Non-Class 1E I & C Power System Bus During Operation CITVA responded to the Bulletin on March 1, 1982. Reviewed in 7.5.3 of the original 1982 SER. Unit 2 Action:

Issue appropriate emergency procedures.B 79-028Possible Malfunction of NAMCO Model EA180 Limit Switches at Elevated Temperatures CTVA: letter dated April 1, 1993NRC: IR 390/391 93-32B 80-001Operability of ADS Valve Pneumatic Supply NABoiling Water ReactorB 80-002Inadequate QA for Nuclear Supplied Equipment NABoiling Water ReactorB 80-003Loss of Charcoal from Standard Type II, 2 Inch, Tray Adsorber Cells CTVA: letter dated March 21, 1980NRC: IR 390/80-15 and 391/80-12B 80-004Analysis of a PWR Main Steam Line Break with Continued Feedwater Addition CIIR 50-390/85-60 and 50-391/85-49 (December 6, 1985) required completion of actions that included determination of temperature profiles inside and outside of containment following a MSLB for Unit 1. Unit 2 Action:

Complete analysis for Unit 2.Page 10 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 80-005Vacuum Condition Resulting in Damage to Chemical Volume Control System Holdup Tanks CIClosed in IR 50-390/84-59 and 50-391/84-45. Unit 2 Action:

Complete surveillance procedures for Unit 2.B 80-006Engineered Safety Feature Reset Control CITVA 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 80-007BWR Jet Pump Assembly Failure NABoiling Water ReactorB 80-008Examination of Containment Liner Penetration Welds CTVA: letter dated July 8, 1980NRC: IR 390/391 81-19B 80-009Hydramotor Actuator Deficiencies CTVA: letter dated January 15, 1981NRC: NUREG/ CR 5291; IR 390/391 85-08; IR 390/85-60 and 391/85-49B 80-010Contamination of Nonradioactive System and Resulting Potential for Unmonitored, Uncontrolled Release of Radioactivity to Environment CIUnit 2 Actions: 2) Include proper monitoring of non-radioactive systems in procedures.B 80-010Contamination of Nonradioactive System and Resulting Potential for Unmonitored, Uncontrolled Release of Radioactivity to Environment CIUnit 2 Actions: 1) Correct deficiencies involving monitoring of systems.B 80-011Masonry Wall Design CINRC 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 80-012Decay Heat Removal System Operability CINRC: IR 390/391 85-08; NUREG/CR 4005

Unit 2 Action:

Implement operating instructions and abnormal operating instructions (AOIs) for RHR.[WAS "NOTE 3."]Page 11 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 80-013Cracking in Core Spray Spargers NABoiling Water ReactorB 80-014Degradation of Scram Discharge Volume Capability NABoiling Water ReactorB 80-015Possible Loss of Emergency Notification System with Loss of Offsite Power CItem concerns a multi-unit issue that was completed for both units.--------------------

NRC: IR 390/391 85-08B 80-016Potential Misapplication of Rosemount, Inc. Models 1151 and 1152 Pressure Transmitters With Either "A" or "D" Output Codes CTVA: letter dated August 29, 1980NRC: IR 390/391 81-17B 80-017Failure of 76 of 185 Control Rods to Fully Insert During a Scram at a BWR NABoiling Water ReactorB 80-018Maintenance of Adequate Minimum Flow Thru Centrifugal Charging Pumps Following Secondary Side High Energy

Rupture CIIR 50-390/85-60 and 50-391/85-49 (Unit 1)Unit 2 Action:

Implement design and procedure changes.B 80-019Mercury-Wetted Matrix Relay in Reactor Protective Systems of Operating Nuclear Power Plants Designed by CE CTVA: letter dated September 4, 1980NRC: NUREG/CR 4933; IR 390/391 81-17B 80-020Failure of Westinghouse Type W-2 Spring Return to Neutral Control Switches CIUnit 2 Action: Modify switches.B 80-021Valve Yokes Supplied by Malcolm Foundry Co., Inc.

CTVA: letter dated May 6, 1981NRC: 390/391 85-08B 80-022Automation Industries, Model 200-520-008 Sealed-Source Connectors NADoes not apply to power reactor.B 80-023Failures of Solenoid Valves Manufactured by Valcor Engineering Corporation CTVA: letter dated March 31, 1981NRC: IR 390/391 81-17; NUREG/CR 5292B 80-024Prevention of Damage Due to Water Leakage Inside Containment (10/17/80 Indian Point 2 Event)

CIUnit 2 Action: Confirm that the reactor cavity can not be flooded, resulting in the partial or total submergence of the reactor vessel unnoticed by the reactor operators.Page 12 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 80-025Operating Problems with Target Rock Safety-Relief Valves at BWRs NABoiling Water ReactorB 81-001Surveillance of Mechanical Snubbers NANRC: IR 390/391 81-17B 81-002Failure of Gate Type Valves to Close Against Differential Pressure CTVA: letter dated September 30, 1983NRC: IR 390/391 84-03B 81-003Flow Blockage of Cooling Water to Safety System Components by Asiatic Clams and Mussels CTVA: letters dated July 21, 1981 and March 21, 1983NRC: IR 390/391 81-17B 82-001Alteration of Radiographs of Welds in Piping Subassemblies CNRC: IR 390/391 85-08B 82-002Degradation of Threaded Fasteners in the Reactor Coolant Pressure Boundary of PWR Plants CITVA: memo dated February 6, 1985NRC: 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 82-003Stress Corrosion Cracking in Thick-Wall, Large Diameter, Stainless Steel, Recirculation System Piping at BWR Plants NABoiling Water ReactorB 82-004Deficiencies in Primary Containment Electrical Penetration Assemblies CTVA: letter dated January 24, 1983NRC: IR 390/83-10 and 391/83-08B 83-001Failure of Trip Breakers (Westinghouse DB-50) to Open on Automatic Trip Signal CNRC: IRs 390/391 85-08 and 390/391 92-13B 83-002Stress Corrosion Cracking in Large-Diameter Stainless Steel Recirculation System Piping at BWR Plants NABoiling Water ReactorB 83-003Check Valve Failures in Raw Water Cooling Systems of Diesel Generators NAAddressed by Inservice Testing for Construction Permit holdersPage 13 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 83-004Failure of the Undervoltage Trip Function of Reactor Trip Breakers CINRC: IR 390/391 85-08

Unit 2 Action: Install new undervoltage attachment with wider grooves on the reactor trip breakers.B 83-005ASME Nuclear Code Pumps and Spare Parts Manufactured by the Hayward Tyler Pump Company CTVA: letter dated September 7, 1983NRC: IR 390/85-03 and 391/85-04; NUREG/CR 5297B 83-006Nonconforming Material Supplied by Tube-Line Facilities CITVA: letter dated February 2, 1984NRC: IR 390/391 84-03; NUREG/CR 4934

NRC SER for both units dated September 23, 1991, provided an alternate acceptance for fittings supplied by Tube-Line. Unit 2 Action: Implement as necessary.

REVISION 04 UPDATE:NRC Inspection Report Nos. 50-390/90-02 and50-391/90-02 found the proposed alternative to ASME code paragraph NA-3451 (a) to be acceptable. It noted that TVA must revise the FSAR to document this deviation from ASME Section IIIrequirements.TVA letter to NRC dated October 11, 2007, stated the Unit 1 exemption is applicable to Unit 2 and was submitted to the NRC as being required for Unit 2 construction.Final action was to incorporate the exemption in the Unit 2 FSAR. This exemption is documented in Unit 2 FSAR Section 3.2 in paragraph 3.2.3.2 and Table 3.2-2a as explained in Note 4. of the table.

04B 83-007Apparently Fraudulent Products Sold by Ray Miller, Inc.

CTVA: letter dated March 22, 1984NRC: IR 390/85-03 and 391/85-04B 83-008Electrical Circuit Breakers With an Undervoltage Trip Feature in Safety-Related Applications Other Than the Reactor Trip System CTVA: letter dated March 29, 1984NRC: IR 390/84-35 and 391/84-33B 84-001Cracks in BWR Mark 1 Containment Vent Headers NABoiling Water ReactorPage 14 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 84-002Failure of GE Type HFA Relays In Use In Class 1E Safety Systems CTVA: letter dated July 10, 1984NRC: IR 390/391 84-42 and IR 390/84-77 and 391/84-54B 84-003Refueling Cavity Water Seal CIReviewed in IR 390/93-11. Unit 2 Action:

Ensure appropriate abnormal operating instructions (AOIs) are used for Unit 2.B 85-001Steam Binding of Auxiliary Feedwater Pumps CITVA: letter dated January 27, 1986NRC: 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 85-002Undervoltage Trip Attachment of Westinghouse DB-50 Type Reactor Trip Breakers CIUnit 2 Action: Install automatic shunt trip on the Westinghouse DS-416 reactor trip breakers on Unit 2.B 85-003Motor-Operated Valve Common Mode Failures During Plant Transients Due to Improper Switch Settings CSuperseded by GL 89-10B 86-001Minimum Flow Logic Problems That Could Disable RHR Pumps NABoiling Water ReactorB 86-002Static "O" Ring Differential Pressure Switches CTVA: letter dated November 20, 1986NRC: IR 390/391/90-24B 86-003Potential Failure of Multiple ECCS Pumps Due to Single Failure of Air-Operated Valve in Minimum Flow Recirculation Line CTVA: letter dated November 14, 1986NRC: IR 390/391/87-03B 86-004Defective Teletherapy Timer That May Not Terminate Treatment Dose NADoes not apply to power reactor.Page 15 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 87-001Thinning of Pipe Walls in Nuclear Power Plants CTVA: letter dated September 18, 1987NRC: NUREG/CR 5287

Closed to GL 89-08B 87-002Fastener Testing to Determine Conformance with Applicable Material Specifications CITVA: letters dated April 15, 1988, July 6, 1988, September 12, 1988, and January 27, 1989NRC: letter dated August 18, 1989

NRC closed in letter dated August 18, 1989.

Unit 2 Action:

Complete for Unit 2, using information used for Unit 1, as applicable.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 03 UPDATE:

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

03B 88-001Defects in Westinghouse Circuit Breakers CTVA: letter dated November 15, 1991NRC: IR 390/391 93-01B 88-002Rapidly Propagating Fatigue Cracks in Steam Generator Tubes CINRC acceptance letter dated June 7, 1990, for both units. Unit 2 Actions: Evaluate E/C data to determine anti-vibration bar penetration depth;perform T/H analysis to identify susceptible tubes; modify, if necessary.B 88-003Inadequate Latch Engagement in HFA Type Latching Relays Manufactured by General Electric (GE) Company CTVA: letter dated April 13, 1992NRC: IR 390/391 92-13B 88-004Potential Safety-Related Pump Loss CINRC acceptance letter dated May 24, 1990, for both units. Unit 2 Actions: Perform calculations, and install check valves to prevent pump to pump interaction.Page 16 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 88-005Nonconforming Materials Supplied by Piping Supplies, Inc. and West Jersey Manufacturing Company CINRC reviewed in Appendix EE of SSER16. Unit 2 Actions:

Complete review to locate installed WJM material, and perform in-situ hardness testing for Unit 2.B 88-006Actions to be Taken for the Transfer of Model No. SPEC 2-T Radiographic Exposure Device NADoes not apply to power reactor.B 88-007Power Oscillations in BWRs NABoiling Water ReactorB 88-008Thermal Stresses in Piping Connected to Reactor Cooling Systems CINRC acceptance letter dated September 19, 1991, for both units. Unit 2 Action: Implement program to prevent thermal stratification.B 88-009Thimble Tube Thinning in Westinghouse Reactors CIReviewed in Appendix EE of SSER16. 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 88-010Nonconforming Molded-Case Circuit Breakers CIUnit 2 Action: Replace those circuits not traceable to a circuit breaker manufacturer.B 88-011Pressurizer Surge Line Thermal Stratification CINRC SER on "Leak-Before-Break" (April 28, 1993) and reviewed in Appendix EE of SSER16. Unit 2 Actions:

Complete modifications to accommodate Surge Line thermal movements, and incorporate a temperature limitation during heatup and cooldown operations into Unit 2 procedures.B 89-001Failure of Westinghouse Steam Generator Tube Mechanical Plugs CINRC acceptance letter dated September 26, 1991 for both units. Unit 2 Action:

Remove SG tube plugs.Page 17 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 89-002Stress Corrosion Cracking of High-Hardness Type 410 Stainless Steel Internal Preloaded Bolting in Anchor Darling Model S350W Swing Check Valves or Valves of Similar Nature CINRC reviewed in Appendix EE of SSER16.

Unit 2 Actions:

Replace the flapper assembly hold-down bolts fabricated on the 14 (12 valves are installed) Atwood and Morrell Mark No. 47W450-53 check valves.Replacement bolts are to be fabricated from ASTM F593 Alloy 630.A review of the remaining Unit 2 safety related swing check valves will be performed.B 89-003Potential Loss of Required Shutdown Margin During Refueling Operations CITVA: letter dated June 19, 1990NRC: 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 90-001Loss of Fill-Oil in Transmitters Manufactured by Rosemount CIUnit 2 Action: Implement applicable recommendations from this Bulletin including identification of potentially defective transmitters and an enhanced surveillance program which monitors transmitters for loss of fill oil.B 90-002Loss of Thermal Margin Caused by Channel Box Bow NABoiling Water ReactorB 91-001Reporting Loss of Criticality Safety Controls NADoes not apply to power reactor.B 92-001Failure of Thermo-Lag 330 Fire Barrier System to Maintain Cabling in Wide Cable Trays and Small Conduits Free From Fire Damage NA--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:

This bulletin was provided for information only to plants with construction permits. See Generic Letter 92-08 for Thermo-lag related actions.

02B 92-002Safety Concerns Related to "End of Life" of Aging Theratronics Teletherapy Units NADoes not apply to power reactor.B 92-003Release of Patients After Brachytherapy NADoes not apply to power reactor.Page 18 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 93-001Release of Patients After Brachytherapy Treatment with Remote Afterloading Devices NADoes not apply to power reactor.B 93-002Debris Plugging of Emergency Core Cooling Suction Strainers CBoiling Water Reactor--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:

In Rev. 01, this was characterized as "NA - BWR only". This Bulletin was provided for Information to holders of construction permits. No WBN response was found. B-93-02 was closed in IR 50-390/94-04 and 50-391/94-04.

02B 93-003Resolution of Issues Related to Reactor Vessel Water Level Instrumentation in BWRs NABoiling Water ReactorB 94-001Potential Fuel Pool Draindown Caused by Inadequate Maintenance Practices at Dresden Unit 1 NAAddressed to holders of licenses for nuclear power reactors that are permanently shut down with spent fuel in the spent fuel poolB 94-002Corrosion Problems in Certain Stainless Steel Packagings Used to Transport Uranium Hexafluoride NADoes not apply to power reactor.B 95-001Quality Assurance Program for Transportation of Radioactive Material NADoes not apply to power reactor.B 95-002Unexpected Clogging of a Residual Heat Removal Pump Strainer While Operating in Suppression Pool Cooling Mode NABoiling Water ReactorPage 19 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 96-001, first partControl Rod Insertion Problems (PWR)

CINRC acceptance letter for Unit 1 dated July 22, 1996 - Initial response for Unit 2 on September 7, 2007. Unit 2 Action:

Issue Emergency Operating Procedure.

REVISION 02 UPDATE:

Unit 2 will load all new RFA-2 fuel for the initial fuel load.

REVISION 03 UPDATE:

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

REVISION 04 UPDATE:

Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.

04B 96-001, last partControl Rod Insertion Problems (PWR)

CINRC acceptance letter for Unit 1 dated July 22, 1996 - Initial response for Unit 2 on September 7, 2007. Unit 2 Action:

and provide core map.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 03 UPDATE:NRC issued the Safety Evaluation (corrected) for Bulletin 1996-001 on May 3, 2010.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 04 UPDATE:

Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.

04Page 20 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 96-002Movement of Heavy Loads over Spent Fuel, Over Fuel in the Reactor, or Over Safety-Related Equipment CINRC closure letter dated May 20, 1998. Unit 2 Action:

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

REVISION 02 UPDATE:

NRC issued the Safety Evaluation for Bulletin 1996-002 on March 4, 2010.

02B 96-003Potential Plugging of ECCS Suction Strainers by Debris in BWRs NABoiling Water ReactorB 96-004Chemical, Galvanic, or Other Reactions in Spent Fuel Storage and Transportation Casks NAInfoB 97-001Potential for Erroneous Calibration, Dose Rate, or Radiation Exposure Measurements with Certain Victoreen Model 530 and 531SI Electrometer/Dosemeters NADoes not apply to power reactor.B 97-002Puncture Testing of Shipping Packages Under 10 CFR Part 71 NADoes not apply to power reactor.Page 21 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 01-001Circumferential Cracking of Reactor Pressure Vessel (RPV) Head Penetration Nozzles CINRC acceptance letter dated November 20, 2001 (Unit 1) - Initial response for Unit 2 on September 7, 2007. Unit 2 Action:

Perform baseline inspection.

REVISION 02 UPDATE:

Unit 2 Actions:

Perform baseline inspection.

Evaluate or repair as necessary.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 03 UPDATE:NRC issued the Safety Evaluation for Bulletin 2001-001 on June 30, 2010.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 04 UPDATE:

Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.

04Page 22 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 02-001RPV Head Degradation and Reactor Coolant Pressure Boundary Integrity CINRC review of Unit 1's 15 day response in letter dated May 20, 2002 - Initial response for Unit 2 on September 7, 2007. Unit 2 Action: Perform baseline inspection.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:Unit 2 Actions:

Perform baseline inspection.

Evaluate or repair as necessary.

REVISION 03 UPDATE:

NRC issued the Safety Evaluation for Bulletin 2002-001 on June 30, 2010.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 04 UPDATE:

Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.

04Page 23 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 02-002RPV Head and Vessel Head Penetration Nozzle Inspection Programs CINRC acceptance letter dated December 20, 2002 (Unit 1) - Initial response for Unit 2 on September 7, 2007. Unit 2 Action:

Perform baseline inspection.

REVISION 02 UPDATE:

Unit 2 Actions:

Perform baseline inspection.

Evaluate or repair as necessary.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 03 UPDATE:NRC issued the Safety Evaluation for Bulletin 2002-002 on June 30, 2010.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 04 UPDATE:

Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.

04B 03-001Potential Impact of Debris Blockage on Emergency Sump Recirculation at PWRs NATVA: letter dated September 7, 2007B 03-002Leakage from RPV Lower Head Penetrations and Reactor Coolant Pressure Boundary Integrity (PWRs)

CINRC acceptance letter dated October 6, 2004 (Unit 1) - Initial response for Unit 2 on September 7, 2007. Unit 2 Action:

Perform baseline inspection.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

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

02Page 24 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 03-003Potentially Deficient 1-inch Valves for Uranium Hexaflouride Cylinders NADoes not apply to power reactor.B 03-004Rebaselining of Data in the Nuclear Management and Safeguards System CTVA: letter dated December 18, 2003

Item concerns a multi-unit issue that was completed for both units.Page 25 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 04-001Inspection of Alloy 82/182/600 Materials Used in the Fabrication of Pressurizer Penetrations and Steam Space Piping Connections at PWRs CIInitial response for Unit 2 on September 7, 2007. Unit 2 Actions:

Provide details of pressurizer and penetrations, and apply Material Stress Improvement Process.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:TVA provided details of the pressurizer and penetrations on September 29, 2008. This letter committed to:Prior to placing the pressurizer in service, TVA will apply theMaterial Stress Improvement Process (MSIP) to the PressurizerPower Operated Relief Valve connections, the safety relief valveconnections, the spray line nozzle and surge line nozzle connections.TVA will perform a bare metal visual (BMV) inspection of the upper pressurizer Alloy 600 locations at the first refueling outage.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 03 UPDATE:April 1, 2010, letter committed to:

TVA will perform NDE prior to and after performance of the MSIP. If circumferential cracking is observed in either pressure boundary or non-pressure boundary portions of any locations covered under the scope of the bulletin, TVA will develop plans to perform an adequate extent-of-condition evaluation, and TVA will discuss those plans with cognizant NRC technical staff prior to starting Unit 2.After performing the BMV inspection during the first refueling outage, if any evidence of apparent reactor coolant pressure boundary leakage is discovered, then NDE capable of determining crack orientation will be performed in order to accurately characterize the flaw, the orientation, and extent. TVA will develop plans to perform an adequate extent of condition evaluation, and plans to possibly expand the scope of NDE to other components in the pressurizer will be discussed with NRC technical staff prior to restarting of Unit 2.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 04 UPDATE:

NRC issued the Safety Evaluation for Bulletin 2004-001 on August 4, 2010.

04Page 26 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVB 05-001Material Control and Accounting at Reactors and Wet Spent Fuel Storage Facilities CTVA: letters dated March 21, 2005 and May 11, 2005

Item concerns a multi-unit issue that was completed for both units.B 05-002Emergency Preparedness and Response Actions for Security-Based Events CTVA: letters dated January 20, 2006 and August 16, 2006.

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

REVISION 05 UPDATE:

The NRC closed this bulletin via letter dated March 25, 2010 (ADAMS Accession No. ML100770549).

05C 76-001Crane Hoist Control Circuit Modifications CSee B 76-007 for additional information.C 76-002Relay Failures - Westinghouse BF (AC) and BFD (DC) Relays CTVA: letter dated November 22, 1976 informed NRC that these relay types are not used in Class 1E circuits.NRC: IR 50/390/76-11 and 50/391/76-11C 76-003Radiation Exposures in Reactor Cavities NAInfoC 76-004Neutron Monitor and Flow Bypass Switch Malfunctions NABoiling Water ReactorC 76-005Hydraulic Shock And Sway Suppressors - Maintenance of Bleed and Lock-Up Velocities on ITT Grinnell's Model Nos. - Fig. 200 And Fig. 201, Catalog Ph-74-R CTVA: letter dated January 7, 1977 informed NRC that no Grinnell shock suppressors or sway braces have been or will be installed at WBN.C 76-006Stress Corrosion Cracks in Stagnant, Low Pressure Stainless Piping Containing Boric Acid Solution at PWRs NAItem was applicable only to units with operating license at the time the item was issued.C 76-007Inadequate Performance by Reactor Operating and Support Staff Members NAItem was applicable only to units with operating license at the time the item was issued.Page 27 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVC 77-001Malfunctions of Limitorque Valve Operators NAInfoC 77-002aPotential Heavy Spring Flooding (CP)

NAItem was applicable only to units with operating license at the time the item was issued.C 77-003Fire Inside a Motor Control Center NAInfoC 77-004Inadequate Lock Assemblies NAInfoC 77-005Fluid Entrapment in Valve Bonnets NAInfoC 77-006Effects of Hydraulic Fluid on Electrical Cables NAInfoC 77-007Short Period During Reactor Startup NABoiling Water ReactorC 77-008Failure of Feedwater Sample Probe NAItem was applicable only to units with operating license at the time the item was issued.C 77-009Improper Fuse Coordination in BWR Standby Liquid Control System Control

Circuits NABoiling Water ReactorC 77-010Vacuum Conditions Resulting in Damage to Liquid Process Tanks NAItem was applicable only to units with operating license at the time the item was issued.C 77-011Leakage of Containment Isolation Valves with Resilient Seats NAInfoC 77-012Dropped Fuel Assemblies at BWR Facilities NABoiling Water ReactorC 77-013Reactor Safety Signals Negated During Testing NAInfoC 77-014Separation of Contaminated Water Systems from Noncontaminated Plant Systems NAInfoC 77-015Degradation of Fuel Oil Flow to the Emergency Diesel Generator NAInfoPage 28 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVC 77-016Emergency Diesel Generator Electrical Trip Lock-Out Features NAInfoC 78-001Loss of Well Logging Source NADoes not apply to power reactor.C 78-002Proper Lubricating Oil for Terry Turbines NAInfoC 78-003Packaging Greater Than Type A Quantities of Low Specific Activity Radioactive Material for Transport NAInfoC 78-004Installation Errors That Could Prevent Closing of Fire Doors NAInfoC 78-005Inadvertent Safety Injection During Cooldown NAInfoC 78-006Potential Common Mode Flooding of ECCS Equipment Rooms at BWR Facilities NAInfoC 78-007Damaged Components of a Bergen-Paterson Series 25000 Hydraulic Test Stand NAInfoC 78-008Environmental Qualification of Safety-Related Electrical Equipment at Nuclear Power Plants NAInfoC 78-009Arcing of General Electric Company Size 2 Contactors NAInfoC 78-010Control of Sealed Sources in Radiation Therapy NADoes not apply to power reactor.C 78-011Recirculation MG Set Overspeed Stops NABoiling Water ReactorC 78-012HPCI Turbine Control Valve Lift Rod Bending NABoiling Water ReactorC 78-013Inoperability of Service Water Pumps NAInfoC 78-014HPCI Turbine Reversing Chamber Hold Down Bolting NABoiling Water ReactorPage 29 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVC 78-015Tilting Disc Check Valves Fail to Close with Gravity in Vertical Position NAInfoC 78-016Limitorque Valve Actuators NAInfoC 78-017Inadequate Guard Training/Qualification and Falsified Training Records NAInfoC 78-018UL Fire Test NAInfoC 78-019Manual Override (Bypass) of Safety System Actuation Signals NAInfoC 79-001Administration of Unauthorized Byproduct Material to Humans NADoes not apply to power reactor.C 79-002Failure of 120 Volt Vital AC Power Supplies NAInfoC 79-003Inadequate Guard Training - Qualification and Falsified Training Records NAInfoC 79-004Loose Locking Nut on Limitorque Valve Operators NAInfoC 79-005Moisture Leakage in Stranded Wire Conductors NAInfoC 79-006Failure to Use Syringe and Bottle Shields in Nuclear Medicine NADoes not apply to power reactor.C 79-007Unexpected Speed Increase of Reactor Recirculation MG Set Resulted in Reactor Power Increase NABoiling Water ReactorC 79-008Attempted Extortion - Low Enriched Uranium NAFuel facilities and operating reactors at the time the item was issuedC 79-009Occurrences of Split or Punctured Regulator Diaphragms in Certain Self Contained Breathing Apparatus NAInfoC 79-010Pipefittings Manufactured from Unacceptable Material NAInfoPage 30 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVC 79-011Design/Construction Interface Problem NAInfoC 79-012Potential Diesel Generator Turbocharger Problem NAInfoC 79-013Replacement of Diesel Fire Pump Starting Contactors NAInfoC 79-014Unauthorized Procurement and Distribution of XE-133 NADoes not apply to power reactor.C 79-015Bursting of High Pressure Hose and Malfunction of Relief Valve O-Ring in Certain Self-Contained Breathing Apparatus NAItem was applicable only to units with operating license at the time the item was issued.C 79-016Excessive Radiation Exposures to Members of the General Public and a Radiographer NADoes not apply to power reactor.C 79-017Contact Problem in SB-12 Switches on General Electric Company Metalclad Circuit Breakers NAInfoC 79-018Proper Installation of Target Rock Safety-Relief Valves NABoiling Water ReactorC 79-019Loose Locking Devices on Ingersoll-Rand Pumps NAInfoC 79-020Failure of GTE Sylvania Relay Type PM Bulletin 7305 Catalog 5U12-11-AC with a 120V AC Coil NAInfoC 79-021Prevention of Unplanned Releases of Radioactivity NAInfoC 79-022Stroke Times for Power Operated Relief Valves NAInfoPage 31 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVC 79-023Motor Starters and Contactors Failed to Operate CThe Circular did not require a response. TVA reported a nonconformance under 10 CFR 50.55e on January 17, 1980, that four motor starters of this type had been located in the 480V control and auxiliary vent boards at WBN. Gould factory representatives supervised the replacement of the carrier assemblies in accordance with the Gould instructions. The starters with replaced carriers were acceptable.NRC IR 50-390/80-03 and 50-391/80-02 reviewed and closed the associated nonconformance reports.

01C 79-024Proper Installation and Calibration of Core Spray Pipe Break Detection Equipment on BWRs NABoiling Water ReactorC 79-025Shock Arrestor Strut Assembly Interference CThe Circular did not require a response.TVA reported a nonconformance under 10 CFR 50.55e onMarch 6, 1980, that a review had determined that nine installed supports had brackets with the potential of hindering full function of the support. Additional supports that were not installed had the same potential problem. TVA initially determined that the supports would be modified in accordance with a vendor approved drawing. TVA subsequently determined that no actual problem existed and no field work was required.NRC IR 50-390/83-15 and 50-391/83-11 reviewed and closed the associated nonconformance reports.

01C 80-001Service Advice for GE Induction Disc Relays NAInfoC 80-002Nuclear Power Plant Staff Work Hours NAInfoC 80-003Protection from Toxic Gas Hazards NAInfoC 80-004Securing of Threaded Locking Devices on Safety-Related Equipment NAInfoC 80-005Emergency Diesel-Generator Lubricating Oil Addition and Onsite Supply NAInfoC 80-006Control and Accountability Systems for Implant Therapy Sources NADoes not apply to power reactor.C 80-007Problems with HPCI Turbine Oil System NABoiling Water ReactorC 80-008BWR Technical Specification Inconsistency - RPS Response Time NABoiling Water ReactorPage 32 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVC 80-009Problems with Plant Internal Communications Systems NAInfoC 80-010Failure to Maintain Environmental Qualification of Equipment NAInfoC 80-011Emergency Diesel Generator Lube Oil Cooler Failures NAInfoC 80-012Valve-Shaft-to-Actuator Key May Fall Out of Place when Mounted Below Horizontal Axis NAInfoC 80-013Grid Strap Damage in Westinghouse Fuel Assemblies NAInfoC 80-014Radioactive Contamination of Plant Demineralized Water System and Resultant Internal Contamination of Personnel NAInfoC 80-015Loss of Reactor Coolant Pump Cooling and Natural Circulation Cooldown NAInfoC 80-016Operational Deficiencies in Rosemount Model 510DU Trip Units and Model 1152 Pressure Transmitters NAInfoC 80-017Fuel Pin Damage Due to Water Jet from Baffle Plate Corner NAInfoC 80-01810 CFR 50.59 Safety Evaluations for Changes to Radioactive Waste Treatment Systems NAInfoC 80-019Noncompliance with License Requirements for Medical Licensees NADoes not apply to power reactor.C 80-020Changes in Safe-Slab Tank Dimensions NAInfoC 80-021Regulation of Refueling Crews NAItem was applicable only to units with operating license at the time the item was issued.C 80-022Confirmation of Employee Qualifications NAInfoC 80-023Potential Defects in Beloit Power Systems Emergency Generators NAInfoPage 33 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVC 80-024AECL Teletherapy Unit Malfunction NADoes not apply to power reactor.C 80-025Case Histories of Radiography Events NADoes not apply to power reactor.C 81-001Design Problems Involving Indicating Pushbutton Switches Manufactured by Honeywell Incorporated NAInfoC 81-002Performance of NRC-Licensed Individuals while on Duty NAItem was applicable only to units with operating license at the time the item was issued.C 81-003Inoperable Seismic Monitoring Instrumentation NAInfoC 81-004The Role of Shift Technical Advisors and Importance of Reporting Operational Events NAInfoC 81-005Self-Aligning Rod End Bushings for Pipe Supports NAInfoC 81-006Potential Deficiency Affecting Certain Foxboro 10 to 50 Milliampere Transmitters NAInfoC 81-007Control of Radioactively Contaminated Material NAInfoC 81-008Foundation Materials NAInfoC 81-009Containment Effluent Water that Bypasses Radioactivity Monitor NAInfoC 81-010Steam Voiding in the Reactor Coolant System During Decay Heat Removal Cooldown NAItem was applicable only to units with operating license at the time the item was issued.C 81-011Inadequate Decay Heat Removal During Reactor Shutdown NABoiling Water ReactorC 81-012Inadequate Periodic Test Procedure of PWR Reactor Protection System NAInfoPage 34 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVC 81-013Torque Switch Electrical Bypass Circuit for Safeguard Service Valve Motors CThe Circular did not require a response.TVA reported a nonconformance under 10 CFR 50.55e on April 4, 1986 (NCR W367-P), that required closing torque switches were found improperly wired. This issue (Torque switch and overload relay bypass capability for active safety related valves) is part of the Electrical Issues Corrective Action Program for WBN Unit 2.

01C 81-014Main Steam Isolation Valve Failures to Close NAInfoC 81-015Unnecessary Radiation Exposures to the Public and Workers During Events Involving Thickness and Level Measuring Devices NAInfoGL 77-001Intrusion Detection Systems Handbook NAInfoGL 77-002Fire Protection Functional Responsibilities NAInfoGL 77-003Transmittal of NUREG-0321, "A Study of the Nuclear Regulatory Commission Quality Assurance Program" NAInfoGL 77-004Shipments of Contaminated Components From NRC Licensed Power Facilities to Vendors & Service Companies NAInfoGL 77-005Nonconformity of Addressees of Items Directed to the Office of Nuclear Reactor Regulation NAInfoGL 77-006Enclosing Questionnaire Related to Steam Generators NAItem was applicable only to units with operating license at the time the item was issued.GL 77-007Reliability of Standby Diesel Generator Units NAItem was applicable only to units with operating license at the time the item was issued.GL 77-008Revised Intrusion Detection Handbook and Entry Control Systems Handbook NAInfoGL 78-001Correction to Letter of 12/15/77 [GL 77-07]

NAItem was applicable only to units with operating license at the time the item was issued.GL 78-002Asymmetric Loads Background and Revised Request for Additional Information CNRC: Reviewed in SSER15 - Appendix C (June 1995). Resolved by approval of leak-before-break analysis.Page 35 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 78-003Request For Information on Cavity Annulus Seal Ring NAItem was applicable only to units with operating license at the time the item was issued.GL 78-004GAO Blanket Clearance for Letter Dated 12/09/77 [GL 77-06]

NAItem was applicable only to units with operating license at the time the item was issued.GL 78-005Internal Distribution of Correspondence - Asking for Comments on Mass Mailing System NAInfoGL 78-006This GL was never issued.

NAGL 78-007This GL was never issued.

NAGL 78-008Enclosing NUREG-0408 Re Mark I Containments, and Granting Exemption from GDC 50 and Enclosing Sample Notice NABoiling Water ReactorGL 78-009Multiple-Subsequent Actuations of Safety/Relief Valves Following an Isolation Event NABoiling Water ReactorGL 78-010Guidance on Radiological Environmental Monitoring NAInfoGL 78-011Guidance on Spent Fuel Pool Modifications NAInfoGL 78-012Notice of Meeting Regarding "Implementation of 10 CFR 73.55 Requirements and Status of Research -"

NAInfoGL 78-013Forwarding of NUREG-0219 NAInfoGL 78-014Transmittal of Draft NUREG-0219 for Comment NAInfoGL 78-015Request for Information on Control of Heavy Loads Near Spent Fuel NASee GL 81-007.GL 78-016Request for Information on Control of Heavy Loads Near Spent Fuel Pools NAInfoGL 78-017Corrected Letter on Heavy Loads Over Spent Fuel NAInfoPage 36 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 78-018Corrected Letter on Heavy Loads Over Spent Fuel NADuplicate of GL 81-007GL 78-019Enclosing Sandia Report SAND 77-0777, "Barrier Technology Handbook" NAInfoGL 78-020Enclosing - "A Systematic Approach to the Conceptual Design of Physical Protection Systems for Nuclear Facilities NAInfoGL 78-021Transmitting NUREG/CR-0181, "Concerning Barrier and Penetration Data Needed for Physical Security System Assessment" NAInfoGL 78-022Revision to Intrusion Detection Systems and Entry Control Systems Handbooks and Nuclear Safeguards Technology Handbook NAInfoGL 78-023Manpower Requirements for Operating Reactors NAInfoGL 78-024Model Appendix I Technical Specifications and Submittal Schedule For BWRs NABoiling Water ReactorGL 78-025This GL was never issued.

NAGL 78-026Excessive Control Rod Guide Tube Wear NAApplies only to Babcock and Wilcox designed plantsGL 78-027Forwarding of NUREG-0181 NAInfoGL 78-028Forwarding pages omitted from 07/11/78 letter [GL 78-24]

NABoiling Water ReactorGL 78-029Notice of PWR Steam Generator Conference NAInfoGL 78-030Forwarding of NUREG-0219 NAInfoGL 78-031Notice of Steam Generator Conference Agenda NAInfoPage 37 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 78-032Reactor Protection System Power Supplies NABoiling Water ReactorGL 78-033Meeting Schedule and Locations For Upgraded Guard Qualification NAInfoGL 78-034Reactor Vessel Atypical Weld Material C See B 78-12.GL 78-035Regional Meetings to Discuss Upgraded Guard Qualifications NAInfoGL 78-036Cessation of Plutonium Shipments by Air Except In NRC Approved Containers NADoes not apply to power reactor.GL 78-037Revised Meeting Schedule & Locations For Upgraded Guard Qualifications NAInfoGL 78-038Forwarding of 2 Tables of Appendix I, Draft Radiological Effluent Technical Specifications, PWR, and NUREG-0133 NAItem was applicable only to units with operating license at the time the item was issued.GL 78-039Forwarding of 2 Tables of Appendix I, Draft Radiological Effluent Technical Specifications, BWR, and NUREG-0133 NABoiling Water ReactorGL 78-040Training & Qualification Program Workshops NAInfoGL 78-041Mark II Generic Acceptance Criteria For Lead Plants NABoiling Water ReactorGL 78-042Training and Qualification Program Workshops NAInfoGL 79-001Interservice Procedures for Instructional Systems Development - TRADOC NAInfoGL 79-002Transmitting Rev. to Entry Control Systems Handbook (SAND 77-1033), Intrusion Detection Handbook (SAND 76-0554), and Barrier Penetration Database NAInfoGL 79-003Offsite Dose Calculation Manual NAInfoPage 38 of 101* = See last page for status code definition.

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

NAInfoGL 79-008Amendment to 10 CFR 73.55 NAInfoGL 79-009Staff Evaluation of Interim Multiple-Consecutive Safety-Relief Valve Actuations NABoiling Water ReactorGL 79-010Transmitting Regulatory Guide 2.6 for Comment NADoes not apply to power reactor.GL 79-011Transmitting "Summary of Operating Experience with Recalculating Steam Generators, January 1979," NUREG-0523 NAInfoGL 79-012ATWS - Enclosing Letter to GE, with NUREG-0460, Vol. 3 NAInfoGL 79-013Schedule for Implementation and Resolution of Mark I Containment Long Term Program NAInfoGL 79-014Pipe Crack Study Group - Enclosing NUREG-0531 and Notice NAInfoGL 79-015Steam Generators - Enclosing Summary of Operating Experience with Recirculating Steam Generators, NUREG-0523 NAInfoGL 79-016Meeting Re Implementation of Physical Security Requirements NAInfoGL 79-017Reliability of Onsite Diesel Generators at Light Water Reactors NAInfoPage 39 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 79-018Westinghouse Two-Loop NSSS NAAddressed to specific plant(s).GL 79-019NRC Staff Review of Responses to Bs 79-06 and 79-06a NAAddressed to specific plant(s).GL 79-020Cracking in Feedwater Lines C See B 79-13.GL 79-021Enclosing NUREG/CR-0660, Enhancement of on Site Emergency Diesel Generator Reliability" NAInfoGL 79-022Enclosing NUREG-0560, "Staff Report on the Generic Assessment of Feedwater Transients in PWRs Designed by B&W" NAApplies only to Babcock and Wilcox designed plantsGL 79-023NRC Staff Review of Responses to B 79-08 NABoiling Water ReactorGL 79-024Multiple Equipment Failures in Safety-Related Systems NAGL 79-24 provided a discussion of an inadvertent reactor scram and safety injection during monthly surveillance tests of the safeguards system at a PWR facility. The GL requested a review to determine if similar errors had or could have occurred at other PWRs. The GL further requested a review of management policies and procedures to assure that multiple equipment failures in safety-related systems will be vigorously pursued and analyzed to identify significant reduction in the ability of safety systems to function as required. A response was requested within 30 days of receipt of the GL with the results of these reviews. TVA does not have a record of receiving or responding to this GL. Thus, TVA concluded that this item was applicable only to PWRs with an operating license at the time the GL was

issued.01GL 79-025Information Required to Review Corporate Capabilities NAInfoGL 79-026Upgraded Standard Technical Specification Bases Program NAInfoGL 79-027Operability Testing of Relief and Safety Relief Valves NABoiling Water ReactorGL 79-028Evaluation of Semi-Scale Small Break Experiment NAInfoGL 79-029Transmitting NUREG-0473, Revision 2, Draft Radiological Effluent Technical Specifications NAInfoPage 40 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 79-030Transmitting NUREG-0472, Revision 2, Draft Radiological Technical Specifications NAInfoGL 79-031Submittal of Copies of Response to 6/29/79 NRC Request [79-25]

NAInfoGL 79-032Transmitting NUREG-0578, "TMI-2 Lessons Learned" NAInfoGL 79-033Transmitting NUREG-0576, "Security Training and Qualification Plans" NAInfoGL 79-034New Physical Security Plans (FR 43280-285)

NADoes not apply to power reactor.GL 79-035Regional Meetings to Discuss Impacts on Emergency Planning NAInfoGL 79-036Adequacy of Station Electric Distribution Systems Voltages CIThis GL tracked compliance with BTP PSB-1, "Adequacy of Station Electric Distribution System Voltages." Unit 2 Action:

Perform verification during the preoperational testing.GL 79-037Amendment to 10 CFR 73.55 Deferral from 8/1/79 to 11/1/79 NAInfoGL 79-038BWR Off-Gas Systems - Enclosing NUREG/CR-0727 NABoiling Water ReactorGL 79-039Transmitting Division 5 Draft Regulatory Guide and Value Impact Statement NADoes not apply to power reactor.GL 79-040Follow-up Actions Resulting from the NRC Staff Reviews Regarding the TMI-2 Accident NAItem was applicable only to units with operating license at the time the item was issued.GL 79-041Compliance with 40 CFR 190, EPA Uranium Fuel Cycle Standard NAInfoGL 79-042Potentially Unreviewed Safety Question on Interaction Between Non-Safety Grade Systems and Safety Grade Systems NAItem was applicable only to units with operating license at the time the item was issued.GL 79-043Reactor Cavity Seal Ring Generic Issue NAAddressed to specific plant(s).Page 41 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 79-044Referencing 6/29/79 Letter Re Multiple Equipment Failures NAItem was applicable only to units with operating license at the time the item was issued.GL 79-045Transmittal of Reports Regarding Foreign Reactor Operating Experiences NAInfoGL 79-046Containment Purge and Venting During Normal Operation - Guidelines for Valve Operability NAItem was applicable only to units with operating license at the time the item was issued.GL 79-047Radiation Training NAInfoGL 79-048Confirmatory Requirements Relating to Condensation Oscillation Loads for the Mark I Containment Long Term Program NABoiling Water ReactorGL 79-049Summary of Meetings Held on 9/18-20/79 to Discuss Potential Unreviewed Safety Question on Systems Interaction for B&W

Pl NAInfoGL 79-050Emergency Plans Submittal Dates NAInfoGL 79-051Follow-up Actions Resulting from the NRC Staff Reviews Regarding the TMI-2 Accident NAGL 79-51 provided follow-up actions resulting from the Three Mile Island Unit 2 accident. GL 79-51 was provided for planning and guidance purposes. Its principal element was a report titled "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations" (NUREG-0573). This GL and the NUREG were superseded by GL 80-90 and NUREG-0737. See GL 80-90 for further information.

01GL 79-052Radioactive Release at North Anna Unit 1 and Lessons Learned NAItem was applicable only to units with operating license at the time the item was issued.GL 79-053ATWS NAInfoGL 79-054Containment Purging and Venting During Normal Operation NAAddressed to specific plant(s).GL 79-055Summary of Meeting Held on October 12, 1979 to Discuss Responses to Bulletins79-05C and 79-06C and HPI Termination Criteria NAInfoGL 79-056Discussion of Lessons Learned Short Term Requirements NAItem was applicable only to units with operating license at the time the item was issued.Page 42 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 79-057Acceptance Criteria for Mark I Long Term Program NABoiling Water ReactorGL 79-058ECCS Calculations on Fuel Cladding NAItem was applicable only to units with operating license at the time the item was issued.GL 79-059This GL was never issued.

NAGL 79-060Discussion of Lessons Learned Short Term Requirements NAInfoGL 79-061Discussion of Lessons Learned Short Term Requirements NAInfoGL 79-062ECCS Calculations on Fuel Cladding NAItem was applicable only to units with operating license at the time the item was issued.Duplicate of GL 79-058GL 79-063Upgraded Emergency Plans CGL 79-63 advised applicants for licenses of proposed rulemaking that NRC concurrence in State and local emergency plans would be a condition for issuing an operating license. TVA responded to GL 79-63 on January 3, 1980, and confirmed the intent to revise the Emergency Plan to address the NRC requirements.

01GL 79-064Suspension of All Operating Licenses (PWRs)NAInfoGL 79-065Radiological Environmental Monitoring Program Requirements - Enclosing Branch Technical Position, Revision 1 NAInfoGL 79-066Additional Information Re 11/09/79 Letter on ECCS Calculations [GL 79-62]

NAInfoGL 79-067Estimates for Evacuation of Various Areas Around Nuclear Power Reactors NAInfoGL 79-068Audit of Small Break LOCA Guidelines NAInfoGL 79-069Cladding Rupture, Swelling, and Coolant Blockage as a Result of a Reactor Accident NAInfoGL 79-070Environmental Monitoring for Direct Radiation NAInfoPage 43 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 80-001NUREG-0630, "Cladding, Swelling and Rupture - Models For LOCA Analysis" NAInfoGL 80-002QA Requirements Regarding Diesel Generator Fuel Oil CTVA: FSAR 9.5.4.2GL 80-003BWR Control Rod Failures NABoiling Water ReactorGL 80-004B 80-01, "Operability of ADS Valve Pneumatic Supply" NABoiling Water ReactorGL 80-005B 79-01b, "Environmental Qualification of Class 1E Equipment" NAInfoGL 80-006Issuance of NUREG-0313, Rev 1, "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping" NABoiling Water ReactorGL 80-007This GL was never issued.

NAGL 80-008B 80-02. "Inadequate Quality Assurance for Nuclear Supplied Equipment" NABoiling Water ReactorGL 80-009Low Level Radioactive Waste Disposal NAItem was applicable only to units with operating license at the time the item was issued.GL 80-010Issuance of NUREG-0588, "Interim Staff Position On Equipment Qualifications of Safety-Related Electrical Equipment" NAInfoGL 80-011B 80-03, "Loss of Charcoal From Standard Type II, 2 Inch, Tray Absorber Cells"CGL 80-11 transmitted Bulletin 80-03. TVA responded to B 80-03 on March 21, 1980. See B 80-03 for further information.

01GL 80-012B 80-04, "Analysis of a PWR Main Steam Line Break With Continued Feedwater Addition"NAInfoGL 80-013Qualification of Safety Related Electrical Equipment NAItem was applicable only to units with operating license at the time the item was issued.Page 44 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 80-014LWR Primary Coolant System Pressure Isolation Valves STVA: FSAR 5.2.7.4NRC: 1.14.2 of SSER 6

NRC reviewed in 1.14.2 of SSER6. Unit 2 Action: Incorporate guidance into Technical Specifications.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.TS Surveillance Requirement 3.4.13.1 verifies RCS operational leakage by performance of an RCS water inventory balance.

02GL 80-015Request for Additional Management and Technical Resources Information NAInfoGL 80-016B 79-01b, "Environmental Qualification of Class 1E Equipment" NAInfoGL 80-017Modifications to BWR Control Rod Drive Systems NABoiling Water ReactorGL 80-018Crystal River 3 Reactor Trip From Approximately 100% Full Power NAApplies only to Babcock and Wilcox designed plantsGL 80-019Resolution of Enhanced Fission Gas Release Concern NAInfoGL 80-020Actions Required From OL Applicants of NSSS Designs by W and CE Resulting From NRC B&O Task Force Review of TMI2 Accident NAInfoGL 80-021B 80-05, "Vacuum Condition Resulting in Damage to Chemical Volume Control System Holdup Tanks" CIClosed in IR 50-390/84-59 and 50-391/84-45. Unit 2 Action:

Complete surveillance procedures for Unit 2.GL 80-022Transmittal of NUREG-0654, "Criteria For Preparation and Evaluation of Radiological Emergency Response Plan" NAInfoGL 80-023Change of Submittal Date For Evaluation Time Estimates NAInfoPage 45 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 80-024Transmittal of Information on NRC "Nuclear Data Link Specifications" NAInfoGL 80-025B 80-06, "Engineering Safety Feature (ESF) Reset Controls" NAInfoGL 80-026Qualifications of Reactor Operators NAInfoGL 80-027B 80-07, "BWR Jet Pump Assembly Failure"NABoiling Water ReactorGL 80-028B 80-08, "Examination of Containment Liner Penetration Welds" CGL 80-28 transmitted Bulletin 80-08. TVA responded to B 80-08 on July 8, 1980. See B 80-08 for further information.

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

01GL 80-033Actions Required From OL Applicants of B&W Designed NSSS Resulting From NRC B&O Task Force Review of TMI2 Accident NAApplies only to Babcock and Wilcox designed plantsGL 80-034Clarification of NRC Requirements for Emergency Response Facilities at Each Site NAInfoGL 80-035Effect of a DC Power Supply Failure on ECCS Performances NABoiling Water ReactorGL 80-036B 80-10, "Contamination of Non-Radioactive System and Resulting Potential For Unmonitored, Uncontrolled Release to Environment" NAInfoPage 46 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 80-037Five Additional TMI-2 Related Requirements to Operating Reactors NAItem was applicable only to units with operating license at the time the item was issued.GL 80-038Summary of Certain Non-Power Reactor Physical Protection Requirements NADoes not apply to power reactor.GL 80-039B 80-11, "Masonry Wall Design" NAInfoGL 80-040Transmittal of NUREG-0654, "Report of the B&O Task Force" and Appropriate NUREG-0626, "Generic Evaluation of FW Transient and Small Break LOCA" NAInfoGL 80-041Summary of Meetings Held on April 22 &23, 1980 With Representatives of the Mark I Owners Group NAInfoGL 80-042B 80-12, "Decay Heat Removal System Operability" NAInfoGL 80-043B 80-13, "Cracking In Core Spray Spargers"NABoiling Water ReactorGL 80-044Reorganization of Functions and Assignments Within ONRR/SSPB NAInfoGL 80-045Fire Protection Rule NAItem was applicable only to units with operating license at the time the item was issued.GL 80-046 andGL 80-047Generic Technical Activity A-12, "Fracture Toughness and Additional Guidance on Potential for Low Fracture toughness and Laminar Tearing on PWR Steam Generator Coolant Pump Supports" CNo response was required for this GL, and NUREG-0577 states that the lamellar tearing aspect of this issue was resolved by the NUREG. Further, the NUREG states that for plants under review, the fracture toughness issue was resolved.GL 80-048Revision to 5/19/80 Letter On Fire Protection [GL 80-45]

NAItem was applicable only to units with operating license at the time the item was issued.GL 80-049Nuclear Safeguards Problems NAInfoGL 80-050Generic Activity A-10, "BWR Cracks" NABoiling Water ReactorGL 80-051On-Site Storage of Low-Level Waste NAItem was applicable only to units with operating license at the time the item was issued.Page 47 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 80-052Five Additional TMI-2 Related Requirements - Erata Sheets to 5/7/80 Letter [GL 80-37]

NAItem was applicable only to units with operating license at the time the item was issued.GL 80-053Decay Heat Removal Capability NAItem was applicable only to units with operating license at the time the item was issued.GL 80-054B 80-14, "Degradation of Scram Discharge Volume Capability" NABoiling Water ReactorGL 80-055B 80-15, "Possible Loss of Hotline With Loss of off-Site Power" NAInfoGL 80-056Commission Memorandum and Order on Equipment Qualification NAInfoGL 80-057Further Commission Guidance For Power Reactor Operating Licenses NUREG-0660 and NUREG-0694 NAInfoGL 80-058B 80-16, "Potential Misapplication of Rosemount Inc. Models 1151/1152 Pressure Transmitters With "A" Or "D" Output Codes" NAInfoGL 80-059Transmittal of Federal Register Notice RE Regional Meetings to Discuss Environmental Qualification of Electrical Equipment NAInfoGL 80-060Request for Information Regarding Evacuation Times NAInfoGL 80-061TMI-2 Lessons Learned NAInfoGL 80-062TMI-2 Lessons Learned NABoiling Water ReactorGL 80-063B 80-17, "Failure of Control Rods to Insert During a Scram at a BWR" NABoiling Water ReactorGL 80-064Scram Discharge Volume Designs NABoiling Water ReactorGL 80-065Request for Estimated Construction Completion and Fuel Load Schedules NAInfoPage 48 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 80-066B 80-17, Supplement 1, "Failure of Control Rods to Insert During a Scram at a BWR"NABoiling Water ReactorGL 80-067Scram Discharge Volume NABoiling Water ReactorGL 80-068B 80-17, Supplement 2, "Failures Revealed by Testing Subsequent to Failure of Control Rods to Insert During a Scram at a BWR" NABoiling Water ReactorGL 80-069B 80-18, "Maintenance of Adequate Minimum Flow Through Centrifugal Charging Pumps Following Secondary Side HELB" NAInfoGL 80-070B 80-19, "Failures of Mercury-Wetted Matrix Relays in RPS of Operating Nuclear Power Plants Designed by GE" NAInfoGL 80-071B 80-20, "Failures of Westinghouse Type W-2 Spring Return to Neutral Control Switches"NAInfoGL 80-072Interim Criteria For Shift Staffing NAInfoGL 80-073"Functional Criteria For Emergency Response Facilities," NUREG-0696 NAInfoGL 80-074Notice of Forthcoming Meeting With Representatives of EPRI to Discuss Program For Resolution of USI A-12, "Fracture Toughness Issue" NAInfoGL 80-075Lessons Learned Tech. Specs.

NAItem was applicable only to units with operating license at the time the item was issued.GL 80-076Notice of Forthcoming Meeting With GE to Discussed Proposed BWR Feedwater Nozzle Leakage Detection System NAInfoPage 49 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 80-077Refueling Water Level - Technical Specifications Changes SUnit 2 Action: Address in Technical Specifications, as appropriate.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

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

02GL 80-078Mark I Containment Long-Term Program NABoiling Water ReactorGL 80-079B 80-17, Supplement 3, "Failures Revealed by Testing Subsequent to Failure of Control Rods to Insert During a Scram At a BWR" NABoiling Water ReactorGL 80-080Preliminary Clarification of TMI Action Plan Requirements NAInfoGL 80-081Preliminary Clarification of TMI Action Plan Requirements - Addendum to 9/5/80 Letter [GL 80-80]

NAInfoGL 80-082B 79-01b, Supplement 2, "Environmental Qualification of Class 1E Equipment" NAInfoGL 80-083Environmental Qualification of Safety-Related Equipment NAInfoGL 80-084BWR Scram System NABoiling Water ReactorGL 80-085Implementation of Guidance From USI A-12, "Potential For LOW Fracture Toughness and Lamellar Tearing On Component Support" NAInfoGL 80-086Notice of Meeting to Discuss Final Resolution of USI A-12 NAInfoGL 80-087Notice of Meeting to Discuss Status of EPRI-Proposed Resolution of the USI A-12 Fracture Toughness Issue NAInfoPage 50 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 80-088Seismic Qualification of Auxiliary Feedwater Systems NAItem was applicable only to units with operating license at the time the item was issued.GL 80-089B 79-01b, Supplement 3, "Environmental Qualification of Class 1E Equipment" NAInfoGL 80-090NUREG-0737, TMI (Prior and future GLs, with the exception of certain discrete scopes, have been screened into NUREG list for those applicable to Watts Bar 2)

CISee NUREG items in this list.GL 80-091ODYN Code Calculation NABoiling Water ReactorGL 80-092B 80-21, "Valve Yokes Supplied by Malcolm Foundry Company, Inc."

CGL 80-92 transmitted Bulletin 80-21. TVA responded to B 80-21 on May 6, 1981. See B 80-21 for further information.

01GL 80-093Emergency Preparedness NADoes not apply to power reactor.GL 80-094Emergency Plan NAInfoGL 80-095Generic Technical Activity A-10, NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking" NABoiling Water ReactorGL 80-096Fire Protection NAAddressed to specific plant(s).GL 80-097B 80-23, "Failures of Solenoid Valves Manufactured by Valcor Engineering Corporation" NAInfoGL 80-098B 80-24, "Prevention of Damage Due to Water Leakage Inside Containment" NAInfoGL 80-099Technical Specifications Revisions For Snubber Surveillance NAInfoGL 80-100Appendix R to 10 CFR 50 Regarding Fire Protection - Federal Register Notice NAItem was applicable only to units with operating license at the time the item was issued.GL 80-101Inservice Inspection Programs NAAddressed to specific plant(s).Page 51 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 80-102Commission Memorandum and Order of May 23, 1980 (Referencing B 79-01b, Supplement 2 - q.2 & 3 - Sept 30, 1980)

NAInfoGL 80-103Fire Protection - Revised Federal Register Notice NAInfoGL 80-104Orders On Environmental Qualification of Safety Related Electrical Equipment NAInfoGL 80-105Implementation of Guidance For USI A-12, "Potential For Low Fracture toughness and Lamellar Tearing On Component Supports" NAInfoGL 80-106Report On ECCS Cladding Models, NUREG-0630 NAInfoGL 80-107BWR Scram Discharge System NABoiling Water ReactorGL 80-108Emergency Planning NAInfoGL 80-109Guidelines For SEP Soil Structure Interaction Reviews NAInfoGL 80-110Periodic Updating of FSARS NAItem was applicable only to units with operating license at the time the item was issued.GL 80-111B 80-17, Supplement 4, "Failure of Control Rods to Insert During a Scram at a BWR"NABoiling Water ReactorGL 80-112B 80-25, "Operating Problems With Target Rock Safety Relief Valves" NAInfoGL 80-113Control of Heavy Loads CSuperseded by GL 81-007.GL 81-001Qualification of Inspection, Examination, Testing and Audit Personnel NAInfoGL 81-002Analysis, Conclusions and Recommendations Concerning Operator Licensing NAInfoPage 52 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 81-003Implementation of NUREG-0313, "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping" NABoiling Water ReactorGL 81-004Emergency Procedures and Training for Station Blackout Events CSuperseded by Station Blackout Rule.GL 81-005Information Regarding The Program For Environmental Qualification of Safety-Related Electrical Equipment NAInfoGL 81-006Periodic Updating of Final Safety Analysis Reports (FSARS)

NAInfoGL 81-007Control of Heavy Loads CI"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 81-008ODYN Code NABoiling Water ReactorGL 81-009BWR Scram Discharge System NABoiling Water ReactorGL 81-010Post-TMI Requirements For The Emergency Operations Facility NAInfoGL 81-011BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking (NUREG-0619)

NABoiling Water ReactorGL 81-012Fire Protection Rule NAItem was applicable only to units with operating license at the time the item was issued.GL 81-013SER For GEXL Correlation For 8X8R Fuel Reload Applications For Appendix D Submittals of The GE topical Report NABoiling Water ReactorPage 53 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 81-014Seismic Qualification of Auxiliary Feedwater Systems CITVA: FSAR 10.4.9

Unit 2 Action: Additional Unit 2 implementing procedures or other activity is required for completion.[WAS "OL."]GL 81-015Environmental Qualification of Class 1E Electrical Equipment - Clarification of Staff's Handling of Proprietary Information NAInfoGL 81-016NUREG-0737, Item I.C.1 SER on Abnormal Transient Operating Guidelines (ATOG)NAApplies only to Babcock and Wilcox designed plantsGL 81-017Functional Criteria for Emergency Response Facilities NAInfoGL 81-018BWR Scram Discharge System - Clarification of Diverse Instrumentation Requirements NABoiling Water ReactorGL 81-019Thermal Shock to Reactor Pressure Vessels NAItem was applicable only to units with operating license at the time the item was issued.GL 81-020Safety Concerns Associated With Pipe Breaks in the BWR Scram System NABoiling Water ReactorGL 81-021Natural Circulation Cooldown CITVA responded December 3, 1981. Unit 2 Action: Issue operating procedures.GL 81-022Engineering Evaluation of the H. B. Robinson Reactor Coolant System Leak on 1/29/81 NAInfoGL 81-023INPO Plant Specific Evaluation Reports NAInfoGL 81-024Multi-Plant Issue B-56, "Control Rods Fail to Fully Insert" NABoiling Water ReactorGL 81-025Change in Implementing Schedule For Submission and Evaluation of Upgraded Emergency Plans NAInfoPage 54 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 81-026Licensing Requirements for Pending Construction Permit and Manufacturing License Applications NAApplicants with pending Construction PermitsGL 81-027Privacy and Proprietary Material in Emergency Plans NAInfoGL 81-028Steam Generator Overfill NAInfoGL 81-029Simulator Examinations NAInfoGL 81-030Safety Concerns Associated With Pipe Breaks in the BWR Scram System NABoiling Water ReactorGL 81-031This GL was never issued.

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

NAGL 81-034Safety Concerns Associated With Pipe Breaks in the BWR Scram System NABoiling Water ReactorGL 81-035Safety Concerns Associated With Pipe Breaks in the BWR Scram System NABoiling Water ReactorGL 81-036Revised Schedule for Completion of TMI Action Plan Item II.D.1, "Relief and Safety Valve Testing" NAInfoGL 81-037ODYN Code Reanalysis Requirements NABoiling Water ReactorGL 81-038Storage of Low Level Radioactive Wastes at Power Reactor Sites NAInfoGL 81-039NRC Volume Reduction Policy NAInfoGL 81-040Qualifications of Reactor Operators NAInfoPage 55 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 82-001New Applications Survey NAInfoGL 82-002Commission Policy on Overtime NAInfoGL 82-003High Burnup MAPLHGR Limits NABoiling Water ReactorGL 82-004Use of INPO See-in Program NAInfoGL 82-005Post-TMI Requirements NAItem was applicable only to units with operating license at the time the item was issued.GL 82-006This GL was never issued.

NAGL 82-007Transmittal of NUREG-0909 Relative to the Ginna Tube Rupture NABoiling Water ReactorGL 82-008Transmittal of NUREG-0909 Relative to the Ginna Tube Rupture NAInfoGL 82-009Environmental Qualification of Safety Related Electrical Equipment NAInfoGL 82-010Post-TMI Requirements NAItem was applicable only to units with operating license at the time the item was issued.GL 82-011Transmittal of NUREG-0916 Relative to the Restart of R. E. Ginna Nuclear Power

Plant NAInfoGL 82-012Nuclear Power Plant Staff Working Hours NAInfoGL 82-013Reactor Operator and Senior Reactor Operator Examinations NAInfoGL 82-014Submittal of Documents to the NRC NAInfoGL 82-015This GL was never issued.

NAGL 82-016NUREG-0737 Technical Specifications NAItem was applicable only to units with operating license at the time the item was issued.Page 56 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 82-017Inconsistency of Requirements Between 50.54(T) and 50.15 NAInfoGL 82-018Reactor Operator and Senior Reactor Operator Requalification Examinations NAInfoGL 82-019Submittal of Copies of Documentation to NRC - Copy Requirements for Emergency Plans and Physical Security Plans NAInfoGL 82-020Guidance for Implementing the Standard Review Plan Rule NAInfoGL 82-021Fire Protection Audits NAInfoGL 82-022Congressional Request for Information Concerning Steam Generator Tube Integrity NAItem was applicable only to units with operating license at the time the item was issued.GL 82-023Inconsistency Between Requirements of 10CFR 73.40(d) and Standard Technical Specifications For Performing Audits of Safeguards Contingency Plans NAInfoGL 82-024Safety Relief Valve Quencher Loads: BWR MARK II and III Containments NABoiling Water ReactorGL 82-025Integrated IAEA Exercise for Physical Inventory at LWRS NAItem was applicable only to units with operating license at the time the item was issued.GL 82-026NUREG-0744, REV. 1, "Pressure Vessel Material Fracture Toughness" NAItem was applicable only to units with operating license at the time the item was issued.GL 82-027Transmittal of NUREG-0763, "Guidelines For Confirmatory In-Plant Tests of Safety-Relief Valve Discharge for BWR Plants"NABoiling Water ReactorPage 57 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 82-028Inadequate Core Cooling Instrumentation System OLICENSE CONDITION - Detectors for Inadequate core cooling (II.F.2)In the original SER, the review of the ICC instrumentation was incomplete. The January 24, 1992, letter superseded the previous responses on this issue. TVA letter for Units 1 and 2 dated January 24, 1992, committed to install Westinghouse ICCM-86 and associated hardware. NRC completed the review for Units 1 and 2 in SSER10. For Unit 2 due to obsolescence of the ICCM-86 system, TVA intends to install the Westinghouse Common Q Post-Accident Monitoring System. Unit 2 Action:

Install Westinghouse Common Q PAM system.GL 82-029This GL was never issued.

NAGL 82-030Filings Related to 10 CFR 50 Production and Utilization Facilities NAInfoGL 82-031This GL was never issued.

NAGL 82-032Draft Steam Generator Report (SAI)

NAItem was applicable only to units with operating license at the time the item was issued.GL 82-033Supplement to NUREG-0737, "Requirements for Emergency Response Capability" CI"Safety Parameter Display System" (SPDS) / "Requirements for Emergency Response Capability" - NRC reviewed in SSER5, SSER6, and 18.2.2 of SSER15. Unit 2 Action:

Install SPDS and have it operational prior to start-up after the first refueling outage.GL 82-034This GL was never issued.

NAGL 82-035This GL was never issued.

NAGL 82-036This GL was never issued.

NAGL 82-037This GL was never issued.

NAGL 82-038Meeting to Discuss Developments for Operator Licensing Examinations NAInfoPage 58 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 82-039Problems With Submittals of Subsequent Information of CURT 73.21 For Licensing Reviews NAInfoGL 83-001Operator Licensing Examination Site Visit NAInfoGL 83-002NUREG-0737 Technical Specifications NABoiling Water ReactorGL 83-003This GL was never issued.

NAGL 83-004Regional Workshops Regarding Supplement 1 to NUREG-0737, "Requirements For Emergency Response Capability" NAInfoGL 83-005Safety Evaluation of "Emergency Procedure Guidelines, Revision 2," June 1982 NABoiling Water ReactorGL 83-006Certificates and Revised Format For Reactor Operator and Senior Reactor Operator Licenses NAInfoGL 83-007The Nuclear Waste Policy Act of 1982 NAInfoGL 83-008Modification of Vacuum Breakers on Mark I Containments NABoiling Water ReactorGL 83-009Review of Combustion Engineering Owners' Group Emergency Procedures Guideline Program NAApplies only to Combustion Engineering designed plantsGL 83-010aResolution of TMI Action Item II.K.3.5., "Automatic Trip of Reactor Coolant Pumps" NAApplies only to Combustion Engineering designed plantsGL 83-010bResolution of TMI Action Item II.K.3.5., "Automatic Trip of Reactor Coolant Pumps" NAApplies only to Combustion Engineering designed plantsPage 59 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 83-010cResolution of TMI Action Item II.K.3.5., "Automatic Trip of Reactor Coolant Pumps" CITVA: letters dated January 5, 1984 and June 25, 1984NRC: letter dated June 8, 1990.

Unit 2 Action: Incorporate emergency response guidelines into applicable procedures.[WAS "NOTE 3."]GL 83-010dResolution of TMI Action Item II.K.3.5., "Automatic Trip of Reactor Coolant Pumps" NAItem was applicable only to units with operating license at the time the item was issued.GL 83-010eResolution of TMI Action Item II.K.3.5., "Automatic Trip of Reactor Coolant Pumps" NAApplies only to Babcock and Wilcox designed plantsGL 83-010fResolution of TMI Action Item II.K.3.5., "Automatic Trip of Reactor Coolant Pumps" NAApplies only to Babcock and Wilcox designed plantsGL 83-011Licensee Qualification for Performing Safety Analyses in Support of Licensing Actions NAItem was applicable only to units with operating license at the time the item was issued.GL 83-012Issuance of NRC FORM 398 - Personal Qualifications Statement - Licensee NAInfoGL 83-013Clarification of Surveillance Requirements for HEPA Filters and Charcoal Absorber Units In Standard Technical Specifications on ESF Cleanup Systems NAInfoGL 83-014Definition of "Key Maintenance Personnel," (Clarification of Generic Letter 82-12)

NAInfoGL 83-015Implementation of Regulatory Guide 1.150, "Ultrasonic Testing of Reactor Vessel Welds During Preservice & Inservice Examinations, Revision 1" NAInfoGL 83-016Transmittal of NUREG-0977 Relative to the ATWS Events at Salem Generating Station, Unit No.1 NAInfoGL 83-016aTransmittal of NUREG-0977 Relative to the ATWS Events at Salem Generating Station, Unit No.1 NAInfoPage 60 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 83-017Integrity of Requalification Examinations for Renewal of Reactor Operator and Senior Reactor Operator Licenses NAInfoGL 83-018NRC Staff Review of the BWR Owners' Group (BWROG) Control Room Survey Program NABoiling Water ReactorGL 83-019New Procedures for Providing Public Notice Concerning Issuance of Amendments to Operating Licenses NAItem was applicable only to units with operating license at the time the item was issued.GL 83-020Integrated Scheduling for Implementation of Plant Modifications NAInfoGL 83-021Clarification of Access Control Procedures for Law Enforcement Visits NAInfoGL 83-022Safety Evaluation of "Emergency Response Guidelines" NAInfoGL 83-023Safety Evaluation of "Emergency Procedure Guidelines" NAApplies only to Combustion Engineering designed plantsGL 83-024TMI Task Action Plan Item I.G.1, "Special Low Power Testing and Training," Recommendations for BWRs NABoiling Water ReactorGL 83-025This GL was never issued.

NAGL 83-026Clarification Of Surveillance Requirements For Diesel Fuel Impurity Level Tests NAInfoGL 83-027Surveillance Intervals in Standard Technical Specifications NAInfoGL 83-028"Required Actions Based on Generic Implications of Salem ATWS Events: 1.2 - Post Trip Review Data and Information Capability CTVA: letters dated November 7, 1983 and December 4, 1987NRC: IR 50-390, 391/86-04Page 61 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 83-028"Required Actions Based on Generic Implications of Salem ATWS Events: 2.1 - Equipment Classification and Vendor Interface (Reactor Trip System Components)

CITVA: letters dated November 7, 1983 and August 24, 1990NRC: letters dated October 20, 1986 and June 18, 1990

Unit 2 Action: Ensure that required information on Critical Structures and Components is properly incorporated into procedures.[WAS "NOTE 3."]GL 83-028"Required Actions Based on Generic Implications of Salem ATWS Events: 2.2 - Equipment Classification and Vendor Interface (All SR Components)" CIUnit 2 Action: Enter engineering component background data in INPO's Equipment Performance and Information Exchange System (EPIX) for Unit 2.GL 83-028"Required Actions Based on Generic Implications of Salem ATWS Events: 3.1 - Post-Maintenance Testing (Reactor Trip System Components)

STVA: letters dated November 7, 1983, January 17, 1986 and November 1, 1993NRC: letters dated December 10, 1985, October 27, 1986, and July 2, 1990; IR 390, 391/86-04

Unit 2 Action: Test and maintenance procedures and Technical Specifications will include post-maintenance operability testing of safety-related components of the reactor trip system.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:

Developmental Revision A of the Unit 2 TS (including the TS Bases) was submitted on March 4, 2009.The Bases for TS Surveillance Requirement 3.0.1 states, in part, "Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2."

02Page 62 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 83-028"Required Actions Based on Generic Implications of Salem ATWS Events: 3.2 - Post-Maintenance Testing (All SR Components)

STVA: letters dated November 7, 1983, January 17, 1986 and November 1, 1993NRC: letters dated December 10, 1985, October 27, 1986, and July 2, 1990; IR 390, 391/86-04

Unit 2 Action: Test and maintenance procedures and Technical Specifications will include post-maintenance operability testing of other (than reactor trip system) safety-related components.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:

Developmental Revision A of the Unit 2 TS (including the TS Bases) was submitted on March 4, 2009.The Bases for TS Surveillance Requirement 3.0.1 states, in part, "Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2."

02GL 83-028"Required Actions Based on Generic Implications of Salem ATWS Events: 4.1 - Reactor Trip System Reliability (Vendor Related Modifications)

CITVA: letter dated May 19, 1986

Unit 2 Action:

Confirm vendor-recommended DS416 breaker modifications are implemented.GL 83-028"Required Actions Based on Generic Implications of Salem ATWS Events: 4.2 - Reactor Trip System Reliability (Preventive Maintenance and Surveillance Program for Reactor Trip Breakers)

STVA: letters dated November 7, 1983, February 10, 1986, and May 19, 1986NRC: letters dated July 26, 1985 and June 18, 1992; SSER 16

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

REVISION 02 UPDATE:

Developmental Revision B of the Unit 2 TS was submitted on February 2, 2010. Item 17. (Reactor Trip Breakers) of TS Table 3.3.1-1 states the requirement for the reactor trip breakers.

02Page 63 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 83-028"Required Actions Based on Generic Implications of Salem ATWS Events: 4.3 - Reactor Trip System Reliability (Automatic Actuation of Shunt Trip Attachment)

CTVA: letters dated November 7, 1983, March 22, 1985NRC: IR 50-390/86-04 and 50-391/86-04; letter dated June 18, 1990GL 83-028"Required Actions Based on Generic Implications of Salem ATWS Events: 4.5 - Reactor Trip System Reliability (Automatic Actuation of Shunt Trip Attachment)

STVA: letters dated November 7, 1983 and July 26, 1985NRC: letters dated June 28, 1990 and October 9, 1990; SSERs 5 and 16

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

REVISION 02 UPDATE:Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.Item 18. (Reactor Trip Breaker Undervoltage and Shunt Trip Mechanisms) of TS Table 3.3.1-1 states the requirement for the shunt trip attachment.

02GL 83-029This GL was never issued.

NAGL 83-030Deletion of Standard Technical Specifications Surveillance Requirement 4.8.1.1.2.d.6 For Diesel Generator Testing NAInfoGL 83-031Safety Evaluation of "Abnormal Transient Operating Guidelines" NAApplies only to Babcock and Wilcox designed plantsGL 83-032NRC Staff Recommendations Regarding Operator Action for Reactor Trip and ATWS NAInfoGL 83-033NRC Positions on Certain Requirements of Appendix R to 10 CFR 50 NAInfoGL 83-034This GL was never issued.

NAGL 83-035Clarification of TMI Action Plan Item II.K.3.31 NAInfoPage 64 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 83-036NUREG-0737 Technical Specifications NABoiling Water ReactorGL 83-037NUREG-0737 Technical Specifications NAItem was applicable only to units with operating license at the time the item was issued.GL 83-038NUREG-0965, "NRC Inventory of Dams" NAInfoGL 83-039Voluntary Survey of Licensed Operators NAInfoGL 83-040Operator Licensing Examination NAInfoGL 83-041Fast Cold Starts of Diesel Generators NAItem was applicable only to units with operating license at the time the item was issued.GL 83-042Clarification to GL 81-07 Regarding Response to NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants" NAInfoGL 83-043Reporting Requirements of 10 CFR 50, Sections 50.72 and 50.73, and Standard Technical Specifications NAInfoGL 83-044Availability of NUREG-1021, "Operator Licensing Examiner Standards" NAInfoGL 84-001NRC Use Of The Terms "Important To Safety" and "Safety Related" NAInfoGL 84-002Notice of Meeting Regarding Facility Staffing NAInfoGL 84-003Availability of NUREG-0933, "A Prioritization of Generic Safety Issues" NAInfoGL 84-004Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops NAInfoGL 84-005Change to NUREG-1021, "Operator Licensing Examiner Standards" NAInfoGL 84-006Operator and Senior Operator License Examination Criteria For Passing Grade NADoes not apply to power reactor.Page 65 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 84-007Procedural Guidance for Pipe Replacement at BWRs NABoiling Water ReactorGL 84-008Interim Procedures for NRC Management of Plant-Specific Backfitting NAInfoGL 84-009Recombiner Capability Requirements of 10 CFR 50.44(c)(3)(ii)

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

NAInfoGL 84-013Technical Specification for Snubbers NAInfoGL 84-014Replacement and Requalification Training Program NAInfoGL 84-015Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability NAInfoGL 84-016Adequacy of On-Shift Operating Experience for Near Term Operating License Applicants NAInfoGL 84-017Annual Meeting to Discuss Recent Developments Regarding Operator Training, Qualifications, and Examinations NAInfoGL 84-018Filing of Applications for Licenses and Amendments NADoes not apply to power reactor.GL 84-019Availability of Supplement 1 to NUREG-0933, "A Prioritization of Generic Safety Issues" NAInfoGL 84-020Scheduling Guidance for Licensee Submittals of Reloads That Involve Unreviewed Safety Questions NAInfoPage 66 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 84-021Long Term Low Power Operation in Pressurized Water Reactors NAInfoGL 84-022This GL was never issued.

NAGL 84-023Reactor Vessel Water Level Instrumentation in BWRs NABoiling Water ReactorGL 84-024Certification of Compliance to 10 CFR 50.49, Environmental Qualification of Electric Equipment Important To Safety For Nuclear Power

Plants CISee Special Program for Environmental Qualification.GL 85-001Fire Protection Policy Steering Committee Report NAOnly issued as draftGL 85-002Recommended Actions Stemming From NRC Integrated Program for the Resolution of Unresolved Safety Issues Regarding Steam Generator Tube Integrity CITVA responded to the GL on June 17, 1985. Unit 2 Action:

Perform SG inspection.GL 85-003Clarification of Equivalent Control Capacity for Standby Liquid Control Systems NABoiling Water ReactorGL 85-004Operating Licensing Examinations NAInfoGL 85-005Inadvertent Boron Dilution Events NAItem was applicable only to units with operating license at the time the item was issued.GL 85-006Quality Assurance Guidance for ATWS Equipment That Is Not Safety-Related NAInfoGL 85-007Implementation of Integrated Schedules for Plant Modifications NAItem was applicable only to units with operating license at the time the item was issued.GL 85-00810 CFR 20.408 Termination Reports - Format NAInfoGL 85-009Technical Specifications For Generic Letter 83-28, Item 4.3 NAInfoGL 85-010Technical Specification For Generic Letter 83-28, Items 4.3 and 4.4 NAApplies only to Babcock and Wilcox designed plantsPage 67 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 85-011Completion of Phase II of "Control of Heavy Loads at Nuclear Power Plants,"NUREG-0612 CSee GL 81-07.GL 85-012Implementation Of TMI Action Item II.K.3.5, "Automatic Trip Of Reactor Coolant Pumps" CI"Implementation of TMI Item II.K.3.5" - Reviewed in 15.5.4 of original 1982 SER; became License Condition 35. The staff determined that their review of Item II.K.3.5 did not have to be completed to support the full power license and considered this license condition resolved in SSER4. The item was further reviewed in Appendix EE of SSER16. Unit 2 Action:

Implement modifications as required.GL 85-013Transmittal Of NUREG-1154 Regarding The Davis-Besse Loss Of Main And Auxiliary Feedwater Event NAInfoGL 85-014Commercial Storage At Power Reactor Sites Of Low Level Radioactive Waste Not Generated By The Utility NAItem was applicable only to units with operating license at the time the item was issued.GL 85-015Information On Deadlines For 10 CFR 50.49, "Environmental Qualification Of Electric Equipment Important To Safety At Nuclear Power Plants" NAItem was applicable only to units with operating license at the time the item was issued.GL 85-016High Boron Concentrations NAInfoGL 85-017Availability Of Supplements 2 and 3 To NUREG-0933, "A Prioritization Of Generic Safety Issues" NAInfoGL 85-018Operator Licensing Examinations NAInfoGL 85-019Reporting Requirements On Primary Coolant Iodine Spikes NAInfoGL 85-020Resolution Of Generic Issue 69: High Pressure Injection/Make-up Nozzle Cracking In Babcock And Wilcox Plants NAApplies only to Babcock and Wilcox designed plantsGL 85-021This GL was never issued.

NAGL 85-022Potential For Loss Of Post-LOCA Recirculation Capability Due To Insulation Debris Blockage NAInfoPage 68 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 86-001Safety Concerns Associated With Pipe Breaks In The BWR Scram System NABoiling Water ReactorGL 86-002Technical Resolution of Generic Issue B-19 - Thermal Hydraulic Stability NABoiling Water ReactorGL 86-003Applications For License Amendments NAInfoGL 86-004Policy Statement On Engineering Expertise On Shift CTVA responded to GL 86-04 on May 29, 1986. TVA provides engineering expertise on shift in the form of a dedicated Shift Technical Advisor (STA) or an STA qualified Senior Reactor Operator.01GL 86-005Implementation Of TMI Action Item II.K.3.5, "Automatic Trip Of Reactor Coolant Pumps" NAApplies only to Babcock and Wilcox designed plantsGL 86-006Implementation Of TMI Action Item II.K.3.5, "Automatic Trip of Reactor Coolant Pumps" NAApplies only to Combustion Engineering designed plantsGL 86-007Transmittal of NUREG-1190 Regarding The San Onofre Unit 1 Loss of Power and Water Hammer Event NAInfoGL 86-008Availability of Supplement 4 to NUREG-0933, "A Prioritization of Generic Safety Issues" NAInfoGL 86-009Technical Resolution of Generic Issue B-59, (N-1) Loop Operation in BWRs and PWRs SN-1 Loop operation was addressed in original 1982 SER (4.4.7). Unit 2 Action: Confirm Technical Specifications prohibit (N-1) Loop Operation.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:

Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.TS LCO 3.4.4 requires that four Reactor Coolant System loops be operable and in operation during Modes 1 and 2.

02GL 86-010Implementation of Fire Protection Requirements NAInfoGL 86-010, S1Fire Endurance Test Acceptance Criteria for Fire Barrier Systems Used to Separate Redundant Safe Shutdown Trains Within the Same Fire Area NAInfoPage 69 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 86-011Distribution of Products Irradiated in Research NADoes not apply to power reactor.GL 86-012Criteria for Unique Purpose Exemption From Conversion From The Use of Heu Fuel NADoes not apply to power reactor.GL 86-013Potential Inconsistency Between Plant Safety Analyses and Technical Specifications NAApplies only to Babcock and Wilcox and Combustion Engineering designed plantsGL 86-014Operator Licensing Examinations NAInfoGL 86-015Information Relating To Compliance With 10 CFR 50.49, "Environmental Qualification of Electric Equipment Important To Safety For Nuclear Power Plants" NAInfoGL 86-016Westinghouse ECCS Evaluation Models NAInfoGL 86-017Availability of NUREG-1169, "Technical Findings Related to Generic Issue C-8, BWR MSIC Leakage And Treatment Methods" NABoiling Water ReactorGL 87-001Public Availability Of The NRC Operator Licensing Examination Question Bank NAInfoGL 87-002 andGL 87-003Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, USI A-46 NAItem was applicable only to units with operating license at the time the item was issued.GL 87-004Temporary Exemption From Provisions Of The FBI Criminal History Rule For Temporary Workers NAItem was applicable only to units with operating license at the time the item was issued.GL 87-005Request for Additional Information on Assessment of License Measures to Mitigate and/or Identify Potential Degradation of Mark I Drywells NABoiling Water ReactorGL 87-006Periodic Verification of Leak Tight Integrity of Pressure Isolation Valves NAItem was applicable only to units with operating license at the time the item was issued.GL 87-007Information Transmittal of Final Rulemaking For Revisions To Operator Licensing - 10 CFR 55 And Confirming Amendments NAInfoPage 70 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 87-008Implementation of 10 CFR 73.55 Miscellaneous Amendments and Search Requirements NAItem was applicable only to units with operating license at the time the item was issued.GL 87-009Sections 3.0 And 4.0 of Standard Tech Specs on Limiting Conditions For Operation And Surveillance Requirements NAInfoGL 87-010Implementation of 10 CFR 73.57, Requirements For FBI Criminal History Checks NAItem was applicable only to units with operating license at the time the item was issued.GL 87-011Relaxation in Arbitrary Intermediate Pipe Rupture Requirements NAInfoGL 87-012Loss of Residual Heat Removal While The Reactor Coolant System is Partially Filled CThis GL was superseded by GL 88-17.GL 87-013Integrity of Requalification Examinations At Non-Power Reactors NADoes not apply to power reactor.GL 87-014Operator Licensing Examinations NAInfoGL 87-015Policy Statement On Deferred Plants NAInfoGL 87-016Transmittal of NUREG-1262, "Answers To Questions On Implementation of 10 CFR 55 On Operators' Licenses" NAInfoGL 88-001NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping NABoiling Water ReactorGL 88-002Integrated Safety Assessment Program II NAItem was applicable only to units with operating license at the time the item was issued.Page 71 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 88-003Resolution of GSI 93, "Steam Binding of Auxiliary Feedwater Pumps" CITVA: letter June 3, 1988. NRC letters dated February 17, 1988 and July 20, 1988NRC: SSER 16

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 88-004Distribution of Gems Irradiated in Research Reactors NADoes not apply to power reactor.GL 88-005Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR plants CINRC acceptance letter dated August 8, 1990 for both units. Unit 2 Action: Implement program.GL 88-006Removal of Organization Charts from Technical Specification Administrative Control Requirements NAInfoGL 88-007Modified Enforcement Policy Relating to 10 CFR 50.49, "Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants"CISee Special Program for Environmental Qualification.GL 88-008Mail Sent or Delivered to the Office of Nuclear Reactor Regulation NAInfoGL 88-009Pilot Testing of Fundamentals Examination NABoiling Water ReactorGL 88-010Purchase of GSA Approved Security Containers NAInfoPage 72 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 88-011NRC Position on Radiation Embrittlement of Reactor Vessel Material and its Impact on Plant Operations SNRC acceptance letter dated June 29, 1989, for both units. Unit 2 Action: Submit Pressure Temperature curves.

REVISION 02 UPDATE:

Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.WCAP-17035-NP "Watts Bar Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation" was submitted with the TS.

02GL 88-012Removal of Fire Protection Requirements from Technical Specification NAInfoGL 88-013Operator Licensing Examinations NAInfoGL 88-014Instrument Air Supply System Problems Affecting Safety-Related Equipment CINRC letter dated July 26, 1990, closing the issue. Unit 2 Action:

Complete Unit 2 implementation.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 04 UPDATE:

The compressed air system is a common system at Watts Bar; therefore, the requirements for this GL have been satisfied for

Unit 2.Watts Bar revised the response in a letter dated July 14, 1995.NRC letter dated July 27, 1995, stated that their conclusion as stated on July 26,1990, had not changed and that their effort was complete.04GL 88-015Electric Power Systems - Inadequate Control Over Design Process NAInfoGL 88-016Removal of Cycle-Specific Parameter Limits from Technical Specifications NAInfoGL 88-017Loss of Decay Heat Removal CINRC acceptance letter dated March 8, 1995 (Unit 1). Unit 2 Action: Implement modifications to provide RCS temperature, RV level and RHR system performance.GL 88-018Plant Record Storage on Optical Disks NAInfoPage 73 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 88-019Use of Deadly Force by Licensee Guards to Prevent Theft of Special Nuclear Material NADoes not apply to power reactor.GL 88-020Individual Plant Examination for Severe Accident Vulnerabilities SUnit 2 Action: Complete evaluation for Unit 2.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:

The Probabilistic Risk Assessment Individual Plant Examination Summary Report was submitted on February 9, 2010.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 04 UPDATE:The Individual Plant Examination of External Events Design Report was submitted on April 30, 2010.

04GL 89-001Implementation of Programmatic and Procedural Controls for Radiological Effluent Technical Specifications NAInfoGL 89-002Actions to Improve the Detection of Counterfeit and Fraudulently Marketed Products CGL 89-02 did not require a response.WBN Unit 2 program for procurement and dedication of materials is based in part on and complies with the guidance of GL 89-02. The program is implemented through project procedures.

01GL 89-003Operator Licensing Examination Schedule NAInfoGL 89-004Guidelines on Developing Acceptable Inservice Testing Programs OVNRC reviewed in 3.9.6 of SSER14 (Unit 1). Unit 2 Action: Submit an ASME Section XI Inservice Test Program for the first ten year interval six months before receiving an Operating License.GL 89-005Pilot Testing of the Fundamentals Examination NAInfoGL 89-006Task Action Plan Item I.D.2 - Safety Parameter Display System - 10 CFR 50.54(f)CI"Safety Parameter Display System" (SPDS) / "Requirements for Emergency Response Capability" - NRC reviewed in SSER5, SSER6, and 18.2.2 of SSER15. Unit 2 Action: Install SPDS and have it operational prior to start-up after the first refueling outage.Page 74 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 89-007Power Reactor Safeguards Contingency Planning for Surface Vehicle Bombs CTVA: letter dated October 31, 1989NRC: memo dated June 26, 1990GL 89-008Erosion/Corrosion-Induced Pipe Wall Thinning CIUnit 1 Flow Accelerated Corrosion Program reviewed in IR 390/94-89 (February 1995). Unit 2 Actions: Prepare procedure, and perform baseline inspections.GL 89-009ASME Section III Component Replacements NAItem was applicable only to units with operating license at the time the item was issued.GL 89-010Safety-Related Motor-Operated Valve Testing and Surveillance CINRC accepted approach in September 14, 1990, letter and reviewed in Appendix EE of SSER16. Unit 2 Action:

Implement pressure testing and surveillance program for safety-related MOVs, satisfying the intent of GL 89-10.GL 89-010 or GL 96-005Involves Main Steam Isolation Valves NABoiling Water ReactorGL 89-011Resolution of Generic Issue 101, "Boiling Water Reactor Water Level Redundancy" NABoiling Water ReactorGL 89-012Operator Licensing Examination NAInfoGL 89-013Service Water System Problems Affecting Safety-Related Equipment CINRC letters dated July 9, 1990 and June 13, 1997, accepting approach. Unit 2 Actions:

1) Implement initial performance testing of the heat exchangers; and 2) Establish eddy current baseline data for the Containment Spray heat exchangers.GL 89-014Line-Item Improvements in Technical Specifications - Removal of 3.25 Limit on Extending Surveillance Intervals NAInfoGL 89-015Emergency Response Data System NAInfoGL 89-016Installation of a Hardened Wetwell Vent NABoiling Water ReactorPage 75 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 89-017Planned Administrative Changes to the NRC Operator Licensing Written Examination Process NAInfoGL 89-018Resolution of Unresolved Safety Issues A-17, "Systems Interactions in Nuclear Power Plants" NAInfoGL 89-019Request for Actions Related to Resolution of Unresolved Safety Issue A-47, "Safety Implication of Control Systems in LWR Nuclear Power Plants" Pursuant to 10 CFR 50.54(f)

CITVA responded by letter dated March 22, 1990. NRC acceptance letter dated October 24, 1990, for both units. Unit 2 Action: Perform evaluation of common mode failures due to fire.GL 89-020Protected Area Long-Term Housekeeping NADoes not apply to power reactor.GL 89-021Request for Information Concerning Status of Implementation of Unresolved Safety Issue (USI) Requirements STVA responded to GL 89-21 with the status of USIs for both units on November 29, 1989. NRC provided an assessment of WBN USI status on May 1, 1990. The NRC assessment included a list of incomplete USIs for WBN. USIs were initially reviewed for WBN in the SER Appendix C. USIs were subsequently reviewed in SSER 15 Appendix C (June 1995) and SSER 16 (September 1995).Unit 2 actions: Provide a status of WBN Unit 2 USIs.Complete implementation of USIs.

REVISION 02 UPDATE:Status of USIs was provided by Enclosure 2 of TVA letter dated September 26, 2008.The applicable USIs are either closed, deleted, or captured in either the SER Framework or the Generic Communications Framework, or they are part of the CAPs and SPs.

02GL 89-022Potential For Increased Roof Loads and Plant Area Flood Runoff Depth At Licensed Nuclear Power Plants Due To Recent Change In Probable Maximum Precipitation Criteria Developed by the National Weather Service CTVA: letter dated December 16, 1981

Answer to informal question provided in TVA letter dated December 16, 1981, and subsequently included in FSAR. GL did not require a response. No further action required.GL 89-023NRC Staff Responses to Questions Pertaining to Implementation of 10 CFR Part 26 NAInfoPage 76 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 90-001Request for Voluntary Participation in NRC Regulatory Impact Survey NAInfoGL 90-002Alternative Requirements for Fuel Assemblies in the Design Features Section of Technical Specifications NAInfoGL 90-003Relaxation of Staff Position in Generic Letter 83-28, Item 2.2 Part 2 "Vendor Interface for Safety-Related Components" NAInfoGL 90-004Request for Information on the Status of Licensee Implementation of GSIs Resolved with Imposition of Requirements or CAs CTVA responded on June 23, 1990GL 90-005Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping NAInfoGL 90-006Resolution of Generic Issues 70, "PORV and Block Valve Reliability," and 94, "Additional LTOP Protection for PWRs" SNRC letter dated January 9, 1991, accepted TVA's response for both units. Unit 2 Actions: 1) Revise operating instruction and surveillance procedure; and 2) Incorporate testing requirements in the Technical Specifications.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:

Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.TS Surveillance Requirement 3.4.11.2 specifies the required testing of each PORV.

02GL 90-007Operator Licensing National Examination Schedule NAInfoGL 90-008Simulation Facility Exemptions NAInfoGL 90-009Alternative Requirements for Snubber Visual Inspection Intervals and Corrective Actions NAInfoGL 91-001Removal of the Schedule for the Withdrawal of Reactor Vessel Material Specimens from Technical Specifications NAInfoPage 77 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 91-002Reporting Mishaps Involving LLW Forms Prepared for Disposal NAItem was applicable only to units with operating license at the time the item was issued.GL 91-003Reporting of Safeguards Events NAInfoGL 91-004Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle NAInfoGL 91-005Licensee Commercial-Grade Procurement and Dedication Programs NAInfoGL 91-006Resolution of Generic Issue A-30, "Adequacy of Safety-Related DC Power Supplies," Pursuant to 10 CFR 50.54(f)

NAItem was applicable only to units with operating license at the time the item was issued.GL 91-007GI-23, "Reactor Coolant Pump Seal Failures" and Its Possible Effect on Station Blackout NAInfoGL 91-008Removal of Component Lists from Technical Specifications NAInfoGL 91-009Modification of Surveillance Interval for the Electrical Protective Assemblies in Power Supplies for the Reactor Protection System NABoiling Water ReactorGL 91-010Explosives Searches at Protected Area Portals NADoes not apply to power reactor.GL 91-011Resolution of Generic Issues A-48, "LCOs for Class 1E Vital Instrument Buses", and 49, "Interlocks and LCOs for Class 1E Tie Breakers," Pursuant to 10 CFR 50.54 NAItem was applicable only to units with operating license at the time the item was issued.GL 91-012Operator Licensing National Examination Schedule NAInfoGL 91-013Request for Information Related to Resolution of Generic Issue 130, "Essential Service Water System Failures@ Multi-Unit Sites" NAAddressed to specific (non-TVA) plants.GL 91-014Emergency Telecommunications NAInfoPage 78 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 91-015Operating Experience Feedback Report, Solenoid-Operated Valve Problems at U.S. Reactors NAInfoGL 91-016Licensed Operators' and Other Nuclear Facility Personnel Fitness for Duty NAInfoGL 91-017Generic Safety Issue 29, "Bolting Degradation or Failure in Nuclear Power Plants" NAInfoGL 91-018Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability NAGL 91-18 has been superseded by RIS 2005-20.GL 91-019Information to Addressees Regarding New Telephone Numbers for NRC Offices Located in One White Flint North NAInfoGL 92-001Reactor Vessel Structural Integrity CBy 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 92-002Resolution of Generic Issue 79, "Unanalyzed Reactor Vessel (PWR) Thermal Stress During Natural Convection Cooldown" NAInfoGL 92-003Compilation of the Current Licensing Basis: Request for Voluntary Participation in Pilot Program NAInfoGL 92-004Resolution of the Issues Related to Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(f)

NABoiling Water ReactorGL 92-005NRC Workshop on the Systematic Assessment of Licensee Performance (SALP) Program NAInfoGL 92-006Operator Licensing National Examination Schedule NAInfoGL 92-007Office of Nuclear Reactor Regulation Reorganization NAInfoPage 79 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 92-008Thermo-Lag 330-1 Fire Barriers OVTVA 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 92-009Limited Participation by NRC in the IAEA International Nuclear Event Scale NAInfoGL 93-001Emergency Response Data System Test Program NAAddressed to specific plant(s).GL 93-002NRC Public Workshop on Commercial Grade Procurement and Dedication NAInfoGL 93-003Verification of Plant Records NAInfoGL 93-004Rod Control System Failure and Withdrawal of Rod Control Cluster Assemblies, 10 CFR 50.54(f)

CINRC letter dated December 9, 1994, accepted TVA commitments for both units. Unit 2 Action: Implement modifications and testing.GL 93-005Line-Item Technical Specifications Improvements to Reduce Surveillance Requirements for Testing During Power Operation NAInfoGL 93-006Research Results on Generic Safety Issue 106, "Piping and the Use of Highly Combustible Gases in Vital Areas" NAInfoGL 93-007Modification of the Technical Specification Administrative Control Requirements for Emergency and Security Plans NAItem was applicable only to units with operating license at the time the item was issued.GL 93-008Relocation of Technical Specification Tables of Instrument Response Time Limits NAItem was applicable only to units with operating license at the time the item was issued.GL 94-001Removal of Accelerated Testing and Special Reporting Requirements for Emergency Diesel Generators NAItem was applicable only to units with operating license at the time the item was issued.Page 80 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 94-002Long-Term Solutions and Upgrade of Interim Operating Recommendations for Thermal-Hydraulic Instabilities in BWRs NABoiling Water ReactorGL 94-003IGSCC of Core Shrouds in BWRs NABoiling Water ReactorGL 94-004Voluntary Reporting of Additional Occupational Radiation Exposure Data NAInfoGL 95-001NRC Staff Technical Position on Fire Protection for Fuel Cycle Facilities NADoes not apply to power reactor.GL 95-002Use of NUMARC/EPRI Report TR-102348, "Guideline on Licensing Digital Upgrades," in Determining the Acceptability of Performing Analog-to-Digital Replacements under 10 CFR 50.59 NAInfoGL 95-003Circumferential Cracking of Steam Generator Tubes CINRC acceptance letter dated May 16, 1997 (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 baseline inspection.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:Unit 2 Action:

Perform baseline inspection.

Evaluate or repair as necessary.

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

02GL 95-004Final Disposition of the Systematic Evaluation Program Lessons-Learned Issues NAInfoGL 95-005Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking CNo specific action or response required by the GL; TVA responded on September 7, 2007.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

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

02Page 81 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 95-006Changes in the Operator Licensing Program NAInfoGL 95-007Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves CIUnit 1 SER for GL 95-07 dated Sept 15, 1999 Unit 2 Actions:

Perform evaluation for pressure locking and thermal binding of safety related power-operated gate valves, and take corrective actions for those valves identified as being susceptible.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 03 UPDATE:

April 1, 2010, letter committed to evaluate missing GL 89-10 motor-operated valves for susceptibility to pressure locking and thermal binding.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 04 UPDATE:

NRC letter dated July 29, 2010, provided RAIs on the GL.

TVA letter dated July 30, 2010, answered the RAIs and provided the following commitments:EDCRs 53292 and 53287 shall be implemented to eliminate the potential for pressure locking prior to startup.Valves 2-FCV-63-25 and -26 will be evaluated for impact due to new parameters from the JOG Topical Report MPR 2524A prior to startup.

NRC issued the Safety Evaluation for GL 1995-007 on August 12, 2010.

04GL 95-00810 CFR 50.54(p) Process for Changes to Security Plans Without Prior NRC Approval NAInfoGL 95-009Monitoring and Training of Shippers and Carriers of Radioactive Materials NAInfoGL 95-010Relocation of Selected Technical Specifications Requirements Related to Instrumentation NAInfoPage 82 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 96-001Testing of Safety-Related Circuits CITVA responded for both units on April 18, 1996. Unit 2 Action:

Implement Recommendations.GL 96-002Reconsideration of Nuclear Power Plant Security Requirements Associated with an Internal Threat NAInfoGL 96-003Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits CINo response requiredUnit 2 Actions: Submit Pressure Temperature limits, andsimilar to Unit 1, upon approval, incorporate into licensee-controlled document.GL 96-004Boraflex Degradation in Spent Fuel Pool Storage Racks NAItem was applicable only to units with operating license at the time the item was issued.GL 96-005Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves CISE of TVA response to GL 96-05 dated July 21, 1999. Unit 2 Actions:

Implement the Joint Owner's Group recommended GL 96-05 MOV PV program, as described in Topical Report No. OG-97-018, and begin testing during the first refueling outage after startup.GL 96-006Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions CINRC letter dated April 6, 1999, accepting TVA response for Unit 1. Unit 2 Action:

Implement modification to provide containment penetration relief.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:

NRC issued the Safety Evaluation for Generic Letter 1996-006 on January 21, 2010.

02GL 96-007Interim Guidance on Transportation of Steam Generators NAItem was applicable only to units with operating license at the time the item was issued.Page 83 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 97-001Degradation of Control Rod Drive Mechanism Nozzle and Other Vessel Closure Head Penetrations CINRC acceptance letter dated November 4, 1999 (Unit 1). Unit 2 Action:

Provide a report to address the inspection program.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 03 UPDATE:NRC issued the Safety Evaluation for Generic Letter 97-001 on June 30, 2010.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 04 UPDATE:

Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.

04GL 97-002Revised Contents of the Monthly Operating Report NAItem was applicable only to units with operating license at the time the item was issued.GL 97-003Annual Financial Update of Surety Requirements for Uranium Recovery Licensees NADoes not apply to power reactor.GL 97-004Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps CINRC acceptance letter dated June 17, 1998 (Unit 1) - Initial response for Unit 2 on September 7, 2007. Unit 2 Actions:

Install new sump strainers, and perform other modification-related activities identical to Unit 1.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:NRC issued the Safety Evaluation for Generic Letter 1997-004 on February 18, 2010.

02Page 84 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 97-005Steam Generator Tube Inspection Techniques CINRC acceptance letter dated September 22, 1998 (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.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:

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

02GL 97-006Degradation of Steam Generator Internals CINRC acceptance letter dated October 19, 1999 (Unit 1) - Initial response for Unit 2 on September 7, 2007. TVA responded to a request for additional information on December 17, 2007. Unit 2 Action:

Perform SG inspections during each refueling outage.

REVISION 02 UPDATE:

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

02GL 98-001Year 2000 Readiness of Computer Systems at Nuclear Power Plants NAItem was applicable only to units with operating license at the time the item was issued.Page 85 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 98-002Loss of Reactor Coolant Inventory and Associated Potential for Loss of Emergency Mitigation Functions While in a Shutdown Condition CIInitial response for Unit 2 on September 7, 2007. Unit 2 Actions:

1) Review the ECCS designs to ensure they do not contain design features which can render them susceptible to common-cause failures; and 2) document the results.

REVISION 02 UPDATE:

NRC issued the Safety Evaluation for Generic Letter 1998-002 on March 3, 2010.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 03 UPDATE:NRC issued the Safety Evaluation for Generic Letter 98-002 on May 11, 2010. This letter noted that it superseded the SE issued by NRC on March 3, 2010.

April 1, 2010, letter committed to ensure that the guidance added to the Unit 1 procedure as a result of the review of NRC GL 98-02 is incorporated into the Unit 2 procedures. Specifically, when decreasing power, valve HCV-74-34, Refueling Water Return (normally locked closed valve) has a hold order placed with specific release criteria before entry into Mode 4 and to remove the hold order before entry into Mode 3 when returning to power.

03GL 98-003NMSS Licensees' and Certificate Holders' Year 2000 Readiness Programs NADoes not apply to power reactor.GL 98-004Potential for Degradation of the ECCS and the Containment Spray System After a LOCA Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment OVNRC closure letter dated November 24, 1999 (Unit 1). - Initial response for Unit 2 on September 7, 2007. Unit 2 Actions: Install new sump strainers, and perform other modification-related activities identical to Unit 1.

REVISION 02 UPDATE:

NRC issued the Safety Evaluation for Generic Letter 1998-004 on February 1, 2010.

02Page 86 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 98-005Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds NABoiling Water ReactorGL 99-001Recent Nuclear Material Safety and Safeguards Decision on Bundling Exempt Quantities NAInfoGL 99-002Laboratory Testing of Nuclear Grade Activated Charcoal NAItem was applicable only to units with operating license at the time the item was issued.GL 03-001Control Room Habitability SInitial response for Unit 2 on September 7, 2007Unit 2 Action: Incorporate TSTF-448 into Technical Specifications.----------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:

NRC issued the Safety Evaluation for Generic Letter 2003-01 on February 1, 2010.

Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.TS Surveillance Requirement 3.7.10.4 requires performance of a Control Room Envelope (CRE) unfiltered air inleakage test in accordance with the CRE Habitability Program.TS 5.7.2.20 provides for the CRE Habitability Program.

These portions of the Unit 2 TS were based on the Unit 1 TS which incorporated TSTF-448 per Amendment 70 (NRC approved A70 on 10/08/2008).

02GL 04-001Requirements for Steam Generator Tube Inspection CINRC acceptance letter dated April 8, 2005 (Unit 1) - Initial response for Unit 2 on September 7, 2007.Unit 2 Action:

Perform baseline inspection.

REVISION 02 UPDATE:

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

02Page 87 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 04-002Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at PWRs OVNRC Audit Report dated February 7, 2007 (Unit 1) - Initial response for Unit 2 on September 7, 2007.Unit 2 Actions:

Install new sump strainers, and perform other modification-related activities identical to Unit 1.GL 06-001Steam Generator Tube Integrity and Associated Technical Specifications SInitial response for Unit 2 on September 7, 2007.Unit 2 Action: Incorporate TSTF-449 into Technical Specifications.----------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:

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

Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.TS 5.7.2.12 is the Steam Generator (SG) Program. This program is implemented to ensure that SG tube integrity is maintained.Unit 2 TS 5.7.2.12 was based on Unit 1 TS 5.7.2.12. Unit 1 TS 5.7.2.1.12 was based on TSTF-449 (NRC approved Unit 1 TS A65 on 1/03/2006).

02GL 06-002Grid Reliability and the Impact on Plant Risk and the Operability of Offsite Power CIInitial response for Unit 2 on September 7, 2007. Unit 2 Action:

Complete the two unit baseline electrical calculations and implementing procedures.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:

NRC issued the Safety Evaluation for Generic Letter 2006-002 on January 20, 2010.

02Page 88 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVGL 06-003Potentially Nonconforming Hemyc and MT Fire Barrier Configurations CITVA does not rely on Hemyc or MT materials to protect electrical and instrumentation cables or equipment that provide safe shutdown capability during a postulated fire. Unit 2 Action: Addressed in CAP/SP. The Fire Protection Corrective Action Program will ensure Unit 2 conforms with NRC requirements and applicable guidelines.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:

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

02GL 07-001Inaccessible or Underground Power Cable Failures That Disable Accident Mitigation Systems or Cause Plant Transients CIInitial response for Unit 2 on September 7, 2007. Unit 2 Action: Complete testing of four additional cables.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:NRC issued the Safety Evaluation for Generic Letter 2007-001 on January 26, 2010.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 04 UPDATE:

NRC Inspection Report 391/2010-603 closed GL 2007-001.

02GL 08-001Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems OInitial response for Unit 2 on October 1, 2008.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:

Unit 2 Actions:

- TVA will provide a submittal within 45 days of completion of the engineering for the ECCS, RHR, and CSS systems.- WBN Unit 2 will complete the required modifications and provide a submittal consistent with the information requested in the GL 90 days prior to fuel load.

02NUREG-0737, I.A.1.1Shift Technical Advisor NANot applicable to WBN per SSER16.Page 89 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVNUREG-0737, I.A.1.2Shift Supervisor Responsibilities NANot applicable to WBN per SSER16.NUREG-0737, I.A.1.3Shift Manning CClosed in SSER16.NUREG-0737, I.A.2.1Immediate Upgrade of RO and SRO Training and Qualifications CClosed in SSER16.NUREG-0737, I.A.2.3Administration of Training Programs CClosed in SSER16.NUREG-0737, I.A.3.1Revise Scope and Criteria for Licensing Exams CClosed in SSER16.NUREG-0737, I.B.1.2 Independent Safety Engineering Group OVLICENSE CONDITION - Independent Safety Engineering 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-0737,I.C.1Short Term Accident and Procedure Review CINRC reviewed in Appendix EE of SSER16. Unit 2 Action:

Implement upgraded Emergency Operating Procedures, including validation and training.NUREG-0737, I.C.2Shift and Relief Turnover Procedures CClosed in SSER16.NUREG-0737, I.C.3Shift Supervisor Responsibility CClosed in SSER16.NUREG-0737, I.C.4Control Room Access CClosed in SSER16.NUREG-0737, I.C.5Feedback of Operating Experience CClosed in SSER16.Page 90 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVNUREG-0737, I.C.6Verify Correct Performance of Operating Activities CClosed in SSER16.NUREG-0737, I.C.7NSSS Vendor Revision of Procedures CIIR 50-390/391 85-08 closed this item for Unit 1, and NRC also reviewed in Appendix EE of SSER16.Unit 2 Action: Revise power ascension and emergency procedures which were reviewed by Westinghouse.NUREG-0737, I.C.8Pilot Monitoring of Selected Emergency Procedures For Near Term Operating Licenses CIIR 50-390/391 85-08 closed this item for Unit 1, and NRC also reviewed in Appendix EE of SSER16.Unit 2 Action: Pilot monitor selected emergency procedures for NTOL.NUREG-0737, I.D.1Control Room Design Review OVNRC reviewed in SSER5, SSER6, SSER15, and Appendix 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-0737, I.D.2Plant-Safety-Parameter-Display Console CINRC 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.NUREG-0737, I.G.1Training During Low-Power Testing CClosed in SSER16.NUREG-0737, II.B.1Reactor Coolant Vent System CILICENSE CONDITION - NUREG-0737, II.B.1, "Reactor Coolant System Vents" - In the original SER, the NRC found TVA's commitment to install reactor coolant vents acceptable pending verification. This was completed for Unit 1 only in SSER5 (IR 390/84-37). Unit 2 Action:

Verify installation of reactor coolant vents.Page 91 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVNUREG-0737, II.B.2Plant Shielding CINRC reviewed in Appendix EE of SSER16. Unit 2 Action:

Complete Design Review of EQ of equipment for spaces/systems which may be used in post accident operations.NUREG-0737, II.B.3Post-Accident Sampling SNRC reviewed in 9.3.2 of SSER16. TVA submitted a 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.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.Rev. 0 of the Unit 1 TS contained 5.7.2.6, "Post Accident Sampling."Amendment 34 to the Unit 1 TS (approved by the NRC on January 14, 2002) deleted 5.7.2.6, "Post Accident Sampling." The markup for Unit 2 Developmental Revision A noted that Unit 2 had deleted 5.7.2.6, "Post Accident Sampling" also.

02NUREG-0737, II.B.4Training for Mitigating Core Damage CClosed in SSER16.NUREG-0737, II.D.1Relief and Safety Valve Test Requirements CINRC reviewed in Technical Evaluation Report attached to Appendix EE of SSER15. Unit 2 Actions: 1) Testing of relief and safety valves; 2) Reanalysis of fluid transient loads for pressurizer relief and safety valve supports and any required modifications; 3) Modifications to pressurizer safety valves, PORVs, PORV block valves and associated piping; and 4) Change motor operated block valves.Page 92 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVNUREG-0737, II.D.3Valve Position Indication CIThe design was reviewed in the original 1982 SER and found acceptable pending confirmation of installation of the acoustic monitoring system. In SSER5 (IR 390/84-35), the staff closed the LICENSE CONDITION for Unit 1 only. Unit 2 Action:

Verify installation of the acoustic monitoring system to PORV to indicate position.NUREG-

0737, II.E.1.1Auxiliary Feedwater System Evaluation, Modifications CIReviewed in Appendix EE of SSER16. Unit 2 Action: Perform Auxiliary Feedwater System analysis as it pertains to system failure and flow rate.NUREG-0737, II.E.1.2Auxiliary Feedwater System Initiation and Flow CINRC: IR 50-390/84-20 and 50-391/84-16; letters datedMarch 29, 1985, and October 31, 1995; SSER 16

Unit 2 Actions: Complete procedures, and qualification testing.NUREG-0737, II.E.3.1Emergency Power For Pressurizer Heaters CINRC: letters dated March 29, 1985, and October 31, 1995; SSER 16-------------------Reviewed in original 1982 SER. Unit 2 Action: Implement procedures and testing.NUREG-0737, II.E.4.1Dedicated Hydrogen Penetrations CNRC: IR 50-390/83-27 and 50-391/83-19; SER (NUREG-0847)Page 93 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVNUREG-0737, II.E.4.2Containment Isolation Dependability STVA: letters dated October 29, 1981, and February 25, 1985NRC: 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 OperationLICENSE 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.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.TS Surveillance Requirement 3.6.3.7 requires verification that the valves are "blocked to restrict the valve from opening

> 50 degrees." 02NUREG-0737, II.F.1.2.A.Accident-Monitoring Instrumentation - Noble Gas CIReviewed in SSER9. Unit 2 Actions:

Install Noble gas, Iodine / particulate sampling, and Containment High Range Monitors.NUREG-0737, II.F.1.2.B.Accident-Monitoring Instrumentation - Iodine/Particulate Sampling CIReviewed in SSER9. Unit 2 Actions:

Install Noble gas, Iodine / particulate sampling, and Containment High Range Monitors.Page 94 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVNUREG-0737, II.F.1.2.C.Accident-Monitoring Instrumentation - Containment High Range Monitoring CIReviewed in SSER9. 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-0737, II.F.1.2.D.Accident-Monitoring Instrumentation - Containment Pressure CIReviewed in SSER9. Unit 2 Action:

Verify installation of containment pressure indication.NUREG-0737, II.F.1.2.E.Accident-Monitoring Instrumentation - Containment Water Level CIReviewed in SSER9. Unit 2 Action:

Verify installation of containment water level monitors.NUREG-0737, II.F.1.2.F.Accident-Monitoring Instrumentation - Containment Hydrogen CIReviewed in SSER9. Unit 2 Action: Verify installation of containment hydrogen accident monitoring instrumentation.NUREG-0737, II.F.2Instrumentation For Detection of Inadequate Core-Cooling OLICENSE CONDITION - Detectors for Inadequate core cooling (II.F.2)In the original SER, the review of the ICC instrumentation was incomplete. The January 24, 1992, letter superseded the previous responses on this issue. TVA letter for Units 1 and 2 dated January 24, 1992, committed to install Westinghouse ICCM-86 and associated hardware. NRC completed the review for Units 1 and 2 in SSER10. For Unit 2 due to obsolescence of the ICCM-86 system, TVA intends to install the Westinghouse Common Q Post-Accident Monitoring System. Unit 2 Action:

Install Westinghouse Common Q PAM system.NUREG-0737, II.G.1Power Supplies For Pressurizer Relief Valves, Block Valves and Level Indicators CIReviewed in original 1982 SER and 8.3.3 of SSER7. Unit 2 Action:

Implement modifications such that PORVS and associated Block Valves are powered from same train but different buses.Page 95 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVNUREG-0737, II.K.1.5Review ESF Valves CNRC: letter dated March 29, 1985; SSER 16NUREG-0737, II.K.1.10Operability Status CIUnit 2 Action: Confirm multi-unit operation will have no impact on administrative procedures with respect to operability status.NUREG-

0737, II.K.1.17Trip Per Low-Level B/S CNRC: letter dated March 29, 1985; SSER 16NUREG-0737, II.K.2.13Effect of High Pressure Injection for Small Break LOCA With No Auxiliary Feedwater CLICENSE CONDITION - Effect of high pressure injection for small break LOCA with no auxiliary feedwater (NUREG-0737, II.K.2.13)In SSER4, the staff concluded that there was reasonable assurance that vessel integrity would be maintained for small breaks with an extended loss of all feedwater and that the USI A-49, "Pressurized Thermal Shock," review did not have to be completed to support the full-power license. They considered this condition resolved.NUREG-

0737, II.K.2.17Voiding in the Reactor Coolant System CLICENSE CONDITION - Voiding in the reactor coolant system (NUREG-0737, II.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-0737, II.K.3.1Auto PORV Isolation CReviewed in SSER5 and resolved based on NRC conclusion that there is no need for an automatic PORV isolation system (NRC letter dated June 29, 1990).NUREG-0737, II.K.3.2Report on PORV Failures CReviewed in SSER5 and resolved based on NRC conclusion that there is no need for an automatic PORV isolation system (NRC letter dated June 29, 1990).Page 96 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVNUREG-0737, II.K.3.3Reporting SV/RV Failures/Challenges S(Action from GL 82-16) - NRC reviewed in Appendix EE of SSER16. Unit 2 Action: Include, as necessary, in Technical Specifications submittal.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.Rev. 0 of the Unit 1 TS contained 5.9.4 (Monthly Operating Reports) which implemented the above commitment for Unit 1.Amendment 57 to the Unit 1 TS (approved by the NRC on March 21, 2005) deleted this section of the TS. The markup for Unit 2 Developmental Revision A noted that Unit 2 will apply this change, and the Unit 2 TS will contain no requirement for Monthly Operating Reports.

02NUREG-0737, II.K.3.5Auto Trip of RCPS CIReviewed in 15.5.4 of original 1982 SER; became License Condition 35. The staff determined that their review of Item II.K.3.5 did not have to be completed to support the full power license and considered this license condition resolved in SSER4. The item was further reviewed in Appendix EE of SSER16. Unit 2 Action:

Implement modifications as required.NUREG-0737, II.K.3.9PID Controller CIReviewed in original 1982 SER. Unit 2 Action:

Set the derivative time constant to zero.Page 97 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVNUREG-0737, II.K.3.10Anticipatory Trip at High Power SNRC: letter dated October 31, 1995; SSER 16

Unit 2 Action: Unit 2 Technical Specifications and surveillance procedures will address this issue.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:

Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.Items 14.a. (Turbine Trip - Low Fluid Oil Pressure) and 14.b. (Turbine Trip - Turbine Stop Valve Closure) of TS Table 3.3.1-1 are the trips of interest. The table and the Bases for these items state that below the P-9 setpoint, these trips do not actuate a reactor trip.Per item 16.d. (Power Range Neutron Flux, P-9) of TSTable 3.3.1-1, the Nominal Trip Setpoint for P-9 is"50% RTP" and the Allowable Value is "< 52.4% RTP."

02NUREG-0737, II.K.3.12Confirm Existence of Anticipatory Reactor Trip Upon Turbine Trip CClosed in SSER16.NUREG-0737, II.K.3.17Report On Outage of Emergency Core Cooling System CLICENSE CONDITION - Report on outage of emergency core cooling system (NUREG-0737, 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 50.73NUREG-0737, II.K.3.25Power On Pump Seals CINRC reviewed and closed in IR 390/84-35 based on Diesel Generator (DG) power to pump sealing cooling system.Unit 2 Action:

Ensure DG power is provided to pump sealing cooling system.Page 98 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVNUREG-0737, II.K.3.30Small Break LOCA Methods CITVA: letter dated October 29, 1981NRC: letters dated March 29, 1985, and July 24, 1986; SSER 16

The staff determined in SSER4 that their review of Items II.K.3.30 and II.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-0737, II.K.3.31Plant Specific Analysis CIThe staff determined in SSER4 that their review of Items II.K.3.30 and II.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-0737, III.A.1.1Emergency Preparedness, Short Term CLICENSE CONDITION - Emergency Preparedness (NUREG-0737, III.A.1, III.A.2)The NRC review of Emergency Preparedness in SSER13 superseded the review in the original 1982 SER. In SSER13, the staff concluded that the WBN Radiological Emergency Plan (REP) provided an adequate planning basis for an acceptable state of onsite emergency preparedness, and the LICENSE CONDITION was deleted. The NRC completed the review of the

REP in SSER20.NUREG-

0737, III.A.1.2Upgrade Emergency Support Facilities CLICENSE CONDITION - Emergency Preparedness (NUREG-0737, III.A.1, III.A.2)The NRC review of Emergency Preparedness in SSER13 superseded the review in the original 1982 SER. In SSER13, the staff concluded that the WBN Radiological Emergency Plan (REP) provided an adequate planning basis for an acceptable state of onsite emergency preparedness, and the LICENSE CONDITION was deleted. The NRC completed the review of the REP in SSER20.Page 99 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVNUREG-0737, III.A.2Emergency Preparedness CLICENSE CONDITION - Emergency Preparedness (NUREG-0737, III.A.1, III.A.2)The NRC review of Emergency Preparedness in SSER13 superseded the review in the original 1982 SER. In SSER13, the staff concluded that the WBN Radiological Emergency Plan (REP) provided an adequate planning basis for an acceptable state of onsite emergency preparedness, and the LICENSE CONDITION was deleted. The NRC completed the review of the REP in SSER20.NUREG-0737, III.D.1.1Primary Coolant Outside Containment SResolved for Unit 1 only in SSER10; reviewed in Appendix EE of SSER16. Unit 2 Actions: Include the waste gas disposal system in the leakage reduction program and incorporate in Unit 2 Technical Specifications.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 02 UPDATE:

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

02NUREG-0737, III.D.3.3In-Plant Iodine Radiation Monitoring CINRC reviewed in Appendix EE of SSER16. Unit 2 Action:

Complete modifications for Unit 2.NUREG-0737, III.D.3.4Control-Room Habitability OVTVA: letter dated October 29, 1981NRC: SSER 16

NRC reviewed in SER and in Appendix EE of SSER16. Unit 2 Action: Complete with CRDR completion.Page 100 of 101* = See last page for status code definition.

ITEMTITLE*ADDITIONAL INFORMATION REVSTATUS 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.

O:OPEN: No action or documentation is provided that shows the staff has reviewed the item for WBN Unit 2.OT: OV: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.OPEN/VALIDATION: The proposed approach has been approved for Watts Bar Unit 1; the same approach is proposed for use on WBN Unit 2 without change.

S:SUBMITTED: Information has been submitted, and is under review by NRC staff.Page 101 of 101* = See last page for status code definition.

Enclosure 4

Generic Communications - Revision 5 Changes GENERIC COMMUNICATIONS: REVISION 5 CHANGESITEMTITLE*ADDITIONAL INFORMATION REVB 07-001Security Officer Attentiveness CItem concerns a multi-unit issue that was completed for both units.--------------------------------------------------------------------------------------------------------------------------------------------------------------------------------

REVISION 05 UPDATE:The NRC closed this bulletin via letter dated March 25, 2010 (ADAMS Accession No. ML100770549).

05STATUS 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.

O:OPEN: No action or documentation is provided that shows the staff has reviewed the item for WBN Unit 2.OT: OV: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.OPEN/VALIDATION: The proposed approach has been approved for Watts Bar Unit 1; the same approach is proposed for use on WBN Unit 2 without change.

S:SUBMITTED: Information has been submitted, and is under review by NRC staff.Page 1 of 1* = See last page for status code definition.

v t e Letter Sequence Other
TAC:MD8314, Clarify Application of Setpoint Methodology for LSSS Functions, Control Room Habitability, Steam Generator Tube Integrity (Open) TAC:ME6311, Clarify Application of Setpoint Methodology for LSSS Functions (Withdrawn, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance... Supplement, Supplement, Supplement Administration Questions 22 May 2008 Answers 2 July 2008 9 November 2010 Results Approval... Other: CNL-14-154, Status of Regulatory Framework for the Completion of Construction and Licensing for Unit 2 - Revision 12 (TAC No. MD6311), and Status of Generic Communications for Unit 2 - Revision 12, CNL-15-137, Status of Regulatory Framework for the Completion of Construction and Licensing for Unit 2 - Revision 13 (TAC No. MD6311) and Status of Generic Communications for Unit 2 - Revision 13, CNL-15-150, Pressurizer PORV Block Valves - Dynamic Testing Alternative, ML080370185, ML081490110, ML081490118, ML082140238, ML082560373, ML082750019, ML101200204, ML101260469, ML101680561, ML101760222, ML102170077, ML102290258, ML102700527, ML103010299, ML110210486, ML11213A130, ML11362A056, ML12234A688, ML13023A315, ML13213A194
MONTHYEAR2010-09-01 [Table View]

ML110210486

Text

Subject:

Reference:

Enclosures:

Navigation menu

Search