ML103140374: Difference between revisions

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| number = ML103140374
| number = ML103140374
| issue date = 10/28/2010
| issue date = 10/28/2010
| title = 2010/10/28 Watts Bar 2 OL - TVA Letter to NRC_10-28-10_Regulatory Framework Rev 4
| title = OL - TVA Letter to NRC_10-28-10_Regulatory Framework Rev 4
| author name =  
| author name =  
| author affiliation = - No Known Affiliation
| author affiliation = - No Known Affiliation
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* REV.                                      ADDITIONAL INFORMATION 2 .1    .3  21  S        02        SRP requirement.
* REV.                                      ADDITIONAL INFORMATION 2 .1    .3  21  S        02        SRP requirement.
Unit 2 Action: Update FSAR for present and projected population over the lifetime of the plant.
Unit 2 Action: Update FSAR for present and projected population over the lifetime of the plant.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Status in SSER21 is Open (NRR).
Status in SSER21 is Open (NRR).
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Part of this amendment revised population information in Section 2.1.3.
Part of this amendment revised population information in Section 2.1.3.
2 .1    .4  21  S        02        "CONCLUSIONS" left open until all items in subsection are closed.
2 .1    .4  21  S        02        "CONCLUSIONS" left open until all items in subsection are closed.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Status in SSER21 is Open (NRR).
Status in SSER21 is Open (NRR).
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2 .2    .1  21  S        02        SRP requirement.
2 .2    .1  21  S        02        SRP requirement.
Unit 2 Action: Update FSAR for potential external hazards and hazardous materials.
Unit 2 Action: Update FSAR for potential external hazards and hazardous materials.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Status in SSER21 is Open (NRR).
Status in SSER21 is Open (NRR).
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* REV.                                      ADDITIONAL INFORMATION 2 .2    .2  21  S        02        SRP requirement.
* REV.                                      ADDITIONAL INFORMATION 2 .2    .2  21  S        02        SRP requirement.
Unit 2 Action: Update FSAR for projected annual number of aircraft flights.
Unit 2 Action: Update FSAR for projected annual number of aircraft flights.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Status in SSER21 is Open (NRR).
Status in SSER21 is Open (NRR).
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Part of this amendment revised information concerning airports and numbers of aircraft flights in 2.2.2.5.
Part of this amendment revised information concerning airports and numbers of aircraft flights in 2.2.2.5.
2 .2    .3  21  S        02        "CONCLUSIONS" left open until all items in subsection are closed.
2 .2    .3  21  S        02        "CONCLUSIONS" left open until all items in subsection are closed.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Status in SSER21 is Open (NRR).
Status in SSER21 is Open (NRR).
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2 .4    .7      0    C                  Approved for both units in SER.
2 .4    .7      0    C                  Approved for both units in SER.
2 .4    .8    21  O          02        CONFIRMATORY ISSUE for design basis groundwater level for ERCW pipeline Amendment 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.
2 .4    .8    21  O          02        CONFIRMATORY ISSUE for design basis groundwater level for ERCW pipeline Amendment 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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Status in SSER21 is Open (NRR).
Status in SSER21 is Open (NRR).
2 .4    .9    21  O          02        SRP requirement.
2 .4    .9    21  O          02        SRP requirement.
Unit 2 Action: Update FSAR for present and projected use of local and regional groundwater.
Unit 2 Action: Update FSAR for present and projected use of local and regional groundwater.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
2 .4    . 10  21    S        02        Staff found flood emergency plan and draft Technical Specifications acceptable in original 1982 SER.
2 .4    . 10  21    S        02        Staff found flood emergency plan and draft Technical Specifications acceptable in original 1982 SER.
Unit 2 Action: Address in Technical Specifications as appropriate.
Unit 2 Action: Address in Technical Specifications as appropriate.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Status in SSER21 is Open (Inspection).
Status in SSER21 is Open (Inspection).
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2 .5    .3    0  C                  Approved for both units in SER.
2 .5    .3    0  C                  Approved for both units in SER.
2 .5    .4  11  C        03        CONFIRMATORY ISSUE for design differential settlement of piping and electrical components Analysis was presented to staff in September 1983. Staff found analysis and results acceptable. Staff closed issue in SSER3.
2 .5    .4  11  C        03        CONFIRMATORY ISSUE for design differential settlement of piping and electrical components Analysis 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 analysis of sheetpile walls Staff performed audit in September 1982, and determined TVA had used reasonable assumptions. Staff closed issue in SSER3.
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CONFIRMATORY ISSUE for material and geometric damping in soil-structure interaction (SSI) analysis 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) analysis Staff performed audit in September 1982, and determined TVA had used reasonable assumptions. Staff closed issue in SSER3.
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OUTSTANDING ISSUE (1) on liquefaction beneath ERCW pipelines and Class 1E electrical conduit.
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.
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.
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FSAR amendment 54-63 was reviewed in SSER9. NRC determined that the conclusions previously issued in the SER and SSER3 remained unchanged.
FSAR amendment 54-63 was reviewed in SSER9. NRC determined that the conclusions previously issued in the SER and SSER3 remained unchanged.
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The Special Program (SP) for Soil Liquefaction was reviewed in SSER11.
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.
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.
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REVISION 03 UPDATE:
REVISION 03 UPDATE:
NRC IR 50-391/2009-605 noted that the Soil Liquefaction SP was closed for Unit 2.
NRC IR 50-391/2009-605 noted that the Soil Liquefaction SP was closed for Unit 2.
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Mechanical System Components", and found the table acceptable.
Mechanical System Components", and found the table acceptable.
3 .2    .1    8  C        01        CONFIRMATORY ISSUE for seismic classification of structures, systems, and components important to safety The 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.
3 .2    .1    8  C        01        CONFIRMATORY ISSUE for seismic classification of structures, systems, and components important to safety The 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 1 Staff verified that required portion of ERCW had been upgraded or replaced satisfactorily in SSER5 and closed this issue.
CONFIRMATORY ISSUE for ERCW upgrade to seismic category 1 Staff 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.
In SSER6, the staff addressed and resolved an issue on Category I boundary.
                                        --------------------
OUTSTANDING ISSUE involving seismic classification of cable trays and conduits In 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.
OUTSTANDING ISSUE involving seismic classification of cable trays and conduits In 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.
Page 6 of 88          * = See last page for status code definition.
Page 6 of 88          * = See last page for status code definition.
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SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                      ADDITIONAL INFORMATION 3 .2    .2  21  CI        02        Section 3.2.2 of SSER3 discusses confirmatory issues for seismic classification and upgrade of ERCW that are already included in 3.2.1.
* REV.                                      ADDITIONAL INFORMATION 3 .2    .2  21  CI        02        Section 3.2.2 of SSER3 discusses confirmatory issues for seismic classification and upgrade of ERCW that are already included in 3.2.1.
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Staff accepted implementation of Heat Code Traceability CAP for Unit 1 in SSER7.
Staff accepted implementation of Heat Code Traceability CAP for Unit 1 in SSER7.
Unit 2 Action: Complete CAP using Unit 1 approach.
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.
Staff reviewed updated information in Amendment 68 on use of codes and standards in SSER9 and stated that prior conclusions were unchanged.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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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.
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.
Unit 2 Action: Complete Special Program using the Unit 1 approach.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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In SSER21, the MELB Special Program was resolved. Completion of MELB SP is tracked under 23.3.8.
In SSER21, the MELB Special Program was resolved. Completion of MELB SP is tracked under 23.3.8.
3 .6    .1  21  O        02        OUTSTANDING ISSUE involving main steam line break (MSLB) outside containment In 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.
3 .6    .1  21  O        02        OUTSTANDING ISSUE involving main steam line break (MSLB) outside containment In 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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Status in SSER21 is Open (NRR).
Status in SSER21 is Open (NRR).
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* REV.                                      ADDITIONAL INFORMATION 3 .7    .0  21  C        03        The 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.
* REV.                                      ADDITIONAL INFORMATION 3 .7    .0  21  C        03        The 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.
Unit 2 Action: Complete CAP using the Unit 1 approach.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
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 .
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.
In SSER21, the Seismic Analysis CAP was resolved. Completion of the Seismic Analysis CAP is tracked under 23.2.16.
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REVISION 03 UPDATE:
REVISION 03 UPDATE:
NRC IR 50-391/2010-602 noted that the Seismic Analysis CAP was closed for Unit 2.
NRC IR 50-391/2010-602 noted that the Seismic Analysis CAP was closed for Unit 2.
3 .7    .1  21  C        03        OUTSTANDING ISSUE involving update of FSAR for seismic design issues The 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.
3 .7    .1  21  C        03        OUTSTANDING ISSUE involving update of FSAR for seismic design issues The 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.
Unit 2 Action: Complete CAP using Unit 1 approach.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
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 .
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.
In SSER21, the Seismic Analysis CAP was resolved. Completion of the Seismic Analysis CAP is tracked under 23.2.16.
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REVISION 03 UPDATE:
REVISION 03 UPDATE:
NRC IR 50-391/2010-602 noted that the Seismic Analysis CAP was closed for Unit 2.
NRC IR 50-391/2010-602 noted that the Seismic Analysis CAP was closed for Unit 2.
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SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                      ADDITIONAL INFORMATION 3 .7    .2  21  C        03        3.7.2.1.2: OUTSTANDING ISSUE involving mass eccentricity In 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.
* REV.                                      ADDITIONAL INFORMATION 3 .7    .2  21  C        03        3.7.2.1.2: OUTSTANDING ISSUE involving mass eccentricity In 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 response The staff considered this item (opened in SSER6) resolved in SSER11 based on audits and inspections since SSER6.
3.7.2.1.2: OUTSTANDING ISSUE involving comparison of Set A vs. Set B response The 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.
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.
In SSER16, the staff discussed the review and acceptability of the NSSS-ICS modeling for seismic analysis.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
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 .
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.
In SSER21, the Seismic Analysis CAP was resolved. Completion of the Seismic Analysis CAP is tracked under 23.2.16.
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REVISION 03 UPDATE:
REVISION 03 UPDATE:
NRC IR 50-391/2010-602 noted that the Seismic Analysis CAP was closed for Unit 2.
NRC IR 50-391/2010-602 noted that the Seismic Analysis CAP was closed for Unit 2.
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SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                      ADDITIONAL INFORMATION 3 .7    .3  21  C        03        OUTSTANDING ISSUE involving number of peak cycles to be used for OBE In 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.
* REV.                                      ADDITIONAL INFORMATION 3 .7    .3  21  C        03        OUTSTANDING ISSUE involving number of peak cycles to be used for OBE In 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 case In 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.
OUTSTANDING ISSUE involving use of code cases, damping factors for conduit and use of worst case, critical case and bounding case In 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.
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.
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.
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.
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.
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 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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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* REV.                                      ADDITIONAL INFORMATION In SSER21, the Seismic Analysis CAP was resolved. Completion of the Seismic Analysis CAP is tracked under 23.2.16.
* REV.                                      ADDITIONAL INFORMATION 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.
In SSER21, the Conduit Supports CAP was resolved. Completion of the Electrical Conduit and Conduit Supports CAP is tracked under 23.2.16.
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REVISION 03 UPDATE:
REVISION 03 UPDATE:
NRC IR 50-391/2010-602 noted that the Seismic Analysis CAP was closed for Unit 2.
NRC IR 50-391/2010-602 noted that the Seismic Analysis CAP was closed for Unit 2.
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3 .8    .0  21  O        02        OUTSTANDING ISSUE involving load combinations and stress allowables In 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.
3 .8    .0  21  O        02        OUTSTANDING ISSUE involving load combinations and stress allowables In 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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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3 .8    .2    7  C        01        The 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 3 .8    .3  21  O        02        The 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.
3 .8    .2    7  C        01        The 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 3 .8    .3  21  O        02        The 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.
The staff reviewed allowable stresses for Category I structural steel and found them acceptable for both units in SSER16.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                ADDITIONAL INFORMATION 3 .9    .3  15  S        02        3.9.3.1: OUTSTANDING ISSUE involving use of experience data to qualify category I(L) piping The 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.
* REV.                                ADDITIONAL INFORMATION 3 .9    .3  15  S        02        3.9.3.1: OUTSTANDING ISSUE involving use of experience data to qualify category I(L) piping The 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)
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.
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 valves The 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.
3.9.3.3: OUTSTANDING ISSUE involving operating characteristics of main steam safety valves The 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.
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 loads The 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: CONFIRMATORY ISSUE involving baseplate flexibility and its effect on anchor bolt loads The 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 supports The 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.
3.9.3.4: OUTSTANDING ISSUE involving stiffness and deflection limits for seismic Category I pipe supports The 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.
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Page 14 of 88        * = See last page for status code definition.
Page 14 of 88        * = See last page for status code definition.


SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                      ADDITIONAL INFORMATION 3.9.3.4: OUTSTANDING ISSUE, staff was awaiting TVA concurrence on their position with respect to margin for critical buckling of pipe supports In 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.
* REV.                                      ADDITIONAL INFORMATION 3.9.3.4: OUTSTANDING ISSUE, staff was awaiting TVA concurrence on their position with respect to margin for critical buckling of pipe supports In 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.
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.
In SSER15, the staff found the response to NUREG-0737, Item II.D.1, "Performance Testing of Relief and Safety Valves," acceptable.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
TVA determined that the Unit 1 MSSV analysis was applicable to Unit 2.
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.
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.
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.
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SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                      ADDITIONAL INFORMATION 3 .9    .6  20    S        02        LICENSE CONDITION on inservice testing of pumps and valves The 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.
* REV.                                      ADDITIONAL INFORMATION 3 .9    .6  20    S        02        LICENSE CONDITION on inservice testing of pumps and valves The 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.
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.
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.
Unit 2 Action: Submit Technical Specifications.
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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 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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.
Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.
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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.
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.
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 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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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3 . 11 . 0    21  CI        02        OUTSTANDING ISSUE - TVA program not submitted at time of SER The EQ program was submitted after issuance of the SER. It was reviewed and found acceptable in SSER15.
3 . 11 . 0    21  CI        02        OUTSTANDING ISSUE - TVA program not submitted at time of SER The EQ program was submitted after issuance of the SER. It was reviewed and found acceptable in SSER15.
Unit 2 Action: Complete EQ Special Program.
Unit 2 Action: Complete EQ Special Program.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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* REV.                                      ADDITIONAL INFORMATION 4 .2    .1  13  S        02        In 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.
* REV.                                      ADDITIONAL INFORMATION 4 .2    .1  13  S        02        In 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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
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) 4 .2    .2    2  S        02        CONFIRMATORY ISSUE on cladding collapse calculations The 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.
FSAR Chapter 4 was updated to address the application of the second generation Robust Fuel Assembly design (RFA-2) 4 .2    .2    2  S        02        CONFIRMATORY ISSUE on cladding collapse calculations The 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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
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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.
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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
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* REV.                                      ADDITIONAL INFORMATION 4 .2    .5    0  S        02        "FUEL DESIGN CONCLUSIONS" left open until all items in subsection are closed.
* REV.                                      ADDITIONAL INFORMATION 4 .2    .5    0  S        02        "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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
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4 .3    .1  13  S        02        In SSER13, NRC reviewed the V5H fuel design and found use of V5H fuel acceptable.
4 .3    .1  13  S        02        In 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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
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4 .3    .2  15  S        02        In SSER13, NRC reviewed the V5H fuel design and found use of V5H fuel acceptable.
4 .3    .2  15  S        02        In 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.
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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
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* REV.                                      ADDITIONAL INFORMATION 4 .3    .3  13  S        02        In SSER13, NRC reviewed the V5H fuel design and found use of V5H fuel acceptable.
* REV.                                      ADDITIONAL INFORMATION 4 .3    .3  13  S        02        In 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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
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4 .3    .4  13  S        02        In SSER13, NRC reviewed the V5H fuel design and found use of V5H fuel acceptable.
4 .3    .4  13  S        02        In 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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
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4 .4    .1    0  C                  Approved for both units in SER.
4 .4    .1    0  C                  Approved for both units in SER.
4 .4    .2  12  S        02        In 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.
4 .4    .2  12  S        02        In 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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
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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.
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.
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 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.
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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.
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.
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.
FSAR Chapter 4 was updated to address the application of RFA-2 fuel.
4 .4    .4  13  S        02        In SSER13, NRC reviewed TVA's responses to a request for additional information concerning fuel rod bowing and crud buildup for WBN Unit 1.
4 .4    .4  13  S        02        In 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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
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4 .4    .7    0  S        02        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).
4 .4    .7    0  S        02        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.
Unit 2 Action: Confirm Technical Specifications prohibit (N-1) Loop Operation.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
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5 .1    .0    6  S        02        The 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).
5 .1    .0    6  S        02        The 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.
Unit 2 Action: Provide additional information for NRC review per 7.2.1.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.
TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.
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5 .2    .1    0  C                  Approved for both units in SER.
5 .2    .1    0  C                  Approved for both units in SER.
5 .2    .2  15  C        01        OUTSTANDING ISSUE on staff review of sensitivity study of required safety valve flow rate versus trip parameter TVA letter dated April 18, 1983, provided the safety valve sizing information and information on differences with the reference plant. Staff closed issue in SSER2.
5 .2    .2  15  C        01        OUTSTANDING ISSUE on staff review of sensitivity study of required safety valve flow rate versus trip parameter TVA 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.
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.
5 .2    .3    0  C                  Approved for both units in SER.
5 .2    .3    0  C                  Approved for both units in SER.
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* REV.                                      ADDITIONAL INFORMATION 5 .2    .4  16  O        03        LICENSE CONDITION - Inservice inspection (ISI) program The 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.
* REV.                                      ADDITIONAL INFORMATION 5 .2    .4  16  O        03        LICENSE CONDITION - Inservice inspection (ISI) program The 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.
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.
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.
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.
Unit 2 Action: Submit Unit 2 PSI program.
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REVISION 03 UPDATE:
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.
Preservice Inspection Plan, Program No. WBN-2 PSI, Revision 3 was submitted to the NRC on June 17, 2010 (ADAMS Accession No. ML101680561).
ML101680561).
5 .2    .5  21  O        02        In 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.
5 .2    .5  21  O        02        In 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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
In SSER21 the status is Open (NRR).
In SSER21 the status is Open (NRR).
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* REV.                                      ADDITIONAL INFORMATION 5 .3    .1  21  O        02        The 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.
* REV.                                      ADDITIONAL INFORMATION 5 .3    .1  21  O        02        The 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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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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.
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.
Unit 2 action: Submit P-T limits.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
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5 .3    .3    0  S        02        OUTSTANDING ISSUE for staff to complete evaluation of Unit 2 after receipt of P-T limits 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.
5 .3    .3    0  S        02        OUTSTANDING ISSUE for staff to complete evaluation of Unit 2 after receipt of P-T limits 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.
Unit 2 action: Submit P-T limits.
Unit 2 action: Submit P-T limits.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
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The staff verified that the alarm had been installed in SSER5, resolving the confirmatory issue.
The staff verified that the alarm had been installed in SSER5, resolving the confirmatory issue.
Unit 2 action: Verify alarm installation.
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 refueling Branch 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.
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 refueling Branch 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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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* REV.                                      ADDITIONAL INFORMATION 5 .4    .5  21  CI        02        LICENSE CONDITION - NUREG-0737, II.B.1, "Reactor Coolant System Vents" - In the original SER, the NRC found TVAs 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.
* REV.                                      ADDITIONAL INFORMATION 5 .4    .5  21  CI        02        LICENSE CONDITION - NUREG-0737, II.B.1, "Reactor Coolant System Vents" - In the original SER, the NRC found TVAs 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.
Unit 2 Action: Verify installation of reactor coolant vents.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (Inspection).
The status in SSER21 is Open (Inspection).
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SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                      ADDITIONAL INFORMATION 6 .2    .1  21  S        02        6.2.1.1: CONFIRMATORY ISSUE involves reviewing analysis that ensures that containment external pressure will not exceed design value of 2.0 psi In 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.
* REV.                                      ADDITIONAL INFORMATION 6 .2    .1  21  S        02        6.2.1.1: CONFIRMATORY ISSUE involves reviewing analysis that ensures that containment external pressure will not exceed design value of 2.0 psi In 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.
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).
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).
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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.
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.
Unit 2 Action: Review Unit 2 Technical Specifications with respect to divider barrier seal surveillance program.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
                                        ----------
Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.
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.
TS 3.6.13 provides the Limiting Condition for Operation for Divider Barrier Integrity.
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SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                      ADDITIONAL INFORMATION 6 .2    .2  21  CI        02        In SSER7, the staff determined that hot standby was an acceptable mode following a main steamline break and the containment cooling system modifications were acceptable.
* REV.                                      ADDITIONAL INFORMATION 6 .2    .2  21  CI        02        In SSER7, the staff determined that hot standby was an acceptable mode following a main steamline break and the containment cooling system modifications were acceptable.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
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 .
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.
In SSER21, the Containment Cooling SP was resolved. Completion of the Containment Cooling SP is tracked under 23.3.2.
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* REV.                                      ADDITIONAL INFORMATION 6 .2    .4  21  S        02        CONFIRMATORY ISSUE to install safety grade isolation valves on 1 chemical feed lines joining feedwater lines to main steam line.
* REV.                                      ADDITIONAL INFORMATION 6 .2    .4  21  S        02        CONFIRMATORY 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.
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 Operation LICENSE CONDITION - Containment isolation dependability In the original 1982 SER, NRC concluded that WBN met all the requirements of NUREG-0737, item II.E.4.2 except subsection (6) concerning containment purging during normal operation. In SSER3, the outstanding issue was closed and the LICENSE CONDITION was left open. NRC completed the review and issued a 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)
OUTSTANDING ISSUE for NRC to complete review of information provided by TVA to address Containment Purging During Normal Plant Operation LICENSE CONDITION - Containment isolation dependability In the original 1982 SER, NRC concluded that WBN met all the requirements of NUREG-0737, item II.E.4.2 except subsection (6) concerning containment purging during normal operation. In SSER3, the outstanding issue was closed and the LICENSE CONDITION was left open. NRC completed the review and issued a 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.
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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (Inspection).
The status in SSER21 is Open (Inspection).
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* REV.                                      ADDITIONAL INFORMATION 6 .2    .5  21  S        04        OUTSTANDING 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 LOCA In 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.
* REV.                                      ADDITIONAL INFORMATION 6 .2    .5  21  S        04        OUTSTANDING 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 LOCA In 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.
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 instrumentation II.F.1 - containment hydrogen In SSER5, NRC closed the LICENSE CONDITION for Unit 1 only (IR 390/84-85).
LICENSE CONDITION - (6f) Accident monitoring instrumentation II.F.1 - containment hydrogen In 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.
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)}.
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)}.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
This amendment deleted the hydrogen recombiners from the Unit 2 FSAR.
This amendment deleted the hydrogen recombiners from the Unit 2 FSAR.
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REVISION 04 UPDATE:
REVISION 04 UPDATE:
EDCR 52329 was initiated to abandon in place Unit 2 hydrogen recombiners.
EDCR 52329 was initiated to abandon in place Unit 2 hydrogen recombiners.
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* REV.                                      ADDITIONAL INFORMATION 6 .2    .6  21  O        02        In 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.
* REV.                                      ADDITIONAL INFORMATION 6 .2    .6  21  O        02        In 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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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6 .3    .0    0  C        01        Approved for both units in SER.
6 .3    .0    0  C        01        Approved for both units in SER.
6 .3    .1  11  S        02        OUTSTANDING ISSUE - involving removal of upper head injection system The 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.
6 .3    .1  11  S        02        OUTSTANDING ISSUE - involving removal of upper head injection system The 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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
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SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                      ADDITIONAL INFORMATION 6 .3    .2  5    S        02        In 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.
* REV.                                      ADDITIONAL INFORMATION 6 .3    .2  5    S        02        In 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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
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.
This amendment revised the FSAR to address the application of RFA-2 fuel.
6 .3    .3  9    S        02        OUTSTANDING ISSUE - involving containment sump screen design In 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.
6 .3    .3  9    S        02        OUTSTANDING ISSUE - involving containment sump screen design In 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-LOCA In 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 staffs initial conclusion that the design to provide protection against sump debris was acceptable.
CONFIRMATORY ISSUE - provide a detailed survey of insulation material that could be debris post-LOCA In 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 staffs 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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
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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.
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.
See 18.1.0 also.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
6 .5    .0    0  C                  Approved for both units in SER.
6 .5    .0    0  C                  Approved for both units in SER.
6 .5    .1  21  O        02        In SSER5, the staff found the Reactor Building Purge Ventilation System acceptable.
6 .5    .1  21  O        02        In SSER5, the staff found the Reactor Building Purge Ventilation System acceptable.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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* REV.                                      ADDITIONAL INFORMATION 6 .6    .0  15  O        03        OUTSTANDING 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 justification NRC 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.
* REV.                                      ADDITIONAL INFORMATION 6 .6    .0  15  O        03        OUTSTANDING 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 justification NRC 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.
Unit 2 action: Submit Unit 2 PSI program.
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REVISION 03 UPDATE:
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.
Preservice Inspection Plan, Program No. WBN-2 PSI, Revision 3 was submitted to the NRC on June 17, 2010 (ADAMS Accession No. ML101680561).
ML101680561).
7 .0    .0    0  C                  Approved for both units in SER.
7 .0    .0    0  C                  Approved for both units in SER.
7 .1    .0    0  C                  Approved for both units in SER.
7 .1    .0    0  C                  Approved for both units in SER.
7 .1    .1  16  S        02        In 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.
7 .1    .1  16  S        02        In 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.
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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.
TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.
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Staff requested discussion of methodology for determining, setting, and evaluating as-found setpoints for drift susceptible instruments.
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).
Unit 2 action: Resolve this issue using the BFN TS-453 precedent (see NRC ML061680008).
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision B of the Unit 2 Technical Specifications (TS) and TS Bases was submitted on February 2, 2010.
Developmental Revision B of the Unit 2 Technical Specifications (TS) and TS Bases was submitted on February 2, 2010.
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7 .2    .1  15  S        02        In 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.
7 .2    .1  15  S        02        In 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.
Unit 2 Action: Provide the additional information for NRC review.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.
TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.
Line 845: Line 674:
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.
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.
Unit 2 Action: Update accident calculation.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
Line 852: Line 679:
Unit 2 Action: Provide the additional information for NRC review.
Unit 2 Action: Provide the additional information for NRC review.
                                         "CONCLUSIONS" left open until all actions in subsection are closed.
                                         "CONCLUSIONS" left open until all actions in subsection are closed.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.
TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.
7 .3    .0  13  S        02        In 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.
7 .3    .0  13  S        02        In 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.
Unit 2 Action: Provide the additional information for NRC review.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.
TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.
Line 867: Line 690:
* REV.                                      ADDITIONAL INFORMATION 7 .3    .1  14  S        02        In 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.
* REV.                                      ADDITIONAL INFORMATION 7 .3    .1  14  S        02        In 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.
Unit 2 Action: Provide the additional information for NRC review.
                                        --------------------
In SSER14, NRC reviewed TVA's FSAR amendment 81 section 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.
In SSER14, NRC reviewed TVA's FSAR amendment 81 section 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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.
TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.
Line 878: Line 698:
7 .3    .5  21  CI        02        CONFIRMATORY 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 reset In 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.
7 .3    .5  21  CI        02        CONFIRMATORY 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 reset In 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.
Unit 2 Action: Perform verification during preoperational testing.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (Inspection).
The status in SSER21 is Open (Inspection).
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Unit 2 Action: Provide the additional information for NRC review.
Unit 2 Action: Provide the additional information for NRC review.
                                         "CONCLUSIONS" left open until all actions in subsection are closed.
                                         "CONCLUSIONS" left open until all actions in subsection are closed.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.
TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.
Line 896: Line 712:
7 .4    .2  21  O        02        By 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.
7 .4    .2  21  O        02        By 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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
Line 910: Line 724:
Unit 2 Actions: Install Noble gas, Iodine / particulate sampling, and Containment High Range Monitors.
Unit 2 Actions: Install Noble gas, Iodine / particulate sampling, and Containment High Range Monitors.
CI in NRC May 28, 2008, letter.
CI in NRC May 28, 2008, letter.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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TVA responded to the Bulletin on March 1, 1982. Reviewed in 7.5.3 of the original 1982 SER.
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.
Unit 2 Action: Issue appropriate emergency procedures.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (Inspection).
The status in SSER21 is Open (Inspection).
7 .5    .4  21  CI        02        "CONCLUSIONS" left CI until all items in subsection are closed.
7 .5    .4  21  CI        02        "CONCLUSIONS" left CI until all items in subsection are closed.
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                                        ----------------------------------------------------------------------------------------------------
REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (Inspection).
The status in SSER21 is Open (Inspection).
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* REV.                                      ADDITIONAL INFORMATION 7 .7    .8  21  S        02        ATWS 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.
* REV.                                      ADDITIONAL INFORMATION 7 .7    .8  21  S        02        ATWS 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.
Unit 2 Action: Address in Technical Specifications as appropriate.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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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.
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.
Unit 2 Action: Verify installation of the acoustic monitoring system to PORV to indicate position. CI in NRC May 28, 2008, letter.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
7 .8    .2  21  CI        02        NUREG-0737, II.E.1.2, "Auxiliary Feedwater System Initiation and Flow Indication" Unit 2 Action: Complete procedures and qualification testing.
7 .8    .2  21  CI        02        NUREG-0737, II.E.1.2, "Auxiliary Feedwater System Initiation and Flow Indication" Unit 2 Action: Complete procedures and qualification testing.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (Inspection).
The status in SSER21 is Open (Inspection).
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* REV.                                      ADDITIONAL INFORMATION 7 .8    .3  21  O          02        NUREG-0737, II.K.3.9, Proportional Integral Derivative Controller Modification - Reviewed in original 1982 SER.
* REV.                                      ADDITIONAL INFORMATION 7 .8    .3  21  O          02        NUREG-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.
Unit 2 Action: Set the derivative time constant to zero.
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                                          ----------------------------------------------------------------------------------------------------
REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (Inspection).
The status in SSER21 is Open (Inspection).
7 .8    .4  21    S        02        NUREG-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.
7 .8    .4  21    S        02        NUREG-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.
Unit 2 Action: Unit 2 Technical Specifications and surveillance procedures will address this issue.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (Inspection).
The status in SSER21 is Open (Inspection).
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SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                      ADDITIONAL INFORMATION 8 .2    .2  21  O        02        8.2.2.1 CONFIRMATORY ISSUE - document additional information in FSAR on control power supplies and distribution system for the Watts Bar Hydro Plant Switchyard In 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.
* REV.                                      ADDITIONAL INFORMATION 8 .2    .2  21  O        02        8.2.2.1 CONFIRMATORY ISSUE - document additional information in FSAR on control power supplies and distribution system for the Watts Bar Hydro Plant Switchyard In 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.
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.
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 Contingencies 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 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.
8.2.2.3 Compliance with GDC 17 for the Duration of the Offsite System Contingencies 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 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 Trip 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 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.
8.2.2.4 Minimizing the Probability of a Two-Unit Trip Following a One-Unit Trip 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 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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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* REV.                                ADDITIONAL INFORMATION 8 .3    .1  20  S        02        8.3 Fifth Diesel Generator In 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.
* REV.                                ADDITIONAL INFORMATION 8 .3    .1  20  S        02        8.3 Fifth Diesel Generator In 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.
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 FSAR In 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.
8.3.1.1: CONFIRMATORY ISSUE - incorporate new design that provides dedicated transformer for each preferred offsite circuit in FSAR In 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 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.
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.
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 report In 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: CONFIRMATORY ISSUE - provide diesel generator reliability qualification test report In 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 temperature In the original 1982 SER, NRC required that the capability of the DGs to start at normal temperature be demonstrated. TVAs August 31, 1983, letter confirmed tests had been performed on a DG identical to those at WBN. In SSER2, NRC closed the issue.
8.3.1.6: LICENSE CONDITION (12) - Diesel generator reliability qualification testing at normal operating temperature In the original 1982 SER, NRC required that the capability of the DGs to start at normal temperature be demonstrated. TVAs 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 By letter dated June 20, 1991, for both units, NRC requested additional Page 46 of 88        * = See last page for status code definition.
8.3.1.7 Possible Interconnection Between Redundant Divisions Through Normal and Alternate Power to the Battery Charger By letter dated June 20, 1991, for both units, NRC requested additional Page 46 of 88        * = See last page for status code definition.


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* REV.                                      ADDITIONAL INFORMATION 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.
* REV.                                      ADDITIONAL INFORMATION 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.
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.
8.3.1.10 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.
8.3.1.11 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 Testing 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 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.
8.3.1.12 The Capability and Independence of Offsite and Onsite Sources When Paralleling During Testing 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 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.
8.3.1.13 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.
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.
Unit 2 Action: Resolve the SP for WBN Unit 2 with the Unit 1 approach.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                ADDITIONAL INFORMATION
* REV.                                ADDITIONAL INFORMATION Revised "SSER18" to "SSER19" item 8.3 above to fix typographical error in Regulatory Framework.
                                        ----------
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.
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.
8.3.1.2: 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.
8.3.1.7: 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.
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.
In SSER21 the Containment Cooling SP was resolved. Completion of the Master Fuse List SP is tracked under 23.3.5.
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SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                ADDITIONAL INFORMATION 8 .3    .2  14  C        01        8.3.2.2: LICENSE CONDITION - DC monitoring and annunciation system In 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.
* REV.                                ADDITIONAL INFORMATION 8 .3    .2  14  C        01        8.3.2.2: LICENSE CONDITION - DC monitoring and annunciation system In 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 FSAR In 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.
8.3.2.4: CONFIRMATORY ISSUE - include diesel generator design analysis in FSAR In 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 Inverters 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 Inverters 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.
                                        ---------------------
8.3.2.5.1 Transfer of Loads Between Power Supplies Associated with the Same Load Group but Different Units 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 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.
8.3.2.5.1 Transfer of Loads Between Power Supplies Associated with the Same Load Group but Different Units 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 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.
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.
8.3.2.7 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 Power 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 Power 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.
Page 49 of 88        * = See last page for status code definition.
Page 49 of 88        * = See last page for status code definition.
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SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                ADDITIONAL INFORMATION 8 .3    .3  21  S        02        8.3.3.1.1: CONFIRMATORY ISSUE involving submergence of electrical equipment as result of a LOCA In 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.
* REV.                                ADDITIONAL INFORMATION 8 .3    .3  21  S        02        8.3.3.1.1: CONFIRMATORY ISSUE involving submergence of electrical equipment as result of a LOCA In 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.
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8.3.3.1.3 Failure Analysis of Circuits Associated with Cables and Cable Splices Unqualified for Submergence 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 reviewed the submergence calculation and closed the issue.
8.3.3.1.3 Failure Analysis of Circuits Associated with Cables and Cable Splices Unqualified for Submergence 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 reviewed the submergence calculation and closed the issue.
Unit 2 Action: Revise calculation for WBN Unit 2.
Unit 2 Action: Revise calculation for WBN Unit 2.
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8.3.3.1.2: CONFIRMATORY ISSUE - verify design for bypass of thermal overload protective device In 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.
8.3.3.1.2: CONFIRMATORY ISSUE - verify design for bypass of thermal overload protective device In 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.
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8.3.3.1.4 Use of Waterproof Splices in Potentially Submersible Sections of Underground Duct Runs 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 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.
8.3.3.1.4 Use of Waterproof Splices in Potentially Submersible Sections of Underground Duct Runs 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 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.
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8.3.3.1.5 Dow Corning RTV-3140 Used to Repair Damaged Kapton Insulated Conductors In 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.
8.3.3.1.5 Dow Corning RTV-3140 Used to Repair Damaged Kapton Insulated Conductors In 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.
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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.
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.
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SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                ADDITIONAL INFORMATION 8.3.3.2: CONFIRMATORY ISSUE - revise FSAR to reflect requirements of shared safety systems In 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.
* REV.                                ADDITIONAL INFORMATION 8.3.3.2: CONFIRMATORY ISSUE - revise FSAR to reflect requirements of shared safety systems In 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.
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8.3.3.2.2 Sharing of AC Distribution Systems and Standby Power Supplies Between Units 1 and 2 In 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.
8.3.3.2.2 Sharing of AC Distribution Systems and Standby Power Supplies Between Units 1 and 2 In 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.
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8.3.3.2.3: CONFIRMATORY ISSUE for design of sharing raceway systems between units In the original SER, NRC indicated that the design for sharing of raceway systems between units would be verified during the electrical drawing review.
8.3.3.2.3: CONFIRMATORY ISSUE for design of sharing raceway systems between units In 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.
The staff confirmed that cable routing was in accordance with accepted separation criteria and closed the issue in SSER2.
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8.3.3.2.4: LICENSE CONDITION - Possible sharing of DC control power to AC switchgear In 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.
8.3.3.2.4: LICENSE CONDITION - Possible sharing of DC control power to AC switchgear In 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.
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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 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.
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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.
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.
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8.3.3.3 Physical Independence (Compliance with GDC 17)
8.3.3.3 Physical Independence (Compliance with GDC 17)
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* REV.                                ADDITIONAL INFORMATION 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.
* REV.                                ADDITIONAL INFORMATION 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.
Unit 2 Action: Incorporate testing requirements into the Unit 2 TRM.
                                        ---------------------
8.3.3.3 Physical Independence (Compliance with GDC 17)
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.
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.
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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.
8.3.3.5.1 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.
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8.3.3.5.2: CONFIRMATORY ISSUE - incorporate commitment to test only one of four diesel generators at one time In 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.
8.3.3.5.2: CONFIRMATORY ISSUE - incorporate commitment to test only one of four diesel generators at one time In 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.
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8.3.3.5.3 Time Constraints for Stability of EDG During No-Load Startup Testing In SSER16, NRC reviewed and approved changes to the no load emergency diesel generator testing surveillance requirements.
8.3.3.5.3 Time Constraints for Stability of EDG During No-Load Startup Testing In 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.
Unit 2 Action: Incorporate into WBN Unit 2 TS surveillances.
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8.3.3.6: CONFIRMATORY ISSUE involving evaluation of penetrations ability to withstand failure of overcurrent protection device In the original 1982 SER, staff required a reevaluation of the penetrations 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: CONFIRMATORY ISSUE involving evaluation of penetrations ability to withstand failure of overcurrent protection device In the original 1982 SER, staff required a reevaluation of the penetrations 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.
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8.3.3.6: LICENSE CONDITION - Testing of reactor coolant pump breakers Page 52 of 88        * = See last page for status code definition.
8.3.3.6: LICENSE CONDITION - Testing of reactor coolant pump breakers Page 52 of 88        * = See last page for status code definition.


SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                      ADDITIONAL INFORMATION In 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.
* REV.                                      ADDITIONAL INFORMATION In 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.
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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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
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.
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.
8.3.3.3: TRM TR 3.8.1 specifies testing of circuit breakers that are used as isolation devices protecting 1E busses from non-qualified loads.
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In SSER15, the staff reviewed TVA's proposed resolution of the Boraflex degradation issue and found it acceptable.
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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
9 .1    .3  21  O          02        In SSER11, the staff reviewed TVA's revised commitment regarding testing of spent fuel pool cooling pumps and found it acceptable.
9 .1    .3  21  O          02        In 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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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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.
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).
Unit 2 Action: Eliminate requirement for Post-Accident Sampling System in Technical Specifications (Identified as CT in NRC letter dated May 28, 2008).
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
                                        ----------
Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.
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."
Rev. 0 of the Unit 1 TS contained 5.7.2.6, "Post Accident Sampling."
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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 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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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9 .5    .2  21  O        02        LICENSE CONDITION - Performance testing of communications system The staff resolved this license condition in SSER5 based on TVAs letter of March 18, 1985 for both units, which described its testing of communications systems.
9 .5    .2  21  O        02        LICENSE CONDITION - Performance testing of communications system The staff resolved this license condition in SSER5 based on TVAs 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.
Unit 2 Action: Perform testing of communication systems on Unit 2.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                  ADDITIONAL INFORMATION 9 .5    .4  12  C        01        9.5.4.1: 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 load In 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.
* REV.                                  ADDITIONAL INFORMATION 9 .5    .4  12  C        01        9.5.4.1: 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 load In 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 with NUREG/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: LICENSE CONDITION - Diesel Generator reliability The staff verified that the modifications necessary to comply with NUREG/CR-0660 had been completed and, as stated above, requirements had been incorporated into operating procedures. Thus, this license condition was resolved in SSER5.
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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.
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.
In SSER5, the staff resolved the issue of the completeness of its review of the emergency diesel engine lubrication oil system.
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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 3 The 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).
9.5.4.1: 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 3 The 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).
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9.5.4.2: CONFIRMATORY ISSUE - provide missile protection for fuel oil storage tank vent lines The staff found TVAs 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.
9.5.4.2: CONFIRMATORY ISSUE - provide missile protection for fuel oil storage tank vent lines The staff found TVAs 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 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.
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.
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* REV.                                      ADDITIONAL INFORMATION 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.
* REV.                                      ADDITIONAL INFORMATION 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.
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.
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The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
9 .5    .5  11  C        01        OUTSTANDING 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 3 The 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.
9 .5    .5  11  C        01        OUTSTANDING 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 3 The 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 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.
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.
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SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                      ADDITIONAL INFORMATION 9 .5    .6  21  O        02        OUTSTANDING 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 3 The 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.
* REV.                                      ADDITIONAL INFORMATION 9 .5    .6  21  O        02        OUTSTANDING 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 3 The 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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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* REV.                                      ADDITIONAL INFORMATION 9 .5    .7  21  O        02        OUTSTANDING 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 features In 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.
* REV.                                      ADDITIONAL INFORMATION 9 .5    .7  21  O        02        OUTSTANDING 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 features In 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.
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 3 The 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.
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 3 The 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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                      ADDITIONAL INFORMATION 9 .5    .8  21  O        02        OUTSTANDING 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 1.26 The 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.
* REV.                                      ADDITIONAL INFORMATION 9 .5    .8  21  O        02        OUTSTANDING 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 1.26 The 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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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10 . 1    .0    0  C                  Approved for both units in SER.
10 . 1    .0    0  C                  Approved for both units in SER.
10 . 2    .0  21  O        02        In SSER5, the staff agreed that the interval between periodic turbine valve testing could be increased for WB from weekly to monthly.
10 . 2    .0  21  O        02        In SSER5, the staff agreed that the interval between periodic turbine valve testing could be increased for WB from weekly to monthly.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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* REV.                                      ADDITIONAL INFORMATION 10 . 3    .1  21  O        02        In 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.
* REV.                                      ADDITIONAL INFORMATION 10 . 3    .1  21  O        02        In 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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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10 . 3    .4    5  S        02        LICENSE CONDITION - Secondary water chemistry monitoring and control program The 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.
10 . 3    .4    5  S        02        LICENSE CONDITION - Secondary water chemistry monitoring and control program The 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.
Unit 2 Action: Take same action for Unit 2.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.
Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.
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10 . 4    .3    0  C                  Approved for both units in SER.
10 . 4    .3    0  C                  Approved for both units in SER.
10 . 4    .4  21  O        02        In 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.
10 . 4    .4  21  O        02        In 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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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Unit 2 Action: Provide system description and information on QA provisions for the Unit 2 Radiation Monitoring System 11 . 6    .0  21  O        02        In 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.
Unit 2 Action: Provide system description and information on QA provisions for the Unit 2 Radiation Monitoring System 11 . 6    .0  21  O        02        In 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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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Noble Gas monitor In 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 staffs concern and resolved the proposed License Condition 6a.
Noble Gas monitor In 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 staffs 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 staffs concern and eliminated the need for 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 staffs concern and eliminated the need for License Condition 6a.
                                        ---------------------
LICENSE CONDITION (6b) - Accident monitoring instrumentation II.F.1 - Iodine particulate sampling See 7.5.2.
LICENSE CONDITION (6b) - Accident monitoring instrumentation II.F.1 - Iodine particulate sampling See 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 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 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 of high-range noble gas monitors on the auxiliary building vent at Watts Bar.
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 of high-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.
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.
The above items were identified as CI by NRC in May 28, 2008, letter.
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Page 67 of 88        * = See last page for status code definition.
Page 67 of 88        * = See last page for status code definition.


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11 . 7    .2  16  S        02        NUREG-0737, III.D.1.1, Primary Coolant Outside Containment - Resolved for Unit 1 only in SSER10; reviewed in Appendix EE of SSER16.
11 . 7    .2  16  S        02        NUREG-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.
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.
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 TVAs 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.
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 TVAs 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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
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12 . 1    .0  21  O        02        In 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.
12 . 1    .0  21  O        02        In 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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                      ADDITIONAL INFORMATION 12 . 2    .0  21  O        02        In 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.
* REV.                                      ADDITIONAL INFORMATION 12 . 2    .0  21  O        02        In 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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
12 . 3    .0  21  O        02        In 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.
12 . 3    .0  21  O        02        In 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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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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 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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                      ADDITIONAL INFORMATION 12 . 5    .0  21  O        02        In 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.
* REV.                                      ADDITIONAL INFORMATION 12 . 5    .0  21  O        02        In 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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
12 . 6    .0  21  O        02        OUTSTANDING ISSUE involving Health Physics Program The staff reviewed TVAs 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 staffs 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.
12 . 6    .0  21  O        02        OUTSTANDING ISSUE involving Health Physics Program The staff reviewed TVAs 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 staffs 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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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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.
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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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Unit 2 Action: Install high range in-containment monitor for Unit 2.
Unit 2 Action: Install high range in-containment monitor for Unit 2.
CI in NRC May 28, 2008, letter.
CI in NRC May 28, 2008, letter.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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NRC reviewed in Appendix EE of SSER16. Identified as CI in NRC letter dated May 28, 2008.
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.
Unit 2 Action: Complete modifications for Unit 2. CI in NRC May 28, 2008, letter.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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13 . 5    .0    0  C                  Approved for both units in SER.
13 . 5    .0    0  C                  Approved for both units in SER.
13 . 5    .1  21  O        02        Approved for both units in SER.
13 . 5    .1  21  O        02        Approved for both units in SER.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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* REV.                                      ADDITIONAL INFORMATION 13 . 5    .2  21  CI        02        OUTSTANDING ISSUE involving operating, maintenance and emergency procedures In the original 1982 SER, this issue was used to track the staffs 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.
* REV.                                      ADDITIONAL INFORMATION 13 . 5    .2  21  CI        02        OUTSTANDING ISSUE involving operating, maintenance and emergency procedures In the original 1982 SER, this issue was used to track the staffs 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.
Unit 2 Action: Issue operating, maintenance and emergency procedures.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (Inspection).
The status in SSER21 is Open (Inspection).
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* REV.                                      ADDITIONAL INFORMATION 13 . 5    .3  21  S        02        LICENSE CONDITION - Report on outage of emergency core cooling system (NUREG-0737, II.K.3.17)
* REV.                                      ADDITIONAL INFORMATION 13 . 5    .3  21  S        02        LICENSE CONDITION - Report on outage of emergency core cooling system (NUREG-0737, II.K.3.17)
In the original 1982 SER, the NRC accepted TVAs 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.
In the original 1982 SER, the NRC accepted TVAs 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)
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.
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.
Unit 2 Action: Include, as necessary, in the Technical Specifications.
CT in NRC May 28, 2008, letter.
CT in NRC May 28, 2008, letter.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (Inspection).
The status in SSER21 is Open (Inspection).
                                        ----------
Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.
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.
Rev. 0 of the Unit 1 TS contained 5.9.4 (Monthly Operating Reports) which implemented the above commitment for Unit 1.
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SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                      ADDITIONAL INFORMATION 13 . 6    .0  21  O        02        OUTSTANDING ISSUE to file appropriate revision to the Physical Security Plan In the original 1982 SER, the staff identified certain outstanding issues with TVAs 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.
* REV.                                      ADDITIONAL INFORMATION 13 . 6    .0  21  O        02        OUTSTANDING ISSUE to file appropriate revision to the Physical Security Plan In the original 1982 SER, the staff identified certain outstanding issues with TVAs 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 - 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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
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* REV.                                  ADDITIONAL INFORMATION 14 . 0    .0  21  S        02        LICENSE CONDITION - Report changes to Initial Test Program In 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 TVAs July 1, 1991, letter to notify NRC within 30 days of any changes to the Startup Test Program made under 10 CFR 50.59.
* REV.                                  ADDITIONAL INFORMATION 14 . 0    .0  21  S        02        LICENSE CONDITION - Report changes to Initial Test Program In 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 TVAs 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.
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 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 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.
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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 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.
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 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.
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 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.
Page 76 of 88        * = See last page for status code definition.
Page 76 of 88        * = See last page for status code definition.
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SER SECTION  SSER #
SER SECTION  SSER #
* REV.                                      ADDITIONAL INFORMATION 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.
* REV.                                      ADDITIONAL INFORMATION 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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (Inspection).
The status in SSER21 is Open (Inspection).
                                          ----------
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
Table 14.2-1 was revised to clarify the testing requirement.
Table 14.2-1 was revised to clarify the testing requirement.
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15 . 2    .0  0    S        02        Approved for both units in SER.
15 . 2    .0  0    S        02        Approved for both units in SER.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
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* REV.                                      ADDITIONAL INFORMATION 15 . 2    .1  14  S        02        In 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.
* REV.                                      ADDITIONAL INFORMATION 15 . 2    .1  14  S        02        In 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.
Unit 2 Action: Provide the additional information for NRC review.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.
TVA responded to the NRC request for additional information on Eagle-21 by letter dated August 25, 2008.
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15 . 2    .3  18  S        02        In 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.
15 . 2    .3  18  S        02        In 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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
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* REV.                                      ADDITIONAL INFORMATION 15 . 2    .4  14    S        02        15.2.4.1 Uncontrolled Rod Cluster Assembly Bank Withdrawal from Zero-Power Condition In 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.
* REV.                                      ADDITIONAL INFORMATION 15 . 2    .4  14    S        02        15.2.4.1 Uncontrolled Rod Cluster Assembly Bank Withdrawal from Zero-Power Condition In 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.
Unit 2 Action: Submit Technical Specifications.
                                          --------------------
15.2.4.4: OUTSTANDING ISSUE for evaluation of Boron dilution and single failure criteria In 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.
15.2.4.4: OUTSTANDING ISSUE for evaluation of Boron dilution and single failure criteria In 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.
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 Ejection In SSER14, NRC accepted a change to the maximum cladding temperature for the rod ejection accident made in FSAR amendment 80.
15.2.4.6 Rod Cluster Control Assembly Ejection In 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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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                                          ----------------------------------------------------------------------------------------------------
REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
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.
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.
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.
Chapter 15 was updated to address the application of RFA-2 fuel.
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* REV.                                      ADDITIONAL INFORMATION 15 . 3    .1  15  S        02        In 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.
* REV.                                      ADDITIONAL INFORMATION 15 . 3    .1  15  S        02        In 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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
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15 . 3    .2  14  S        02        In 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.
15 . 3    .2  14  S        02        In 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: Perform analysis.
                                        --------------------
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
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.
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.
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..
Chapter 15 was updated to address the application of RFA-2 fuel..
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* REV.                                      ADDITIONAL INFORMATION 15 . 3    .3  14  S        02        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.
* REV.                                      ADDITIONAL INFORMATION 15 . 3    .3  14  S        02        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: Perform analysis.
                                        --------------------
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
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.
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.
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.
Chapter 15 was updated to address the application of RFA-2 fuel.
15 . 3    .4  14  S        02        In SSER14, NRC reviewed this section based on VANTAGE 5H fuel and found it acceptable.
15 . 3    .4  14  S        02        In 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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
Line 1,658: Line 1,300:
15 . 3    .5  14  S        02        In SSER14, NRC reviewed this section based on VANTAGE 5H fuel and found it acceptable.
15 . 3    .5  14  S        02        In 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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
Line 1,670: Line 1,310:
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.
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.
CI in May 28, 2008, NRC letter.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (Inspection).
The status in SSER21 is Open (Inspection).
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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.
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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
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* REV.                                      ADDITIONAL INFORMATION 15 . 4    .2  15  S        02        In SSER15, NRC reviewed revised short term atmospheric relative concentration factors. The conclusions reached in the SER and supplements remained unchanged.
* REV.                                      ADDITIONAL INFORMATION 15 . 4    .2  15  S        02        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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
Line 1,699: Line 1,333:
: 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.
: 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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
Line 1,709: Line 1,341:
* REV.                                      ADDITIONAL INFORMATION 15 . 4    .4  15  S        02        In SSER15, NRC reviewed revised short term atmospheric relative concentration factors. The conclusions reached in the SER and supplements remained unchanged.
* REV.                                      ADDITIONAL INFORMATION 15 . 4    .4  15  S        02        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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
Line 1,716: Line 1,346:
15 . 4    .5  15  S        02        In 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.
15 . 4    .5  15  S        02        In 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.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
Line 1,723: Line 1,351:
15 . 4    .6    0  S        02        Approved for both units in SER.
15 . 4    .6    0  S        02        Approved for both units in SER.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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                                        ----------------------------------------------------------------------------------------------------
REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
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15 . 4    .7    0  S        02        Approved for both units in SER.
15 . 4    .7    0  S        02        Approved for both units in SER.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
Unit 2 action: Use Westinghouse RFA-2 fuel as currently installed in Unit 1 for the initial cycle.
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                                        ----------------------------------------------------------------------------------------------------
REVISION 02 UPDATE:
REVISION 02 UPDATE:
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
Amendment 97 to the Unit 2 FSAR was submitted on January 21, 2010.
Line 1,747: Line 1,371:
15 . 5    .4  21  CI        02        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.
15 . 5    .4  21  CI        02        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.
Unit 2 Action: Implement modifications as required.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Status in SSER21 is Open (Inspection).
Status in SSER21 is Open (Inspection).
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* REV.                                      ADDITIONAL INFORMATION 15 . 5    .5  21    S        02        NUREG-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.
* REV.                                      ADDITIONAL INFORMATION 15 . 5    .5  21    S        02        NUREG-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.
Unit 2 Action: Complete analysis for Unit 2.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Status in SSER21 is Open (Inspection).
Status in SSER21 is Open (Inspection).
                                          ----------
Unit 2 FSAR Amendment 97 was submitted on January 11, 2010.
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.
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.
Line 1,766: Line 1,385:
15 . 6    .1    0    C                  Approved for both units in SER.
15 . 6    .1    0    C                  Approved for both units in SER.
16 . 0    .0        S        02        Unit 2 Action: Submit Technical Specifications.
16 . 0    .0        S        02        Unit 2 Action: Submit Technical Specifications.
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                                          ----------------------------------------------------------------------------------------------------
REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision A of the Unit 2 Technical Specifications was submitted on March 4, 2009.
Developmental Revision A of the Unit 2 Technical Specifications was submitted on March 4, 2009.
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18 . 1    .0  21  CI        02        NUREG-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.
18 . 1    .0  21  CI        02        NUREG-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.
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.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
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.
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.
In SSER21, the Detailed Control Room Design Review (CRDR) Special Program was resolved. Completion of CRDR is tracked under 23.3.3.
Page 87 of 88          * = See last page for status code definition.
Page 87 of 88          * = See last page for status code definition.
Line 1,797: Line 1,410:
SER SECTION            SSER #
SER SECTION            SSER #
* REV.                                      ADDITIONAL INFORMATION 18 . 2    .0            21        CI          02      "CONCLUSIONS" left open until all items in subsection are closed.
* REV.                                      ADDITIONAL INFORMATION 18 . 2    .0            21        CI          02      "CONCLUSIONS" left open until all items in subsection are closed.
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                                                        ----------------------------------------------------------------------------------------------------
REVISION 02 UPDATE:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
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.
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.
In SSER21, the Detailed Control Room Design Review (CRDR) Special Program was resolved.
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.
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.
Line 1,821: Line 1,430:
* REV.                                      ADDITIONAL INFORMATION 6 .2      .5    21  S        04        OUTSTANDING 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 LOCA In 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.
* REV.                                      ADDITIONAL INFORMATION 6 .2      .5    21  S        04        OUTSTANDING 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 LOCA In 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.
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 instrumentation II.F.1 - containment hydrogen In SSER5, NRC closed the LICENSE CONDITION for Unit 1 only (IR 390/84-85).
LICENSE CONDITION - (6f) Accident monitoring instrumentation II.F.1 - containment hydrogen In 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.
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)}.
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:
REVISION 02 UPDATE:
The status in SSER21 is Open (NRR).
The status in SSER21 is Open (NRR).
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
Amendment 95 to the Unit 2 FSAR was submitted on November 24, 2009.
This amendment deleted the hydrogen recombiners from the Unit 2 FSAR.
This amendment deleted the hydrogen recombiners from the Unit 2 FSAR.
                                          ----------------------------------------------------------------------------------------------------
                                          ----------------------------------------------------------------------------------------------------
REVISION 04 UPDATE:
REVISION 04 UPDATE:
EDCR 52329 was initiated to abandon in place Unit 2 hydrogen recombiners.
EDCR 52329 was initiated to abandon in place Unit 2 hydrogen recombiners.
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Enclosure 3 Generic Communications - Master Table
Enclosure 3 Generic Communications - Master Table


GENERIC COMMUNICATIONS: MASTER TABLE
GENERIC COMMUNICATIONS: MASTER TABLE ITEM                        TITLE                REV                      ADDITIONAL INFORMATION B 71-002    PWR Reactor Trip Circuit Breakers        NA    Addressed to specific plant(s).
* ITEM                        TITLE                REV                      ADDITIONAL INFORMATION B 71-002    PWR Reactor Trip Circuit Breakers        NA    Addressed to specific plant(s).
B 71-003    Catastrophic Failure of Main Steam Line  NA    Addressed to specific plant(s).
B 71-003    Catastrophic Failure of Main Steam Line  NA    Addressed to specific plant(s).
Relief Valve Headers B 72-001    Failed Hangers for Emergency Core        NA    Addressed to specific plant(s).
Relief Valve Headers B 72-001    Failed Hangers for Emergency Core        NA    Addressed to specific plant(s).
Line 1,859: Line 1,461:
Injection Signal on Both Units of a Dual Unit Facility B 72-003    Limitorque Valve Operator Failures      NA    Addressed to specific plant(s).
Injection Signal on Both Units of a Dual Unit Facility B 72-003    Limitorque Valve Operator Failures      NA    Addressed to specific plant(s).
B 73-001    Faulty Overcurrent Trip Delay Device in  C    TVA: letter dated April 4, 1973 Circuit Breakers for Engineered Safety Systems                                        NRC: IR 390/391 75-5 B 73-002    Malfunction of Containment Purge Supply  C    TVA: letter dated August 22, 1973 Valve Switch NRC: IR 390/391 75-5 B 73-003    Defective Hydraulic Snubbers and          C    TVA: letter dated February 7, 1985 Restraints NRC: IR 390/391 85-08 B 73-004    Defective Bergen-Patterson Hydraulic      C    TVA: memo dated February 7, 1985 Shock Absorbers NRC: IR 390/391 85-08 B 73-005    Manufacturing Defect in BWR Control      NA    Boiling Water Reactor Rods B 73-006    Inadvertent Criticality in a BWR        NA    Boiling Water Reactor B 74-001    Valve Deficiencies                        C    TVA: letter dated April 15, 1974 NRC: IR 390/391 75-5 B 74-002    Truck Strike Possibility                NA    Info Page 1 of 95                                            * = See last page for status code definition.
B 73-001    Faulty Overcurrent Trip Delay Device in  C    TVA: letter dated April 4, 1973 Circuit Breakers for Engineered Safety Systems                                        NRC: IR 390/391 75-5 B 73-002    Malfunction of Containment Purge Supply  C    TVA: letter dated August 22, 1973 Valve Switch NRC: IR 390/391 75-5 B 73-003    Defective Hydraulic Snubbers and          C    TVA: letter dated February 7, 1985 Restraints NRC: IR 390/391 85-08 B 73-004    Defective Bergen-Patterson Hydraulic      C    TVA: memo dated February 7, 1985 Shock Absorbers NRC: IR 390/391 85-08 B 73-005    Manufacturing Defect in BWR Control      NA    Boiling Water Reactor Rods B 73-006    Inadvertent Criticality in a BWR        NA    Boiling Water Reactor B 74-001    Valve Deficiencies                        C    TVA: letter dated April 15, 1974 NRC: IR 390/391 75-5 B 74-002    Truck Strike Possibility                NA    Info Page 1 of 95                                            * = See last page for status code definition.
* ITEM                        TITLE                  REV                        ADDITIONAL INFORMATION B 74-003    Failure of Structural or Seismic Support  CI    TVA: memo dated January 22, 1985 Bolts on Class I Components NRC: IR 390/391 85-08
 
                                                              --------------------
ITEM                        TITLE                  REV                        ADDITIONAL INFORMATION B 74-003    Failure of Structural or Seismic Support  CI    TVA: memo dated January 22, 1985 Bolts on Class I Components NRC: IR 390/391 85-08 Approach accepted in IR 50-390/85-08 and 50-391/85-08 (March 29, 1985).
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.
Unit 2 Action: Implement per NUREG-0577 as was done for Unit 1.
B 74-004    Malfunction of Target Rock Safety Relief  NA    Boiling Water Reactor Valves B 74-005    Shipment of an Improperly Shielded        NA    Does not apply to power reactor.
B 74-004    Malfunction of Target Rock Safety Relief  NA    Boiling Water Reactor Valves B 74-005    Shipment of an Improperly Shielded        NA    Does not apply to power reactor.
Source B 74-006    Defective Westinghouse Type W-2            C    TVA: letter dated October 18, 1974 Control Switch Component NRC: IR 390/391 75-6 B 74-007    Personnel Exposure - Irradiation Facility NA    Does not apply to power reactor.
Source B 74-006    Defective Westinghouse Type W-2            C    TVA: letter dated October 18, 1974 Control Switch Component NRC: IR 390/391 75-6 B 74-007    Personnel Exposure - Irradiation Facility NA    Does not apply to power reactor.
B 74-008    Deficiency in the ITE Molded Case Circuit  C    TVA: letter dated August 21, 1974 Breakers, Type HE-3 NRC: IR 390/391 75-5 B 74-009    Deficiency in GE Model 4KV Magne-Blast    C    TVA: letter dated September 20, 1974 Circuit Breakers NRC: IR 390/391 76-6 B 74-010    Failures in 4-Inch Bypass Pipe at        NA    Boiling Water Reactor Dresden 2 B 74-011    Improper Wiring of Safety Injection Logic  C    NRC: IR 390/391 75-6 at Zion 1 & 2 B 74-012    Incorrect Coils in Westinghouse Type SG    C    NRC: IR 390/391 75-5 Relays at Trojan B 74-013    Improper Factory Wiring on GE Motor        C    TVA: letter dated December 24, 1974 Control Centers at Fort Calhoun NRC: IR 390/391 75-5 B 74-014    BWR Relief Valve Discharge to            NA    Boiling Water Reactor Suppression Pool Page 2 of 95                                            * = See last page for status code definition.
B 74-008    Deficiency in the ITE Molded Case Circuit  C    TVA: letter dated August 21, 1974 Breakers, Type HE-3 NRC: IR 390/391 75-5 B 74-009    Deficiency in GE Model 4KV Magne-Blast    C    TVA: letter dated September 20, 1974 Circuit Breakers NRC: IR 390/391 76-6 B 74-010    Failures in 4-Inch Bypass Pipe at        NA    Boiling Water Reactor Dresden 2 B 74-011    Improper Wiring of Safety Injection Logic  C    NRC: IR 390/391 75-6 at Zion 1 & 2 B 74-012    Incorrect Coils in Westinghouse Type SG    C    NRC: IR 390/391 75-5 Relays at Trojan B 74-013    Improper Factory Wiring on GE Motor        C    TVA: letter dated December 24, 1974 Control Centers at Fort Calhoun NRC: IR 390/391 75-5 B 74-014    BWR Relief Valve Discharge to            NA    Boiling Water Reactor Suppression Pool Page 2 of 95                                            * = See last page for status code definition.
* ITEM                        TITLE                  REV                        ADDITIONAL INFORMATION B 74-015    Misapplication of Cutler-Hammer Three      CI    TVA: letter dated May 5, 1975 Position Maintained Switch Model No.
 
10250T                                          NRC: IR 390/391 75-5
ITEM                        TITLE                  REV                        ADDITIONAL INFORMATION B 74-015    Misapplication of Cutler-Hammer Three      CI    TVA: letter dated May 5, 1975 Position Maintained Switch Model No.
                                                              --------------------
10250T                                          NRC: IR 390/391 75-5 Unit 2 Action: Install modified A3 Cutler-Hammer 10250T switches.
Unit 2 Action: Install modified A3 Cutler-Hammer 10250T switches.
B 74-016    Improper Machining of Pistons in Colt      C    TVA: letter dated January 2, 1975 Industries (Fairbanks-Morse)
B 74-016    Improper Machining of Pistons in Colt      C    TVA: letter dated January 2, 1975 Industries (Fairbanks-Morse)
Diesel-Generators                                NRC: IR 390/391 75-3 B 75-001    Through-Wall Cracks in Core Spray        NA    Boiling Water Reactor Piping at Dresden-2 B 75-002    Defective Radionics Radiograph            NA    Does not apply to power reactor.
Diesel-Generators                                NRC: IR 390/391 75-3 B 75-001    Through-Wall Cracks in Core Spray        NA    Boiling Water Reactor Piping at Dresden-2 B 75-002    Defective Radionics Radiograph            NA    Does not apply to power reactor.
Exposure Devices and Source Changers B 75-003    Incorrect Lower Disc Spring and            CI    TVA: letter dated May 16, 1975 Clearance Dimension in Series 8300 and 8302 ASCO Solenoid Valves                        NRC: IR 390/391 75-6
Exposure Devices and Source Changers B 75-003    Incorrect Lower Disc Spring and            CI    TVA: letter dated May 16, 1975 Clearance Dimension in Series 8300 and 8302 ASCO Solenoid Valves                        NRC: IR 390/391 75-6 NRC accepted in IR 50-390/75-6 and 50-391/75-6 (August 21, 1975).
                                                              --------------------
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.
Unit 2 Action: Modify valves not modified at factory.
B 75-004    Cable Fire at BFNPP                        CI    NRC: IR 390/391 85-08 Closed to Fire Protection CAP
B 75-004    Cable Fire at BFNPP                        CI    NRC: IR 390/391 85-08 Closed to Fire Protection CAP Part of Fire Protection CAP B 75-005    Operability of Category I Hydraulic Shock  CI    TVA: letter dated June 16, 1975 and Sway Suppressors NRC: IR 390/391 75-6 NRC accepted in IR 50-390/75-6 and 50-391/75-6 (August 21, 1975).
                                                              --------------------
Part of Fire Protection CAP B 75-005    Operability of Category I Hydraulic Shock  CI    TVA: letter dated June 16, 1975 and Sway Suppressors NRC: IR 390/391 75-6
                                                              --------------------
NRC accepted in IR 50-390/75-6 and 50-391/75-6 (August 21, 1975).
Unit 2 Action: Install proper suppressors.
Unit 2 Action: Install proper suppressors.
B 75-006    Defective Westinghouse Type OT-2          CI    TVA: letter dated July 31, 1975 Control Switches NRC: IR 390/85-25 and 391/85-20
B 75-006    Defective Westinghouse Type OT-2          CI    TVA: letter dated July 31, 1975 Control Switches NRC: IR 390/85-25 and 391/85-20 Unit 2 Action: Inspect Westinghouse Type OT-2 control switches.
                                                              --------------------
Unit 2 Action: Inspect Westinghouse Type OT-2 control switches.
[WAS "NOTE 3."]
[WAS "NOTE 3."]
Page 3 of 95                                            * = See last page for status code definition.
Page 3 of 95                                            * = See last page for status code definition.
* ITEM                        TITLE                      REV                          ADDITIONAL INFORMATION B 75-007    Exothermic Reaction in Radwaste              NA      Does not apply to power reactor.
 
Shipment B 75-008    PWR Pressure Instrumentation                  S      NRC: IR 390/391 85-08
ITEM                        TITLE                      REV                          ADDITIONAL INFORMATION B 75-007    Exothermic Reaction in Radwaste              NA      Does not apply to power reactor.
                                                                  -------------------
Shipment B 75-008    PWR Pressure Instrumentation                  S      NRC: IR 390/391 85-08 02    Unit 2 Action: Ensure that Technical Specifications and Site Operating Instructions address importance of maintaining temperature and pressure within prescribed limits.
02    Unit 2 Action: Ensure that Technical Specifications and Site Operating Instructions address importance of maintaining temperature and pressure within prescribed limits.
                                                                  ----------------------------------------------------------------------------------------
                                                                  ----------------------------------------------------------------------------------------
REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
Line 1,899: Line 1,488:
B 76-003    Relay Malfunctions - GE Type STD              C      TVA: letter dated May 17, 1976 Relays NRC: IR 390/391 76-6 B 76-004    Cracks in Cold Worked Piping at BWRs        NA      Boiling Water Reactor B 76-005    Relay Failures - Westinghouse BFD            C      TVA: letter dated June 7, 1976 Relays NRC: IR 390/391 85-08 B 76-006    Diaphragm Failures in Air Operated            C      TVA: memo dated January 25, 1985 Auxiliary Actuators for Safety/Relief Valves NRC: IR 390/391 85-08 B 76-007    Crane Hoist Control Circuit Modifications    C      TVA: letter dated October 29, 1976 NRC: IR 390/391 85-08 B 76-008    Teletherapy Units                            NA      Does not apply to power reactor.
B 76-003    Relay Malfunctions - GE Type STD              C      TVA: letter dated May 17, 1976 Relays NRC: IR 390/391 76-6 B 76-004    Cracks in Cold Worked Piping at BWRs        NA      Boiling Water Reactor B 76-005    Relay Failures - Westinghouse BFD            C      TVA: letter dated June 7, 1976 Relays NRC: IR 390/391 85-08 B 76-006    Diaphragm Failures in Air Operated            C      TVA: memo dated January 25, 1985 Auxiliary Actuators for Safety/Relief Valves NRC: IR 390/391 85-08 B 76-007    Crane Hoist Control Circuit Modifications    C      TVA: letter dated October 29, 1976 NRC: IR 390/391 85-08 B 76-008    Teletherapy Units                            NA      Does not apply to power reactor.
Page 4 of 95                                                * = See last page for status code definition.
Page 4 of 95                                                * = See last page for status code definition.
* ITEM                      TITLE                REV                          ADDITIONAL INFORMATION B 77-001    Pneumatic Time Delay Relay Setpoint      C      TVA: letter dated July 1, 1977 Drift NRC: IR 390/391 85-08 B 77-002    Potential Failure Mechanism in Certain  C      TVA: letter dated November 11, 1977 Westinghouse AR Relays with Latch Attachments                                    NRC: IR 390/391 85-08 B 77-003    On-Line Testing of the Westinghouse      CI    Unit 2 Action: Include necessary periodic testing in test Solid State Protection System                  procedures.
 
ITEM                      TITLE                REV                          ADDITIONAL INFORMATION B 77-001    Pneumatic Time Delay Relay Setpoint      C      TVA: letter dated July 1, 1977 Drift NRC: IR 390/391 85-08 B 77-002    Potential Failure Mechanism in Certain  C      TVA: letter dated November 11, 1977 Westinghouse AR Relays with Latch Attachments                                    NRC: IR 390/391 85-08 B 77-003    On-Line Testing of the Westinghouse      CI    Unit 2 Action: Include necessary periodic testing in test Solid State Protection System                  procedures.
B 77-004    Calculation Error Affecting The Design  S      TVA: letter dated January 23, 1978 Performance of a System for Controlling pH of Containment Sump Water            02    NRC: IR 390/78-11 and 391/78-09 Following a LOCA                                ------------------------
B 77-004    Calculation Error Affecting The Design  S      TVA: letter dated January 23, 1978 Performance of a System for Controlling pH of Containment Sump Water            02    NRC: IR 390/78-11 and 391/78-09 Following a LOCA                                ------------------------
Unit 2 Action: Ensure Technical Specifications includes limit on Boron concentration.
Unit 2 Action: Ensure Technical Specifications includes limit on Boron concentration.
                                                            ----------------------------------------------------------------------------------------
                                                            ----------------------------------------------------------------------------------------
REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
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.
TS Surveillance Requirement 3.6.11.5 requires verification that the boron concentration is within a specified range.
B 77-005    Electrical Connector Assemblies          C      TVA: letter dated January 17, 1978 and B 77-005 A                                                  NRC: IR 390/78-11 and 391/78-09 B 77-006    Potential Problems with Containment      C      Item was applicable only to units with operating license at the Electrical Penetration Assemblies              time the item was issued.
B 77-005    Electrical Connector Assemblies          C      TVA: letter dated January 17, 1978 and B 77-005 A                                                  NRC: IR 390/78-11 and 391/78-09 B 77-006    Potential Problems with Containment      C      Item was applicable only to units with operating license at the Electrical Penetration Assemblies              time the item was issued.
                                                            --------------------
NRC: IR 390/391 85-08 B 77-007    Containment Electrical Penetration      C      TVA: letter dated January 20, 1978 Assemblies at Nuclear Power Plants Under Construction                              NRC: IR 390/78-11 and 391/78-09 B 77-008    Assurance of Safety and Safeguards      C      Item concerns a multi-unit issue that was completed for both During an Emergency - Locking Systems          units.
NRC: IR 390/391 85-08 B 77-007    Containment Electrical Penetration      C      TVA: letter dated January 20, 1978 Assemblies at Nuclear Power Plants Under Construction                              NRC: IR 390/78-11 and 391/78-09 B 77-008    Assurance of Safety and Safeguards      C      Item concerns a multi-unit issue that was completed for both During an Emergency - Locking Systems          units.
                                                            --------------------
TVA: letter dated March 1, 1978 NRC: IR 390/78-11 and 391/78-09 B 78-001    Flammable Contact - Arm Retainers in    C      TVA: letter dated March 20, 1978 GE CR120A Relays NRC: IR 390/78-11 and 391/78-09 Page 5 of 95                                            * = See last page for status code definition.
TVA: letter dated March 1, 1978 NRC: IR 390/78-11 and 391/78-09 B 78-001    Flammable Contact - Arm Retainers in    C      TVA: letter dated March 20, 1978 GE CR120A Relays NRC: IR 390/78-11 and 391/78-09 Page 5 of 95                                            * = See last page for status code definition.
* ITEM                        TITLE                    REV                        ADDITIONAL INFORMATION B 78-002    Terminal Block Qualification                C    TVA: letter dated March 1, 1978 NRC: IR 390/78-11 and 391/78-09 B 78-003    Potential Explosive Gas Mixture            NA    Boiling Water Reactor Accumulations Associated with BWR Offgas System Operations B 78-004    Environmental Qualification of Certain      CI    TVA: letter dated December 19, 1978 Stem Mounted Limit Switches Inside Reactor Containment                                NRC: IR 390/82-13 and 391/82-10 Closed to EQ Program
 
                                                                --------------------
ITEM                        TITLE                    REV                        ADDITIONAL INFORMATION B 78-002    Terminal Block Qualification                C    TVA: letter dated March 1, 1978 NRC: IR 390/78-11 and 391/78-09 B 78-003    Potential Explosive Gas Mixture            NA    Boiling Water Reactor Accumulations Associated with BWR Offgas System Operations B 78-004    Environmental Qualification of Certain      CI    TVA: letter dated December 19, 1978 Stem Mounted Limit Switches Inside Reactor Containment                                NRC: IR 390/82-13 and 391/82-10 Closed to EQ Program IR 50-390/82-13 and 50-391/82-10 (April 22, 1982) accepted approach.
IR 50-390/82-13 and 50-391/82-10 (April 22, 1982) accepted approach.
Unit 2 Action: Ensure NAMCO switches have been replaced.
Unit 2 Action: Ensure NAMCO switches have been replaced.
B 78-005    Malfunctioning of Circuit Breaker Auxiliary  C    TVA: letter dated June 12, 1978 Contact Mechanism - GE Model CR105X                                            NRC: IR 390/78-17 and 391/78-15 B 78-006    Defective Cutler-Hammer Type M Relays        C    NRC: IR 390/78-22 and 391/78-19 With DC Coils B 78-007    Protection Afforded by Air-Line            NA    Item was applicable only to units with operating license at the Respirators and Supplied-Air Hoods                time the item was issued.
B 78-005    Malfunctioning of Circuit Breaker Auxiliary  C    TVA: letter dated June 12, 1978 Contact Mechanism - GE Model CR105X                                            NRC: IR 390/78-17 and 391/78-15 B 78-006    Defective Cutler-Hammer Type M Relays        C    NRC: IR 390/78-22 and 391/78-19 With DC Coils B 78-007    Protection Afforded by Air-Line            NA    Item was applicable only to units with operating license at the Respirators and Supplied-Air Hoods                time the item was issued.
B 78-008    Radiation Levels from Fuel Element          NA    Item was applicable only to units with operating license at the Transfer Tubes                                    time the item was issued.
B 78-008    Radiation Levels from Fuel Element          NA    Item was applicable only to units with operating license at the Transfer Tubes                                    time the item was issued.
                                                                --------------------
NRC: IR 390/391 85-08 B 78-009    BWR Drywell Leakage Paths Associated        NA    Boiling Water Reactor with Inadequate Drywell Closures B 78-010    Bergen-Patterson Hydraulic Shock            C    TVA: letter dated August 14, 1978 Suppressor Accumulator Spring Coils NRC: IR 390/78-22 and 391/78-19 B 78-011    Examination of Mark I Containment Torus    NA    Boiling Water Reactor Welds B 78-012    Atypical Weld Material in Reactor            C    TVA: Westinghouse letter dated October 29, 1979 Pressure Vessel Welds NRC: IR 390/391 81-04 Page 6 of 95                                              * = See last page for status code definition.
NRC: IR 390/391 85-08 B 78-009    BWR Drywell Leakage Paths Associated        NA    Boiling Water Reactor with Inadequate Drywell Closures B 78-010    Bergen-Patterson Hydraulic Shock            C    TVA: letter dated August 14, 1978 Suppressor Accumulator Spring Coils NRC: IR 390/78-22 and 391/78-19 B 78-011    Examination of Mark I Containment Torus    NA    Boiling Water Reactor Welds B 78-012    Atypical Weld Material in Reactor            C    TVA: Westinghouse letter dated October 29, 1979 Pressure Vessel Welds NRC: IR 390/391 81-04 Page 6 of 95                                              * = See last page for status code definition.
* ITEM                        TITLE                    REV                        ADDITIONAL INFORMATION B 78-013    Failures in Source Heads Kay Ray, Inc.      NA    Does not apply to power reactor.
 
ITEM                        TITLE                    REV                        ADDITIONAL INFORMATION B 78-013    Failures in Source Heads Kay Ray, Inc.      NA    Does not apply to power reactor.
Gauges Models 7050, 7050B, 7051, 7051B, 7060, 7060B, 7061 and 7061B B 78-014    Deterioration of Buna-N Components in      NA    Boiling Water Reactor ASCO Solenoids B 79-001    Environmental Qualification of Class 1E      C    NRC: IR 390/80-06 and 391/80-05 Equipment B 79-002    Pipe Support Base Plate Designs Using        CI    NRC review of HAAUP Program in NUREG-1232, SSER6, and Concrete Expansion Anchor Bolts                    SSER8.
Gauges Models 7050, 7050B, 7051, 7051B, 7060, 7060B, 7061 and 7061B B 78-014    Deterioration of Buna-N Components in      NA    Boiling Water Reactor ASCO Solenoids B 79-001    Environmental Qualification of Class 1E      C    NRC: IR 390/80-06 and 391/80-05 Equipment B 79-002    Pipe Support Base Plate Designs Using        CI    NRC review of HAAUP Program in NUREG-1232, SSER6, and Concrete Expansion Anchor Bolts                    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.
Unit 2 Actions: Addressed in CAP/SP. Conduct a complete review of affected support calculations, and perform the necessary revisions to design documents and field modifications to achieve compliance.
B 79-003    Longitudinal Weld Defects in ASME            C    TVA: letter dated July 16, 1981 SA-312 Type 304 SS Pipe Spools Manufactured by Youngstown Welding &              NRC: IRs 390/82-21 and 391/82-17; 390/84-35 and Engineering                                        391/84-33 B 79-004    Incorrect Weights for Swing Check            C    TVA: letter dated October 20, 1980 Valves Manufactured by Velan Engineering Corporation                            NRC: IR 390/83-15 and 391/83-11 B 79-005    Nuclear Incident at TMI                    NA    Applies only to Babcock and Wilcox designed plants B 79-006    Review of Operational Errors and System      C    NRC: IR 390/80-06 and 391/80-05 Misalignments Identified During the Three Mile Island Incident B 79-007    Seismic Stress Analysis of Safety-Related    C    TVA: letter dated May 31, 1979 Piping NRC: IR 390/79-30 and 391/79-25 B 79-008    Events Relevant to BWRs Identified          NA    Boiling Water Reactor During TMI Incident B 79-009    Failure of GE Type AK-2 Circuit Breaker      CI    TVA: letter dated June 20, 1979 in Safety Related Systems
B 79-003    Longitudinal Weld Defects in ASME            C    TVA: letter dated July 16, 1981 SA-312 Type 304 SS Pipe Spools Manufactured by Youngstown Welding &              NRC: IRs 390/82-21 and 391/82-17; 390/84-35 and Engineering                                        391/84-33 B 79-004    Incorrect Weights for Swing Check            C    TVA: letter dated October 20, 1980 Valves Manufactured by Velan Engineering Corporation                            NRC: IR 390/83-15 and 391/83-11 B 79-005    Nuclear Incident at TMI                    NA    Applies only to Babcock and Wilcox designed plants B 79-006    Review of Operational Errors and System      C    NRC: IR 390/80-06 and 391/80-05 Misalignments Identified During the Three Mile Island Incident B 79-007    Seismic Stress Analysis of Safety-Related    C    TVA: letter dated May 31, 1979 Piping NRC: IR 390/79-30 and 391/79-25 B 79-008    Events Relevant to BWRs Identified          NA    Boiling Water Reactor During TMI Incident B 79-009    Failure of GE Type AK-2 Circuit Breaker      CI    TVA: letter dated June 20, 1979 in Safety Related Systems Unit 2 Action: Complete preservice preventive maintenance on AK-2 Circuit Breakers.
                                                                --------------------
Unit 2 Action: Complete preservice preventive maintenance on AK-2 Circuit Breakers.
[WAS "NOTE 3."]
[WAS "NOTE 3."]
B 79-010    Requalification Training Program Statistics NA    Item was applicable only to units with operating license at the time the item was issued.
B 79-010    Requalification Training Program Statistics NA    Item was applicable only to units with operating license at the time the item was issued.
Page 7 of 95                                              * = See last page for status code definition.
Page 7 of 95                                              * = See last page for status code definition.
* ITEM                        TITLE                  REV                        ADDITIONAL INFORMATION B 79-011    Faulty Overcurrent Trip Device in Circuit  C    TVA: letter dated July 20, 1979 Breakers for Engineering Safety Systems NRC: IR 390/79-30 and 391/79-25 B 79-012    Short Period Scrams at BWR Facilities    NA    Boiling Water Reactor B 79-013    Cracking in Feedwater Piping              C    Item was applicable only to units with operating license at the time the item was issued.
 
                                                              --------------------
ITEM                        TITLE                  REV                        ADDITIONAL INFORMATION B 79-011    Faulty Overcurrent Trip Device in Circuit  C    TVA: letter dated July 20, 1979 Breakers for Engineering Safety Systems NRC: IR 390/79-30 and 391/79-25 B 79-012    Short Period Scrams at BWR Facilities    NA    Boiling Water Reactor B 79-013    Cracking in Feedwater Piping              C    Item was applicable only to units with operating license at the time the item was issued.
TVA: letter dated December 1, 1983 NRC: IR 390/391 85-08 B 79-014    Seismic Analysis for As-Built              CI    NRC review of HAAUP Program in NUREG-1232, SSER6, and Safety-Related Piping Systems                    SSER8.
TVA: letter dated December 1, 1983 NRC: IR 390/391 85-08 B 79-014    Seismic Analysis for As-Built              CI    NRC review of HAAUP Program in NUREG-1232, SSER6, and Safety-Related Piping Systems                    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.
Unit 2 Actions: Addressed in CAP/SP. Initiate a Unit 2 hanger walkdown and hanger analysis program similar to the program for Unit 1. Complete re-analysis of piping and associated supports as necessary. Perform modifications as required by re-analysis.
B 79-015    Deep Draft Pump Deficiencies              C    TVA: letter dated January 24, 1992 NRC: IR 390/391 95-70 B 79-016    Vital Area Access Controls                NA    Item was applicable only to units with operating license at the time the item was issued.
B 79-015    Deep Draft Pump Deficiencies              C    TVA: letter dated January 24, 1992 NRC: IR 390/391 95-70 B 79-016    Vital Area Access Controls                NA    Item was applicable only to units with operating license at the time the item was issued.
                                                              --------------------
NRC: IR 390/80-06 and 391/80-05 B 79-017    Pipe Cracks in Stagnant Borated Water    NA    Item was applicable only to units with operating license at the Systems at PWR Plants                            time the item was issued.
NRC: IR 390/80-06 and 391/80-05 B 79-017    Pipe Cracks in Stagnant Borated Water    NA    Item was applicable only to units with operating license at the Systems at PWR Plants                            time the item was issued.
NRC: IR 390/80-06 and 391/80-05; NUREG/ CR 5286 B 79-018    Audibility Problems Encountered on        NA    Item was applicable only to units with operating license at the Evacuation of Personnel from                    time the item was issued.
NRC: IR 390/80-06 and 391/80-05; NUREG/ CR 5286 B 79-018    Audibility Problems Encountered on        NA    Item was applicable only to units with operating license at the Evacuation of Personnel from                    time the item was issued.
High-Noise Areas
High-Noise Areas NRC: IR 390/80-06 and 391/80-05 B 79-019    Packaging of Low-Level Radioactive        NA    Item was applicable only to units with operating license at the Waste for Transport and Burial                  time the item was issued.
                                                              --------------------
NRC: IR 390/80-06 and 391/80-05 B 79-019    Packaging of Low-Level Radioactive        NA    Item was applicable only to units with operating license at the Waste for Transport and Burial                  time the item was issued.
                                                              --------------------
NRC: IR 390/80-06 and 391/80-05 Page 8 of 95                                            * = See last page for status code definition.
NRC: IR 390/80-06 and 391/80-05 Page 8 of 95                                            * = See last page for status code definition.
* ITEM                        TITLE                REV                        ADDITIONAL INFORMATION B 79-020    Packaging, Transport and Burial of      NA    Item was applicable only to units with operating license at the Low-Level Radioactive Waste                    time the item was issued.
 
                                                            --------------------
ITEM                        TITLE                REV                        ADDITIONAL INFORMATION B 79-020    Packaging, Transport and Burial of      NA    Item was applicable only to units with operating license at the Low-Level Radioactive Waste                    time the item was issued.
NRC: IR 390/80-06 and 391/80-05 B 79-021    Temperature Effects on Level              CI    Reviewed in 7.2.5 of both the original 1982 SER and SSER14.
NRC: IR 390/80-06 and 391/80-05 B 79-021    Temperature Effects on Level              CI    Reviewed in 7.2.5 of both the original 1982 SER and SSER14.
Measurements Unit 2 Action: Update accident calculation.
Measurements Unit 2 Action: Update accident calculation.
                                                            -------------------
CONFIRMATORY ISSUE - address IEB 79-21 to alleviate temperature dependence problem associated with measuring SG water level In SSER14, NRC concurred with TVA's assessment to not insulate the steam generator water level instrument reference leg.
CONFIRMATORY ISSUE - address IEB 79-21 to alleviate temperature dependence problem associated with measuring SG water level In SSER14, NRC concurred with TVA's assessment to not insulate the steam generator water level instrument reference leg.
Unit 2 Action: Update accident calculation.
Unit 2 Action: Update accident calculation.
B 79-022    Possible Leakage of Tubes of Tritium Gas NA    Does not apply to power reactor.
B 79-022    Possible Leakage of Tubes of Tritium Gas NA    Does not apply to power reactor.
Used in Time Pieces for Luminosity
Used in Time Pieces for Luminosity NRC: IR 390/80-06 and 391/80-05 B 79-023    Potential Failure of Emergency Diesel    C    TVA: letter dated October 29, 1979 Generator Field Exciter Transformer NRC: IR 390/80-06 and 391/80-05 B 79-024    Frozen Lines                              CI    Unit 2 Actions: Insulate the section of piping in the containment spray full-flow test line that is exposed to outside air. Confirm installation of heat tracing on the sensing lines off the feedwater flow elements.
                                                            --------------------
NRC: IR 390/80-06 and 391/80-05 B 79-023    Potential Failure of Emergency Diesel    C    TVA: letter dated October 29, 1979 Generator Field Exciter Transformer NRC: IR 390/80-06 and 391/80-05 B 79-024    Frozen Lines                              CI    Unit 2 Actions: Insulate the section of piping in the containment spray full-flow test line that is exposed to outside air. Confirm installation of heat tracing on the sensing lines off the feedwater flow elements.
B 79-025    Failures of Westinghouse BFD Relays in    C    TVA: letter dated January 4, 1980 Safety-Related Systems NRC: IR 390/80-03 and 391/80-02 B 79-026    Boron Loss from BWR Control Blades      NA    Boiling Water Reactor B 79-027    Loss of Non-Class 1E I & C Power          CI    TVA responded to the Bulletin on March 1, 1982. Reviewed in System Bus During Operation                    7.5.3 of the original 1982 SER.
B 79-025    Failures of Westinghouse BFD Relays in    C    TVA: letter dated January 4, 1980 Safety-Related Systems NRC: IR 390/80-03 and 391/80-02 B 79-026    Boron Loss from BWR Control Blades      NA    Boiling Water Reactor B 79-027    Loss of Non-Class 1E I & C Power          CI    TVA responded to the Bulletin on March 1, 1982. Reviewed in System Bus During Operation                    7.5.3 of the original 1982 SER.
Unit 2 Action: Issue appropriate emergency procedures.
Unit 2 Action: Issue appropriate emergency procedures.
B 79-028    Possible Malfunction of NAMCO Model      C    TVA: letter dated April 1, 1993 EA180 Limit Switches at Elevated Temperatures                                    NRC: IR 390/391 93-32 B 80-001    Operability of ADS Valve Pneumatic      NA    Boiling Water Reactor Supply Page 9 of 95                                            * = See last page for status code definition.
B 79-028    Possible Malfunction of NAMCO Model      C    TVA: letter dated April 1, 1993 EA180 Limit Switches at Elevated Temperatures                                    NRC: IR 390/391 93-32 B 80-001    Operability of ADS Valve Pneumatic      NA    Boiling Water Reactor Supply Page 9 of 95                                            * = See last page for status code definition.
* ITEM                        TITLE                    REV                      ADDITIONAL INFORMATION B 80-002      Inadequate QA for Nuclear Supplied        NA    Boiling Water Reactor Equipment B 80-003      Loss of Charcoal from Standard Type II, 2  C    TVA: letter dated March 21, 1980 Inch, Tray Adsorber Cells NRC: IR 390/80-15 and 391/80-12 B 80-004      Analysis of a PWR Main Steam Line          CI    IR 50-390/85-60 and 50-391/85-49 (December 6, 1985) required Break with Continued Feedwater Addition          completion of actions that included determination of temperature profiles inside and outside of containment following a MSLB for Unit 1.
 
ITEM                        TITLE                    REV                      ADDITIONAL INFORMATION B 80-002      Inadequate QA for Nuclear Supplied        NA    Boiling Water Reactor Equipment B 80-003      Loss of Charcoal from Standard Type II, 2  C    TVA: letter dated March 21, 1980 Inch, Tray Adsorber Cells NRC: IR 390/80-15 and 391/80-12 B 80-004      Analysis of a PWR Main Steam Line          CI    IR 50-390/85-60 and 50-391/85-49 (December 6, 1985) required Break with Continued Feedwater Addition          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.
Unit 2 Action: Complete analysis for Unit 2.
B 80-005      Vacuum Condition Resulting in Damage      CI    Closed in IR 50-390/84-59 and 50-391/84-45.
B 80-005      Vacuum Condition Resulting in Damage      CI    Closed in IR 50-390/84-59 and 50-391/84-45.
Line 1,967: Line 1,544:
Unit 2 Action: Complete implementation for Unit 2.
Unit 2 Action: Complete implementation for Unit 2.
Page 10 of 95                                            * = See last page for status code definition.
Page 10 of 95                                            * = See last page for status code definition.
* ITEM                        TITLE                  REV                        ADDITIONAL INFORMATION B 80-012      Decay Heat Removal System Operability      CI    NRC: IR 390/391 85-08; NUREG/CR 4005
 
                                                                --------------------
ITEM                        TITLE                  REV                        ADDITIONAL INFORMATION B 80-012      Decay Heat Removal System Operability      CI    NRC: IR 390/391 85-08; NUREG/CR 4005 Unit 2 Action: Implement operating instructions and abnormal operating instructions (AOIs) for RHR.
Unit 2 Action: Implement operating instructions and abnormal operating instructions (AOIs) for RHR.
[WAS "NOTE 3."]
[WAS "NOTE 3."]
B 80-013      Cracking in Core Spray Spargers            NA    Boiling Water Reactor B 80-014      Degradation of Scram Discharge Volume      NA    Boiling Water Reactor Capability B 80-015      Possible Loss of Emergency Notification    C    Item concerns a multi-unit issue that was completed for both System with Loss of Offsite Power                units.
B 80-013      Cracking in Core Spray Spargers            NA    Boiling Water Reactor B 80-014      Degradation of Scram Discharge Volume      NA    Boiling Water Reactor Capability B 80-015      Possible Loss of Emergency Notification    C    Item concerns a multi-unit issue that was completed for both System with Loss of Offsite Power                units.
                                                                --------------------
NRC: IR 390/391 85-08 B 80-016      Potential Misapplication of Rosemount,      C    TVA: letter dated August 29, 1980 Inc. Models 1151 and 1152 Pressure Transmitters With Either A or D Output        NRC: IR 390/391 81-17 Codes B 80-017      Failure of 76 of 185 Control Rods to Fully NA    Boiling Water Reactor Insert During a Scram at a BWR B 80-018      Maintenance of Adequate Minimum Flow        CI    IR 50-390/85-60 and 50-391/85-49 (Unit 1)
NRC: IR 390/391 85-08 B 80-016      Potential Misapplication of Rosemount,      C    TVA: letter dated August 29, 1980 Inc. Models 1151 and 1152 Pressure Transmitters With Either A or D Output        NRC: IR 390/391 81-17 Codes B 80-017      Failure of 76 of 185 Control Rods to Fully NA    Boiling Water Reactor Insert During a Scram at a BWR B 80-018      Maintenance of Adequate Minimum Flow        CI    IR 50-390/85-60 and 50-391/85-49 (Unit 1)
Thru Centrifugal Charging Pumps Following Secondary Side High Energy              Unit 2 Action: Implement design and procedure changes.
Thru Centrifugal Charging Pumps Following Secondary Side High Energy              Unit 2 Action: Implement design and procedure changes.
Line 1,979: Line 1,554:
NRC: 390/391 85-08 B 80-022      Automation Industries, Model 200-520-      NA    Does not apply to power reactor.
NRC: 390/391 85-08 B 80-022      Automation Industries, Model 200-520-      NA    Does not apply to power reactor.
008 Sealed-Source Connectors B 80-023      Failures of Solenoid Valves Manufactured    C    TVA: letter dated March 31, 1981 by Valcor Engineering Corporation NRC: IR 390/391 81-17; NUREG/CR 5292 Page 11 of 95                                              * = See last page for status code definition.
008 Sealed-Source Connectors B 80-023      Failures of Solenoid Valves Manufactured    C    TVA: letter dated March 31, 1981 by Valcor Engineering Corporation NRC: IR 390/391 81-17; NUREG/CR 5292 Page 11 of 95                                              * = See last page for status code definition.
* ITEM                          TITLE                REV                        ADDITIONAL INFORMATION B 80-024      Prevention of Damage Due to Water        CI    Unit 2 Action: Confirm that the reactor cavity can not be flooded, Leakage Inside Containment (10/17/80            resulting in the partial or total submergence of the reactor vessel Indian Point 2 Event)                          unnoticed by the reactor operators.
 
B 80-025      Operating Problems with Target Rock      NA    Boiling Water Reactor Safety-Relief Valves at BWRs B 81-001      Surveillance of Mechanical Snubbers      NA    NRC: IR 390/391 81-17 B 81-002      Failure of Gate Type Valves to Close      C    TVA: letter dated September 30, 1983 Against Differential Pressure NRC: IR 390/391 84-03 B 81-003      Flow Blockage of Cooling Water to Safety  C    TVA: letters dated July 21, 1981 and March 21, 1983 System Components by Asiatic Clams and Mussels                                    NRC: IR 390/391 81-17 B 82-001      Alteration of Radiographs of Welds in    C    NRC: IR 390/391 85-08 Piping Subassemblies B 82-002      Degradation of Threaded Fasteners in the  CI    TVA: memo dated February 6, 1985 Reactor Coolant Pressure Boundary of PWR Plants                                      NRC: IR 390/391 85-08
ITEM                          TITLE                REV                        ADDITIONAL INFORMATION B 80-024      Prevention of Damage Due to Water        CI    Unit 2 Action: Confirm that the reactor cavity can not be flooded, Leakage Inside Containment (10/17/80            resulting in the partial or total submergence of the reactor vessel Indian Point 2 Event)                          unnoticed by the reactor operators.
                                                              --------------------
B 80-025      Operating Problems with Target Rock      NA    Boiling Water Reactor Safety-Relief Valves at BWRs B 81-001      Surveillance of Mechanical Snubbers      NA    NRC: IR 390/391 81-17 B 81-002      Failure of Gate Type Valves to Close      C    TVA: letter dated September 30, 1983 Against Differential Pressure NRC: IR 390/391 84-03 B 81-003      Flow Blockage of Cooling Water to Safety  C    TVA: letters dated July 21, 1981 and March 21, 1983 System Components by Asiatic Clams and Mussels                                    NRC: IR 390/391 81-17 B 82-001      Alteration of Radiographs of Welds in    C    NRC: IR 390/391 85-08 Piping Subassemblies B 82-002      Degradation of Threaded Fasteners in the  CI    TVA: memo dated February 6, 1985 Reactor Coolant Pressure Boundary of PWR Plants                                      NRC: IR 390/391 85-08 Approach accepted in IR 50-390/85-08 and 50-391/85-08 (March 29, 1985).
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.
Unit 2 Action: Implement same approach as Unit 1.
B 82-003      Stress Corrosion Cracking in Thick-Wall, NA    Boiling Water Reactor Large Diameter, Stainless Steel, Recirculation System Piping at BWR Plants B 82-004      Deficiencies in Primary Containment      C    TVA: letter dated January 24, 1983 Electrical Penetration Assemblies NRC: IR 390/83-10 and 391/83-08 B 83-001      Failure of Trip Breakers (Westinghouse    C    NRC: IRs 390/391 85-08 and 390/391 92-13 DB-50) to Open on Automatic Trip Signal B 83-002      Stress Corrosion Cracking in Large-      NA    Boiling Water Reactor Diameter Stainless Steel Recirculation System Piping at BWR Plants B 83-003      Check Valve Failures in Raw Water        NA    Addressed by Inservice Testing for Construction Permit holders Cooling Systems of Diesel Generators Page 12 of 95                                            * = See last page for status code definition.
B 82-003      Stress Corrosion Cracking in Thick-Wall, NA    Boiling Water Reactor Large Diameter, Stainless Steel, Recirculation System Piping at BWR Plants B 82-004      Deficiencies in Primary Containment      C    TVA: letter dated January 24, 1983 Electrical Penetration Assemblies NRC: IR 390/83-10 and 391/83-08 B 83-001      Failure of Trip Breakers (Westinghouse    C    NRC: IRs 390/391 85-08 and 390/391 92-13 DB-50) to Open on Automatic Trip Signal B 83-002      Stress Corrosion Cracking in Large-      NA    Boiling Water Reactor Diameter Stainless Steel Recirculation System Piping at BWR Plants B 83-003      Check Valve Failures in Raw Water        NA    Addressed by Inservice Testing for Construction Permit holders Cooling Systems of Diesel Generators Page 12 of 95                                            * = See last page for status code definition.
* ITEM                          TITLE                REV                          ADDITIONAL INFORMATION B 83-004      Failure of the Undervoltage Trip Function  CI    NRC: IR 390/391 85-08 of Reactor Trip Breakers
 
                                                                --------------------
ITEM                          TITLE                REV                          ADDITIONAL INFORMATION B 83-004      Failure of the Undervoltage Trip Function  CI    NRC: IR 390/391 85-08 of Reactor Trip Breakers Unit 2 Action: Install new undervoltage attachment with wider grooves on the reactor trip breakers.
Unit 2 Action: Install new undervoltage attachment with wider grooves on the reactor trip breakers.
B 83-005      ASME Nuclear Code Pumps and Spare          C      TVA: letter dated September 7, 1983 Parts Manufactured by the Hayward Tyler Pump Company                                      NRC: IR 390/85-03 and 391/85-04; NUREG/CR 5297 B 83-006      Nonconforming Material Supplied by        CI    TVA: letter dated February 2, 1984 Tube-Line Facilities 04    NRC: IR 390/391 84-03; NUREG/CR 4934 NRC SER for both units dated September 23, 1991, provided an alternate acceptance for fittings supplied by Tube-Line.
B 83-005      ASME Nuclear Code Pumps and Spare          C      TVA: letter dated September 7, 1983 Parts Manufactured by the Hayward Tyler Pump Company                                      NRC: IR 390/85-03 and 391/85-04; NUREG/CR 5297 B 83-006      Nonconforming Material Supplied by        CI    TVA: letter dated February 2, 1984 Tube-Line Facilities 04    NRC: IR 390/391 84-03; NUREG/CR 4934
                                                                --------------------
NRC SER for both units dated September 23, 1991, provided an alternate acceptance for fittings supplied by Tube-Line.
Unit 2 Action: Implement as necessary.
Unit 2 Action: Implement as necessary.
                                                                ----------------------------------------------------------------------------------------
                                                                ----------------------------------------------------------------------------------------
REVISION 04 UPDATE:
REVISION 04 UPDATE:
NRC Inspection Report Nos. 50-390/90-02 and 50-391/90-02 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 III requirements.
NRC Inspection Report Nos. 50-390/90-02 and 50-391/90-02 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 III requirements.
Line 2,001: Line 1,570:
B 83-007      Apparently Fraudulent Products Sold by    C      TVA: letter dated March 22, 1984 Ray Miller, Inc.
B 83-007      Apparently Fraudulent Products Sold by    C      TVA: letter dated March 22, 1984 Ray Miller, Inc.
NRC: IR 390/85-03 and 391/85-04 B 83-008      Electrical Circuit Breakers With an        C      TVA: letter dated March 29, 1984 Undervoltage Trip Feature in Safety-Related Applications Other Than            NRC: IR 390/84-35 and 391/84-33 the Reactor Trip System B 84-001      Cracks in BWR Mark 1 Containment Vent    NA      Boiling Water Reactor Headers B 84-002      Failure of GE Type HFA Relays In Use In    C      TVA: letter dated July 10, 1984 Class 1E Safety Systems NRC: IR 390/391 84-42 and IR 390/84-77 and 391/84-54 Page 13 of 95                                              * = See last page for status code definition.
NRC: IR 390/85-03 and 391/85-04 B 83-008      Electrical Circuit Breakers With an        C      TVA: letter dated March 29, 1984 Undervoltage Trip Feature in Safety-Related Applications Other Than            NRC: IR 390/84-35 and 391/84-33 the Reactor Trip System B 84-001      Cracks in BWR Mark 1 Containment Vent    NA      Boiling Water Reactor Headers B 84-002      Failure of GE Type HFA Relays In Use In    C      TVA: letter dated July 10, 1984 Class 1E Safety Systems NRC: IR 390/391 84-42 and IR 390/84-77 and 391/84-54 Page 13 of 95                                              * = See last page for status code definition.
* ITEM                          TITLE                REV                        ADDITIONAL INFORMATION B 84-003      Refueling Cavity Water Seal              CI    Reviewed in IR 390/93-11.
 
ITEM                          TITLE                REV                        ADDITIONAL INFORMATION B 84-003      Refueling Cavity Water Seal              CI    Reviewed in IR 390/93-11.
Unit 2 Action: Ensure appropriate abnormal operating instructions (AOIs) are used for Unit 2.
Unit 2 Action: Ensure appropriate abnormal operating instructions (AOIs) are used for Unit 2.
B 85-001      Steam Binding of Auxiliary Feedwater      CI    TVA: letter dated January 27, 1986 Pumps NRC: IR 390/391 90-20
B 85-001      Steam Binding of Auxiliary Feedwater      CI    TVA: letter dated January 27, 1986 Pumps NRC: IR 390/391 90-20 NRC accepted approach in letter dated July 20, 1988, and reviewed response in Appendix EE of SSER16.
                                                              --------------------
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.
Unit 2 Action: Procedures and hardware will be in place to ensure recognition of indications of steam binding and maintenance of system operability until check valves are repaired and back leakage stopped.
B 85-002      Undervoltage Trip Attachment of          CI    Unit 2 Action: Install automatic shunt trip on the Westinghouse Westinghouse DB-50 Type Reactor Trip            DS-416 reactor trip breakers on Unit 2.
B 85-002      Undervoltage Trip Attachment of          CI    Unit 2 Action: Install automatic shunt trip on the Westinghouse Westinghouse DB-50 Type Reactor Trip            DS-416 reactor trip breakers on Unit 2.
Breakers B 85-003      Motor-Operated Valve Common Mode          C    Superseded by GL 89-10 Failures During Plant Transients Due to Improper Switch Settings B 86-001      Minimum Flow Logic Problems That        NA    Boiling Water Reactor Could Disable RHR Pumps B 86-002      Static "O" Ring Differential Pressure    C    TVA: letter dated November 20, 1986 Switches NRC: IR 390/391/90-24 B 86-003      Potential Failure of Multiple ECCS Pumps  C    TVA: letter dated November 14, 1986 Due to Single Failure of Air-Operated Valve in Minimum Flow Recirculation Line        NRC: IR 390/391/87-03 B 86-004      Defective Teletherapy Timer That May    NA    Does not apply to power reactor.
Breakers B 85-003      Motor-Operated Valve Common Mode          C    Superseded by GL 89-10 Failures During Plant Transients Due to Improper Switch Settings B 86-001      Minimum Flow Logic Problems That        NA    Boiling Water Reactor Could Disable RHR Pumps B 86-002      Static "O" Ring Differential Pressure    C    TVA: letter dated November 20, 1986 Switches NRC: IR 390/391/90-24 B 86-003      Potential Failure of Multiple ECCS Pumps  C    TVA: letter dated November 14, 1986 Due to Single Failure of Air-Operated Valve in Minimum Flow Recirculation Line        NRC: IR 390/391/87-03 B 86-004      Defective Teletherapy Timer That May    NA    Does not apply to power reactor.
Not Terminate Treatment Dose B 87-001      Thinning of Pipe Walls in Nuclear Power  C    TVA: letter dated September 18, 1987 Plants NRC: NUREG/CR 5287
Not Terminate Treatment Dose B 87-001      Thinning of Pipe Walls in Nuclear Power  C    TVA: letter dated September 18, 1987 Plants NRC: NUREG/CR 5287 Closed to GL 89-08 Page 14 of 95                                            * = See last page for status code definition.
                                                              --------------------
 
Closed to GL 89-08 Page 14 of 95                                            * = See last page for status code definition.
ITEM                        TITLE                  REV                          ADDITIONAL INFORMATION B 87-002      Fastener Testing to Determine            CI    TVA: letters dated April 15, 1988, July 6, 1988, Conformance with Applicable Material            September 12, 1988, and January 27, 1989 Specifications                            03 NRC: letter dated August 18, 1989 NRC closed in letter dated August 18, 1989.
* ITEM                        TITLE                  REV                          ADDITIONAL INFORMATION B 87-002      Fastener Testing to Determine            CI    TVA: letters dated April 15, 1988, July 6, 1988, Conformance with Applicable Material            September 12, 1988, and January 27, 1989 Specifications                            03 NRC: letter dated August 18, 1989
                                                              --------------------
NRC closed in letter dated August 18, 1989.
Unit 2 Action: Complete for Unit 2, using information used for Unit 1, as applicable.
Unit 2 Action: Complete for Unit 2, using information used for Unit 1, as applicable.
                                                              ----------------------------------------------------------------------------------------
                                                              ----------------------------------------------------------------------------------------
REVISION 03 UPDATE:
REVISION 03 UPDATE:
Unit 2 has completed fastener testing as required by this Bulletin.
Unit 2 has completed fastener testing as required by this Bulletin.
Line 2,030: Line 1,593:
Unit 2 Action: Implement program to prevent thermal stratification.
Unit 2 Action: Implement program to prevent thermal stratification.
Page 15 of 95                                            * = See last page for status code definition.
Page 15 of 95                                            * = See last page for status code definition.
* ITEM                          TITLE              REV                        ADDITIONAL INFORMATION B 88-009      Thimble Tube Thinning in Westinghouse  CI    Reviewed in Appendix EE of SSER16.
 
ITEM                          TITLE              REV                        ADDITIONAL INFORMATION B 88-009      Thimble Tube Thinning in Westinghouse  CI    Reviewed in Appendix EE of SSER16.
Reactors Unit 2 Action: TVA letter dated March 11, 1994, for both units committed to establish a program and inspect the thimble tubes during the first refueling outage.
Reactors Unit 2 Action: TVA letter dated March 11, 1994, for both units committed to establish a program and inspect the thimble tubes during the first refueling outage.
B 88-010      Nonconforming Molded-Case Circuit      CI    Unit 2 Action: Replace those circuits not traceable to a circuit Breakers                                      breaker manufacturer.
B 88-010      Nonconforming Molded-Case Circuit      CI    Unit 2 Action: Replace those circuits not traceable to a circuit Breakers                                      breaker manufacturer.
Line 2,039: Line 1,603:
B 89-002      Stress Corrosion Cracking of            CI    NRC reviewed in Appendix EE of SSER16.
B 89-002      Stress Corrosion Cracking of            CI    NRC reviewed in Appendix EE of SSER16.
High-Hardness Type 410 Stainless Steel Internal Preloaded Bolting in Anchor          Unit 2 Actions: Replace the flapper assembly hold-down bolts Darling Model S350W Swing Check              fabricated on the 14 (12 valves are installed) Atwood and Morrell Valves or Valves of Similar Nature            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.
High-Hardness Type 410 Stainless Steel Internal Preloaded Bolting in Anchor          Unit 2 Actions: Replace the flapper assembly hold-down bolts Darling Model S350W Swing Check              fabricated on the 14 (12 valves are installed) Atwood and Morrell Valves or Valves of Similar Nature            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 89-003      Potential Loss of Required Shutdown    CI    TVA: letter dated June 19, 1990 Margin During Refueling Operations NRC: IR 390/391 94-04 and letter dated June 22, 1990
B 89-003      Potential Loss of Required Shutdown    CI    TVA: letter dated June 19, 1990 Margin During Refueling Operations NRC: IR 390/391 94-04 and letter dated June 22, 1990 NRC acceptance letter dated June 22, 1990.
                                                            --------------------
NRC acceptance letter dated June 22, 1990.
Unit 2 Action: Ensure that requirements for fuel assembly configuration, fuel loading and training are included in Unit 2.
Unit 2 Action: Ensure that requirements for fuel assembly configuration, fuel loading and training are included in Unit 2.
B 90-001      Loss of Fill-Oil in Transmitters        CI    Unit 2 Action: Implement applicable recommendations from this Manufactured by Rosemount                    Bulletin including identification of potentially defective transmitters and an enhanced surveillance program which monitors transmitters for loss of fill oil.
B 90-001      Loss of Fill-Oil in Transmitters        CI    Unit 2 Action: Implement applicable recommendations from this Manufactured by Rosemount                    Bulletin including identification of potentially defective transmitters and an enhanced surveillance program which monitors transmitters for loss of fill oil.
B 90-002      Loss of Thermal Margin Caused by      NA    Boiling Water Reactor Channel Box Bow B 91-001      Reporting Loss of Criticality Safety  NA    Does not apply to power reactor.
B 90-002      Loss of Thermal Margin Caused by      NA    Boiling Water Reactor Channel Box Bow B 91-001      Reporting Loss of Criticality Safety  NA    Does not apply to power reactor.
Controls Page 16 of 95                                          * = See last page for status code definition.
Controls Page 16 of 95                                          * = See last page for status code definition.
* ITEM                        TITLE                  REV                          ADDITIONAL INFORMATION B 92-001      Failure of Thermo-Lag 330 Fire Barrier  NA      ----------------------------------------------------------------------------------------
 
ITEM                        TITLE                  REV                          ADDITIONAL INFORMATION B 92-001      Failure of Thermo-Lag 330 Fire Barrier  NA      ----------------------------------------------------------------------------------------
System to Maintain Cabling in Wide              ----------------------------------------------------------------------------------------
System to Maintain Cabling in Wide              ----------------------------------------------------------------------------------------
Cable Trays and Small Conduits Free      02 From Fire Damage                                REVISION 02 UPDATE:
Cable Trays and Small Conduits Free      02 From Fire Damage                                REVISION 02 UPDATE:
Line 2,054: Line 1,617:
B 93-001      Release of Patients After Brachytherapy  NA      Does not apply to power reactor.
B 93-001      Release of Patients After Brachytherapy  NA      Does not apply to power reactor.
Treatment with Remote Afterloading Devices B 93-002      Debris Plugging of Emergency Core        C      Boiling Water Reactor Cooling Suction Strainers 02    ----------------------------------------------------------------------------------------
Treatment with Remote Afterloading Devices B 93-002      Debris Plugging of Emergency Core        C      Boiling Water Reactor Cooling Suction Strainers 02    ----------------------------------------------------------------------------------------
                                                              ----------------------------------------------------------------------------------------
REVISION 02 UPDATE:
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.
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.
Line 2,060: Line 1,622:
Steel Packagings Used to Transport Uranium Hexafluoride B 95-001      Quality Assurance Program for            NA      Does not apply to power reactor.
Steel Packagings Used to Transport Uranium Hexafluoride B 95-001      Quality Assurance Program for            NA      Does not apply to power reactor.
Transportation of Radioactive Material B 95-002      Unexpected Clogging of a Residual Heat  NA      Boiling Water Reactor Removal Pump Strainer While Operating in Suppression Pool Cooling Mode Page 17 of 95                                            * = See last page for status code definition.
Transportation of Radioactive Material B 95-002      Unexpected Clogging of a Residual Heat  NA      Boiling Water Reactor Removal Pump Strainer While Operating in Suppression Pool Cooling Mode Page 17 of 95                                            * = See last page for status code definition.
* ITEM                      TITLE                REV                          ADDITIONAL INFORMATION B 96-001,    Control Rod Insertion Problems (PWR)    CI    NRC acceptance letter for Unit 1 dated July 22, 1996 - Initial first part                                                    response for Unit 2 on September 7, 2007.
 
ITEM                      TITLE                REV                          ADDITIONAL INFORMATION B 96-001,    Control Rod Insertion Problems (PWR)    CI    NRC acceptance letter for Unit 1 dated July 22, 1996 - Initial first part                                                    response for Unit 2 on September 7, 2007.
04 Unit 2 Action: Issue Emergency Operating Procedure.
04 Unit 2 Action: Issue Emergency Operating Procedure.
                                                              ----------------------------------------------------------------------------------------
                                                              ----------------------------------------------------------------------------------------
REVISION 02 UPDATE:
REVISION 02 UPDATE:
Unit 2 will load all new RFA-2 fuel for the initial fuel load.
Unit 2 will load all new RFA-2 fuel for the initial fuel load.
                                                              ----------------------------------------------------------------------------------------
                                                              ----------------------------------------------------------------------------------------
REVISION 03 UPDATE:
REVISION 03 UPDATE:
NRC issued the Safety Evaluation (corrected) for Bulletin 1996-001 on May 3, 2010.
NRC issued the Safety Evaluation (corrected) for Bulletin 1996-001 on May 3, 2010.
                                                              ----------------------------------------------------------------------------------------
                                                              ----------------------------------------------------------------------------------------
REVISION 04 UPDATE:
REVISION 04 UPDATE:
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
B 96-001,    Control Rod Insertion Problems (PWR)    CI    NRC acceptance letter for Unit 1 dated July 22, 1996 - Initial last part                                                    response for Unit 2 on September 7, 2007.
B 96-001,    Control Rod Insertion Problems (PWR)    CI    NRC acceptance letter for Unit 1 dated July 22, 1996 - Initial last part                                                    response for Unit 2 on September 7, 2007.
04 Unit 2 Action: and provide core map.
04 Unit 2 Action: and provide core map.
                                                              ----------------------------------------------------------------------------------------
                                                              ----------------------------------------------------------------------------------------
REVISION 03 UPDATE:
REVISION 03 UPDATE:
NRC issued the Safety Evaluation (corrected) for Bulletin 1996-001 on May 3, 2010.
NRC issued the Safety Evaluation (corrected) for Bulletin 1996-001 on May 3, 2010.
                                                              ----------------------------------------------------------------------------------------
                                                              ----------------------------------------------------------------------------------------
REVISION 04 UPDATE:
REVISION 04 UPDATE:
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
B 96-002      Movement of Heavy Loads over Spent      CI    NRC closure letter dated May 20, 1998.
B 96-002      Movement of Heavy Loads over Spent      CI    NRC closure letter dated May 20, 1998.
Fuel, Over Fuel in the Reactor, or Over Safety-Related Equipment                02    Unit 2 Action: Unit 2 Heavy Loads Program will be in compliance with NUREG-0612.
Fuel, Over Fuel in the Reactor, or Over Safety-Related Equipment                02    Unit 2 Action: Unit 2 Heavy Loads Program will be in compliance with NUREG-0612.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
NRC issued the Safety Evaluation for Bulletin 1996-002 on March 4, 2010.
NRC issued the Safety Evaluation for Bulletin 1996-002 on March 4, 2010.
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* ITEM                          TITLE                  REV                          ADDITIONAL INFORMATION B 96-003      Potential Plugging of ECCS Suction        NA      Boiling Water Reactor Strainers by Debris in BWRs B 96-004      Chemical, Galvanic, or Other Reactions in NA      Info Spent Fuel Storage and Transportation Casks B 97-001      Potential for Erroneous Calibration, Dose NA      Does not apply to power reactor.
 
ITEM                          TITLE                  REV                          ADDITIONAL INFORMATION B 96-003      Potential Plugging of ECCS Suction        NA      Boiling Water Reactor Strainers by Debris in BWRs B 96-004      Chemical, Galvanic, or Other Reactions in NA      Info Spent Fuel Storage and Transportation Casks B 97-001      Potential for Erroneous Calibration, Dose NA      Does not apply to power reactor.
Rate, or Radiation Exposure Measurements with Certain Victoreen Model 530 and 531SI Electrometer/Dosemeters B 97-002      Puncture Testing of Shipping Packages    NA      Does not apply to power reactor.
Rate, or Radiation Exposure Measurements with Certain Victoreen Model 530 and 531SI Electrometer/Dosemeters B 97-002      Puncture Testing of Shipping Packages    NA      Does not apply to power reactor.
Under 10 CFR Part 71 B 01-001      Circumferential Cracking of Reactor        CI    NRC acceptance letter dated November 20, 2001 (Unit 1) -
Under 10 CFR Part 71 B 01-001      Circumferential Cracking of Reactor        CI    NRC acceptance letter dated November 20, 2001 (Unit 1) -
Pressure Vessel (RPV) Head Penetration            Initial response for Unit 2 on September 7, 2007.
Pressure Vessel (RPV) Head Penetration            Initial response for Unit 2 on September 7, 2007.
Nozzles                                    04 Unit 2 Action: Perform baseline inspection.
Nozzles                                    04 Unit 2 Action: Perform baseline inspection.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Unit 2 Action: Perform baseline inspection. Evaluate or repair as necessary.
Unit 2 Action: Perform baseline inspection. Evaluate or repair as necessary.
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REVISION 03 UPDATE:
REVISION 03 UPDATE:
NRC issued the Safety Evaluation for Bulletin 2001-001 on June 30, 2010.
NRC issued the Safety Evaluation for Bulletin 2001-001 on June 30, 2010.
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REVISION 04 UPDATE:
REVISION 04 UPDATE:
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
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* ITEM                        TITLE                REV                          ADDITIONAL INFORMATION B 02-001      RPV Head Degradation and Reactor        CI    NRC review of Unit 1's 15 day response in letter dated Coolant Pressure Boundary Integrity            May 20, 2002 - Initial response for Unit 2 on 04    September 7, 2007.
 
ITEM                        TITLE                REV                          ADDITIONAL INFORMATION B 02-001      RPV Head Degradation and Reactor        CI    NRC review of Unit 1's 15 day response in letter dated Coolant Pressure Boundary Integrity            May 20, 2002 - Initial response for Unit 2 on 04    September 7, 2007.
Unit 2 Action: Perform baseline inspection.
Unit 2 Action: Perform baseline inspection.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Unit 2 Action: Perform baseline inspection. Evaluate or repair as necessary.
Unit 2 Action: Perform baseline inspection. Evaluate or repair as necessary.
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REVISION 03 UPDATE:
REVISION 03 UPDATE:
NRC issued the Safety Evaluation for Bulletin 2002-001 on June 30, 2010.
NRC issued the Safety Evaluation for Bulletin 2002-001 on June 30, 2010.
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REVISION 04 UPDATE:
REVISION 04 UPDATE:
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
Line 2,126: Line 1,667:
Nozzle Inspection Programs                    Initial response for Unit 2 on September 7, 2007.
Nozzle Inspection Programs                    Initial response for Unit 2 on September 7, 2007.
04 Unit 2 Action: Perform baseline inspection.
04 Unit 2 Action: Perform baseline inspection.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Unit 2 Action: Perform baseline inspection. Evaluate or repair as necessary.
Unit 2 Action: Perform baseline inspection. Evaluate or repair as necessary.
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REVISION 03 UPDATE:
REVISION 03 UPDATE:
NRC issued the Safety Evaluation for Bulletin 2002-002 on June 30, 2010.
NRC issued the Safety Evaluation for Bulletin 2002-002 on June 30, 2010.
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REVISION 04 UPDATE:
REVISION 04 UPDATE:
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
B 03-001      Potential Impact of Debris Blockage on NA      TVA: letter dated September 7, 2007 Emergency Sump Recirculation at PWRs Page 20 of 95                                          * = See last page for status code definition.
B 03-001      Potential Impact of Debris Blockage on NA      TVA: letter dated September 7, 2007 Emergency Sump Recirculation at PWRs Page 20 of 95                                          * = See last page for status code definition.
* ITEM                          TITLE                REV                          ADDITIONAL INFORMATION B 03-002      Leakage from RPV Lower Head              CI    NRC acceptance letter dated October 6, 2004 (Unit 1) - Initial Penetrations and Reactor Coolant                response for Unit 2 on September 7, 2007.
 
ITEM                          TITLE                REV                          ADDITIONAL INFORMATION B 03-002      Leakage from RPV Lower Head              CI    NRC acceptance letter dated October 6, 2004 (Unit 1) - Initial Penetrations and Reactor Coolant                response for Unit 2 on September 7, 2007.
Pressure Boundary Integrity (PWRs)      02 Unit 2 Action: Perform baseline inspection.
Pressure Boundary Integrity (PWRs)      02 Unit 2 Action: Perform baseline inspection.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
NRC issued the Safety Evaluation for Bulletin 2003-002 on January 21, 2010.
NRC issued the Safety Evaluation for Bulletin 2003-002 on January 21, 2010.
Unit 2 Action: Perform baseline inspection. Evaluate or repair as necessary.
Unit 2 Action: Perform baseline inspection. Evaluate or repair as necessary.
B 03-003      Potentially Deficient 1-inch Valves for NA      Does not apply to power reactor.
B 03-003      Potentially Deficient 1-inch Valves for NA      Does not apply to power reactor.
Uranium Hexaflouride Cylinders B 03-004      Rebaselining of Data in the Nuclear      C      TVA: letter dated December 18, 2003 Management and Safeguards System
Uranium Hexaflouride Cylinders B 03-004      Rebaselining of Data in the Nuclear      C      TVA: letter dated December 18, 2003 Management and Safeguards System Item concerns a multi-unit issue that was completed for both units.
                                                              --------------------
Item concerns a multi-unit issue that was completed for both units.
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* ITEM                          TITLE                REV                          ADDITIONAL INFORMATION B 04-001      Inspection of Alloy 82/182/600 Materials  CI    Initial response for Unit 2 on September 7, 2007.
 
ITEM                          TITLE                REV                          ADDITIONAL INFORMATION B 04-001      Inspection of Alloy 82/182/600 Materials  CI    Initial response for Unit 2 on September 7, 2007.
Used in the Fabrication of Pressurizer Penetrations and Steam Space Piping      04    Unit 2 Actions: Provide details of pressurizer and penetrations Connections at PWRs                              and apply Material Stress Improvement Process.
Used in the Fabrication of Pressurizer Penetrations and Steam Space Piping      04    Unit 2 Actions: Provide details of pressurizer and penetrations Connections at PWRs                              and apply Material Stress Improvement Process.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
TVA provided details of the pressurizer and penetrations on September 29, 2008. This letter committed to:
TVA provided details of the pressurizer and penetrations on September 29, 2008. This letter committed to:
Prior to placing the pressurizer in service, TVA will apply the Material Stress Improvement Process (MSIP) to the Pressurizer Power Operated Relief Valve connections, the safety relief valve connections, the spray line nozzle and surge line nozzle connections.
Prior to placing the pressurizer in service, TVA will apply the Material Stress Improvement Process (MSIP) to the Pressurizer Power Operated Relief Valve connections, the safety relief valve connections, the spray line nozzle and surge line nozzle connections.
TVA will perform a bare metal visual (BMV) inspection of the upper pressurizer Alloy 600 locations at the first refueling outage.
TVA will perform a bare metal visual (BMV) inspection of the upper pressurizer Alloy 600 locations at the first refueling outage.
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REVISION 03 UPDATE:
REVISION 03 UPDATE:
April 1, 2010, letter committed to:
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.
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.
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.
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REVISION 04 UPDATE:
REVISION 04 UPDATE:
NRC issued the Safety Evaluation for Bulletin 2004-001 on August 4, 2010.
NRC issued the Safety Evaluation for Bulletin 2004-001 on August 4, 2010.
Line 2,172: Line 1,699:
Item concerns a multi-unit issue that was completed for both units.
Item concerns a multi-unit issue that was completed for both units.
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* ITEM                        TITLE                  REV                        ADDITIONAL INFORMATION B 05-002      Emergency Preparedness and Response        C    TVA: letters dated January 20, 2006 and August 16, 2006.
 
Actions for Security-Based Events
ITEM                        TITLE                  REV                        ADDITIONAL INFORMATION B 05-002      Emergency Preparedness and Response        C    TVA: letters dated January 20, 2006 and August 16, 2006.
                                                              --------------------
Actions for Security-Based Events Item concerns a multi-unit issue that was completed for both units.
Item concerns a multi-unit issue that was completed for both units.
B 07-001      Security Officer Attentiveness            OV    Item concerns a multi-unit issue that was completed for both units.
B 07-001      Security Officer Attentiveness            OV    Item concerns a multi-unit issue that was completed for both units.
C 76-001      Crane Hoist Control Circuit Modifications  C    See B 76-007 for additional information.
C 76-001      Crane Hoist Control Circuit Modifications  C    See B 76-007 for additional information.
Line 2,184: Line 1,710:
C 77-001      Malfunctions of Limitorque Valve          NA    Info Operators C 77-002a    Potential Heavy Spring Flooding (CP)      NA    Item was applicable only to units with operating license at the time the item was issued.
C 77-001      Malfunctions of Limitorque Valve          NA    Info Operators C 77-002a    Potential Heavy Spring Flooding (CP)      NA    Item was applicable only to units with operating license at the time the item was issued.
C 77-003      Fire Inside a Motor Control Center        NA    Info Page 23 of 95                                            * = See last page for status code definition.
C 77-003      Fire Inside a Motor Control Center        NA    Info Page 23 of 95                                            * = See last page for status code definition.
* ITEM                        TITLE                  REV                      ADDITIONAL INFORMATION C 77-004      Inadequate Lock Assemblies              NA    Info C 77-005      Fluid Entrapment in Valve Bonnets        NA    Info C 77-006      Effects of Hydraulic Fluid on Electrical NA    Info Cables C 77-007      Short Period During Reactor Startup      NA    Boiling Water Reactor C 77-008      Failure of Feedwater Sample Probe        NA    Item was applicable only to units with operating license at the time the item was issued.
 
ITEM                        TITLE                  REV                      ADDITIONAL INFORMATION C 77-004      Inadequate Lock Assemblies              NA    Info C 77-005      Fluid Entrapment in Valve Bonnets        NA    Info C 77-006      Effects of Hydraulic Fluid on Electrical NA    Info Cables C 77-007      Short Period During Reactor Startup      NA    Boiling Water Reactor C 77-008      Failure of Feedwater Sample Probe        NA    Item was applicable only to units with operating license at the time the item was issued.
C 77-009      Improper Fuse Coordination in BWR        NA    Boiling Water Reactor Standby Liquid Control System Control Circuits C 77-010      Vacuum Conditions Resulting in Damage    NA    Item was applicable only to units with operating license at the to Liquid Process Tanks                        time the item was issued.
C 77-009      Improper Fuse Coordination in BWR        NA    Boiling Water Reactor Standby Liquid Control System Control Circuits C 77-010      Vacuum Conditions Resulting in Damage    NA    Item was applicable only to units with operating license at the to Liquid Process Tanks                        time the item was issued.
C 77-011      Leakage of Containment Isolation Valves  NA    Info with Resilient Seats C 77-012      Dropped Fuel Assemblies at BWR          NA    Boiling Water Reactor Facilities C 77-013      Reactor Safety Signals Negated During    NA    Info Testing C 77-014      Separation of Contaminated Water        NA    Info Systems from Noncontaminated Plant Systems C 77-015      Degradation of Fuel Oil Flow to the      NA    Info Emergency Diesel Generator C 77-016      Emergency Diesel Generator Electrical    NA    Info Trip Lock-Out Features Page 24 of 95                                            * = See last page for status code definition.
C 77-011      Leakage of Containment Isolation Valves  NA    Info with Resilient Seats C 77-012      Dropped Fuel Assemblies at BWR          NA    Boiling Water Reactor Facilities C 77-013      Reactor Safety Signals Negated During    NA    Info Testing C 77-014      Separation of Contaminated Water        NA    Info Systems from Noncontaminated Plant Systems C 77-015      Degradation of Fuel Oil Flow to the      NA    Info Emergency Diesel Generator C 77-016      Emergency Diesel Generator Electrical    NA    Info Trip Lock-Out Features Page 24 of 95                                            * = See last page for status code definition.
* ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION C 78-001      Loss of Well Logging Source              NA    Does not apply to power reactor.
 
ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION C 78-001      Loss of Well Logging Source              NA    Does not apply to power reactor.
C 78-002      Proper Lubricating Oil for Terry Turbines NA    Info C 78-003      Packaging Greater Than Type A            NA    Info Quantities of Low Specific Activity Radioactive Material for Transport C 78-004      Installation Errors That Could Prevent    NA    Info Closing of Fire Doors C 78-005      Inadvertent Safety Injection During      NA    Info Cooldown C 78-006      Potential Common Mode Flooding of        NA    Info ECCS Equipment Rooms at BWR Facilities C 78-007      Damaged Components of a                  NA    Info Bergen-Paterson Series 25000 Hydraulic Test Stand C 78-008      Environmental Qualification of            NA    Info Safety-Related Electrical Equipment at Nuclear Power Plants C 78-009      Arcing of General Electric Company        NA    Info Size 2 Contactors C 78-010      Control of Sealed Sources in Radiation    NA    Does not apply to power reactor.
C 78-002      Proper Lubricating Oil for Terry Turbines NA    Info C 78-003      Packaging Greater Than Type A            NA    Info Quantities of Low Specific Activity Radioactive Material for Transport C 78-004      Installation Errors That Could Prevent    NA    Info Closing of Fire Doors C 78-005      Inadvertent Safety Injection During      NA    Info Cooldown C 78-006      Potential Common Mode Flooding of        NA    Info ECCS Equipment Rooms at BWR Facilities C 78-007      Damaged Components of a                  NA    Info Bergen-Paterson Series 25000 Hydraulic Test Stand C 78-008      Environmental Qualification of            NA    Info Safety-Related Electrical Equipment at Nuclear Power Plants C 78-009      Arcing of General Electric Company        NA    Info Size 2 Contactors C 78-010      Control of Sealed Sources in Radiation    NA    Does not apply to power reactor.
Therapy C 78-011      Recirculation MG Set Overspeed Stops      NA    Boiling Water Reactor C 78-012      HPCI Turbine Control Valve Lift Rod      NA    Boiling Water Reactor Bending C 78-013      Inoperability of Service Water Pumps      NA    Info Page 25 of 95                                            * = See last page for status code definition.
Therapy C 78-011      Recirculation MG Set Overspeed Stops      NA    Boiling Water Reactor C 78-012      HPCI Turbine Control Valve Lift Rod      NA    Boiling Water Reactor Bending C 78-013      Inoperability of Service Water Pumps      NA    Info Page 25 of 95                                            * = See last page for status code definition.
* ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION C 78-014      HPCI Turbine Reversing Chamber Hold      NA    Boiling Water Reactor Down Bolting C 78-015      Tilting Disc Check Valves Fail to Close  NA    Info with Gravity in Vertical Position C 78-016      Limitorque Valve Actuators                NA    Info C 78-017      Inadequate Guard Training/Qualification  NA    Info and Falsified Training Records C 78-018      UL Fire Test                              NA    Info C 78-019      Manual Override (Bypass) of Safety        NA    Info System Actuation Signals C 79-001      Administration of Unauthorized Byproduct  NA    Does not apply to power reactor.
 
ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION C 78-014      HPCI Turbine Reversing Chamber Hold      NA    Boiling Water Reactor Down Bolting C 78-015      Tilting Disc Check Valves Fail to Close  NA    Info with Gravity in Vertical Position C 78-016      Limitorque Valve Actuators                NA    Info C 78-017      Inadequate Guard Training/Qualification  NA    Info and Falsified Training Records C 78-018      UL Fire Test                              NA    Info C 78-019      Manual Override (Bypass) of Safety        NA    Info System Actuation Signals C 79-001      Administration of Unauthorized Byproduct  NA    Does not apply to power reactor.
Material to Humans C 79-002      Failure of 120 Volt Vital AC Power        NA    Info Supplies C 79-003      Inadequate Guard Training -              NA    Info Qualification and Falsified Training Records C 79-004      Loose Locking Nut on Limitorque Valve    NA    Info Operators C 79-005      Moisture Leakage in Stranded Wire        NA    Info Conductors C 79-006      Failure to Use Syringe and Bottle Shields NA    Does not apply to power reactor.
Material to Humans C 79-002      Failure of 120 Volt Vital AC Power        NA    Info Supplies C 79-003      Inadequate Guard Training -              NA    Info Qualification and Falsified Training Records C 79-004      Loose Locking Nut on Limitorque Valve    NA    Info Operators C 79-005      Moisture Leakage in Stranded Wire        NA    Info Conductors C 79-006      Failure to Use Syringe and Bottle Shields NA    Does not apply to power reactor.
in Nuclear Medicine C 79-007      Unexpected Speed Increase of Reactor      NA    Boiling Water Reactor Recirculation MG Set Resulted in Reactor Power Increase Page 26 of 95                                            * = See last page for status code definition.
in Nuclear Medicine C 79-007      Unexpected Speed Increase of Reactor      NA    Boiling Water Reactor Recirculation MG Set Resulted in Reactor Power Increase Page 26 of 95                                            * = See last page for status code definition.
* ITEM                          TITLE                REV                      ADDITIONAL INFORMATION C 79-008      Attempted Extortion - Low Enriched      NA    Fuel facilities and operating reactors at the time the item was Uranium                                        issued C 79-009      Occurrences of Split or Punctured      NA    Info Regulator Diaphragms in Certain Self Contained Breathing Apparatus C 79-010      Pipefittings Manufactured from          NA    Info Unacceptable Material C 79-011      Design/Construction Interface Problem  NA    Info C 79-012      Potential Diesel Generator Turbocharger NA    Info Problem C 79-013      Replacement of Diesel Fire Pump        NA    Info Starting Contactors C 79-014      Unauthorized Procurement and            NA    Does not apply to power reactor.
 
ITEM                          TITLE                REV                      ADDITIONAL INFORMATION C 79-008      Attempted Extortion - Low Enriched      NA    Fuel facilities and operating reactors at the time the item was Uranium                                        issued C 79-009      Occurrences of Split or Punctured      NA    Info Regulator Diaphragms in Certain Self Contained Breathing Apparatus C 79-010      Pipefittings Manufactured from          NA    Info Unacceptable Material C 79-011      Design/Construction Interface Problem  NA    Info C 79-012      Potential Diesel Generator Turbocharger NA    Info Problem C 79-013      Replacement of Diesel Fire Pump        NA    Info Starting Contactors C 79-014      Unauthorized Procurement and            NA    Does not apply to power reactor.
Distribution of XE-133 C 79-015      Bursting of High Pressure Hose and      NA    Item was applicable only to units with operating license at the Malfunction of Relief Valve O-Ring in          time the item was issued.
Distribution of XE-133 C 79-015      Bursting of High Pressure Hose and      NA    Item was applicable only to units with operating license at the Malfunction of Relief Valve O-Ring in          time the item was issued.
Certain Self-Contained Breathing Apparatus C 79-016      Excessive Radiation Exposures to        NA    Does not apply to power reactor.
Certain Self-Contained Breathing Apparatus C 79-016      Excessive Radiation Exposures to        NA    Does not apply to power reactor.
Members of the General Public and a Radiographer C 79-017      Contact Problem in SB-12 Switches on    NA    Info General Electric Company Metalclad Circuit Breakers C 79-018      Proper Installation of Target Rock      NA    Boiling Water Reactor Safety-Relief Valves C 79-019      Loose Locking Devices on Ingersoll-Rand NA    Info Pumps C 79-020      Failure of GTE Sylvania Relay Type PM  NA    Info Bulletin 7305 Catalog 5U12-11-AC with a 120V AC Coil Page 27 of 95                                          * = See last page for status code definition.
Members of the General Public and a Radiographer C 79-017      Contact Problem in SB-12 Switches on    NA    Info General Electric Company Metalclad Circuit Breakers C 79-018      Proper Installation of Target Rock      NA    Boiling Water Reactor Safety-Relief Valves C 79-019      Loose Locking Devices on Ingersoll-Rand NA    Info Pumps C 79-020      Failure of GTE Sylvania Relay Type PM  NA    Info Bulletin 7305 Catalog 5U12-11-AC with a 120V AC Coil Page 27 of 95                                          * = See last page for status code definition.
* ITEM                          TITLE                    REV                      ADDITIONAL INFORMATION C 79-021      Prevention of Unplanned Releases of        NA      Info Radioactivity C 79-022      Stroke Times for Power Operated Relief      NA      Info Valves C 79-023      Motor Starters and Contactors Failed to      C      The Circular did not require a response.
 
ITEM                          TITLE                    REV                      ADDITIONAL INFORMATION C 79-021      Prevention of Unplanned Releases of        NA      Info Radioactivity C 79-022      Stroke Times for Power Operated Relief      NA      Info Valves C 79-023      Motor Starters and Contactors Failed to      C      The Circular did not require a response.
Operate 01    TVA reported a nonconformance under 10 CFR 50.55e on January 17, 1980, that four motor starters of this type had been located in the 480V control and auxiliary vent boards at WBN.
Operate 01    TVA reported a nonconformance under 10 CFR 50.55e on January 17, 1980, that four motor starters of this type had been located in the 480V control and auxiliary vent boards at WBN.
Gould factory representatives supervised the replacement of the carrier assemblies in accordance with the Gould instructions.
Gould factory representatives supervised the replacement of the carrier assemblies in accordance with the Gould instructions.
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NRC IR 50-390/83-15 and 50-391/83-11 reviewed and closed the associated nonconformance reports.
NRC IR 50-390/83-15 and 50-391/83-11 reviewed and closed the associated nonconformance reports.
C 80-001      Service Advice for GE Induction Disc        NA      Info Relays C 80-002      Nuclear Power Plant Staff Work Hours        NA      Info C 80-003      Protection from Toxic Gas Hazards          NA      Info C 80-004      Securing of Threaded Locking Devices on    NA      Info Safety-Related Equipment C 80-005      Emergency Diesel-Generator Lubricating      NA      Info Oil Addition and Onsite Supply Page 28 of 95                                                * = See last page for status code definition.
C 80-001      Service Advice for GE Induction Disc        NA      Info Relays C 80-002      Nuclear Power Plant Staff Work Hours        NA      Info C 80-003      Protection from Toxic Gas Hazards          NA      Info C 80-004      Securing of Threaded Locking Devices on    NA      Info Safety-Related Equipment C 80-005      Emergency Diesel-Generator Lubricating      NA      Info Oil Addition and Onsite Supply Page 28 of 95                                                * = See last page for status code definition.
* ITEM                          TITLE                REV                      ADDITIONAL INFORMATION C 80-006      Control and Accountability Systems for  NA    Does not apply to power reactor.
 
ITEM                          TITLE                REV                      ADDITIONAL INFORMATION C 80-006      Control and Accountability Systems for  NA    Does not apply to power reactor.
Implant Therapy Sources C 80-007      Problems with HPCI Turbine Oil System    NA    Boiling Water Reactor C 80-008      BWR Technical Specification              NA    Boiling Water Reactor Inconsistency - RPS Response Time C 80-009      Problems with Plant Internal            NA    Info Communications Systems C 80-010      Failure to Maintain Environmental        NA    Info Qualification of Equipment C 80-011      Emergency Diesel Generator Lube Oil      NA    Info Cooler Failures C 80-012      Valve-Shaft-to-Actuator Key May Fall Out NA    Info of Place when Mounted Below Horizontal Axis C 80-013      Grid Strap Damage in Westinghouse Fuel  NA    Info Assemblies C 80-014      Radioactive Contamination of Plant      NA    Info Demineralized Water System and Resultant Internal Contamination of Personnel C 80-015      Loss of Reactor Coolant Pump Cooling    NA    Info and Natural Circulation Cooldown C 80-016      Operational Deficiencies in Rosemount    NA    Info Model 510DU Trip Units and Model 1152 Pressure Transmitters C 80-017      Fuel Pin Damage Due to Water Jet from    NA    Info Baffle Plate Corner C 80-018      10 CFR 50.59 Safety Evaluations for      NA    Info Changes to Radioactive Waste Treatment Systems Page 29 of 95                                            * = See last page for status code definition.
Implant Therapy Sources C 80-007      Problems with HPCI Turbine Oil System    NA    Boiling Water Reactor C 80-008      BWR Technical Specification              NA    Boiling Water Reactor Inconsistency - RPS Response Time C 80-009      Problems with Plant Internal            NA    Info Communications Systems C 80-010      Failure to Maintain Environmental        NA    Info Qualification of Equipment C 80-011      Emergency Diesel Generator Lube Oil      NA    Info Cooler Failures C 80-012      Valve-Shaft-to-Actuator Key May Fall Out NA    Info of Place when Mounted Below Horizontal Axis C 80-013      Grid Strap Damage in Westinghouse Fuel  NA    Info Assemblies C 80-014      Radioactive Contamination of Plant      NA    Info Demineralized Water System and Resultant Internal Contamination of Personnel C 80-015      Loss of Reactor Coolant Pump Cooling    NA    Info and Natural Circulation Cooldown C 80-016      Operational Deficiencies in Rosemount    NA    Info Model 510DU Trip Units and Model 1152 Pressure Transmitters C 80-017      Fuel Pin Damage Due to Water Jet from    NA    Info Baffle Plate Corner C 80-018      10 CFR 50.59 Safety Evaluations for      NA    Info Changes to Radioactive Waste Treatment Systems Page 29 of 95                                            * = See last page for status code definition.
* ITEM                        TITLE                  REV                      ADDITIONAL INFORMATION C 80-019      Noncompliance with License                NA    Does not apply to power reactor.
 
ITEM                        TITLE                  REV                      ADDITIONAL INFORMATION C 80-019      Noncompliance with License                NA    Does not apply to power reactor.
Requirements for Medical Licensees C 80-020      Changes in Safe-Slab Tank Dimensions      NA    Info C 80-021      Regulation of Refueling Crews            NA    Item was applicable only to units with operating license at the time the item was issued.
Requirements for Medical Licensees C 80-020      Changes in Safe-Slab Tank Dimensions      NA    Info C 80-021      Regulation of Refueling Crews            NA    Item was applicable only to units with operating license at the time the item was issued.
C 80-022      Confirmation of Employee Qualifications  NA    Info C 80-023      Potential Defects in Beloit Power Systems NA    Info Emergency Generators C 80-024      AECL Teletherapy Unit Malfunction        NA    Does not apply to power reactor.
C 80-022      Confirmation of Employee Qualifications  NA    Info C 80-023      Potential Defects in Beloit Power Systems NA    Info Emergency Generators C 80-024      AECL Teletherapy Unit Malfunction        NA    Does not apply to power reactor.
Line 2,214: Line 1,747:
C 81-001      Design Problems Involving Indicating      NA    Info Pushbutton Switches Manufactured by Honeywell Incorporated C 81-002      Performance of NRC-Licensed Individuals  NA    Item was applicable only to units with operating license at the while on Duty                                    time the item was issued.
C 81-001      Design Problems Involving Indicating      NA    Info Pushbutton Switches Manufactured by Honeywell Incorporated C 81-002      Performance of NRC-Licensed Individuals  NA    Item was applicable only to units with operating license at the while on Duty                                    time the item was issued.
C 81-003      Inoperable Seismic Monitoring            NA    Info Instrumentation C 81-004      The Role of Shift Technical Advisors and  NA    Info Importance of Reporting Operational Events C 81-005      Self-Aligning Rod End Bushings for Pipe  NA    Info Supports C 81-006      Potential Deficiency Affecting Certain    NA    Info Foxboro 10 to 50 Milliampere Transmitters Page 30 of 95                                            * = See last page for status code definition.
C 81-003      Inoperable Seismic Monitoring            NA    Info Instrumentation C 81-004      The Role of Shift Technical Advisors and  NA    Info Importance of Reporting Operational Events C 81-005      Self-Aligning Rod End Bushings for Pipe  NA    Info Supports C 81-006      Potential Deficiency Affecting Certain    NA    Info Foxboro 10 to 50 Milliampere Transmitters Page 30 of 95                                            * = See last page for status code definition.
* ITEM                        TITLE                    REV                      ADDITIONAL INFORMATION C 81-007      Control of Radioactively Contaminated      NA      Info Material C 81-008      Foundation Materials                        NA      Info C 81-009      Containment Effluent Water that            NA      Info Bypasses Radioactivity Monitor C 81-010      Steam Voiding in the Reactor Coolant        NA      Item was applicable only to units with operating license at the System During Decay Heat Removal                    time the item was issued.
 
ITEM                        TITLE                    REV                      ADDITIONAL INFORMATION C 81-007      Control of Radioactively Contaminated      NA      Info Material C 81-008      Foundation Materials                        NA      Info C 81-009      Containment Effluent Water that            NA      Info Bypasses Radioactivity Monitor C 81-010      Steam Voiding in the Reactor Coolant        NA      Item was applicable only to units with operating license at the System During Decay Heat Removal                    time the item was issued.
Cooldown C 81-011      Inadequate Decay Heat Removal During        NA      Boiling Water Reactor Reactor Shutdown C 81-012      Inadequate Periodic Test Procedure of      NA      Info PWR Reactor Protection System C 81-013      Torque Switch Electrical Bypass Circuit      C      The Circular did not require a response.
Cooldown C 81-011      Inadequate Decay Heat Removal During        NA      Boiling Water Reactor Reactor Shutdown C 81-012      Inadequate Periodic Test Procedure of      NA      Info PWR Reactor Protection System C 81-013      Torque Switch Electrical Bypass Circuit      C      The Circular did not require a response.
for Safeguard Service Valve Motors 01    TVA reported a nonconformance under 10 CFR 50.55e on April 4, 1986 (NCR W367-P), that required closing torque switches were 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.
for Safeguard Service Valve Motors 01    TVA reported a nonconformance under 10 CFR 50.55e on April 4, 1986 (NCR W367-P), that required closing torque switches were 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.
C 81-014      Main Steam Isolation Valve Failures to      NA      Info Close C 81-015      Unnecessary Radiation Exposures to the      NA      Info Public and Workers During Events Involving Thickness and Level Measuring Devices GL 77-001    Intrusion Detection Systems Handbook        NA      Info GL 77-002    Fire Protection Functional Responsibilities NA      Info GL 77-003    Transmittal of NUREG-0321, A Study of      NA      Info the Nuclear Regulatory Commission Quality Assurance Program Page 31 of 95                                                * = See last page for status code definition.
C 81-014      Main Steam Isolation Valve Failures to      NA      Info Close C 81-015      Unnecessary Radiation Exposures to the      NA      Info Public and Workers During Events Involving Thickness and Level Measuring Devices GL 77-001    Intrusion Detection Systems Handbook        NA      Info GL 77-002    Fire Protection Functional Responsibilities NA      Info GL 77-003    Transmittal of NUREG-0321, A Study of      NA      Info the Nuclear Regulatory Commission Quality Assurance Program Page 31 of 95                                                * = See last page for status code definition.
* ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 77-004    Shipments of Contaminated Components        NA    Info From NRC Licensed Power Facilities to Vendors & Service Companies GL 77-005    Nonconformity of Addressees of Items        NA    Info Directed to the Office of Nuclear Reactor Regulation GL 77-006    Enclosing Questionnaire Related to          NA    Item was applicable only to units with operating license at the Steam Generators                                  time the item was issued.
 
ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 77-004    Shipments of Contaminated Components        NA    Info From NRC Licensed Power Facilities to Vendors & Service Companies GL 77-005    Nonconformity of Addressees of Items        NA    Info Directed to the Office of Nuclear Reactor Regulation GL 77-006    Enclosing Questionnaire Related to          NA    Item was applicable only to units with operating license at the Steam Generators                                  time the item was issued.
GL 77-007    Reliability of Standby Diesel Generator    NA    Item was applicable only to units with operating license at the Units                                              time the item was issued.
GL 77-007    Reliability of Standby Diesel Generator    NA    Item was applicable only to units with operating license at the Units                                              time the item was issued.
GL 77-008    Revised Intrusion Detection Handbook        NA    Info and Entry Control Systems Handbook GL 78-001    Correction to Letter of 12/15/77 [GL 77-07] NA    Item was applicable only to units with operating license at the time the item was issued.
GL 77-008    Revised Intrusion Detection Handbook        NA    Info and Entry Control Systems Handbook GL 78-001    Correction to Letter of 12/15/77 [GL 77-07] NA    Item was applicable only to units with operating license at the time the item was issued.
Line 2,227: Line 1,762:
GL 78-005    Internal Distribution of Correspondence    NA    Info
GL 78-005    Internal Distribution of Correspondence    NA    Info
               - Asking for Comments on Mass Mailing System GL 78-006    This GL was never issued.                  NA GL 78-007    This GL was never issued.                  NA GL 78-008    Enclosing NUREG-0408 Re Mark I              NA    Boiling Water Reactor Containments, and Granting Exemption from GDC 50 and Enclosing Sample Notice Page 32 of 95                                              * = See last page for status code definition.
               - Asking for Comments on Mass Mailing System GL 78-006    This GL was never issued.                  NA GL 78-007    This GL was never issued.                  NA GL 78-008    Enclosing NUREG-0408 Re Mark I              NA    Boiling Water Reactor Containments, and Granting Exemption from GDC 50 and Enclosing Sample Notice Page 32 of 95                                              * = See last page for status code definition.
* ITEM                        TITLE                  REV                      ADDITIONAL INFORMATION GL 78-009    Multiple-Subsequent Actuations of        NA    Boiling Water Reactor Safety/Relief Valves Following an Isolation Event GL 78-010    Guidance on Radiological Environmental    NA    Info Monitoring GL 78-011    Guidance on Spent Fuel Pool              NA    Info Modifications GL 78-012    Notice of Meeting Regarding              NA    Info Implementation of 10 CFR 73.55 Requirements and Status of Research GL 78-013    Forwarding of NUREG-0219                  NA    Info GL 78-014    Transmittal of Draft NUREG-0219 for      NA    Info Comment GL 78-015    Request for Information on Control of    NA    See GL 81-007.
 
ITEM                        TITLE                  REV                      ADDITIONAL INFORMATION GL 78-009    Multiple-Subsequent Actuations of        NA    Boiling Water Reactor Safety/Relief Valves Following an Isolation Event GL 78-010    Guidance on Radiological Environmental    NA    Info Monitoring GL 78-011    Guidance on Spent Fuel Pool              NA    Info Modifications GL 78-012    Notice of Meeting Regarding              NA    Info Implementation of 10 CFR 73.55 Requirements and Status of Research GL 78-013    Forwarding of NUREG-0219                  NA    Info GL 78-014    Transmittal of Draft NUREG-0219 for      NA    Info Comment GL 78-015    Request for Information on Control of    NA    See GL 81-007.
Heavy Loads Near Spent Fuel GL 78-016    Request for Information on Control of    NA    Info Heavy Loads Near Spent Fuel Pools GL 78-017    Corrected Letter on Heavy Loads Over      NA    Info Spent Fuel GL 78-018    Corrected Letter on Heavy Loads Over      NA    Duplicate of GL 81-007 Spent Fuel GL 78-019    Enclosing Sandia Report SAND 77-0777,    NA    Info Barrier Technology Handbook GL 78-020    Enclosing - A Systematic Approach to    NA    Info the Conceptual Design of Physical Protection Systems for Nuclear Facilities GL 78-021    Transmitting NUREG/CR-0181,              NA    Info Concerning Barrier and Penetration Data Needed for Physical Security System Assessment Page 33 of 95                                            * = See last page for status code definition.
Heavy Loads Near Spent Fuel GL 78-016    Request for Information on Control of    NA    Info Heavy Loads Near Spent Fuel Pools GL 78-017    Corrected Letter on Heavy Loads Over      NA    Info Spent Fuel GL 78-018    Corrected Letter on Heavy Loads Over      NA    Duplicate of GL 81-007 Spent Fuel GL 78-019    Enclosing Sandia Report SAND 77-0777,    NA    Info Barrier Technology Handbook GL 78-020    Enclosing - A Systematic Approach to    NA    Info the Conceptual Design of Physical Protection Systems for Nuclear Facilities GL 78-021    Transmitting NUREG/CR-0181,              NA    Info Concerning Barrier and Penetration Data Needed for Physical Security System Assessment Page 33 of 95                                            * = See last page for status code definition.
* ITEM                          TITLE                REV                      ADDITIONAL INFORMATION GL 78-022    Revision to Intrusion Detection Systems NA    Info and Entry Control Systems Handbooks and Nuclear Safeguards Technology Handbook GL 78-023    Manpower Requirements for Operating    NA    Info Reactors GL 78-024    Model Appendix I Technical              NA    Boiling Water Reactor Specifications and Submittal Schedule For BWRs GL 78-025    This GL was never issued.              NA GL 78-026    Excessive Control Rod Guide Tube Wear  NA    Applies only to Babcock and Wilcox designed plants GL 78-027    Forwarding of NUREG-0181                NA    Info GL 78-028    Forwarding pages omitted from 07/11/78  NA    Boiling Water Reactor letter [GL 78-24]
 
ITEM                          TITLE                REV                      ADDITIONAL INFORMATION GL 78-022    Revision to Intrusion Detection Systems NA    Info and Entry Control Systems Handbooks and Nuclear Safeguards Technology Handbook GL 78-023    Manpower Requirements for Operating    NA    Info Reactors GL 78-024    Model Appendix I Technical              NA    Boiling Water Reactor Specifications and Submittal Schedule For BWRs GL 78-025    This GL was never issued.              NA GL 78-026    Excessive Control Rod Guide Tube Wear  NA    Applies only to Babcock and Wilcox designed plants GL 78-027    Forwarding of NUREG-0181                NA    Info GL 78-028    Forwarding pages omitted from 07/11/78  NA    Boiling Water Reactor letter [GL 78-24]
GL 78-029    Notice of PWR Steam Generator          NA    Info Conference GL 78-030    Forwarding of NUREG-0219                NA    Info GL 78-031    Notice of Steam Generator Conference    NA    Info Agenda GL 78-032    Reactor Protection System Power        NA    Boiling Water Reactor Supplies GL 78-033    Meeting Schedule and Locations For      NA    Info Upgraded Guard Qualification GL 78-034    Reactor Vessel Atypical Weld Material    C    See B 78-12.
GL 78-029    Notice of PWR Steam Generator          NA    Info Conference GL 78-030    Forwarding of NUREG-0219                NA    Info GL 78-031    Notice of Steam Generator Conference    NA    Info Agenda GL 78-032    Reactor Protection System Power        NA    Boiling Water Reactor Supplies GL 78-033    Meeting Schedule and Locations For      NA    Info Upgraded Guard Qualification GL 78-034    Reactor Vessel Atypical Weld Material    C    See B 78-12.
Page 34 of 95                                          * = See last page for status code definition.
Page 34 of 95                                          * = See last page for status code definition.
* ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 78-035    Regional Meetings to Discuss Upgraded    NA    Info Guard Qualifications GL 78-036    Cessation of Plutonium Shipments by Air  NA    Does not apply to power reactor.
 
ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 78-035    Regional Meetings to Discuss Upgraded    NA    Info Guard Qualifications GL 78-036    Cessation of Plutonium Shipments by Air  NA    Does not apply to power reactor.
Except In NRC Approved Containers GL 78-037    Revised Meeting Schedule & Locations      NA    Info For Upgraded Guard Qualifications GL 78-038    Forwarding of 2 Tables of Appendix I,    NA    Item was applicable only to units with operating license at the Draft Radiological Effluent Technical            time the item was issued.
Except In NRC Approved Containers GL 78-037    Revised Meeting Schedule & Locations      NA    Info For Upgraded Guard Qualifications GL 78-038    Forwarding of 2 Tables of Appendix I,    NA    Item was applicable only to units with operating license at the Draft Radiological Effluent Technical            time the item was issued.
Specifications, PWR, and NUREG-0133 GL 78-039    Forwarding of 2 Tables of Appendix I,    NA    Boiling Water Reactor Draft Radiological Effluent Technical Specifications, BWR, and NUREG-0133 GL 78-040    Training & Qualification Program          NA    Info Workshops GL 78-041    Mark II Generic Acceptance Criteria For  NA    Boiling Water Reactor Lead Plants GL 78-042    Training and Qualification Program        NA    Info Workshops GL 79-001    Interservice Procedures for Instructional NA    Info Systems Development - TRADOC GL 79-002    Transmitting Rev. to Entry Control        NA    Info Systems Handbook (SAND 77-1033),
Specifications, PWR, and NUREG-0133 GL 78-039    Forwarding of 2 Tables of Appendix I,    NA    Boiling Water Reactor Draft Radiological Effluent Technical Specifications, BWR, and NUREG-0133 GL 78-040    Training & Qualification Program          NA    Info Workshops GL 78-041    Mark II Generic Acceptance Criteria For  NA    Boiling Water Reactor Lead Plants GL 78-042    Training and Qualification Program        NA    Info Workshops GL 79-001    Interservice Procedures for Instructional NA    Info Systems Development - TRADOC GL 79-002    Transmitting Rev. to Entry Control        NA    Info Systems Handbook (SAND 77-1033),
Intrusion Detection Handbook (SAND 76-0554), and Barrier Penetration Database GL 79-003    Offsite Dose Calculation Manual          NA    Info GL 79-004    Referencing 4/14/78 Letter -              NA    Info Modifications to NRC Guidance "Review and Acceptance of Spent Fuel Pool Storage and Handling" GL 79-005    Information Relating to Categorization of NA    Info Recent Regulatory Guides by the Regulatory Requirements Review Committee Page 35 of 95                                            * = See last page for status code definition.
Intrusion Detection Handbook (SAND 76-0554), and Barrier Penetration Database GL 79-003    Offsite Dose Calculation Manual          NA    Info GL 79-004    Referencing 4/14/78 Letter -              NA    Info Modifications to NRC Guidance "Review and Acceptance of Spent Fuel Pool Storage and Handling" GL 79-005    Information Relating to Categorization of NA    Info Recent Regulatory Guides by the Regulatory Requirements Review Committee Page 35 of 95                                            * = See last page for status code definition.
* ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 79-006    Contents of the Offsite Dose Calculation  NA    Info Manual GL 79-007    Seismic (SSE) and LOCA Responses          NA    Info (NUREG-0484)
 
ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 79-006    Contents of the Offsite Dose Calculation  NA    Info Manual GL 79-007    Seismic (SSE) and LOCA Responses          NA    Info (NUREG-0484)
GL 79-008    Amendment to 10 CFR 73.55                  NA    Info GL 79-009    Staff Evaluation of Interim                NA    Boiling Water Reactor Multiple-Consecutive Safety-Relief Valve Actuations GL 79-010    Transmitting Regulatory Guide 2.6 for      NA    Does not apply to power reactor.
GL 79-008    Amendment to 10 CFR 73.55                  NA    Info GL 79-009    Staff Evaluation of Interim                NA    Boiling Water Reactor Multiple-Consecutive Safety-Relief Valve Actuations GL 79-010    Transmitting Regulatory Guide 2.6 for      NA    Does not apply to power reactor.
Comment GL 79-011    Transmitting "Summary of Operating        NA    Info Experience with Recalculating Steam Generators, January 1979," NUREG-0523 GL 79-012    ATWS - Enclosing Letter to GE, with        NA    Info NUREG-0460, Vol. 3 GL 79-013    Schedule for Implementation and            NA    Info Resolution of Mark I Containment Long Term Program GL 79-014    Pipe Crack Study Group - Enclosing        NA    Info NUREG-0531 and Notice GL 79-015    Steam Generators - Enclosing Summary      NA    Info of Operating Experience with Recirculating Steam Generators, NUREG-0523 GL 79-016    Meeting Re Implementation of Physical      NA    Info Security Requirements GL 79-017    Reliability of Onsite Diesel Generators at NA    Info Light Water Reactors GL 79-018    Westinghouse Two-Loop NSSS                NA    Addressed to specific plant(s).
Comment GL 79-011    Transmitting "Summary of Operating        NA    Info Experience with Recalculating Steam Generators, January 1979," NUREG-0523 GL 79-012    ATWS - Enclosing Letter to GE, with        NA    Info NUREG-0460, Vol. 3 GL 79-013    Schedule for Implementation and            NA    Info Resolution of Mark I Containment Long Term Program GL 79-014    Pipe Crack Study Group - Enclosing        NA    Info NUREG-0531 and Notice GL 79-015    Steam Generators - Enclosing Summary      NA    Info of Operating Experience with Recirculating Steam Generators, NUREG-0523 GL 79-016    Meeting Re Implementation of Physical      NA    Info Security Requirements GL 79-017    Reliability of Onsite Diesel Generators at NA    Info Light Water Reactors GL 79-018    Westinghouse Two-Loop NSSS                NA    Addressed to specific plant(s).
Page 36 of 95                                              * = See last page for status code definition.
Page 36 of 95                                              * = See last page for status code definition.
* ITEM                        TITLE                  REV                      ADDITIONAL INFORMATION GL 79-019    NRC Staff Review of Responses to        NA      Addressed to specific plant(s).
 
ITEM                        TITLE                  REV                      ADDITIONAL INFORMATION GL 79-019    NRC Staff Review of Responses to        NA      Addressed to specific plant(s).
Bs 79-06 and 79-06a GL 79-020    Cracking in Feedwater Lines              C      See B 79-13.
Bs 79-06 and 79-06a GL 79-020    Cracking in Feedwater Lines              C      See B 79-13.
GL 79-021    Enclosing NUREG/CR-0660,                NA      Info Enhancement of on Site Emergency Diesel Generator Reliability" GL 79-022    Enclosing NUREG-0560, "Staff Report on  NA      Applies only to Babcock and Wilcox designed plants the Generic Assessment of Feedwater Transients in PWRs Designed by B&W GL 79-023    NRC Staff Review of Responses to B 79-  NA      Boiling Water Reactor 08 GL 79-024    Multiple Equipment Failures in          NA      GL 79-24 provided a discussion of an inadvertent reactor scram Safety-Related Systems                          and safety injection during monthly surveillance tests of the 01    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.
GL 79-021    Enclosing NUREG/CR-0660,                NA      Info Enhancement of on Site Emergency Diesel Generator Reliability" GL 79-022    Enclosing NUREG-0560, "Staff Report on  NA      Applies only to Babcock and Wilcox designed plants the Generic Assessment of Feedwater Transients in PWRs Designed by B&W GL 79-023    NRC Staff Review of Responses to B 79-  NA      Boiling Water Reactor 08 GL 79-024    Multiple Equipment Failures in          NA      GL 79-24 provided a discussion of an inadvertent reactor scram Safety-Related Systems                          and safety injection during monthly surveillance tests of the 01    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.
GL 79-025    Information Required to Review          NA      Info Corporate Capabilities GL 79-026    Upgraded Standard Technical              NA      Info Specification Bases Program GL 79-027    Operability Testing of Relief and Safety NA      Boiling Water Reactor Relief Valves GL 79-028    Evaluation of Semi-Scale Small Break    NA      Info Experiment GL 79-029    Transmitting NUREG-0473, Revision 2,    NA      Info Draft Radiological Effluent Technical Specifications Page 37 of 95                                            * = See last page for status code definition.
GL 79-025    Information Required to Review          NA      Info Corporate Capabilities GL 79-026    Upgraded Standard Technical              NA      Info Specification Bases Program GL 79-027    Operability Testing of Relief and Safety NA      Boiling Water Reactor Relief Valves GL 79-028    Evaluation of Semi-Scale Small Break    NA      Info Experiment GL 79-029    Transmitting NUREG-0473, Revision 2,    NA      Info Draft Radiological Effluent Technical Specifications Page 37 of 95                                            * = See last page for status code definition.
* ITEM                        TITLE                  REV                      ADDITIONAL INFORMATION GL 79-030    Transmitting NUREG-0472, Revision 2,      NA    Info Draft Radiological Technical Specifications GL 79-031    Submittal of Copies of Response to        NA    Info 6/29/79 NRC Request [79-25]
 
ITEM                        TITLE                  REV                      ADDITIONAL INFORMATION GL 79-030    Transmitting NUREG-0472, Revision 2,      NA    Info Draft Radiological Technical Specifications GL 79-031    Submittal of Copies of Response to        NA    Info 6/29/79 NRC Request [79-25]
GL 79-032    Transmitting NUREG-0578, "TMI-2          NA    Info Lessons Learned" GL 79-033    Transmitting NUREG-0576, Security        NA    Info Training and Qualification Plans GL 79-034    New Physical Security Plans              NA    Does not apply to power reactor.
GL 79-032    Transmitting NUREG-0578, "TMI-2          NA    Info Lessons Learned" GL 79-033    Transmitting NUREG-0576, Security        NA    Info Training and Qualification Plans GL 79-034    New Physical Security Plans              NA    Does not apply to power reactor.
(FR 43280-285)
(FR 43280-285)
Line 2,253: Line 1,794:
Accident GL 79-041    Compliance with 40 CFR 190, EPA          NA    Info Uranium Fuel Cycle Standard GL 79-042    Potentially Unreviewed Safety Question    NA    Item was applicable only to units with operating license at the on Interaction Between Non-Safety Grade          time the item was issued.
Accident GL 79-041    Compliance with 40 CFR 190, EPA          NA    Info Uranium Fuel Cycle Standard GL 79-042    Potentially Unreviewed Safety Question    NA    Item was applicable only to units with operating license at the on Interaction Between Non-Safety Grade          time the item was issued.
Systems and Safety Grade Systems Page 38 of 95                                            * = See last page for status code definition.
Systems and Safety Grade Systems Page 38 of 95                                            * = See last page for status code definition.
* ITEM                        TITLE                  REV                      ADDITIONAL INFORMATION GL 79-043    Reactor Cavity Seal Ring Generic Issue  NA      Addressed to specific plant(s).
 
ITEM                        TITLE                  REV                      ADDITIONAL INFORMATION GL 79-043    Reactor Cavity Seal Ring Generic Issue  NA      Addressed to specific plant(s).
GL 79-044    Referencing 6/29/79 Letter Re Multiple  NA      Item was applicable only to units with operating license at the Equipment Failures                              time the item was issued.
GL 79-044    Referencing 6/29/79 Letter Re Multiple  NA      Item was applicable only to units with operating license at the Equipment Failures                              time the item was issued.
GL 79-045    Transmittal of Reports Regarding Foreign NA      Info Reactor Operating Experiences GL 79-046    Containment Purge and Venting During    NA      Item was applicable only to units with operating license at the Normal Operation - Guidelines for Valve          time the item was issued.
GL 79-045    Transmittal of Reports Regarding Foreign NA      Info Reactor Operating Experiences GL 79-046    Containment Purge and Venting During    NA      Item was applicable only to units with operating license at the Normal Operation - Guidelines for Valve          time the item was issued.
Line 2,260: Line 1,802:
GL 79-053    ATWS                                    NA      Info GL 79-054    Containment Purging and Venting During  NA      Addressed to specific plant(s).
GL 79-053    ATWS                                    NA      Info GL 79-054    Containment Purging and Venting During  NA      Addressed to specific plant(s).
Normal Operation Page 39 of 95                                            * = See last page for status code definition.
Normal Operation Page 39 of 95                                            * = See last page for status code definition.
* ITEM                        TITLE                  REV                      ADDITIONAL INFORMATION GL 79-055    Summary of Meeting Held on                NA      Info October 12, 1979 to Discuss Responses to Bulletins 79-05C and 79-06C and HPI Termination Criteria GL 79-056    Discussion of Lessons Learned Short      NA      Item was applicable only to units with operating license at the Term Requirements                                time the item was issued.
 
ITEM                        TITLE                  REV                      ADDITIONAL INFORMATION GL 79-055    Summary of Meeting Held on                NA      Info October 12, 1979 to Discuss Responses to Bulletins 79-05C and 79-06C and HPI Termination Criteria GL 79-056    Discussion of Lessons Learned Short      NA      Item was applicable only to units with operating license at the Term Requirements                                time the item was issued.
GL 79-057    Acceptance Criteria for Mark I Long Term  NA      Boiling Water Reactor Program GL 79-058    ECCS Calculations on Fuel Cladding        NA      Item was applicable only to units with operating license at the time the item was issued.
GL 79-057    Acceptance Criteria for Mark I Long Term  NA      Boiling Water Reactor Program GL 79-058    ECCS Calculations on Fuel Cladding        NA      Item was applicable only to units with operating license at the time the item was issued.
GL 79-059    This GL was never issued.                NA GL 79-060    Discussion of Lessons Learned Short      NA      Info Term Requirements GL 79-061    Discussion of Lessons Learned Short      NA      Info Term Requirements GL 79-062    ECCS Calculations on Fuel Cladding        NA      Item was applicable only to units with operating license at the time the item was issued.
GL 79-059    This GL was never issued.                NA GL 79-060    Discussion of Lessons Learned Short      NA      Info Term Requirements GL 79-061    Discussion of Lessons Learned Short      NA      Info Term Requirements GL 79-062    ECCS Calculations on Fuel Cladding        NA      Item was applicable only to units with operating license at the time the item was issued.
Line 2,267: Line 1,810:
GL 79-065    Radiological Environmental Monitoring    NA      Info Program Requirements - Enclosing Branch Technical Position, Revision 1 GL 79-066    Additional Information Re 11/09/79 Letter NA      Info on ECCS Calculations [GL 79-62]
GL 79-065    Radiological Environmental Monitoring    NA      Info Program Requirements - Enclosing Branch Technical Position, Revision 1 GL 79-066    Additional Information Re 11/09/79 Letter NA      Info on ECCS Calculations [GL 79-62]
GL 79-067    Estimates for Evacuation of Various      NA      Info Areas Around Nuclear Power Reactors Page 40 of 95                                              * = See last page for status code definition.
GL 79-067    Estimates for Evacuation of Various      NA      Info Areas Around Nuclear Power Reactors Page 40 of 95                                              * = See last page for status code definition.
* ITEM                        TITLE                  REV                      ADDITIONAL INFORMATION GL 79-068    Audit of Small Break LOCA Guidelines      NA    Info GL 79-069    Cladding Rupture, Swelling, and Coolant  NA    Info Blockage as a Result of a Reactor Accident GL 79-070    Environmental Monitoring for Direct      NA    Info Radiation GL 80-001    NUREG-0630, "Cladding, Swelling and      NA    Info Rupture - Models For LOCA Analysis" GL 80-002    QA Requirements Regarding Diesel          C    TVA: FSAR 9.5.4.2 Generator Fuel Oil GL 80-003    BWR Control Rod Failures                  NA    Boiling Water Reactor GL 80-004    B 80-01, Operability of ADS Valve        NA    Boiling Water Reactor Pneumatic Supply GL 80-005    B 79-01b, Environmental Qualification of NA    Info Class 1E Equipment GL 80-006    Issuance of NUREG-0313, Rev 1,            NA    Boiling Water Reactor "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping GL 80-007    This GL was never issued.                NA GL 80-008    B 80-02. "Inadequate Quality Assurance    NA    Boiling Water Reactor for Nuclear Supplied Equipment" GL 80-009    Low Level Radioactive Waste Disposal      NA    Item was applicable only to units with operating license at the time the item was issued.
 
ITEM                        TITLE                  REV                      ADDITIONAL INFORMATION GL 79-068    Audit of Small Break LOCA Guidelines      NA    Info GL 79-069    Cladding Rupture, Swelling, and Coolant  NA    Info Blockage as a Result of a Reactor Accident GL 79-070    Environmental Monitoring for Direct      NA    Info Radiation GL 80-001    NUREG-0630, "Cladding, Swelling and      NA    Info Rupture - Models For LOCA Analysis" GL 80-002    QA Requirements Regarding Diesel          C    TVA: FSAR 9.5.4.2 Generator Fuel Oil GL 80-003    BWR Control Rod Failures                  NA    Boiling Water Reactor GL 80-004    B 80-01, Operability of ADS Valve        NA    Boiling Water Reactor Pneumatic Supply GL 80-005    B 79-01b, Environmental Qualification of NA    Info Class 1E Equipment GL 80-006    Issuance of NUREG-0313, Rev 1,            NA    Boiling Water Reactor "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping GL 80-007    This GL was never issued.                NA GL 80-008    B 80-02. "Inadequate Quality Assurance    NA    Boiling Water Reactor for Nuclear Supplied Equipment" GL 80-009    Low Level Radioactive Waste Disposal      NA    Item was applicable only to units with operating license at the time the item was issued.
GL 80-010    Issuance of NUREG-0588, "Interim Staff    NA    Info Position On Equipment Qualifications of Safety-Related Electrical Equipment Page 41 of 95                                            * = See last page for status code definition.
GL 80-010    Issuance of NUREG-0588, "Interim Staff    NA    Info Position On Equipment Qualifications of Safety-Related Electrical Equipment Page 41 of 95                                            * = See last page for status code definition.
* ITEM                          TITLE                  REV                          ADDITIONAL INFORMATION GL 80-011    B 80-03, Loss of Charcoal From            C      GL 80-11 transmitted Bulletin 80-03. TVA responded to B 80-03 Standard Type II, 2 Inch, Tray Absorber            on March 21, 1980. See B 80-03 for further information.
 
ITEM                          TITLE                  REV                          ADDITIONAL INFORMATION GL 80-011    B 80-03, Loss of Charcoal From            C      GL 80-11 transmitted Bulletin 80-03. TVA responded to B 80-03 Standard Type II, 2 Inch, Tray Absorber            on March 21, 1980. See B 80-03 for further information.
Cells                                      01 GL 80-012    B 80-04, Analysis of a PWR Main Steam    NA      Info Line Break With Continued Feedwater Addition GL 80-013    Qualification of Safety Related Electrical NA      Item was applicable only to units with operating license at the Equipment                                          time the item was issued.
Cells                                      01 GL 80-012    B 80-04, Analysis of a PWR Main Steam    NA      Info Line Break With Continued Feedwater Addition GL 80-013    Qualification of Safety Related Electrical NA      Item was applicable only to units with operating license at the Equipment                                          time the item was issued.
GL 80-014    LWR Primary Coolant System Pressure        S      TVA: FSAR 5.2.7.4 Isolation Valves 02    NRC: 1.14.2 of SSER 6
GL 80-014    LWR Primary Coolant System Pressure        S      TVA: FSAR 5.2.7.4 Isolation Valves 02    NRC: 1.14.2 of SSER 6 NRC reviewed in 1.14.2 of SSER6.
                                                                ----------
NRC reviewed in 1.14.2 of SSER6.
Unit 2 Action: Incorporate guidance into Technical Specifications.
Unit 2 Action: Incorporate guidance into Technical Specifications.
                                                                ----------------------------------------------------------------------------------------
                                                                ----------------------------------------------------------------------------------------
REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
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.
TS Surveillance Requirement 3.4.13.1 verifies RCS operational leakage by performance of an RCS water inventory balance.
GL 80-015    Request for Additional Management and      NA      Info Technical Resources Information GL 80-016    B 79-01b, Environmental Qualification of  NA      Info Class 1E Equipment GL 80-017    Modifications to BWR Control Rod Drive    NA      Boiling Water Reactor Systems GL 80-018    Crystal River 3 Reactor Trip From          NA      Applies only to Babcock and Wilcox designed plants Approximately 100% Full Power GL 80-019    Resolution of Enhanced Fission Gas        NA      Info Release Concern GL 80-020    Actions Required From OL Applicants of    NA      Info NSSS Designs by W and CE Resulting From NRC B&O Task Force Review of TMI2 Accident Page 42 of 95                                              * = See last page for status code definition.
GL 80-015    Request for Additional Management and      NA      Info Technical Resources Information GL 80-016    B 79-01b, Environmental Qualification of  NA      Info Class 1E Equipment GL 80-017    Modifications to BWR Control Rod Drive    NA      Boiling Water Reactor Systems GL 80-018    Crystal River 3 Reactor Trip From          NA      Applies only to Babcock and Wilcox designed plants Approximately 100% Full Power GL 80-019    Resolution of Enhanced Fission Gas        NA      Info Release Concern GL 80-020    Actions Required From OL Applicants of    NA      Info NSSS Designs by W and CE Resulting From NRC B&O Task Force Review of TMI2 Accident Page 42 of 95                                              * = See last page for status code definition.
* ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 80-021    B 80-05, Vacuum Condition Resulting in    CI    Closed in IR 50-390/84-59 and 50-391/84-45.
 
ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 80-021    B 80-05, Vacuum Condition Resulting in    CI    Closed in IR 50-390/84-59 and 50-391/84-45.
Damage to Chemical Volume Control System Holdup Tanks                              Unit 2 Action: Complete surveillance procedures for Unit 2.
Damage to Chemical Volume Control System Holdup Tanks                              Unit 2 Action: Complete surveillance procedures for Unit 2.
GL 80-022    Transmittal of NUREG-0654, "Criteria For  NA      Info Preparation and Evaluation of Radiological Emergency Response Plan GL 80-023    Change of Submittal Date For Evaluation    NA      Info Time Estimates GL 80-024    Transmittal of Information on NRC          NA      Info "Nuclear Data Link Specifications" GL 80-025    B 80-06, Engineering Safety Feature      NA      Info (ESF) Reset Controls GL 80-026    Qualifications of Reactor Operators        NA      Info GL 80-027    B 80-07, BWR Jet Pump Assembly            NA      Boiling Water Reactor Failure GL 80-028    B 80-08, Examination of Containment        C      GL 80-28 transmitted Bulletin 80-08. TVA responded to Liner Penetration Welds                          B 80-08 on July 8, 1980. See B 80-08 for further information.
GL 80-022    Transmittal of NUREG-0654, "Criteria For  NA      Info Preparation and Evaluation of Radiological Emergency Response Plan GL 80-023    Change of Submittal Date For Evaluation    NA      Info Time Estimates GL 80-024    Transmittal of Information on NRC          NA      Info "Nuclear Data Link Specifications" GL 80-025    B 80-06, Engineering Safety Feature      NA      Info (ESF) Reset Controls GL 80-026    Qualifications of Reactor Operators        NA      Info GL 80-027    B 80-07, BWR Jet Pump Assembly            NA      Boiling Water Reactor Failure GL 80-028    B 80-08, Examination of Containment        C      GL 80-28 transmitted Bulletin 80-08. TVA responded to Liner Penetration Welds                          B 80-08 on July 8, 1980. See B 80-08 for further information.
Line 2,288: Line 1,830:
Masonry Walls Employed by Plants                  TVA provided the information requested by letters dated Under CP and OL Review                      01    February 12, 1981, for reinforced walls and August 20, 1981, for nonreinforced walls. TVA provided a final response on January 22, 1982. See B 80-11 for further information.
Masonry Walls Employed by Plants                  TVA provided the information requested by letters dated Under CP and OL Review                      01    February 12, 1981, for reinforced walls and August 20, 1981, for nonreinforced walls. TVA provided a final response on January 22, 1982. See B 80-11 for further information.
GL 80-033    Actions Required From OL Applicants of    NA      Applies only to Babcock and Wilcox designed plants B&W Designed NSSS Resulting From NRC B&O Task Force Review of TMI2 Accident Page 43 of 95                                              * = See last page for status code definition.
GL 80-033    Actions Required From OL Applicants of    NA      Applies only to Babcock and Wilcox designed plants B&W Designed NSSS Resulting From NRC B&O Task Force Review of TMI2 Accident Page 43 of 95                                              * = See last page for status code definition.
* ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 80-034    Clarification of NRC Requirements for      NA    Info Emergency Response Facilities at Each Site GL 80-035    Effect of a DC Power Supply Failure on    NA    Boiling Water Reactor ECCS Performances GL 80-036    B 80-10, Contamination of                NA    Info Non-Radioactive System and Resulting Potential For Unmonitored, Uncontrolled Release to Environment GL 80-037    Five Additional TMI-2 Related              NA    Item was applicable only to units with operating license at the Requirements to Operating Reactors                time the item was issued.
 
ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 80-034    Clarification of NRC Requirements for      NA    Info Emergency Response Facilities at Each Site GL 80-035    Effect of a DC Power Supply Failure on    NA    Boiling Water Reactor ECCS Performances GL 80-036    B 80-10, Contamination of                NA    Info Non-Radioactive System and Resulting Potential For Unmonitored, Uncontrolled Release to Environment GL 80-037    Five Additional TMI-2 Related              NA    Item was applicable only to units with operating license at the Requirements to Operating Reactors                time the item was issued.
GL 80-038    Summary of Certain Non-Power Reactor      NA    Does not apply to power reactor.
GL 80-038    Summary of Certain Non-Power Reactor      NA    Does not apply to power reactor.
Physical Protection Requirements GL 80-039    B 80-11, Masonry Wall Design            NA    Info GL 80-040    Transmittal of NUREG-0654, "Report of      NA    Info the B&O Task Force and Appropriate NUREG-0626, "Generic Evaluation of FW Transient and Small Break LOCA GL 80-041    Summary of Meetings Held on                NA    Info April 22 &23, 1980 With Representatives of the Mark I Owners Group GL 80-042    B 80-12, Decay Heat Removal System        NA    Info Operability GL 80-043    B 80-13, Cracking In Core Spray          NA    Boiling Water Reactor Spargers GL 80-044    Reorganization of Functions and            NA    Info Assignments Within ONRR/SSPB GL 80-045    Fire Protection Rule                      NA    Item was applicable only to units with operating license at the time the item was issued.
Physical Protection Requirements GL 80-039    B 80-11, Masonry Wall Design            NA    Info GL 80-040    Transmittal of NUREG-0654, "Report of      NA    Info the B&O Task Force and Appropriate NUREG-0626, "Generic Evaluation of FW Transient and Small Break LOCA GL 80-041    Summary of Meetings Held on                NA    Info April 22 &23, 1980 With Representatives of the Mark I Owners Group GL 80-042    B 80-12, Decay Heat Removal System        NA    Info Operability GL 80-043    B 80-13, Cracking In Core Spray          NA    Boiling Water Reactor Spargers GL 80-044    Reorganization of Functions and            NA    Info Assignments Within ONRR/SSPB GL 80-045    Fire Protection Rule                      NA    Item was applicable only to units with operating license at the time the item was issued.
GL 80-046    Generic Technical Activity A-12, "Fracture  C    No response was required for this GL, and NUREG-0577 states and        Toughness and Additional Guidance on              that the lamellar tearing aspect of this issue was resolved by the GL 80-047    Potential for Low Fracture toughness and          NUREG. Further, the NUREG states that for plants under review, Laminar Tearing on PWR Steam                      the fracture toughness issue was resolved.
GL 80-046    Generic Technical Activity A-12, "Fracture  C    No response was required for this GL, and NUREG-0577 states and        Toughness and Additional Guidance on              that the lamellar tearing aspect of this issue was resolved by the GL 80-047    Potential for Low Fracture toughness and          NUREG. Further, the NUREG states that for plants under review, Laminar Tearing on PWR Steam                      the fracture toughness issue was resolved.
Generator Coolant Pump Supports" Page 44 of 95                                              * = See last page for status code definition.
Generator Coolant Pump Supports" Page 44 of 95                                              * = See last page for status code definition.
* ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 80-048    Revision to 5/19/80 Letter On Fire        NA    Item was applicable only to units with operating license at the Protection [GL 80-45]                            time the item was issued.
 
ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 80-048    Revision to 5/19/80 Letter On Fire        NA    Item was applicable only to units with operating license at the Protection [GL 80-45]                            time the item was issued.
GL 80-049    Nuclear Safeguards Problems              NA    Info GL 80-050    Generic Activity A-10, "BWR Cracks"      NA    Boiling Water Reactor GL 80-051    On-Site Storage of Low-Level Waste        NA    Item was applicable only to units with operating license at the time the item was issued.
GL 80-049    Nuclear Safeguards Problems              NA    Info GL 80-050    Generic Activity A-10, "BWR Cracks"      NA    Boiling Water Reactor GL 80-051    On-Site Storage of Low-Level Waste        NA    Item was applicable only to units with operating license at the time the item was issued.
GL 80-052    Five Additional TMI-2 Related            NA    Item was applicable only to units with operating license at the Requirements - Erata Sheets to 5/7/80            time the item was issued.
GL 80-052    Five Additional TMI-2 Related            NA    Item was applicable only to units with operating license at the Requirements - Erata Sheets to 5/7/80            time the item was issued.
Line 2,299: Line 1,843:
GL 80-053    Decay Heat Removal Capability            NA    Item was applicable only to units with operating license at the time the item was issued.
GL 80-053    Decay Heat Removal Capability            NA    Item was applicable only to units with operating license at the time the item was issued.
GL 80-054    B 80-14, Degradation of Scram            NA    Boiling Water Reactor Discharge Volume Capability GL 80-055    B 80-15, Possible Loss of Hotline With  NA    Info Loss of off-Site Power GL 80-056    Commission Memorandum and Order on        NA    Info Equipment Qualification GL 80-057    Further Commission Guidance For Power    NA    Info Reactor Operating Licenses NUREG-0660 and NUREG-0694 GL 80-058    B 80-16, Potential Misapplication of    NA    Info Rosemount Inc. Models 1151/1152 Pressure Transmitters With "A" Or "D" Output Codes GL 80-059    Transmittal of Federal Register Notice RE NA    Info Regional Meetings to Discuss Environmental Qualification of Electrical Equipment GL 80-060    Request for Information Regarding        NA    Info Evacuation Times Page 45 of 95                                            * = See last page for status code definition.
GL 80-054    B 80-14, Degradation of Scram            NA    Boiling Water Reactor Discharge Volume Capability GL 80-055    B 80-15, Possible Loss of Hotline With  NA    Info Loss of off-Site Power GL 80-056    Commission Memorandum and Order on        NA    Info Equipment Qualification GL 80-057    Further Commission Guidance For Power    NA    Info Reactor Operating Licenses NUREG-0660 and NUREG-0694 GL 80-058    B 80-16, Potential Misapplication of    NA    Info Rosemount Inc. Models 1151/1152 Pressure Transmitters With "A" Or "D" Output Codes GL 80-059    Transmittal of Federal Register Notice RE NA    Info Regional Meetings to Discuss Environmental Qualification of Electrical Equipment GL 80-060    Request for Information Regarding        NA    Info Evacuation Times Page 45 of 95                                            * = See last page for status code definition.
* ITEM                          TITLE                    REV                      ADDITIONAL INFORMATION GL 80-061    TMI-2 Lessons Learned                      NA    Info GL 80-062    TMI-2 Lessons Learned                      NA    Boiling Water Reactor GL 80-063    B 80-17, Failure of Control Rods to Insert NA    Boiling Water Reactor During a Scram at a BWR GL 80-064    Scram Discharge Volume Designs              NA    Boiling Water Reactor GL 80-065    Request for Estimated Construction          NA    Info Completion and Fuel Load Schedules GL 80-066    B 80-17, Supplement 1, Failure of          NA    Boiling Water Reactor Control Rods to Insert During a Scram at a BWR GL 80-067    Scram Discharge Volume                      NA    Boiling Water Reactor GL 80-068    B 80-17, Supplement 2, Failures            NA    Boiling Water Reactor Revealed by Testing Subsequent to Failure of Control Rods to Insert During a Scram at a BWR GL 80-069    B 80-18, Maintenance of Adequate          NA    Info Minimum Flow Through Centrifugal Charging Pumps Following Secondary Side HELB GL 80-070    B 80-19, Failures of Mercury-Wetted        NA    Info Matrix Relays in RPS of Operating Nuclear Power Plants Designed by GE GL 80-071    B 80-20, Failures of Westinghouse Type    NA    Info W-2 Spring Return to Neutral Control Switches GL 80-072    Interim Criteria For Shift Staffing        NA    Info GL 80-073    "Functional Criteria For Emergency          NA    Info Response Facilities, NUREG-0696 Page 46 of 95                                              * = See last page for status code definition.
 
* ITEM                          TITLE                  REV                          ADDITIONAL INFORMATION GL 80-074    Notice of Forthcoming Meeting With        NA      Info Representatives of EPRI to Discuss Program For Resolution of USI A-12, Fracture Toughness Issue GL 80-075    Lessons Learned Tech. Specs.              NA      Item was applicable only to units with operating license at the time the item was issued.
ITEM                          TITLE                    REV                      ADDITIONAL INFORMATION GL 80-061    TMI-2 Lessons Learned                      NA    Info GL 80-062    TMI-2 Lessons Learned                      NA    Boiling Water Reactor GL 80-063    B 80-17, Failure of Control Rods to Insert NA    Boiling Water Reactor During a Scram at a BWR GL 80-064    Scram Discharge Volume Designs              NA    Boiling Water Reactor GL 80-065    Request for Estimated Construction          NA    Info Completion and Fuel Load Schedules GL 80-066    B 80-17, Supplement 1, Failure of          NA    Boiling Water Reactor Control Rods to Insert During a Scram at a BWR GL 80-067    Scram Discharge Volume                      NA    Boiling Water Reactor GL 80-068    B 80-17, Supplement 2, Failures            NA    Boiling Water Reactor Revealed by Testing Subsequent to Failure of Control Rods to Insert During a Scram at a BWR GL 80-069    B 80-18, Maintenance of Adequate          NA    Info Minimum Flow Through Centrifugal Charging Pumps Following Secondary Side HELB GL 80-070    B 80-19, Failures of Mercury-Wetted        NA    Info Matrix Relays in RPS of Operating Nuclear Power Plants Designed by GE GL 80-071    B 80-20, Failures of Westinghouse Type    NA    Info W-2 Spring Return to Neutral Control Switches GL 80-072    Interim Criteria For Shift Staffing        NA    Info GL 80-073    "Functional Criteria For Emergency          NA    Info Response Facilities, NUREG-0696 Page 46 of 95                                              * = See last page for status code definition.
 
ITEM                          TITLE                  REV                          ADDITIONAL INFORMATION GL 80-074    Notice of Forthcoming Meeting With        NA      Info Representatives of EPRI to Discuss Program For Resolution of USI A-12, Fracture Toughness Issue GL 80-075    Lessons Learned Tech. Specs.              NA      Item was applicable only to units with operating license at the time the item was issued.
GL 80-076    Notice of Forthcoming Meeting With GE      NA      Info to Discussed Proposed BWR Feedwater Nozzle Leakage Detection System GL 80-077    Refueling Water Level - Technical          S      Unit 2 Action: Address in Technical Specifications, as Specifications Changes                            appropriate.
GL 80-076    Notice of Forthcoming Meeting With GE      NA      Info to Discussed Proposed BWR Feedwater Nozzle Leakage Detection System GL 80-077    Refueling Water Level - Technical          S      Unit 2 Action: Address in Technical Specifications, as Specifications Changes                            appropriate.
02
02 REVISION 02 UPDATE:
                                                                ----------------------------------------------------------------------------------------
                                                                ----------------------------------------------------------------------------------------
REVISION 02 UPDATE:
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
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.
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.
GL 80-078    Mark I Containment Long-Term Program      NA      Boiling Water Reactor GL 80-079    B 80-17, Supplement 3, Failures          NA      Boiling Water Reactor Revealed by Testing Subsequent to Failure of Control Rods to Insert During a Scram At a BWR GL 80-080    Preliminary Clarification of TMI Action    NA      Info Plan Requirements GL 80-081    Preliminary Clarification of TMI Action    NA      Info Plan Requirements - Addendum to 9/5/80 Letter [GL 80-80]
GL 80-078    Mark I Containment Long-Term Program      NA      Boiling Water Reactor GL 80-079    B 80-17, Supplement 3, Failures          NA      Boiling Water Reactor Revealed by Testing Subsequent to Failure of Control Rods to Insert During a Scram At a BWR GL 80-080    Preliminary Clarification of TMI Action    NA      Info Plan Requirements GL 80-081    Preliminary Clarification of TMI Action    NA      Info Plan Requirements - Addendum to 9/5/80 Letter [GL 80-80]
GL 80-082    B 79-01b, Supplement 2, Environmental    NA      Info Qualification of Class 1E Equipment GL 80-083    Environmental Qualification of            NA      Info Safety-Related Equipment GL 80-084    BWR Scram System                          NA      Boiling Water Reactor Page 47 of 95                                              * = See last page for status code definition.
GL 80-082    B 79-01b, Supplement 2, Environmental    NA      Info Qualification of Class 1E Equipment GL 80-083    Environmental Qualification of            NA      Info Safety-Related Equipment GL 80-084    BWR Scram System                          NA      Boiling Water Reactor Page 47 of 95                                              * = See last page for status code definition.
* ITEM                          TITLE                REV                      ADDITIONAL INFORMATION GL 80-085    Implementation of Guidance From          NA      Info USI A-12, Potential For LOW Fracture Toughness and Lamellar Tearing On Component Support GL 80-086    Notice of Meeting to Discuss Final        NA      Info Resolution of USI A-12 GL 80-087    Notice of Meeting to Discuss Status of    NA      Info EPRI-Proposed Resolution of the USI A-12 Fracture Toughness Issue GL 80-088    Seismic Qualification of Auxiliary        NA      Item was applicable only to units with operating license at the Feedwater Systems                                time the item was issued.
 
ITEM                          TITLE                REV                      ADDITIONAL INFORMATION GL 80-085    Implementation of Guidance From          NA      Info USI A-12, Potential For LOW Fracture Toughness and Lamellar Tearing On Component Support GL 80-086    Notice of Meeting to Discuss Final        NA      Info Resolution of USI A-12 GL 80-087    Notice of Meeting to Discuss Status of    NA      Info EPRI-Proposed Resolution of the USI A-12 Fracture Toughness Issue GL 80-088    Seismic Qualification of Auxiliary        NA      Item was applicable only to units with operating license at the Feedwater Systems                                time the item was issued.
GL 80-089    B 79-01b, Supplement 3, Environmental    NA      Info Qualification of Class 1E Equipment GL 80-090    NUREG-0737, TMI (Prior and future GLs,    CI    See NUREG items in this list.
GL 80-089    B 79-01b, Supplement 3, Environmental    NA      Info Qualification of Class 1E Equipment GL 80-090    NUREG-0737, TMI (Prior and future GLs,    CI    See NUREG items in this list.
with the exception of certain discrete scopes, have been screened into NUREG list for those applicable to Watts Bar 2)
with the exception of certain discrete scopes, have been screened into NUREG list for those applicable to Watts Bar 2)
Line 2,317: Line 1,861:
GL 80-094    Emergency Plan                            NA      Info GL 80-095    Generic Technical Activity A-10,          NA      Boiling Water Reactor NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking GL 80-096    Fire Protection                          NA      Addressed to specific plant(s).
GL 80-094    Emergency Plan                            NA      Info GL 80-095    Generic Technical Activity A-10,          NA      Boiling Water Reactor NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking GL 80-096    Fire Protection                          NA      Addressed to specific plant(s).
GL 80-097    B 80-23, Failures of Solenoid Valves    NA      Info Manufactured by Valcor Engineering Corporation Page 48 of 95                                              * = See last page for status code definition.
GL 80-097    B 80-23, Failures of Solenoid Valves    NA      Info Manufactured by Valcor Engineering Corporation Page 48 of 95                                              * = See last page for status code definition.
* ITEM                        TITLE                  REV                      ADDITIONAL INFORMATION GL 80-098    B 80-24, Prevention of Damage Due to    NA    Info Water Leakage Inside Containment GL 80-099    Technical Specifications Revisions For  NA    Info Snubber Surveillance GL 80-100    Appendix R to 10 CFR 50 Regarding Fire  NA    Item was applicable only to units with operating license at the Protection - Federal Register Notice            time the item was issued.
 
ITEM                        TITLE                  REV                      ADDITIONAL INFORMATION GL 80-098    B 80-24, Prevention of Damage Due to    NA    Info Water Leakage Inside Containment GL 80-099    Technical Specifications Revisions For  NA    Info Snubber Surveillance GL 80-100    Appendix R to 10 CFR 50 Regarding Fire  NA    Item was applicable only to units with operating license at the Protection - Federal Register Notice            time the item was issued.
GL 80-101    Inservice Inspection Programs            NA    Addressed to specific plant(s).
GL 80-101    Inservice Inspection Programs            NA    Addressed to specific plant(s).
GL 80-102    Commission Memorandum and Order of      NA    Info May 23, 1980 (Referencing B 79-01b, Supplement 2 - q.2 & 3 - Sept 30, 1980)
GL 80-102    Commission Memorandum and Order of      NA    Info May 23, 1980 (Referencing B 79-01b, Supplement 2 - q.2 & 3 - Sept 30, 1980)
GL 80-103    Fire Protection - Revised Federal        NA    Info Register Notice GL 80-104    Orders On Environmental Qualification of NA    Info Safety Related Electrical Equipment GL 80-105    Implementation of Guidance For USI      NA    Info A-12, Potential For Low Fracture toughness and Lamellar Tearing On Component Supports GL 80-106    Report On ECCS Cladding Models,          NA    Info NUREG-0630 GL 80-107    BWR Scram Discharge System              NA    Boiling Water Reactor GL 80-108    Emergency Planning                      NA    Info GL 80-109    Guidelines For SEP Soil Structure        NA    Info Interaction Reviews GL 80-110    Periodic Updating of FSARS              NA    Item was applicable only to units with operating license at the time the item was issued.
GL 80-103    Fire Protection - Revised Federal        NA    Info Register Notice GL 80-104    Orders On Environmental Qualification of NA    Info Safety Related Electrical Equipment GL 80-105    Implementation of Guidance For USI      NA    Info A-12, Potential For Low Fracture toughness and Lamellar Tearing On Component Supports GL 80-106    Report On ECCS Cladding Models,          NA    Info NUREG-0630 GL 80-107    BWR Scram Discharge System              NA    Boiling Water Reactor GL 80-108    Emergency Planning                      NA    Info GL 80-109    Guidelines For SEP Soil Structure        NA    Info Interaction Reviews GL 80-110    Periodic Updating of FSARS              NA    Item was applicable only to units with operating license at the time the item was issued.
Page 49 of 95                                            * = See last page for status code definition.
Page 49 of 95                                            * = See last page for status code definition.
* ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 80-111    B 80-17, Supplement 4, Failure of        NA    Boiling Water Reactor Control Rods to Insert During a Scram at a BWR GL 80-112    B 80-25, Operating Problems With          NA    Info Target Rock Safety Relief Valves GL 80-113    Control of Heavy Loads                      C    Superseded by GL 81-007.
 
ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 80-111    B 80-17, Supplement 4, Failure of        NA    Boiling Water Reactor Control Rods to Insert During a Scram at a BWR GL 80-112    B 80-25, Operating Problems With          NA    Info Target Rock Safety Relief Valves GL 80-113    Control of Heavy Loads                      C    Superseded by GL 81-007.
GL 81-001    Qualification of Inspection, Examination,  NA    Info Testing and Audit Personnel GL 81-002    Analysis, Conclusions and                  NA    Info Recommendations Concerning Operator Licensing GL 81-003    Implementation of NUREG-0313,              NA    Boiling Water Reactor Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping GL 81-004    Emergency Procedures and Training for      C    Superseded by Station Blackout Rule.
GL 81-001    Qualification of Inspection, Examination,  NA    Info Testing and Audit Personnel GL 81-002    Analysis, Conclusions and                  NA    Info Recommendations Concerning Operator Licensing GL 81-003    Implementation of NUREG-0313,              NA    Boiling Water Reactor Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping GL 81-004    Emergency Procedures and Training for      C    Superseded by Station Blackout Rule.
Station Blackout Events GL 81-005    Information Regarding The Program For      NA    Info Environmental Qualification of Safety-Related Electrical Equipment GL 81-006    Periodic Updating of Final Safety Analysis NA    Info Reports (FSARS)
Station Blackout Events GL 81-005    Information Regarding The Program For      NA    Info Environmental Qualification of Safety-Related Electrical Equipment GL 81-006    Periodic Updating of Final Safety Analysis NA    Info Reports (FSARS)
Line 2,330: Line 1,876:
Unit 2 Action: Unit 2 Heavy Loads Program will be in compliance with NUREG-0612.
Unit 2 Action: Unit 2 Heavy Loads Program will be in compliance with NUREG-0612.
GL 81-008    ODYN Code                                  NA    Boiling Water Reactor Page 50 of 95                                              * = See last page for status code definition.
GL 81-008    ODYN Code                                  NA    Boiling Water Reactor Page 50 of 95                                              * = See last page for status code definition.
* ITEM                          TITLE                  REV                        ADDITIONAL INFORMATION GL 81-009    BWR Scram Discharge System                  NA    Boiling Water Reactor GL 81-010    Post-TMI Requirements For The              NA    Info Emergency Operations Facility GL 81-011    BWR Feedwater Nozzle and Control Rod        NA    Boiling Water Reactor Drive Return Line Nozzle Cracking (NUREG-0619)
 
ITEM                          TITLE                  REV                        ADDITIONAL INFORMATION GL 81-009    BWR Scram Discharge System                  NA    Boiling Water Reactor GL 81-010    Post-TMI Requirements For The              NA    Info Emergency Operations Facility GL 81-011    BWR Feedwater Nozzle and Control Rod        NA    Boiling Water Reactor Drive Return Line Nozzle Cracking (NUREG-0619)
GL 81-012    Fire Protection Rule                        NA    Item was applicable only to units with operating license at the time the item was issued.
GL 81-012    Fire Protection Rule                        NA    Item was applicable only to units with operating license at the time the item was issued.
GL 81-013    SER For GEXL Correlation For 8X8R          NA    Boiling Water Reactor Fuel Reload Applications For Appendix D Submittals of The GE topical Report GL 81-014    Seismic Qualification of Auxiliary          CI    TVA: FSAR 10.4.9 Feedwater Systems
GL 81-013    SER For GEXL Correlation For 8X8R          NA    Boiling Water Reactor Fuel Reload Applications For Appendix D Submittals of The GE topical Report GL 81-014    Seismic Qualification of Auxiliary          CI    TVA: FSAR 10.4.9 Feedwater Systems Unit 2 Action: Additional Unit 2 implementing procedures or other activity is required for completion.
                                                                --------------------
Unit 2 Action: Additional Unit 2 implementing procedures or other activity is required for completion.
[WAS "OL."]
[WAS "OL."]
GL 81-015    Environmental Qualification of Class 1E    NA    Info Electrical Equipment - Clarification of Staffs Handling of Proprietary Information GL 81-016    NUREG-0737, Item I.C.1 SER on              NA    Applies only to Babcock and Wilcox designed plants Abnormal Transient Operating Guidelines (ATOG)
GL 81-015    Environmental Qualification of Class 1E    NA    Info Electrical Equipment - Clarification of Staffs Handling of Proprietary Information GL 81-016    NUREG-0737, Item I.C.1 SER on              NA    Applies only to Babcock and Wilcox designed plants Abnormal Transient Operating Guidelines (ATOG)
GL 81-017    Functional Criteria for Emergency          NA    Info Response Facilities GL 81-018    BWR Scram Discharge System -                NA    Boiling Water Reactor Clarification of Diverse Instrumentation Requirements GL 81-019    Thermal Shock to Reactor Pressure          NA    Item was applicable only to units with operating license at the Vessels                                            time the item was issued.
GL 81-017    Functional Criteria for Emergency          NA    Info Response Facilities GL 81-018    BWR Scram Discharge System -                NA    Boiling Water Reactor Clarification of Diverse Instrumentation Requirements GL 81-019    Thermal Shock to Reactor Pressure          NA    Item was applicable only to units with operating license at the Vessels                                            time the item was issued.
GL 81-020    Safety Concerns Associated With Pipe        NA    Boiling Water Reactor Breaks in the BWR Scram System Page 51 of 95                                              * = See last page for status code definition.
GL 81-020    Safety Concerns Associated With Pipe        NA    Boiling Water Reactor Breaks in the BWR Scram System Page 51 of 95                                              * = See last page for status code definition.
* ITEM                        TITLE                    REV                      ADDITIONAL INFORMATION GL 81-021    Natural Circulation Cooldown                CI    TVA responded December 3, 1981.
 
ITEM                        TITLE                    REV                      ADDITIONAL INFORMATION GL 81-021    Natural Circulation Cooldown                CI    TVA responded December 3, 1981.
Unit 2 Action: Issue operating procedures.
Unit 2 Action: Issue operating procedures.
GL 81-022    Engineering Evaluation of the              NA    Info H. B. Robinson Reactor Coolant System Leak on 1/29/81 GL 81-023    INPO Plant Specific Evaluation Reports    NA    Info GL 81-024    Multi-Plant Issue B-56, Control Rods Fail NA    Boiling Water Reactor to Fully Insert GL 81-025    Change in Implementing Schedule For        NA    Info Submission and Evaluation of Upgraded Emergency Plans GL 81-026    Licensing Requirements for Pending        NA    Applicants with pending Construction Permits Construction Permit and Manufacturing License Applications GL 81-027    Privacy and Proprietary Material in        NA    Info Emergency Plans GL 81-028    Steam Generator Overfill                  NA    Info GL 81-029    Simulator Examinations                    NA    Info GL 81-030    Safety Concerns Associated With Pipe      NA    Boiling Water Reactor Breaks in the BWR Scram System GL 81-031    This GL was never issued.                  NA GL 81-032    NUREG-0737, Item II.K.3.44, Evaluation    NA    Boiling Water Reactor of Anticipated Transients Combined With Single Failure GL 81-033    This GL was never issued.                  NA Page 52 of 95                                              * = See last page for status code definition.
GL 81-022    Engineering Evaluation of the              NA    Info H. B. Robinson Reactor Coolant System Leak on 1/29/81 GL 81-023    INPO Plant Specific Evaluation Reports    NA    Info GL 81-024    Multi-Plant Issue B-56, Control Rods Fail NA    Boiling Water Reactor to Fully Insert GL 81-025    Change in Implementing Schedule For        NA    Info Submission and Evaluation of Upgraded Emergency Plans GL 81-026    Licensing Requirements for Pending        NA    Applicants with pending Construction Permits Construction Permit and Manufacturing License Applications GL 81-027    Privacy and Proprietary Material in        NA    Info Emergency Plans GL 81-028    Steam Generator Overfill                  NA    Info GL 81-029    Simulator Examinations                    NA    Info GL 81-030    Safety Concerns Associated With Pipe      NA    Boiling Water Reactor Breaks in the BWR Scram System GL 81-031    This GL was never issued.                  NA GL 81-032    NUREG-0737, Item II.K.3.44, Evaluation    NA    Boiling Water Reactor of Anticipated Transients Combined With Single Failure GL 81-033    This GL was never issued.                  NA Page 52 of 95                                              * = See last page for status code definition.
* ITEM                          TITLE                    REV                      ADDITIONAL INFORMATION GL 81-034    Safety Concerns Associated With Pipe        NA    Boiling Water Reactor Breaks in the BWR Scram System GL 81-035    Safety Concerns Associated With Pipe        NA    Boiling Water Reactor Breaks in the BWR Scram System GL 81-036    Revised Schedule for Completion of TMI      NA    Info Action Plan Item II.D.1, Relief and Safety Valve Testing GL 81-037    ODYN Code Reanalysis Requirements          NA    Boiling Water Reactor GL 81-038    Storage of Low Level Radioactive Wastes    NA    Info at Power Reactor Sites GL 81-039    NRC Volume Reduction Policy                NA    Info GL 81-040    Qualifications of Reactor Operators        NA    Info GL 82-001    New Applications Survey                    NA    Info GL 82-002    Commission Policy on Overtime              NA    Info GL 82-003    High Burnup MAPLHGR Limits                  NA    Boiling Water Reactor GL 82-004    Use of INPO See-in Program                  NA    Info GL 82-005    Post-TMI Requirements                      NA    Item was applicable only to units with operating license at the time the item was issued.
 
ITEM                          TITLE                    REV                      ADDITIONAL INFORMATION GL 81-034    Safety Concerns Associated With Pipe        NA    Boiling Water Reactor Breaks in the BWR Scram System GL 81-035    Safety Concerns Associated With Pipe        NA    Boiling Water Reactor Breaks in the BWR Scram System GL 81-036    Revised Schedule for Completion of TMI      NA    Info Action Plan Item II.D.1, Relief and Safety Valve Testing GL 81-037    ODYN Code Reanalysis Requirements          NA    Boiling Water Reactor GL 81-038    Storage of Low Level Radioactive Wastes    NA    Info at Power Reactor Sites GL 81-039    NRC Volume Reduction Policy                NA    Info GL 81-040    Qualifications of Reactor Operators        NA    Info GL 82-001    New Applications Survey                    NA    Info GL 82-002    Commission Policy on Overtime              NA    Info GL 82-003    High Burnup MAPLHGR Limits                  NA    Boiling Water Reactor GL 82-004    Use of INPO See-in Program                  NA    Info GL 82-005    Post-TMI Requirements                      NA    Item was applicable only to units with operating license at the time the item was issued.
GL 82-006    This GL was never issued.                  NA Page 53 of 95                                              * = See last page for status code definition.
GL 82-006    This GL was never issued.                  NA Page 53 of 95                                              * = See last page for status code definition.
* ITEM                          TITLE                REV                      ADDITIONAL INFORMATION GL 82-007    Transmittal of NUREG-0909 Relative to    NA    Boiling Water Reactor the Ginna Tube Rupture GL 82-008    Transmittal of NUREG-0909 Relative to    NA    Info the Ginna Tube Rupture GL 82-009    Environmental Qualification of Safety    NA    Info Related Electrical Equipment GL 82-010    Post-TMI Requirements                    NA    Item was applicable only to units with operating license at the time the item was issued.
 
ITEM                          TITLE                REV                      ADDITIONAL INFORMATION GL 82-007    Transmittal of NUREG-0909 Relative to    NA    Boiling Water Reactor the Ginna Tube Rupture GL 82-008    Transmittal of NUREG-0909 Relative to    NA    Info the Ginna Tube Rupture GL 82-009    Environmental Qualification of Safety    NA    Info Related Electrical Equipment GL 82-010    Post-TMI Requirements                    NA    Item was applicable only to units with operating license at the time the item was issued.
GL 82-011    Transmittal of NUREG-0916 Relative to    NA    Info the Restart of R. E. Ginna Nuclear Power Plant GL 82-012    Nuclear Power Plant Staff Working Hours  NA    Info GL 82-013    Reactor Operator and Senior Reactor      NA    Info Operator Examinations GL 82-014    Submittal of Documents to the NRC        NA    Info GL 82-015    This GL was never issued.                NA GL 82-016    NUREG-0737 Technical Specifications      NA    Item was applicable only to units with operating license at the time the item was issued.
GL 82-011    Transmittal of NUREG-0916 Relative to    NA    Info the Restart of R. E. Ginna Nuclear Power Plant GL 82-012    Nuclear Power Plant Staff Working Hours  NA    Info GL 82-013    Reactor Operator and Senior Reactor      NA    Info Operator Examinations GL 82-014    Submittal of Documents to the NRC        NA    Info GL 82-015    This GL was never issued.                NA GL 82-016    NUREG-0737 Technical Specifications      NA    Item was applicable only to units with operating license at the time the item was issued.
GL 82-017    Inconsistency of Requirements Between    NA    Info 50.54(T) and 50.15 GL 82-018    Reactor Operator and Senior Reactor      NA    Info Operator Requalification Examinations GL 82-019    Submittal of Copies of Documentation to  NA    Info NRC - Copy Requirements for Emergency Plans and Physical Security Plans Page 54 of 95                                            * = See last page for status code definition.
GL 82-017    Inconsistency of Requirements Between    NA    Info 50.54(T) and 50.15 GL 82-018    Reactor Operator and Senior Reactor      NA    Info Operator Requalification Examinations GL 82-019    Submittal of Copies of Documentation to  NA    Info NRC - Copy Requirements for Emergency Plans and Physical Security Plans Page 54 of 95                                            * = See last page for status code definition.
* ITEM                          TITLE                REV                      ADDITIONAL INFORMATION GL 82-020    Guidance for Implementing the Standard  NA    Info Review Plan Rule GL 82-021    Fire Protection Audits                  NA    Info GL 82-022    Congressional Request for Information  NA    Item was applicable only to units with operating license at the Concerning Steam Generator Tube                time the item was issued.
 
ITEM                          TITLE                REV                      ADDITIONAL INFORMATION GL 82-020    Guidance for Implementing the Standard  NA    Info Review Plan Rule GL 82-021    Fire Protection Audits                  NA    Info GL 82-022    Congressional Request for Information  NA    Item was applicable only to units with operating license at the Concerning Steam Generator Tube                time the item was issued.
Integrity GL 82-023    Inconsistency Between Requirements of  NA    Info 10CFR 73.40(d) and Standard Technical Specifications For Performing Audits of Safeguards Contingency Plans GL 82-024    Safety Relief Valve Quencher Loads:    NA    Boiling Water Reactor BWR MARK II and III Containments GL 82-025    Integrated IAEA Exercise for Physical  NA    Item was applicable only to units with operating license at the Inventory at LWRS                              time the item was issued.
Integrity GL 82-023    Inconsistency Between Requirements of  NA    Info 10CFR 73.40(d) and Standard Technical Specifications For Performing Audits of Safeguards Contingency Plans GL 82-024    Safety Relief Valve Quencher Loads:    NA    Boiling Water Reactor BWR MARK II and III Containments GL 82-025    Integrated IAEA Exercise for Physical  NA    Item was applicable only to units with operating license at the Inventory at LWRS                              time the item was issued.
GL 82-026    NUREG-0744, REV. 1, Pressure Vessel    NA    Item was applicable only to units with operating license at the Material Fracture Toughness                  time the item was issued.
GL 82-026    NUREG-0744, REV. 1, Pressure Vessel    NA    Item was applicable only to units with operating license at the Material Fracture Toughness                  time the item was issued.
Line 2,354: Line 1,903:
Unit 2 Action: Install Westinghouse Common Q PAM system.
Unit 2 Action: Install Westinghouse Common Q PAM system.
GL 82-029    This GL was never issued.              NA GL 82-030    Filings Related to 10 CFR 50 Production NA    Info and Utilization Facilities Page 55 of 95                                          * = See last page for status code definition.
GL 82-029    This GL was never issued.              NA GL 82-030    Filings Related to 10 CFR 50 Production NA    Info and Utilization Facilities Page 55 of 95                                          * = See last page for status code definition.
* ITEM                        TITLE                    REV                      ADDITIONAL INFORMATION GL 82-031    This GL was never issued.                NA GL 82-032    Draft Steam Generator Report (SAI)        NA    Item was applicable only to units with operating license at the time the item was issued.
 
ITEM                        TITLE                    REV                      ADDITIONAL INFORMATION GL 82-031    This GL was never issued.                NA GL 82-032    Draft Steam Generator Report (SAI)        NA    Item was applicable only to units with operating license at the time the item was issued.
GL 82-033    Supplement to NUREG-0737,                  CI    Safety Parameter Display System (SPDS) / Requirements for Requirements for Emergency Response            Emergency Response Capability - NRC reviewed in SSER5, Capability                                      SSER6, and 18.2.2 of SSER15.
GL 82-033    Supplement to NUREG-0737,                  CI    Safety Parameter Display System (SPDS) / Requirements for Requirements for Emergency Response            Emergency Response Capability - NRC reviewed in SSER5, Capability                                      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.
Unit 2 Action: Install SPDS and have it operational prior to start-up after the first refueling outage.
GL 82-034    This GL was never issued.                NA GL 82-035    This GL was never issued.                NA GL 82-036    This GL was never issued.                NA GL 82-037    This GL was never issued.                NA GL 82-038    Meeting to Discuss Developments for      NA    Info Operator Licensing Examinations GL 82-039    Problems With Submittals of Subsequent    NA    Info Information of CURT 73.21 For Licensing Reviews GL 83-001    Operator Licensing Examination Site Visit NA    Info GL 83-002    NUREG-0737 Technical Specifications      NA    Boiling Water Reactor GL 83-003    This GL was never issued.                NA GL 83-004    Regional Workshops Regarding              NA    Info Supplement 1 to NUREG-0737, Requirements For Emergency Response Capability Page 56 of 95                                            * = See last page for status code definition.
GL 82-034    This GL was never issued.                NA GL 82-035    This GL was never issued.                NA GL 82-036    This GL was never issued.                NA GL 82-037    This GL was never issued.                NA GL 82-038    Meeting to Discuss Developments for      NA    Info Operator Licensing Examinations GL 82-039    Problems With Submittals of Subsequent    NA    Info Information of CURT 73.21 For Licensing Reviews GL 83-001    Operator Licensing Examination Site Visit NA    Info GL 83-002    NUREG-0737 Technical Specifications      NA    Boiling Water Reactor GL 83-003    This GL was never issued.                NA GL 83-004    Regional Workshops Regarding              NA    Info Supplement 1 to NUREG-0737, Requirements For Emergency Response Capability Page 56 of 95                                            * = See last page for status code definition.
* ITEM                        TITLE                  REV                        ADDITIONAL INFORMATION GL 83-005    Safety Evaluation of "Emergency          NA    Boiling Water Reactor Procedure Guidelines, Revision 2,"
 
ITEM                        TITLE                  REV                        ADDITIONAL INFORMATION GL 83-005    Safety Evaluation of "Emergency          NA    Boiling Water Reactor Procedure Guidelines, Revision 2,"
June 1982 GL 83-006    Certificates and Revised Format For      NA    Info Reactor Operator and Senior Reactor Operator Licenses GL 83-007    The Nuclear Waste Policy Act of 1982    NA    Info GL 83-008    Modification of Vacuum Breakers on Mark  NA    Boiling Water Reactor I Containments GL 83-009    Review of Combustion Engineering        NA    Applies only to Combustion Engineering designed plants Owners' Group Emergency Procedures Guideline Program GL 83-010a    Resolution of TMI Action Item II.K.3.5., NA    Applies only to Combustion Engineering designed plants "Automatic Trip of Reactor Coolant Pumps" GL 83-010b    Resolution of TMI Action Item II.K.3.5., NA    Applies only to Combustion Engineering designed plants "Automatic Trip of Reactor Coolant Pumps" GL 83-010c    Resolution of TMI Action Item II.K.3.5.,  CI    TVA: letters dated January 5, 1984 and June 25, 1984 "Automatic Trip of Reactor Coolant Pumps"                                          NRC: letter dated June 8, 1990.
June 1982 GL 83-006    Certificates and Revised Format For      NA    Info Reactor Operator and Senior Reactor Operator Licenses GL 83-007    The Nuclear Waste Policy Act of 1982    NA    Info GL 83-008    Modification of Vacuum Breakers on Mark  NA    Boiling Water Reactor I Containments GL 83-009    Review of Combustion Engineering        NA    Applies only to Combustion Engineering designed plants Owners' Group Emergency Procedures Guideline Program GL 83-010a    Resolution of TMI Action Item II.K.3.5., NA    Applies only to Combustion Engineering designed plants "Automatic Trip of Reactor Coolant Pumps" GL 83-010b    Resolution of TMI Action Item II.K.3.5., NA    Applies only to Combustion Engineering designed plants "Automatic Trip of Reactor Coolant Pumps" GL 83-010c    Resolution of TMI Action Item II.K.3.5.,  CI    TVA: letters dated January 5, 1984 and June 25, 1984 "Automatic Trip of Reactor Coolant Pumps"                                          NRC: letter dated June 8, 1990.
                                                              --------------------
Unit 2 Action: Incorporate emergency response guidelines into applicable procedures.
Unit 2 Action: Incorporate emergency response guidelines into applicable procedures.
[WAS "NOTE 3."]
[WAS "NOTE 3."]
Line 2,366: Line 1,916:
Pumps" GL 83-010e    Resolution of TMI Action Item II.K.3.5., NA    Applies only to Babcock and Wilcox designed plants "Automatic Trip of Reactor Coolant Pumps" GL 83-010f    Resolution of TMI Action Item II.K.3.5., NA    Applies only to Babcock and Wilcox designed plants "Automatic Trip of Reactor Coolant Pumps" GL 83-011    Licensee Qualification for Performing    NA    Item was applicable only to units with operating license at the Safety Analyses in Support of Licensing        time the item was issued.
Pumps" GL 83-010e    Resolution of TMI Action Item II.K.3.5., NA    Applies only to Babcock and Wilcox designed plants "Automatic Trip of Reactor Coolant Pumps" GL 83-010f    Resolution of TMI Action Item II.K.3.5., NA    Applies only to Babcock and Wilcox designed plants "Automatic Trip of Reactor Coolant Pumps" GL 83-011    Licensee Qualification for Performing    NA    Item was applicable only to units with operating license at the Safety Analyses in Support of Licensing        time the item was issued.
Actions Page 57 of 95                                            * = See last page for status code definition.
Actions Page 57 of 95                                            * = See last page for status code definition.
* ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 83-012    Issuance of NRC FORM 398 - Personal        NA    Info Qualifications Statement - Licensee GL 83-013    Clarification of Surveillance Requirements NA    Info for HEPA Filters and Charcoal Absorber Units In Standard Technical Specifications on ESF Cleanup Systems GL 83-014    Definition of "Key Maintenance            NA    Info Personnel," (Clarification of Generic Letter 82-12)
 
ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 83-012    Issuance of NRC FORM 398 - Personal        NA    Info Qualifications Statement - Licensee GL 83-013    Clarification of Surveillance Requirements NA    Info for HEPA Filters and Charcoal Absorber Units In Standard Technical Specifications on ESF Cleanup Systems GL 83-014    Definition of "Key Maintenance            NA    Info Personnel," (Clarification of Generic Letter 82-12)
GL 83-015    Implementation of Regulatory Guide        NA    Info 1.150, "Ultrasonic Testing of Reactor Vessel Welds During Preservice &
GL 83-015    Implementation of Regulatory Guide        NA    Info 1.150, "Ultrasonic Testing of Reactor Vessel Welds During Preservice &
Inservice Examinations, Revision 1 GL 83-016    Transmittal of NUREG-0977 Relative to      NA    Info the ATWS Events at Salem Generating Station, Unit No.1 GL 83-016a    Transmittal of NUREG-0977 Relative to      NA    Info the ATWS Events at Salem Generating Station, Unit No.1 GL 83-017    Integrity of Requalification Examinations  NA    Info for Renewal of Reactor Operator and Senior Reactor Operator Licenses GL 83-018    NRC Staff Review of the BWR Owners'        NA    Boiling Water Reactor Group (BWROG) Control Room Survey Program GL 83-019    New Procedures for Providing Public        NA    Item was applicable only to units with operating license at the Notice Concerning Issuance of                    time the item was issued.
Inservice Examinations, Revision 1 GL 83-016    Transmittal of NUREG-0977 Relative to      NA    Info the ATWS Events at Salem Generating Station, Unit No.1 GL 83-016a    Transmittal of NUREG-0977 Relative to      NA    Info the ATWS Events at Salem Generating Station, Unit No.1 GL 83-017    Integrity of Requalification Examinations  NA    Info for Renewal of Reactor Operator and Senior Reactor Operator Licenses GL 83-018    NRC Staff Review of the BWR Owners'        NA    Boiling Water Reactor Group (BWROG) Control Room Survey Program GL 83-019    New Procedures for Providing Public        NA    Item was applicable only to units with operating license at the Notice Concerning Issuance of                    time the item was issued.
Amendments to Operating Licenses GL 83-020    Integrated Scheduling for Implementation  NA    Info of Plant Modifications GL 83-021    Clarification of Access Control            NA    Info Procedures for Law Enforcement Visits GL 83-022    Safety Evaluation of "Emergency            NA    Info Response Guidelines" GL 83-023    Safety Evaluation of "Emergency            NA    Applies only to Combustion Engineering designed plants Procedure Guidelines" Page 58 of 95                                              * = See last page for status code definition.
Amendments to Operating Licenses GL 83-020    Integrated Scheduling for Implementation  NA    Info of Plant Modifications GL 83-021    Clarification of Access Control            NA    Info Procedures for Law Enforcement Visits GL 83-022    Safety Evaluation of "Emergency            NA    Info Response Guidelines" GL 83-023    Safety Evaluation of "Emergency            NA    Applies only to Combustion Engineering designed plants Procedure Guidelines" Page 58 of 95                                              * = See last page for status code definition.
* ITEM                          TITLE              REV                        ADDITIONAL INFORMATION GL 83-024    TMI Task Action Plan Item I.G.1,      NA    Boiling Water Reactor "Special Low Power Testing and Training," Recommendations for BWRs GL 83-025    This GL was never issued.              NA GL 83-026    Clarification Of Surveillance          NA    Info Requirements For Diesel Fuel Impurity Level Tests GL 83-027    Surveillance Intervals in Standard    NA    Info Technical Specifications GL 83-028    "Required Actions Based on Generic      C    TVA: letters dated November 7, 1983 and Implications of Salem ATWS Events:            December 4, 1987 1.2 - Post Trip Review Data and              NRC: IR 50-390, 391/86-04 Information Capability GL 83-028    "Required Actions Based on Generic      CI    TVA: letters dated November 7, 1983 and August 24, 1990 Implications of Salem ATWS Events:
 
ITEM                          TITLE              REV                        ADDITIONAL INFORMATION GL 83-024    TMI Task Action Plan Item I.G.1,      NA    Boiling Water Reactor "Special Low Power Testing and Training," Recommendations for BWRs GL 83-025    This GL was never issued.              NA GL 83-026    Clarification Of Surveillance          NA    Info Requirements For Diesel Fuel Impurity Level Tests GL 83-027    Surveillance Intervals in Standard    NA    Info Technical Specifications GL 83-028    "Required Actions Based on Generic      C    TVA: letters dated November 7, 1983 and Implications of Salem ATWS Events:            December 4, 1987 1.2 - Post Trip Review Data and              NRC: IR 50-390, 391/86-04 Information Capability GL 83-028    "Required Actions Based on Generic      CI    TVA: letters dated November 7, 1983 and August 24, 1990 Implications of Salem ATWS Events:
NRC: letters dated October 20, 1986 and June 18, 1990 2.1 - Equipment Classification and Vendor Interface (Reactor Trip        --------------------
NRC: letters dated October 20, 1986 and June 18, 1990 2.1 - Equipment Classification and Vendor Interface (Reactor Trip        --------------------
System Components)
System Components)
Line 2,378: Line 1,930:
2.2 - Equipment Classification and Vendor Interface (All SR Components)"
2.2 - Equipment Classification and Vendor Interface (All SR Components)"
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* ITEM                        TITLE                REV                          ADDITIONAL INFORMATION GL 83-028    "Required Actions Based on Generic      S      TVA: letters dated November 7, 1983, January 17, 1986 and Implications of Salem ATWS Events:              November 1, 1993 02 3.1 - Post-Maintenance Testing                  NRC: letters dated December 10, 1985, October 27, 1986, and (Reactor Trip System Components)        July 2, 1990; IR 390, 391/86-04
 
                                                              --------------------
ITEM                        TITLE                REV                          ADDITIONAL INFORMATION GL 83-028    "Required Actions Based on Generic      S      TVA: letters dated November 7, 1983, January 17, 1986 and Implications of Salem ATWS Events:              November 1, 1993 02 3.1 - Post-Maintenance Testing                  NRC: letters dated December 10, 1985, October 27, 1986, and (Reactor Trip System Components)        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.
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.
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                                                              ----------------------------------------------------------------------------------------
REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision A of the Unit 2 TS (including the TS Bases) was submitted on March 4, 2009.
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.
The Bases for TS Surveillance Requirement 3.0.1 states, in part, Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2.
GL 83-028    "Required Actions Based on Generic      S      TVA: letters dated November 7, 1983, January 17, 1986 and Implications of Salem ATWS Events:              November 1, 1993 02 3.2 - Post-Maintenance Testing (All SR          NRC: letters dated December 10, 1985, October 27, 1986, and Components)                              July 2, 1990; IR 390, 391/86-04
GL 83-028    "Required Actions Based on Generic      S      TVA: letters dated November 7, 1983, January 17, 1986 and Implications of Salem ATWS Events:              November 1, 1993 02 3.2 - Post-Maintenance Testing (All SR          NRC: letters dated December 10, 1985, October 27, 1986, and Components)                              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.
                                                              --------------------
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:
REVISION 02 UPDATE:
Developmental Revision A of the Unit 2 TS (including the TS Bases) was submitted on March 4, 2009.
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.
The Bases for TS Surveillance Requirement 3.0.1 states, in part, Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2.
GL 83-028    "Required Actions Based on Generic      CI    TVA: letter dated May 19, 1986 Implications of Salem ATWS Events:
GL 83-028    "Required Actions Based on Generic      CI    TVA: letter dated May 19, 1986 Implications of Salem ATWS Events:
                                                              --------------------
4.1 - Reactor Trip System Reliability (Vendor Related Modifications)          Unit 2 Action: Confirm vendor-recommended DS416 breaker modifications are implemented.
4.1 - Reactor Trip System Reliability (Vendor Related Modifications)          Unit 2 Action: Confirm vendor-recommended DS416 breaker modifications are implemented.
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* ITEM                        TITLE                    REV                          ADDITIONAL INFORMATION GL 83-028    "Required Actions Based on Generic          S      TVA: letters dated November 7, 1983, February 10, 1986, and Implications of Salem ATWS Events:                May 19, 1986 02 4.2 - Reactor Trip System Reliability              NRC: letters dated July 26, 1985 and June 18, 1992; SSER 16 (Preventive Maintenance and Surveillance Program for Reactor            --------------------
 
ITEM                        TITLE                    REV                          ADDITIONAL INFORMATION GL 83-028    "Required Actions Based on Generic          S      TVA: letters dated November 7, 1983, February 10, 1986, and Implications of Salem ATWS Events:                May 19, 1986 02 4.2 - Reactor Trip System Reliability              NRC: letters dated July 26, 1985 and June 18, 1992; SSER 16 (Preventive Maintenance and Surveillance Program for Reactor            --------------------
Trip Breakers)
Trip Breakers)
Unit 2 Action: Ensure maintenance instruction procedure and Technical Specifications support reliable reactor trip breaker operation.
Unit 2 Action: Ensure maintenance instruction procedure and Technical Specifications support reliable reactor trip breaker operation.
                                                                ----------------------------------------------------------------------------------------
                                                                ----------------------------------------------------------------------------------------
REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision B of the Unit 2 TS was submitted on February 2, 2010.
Developmental Revision B of the Unit 2 TS was submitted on February 2, 2010.
Line 2,412: Line 1,955:
Attachment)
Attachment)
Unit 2 Action: Address in Technical Specifications, as appropriate.
Unit 2 Action: Address in Technical Specifications, as appropriate.
                                                                ----------------------------------------------------------------------------------------
                                                                ----------------------------------------------------------------------------------------
REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
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.
Item 18. (Reactor Trip Breaker Undervoltage and Shunt Trip Mechanisms) of TS Table 3.3.1-1 states the requirement for the shunt trip attachment.
GL 83-029    This GL was never issued.                  NA GL 83-030    Deletion of Standard Technical            NA      Info Specifications Surveillance Requirement 4.8.1.1.2.d.6 For Diesel Generator Testing Page 61 of 95                                              * = See last page for status code definition.
GL 83-029    This GL was never issued.                  NA GL 83-030    Deletion of Standard Technical            NA      Info Specifications Surveillance Requirement 4.8.1.1.2.d.6 For Diesel Generator Testing Page 61 of 95                                              * = See last page for status code definition.
* ITEM                          TITLE                REV                      ADDITIONAL INFORMATION GL 83-031    Safety Evaluation of "Abnormal Transient NA    Applies only to Babcock and Wilcox designed plants Operating Guidelines" GL 83-032    NRC Staff Recommendations Regarding      NA    Info Operator Action for Reactor Trip and ATWS GL 83-033    NRC Positions on Certain Requirements    NA    Info of Appendix R to 10 CFR 50 GL 83-034    This GL was never issued.                NA GL 83-035    Clarification of TMI Action Plan Item    NA    Info II.K.3.31 GL 83-036    NUREG-0737 Technical Specifications      NA    Boiling Water Reactor GL 83-037    NUREG-0737 Technical Specifications      NA    Item was applicable only to units with operating license at the time the item was issued.
 
ITEM                          TITLE                REV                      ADDITIONAL INFORMATION GL 83-031    Safety Evaluation of "Abnormal Transient NA    Applies only to Babcock and Wilcox designed plants Operating Guidelines" GL 83-032    NRC Staff Recommendations Regarding      NA    Info Operator Action for Reactor Trip and ATWS GL 83-033    NRC Positions on Certain Requirements    NA    Info of Appendix R to 10 CFR 50 GL 83-034    This GL was never issued.                NA GL 83-035    Clarification of TMI Action Plan Item    NA    Info II.K.3.31 GL 83-036    NUREG-0737 Technical Specifications      NA    Boiling Water Reactor GL 83-037    NUREG-0737 Technical Specifications      NA    Item was applicable only to units with operating license at the time the item was issued.
GL 83-038    NUREG-0965, "NRC Inventory of Dams"      NA    Info GL 83-039    Voluntary Survey of Licensed Operators  NA    Info GL 83-040    Operator Licensing Examination          NA    Info GL 83-041    Fast Cold Starts of Diesel Generators    NA    Item was applicable only to units with operating license at the time the item was issued.
GL 83-038    NUREG-0965, "NRC Inventory of Dams"      NA    Info GL 83-039    Voluntary Survey of Licensed Operators  NA    Info GL 83-040    Operator Licensing Examination          NA    Info GL 83-041    Fast Cold Starts of Diesel Generators    NA    Item was applicable only to units with operating license at the time the item was issued.
GL 83-042    Clarification to GL 81-07 Regarding      NA    Info Response to NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants" GL 83-043    Reporting Requirements of 10 CFR 50,    NA    Info Sections 50.72 and 50.73, and Standard Technical Specifications Page 62 of 95                                            * = See last page for status code definition.
GL 83-042    Clarification to GL 81-07 Regarding      NA    Info Response to NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants" GL 83-043    Reporting Requirements of 10 CFR 50,    NA    Info Sections 50.72 and 50.73, and Standard Technical Specifications Page 62 of 95                                            * = See last page for status code definition.
* ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 83-044    Availability of NUREG-1021, "Operator      NA    Info Licensing Examiner Standards GL 84-001    NRC Use Of The Terms "Important To        NA    Info Safety" and "Safety Related" GL 84-002    Notice of Meeting Regarding Facility      NA    Info Staffing GL 84-003    Availability of NUREG-0933, "A            NA    Info Prioritization of Generic Safety Issues" GL 84-004    Safety Evaluation of Westinghouse          NA    Info Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops GL 84-005    Change to NUREG-1021, "Operator            NA    Info Licensing Examiner Standards" GL 84-006    Operator and Senior Operator License      NA    Does not apply to power reactor.
 
ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 83-044    Availability of NUREG-1021, "Operator      NA    Info Licensing Examiner Standards GL 84-001    NRC Use Of The Terms "Important To        NA    Info Safety" and "Safety Related" GL 84-002    Notice of Meeting Regarding Facility      NA    Info Staffing GL 84-003    Availability of NUREG-0933, "A            NA    Info Prioritization of Generic Safety Issues" GL 84-004    Safety Evaluation of Westinghouse          NA    Info Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops GL 84-005    Change to NUREG-1021, "Operator            NA    Info Licensing Examiner Standards" GL 84-006    Operator and Senior Operator License      NA    Does not apply to power reactor.
Examination Criteria For Passing Grade GL 84-007    Procedural Guidance for Pipe              NA    Boiling Water Reactor Replacement at BWRs GL 84-008    Interim Procedures for NRC Management      NA    Info of Plant-Specific Backfitting GL 84-009    Recombiner Capability Requirements of      NA    Boiling Water Reactor 10 CFR 50.44(c)(3)(ii)
Examination Criteria For Passing Grade GL 84-007    Procedural Guidance for Pipe              NA    Boiling Water Reactor Replacement at BWRs GL 84-008    Interim Procedures for NRC Management      NA    Info of Plant-Specific Backfitting GL 84-009    Recombiner Capability Requirements of      NA    Boiling Water Reactor 10 CFR 50.44(c)(3)(ii)
GL 84-010    Administration of Operating Tests Prior to NA    Info Initial Criticality GL 84-011    Inspection of BWR Stainless Steel Piping  NA    Boiling Water Reactor GL 84-012    Compliance With 10 CFR Part 61 and        NA    Info Implementation of Radiological Effluent Technical Specifications (RETs) and Attendant Process Control Program (PCP)
GL 84-010    Administration of Operating Tests Prior to NA    Info Initial Criticality GL 84-011    Inspection of BWR Stainless Steel Piping  NA    Boiling Water Reactor GL 84-012    Compliance With 10 CFR Part 61 and        NA    Info Implementation of Radiological Effluent Technical Specifications (RETs) and Attendant Process Control Program (PCP)
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* ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 84-013    Technical Specification for Snubbers      NA    Info GL 84-014    Replacement and Requalification Training  NA    Info Program GL 84-015    Proposed Staff Actions to Improve and      NA    Info Maintain Diesel Generator Reliability GL 84-016    Adequacy of On-Shift Operating            NA    Info Experience for Near Term Operating License Applicants GL 84-017    Annual Meeting to Discuss Recent          NA    Info Developments Regarding Operator Training, Qualifications, and Examinations GL 84-018    Filing of Applications for Licenses and    NA    Does not apply to power reactor.
 
ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 84-013    Technical Specification for Snubbers      NA    Info GL 84-014    Replacement and Requalification Training  NA    Info Program GL 84-015    Proposed Staff Actions to Improve and      NA    Info Maintain Diesel Generator Reliability GL 84-016    Adequacy of On-Shift Operating            NA    Info Experience for Near Term Operating License Applicants GL 84-017    Annual Meeting to Discuss Recent          NA    Info Developments Regarding Operator Training, Qualifications, and Examinations GL 84-018    Filing of Applications for Licenses and    NA    Does not apply to power reactor.
Amendments GL 84-019    Availability of Supplement 1 to            NA    Info NUREG-0933, "A Prioritization of Generic Safety Issues" GL 84-020    Scheduling Guidance for Licensee          NA    Info Submittals of Reloads That Involve Unreviewed Safety Questions GL 84-021    Long Term Low Power Operation in          NA    Info Pressurized Water Reactors GL 84-022    This GL was never issued.                  NA GL 84-023    Reactor Vessel Water Level                NA    Boiling Water Reactor Instrumentation in BWRs GL 84-024    Certification of Compliance to              CI    See Special Program for Environmental Qualification.
Amendments GL 84-019    Availability of Supplement 1 to            NA    Info NUREG-0933, "A Prioritization of Generic Safety Issues" GL 84-020    Scheduling Guidance for Licensee          NA    Info Submittals of Reloads That Involve Unreviewed Safety Questions GL 84-021    Long Term Low Power Operation in          NA    Info Pressurized Water Reactors GL 84-022    This GL was never issued.                  NA GL 84-023    Reactor Vessel Water Level                NA    Boiling Water Reactor Instrumentation in BWRs GL 84-024    Certification of Compliance to              CI    See Special Program for Environmental Qualification.
10 CFR 50.49, Environmental Qualification of Electric Equipment Important To Safety For Nuclear Power Plants GL 85-001    Fire Protection Policy Steering Committee  NA    Only issued as draft Report Page 64 of 95                                              * = See last page for status code definition.
10 CFR 50.49, Environmental Qualification of Electric Equipment Important To Safety For Nuclear Power Plants GL 85-001    Fire Protection Policy Steering Committee  NA    Only issued as draft Report Page 64 of 95                                              * = See last page for status code definition.
* ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 85-002    Recommended Actions Stemming From          CI    TVA responded to the GL on June 17, 1985.
 
ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 85-002    Recommended Actions Stemming From          CI    TVA responded to the GL on June 17, 1985.
NRC Integrated Program for the Resolution of Unresolved Safety Issues            Unit 2 Action: Perform SG inspection.
NRC Integrated Program for the Resolution of Unresolved Safety Issues            Unit 2 Action: Perform SG inspection.
Regarding Steam Generator Tube Integrity GL 85-003    Clarification of Equivalent Control        NA    Boiling Water Reactor Capacity for Standby Liquid Control Systems GL 85-004    Operating Licensing Examinations          NA    Info GL 85-005    Inadvertent Boron Dilution Events          NA    Item was applicable only to units with operating license at the time the item was issued.
Regarding Steam Generator Tube Integrity GL 85-003    Clarification of Equivalent Control        NA    Boiling Water Reactor Capacity for Standby Liquid Control Systems GL 85-004    Operating Licensing Examinations          NA    Info GL 85-005    Inadvertent Boron Dilution Events          NA    Item was applicable only to units with operating license at the time the item was issued.
Line 2,436: Line 1,981:
Unit 2 Action: Implement modifications as required.
Unit 2 Action: Implement modifications as required.
GL 85-013    Transmittal Of NUREG-1154 Regarding        NA    Info The Davis-Besse Loss Of Main And Auxiliary Feedwater Event Page 65 of 95                                              * = See last page for status code definition.
GL 85-013    Transmittal Of NUREG-1154 Regarding        NA    Info The Davis-Besse Loss Of Main And Auxiliary Feedwater Event Page 65 of 95                                              * = See last page for status code definition.
* ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 85-014    Commercial Storage At Power Reactor        NA      Item was applicable only to units with operating license at the Sites Of Low Level Radioactive Waste              time the item was issued.
 
ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 85-014    Commercial Storage At Power Reactor        NA      Item was applicable only to units with operating license at the Sites Of Low Level Radioactive Waste              time the item was issued.
Not Generated By The Utility GL 85-015    Information On Deadlines For              NA      Item was applicable only to units with operating license at the 10 CFR 50.49, "Environmental                      time the item was issued.
Not Generated By The Utility GL 85-015    Information On Deadlines For              NA      Item was applicable only to units with operating license at the 10 CFR 50.49, "Environmental                      time the item was issued.
Qualification Of Electric Equipment Important To Safety At Nuclear Power Plants" GL 85-016    High Boron Concentrations                  NA      Info GL 85-017    Availability Of Supplements 2 and 3 To    NA      Info NUREG-0933, "A Prioritization Of Generic Safety Issues" GL 85-018    Operator Licensing Examinations            NA      Info GL 85-019    Reporting Requirements On Primary          NA      Info Coolant Iodine Spikes GL 85-020    Resolution Of Generic Issue 69: High      NA      Applies only to Babcock and Wilcox designed plants Pressure Injection/Make-up Nozzle Cracking In Babcock And Wilcox Plants GL 85-021    This GL was never issued.                  NA GL 85-022    Potential For Loss Of Post-LOCA            NA      Info Recirculation Capability Due To Insulation Debris Blockage GL 86-001    Safety Concerns Associated With Pipe      NA      Boiling Water Reactor Breaks In The BWR Scram System GL 86-002    Technical Resolution of Generic Issue      NA      Boiling Water Reactor B Thermal Hydraulic Stability GL 86-003    Applications For License Amendments        NA      Info GL 86-004    Policy Statement On Engineering            C      TVA responded to GL 86-04 on May 29, 1986. TVA provides Expertise On Shift                                engineering expertise on shift in the form of a dedicated Shift 01    Technical Advisor (STA) or an STA qualified Senior Reactor Operator.
Qualification Of Electric Equipment Important To Safety At Nuclear Power Plants" GL 85-016    High Boron Concentrations                  NA      Info GL 85-017    Availability Of Supplements 2 and 3 To    NA      Info NUREG-0933, "A Prioritization Of Generic Safety Issues" GL 85-018    Operator Licensing Examinations            NA      Info GL 85-019    Reporting Requirements On Primary          NA      Info Coolant Iodine Spikes GL 85-020    Resolution Of Generic Issue 69: High      NA      Applies only to Babcock and Wilcox designed plants Pressure Injection/Make-up Nozzle Cracking In Babcock And Wilcox Plants GL 85-021    This GL was never issued.                  NA GL 85-022    Potential For Loss Of Post-LOCA            NA      Info Recirculation Capability Due To Insulation Debris Blockage GL 86-001    Safety Concerns Associated With Pipe      NA      Boiling Water Reactor Breaks In The BWR Scram System GL 86-002    Technical Resolution of Generic Issue      NA      Boiling Water Reactor B Thermal Hydraulic Stability GL 86-003    Applications For License Amendments        NA      Info GL 86-004    Policy Statement On Engineering            C      TVA responded to GL 86-04 on May 29, 1986. TVA provides Expertise On Shift                                engineering expertise on shift in the form of a dedicated Shift 01    Technical Advisor (STA) or an STA qualified Senior Reactor Operator.
Page 66 of 95                                              * = See last page for status code definition.
Page 66 of 95                                              * = See last page for status code definition.
* ITEM                          TITLE                  REV                          ADDITIONAL INFORMATION GL 86-005    Implementation Of TMI Action Item        NA      Applies only to Babcock and Wilcox designed plants II.K.3.5, "Automatic Trip Of Reactor Coolant Pumps" GL 86-006    Implementation Of TMI Action Item        NA      Applies only to Combustion Engineering designed plants II.K.3.5, "Automatic Trip of Reactor Coolant Pumps" GL 86-007    Transmittal of NUREG-1190 Regarding      NA      Info The San Onofre Unit 1 Loss of Power and Water Hammer Event GL 86-008    Availability of Supplement 4 to          NA      Info NUREG-0933, "A Prioritization of Generic Safety Issues" GL 86-009    Technical Resolution of Generic Issue      S      N-1 Loop operation was addressed in original 1982 SER (4.4.7).
 
ITEM                          TITLE                  REV                          ADDITIONAL INFORMATION GL 86-005    Implementation Of TMI Action Item        NA      Applies only to Babcock and Wilcox designed plants II.K.3.5, "Automatic Trip Of Reactor Coolant Pumps" GL 86-006    Implementation Of TMI Action Item        NA      Applies only to Combustion Engineering designed plants II.K.3.5, "Automatic Trip of Reactor Coolant Pumps" GL 86-007    Transmittal of NUREG-1190 Regarding      NA      Info The San Onofre Unit 1 Loss of Power and Water Hammer Event GL 86-008    Availability of Supplement 4 to          NA      Info NUREG-0933, "A Prioritization of Generic Safety Issues" GL 86-009    Technical Resolution of Generic Issue      S      N-1 Loop operation was addressed in original 1982 SER (4.4.7).
B-59, (N-1) Loop Operation in BWRs and PWRs                                      02    Unit 2 Action: Confirm Technical Specifications prohibit (N-1) Loop Operation.
B-59, (N-1) Loop Operation in BWRs and PWRs                                      02    Unit 2 Action: Confirm Technical Specifications prohibit (N-1) Loop Operation.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
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Research GL 86-012    Criteria for Unique Purpose Exemption    NA      Does not apply to power reactor.
Research GL 86-012    Criteria for Unique Purpose Exemption    NA      Does not apply to power reactor.
From Conversion From The Use of Heu Fuel GL 86-013    Potential Inconsistency Between Plant    NA      Applies only to Babcock and Wilcox and Combustion Safety Analyses and Technical                    Engineering designed plants Specifications GL 86-014    Operator Licensing Examinations          NA      Info Page 67 of 95                                              * = See last page for status code definition.
From Conversion From The Use of Heu Fuel GL 86-013    Potential Inconsistency Between Plant    NA      Applies only to Babcock and Wilcox and Combustion Safety Analyses and Technical                    Engineering designed plants Specifications GL 86-014    Operator Licensing Examinations          NA      Info Page 67 of 95                                              * = See last page for status code definition.
* ITEM                          TITLE                REV                      ADDITIONAL INFORMATION GL 86-015    Information Relating To Compliance With  NA    Info 10 CFR 50.49, "Environmental Qualification of Electric Equipment Important To Safety For Nuclear Power Plants" GL 86-016    Westinghouse ECCS Evaluation Models      NA    Info GL 86-017    Availability of NUREG-1169, "Technical    NA    Boiling Water Reactor Findings Related to Generic Issue C-8, BWR MSIC Leakage And Treatment Methods" GL 87-001    Public Availability Of The NRC Operator  NA    Info Licensing Examination Question Bank GL 87-002    Verification of Seismic Adequacy of      NA    Item was applicable only to units with operating license at the and      Mechanical and Electrical Equipment in          time the item was issued.
 
ITEM                          TITLE                REV                      ADDITIONAL INFORMATION GL 86-015    Information Relating To Compliance With  NA    Info 10 CFR 50.49, "Environmental Qualification of Electric Equipment Important To Safety For Nuclear Power Plants" GL 86-016    Westinghouse ECCS Evaluation Models      NA    Info GL 86-017    Availability of NUREG-1169, "Technical    NA    Boiling Water Reactor Findings Related to Generic Issue C-8, BWR MSIC Leakage And Treatment Methods" GL 87-001    Public Availability Of The NRC Operator  NA    Info Licensing Examination Question Bank GL 87-002    Verification of Seismic Adequacy of      NA    Item was applicable only to units with operating license at the and      Mechanical and Electrical Equipment in          time the item was issued.
GL 87-003    Operating Reactors, USI A-46 GL 87-004    Temporary Exemption From Provisions Of    NA    Item was applicable only to units with operating license at the The FBI Criminal History Rule For                time the item was issued.
GL 87-003    Operating Reactors, USI A-46 GL 87-004    Temporary Exemption From Provisions Of    NA    Item was applicable only to units with operating license at the The FBI Criminal History Rule For                time the item was issued.
Temporary Workers GL 87-005    Request for Additional Information on    NA    Boiling Water Reactor Assessment of License Measures to Mitigate and/or Identify Potential Degradation of Mark I Drywells GL 87-006    Periodic Verification of Leak Tight      NA    Item was applicable only to units with operating license at the Integrity of Pressure Isolation Valves          time the item was issued.
Temporary Workers GL 87-005    Request for Additional Information on    NA    Boiling Water Reactor Assessment of License Measures to Mitigate and/or Identify Potential Degradation of Mark I Drywells GL 87-006    Periodic Verification of Leak Tight      NA    Item was applicable only to units with operating license at the Integrity of Pressure Isolation Valves          time the item was issued.
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Requirements GL 87-009    Sections 3.0 And 4.0 of Standard Tech    NA    Info Specs on Limiting Conditions For Operation And Surveillance Requirements GL 87-010    Implementation of 10 CFR 73.57,          NA    Item was applicable only to units with operating license at the Requirements For FBI Criminal History            time the item was issued.
Requirements GL 87-009    Sections 3.0 And 4.0 of Standard Tech    NA    Info Specs on Limiting Conditions For Operation And Surveillance Requirements GL 87-010    Implementation of 10 CFR 73.57,          NA    Item was applicable only to units with operating license at the Requirements For FBI Criminal History            time the item was issued.
Checks GL 87-011    Relaxation in Arbitrary Intermediate Pipe NA    Info Rupture Requirements Page 68 of 95                                            * = See last page for status code definition.
Checks GL 87-011    Relaxation in Arbitrary Intermediate Pipe NA    Info Rupture Requirements Page 68 of 95                                            * = See last page for status code definition.
* ITEM                        TITLE                  REV                        ADDITIONAL INFORMATION GL 87-012    Loss of Residual Heat Removal While        C    This GL was superseded by GL 88-17.
 
ITEM                        TITLE                  REV                        ADDITIONAL INFORMATION GL 87-012    Loss of Residual Heat Removal While        C    This GL was superseded by GL 88-17.
The Reactor Coolant System is Partially Filled GL 87-013    Integrity of Requalification Examinations NA    Does not apply to power reactor.
The Reactor Coolant System is Partially Filled GL 87-013    Integrity of Requalification Examinations NA    Does not apply to power reactor.
At Non-Power Reactors GL 87-014    Operator Licensing Examinations          NA    Info GL 87-015    Policy Statement On Deferred Plants      NA    Info GL 87-016    Transmittal of NUREG-1262, "Answers To    NA    Info Questions On Implementation of 10 CFR 55 On Operators' Licenses" GL 88-001    NRC Position on IGSCC in BWR              NA    Boiling Water Reactor Austenitic Stainless Steel Piping GL 88-002    Integrated Safety Assessment Program II  NA    Item was applicable only to units with operating license at the time the item was issued.
At Non-Power Reactors GL 87-014    Operator Licensing Examinations          NA    Info GL 87-015    Policy Statement On Deferred Plants      NA    Info GL 87-016    Transmittal of NUREG-1262, "Answers To    NA    Info Questions On Implementation of 10 CFR 55 On Operators' Licenses" GL 88-001    NRC Position on IGSCC in BWR              NA    Boiling Water Reactor Austenitic Stainless Steel Piping GL 88-002    Integrated Safety Assessment Program II  NA    Item was applicable only to units with operating license at the time the item was issued.
GL 88-003    Resolution of GSI 93, Steam Binding of    CI    TVA: letter June 3, 1988. NRC letters dated Auxiliary Feedwater Pumps                      February 17, 1988 and July 20, 1988 NRC: SSER 16
GL 88-003    Resolution of GSI 93, Steam Binding of    CI    TVA: letter June 3, 1988. NRC letters dated Auxiliary Feedwater Pumps                      February 17, 1988 and July 20, 1988 NRC: SSER 16 NRC accepted approach in letter dated July 20, 1988, and reviewed response in Appendix EE of SSER16.
                                                              --------------------
NRC accepted approach in letter dated July 20, 1988, and reviewed response in Appendix EE of SSER16.
Unit 2 Action: Procedures and hardware will be in place to ensure recognition of indications of steam binding and maintenance of system operability until check valves are repaired and back leakage stopped.
Unit 2 Action: Procedures and hardware will be in place to ensure recognition of indications of steam binding and maintenance of system operability until check valves are repaired and back leakage stopped.
GL 88-004    Distribution of Gems Irradiated in        NA    Does not apply to power reactor.
GL 88-004    Distribution of Gems Irradiated in        NA    Does not apply to power reactor.
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Reactor Pressure Boundary Components in PWR plants                                    Unit 2 Action: Implement program.
Reactor Pressure Boundary Components in PWR plants                                    Unit 2 Action: Implement program.
GL 88-006    Removal of Organization Charts from      NA    Info Technical Specification Administrative Control Requirements Page 69 of 95                                            * = See last page for status code definition.
GL 88-006    Removal of Organization Charts from      NA    Info Technical Specification Administrative Control Requirements Page 69 of 95                                            * = See last page for status code definition.
* ITEM                          TITLE                  REV                          ADDITIONAL INFORMATION GL 88-007    Modified Enforcement Policy Relating to    CI    See Special Program for Environmental Qualification.
 
ITEM                          TITLE                  REV                          ADDITIONAL INFORMATION GL 88-007    Modified Enforcement Policy Relating to    CI    See Special Program for Environmental Qualification.
10 CFR 50.49, Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants GL 88-008    Mail Sent or Delivered to the Office of  NA      Info Nuclear Reactor Regulation GL 88-009    Pilot Testing of Fundamentals            NA      Boiling Water Reactor Examination GL 88-010    Purchase of GSA Approved Security        NA      Info Containers GL 88-011    NRC Position on Radiation Embrittlement    S      NRC acceptance letter dated June 29, 1989, for both units.
10 CFR 50.49, Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants GL 88-008    Mail Sent or Delivered to the Office of  NA      Info Nuclear Reactor Regulation GL 88-009    Pilot Testing of Fundamentals            NA      Boiling Water Reactor Examination GL 88-010    Purchase of GSA Approved Security        NA      Info Containers GL 88-011    NRC Position on Radiation Embrittlement    S      NRC acceptance letter dated June 29, 1989, for both units.
of Reactor Vessel Material and its Impact on Plant Operations                        02    Unit 2 Action: Submit Pressure Temperature curves.
of Reactor Vessel Material and its Impact on Plant Operations                        02    Unit 2 Action: Submit Pressure Temperature curves.
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                                                                ----------------------------------------------------------------------------------------
REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
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GL 88-012    Removal of Fire Protection Requirements  NA      Info from Technical Specification GL 88-013    Operator Licensing Examinations          NA      Info GL 88-014    Instrument Air Supply System Problems      CI    NRC letter dated July 26, 1990, closing the issue.
GL 88-012    Removal of Fire Protection Requirements  NA      Info from Technical Specification GL 88-013    Operator Licensing Examinations          NA      Info GL 88-014    Instrument Air Supply System Problems      CI    NRC letter dated July 26, 1990, closing the issue.
Affecting Safety-Related Equipment 04    Unit 2 Action: Complete Unit 2 implementation.
Affecting Safety-Related Equipment 04    Unit 2 Action: Complete Unit 2 implementation.
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REVISION 04 UPDATE:
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.
The compressed air system is a common system at Watts Bar; therefore, the requirements for this GL have been satisfied for Unit 2.
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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.
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.
Page 70 of 95                                              * = See last page for status code definition.
Page 70 of 95                                              * = See last page for status code definition.
* ITEM                        TITLE                REV                          ADDITIONAL INFORMATION GL 88-015    Electric Power Systems - Inadequate    NA      Info Control Over Design Process GL 88-016    Removal of Cycle-Specific Parameter    NA      Info Limits from Technical Specifications GL 88-017    Loss of Decay Heat Removal              CI    NRC acceptance letter dated March 8, 1995 (Unit 1).
 
ITEM                        TITLE                REV                          ADDITIONAL INFORMATION GL 88-015    Electric Power Systems - Inadequate    NA      Info Control Over Design Process GL 88-016    Removal of Cycle-Specific Parameter    NA      Info Limits from Technical Specifications GL 88-017    Loss of Decay Heat Removal              CI    NRC acceptance letter dated March 8, 1995 (Unit 1).
Unit 2 Action: Implement modifications to provide RCS temperature, RV level and RHR system performance.
Unit 2 Action: Implement modifications to provide RCS temperature, RV level and RHR system performance.
GL 88-018    Plant Record Storage on Optical Disks  NA      Info GL 88-019    Use of Deadly Force by Licensee Guards  NA      Does not apply to power reactor.
GL 88-018    Plant Record Storage on Optical Disks  NA      Info GL 88-019    Use of Deadly Force by Licensee Guards  NA      Does not apply to power reactor.
to Prevent Theft of Special Nuclear Material GL 88-020    Individual Plant Examination for Severe  S      Unit 2 Action: Complete evaluation for Unit 2.
to Prevent Theft of Special Nuclear Material GL 88-020    Individual Plant Examination for Severe  S      Unit 2 Action: Complete evaluation for Unit 2.
Accident Vulnerabilities 04    ----------------------------------------------------------------------------------------
Accident Vulnerabilities 04    ----------------------------------------------------------------------------------------
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
The Probabilistic Risk Assessment Individual Plant Examination Summary Report was submitted on February 9, 2010.
The Probabilistic Risk Assessment Individual Plant Examination Summary Report was submitted on February 9, 2010.
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REVISION 04 UPDATE:
REVISION 04 UPDATE:
The Individual Plant Examination of External Events Design Report was submitted on April 30, 2010.
The Individual Plant Examination of External Events Design Report was submitted on April 30, 2010.
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Inservice Testing Programs Unit 2 Action: Submit an ASME Section XI Inservice Test Program for the first ten year interval six months before receiving an Operating License.
Inservice Testing Programs Unit 2 Action: Submit an ASME Section XI Inservice Test Program for the first ten year interval six months before receiving an Operating License.
Page 71 of 95                                            * = See last page for status code definition.
Page 71 of 95                                            * = See last page for status code definition.
* ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 89-005    Pilot Testing of the Fundamentals        NA    Info Examination GL 89-006    Task Action Plan Item I.D.2 - Safety      CI    Safety Parameter Display System (SPDS) / Requirements for Parameter Display System - 10 CFR                Emergency Response Capability - NRC reviewed in SSER5, 50.54(f)                                        SSER6, and 18.2.2 of SSER15.
 
ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 89-005    Pilot Testing of the Fundamentals        NA    Info Examination GL 89-006    Task Action Plan Item I.D.2 - Safety      CI    Safety Parameter Display System (SPDS) / Requirements for Parameter Display System - 10 CFR                Emergency Response Capability - NRC reviewed in SSER5, 50.54(f)                                        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.
Unit 2 Action: Install SPDS and have it operational prior to start-up after the first refueling outage.
GL 89-007    Power Reactor Safeguards Contingency      C    TVA: letter dated October 31, 1989 Planning for Surface Vehicle Bombs NRC: memo dated June 26, 1990 GL 89-008    Erosion/Corrosion-Induced Pipe Wall        CI    Unit 1 Flow Accelerated Corrosion Program reviewed in Thinning                                        IR 390/94-89 (February 1995).
GL 89-007    Power Reactor Safeguards Contingency      C    TVA: letter dated October 31, 1989 Planning for Surface Vehicle Bombs NRC: memo dated June 26, 1990 GL 89-008    Erosion/Corrosion-Induced Pipe Wall        CI    Unit 1 Flow Accelerated Corrosion Program reviewed in Thinning                                        IR 390/94-89 (February 1995).
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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.
Unit 2 Actions: 1) Implement initial performance testing of the heat exchangers; and 2) Establish eddy current baseline data for the Containment Spray heat exchangers.
GL 89-014    Line-Item Improvements in Technical      NA    Info Specifications - Removal of 3.25 Limit on Extending Surveillance Intervals Page 72 of 95                                            * = See last page for status code definition.
GL 89-014    Line-Item Improvements in Technical      NA    Info Specifications - Removal of 3.25 Limit on Extending Surveillance Intervals Page 72 of 95                                            * = See last page for status code definition.
* ITEM                          TITLE                  REV                          ADDITIONAL INFORMATION GL 89-015    Emergency Response Data System            NA      Info GL 89-016    Installation of a Hardened Wetwell Vent  NA      Boiling Water Reactor GL 89-017    Planned Administrative Changes to the    NA      Info NRC Operator Licensing Written Examination Process GL 89-018    Resolution of Unresolved Safety Issues    NA      Info A-17, "Systems Interactions in Nuclear Power Plants" GL 89-019    Request for Actions Related to Resolution  CI    TVA responded by letter dated March 22, 1990. NRC of Unresolved Safety Issue A-47, "Safety          acceptance letter dated October 24, 1990, for both units.
 
ITEM                          TITLE                  REV                          ADDITIONAL INFORMATION GL 89-015    Emergency Response Data System            NA      Info GL 89-016    Installation of a Hardened Wetwell Vent  NA      Boiling Water Reactor GL 89-017    Planned Administrative Changes to the    NA      Info NRC Operator Licensing Written Examination Process GL 89-018    Resolution of Unresolved Safety Issues    NA      Info A-17, "Systems Interactions in Nuclear Power Plants" GL 89-019    Request for Actions Related to Resolution  CI    TVA responded by letter dated March 22, 1990. NRC of Unresolved Safety Issue A-47, "Safety          acceptance letter dated October 24, 1990, for both units.
Implication of Control Systems in LWR Nuclear Power Plants" Pursuant to                Unit 2 Action: Perform evaluation of common mode failures due 10 CFR 50.54(f)                                  to fire.
Implication of Control Systems in LWR Nuclear Power Plants" Pursuant to                Unit 2 Action: Perform evaluation of common mode failures due 10 CFR 50.54(f)                                  to fire.
GL 89-020    Protected Area Long-Term Housekeeping    NA      Does not apply to power reactor.
GL 89-020    Protected Area Long-Term Housekeeping    NA      Does not apply to power reactor.
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Unit 2 actions: Provide a status of WBN Unit 2 USIs.
Unit 2 actions: Provide a status of WBN Unit 2 USIs.
Complete implementation of USIs.
Complete implementation of USIs.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Status of USIs was provided by Enclosure 2 of TVA letter dated September 26, 2008.
Status of USIs was provided by Enclosure 2 of TVA letter dated September 26, 2008.
Line 2,525: Line 2,065:
Recent Change In Probable Maximum Precipitation Criteria Developed by the          Answer to informal question provided in TVA letter dated National Weather Service                          December 16, 1981, and subsequently included in FSAR. GL did not require a response. No further action required.
Recent Change In Probable Maximum Precipitation Criteria Developed by the          Answer to informal question provided in TVA letter dated National Weather Service                          December 16, 1981, and subsequently included in FSAR. GL did not require a response. No further action required.
Page 73 of 95                                              * = See last page for status code definition.
Page 73 of 95                                              * = See last page for status code definition.
* ITEM                          TITLE                  REV                          ADDITIONAL INFORMATION GL 89-023    NRC Staff Responses to Questions          NA      Info Pertaining to Implementation of 10 CFR Part 26 GL 90-001    Request for Voluntary Participation in    NA      Info NRC Regulatory Impact Survey GL 90-002    Alternative Requirements for Fuel          NA      Info Assemblies in the Design Features Section of Technical Specifications GL 90-003    Relaxation of Staff Position in Generic    NA      Info Letter 83-28, Item 2.2 Part 2 "Vendor Interface for Safety-Related Components" GL 90-004    Request for Information on the Status of    C      TVA responded on June 23, 1990 Licensee Implementation of GSIs Resolved with Imposition of Requirements or CAs GL 90-005    Guidance for Performing Temporary          NA      Info Non-Code Repair of ASME Code Class 1, 2, and 3 Piping GL 90-006    Resolution of Generic Issues 70, "PORV      S      NRC letter dated January 9, 1991, accepted TVAs response for and Block Valve Reliability," and 94,              both units.
 
ITEM                          TITLE                  REV                          ADDITIONAL INFORMATION GL 89-023    NRC Staff Responses to Questions          NA      Info Pertaining to Implementation of 10 CFR Part 26 GL 90-001    Request for Voluntary Participation in    NA      Info NRC Regulatory Impact Survey GL 90-002    Alternative Requirements for Fuel          NA      Info Assemblies in the Design Features Section of Technical Specifications GL 90-003    Relaxation of Staff Position in Generic    NA      Info Letter 83-28, Item 2.2 Part 2 "Vendor Interface for Safety-Related Components" GL 90-004    Request for Information on the Status of    C      TVA responded on June 23, 1990 Licensee Implementation of GSIs Resolved with Imposition of Requirements or CAs GL 90-005    Guidance for Performing Temporary          NA      Info Non-Code Repair of ASME Code Class 1, 2, and 3 Piping GL 90-006    Resolution of Generic Issues 70, "PORV      S      NRC letter dated January 9, 1991, accepted TVAs response for and Block Valve Reliability," and 94,              both units.
               "Additional LTOP Protection for PWRs"      02 Unit 2 Actions: 1) Revise operating instruction and surveillance procedure; and 2) Incorporate testing requirements in the Technical Specifications.
               "Additional LTOP Protection for PWRs"      02 Unit 2 Actions: 1) Revise operating instruction and surveillance procedure; and 2) Incorporate testing requirements in the Technical Specifications.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.
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.
TS Surveillance Requirement 3.4.11.2 specifies the required testing of each PORV.
GL 90-007    Operator Licensing National Examination    NA      Info Schedule GL 90-008    Simulation Facility Exemptions            NA      Info GL 90-009    Alternative Requirements for Snubber      NA      Info Visual Inspection Intervals and Corrective Actions Page 74 of 95                                              * = See last page for status code definition.
GL 90-007    Operator Licensing National Examination    NA      Info Schedule GL 90-008    Simulation Facility Exemptions            NA      Info GL 90-009    Alternative Requirements for Snubber      NA      Info Visual Inspection Intervals and Corrective Actions Page 74 of 95                                              * = See last page for status code definition.
* ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 91-001    Removal of the Schedule for the          NA    Info Withdrawal of Reactor Vessel Material Specimens from Technical Specifications GL 91-002    Reporting Mishaps Involving LLW Forms    NA    Item was applicable only to units with operating license at the Prepared for Disposal                            time the item was issued.
 
ITEM                          TITLE                  REV                      ADDITIONAL INFORMATION GL 91-001    Removal of the Schedule for the          NA    Info Withdrawal of Reactor Vessel Material Specimens from Technical Specifications GL 91-002    Reporting Mishaps Involving LLW Forms    NA    Item was applicable only to units with operating license at the Prepared for Disposal                            time the item was issued.
GL 91-003    Reporting of Safeguards Events            NA    Info GL 91-004    Changes in Technical Specification        NA    Info Surveillance Intervals to Accommodate a 24-Month Fuel Cycle GL 91-005    Licensee Commercial-Grade                NA    Info Procurement and Dedication Programs GL 91-006    Resolution of Generic Issue A-30,        NA    Item was applicable only to units with operating license at the Adequacy of Safety-Related DC Power            time the item was issued.
GL 91-003    Reporting of Safeguards Events            NA    Info GL 91-004    Changes in Technical Specification        NA    Info Surveillance Intervals to Accommodate a 24-Month Fuel Cycle GL 91-005    Licensee Commercial-Grade                NA    Info Procurement and Dedication Programs GL 91-006    Resolution of Generic Issue A-30,        NA    Item was applicable only to units with operating license at the Adequacy of Safety-Related DC Power            time the item was issued.
Supplies," Pursuant to 10 CFR 50.54(f)
Supplies," Pursuant to 10 CFR 50.54(f)
Line 2,541: Line 2,081:
Resolution of Generic Issue 130, Essential Service Water System Failures
Resolution of Generic Issue 130, Essential Service Water System Failures
               @ Multi-Unit Sites Page 75 of 95                                            * = See last page for status code definition.
               @ Multi-Unit Sites Page 75 of 95                                            * = See last page for status code definition.
* ITEM                          TITLE              REV                      ADDITIONAL INFORMATION GL 91-014    Emergency Telecommunications            NA    Info GL 91-015    Operating Experience Feedback Report,  NA    Info Solenoid-Operated Valve Problems at U.S. Reactors GL 91-016    Licensed Operators' and Other Nuclear  NA    Info Facility Personnel Fitness for Duty GL 91-017    Generic Safety Issue 29, "Bolting      NA    Info Degradation or Failure in Nuclear Power Plants" GL 91-018    Information to Licensees Regarding Two  NA    GL 91-18 has been superseded by RIS 2005-20.
 
ITEM                          TITLE              REV                      ADDITIONAL INFORMATION GL 91-014    Emergency Telecommunications            NA    Info GL 91-015    Operating Experience Feedback Report,  NA    Info Solenoid-Operated Valve Problems at U.S. Reactors GL 91-016    Licensed Operators' and Other Nuclear  NA    Info Facility Personnel Fitness for Duty GL 91-017    Generic Safety Issue 29, "Bolting      NA    Info Degradation or Failure in Nuclear Power Plants" GL 91-018    Information to Licensees Regarding Two  NA    GL 91-18 has been superseded by RIS 2005-20.
NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability GL 91-019    Information to Addressees Regarding    NA    Info New Telephone Numbers for NRC Offices Located in One White Flint North GL 92-001    Reactor Vessel Structural Integrity      C    By letter dated May 11, 1994, for both units NRC confirmed TVA had provided the information requested in GL 92-01. NRC issued GL 92-01 revision 1, supplement 1 on May 19, 1995.
NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability GL 91-019    Information to Addressees Regarding    NA    Info New Telephone Numbers for NRC Offices Located in One White Flint North GL 92-001    Reactor Vessel Structural Integrity      C    By letter dated May 11, 1994, for both units NRC confirmed TVA had provided the information requested in GL 92-01. NRC issued GL 92-01 revision 1, supplement 1 on May 19, 1995.
By letter dated July 26, 1996, NRC closed GL 92-01, Revision 1, Supplement 1 for both Watts Bar units.
By letter dated July 26, 1996, NRC closed GL 92-01, Revision 1, Supplement 1 for both Watts Bar units.
Line 2,547: Line 2,088:
Thermal Stress During Natural Convection Cooldown" GL 92-003    Compilation of the Current Licensing    NA    Info Basis: Request for Voluntary Participation in Pilot Program GL 92-004    Resolution of the Issues Related to    NA    Boiling Water Reactor Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(f)
Thermal Stress During Natural Convection Cooldown" GL 92-003    Compilation of the Current Licensing    NA    Info Basis: Request for Voluntary Participation in Pilot Program GL 92-004    Resolution of the Issues Related to    NA    Boiling Water Reactor Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(f)
GL 92-005    NRC Workshop on the Systematic          NA    Info Assessment of Licensee Performance (SALP) Program GL 92-006    Operator Licensing National Examination NA    Info Schedule GL 92-007    Office of Nuclear Reactor Regulation    NA    Info Reorganization Page 76 of 95                                          * = See last page for status code definition.
GL 92-005    NRC Workshop on the Systematic          NA    Info Assessment of Licensee Performance (SALP) Program GL 92-006    Operator Licensing National Examination NA    Info Schedule GL 92-007    Office of Nuclear Reactor Regulation    NA    Info Reorganization Page 76 of 95                                          * = See last page for status code definition.
* ITEM                          TITLE                    REV                      ADDITIONAL INFORMATION GL 92-008    Thermo-Lag 330-1 Fire Barriers              OV    TVA configurations for Thermo-Lag 330-1 were reviewed in SSER18 and accepted in NRC letter dated January 6, 1998 (includes a supplemental SE).
 
ITEM                          TITLE                    REV                      ADDITIONAL INFORMATION GL 92-008    Thermo-Lag 330-1 Fire Barriers              OV    TVA configurations for Thermo-Lag 330-1 were reviewed in SSER18 and accepted in NRC letter dated January 6, 1998 (includes a supplemental SE).
Unit 2 Actions: 1) Review Watts Bar design and installation requirements for Thermolag 330-1 fire barrier system and evaluate the Thermolag currently installed in Unit 2.
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.
: 2) Remove and replace, as required, or prepare an approved deviation.
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Limits GL 94-001    Removal of Accelerated Testing and          NA    Item was applicable only to units with operating license at the Special Reporting Requirements for                time the item was issued.
Limits GL 94-001    Removal of Accelerated Testing and          NA    Item was applicable only to units with operating license at the Special Reporting Requirements for                time the item was issued.
Emergency Diesel Generators GL 94-002    Long-Term Solutions and Upgrade of          NA    Boiling Water Reactor Interim Operating Recommendations for Thermal-Hydraulic Instabilities in BWRs Page 77 of 95                                              * = See last page for status code definition.
Emergency Diesel Generators GL 94-002    Long-Term Solutions and Upgrade of          NA    Boiling Water Reactor Interim Operating Recommendations for Thermal-Hydraulic Instabilities in BWRs Page 77 of 95                                              * = See last page for status code definition.
* ITEM                          TITLE                REV                          ADDITIONAL INFORMATION GL 94-003    IGSCC of Core Shrouds in BWRs          NA      Boiling Water Reactor GL 94-004    Voluntary Reporting of Additional      NA      Info Occupational Radiation Exposure Data GL 95-001    NRC Staff Technical Position on Fire    NA      Does not apply to power reactor.
 
ITEM                          TITLE                REV                          ADDITIONAL INFORMATION GL 94-003    IGSCC of Core Shrouds in BWRs          NA      Boiling Water Reactor GL 94-004    Voluntary Reporting of Additional      NA      Info Occupational Radiation Exposure Data GL 95-001    NRC Staff Technical Position on Fire    NA      Does not apply to power reactor.
Protection for Fuel Cycle Facilities GL 95-002    Use of NUMARC/EPRI Report TR-          NA      Info 102348, "Guideline on Licensing Digital Upgrades," in Determining the Acceptability of Performing Analog-to-Digital Replacements under 10 CFR 50.59 GL 95-003    Circumferential Cracking of Steam        CI    NRC acceptance letter dated May 16, 1997 (Unit 1) - Initial Generator Tubes                                response for Unit 2 on September 7, 2007. TVA responded to a 02    request for additional information on December 17, 2007.
Protection for Fuel Cycle Facilities GL 95-002    Use of NUMARC/EPRI Report TR-          NA      Info 102348, "Guideline on Licensing Digital Upgrades," in Determining the Acceptability of Performing Analog-to-Digital Replacements under 10 CFR 50.59 GL 95-003    Circumferential Cracking of Steam        CI    NRC acceptance letter dated May 16, 1997 (Unit 1) - Initial Generator Tubes                                response for Unit 2 on September 7, 2007. TVA responded to a 02    request for additional information on December 17, 2007.
Unit 2 Action: Perform baseline inspection.
Unit 2 Action: Perform baseline inspection.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Unit 2 Action: Perform baseline inspection. Evaluate or repair as necessary.
Unit 2 Action: Perform baseline inspection. Evaluate or repair as necessary.
Line 2,568: Line 2,109:
GL 95-004    Final Disposition of the Systematic    NA      Info Evaluation Program Lessons-Learned Issues GL 95-005    Voltage-Based Repair Criteria for        C      No specific action or response required by the GL; TVA Westinghouse Steam Generator Tubes              responded on September 7, 2007.
GL 95-004    Final Disposition of the Systematic    NA      Info Evaluation Program Lessons-Learned Issues GL 95-005    Voltage-Based Repair Criteria for        C      No specific action or response required by the GL; TVA Westinghouse Steam Generator Tubes              responded on September 7, 2007.
Affected by Outside Diameter Stress      02 Corrosion Cracking                              ----------------------------------------------------------------------------------------
Affected by Outside Diameter Stress      02 Corrosion Cracking                              ----------------------------------------------------------------------------------------
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REVISION 02 UPDATE:
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.
On January 21, 2010, NRC issued the Safety Evaluation for the following Generic Letters: 1995-03, 1995-05, 1997-05, 1997-06, 2004-01, and 2006-01.
GL 95-006    Changes in the Operator Licensing      NA      Info Program Page 78 of 95                                            * = See last page for status code definition.
GL 95-006    Changes in the Operator Licensing      NA      Info Program Page 78 of 95                                            * = See last page for status code definition.
* ITEM                          TITLE                REV                          ADDITIONAL INFORMATION GL 95-007    Pressure Locking and Thermal Binding of  CI    Unit 1 SER for GL 95-07 dated Sept 15, 1999 Safety-Related Power-Operated Gate Valves                                  04    Unit 2 Action: Perform evaluation for pressure locking and thermal binding of safety related power-operated gate valves and take corrective actions for those valves identified as being susceptible.
 
                                                              ----------------------------------------------------------------------------------------
ITEM                          TITLE                REV                          ADDITIONAL INFORMATION GL 95-007    Pressure Locking and Thermal Binding of  CI    Unit 1 SER for GL 95-07 dated Sept 15, 1999 Safety-Related Power-Operated Gate Valves                                  04    Unit 2 Action: Perform evaluation for pressure locking and thermal binding of safety related power-operated gate valves and take corrective actions for those valves identified as being susceptible.
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REVISION 03 UPDATE:
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.
April 1, 2010, letter committed to evaluate missing GL 89-10 motor-operated valves for susceptibility to pressure locking and thermal binding.
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REVISION 04 UPDATE:
REVISION 04 UPDATE:
NRC letter dated July 29, 2010, provided RAIs on the GL.
NRC letter dated July 29, 2010, provided RAIs on the GL.
Line 2,584: Line 2,121:
EDCRs 53292 and 53287 shall be implemented to eliminate the potential for pressure locking prior to startup.
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.
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.
NRC issued the Safety Evaluation for GL 1995-007 on August 12, 2010.
GL 95-008    10 CFR 50.54(p) Process for Changes to  NA      Info Security Plans Without Prior NRC Approval GL 95-009    Monitoring and Training of Shippers and NA      Info Carriers of Radioactive Materials GL 95-010    Relocation of Selected Technical        NA      Info Specifications Requirements Related to Instrumentation GL 96-001    Testing of Safety-Related Circuits      CI    TVA responded for both units on April 18, 1996.
GL 95-008    10 CFR 50.54(p) Process for Changes to  NA      Info Security Plans Without Prior NRC Approval GL 95-009    Monitoring and Training of Shippers and NA      Info Carriers of Radioactive Materials GL 95-010    Relocation of Selected Technical        NA      Info Specifications Requirements Related to Instrumentation GL 96-001    Testing of Safety-Related Circuits      CI    TVA responded for both units on April 18, 1996.
Unit 2 Action: Implement Recommendations.
Unit 2 Action: Implement Recommendations.
GL 96-002    Reconsideration of Nuclear Power Plant  NA      Info Security Requirements Associated with an Internal Threat Page 79 of 95                                            * = See last page for status code definition.
GL 96-002    Reconsideration of Nuclear Power Plant  NA      Info Security Requirements Associated with an Internal Threat Page 79 of 95                                            * = See last page for status code definition.
* ITEM                          TITLE                REV                          ADDITIONAL INFORMATION GL 96-003    Relocation of the Pressure Temperature  CI    No response required Limit Curves and Low Temperature Overpressure Protection System Limits          Unit 2 Action: Submit Pressure Temperature limits and similar to Unit 1, upon approval, incorporate into licensee-controlled document.
 
ITEM                          TITLE                REV                          ADDITIONAL INFORMATION GL 96-003    Relocation of the Pressure Temperature  CI    No response required Limit Curves and Low Temperature Overpressure Protection System Limits          Unit 2 Action: Submit Pressure Temperature limits and similar to Unit 1, upon approval, incorporate into licensee-controlled document.
GL 96-004    Boraflex Degradation in Spent Fuel Pool NA      Item was applicable only to units with operating license at the Storage Racks                                  time the item was issued.
GL 96-004    Boraflex Degradation in Spent Fuel Pool NA      Item was applicable only to units with operating license at the Storage Racks                                  time the item was issued.
GL 96-005    Periodic Verification of Design-Basis    CI    SE of TVA response to GL 96-05 dated July 21, 1999.
GL 96-005    Periodic Verification of Design-Basis    CI    SE of TVA response to GL 96-05 dated July 21, 1999.
Line 2,595: Line 2,132:
GL 96-006    Assurance of Equipment Operability and  CI    NRC letter dated April 6, 1999, accepting TVA response for Containment Integrity During                    Unit 1.
GL 96-006    Assurance of Equipment Operability and  CI    NRC letter dated April 6, 1999, accepting TVA response for Containment Integrity During                    Unit 1.
Design-Basis Accident Conditions        02 Unit 2 Action: Implement modification to provide containment penetration relief.
Design-Basis Accident Conditions        02 Unit 2 Action: Implement modification to provide containment penetration relief.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
NRC issued the Safety Evaluation for Generic Letter 1996-006 on January 21, 2010.
NRC issued the Safety Evaluation for Generic Letter 1996-006 on January 21, 2010.
Line 2,602: Line 2,137:
GL 97-001    Degradation of Control Rod Drive        CI    NRC acceptance letter dated November 4, 1999 (Unit 1).
GL 97-001    Degradation of Control Rod Drive        CI    NRC acceptance letter dated November 4, 1999 (Unit 1).
Mechanism Nozzle and Other Vessel Closure Head Penetrations                04    Unit 2 Action: Provide a report to address the inspection program.
Mechanism Nozzle and Other Vessel Closure Head Penetrations                04    Unit 2 Action: Provide a report to address the inspection program.
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REVISION 03 UPDATE:
REVISION 03 UPDATE:
NRC issued the Safety Evaluation for Generic Letter 97-001 on June 30, 2010.
NRC issued the Safety Evaluation for Generic Letter 97-001 on June 30, 2010.
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                                                              ----------------------------------------------------------------------------------------
REVISION 04 UPDATE:
REVISION 04 UPDATE:
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
GL 97-002    Revised Contents of the Monthly        NA      Item was applicable only to units with operating license at the Operating Report                                time the item was issued.
GL 97-002    Revised Contents of the Monthly        NA      Item was applicable only to units with operating license at the Operating Report                                time the item was issued.
Page 80 of 95                                            * = See last page for status code definition.
Page 80 of 95                                            * = See last page for status code definition.
* ITEM                        TITLE                  REV                          ADDITIONAL INFORMATION GL 97-003    Annual Financial Update of Surety        NA      Does not apply to power reactor.
 
ITEM                        TITLE                  REV                          ADDITIONAL INFORMATION GL 97-003    Annual Financial Update of Surety        NA      Does not apply to power reactor.
Requirements for Uranium Recovery Licensees GL 97-004    Assurance of Sufficient Net Positive      CI    NRC acceptance letter dated June 17, 1998 (Unit 1) - Initial Suction Head for Emergency Core                  response for Unit 2 on September 7, 2007.
Requirements for Uranium Recovery Licensees GL 97-004    Assurance of Sufficient Net Positive      CI    NRC acceptance letter dated June 17, 1998 (Unit 1) - Initial Suction Head for Emergency Core                  response for Unit 2 on September 7, 2007.
Cooling and Containment Heat Removal      02 Pumps                                            Unit 2 Actions: Install new sump strainers, and perform other modification-related activities identical to Unit 1.
Cooling and Containment Heat Removal      02 Pumps                                            Unit 2 Actions: Install new sump strainers, and perform other modification-related activities identical to Unit 1.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
NRC issued the Safety Evaluation for Generic Letter 1997-004 on February 18, 2010.
NRC issued the Safety Evaluation for Generic Letter 1997-004 on February 18, 2010.
Line 2,622: Line 2,152:
Techniques                                      Initial response for Unit 2 on September 7, 2007.
Techniques                                      Initial response for Unit 2 on September 7, 2007.
02 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.
02 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.
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REVISION 02 UPDATE:
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.
On January 21, 2010, NRC issued the Safety Evaluation for the following Generic Letters: 1995-03, 1995-05, 1997-05, 1997-06, 2004-01, and 2006-01.
GL 97-006    Degradation of Steam Generator Internals  CI    NRC acceptance letter dated October 19, 1999 (Unit 1) - Initial response for Unit 2 on September 7, 2007. TVA responded to a 02    request for additional information on December 17, 2007.
GL 97-006    Degradation of Steam Generator Internals  CI    NRC acceptance letter dated October 19, 1999 (Unit 1) - Initial response for Unit 2 on September 7, 2007. TVA responded to a 02    request for additional information on December 17, 2007.
Unit 2 Action: Perform SG inspections during each refueling outage.
Unit 2 Action: Perform SG inspections during each refueling outage.
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REVISION 02 UPDATE:
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.
On January 21, 2010, NRC issued the Safety Evaluation for the following Generic Letters: 1995-03, 1995-05, 1997-05, 1997-06, 2004-01, and 2006-01.
GL 98-001    Year 2000 Readiness of Computer          NA      Item was applicable only to units with operating license at the Systems at Nuclear Power Plants                  time the item was issued.
GL 98-001    Year 2000 Readiness of Computer          NA      Item was applicable only to units with operating license at the Systems at Nuclear Power Plants                  time the item was issued.
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* ITEM                        TITLE                  REV                          ADDITIONAL INFORMATION GL 98-002    Loss of Reactor Coolant Inventory and    CI    Initial response for Unit 2 on September 7, 2007.
 
ITEM                        TITLE                  REV                          ADDITIONAL INFORMATION GL 98-002    Loss of Reactor Coolant Inventory and    CI    Initial response for Unit 2 on September 7, 2007.
Associated Potential for Loss of Emergency Mitigation Functions While in  03    Unit 2 Actions: 1) Review the ECCS designs to ensure they do a Shutdown Condition                            not contain design features which can render them susceptible to common-cause failures; and 2) document the results.
Associated Potential for Loss of Emergency Mitigation Functions While in  03    Unit 2 Actions: 1) Review the ECCS designs to ensure they do a Shutdown Condition                            not contain design features which can render them susceptible to common-cause failures; and 2) document the results.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
NRC issued the Safety Evaluation for Generic Letter 1998-002 on March 3, 2010.
NRC issued the Safety Evaluation for Generic Letter 1998-002 on March 3, 2010.
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REVISION 03 UPDATE:
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.
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.
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.
GL 98-003    NMSS Licensees' and Certificate Holders' NA      Does not apply to power reactor.
GL 98-003    NMSS Licensees' and Certificate Holders' NA      Does not apply to power reactor.
Year 2000 Readiness Programs GL 98-004    Potential for Degradation of the ECCS    OV      NRC closure letter dated November 24, 1999 (Unit 1). - Initial and the Containment Spray System After          response for Unit 2 on September 7, 2007.
Year 2000 Readiness Programs GL 98-004    Potential for Degradation of the ECCS    OV      NRC closure letter dated November 24, 1999 (Unit 1). - Initial and the Containment Spray System After          response for Unit 2 on September 7, 2007.
a LOCA Because of Construction and        02 Protective Coating Deficiencies and              Unit 2 Actions: Install new sump strainers, and perform other Foreign Material in Containment                  modification-related activities identical to Unit 1.
a LOCA Because of Construction and        02 Protective Coating Deficiencies and              Unit 2 Actions: Install new sump strainers, and perform other Foreign Material in Containment                  modification-related activities identical to Unit 1.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
NRC issued the Safety Evaluation for Generic Letter 1998-004 on February 1, 2010.
NRC issued the Safety Evaluation for Generic Letter 1998-004 on February 1, 2010.
GL 98-005    Boiling Water Reactor Licensees Use of  NA      Boiling Water Reactor the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds GL 99-001    Recent Nuclear Material Safety and      NA      Info Safeguards Decision on Bundling Exempt Quantities Page 82 of 95                                            * = See last page for status code definition.
GL 98-005    Boiling Water Reactor Licensees Use of  NA      Boiling Water Reactor the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds GL 99-001    Recent Nuclear Material Safety and      NA      Info Safeguards Decision on Bundling Exempt Quantities Page 82 of 95                                            * = See last page for status code definition.
* ITEM                        TITLE                REV                          ADDITIONAL INFORMATION GL 99-002    Laboratory Testing of Nuclear Grade    NA      Item was applicable only to units with operating license at the Activated Charcoal                            time the item was issued.
 
ITEM                        TITLE                REV                          ADDITIONAL INFORMATION GL 99-002    Laboratory Testing of Nuclear Grade    NA      Item was applicable only to units with operating license at the Activated Charcoal                            time the item was issued.
GL 03-001    Control Room Habitability              S      Initial response for Unit 2 on September 7, 2007 02    Unit 2 Action: Incorporate TSTF-448 into Technical Specifications.
GL 03-001    Control Room Habitability              S      Initial response for Unit 2 on September 7, 2007 02    Unit 2 Action: Incorporate TSTF-448 into Technical Specifications.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
NRC issued the Safety Evaluation for Generic Letter 2003-01 on February 1, 2010.
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.
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 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.
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GL 04-001    Requirements for Steam Generator Tube  CI    NRC acceptance letter dated April 8, 2005 (Unit 1) - Initial Inspection                                    response for Unit 2 on September 7, 2007.
GL 04-001    Requirements for Steam Generator Tube  CI    NRC acceptance letter dated April 8, 2005 (Unit 1) - Initial Inspection                                    response for Unit 2 on September 7, 2007.
02 Unit 2 Action: Perform baseline inspection.
02 Unit 2 Action: Perform baseline inspection.
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REVISION 02 UPDATE:
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.
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.
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Basis Accidents at PWRs Unit 2 Actions: Install new sump strainers, and perform other modification-related activities identical to Unit 1.
Basis Accidents at PWRs Unit 2 Actions: Install new sump strainers, and perform other modification-related activities identical to Unit 1.
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* ITEM                          TITLE                  REV                          ADDITIONAL INFORMATION GL 06-001    Steam Generator Tube Integrity and        S      Initial response for Unit 2 on September 7, 2007.
 
ITEM                          TITLE                  REV                          ADDITIONAL INFORMATION GL 06-001    Steam Generator Tube Integrity and        S      Initial response for Unit 2 on September 7, 2007.
Associated Technical Specifications 02    Unit 2 Action: Incorporate TSTF-449 into Technical Specifications.
Associated Technical Specifications 02    Unit 2 Action: Incorporate TSTF-449 into Technical Specifications.
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REVISION 02 UPDATE:
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.
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.
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.
TS 5.7.2.12 is the Steam Generator (SG) Program. This program is implemented to ensure that SG tube integrity is maintained.
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GL 06-002    Grid Reliability and the Impact on Plant  CI    Initial response for Unit 2 on September 7, 2007.
GL 06-002    Grid Reliability and the Impact on Plant  CI    Initial response for Unit 2 on September 7, 2007.
Risk and the Operability of Offsite Power 02    Unit 2 Action: Complete the two unit baseline electrical calculations and implementing procedures.
Risk and the Operability of Offsite Power 02    Unit 2 Action: Complete the two unit baseline electrical calculations and implementing procedures.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
NRC issued the Safety Evaluation for Generic Letter 2006-002 on January 20, 2010.
NRC issued the Safety Evaluation for Generic Letter 2006-002 on January 20, 2010.
GL 06-003    Potentially Nonconforming Hemyc and        CI    TVA does not rely on Hemyc or MT materials to protect electrical MT Fire Barrier Configurations                    and instrumentation cables or equipment that provide safe 02    shutdown capability during a postulated fire.
GL 06-003    Potentially Nonconforming Hemyc and        CI    TVA does not rely on Hemyc or MT materials to protect electrical MT Fire Barrier Configurations                    and instrumentation cables or equipment that provide safe 02    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.
Unit 2 Action: Addressed in CAP/SP. The Fire Protection Corrective Action Program will ensure Unit 2 conforms with NRC requirements and applicable guidelines.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
NRC issued the Safety Evaluation for Generic Letter 2006-003 on February 25, 2010.
NRC issued the Safety Evaluation for Generic Letter 2006-003 on February 25, 2010.
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* ITEM                        TITLE                REV                          ADDITIONAL INFORMATION GL 07-001    Inaccessible or Underground Power        CI    Initial response for Unit 2 on September 7, 2007.
 
ITEM                        TITLE                REV                          ADDITIONAL INFORMATION GL 07-001    Inaccessible or Underground Power        CI    Initial response for Unit 2 on September 7, 2007.
Cable Failures That Disable Accident Mitigation Systems or Cause Plant        02    Unit 2 Action: Complete testing of four additional cables.
Cable Failures That Disable Accident Mitigation Systems or Cause Plant        02    Unit 2 Action: Complete testing of four additional cables.
Transients
Transients REVISION 02 UPDATE:
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REVISION 02 UPDATE:
NRC issued the Safety Evaluation for Generic Letter 2007-001 on January 26, 2010.
NRC issued the Safety Evaluation for Generic Letter 2007-001 on January 26, 2010.
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REVISION 04 UPDATE:
REVISION 04 UPDATE:
NRC Inspection Report 391/2010-603 closed GL 2007-001.
NRC Inspection Report 391/2010-603 closed GL 2007-001.
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: 0737, I.A.2.3 NUREG-        Revise Scope and Criteria for Licensing  C      Closed in SSER16.
: 0737, I.A.2.3 NUREG-        Revise Scope and Criteria for Licensing  C      Closed in SSER16.
0737,        Exams I.A.3.1 Page 85 of 95                                            * = See last page for status code definition.
0737,        Exams I.A.3.1 Page 85 of 95                                            * = See last page for status code definition.
* ITEM                        TITLE                REV                      ADDITIONAL INFORMATION NUREG-        Independent Safety Engineering Group    OV    LICENSE CONDITION - Independent Safety Engineering Group 0737,                                                        (ISEG) (NUREG-0737, I.B.1.2)
 
ITEM                        TITLE                REV                      ADDITIONAL INFORMATION NUREG-        Independent Safety Engineering Group    OV    LICENSE CONDITION - Independent Safety Engineering Group 0737,                                                        (ISEG) (NUREG-0737, I.B.1.2)
I.B.1.2 Resolved for Unit 1 only in SSER8.
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.
Unit 2 action: Implement the alternate ISEG that was approved for the rest of the TVA units including WBN Unit 1 by NRC on August 26, 1999. The function will be performed by the site engineering organizations.
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: 0737, I.D.2                                                        Unit 2 Action: Install SPDS and have it operational prior to start-up after the first refueling outage.
: 0737, I.D.2                                                        Unit 2 Action: Install SPDS and have it operational prior to start-up after the first refueling outage.
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* ITEM                        TITLE            REV                          ADDITIONAL INFORMATION NUREG-        Training During Low-Power Testing    C      Closed in SSER16.
 
ITEM                        TITLE            REV                          ADDITIONAL INFORMATION NUREG-        Training During Low-Power Testing    C      Closed in SSER16.
: 0737, I.G.1 NUREG-        Reactor Coolant Vent System          CI    LICENSE CONDITION - NUREG-0737, II.B.1, "Reactor Coolant 0737,                                                    System Vents" - In the original SER, the NRC found TVAs II.B.1                                                    commitment to install reactor coolant vents acceptable pending verification. This was completed for Unit 1 only in SSER5 (IR 390/84-37).
: 0737, I.G.1 NUREG-        Reactor Coolant Vent System          CI    LICENSE CONDITION - NUREG-0737, II.B.1, "Reactor Coolant 0737,                                                    System Vents" - In the original SER, the NRC found TVAs II.B.1                                                    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.
Unit 2 Action: Verify installation of reactor coolant vents.
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NUREG-        Post-Accident Sampling              S      NRC reviewed in 9.3.2 of SSER16. TVA submitted a TS 0737,                                                    improvement to eliminate requirements for the Post Accident II.B.3                                            02    Sampling System using the Consolidated Line Item Improvement Process in a letter dated October 31, 2001.
NUREG-        Post-Accident Sampling              S      NRC reviewed in 9.3.2 of SSER16. TVA submitted a TS 0737,                                                    improvement to eliminate requirements for the Post Accident II.B.3                                            02    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.
Unit 2 Actions: Unit 2 Technical Specifications will eliminate requirements for the Post-Accident Sampling System.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.
Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.
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: 3) Modifications to pressurizer safety valves, PORVs, PORV block valves and associated piping; and 4) Change motor operated block valves.
: 3) Modifications to pressurizer safety valves, PORVs, PORV block valves and associated piping; and 4) Change motor operated block valves.
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* ITEM                      TITLE                  REV                        ADDITIONAL INFORMATION NUREG-        Valve Position Indication                  CI    The design was reviewed in the original 1982 SER and found 0737,                                                          acceptable pending confirmation of installation of the acoustic II.D.3                                                        monitoring system. In SSER5 (IR 390/84-35), the staff closed the LICENSE CONDITION for Unit 1 only.
 
ITEM                      TITLE                  REV                        ADDITIONAL INFORMATION NUREG-        Valve Position Indication                  CI    The design was reviewed in the original 1982 SER and found 0737,                                                          acceptable pending confirmation of installation of the acoustic II.D.3                                                        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.
Unit 2 Action: Verify installation of the acoustic monitoring system to PORV to indicate position.
NUREG-        Auxiliary Feedwater System Evaluation,    CI    Reviewed in Appendix EE of SSER16.
NUREG-        Auxiliary Feedwater System Evaluation,    CI    Reviewed in Appendix EE of SSER16.
0737,        Modifications II.E.1.1                                                      Unit 2 Action: Perform Auxiliary Feedwater System analysis as it pertains to system failure and flow rate.
0737,        Modifications II.E.1.1                                                      Unit 2 Action: Perform Auxiliary Feedwater System analysis as it pertains to system failure and flow rate.
NUREG-        Auxiliary Feedwater System Initiation and  CI    NRC: IR 50-390/84-20 and 50-391/84-16; letters dated 0737,        Flow                                            March 29, 1985, and October 31, 1995; SSER 16 II.E.1.2
NUREG-        Auxiliary Feedwater System Initiation and  CI    NRC: IR 50-390/84-20 and 50-391/84-16; letters dated 0737,        Flow                                            March 29, 1985, and October 31, 1995; SSER 16 II.E.1.2 Unit 2 Action: Complete procedures and qualification testing.
                                                              --------------------
NUREG-        Emergency Power For Pressurizer            CI    NRC: letters dated March 29, 1985, and October 31, 1995; 0737,        Heaters                                          SSER 16 II.E.3.1 Reviewed in original 1982 SER.
Unit 2 Action: Complete procedures and qualification testing.
NUREG-        Emergency Power For Pressurizer            CI    NRC: letters dated March 29, 1985, and October 31, 1995; 0737,        Heaters                                          SSER 16 II.E.3.1
                                                              -------------------
Reviewed in original 1982 SER.
Unit 2 Action: Implement procedures and testing.
Unit 2 Action: Implement procedures and testing.
NUREG-        Dedicated Hydrogen Penetrations            C    NRC: IR 50-390/83-27 and 50-391/83-19; SER (NUREG-0847)
NUREG-        Dedicated Hydrogen Penetrations            C    NRC: IR 50-390/83-27 and 50-391/83-19; SER (NUREG-0847)
: 0737, II.E.4.1 Page 88 of 95                                            * = See last page for status code definition.
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* ITEM                        TITLE              REV                          ADDITIONAL INFORMATION NUREG-        Containment Isolation Dependability    S      TVA: letters dated October 29, 1981, and 0737,                                                      February 25, 1985 II.E.4.2                                            02 NRC: letters dated March 29, 1985, July 12, 1990 and October 31, 1995; SSER 16.
 
                                                            --------------------------------------------------------------------------------------
ITEM                        TITLE              REV                          ADDITIONAL INFORMATION NUREG-        Containment Isolation Dependability    S      TVA: letters dated October 29, 1981, and 0737,                                                      February 25, 1985 II.E.4.2                                            02 NRC: letters dated March 29, 1985, July 12, 1990 and October 31, 1995; SSER 16.
OUTSTANDING ISSUE for NRC to complete review of information provided by TVA to address Containment Purging During Normal Plant Operation LICENSE CONDITION - Containment isolation dependability In the original 1982 SER, NRC concluded that WBN met all the requirements of NUREG-0737, item II.E.4.2 except subsection (6) concerning containment purging during normal operation. In SSER3, the outstanding issue was closed and the LICENSE CONDITION was left open.
OUTSTANDING ISSUE for NRC to complete review of information provided by TVA to address Containment Purging During Normal Plant Operation LICENSE CONDITION - Containment isolation dependability In the original 1982 SER, NRC concluded that WBN met all the requirements of NUREG-0737, item II.E.4.2 except subsection (6) concerning containment purging during normal operation. In SSER3, the outstanding issue was closed and the LICENSE CONDITION was left open.
NRC completed the review and issued a Technical Evaluation Report for both units on July 12, 1990. NRC concluded that the isolation valves can close against the buildup of pressure in the event of a design basis accident if the lower containment isolation valves are physically blocked to an opening angle of 50 degrees or less. (SSER5)
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.
Unit 2 Action: Reflect valve opening restriction in the Technical Specifications.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
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NUREG-        Accident-Monitoring Instrumentation -  CI    Reviewed in SSER9.
NUREG-        Accident-Monitoring Instrumentation -  CI    Reviewed in SSER9.
0737,        Containment High Range Monitoring II.F.1.2.C.                                                Unit 2 Actions: Install Noble gas, Iodine / particulate sampling, and Containment High Range Monitors.
0737,        Containment High Range Monitoring II.F.1.2.C.                                                Unit 2 Actions: Install Noble gas, Iodine / particulate sampling, and Containment High Range Monitors.
                                                            --------------------
Unit 2 Action: Install high range in-containment monitor for Unit 2.
Unit 2 Action: Install high range in-containment monitor for Unit 2.
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* ITEM                        TITLE                    REV                      ADDITIONAL INFORMATION NUREG-        Accident-Monitoring Instrumentation -        CI    Reviewed in SSER9.
 
ITEM                        TITLE                    REV                      ADDITIONAL INFORMATION NUREG-        Accident-Monitoring Instrumentation -        CI    Reviewed in SSER9.
0737,        Containment Pressure II.F.1.2.D.                                                      Unit 2 Action: Verify installation of containment pressure indication.
0737,        Containment Pressure II.F.1.2.D.                                                      Unit 2 Action: Verify installation of containment pressure indication.
NUREG-        Accident-Monitoring Instrumentation -        CI    Reviewed in SSER9.
NUREG-        Accident-Monitoring Instrumentation -        CI    Reviewed in SSER9.
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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.
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.
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* ITEM                      TITLE              REV                          ADDITIONAL INFORMATION NUREG-        Voiding in the Reactor Coolant System  C      LICENSE CONDITION - Voiding in the reactor coolant system 0737,                                                      (NUREG-0737, II.K.2.17)
 
ITEM                      TITLE              REV                          ADDITIONAL INFORMATION NUREG-        Voiding in the Reactor Coolant System  C      LICENSE CONDITION - Voiding in the reactor coolant system 0737,                                                      (NUREG-0737, II.K.2.17)
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.
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-        Auto PORV Isolation                    C      Reviewed in SSER5 and resolved based on NRC conclusion that 0737,                                                      there is no need for an automatic PORV isolation system (NRC II.K.3.1                                                    letter dated June 29, 1990).
NUREG-        Auto PORV Isolation                    C      Reviewed in SSER5 and resolved based on NRC conclusion that 0737,                                                      there is no need for an automatic PORV isolation system (NRC II.K.3.1                                                    letter dated June 29, 1990).
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NUREG-        Reporting SV/RV Failures/Challenges    S      (Action from GL 82-16) - NRC reviewed in Appendix EE of 0737,                                                      SSER16.
NUREG-        Reporting SV/RV Failures/Challenges    S      (Action from GL 82-16) - NRC reviewed in Appendix EE of 0737,                                                      SSER16.
II.K.3.3                                            02 Unit 2 Action: Include, as necessary, in Technical Specifications submittal.
II.K.3.3                                            02 Unit 2 Action: Include, as necessary, in Technical Specifications submittal.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.
Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.
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: 0737, II.K.3.9                                                    Unit 2 Action: Set the derivative time constant to zero.
: 0737, II.K.3.9                                                    Unit 2 Action: Set the derivative time constant to zero.
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* ITEM                        TITLE                  REV                          ADDITIONAL INFORMATION NUREG-        Anticipatory Trip at High Power            S      NRC: letter dated October 31, 1995; SSER 16
 
ITEM                        TITLE                  REV                          ADDITIONAL INFORMATION NUREG-        Anticipatory Trip at High Power            S      NRC: letter dated October 31, 1995; SSER 16
: 0737, II.K.3.10                                                02    --------------------
: 0737, II.K.3.10                                                02    --------------------
Unit 2 Action: Unit 2 Technical Specifications and surveillance procedures will address this issue.
Unit 2 Action: Unit 2 Technical Specifications and surveillance procedures will address this issue.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.
Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.
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II.K.3.25 Unit 2 Action: Ensure DG power is provided to pump sealing cooling system.
II.K.3.25 Unit 2 Action: Ensure DG power is provided to pump sealing cooling system.
NUREG-        Small Break LOCA Methods                  CI    TVA: letter dated October 29, 1981
NUREG-        Small Break LOCA Methods                  CI    TVA: letter dated October 29, 1981
: 0737, II.K.3.30                                                      NRC: letters dated March 29, 1985, and July 24, 1986; SSER 16
: 0737, II.K.3.30                                                      NRC: letters dated March 29, 1985, and July 24, 1986; SSER 16 The staff determined in SSER4 that their review of Items 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.
                                                                --------------------
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.
Unit 2 Action: Complete analysis for Unit 2.
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* ITEM                      TITLE              REV                      ADDITIONAL INFORMATION NUREG-        Plant Specific Analysis              CI    The staff determined in SSER4 that their review of Items II.K.3.30 0737,                                                    and II.K.3.31 did not have to be completed to support the II.K.3.31                                                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.
 
ITEM                      TITLE              REV                      ADDITIONAL INFORMATION NUREG-        Plant Specific Analysis              CI    The staff determined in SSER4 that their review of Items II.K.3.30 0737,                                                    and II.K.3.31 did not have to be completed to support the II.K.3.31                                                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.
Unit 2 Action: Complete analysis for Unit 2.
NUREG-        Emergency Preparedness, Short Term    C    LICENSE CONDITION - Emergency Preparedness (NUREG-0737,                                                    0737, III.A.1, III.A.2)
NUREG-        Emergency Preparedness, Short Term    C    LICENSE CONDITION - Emergency Preparedness (NUREG-0737,                                                    0737, III.A.1, III.A.2)
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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.
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.
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* ITEM                      TITLE              REV                          ADDITIONAL INFORMATION NUREG-        Primary Coolant Outside Containment  S      Resolved for Unit 1 only in SSER10; reviewed in Appendix EE of 0737,                                                      SSER16.
 
ITEM                      TITLE              REV                          ADDITIONAL INFORMATION NUREG-        Primary Coolant Outside Containment  S      Resolved for Unit 1 only in SSER10; reviewed in Appendix EE of 0737,                                                      SSER16.
III.D.1.1                                          02 Unit 2 Actions: Include the waste gas disposal system in the leakage reduction program and incorporate in Unit 2 Technical Specifications.
III.D.1.1                                          02 Unit 2 Actions: Include the waste gas disposal system in the leakage reduction program and incorporate in Unit 2 Technical Specifications.
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REVISION 02 UPDATE:
REVISION 02 UPDATE:
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.
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: 0737, III.D.3.3                                                  Unit 2 Action: Complete modifications for Unit 2.
: 0737, III.D.3.3                                                  Unit 2 Action: Complete modifications for Unit 2.
NUREG-        Control-Room Habitability            OV      TVA: letter dated October 29, 1981
NUREG-        Control-Room Habitability            OV      TVA: letter dated October 29, 1981
: 0737, III.D.3.4                                                  NRC: SSER 16
: 0737, III.D.3.4                                                  NRC: SSER 16 NRC reviewed in SER and in Appendix EE of SSER16.
                                                          --------------------
NRC reviewed in SER and in Appendix EE of SSER16.
Unit 2 Action: Complete with CRDR completion.
Unit 2 Action: Complete with CRDR completion.
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* ITEM                          TITLE                        REV                      ADDITIONAL INFORMATION STATUS CODE DEFINITIONS C: CLOSED: Previous staff review of NUREG-0847 and/or supplements has closed the item either for both units at WBN or explicitly for WBN Unit 2.
 
ITEM                          TITLE                        REV                      ADDITIONAL INFORMATION STATUS CODE DEFINITIONS C: CLOSED: Previous staff review of NUREG-0847 and/or supplements has closed the item either for both units at WBN or explicitly for WBN Unit 2.
CI: CLOSED/IMPLEMENTATION: Staff has approved either for both units at WBN or explicitly for WBN Unit 2; there is no change to the approved design; and implementation is recommended through Regional Inspection.
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.
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.
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Enclosure 4 Generic Communications - Revision 4 Changes
Enclosure 4 Generic Communications - Revision 4 Changes


GENERIC COMMUNICATIONS: REVISION 4 CHANGES
GENERIC COMMUNICATIONS: REVISION 4 CHANGES ITEM                    TITLE            REV                          ADDITIONAL INFORMATION B 83-006    Nonconforming Material Supplied by    CI    TVA: letter dated February 2, 1984 Tube-Line Facilities 04    NRC: IR 390/391 84-03; NUREG/CR 4934 NRC SER for both units dated September 23, 1991, provided an alternate acceptance for fittings supplied by Tube-Line.
* ITEM                    TITLE            REV                          ADDITIONAL INFORMATION B 83-006    Nonconforming Material Supplied by    CI    TVA: letter dated February 2, 1984 Tube-Line Facilities 04    NRC: IR 390/391 84-03; NUREG/CR 4934
                                                        --------------------
NRC SER for both units dated September 23, 1991, provided an alternate acceptance for fittings supplied by Tube-Line.
Unit 2 Action: Implement as necessary.
Unit 2 Action: Implement as necessary.
                                                        ----------------------------------------------------------------------------------------
                                                        ----------------------------------------------------------------------------------------
REVISION 04 UPDATE:
REVISION 04 UPDATE:
NRC Inspection Report Nos. 50-390/90-02 and 50-391/90-02 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 III requirements.
NRC Inspection Report Nos. 50-390/90-02 and 50-391/90-02 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 III requirements.
Line 2,899: Line 2,390:
B 96-001,  Control Rod Insertion Problems (PWR)  CI    NRC acceptance letter for Unit 1 dated July 22, 1996 - Initial first part                                              response for Unit 2 on September 7, 2007.
B 96-001,  Control Rod Insertion Problems (PWR)  CI    NRC acceptance letter for Unit 1 dated July 22, 1996 - Initial first part                                              response for Unit 2 on September 7, 2007.
04 Unit 2 Action: Issue Emergency Operating Procedure.
04 Unit 2 Action: Issue Emergency Operating Procedure.
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                                                        ----------------------------------------------------------------------------------------
REVISION 02 UPDATE:
REVISION 02 UPDATE:
Unit 2 will load all new RFA-2 fuel for the initial fuel load.
Unit 2 will load all new RFA-2 fuel for the initial fuel load.
                                                        ----------------------------------------------------------------------------------------
                                                        ----------------------------------------------------------------------------------------
REVISION 03 UPDATE:
REVISION 03 UPDATE:
NRC issued the Safety Evaluation (corrected) for Bulletin 1996-001 on May 3, 2010.
NRC issued the Safety Evaluation (corrected) for Bulletin 1996-001 on May 3, 2010.
                                                        ----------------------------------------------------------------------------------------
                                                        ----------------------------------------------------------------------------------------
REVISION 04 UPDATE:
REVISION 04 UPDATE:
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
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* ITEM                    TITLE                REV                          ADDITIONAL INFORMATION B 96-001,  Control Rod Insertion Problems (PWR)    CI    NRC acceptance letter for Unit 1 dated July 22, 1996 - Initial last part                                                  response for Unit 2 on September 7, 2007.
 
ITEM                    TITLE                REV                          ADDITIONAL INFORMATION B 96-001,  Control Rod Insertion Problems (PWR)    CI    NRC acceptance letter for Unit 1 dated July 22, 1996 - Initial last part                                                  response for Unit 2 on September 7, 2007.
04 Unit 2 Action: and provide core map.
04 Unit 2 Action: and provide core map.
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REVISION 03 UPDATE:
REVISION 03 UPDATE:
NRC issued the Safety Evaluation (corrected) for Bulletin 1996-001 on May 3, 2010.
NRC issued the Safety Evaluation (corrected) for Bulletin 1996-001 on May 3, 2010.
                                                          ----------------------------------------------------------------------------------------
                                                          ----------------------------------------------------------------------------------------
REVISION 04 UPDATE:
REVISION 04 UPDATE:
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
Line 2,925: Line 2,407:
Pressure Vessel (RPV) Head Penetration        Initial response for Unit 2 on September 7, 2007.
Pressure Vessel (RPV) Head Penetration        Initial response for Unit 2 on September 7, 2007.
Nozzles                                04 Unit 2 Action: Perform baseline inspection.
Nozzles                                04 Unit 2 Action: Perform baseline inspection.
                                                          ----------------------------------------------------------------------------------------
                                                          ----------------------------------------------------------------------------------------
REVISION 02 UPDATE:
REVISION 02 UPDATE:
Unit 2 Action: Perform baseline inspection. Evaluate or repair as necessary.
Unit 2 Action: Perform baseline inspection. Evaluate or repair as necessary.
                                                          ----------------------------------------------------------------------------------------
                                                          ----------------------------------------------------------------------------------------
REVISION 03 UPDATE:
REVISION 03 UPDATE:
NRC issued the Safety Evaluation for Bulletin 2001-001 on June 30, 2010.
NRC issued the Safety Evaluation for Bulletin 2001-001 on June 30, 2010.
                                                          ----------------------------------------------------------------------------------------
                                                          ----------------------------------------------------------------------------------------
REVISION 04 UPDATE:
REVISION 04 UPDATE:
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
Page 2 of 7                                          * = See last page for status code definition.
Page 2 of 7                                          * = See last page for status code definition.
* ITEM                    TITLE              REV                          ADDITIONAL INFORMATION B 02-001    RPV Head Degradation and Reactor      CI    NRC review of Unit 1's 15 day response in letter dated Coolant Pressure Boundary Integrity          May 20, 2002 - Initial response for Unit 2 on 04    September 7, 2007.
 
ITEM                    TITLE              REV                          ADDITIONAL INFORMATION B 02-001    RPV Head Degradation and Reactor      CI    NRC review of Unit 1's 15 day response in letter dated Coolant Pressure Boundary Integrity          May 20, 2002 - Initial response for Unit 2 on 04    September 7, 2007.
Unit 2 Action: Perform baseline inspection.
Unit 2 Action: Perform baseline inspection.
                                                        ----------------------------------------------------------------------------------------
                                                        ----------------------------------------------------------------------------------------
REVISION 02 UPDATE:
REVISION 02 UPDATE:
Unit 2 Action: Perform baseline inspection. Evaluate or repair as necessary.
Unit 2 Action: Perform baseline inspection. Evaluate or repair as necessary.
                                                        ----------------------------------------------------------------------------------------
                                                        ----------------------------------------------------------------------------------------
REVISION 03 UPDATE:
REVISION 03 UPDATE:
NRC issued the Safety Evaluation for Bulletin 2002-001 on June 30, 2010.
NRC issued the Safety Evaluation for Bulletin 2002-001 on June 30, 2010.
                                                        ----------------------------------------------------------------------------------------
                                                        ----------------------------------------------------------------------------------------
REVISION 04 UPDATE:
REVISION 04 UPDATE:
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
Line 2,955: Line 2,426:
Nozzle Inspection Programs                  Initial response for Unit 2 on September 7, 2007.
Nozzle Inspection Programs                  Initial response for Unit 2 on September 7, 2007.
04 Unit 2 Action: Perform baseline inspection.
04 Unit 2 Action: Perform baseline inspection.
                                                        ---------------------------------------------------------------------------------------
                                                        ---------------------------------------------------------------------------------------
REVISION 02 UPDATE:
REVISION 02 UPDATE:
Unit 2 Action: Perform baseline inspection. Evaluate or repair as necessary.
Unit 2 Action: Perform baseline inspection. Evaluate or repair as necessary.
                                                        ----------------------------------------------------------------------------------------
                                                        ----------------------------------------------------------------------------------------
REVISION 03 UPDATE:
REVISION 03 UPDATE:
NRC issued the Safety Evaluation for Bulletin 2002-002 on June 30, 2010.
NRC issued the Safety Evaluation for Bulletin 2002-002 on June 30, 2010.
                                                        ----------------------------------------------------------------------------------------
                                                        ----------------------------------------------------------------------------------------
REVISION 04 UPDATE:
REVISION 04 UPDATE:
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
Page 3 of 7                                        * = See last page for status code definition.
Page 3 of 7                                        * = See last page for status code definition.
* ITEM                        TITLE                REV                          ADDITIONAL INFORMATION B 04-001    Inspection of Alloy 82/182/600 Materials  CI    Initial response for Unit 2 on September 7, 2007.
 
ITEM                        TITLE                REV                          ADDITIONAL INFORMATION B 04-001    Inspection of Alloy 82/182/600 Materials  CI    Initial response for Unit 2 on September 7, 2007.
Used in the Fabrication of Pressurizer Penetrations and Steam Space Piping      04    Unit 2 Actions: Provide details of pressurizer and penetrations Connections at PWRs                              and apply Material Stress Improvement Process.
Used in the Fabrication of Pressurizer Penetrations and Steam Space Piping      04    Unit 2 Actions: Provide details of pressurizer and penetrations Connections at PWRs                              and apply Material Stress Improvement Process.
                                                            ----------------------------------------------------------------------------------------
                                                            ----------------------------------------------------------------------------------------
REVISION 02 UPDATE:
REVISION 02 UPDATE:
TVA provided details of the pressurizer and penetrations on September 29, 2008. This letter committed to:
TVA provided details of the pressurizer and penetrations on September 29, 2008. This letter committed to:
Prior to placing the pressurizer in service, TVA will apply the Material Stress Improvement Process (MSIP) to the Pressurizer Power Operated Relief Valve connections, the safety relief valve connections, the spray line nozzle and surge line nozzle connections.
Prior to placing the pressurizer in service, TVA will apply the Material Stress Improvement Process (MSIP) to the Pressurizer Power Operated Relief Valve connections, the safety relief valve connections, the spray line nozzle and surge line nozzle connections.
TVA will perform a bare metal visual (BMV) inspection of the upper pressurizer Alloy 600 locations at the first refueling outage.
TVA will perform a bare metal visual (BMV) inspection of the upper pressurizer Alloy 600 locations at the first refueling outage.
                                                            ----------------------------------------------------------------------------------------
                                                            ----------------------------------------------------------------------------------------
REVISION 03 UPDATE:
REVISION 03 UPDATE:
April 1, 2010, letter committed to:
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.
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.
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:
REVISION 04 UPDATE:
NRC issued the Safety Evaluation for Bulletin 2004-001 on August 4, 2010.
NRC issued the Safety Evaluation for Bulletin 2004-001 on August 4, 2010.
Page 4 of 7                                            * = See last page for status code definition.
Page 4 of 7                                            * = See last page for status code definition.
* ITEM                      TITLE                REV                          ADDITIONAL INFORMATION GL 88-014  Instrument Air Supply System Problems    CI    NRC letter dated July 26, 1990, closing the issue.
 
ITEM                      TITLE                REV                          ADDITIONAL INFORMATION GL 88-014  Instrument Air Supply System Problems    CI    NRC letter dated July 26, 1990, closing the issue.
Affecting Safety-Related Equipment 04    Unit 2 Action: Complete Unit 2 implementation.
Affecting Safety-Related Equipment 04    Unit 2 Action: Complete Unit 2 implementation.
                                                            ----------------------------------------------------------------------------------------
                                                            ----------------------------------------------------------------------------------------
REVISION 04 UPDATE:
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.
The compressed air system is a common system at Watts Bar; therefore, the requirements for this GL have been satisfied for Unit 2.
Line 2,997: Line 2,456:
GL 88-020  Individual Plant Examination for Severe  S      Unit 2 Action: Complete evaluation for Unit 2.
GL 88-020  Individual Plant Examination for Severe  S      Unit 2 Action: Complete evaluation for Unit 2.
Accident Vulnerabilities 04    ----------------------------------------------------------------------------------------
Accident Vulnerabilities 04    ----------------------------------------------------------------------------------------
                                                            ----------------------------------------------------------------------------------------
REVISION 02 UPDATE:
REVISION 02 UPDATE:
The Probabilistic Risk Assessment Individual Plant Examination Summary Report was submitted on February 9, 2010.
The Probabilistic Risk Assessment Individual Plant Examination Summary Report was submitted on February 9, 2010.
                                                            ----------------------------------------------------------------------------------------
                                                            ----------------------------------------------------------------------------------------
REVISION 04 UPDATE:
REVISION 04 UPDATE:
The Individual Plant Examination of External Events Design Report was submitted on April 30, 2010.
The Individual Plant Examination of External Events Design Report was submitted on April 30, 2010.
Page 5 of 7                                            * = See last page for status code definition.
Page 5 of 7                                            * = See last page for status code definition.
* ITEM                      TITLE                  REV                          ADDITIONAL INFORMATION GL 95-007  Pressure Locking and Thermal Binding of  CI    Unit 1 SER for GL 95-07 dated Sept 15, 1999 Safety-Related Power-Operated Gate Valves                                  04    Unit 2 Action: Perform evaluation for pressure locking and thermal binding of safety related power-operated gate valves and take corrective actions for those valves identified as being susceptible.
 
                                                            ----------------------------------------------------------------------------------------
ITEM                      TITLE                  REV                          ADDITIONAL INFORMATION GL 95-007  Pressure Locking and Thermal Binding of  CI    Unit 1 SER for GL 95-07 dated Sept 15, 1999 Safety-Related Power-Operated Gate Valves                                  04    Unit 2 Action: Perform evaluation for pressure locking and thermal binding of safety related power-operated gate valves and take corrective actions for those valves identified as being susceptible.
                                                            ----------------------------------------------------------------------------------------
REVISION 03 UPDATE:
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.
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:
REVISION 04 UPDATE:
NRC letter dated July 29, 2010, provided RAIs on the GL.
NRC letter dated July 29, 2010, provided RAIs on the GL.
Line 3,017: Line 2,470:
EDCRs 53292 and 53287 shall be implemented to eliminate the potential for pressure locking prior to startup.
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.
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.
NRC issued the Safety Evaluation for GL 1995-007 on August 12, 2010.
GL 97-001  Degradation of Control Rod Drive        CI    NRC acceptance letter dated November 4, 1999 (Unit 1).
GL 97-001  Degradation of Control Rod Drive        CI    NRC acceptance letter dated November 4, 1999 (Unit 1).
Mechanism Nozzle and Other Vessel Closure Head Penetrations                04    Unit 2 Action: Provide a report to address the inspection program.
Mechanism Nozzle and Other Vessel Closure Head Penetrations                04    Unit 2 Action: Provide a report to address the inspection program.
                                                            ----------------------------------------------------------------------------------------
                                                            ----------------------------------------------------------------------------------------
REVISION 03 UPDATE:
REVISION 03 UPDATE:
NRC issued the Safety Evaluation for Generic Letter 97-001 on June 30, 2010.
NRC issued the Safety Evaluation for Generic Letter 97-001 on June 30, 2010.
                                                            ----------------------------------------------------------------------------------------
                                                            ----------------------------------------------------------------------------------------
REVISION 04 UPDATE:
REVISION 04 UPDATE:
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
Corrected status from "OV" to "CI" due to NRC issuance of Safety Evaluation as noted in Revision 03 update.
Page 6 of 7                                            * = See last page for status code definition.
Page 6 of 7                                            * = See last page for status code definition.
* ITEM                          TITLE                        REV                      ADDITIONAL INFORMATION STATUS CODE DEFINITIONS C: CLOSED: Previous staff review of NUREG-0847 and/or supplements has closed the item either for both units at WBN or explicitly for WBN Unit 2.
 
ITEM                          TITLE                        REV                      ADDITIONAL INFORMATION STATUS CODE DEFINITIONS C: CLOSED: Previous staff review of NUREG-0847 and/or supplements has closed the item either for both units at WBN or explicitly for WBN Unit 2.
CI: CLOSED/IMPLEMENTATION: Staff has approved either for both units at WBN or explicitly for WBN Unit 2; there is no change to the approved design; and implementation is recommended through Regional Inspection.
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.
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.

Latest revision as of 12:31, 11 March 2020

OL - TVA Letter to NRC_10-28-10_Regulatory Framework Rev 4
ML103140374
Person / Time
Site: Watts Bar Tennessee Valley Authority icon.png
Issue date: 10/28/2010
From:
- No Known Affiliation
To:
Division of Operating Reactor Licensing
References
Download: ML103140374 (201)
v • d • e


Text

WBN2Public Resource From: Boyd, Desiree L [dlboyd@tva.gov]

Sent: Thursday, October 28, 2010 12:43 PM To: Wiebe, Joel; Poole, Justin; Raghavan, Rags; Milano, Patrick; Campbell, Stephen Cc: Crouch, William D; Hamill, Carol L; Boyd, Desiree L; Elton, Thomas L

Subject:

TVA letter to NRC_10-28-10_Regulatory Framework Rev 4 Attachments: 10-28-10_Regulatory Framework Rev 4_NRC Copy.pdf Please see attached letter that was sent to NRC today.

~*~*~*~*~*~*~*~*~*~*~*~*~*~*~

Désireé L. Boyd WBN 2 Licensing Support Sun Technical Services dlboyd@tva.gov 4233658764

~*~*~*~*~*~*~*~*~*~*~*~*~*~*~

1

Hearing Identifier: Watts_Bar_2_Operating_LA_Public Email Number: 166 Mail Envelope Properties (7AB41F650F76BD44B5BCAB7C0CCABFAF1764BAFA)

Subject:

TVA letter to NRC_10-28-10_Regulatory Framework Rev 4 Sent Date: 10/28/2010 12:42:36 PM Received Date: 10/28/2010 12:48:25 PM From: Boyd, Desiree L Created By: dlboyd@tva.gov Recipients:

"Crouch, William D" <wdcrouch@tva.gov>

Tracking Status: None "Hamill, Carol L" <clhamill@tva.gov>

Tracking Status: None "Boyd, Desiree L" <dlboyd@tva.gov>

Tracking Status: None "Elton, Thomas L" <tlelton@tva.gov>

Tracking Status: None "Wiebe, Joel" <Joel.Wiebe@nrc.gov>

Tracking Status: None "Poole, Justin" <Justin.Poole@nrc.gov>

Tracking Status: None "Raghavan, Rags" <Rags.Raghavan@nrc.gov>

Tracking Status: None "Milano, Patrick" <Patrick.Milano@nrc.gov>

Tracking Status: None "Campbell, Stephen" <Stephen.Campbell@nrc.gov>

Tracking Status: None Post Office: TVANUCXVS2.main.tva.gov Files Size Date & Time MESSAGE 304 10/28/2010 12:48:25 PM 10-28-10_Regulatory Framework Rev 4_NRC Copy.pdf 1576291 Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:

Recipients Received:

Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 October 28, 2010 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 4 (TAC No. MD6311), and Status of Generic Communications for Unit 2 -

Revision 4 (TAC No. MD8314)

Reference:

1. Letter from TVA to NRC dated July 30, 2010, Watts Bar Nuclear Plant (WBN) Unit 2 - Status of Regulatory Framework for the Completion of Construction and Licensing for Unit 2 - Revision 3 (TAC No. MD6311), and Status of Generic Communications for Unit 2 - Revision 3 (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. TVAs last revision to these two status updates, Revision 3, was submitted on July 30, 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 Revision 4.

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 Revision 4.

U.S. Nuclear Regulatory Commission Page 3 October 28, 2010

Enclosures:

1. SER and Supplements Review Matrix - Master Table
2. SER and Supplements Review Matrix - Revision 4 Changes
3. Generic Communications - Master Table
4. Generic Communications - Revision 4 Changes 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 Supplements Review Matrix - Master Table

SAFETY EVALUATION REPORT AND SUPPLEMENTS (NUREG-0847) REVIEW MATRIX:

MASTER TABLE SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 1 .0 .0 NA Overview only 1 .1 .0 NA Overview only 1 .1 .1 NA Overview only 1 .1 .2 NA Overview only 1 .1 .3 NA Overview only 1 .1 .4 NA Overview only 1 .2 .0 NA Overview only 1 .3 .0 NA Overview only 1 .3 .1 NA Overview only 1 .3 .2 NA Overview only 1 .4 .0 NA Overview only 1 .5 .0 NA Overview only 1 .6 .0 NA Overview only 1 .7 .0 NA Overview only 1 .8 .0 NA Overview only 1 .9 .0 NA Overview only 1 . 10 . 0 NA Overview only 2 .0 .0 0 C Approved for both units in SER.

2 .1 .0 0 C Approved for both units in SER.

2 .1 .1 0 C Approved for both units in SER.

2 .1 .2 0 C Approved for both units in SER.

Page 1 of 88 * = See last page for status code definition.

SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 2 .1 .3 21 S 02 SRP 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.

2 .1 .4 21 S 02 "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.

2 .2 .0 0 C Approved for both units in SER.

2 .2 .1 21 S 02 SRP 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.

Page 2 of 88 * = See last page for status code definition.

SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 2 .2 .2 21 S 02 SRP 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.

2 .2 .3 21 S 02 "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.

2 .3 .0 0 C Approved for both units in SER.

2 .3 .1 0 C Approved for both units in SER.

2 .3 .2 0 C Approved for both units in SER.

2 .3 .3 0 C See 13.3.3 (Emergency Preparedness Evaluation Conclusions).

2 .3 .4 14 C 01 TVA updated information on portions of the metrology program in FSAR amendment 83. This was reviewed and found acceptable in SSER14.

2 .3 .5 14 C 01 TVA updated information on portions of the metrology program in FSAR amendment 83. This was reviewed and found acceptable in SSER14.

2 .4 .0 0 C Approved for both units in SER.

2 .4 .1 0 C Approved for both units in SER.

2 .4 .2 0 C Approved for both units in SER.

2 .4 .3 0 O 02 REVISION 02 UPDATE:

Approved for both units in SER.

2 .4 .4 0 C Approved for both units in SER.

Page 3 of 88 * = See last page for status code definition.

SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 2 .4 .5 0 C GL 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.

2 .4 .6 0 C Approved for both units in SER.

2 .4 .7 0 C Approved for both units in SER.

2 .4 .8 21 O 02 CONFIRMATORY ISSUE for design basis groundwater level for ERCW pipeline Amendment 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).

2 .4 .9 21 O 02 SRP 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).

2 .4 . 10 21 S 02 Staff 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.

2 .4 . 11 NA Addressed in 2.4.6.

2 .4 . 12 NA Addressed in 2.4.7.

2 .4 . 13 NA Addressed in 2.4.9.

2 .4 . 14 NA Addressed in 2.4.10.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 2 .5 .0 0 C Approved for both units in SER.

2 .5 .1 0 C Approved for both units in SER.

2 .5 .2 0 C Approved for both units in SER.

2 .5 .3 0 C Approved for both units in SER.

2 .5 .4 11 C 03 CONFIRMATORY ISSUE for design differential settlement of piping and electrical components Analysis 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) analysis Staff 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.

2 .5 .5 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 2 .5 .6 0 C Approved for both units in SER.

2 .6 .0 0 C Approved for both units in SER.

3 .0 .0 0 C Approved for both units in SER.

3 .1 .0 0 C Approved for both units in SER.

3 .1 .1 0 C Approved for both units in SER.

3 .1 .2 0 C Approved for both units in SER.

3 .2 .0 14 C 01 In SSER14, the staff reviewed revisions to Table 3.2-2, "Summary of Criteria -

Mechanical System Components", and found the table acceptable.

3 .2 .1 8 C 01 CONFIRMATORY ISSUE for seismic classification of structures, systems, and components important to safety The 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 1 Staff 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 conduits In 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 3 .2 .2 21 CI 02 Section 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.

3 .3 .0 0 C Approved for both units in SER.

3 .3 .1 0 C Approved for both units in SER.

3 .3 .2 0 C Approved for both units in SER.

3 .4 .0 0 C Approved for both units in SER.

3 .4 .1 0 C Approved for both units in SER.

3 .4 .2 NA Addressed in 3.4.1.

3 .5 .0 0 C Approved for both units in SER.

3 .5 .1 14 C 01 In 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.

3 .5 .2 2 C CONFIRMATORY ISSUE for modifications to protect Diesel Generators TVA 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 3 .5 .3 0 C 01 Approved for both units in SER.

3 .6 .0 21 CI 02 In SSER6, the staff accepted TVA approaches 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.

3 .6 .1 21 O 02 OUTSTANDING ISSUE involving main steam line break (MSLB) outside containment In 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).

3 .6 .2 14 C 01 The 3.6.2 discussion in SSER14 on response spectra for the SCV refers to the evaluation provided in 3.6.1.

3 .6 .3 12 C 01 New 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 3 .7 .0 21 C 03 The 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.

3 .7 .1 21 C 03 OUTSTANDING ISSUE involving update of FSAR for seismic design issues The 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 3 .7 .2 21 C 03 3.7.2.1.2: OUTSTANDING ISSUE involving mass eccentricity In 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 response The 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 3 .7 .3 21 C 03 OUTSTANDING ISSUE involving number of peak cycles to be used for OBE In 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 case In 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).

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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 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.

3 .7 .4 0 C Approved for both units in SER.

3 .8 .0 21 O 02 OUTSTANDING ISSUE involving load combinations and stress allowables In 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).

3 .8 .1 3 C 01 CONFIRMATORY ISSUE - verify buckling methodology In 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.

3 .8 .2 7 C 01 The 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 3 .8 .3 21 O 02 The 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).

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 3 .8 .4 0 C Approved for both units in SER.

3 .9 .0 0 C Approved for both units in SER.

3 .9 .1 13 OV 01 OUTSTANDING ISSUE involving assumption in piping analysis for water-hammer due to check valve slam In SSER6, the NRC expressed concern regarding TVAs 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.

3 .9 .2 14 C 01 The staff reviewed "Pre-operational Vibration and Dynamic Effects Testing on Piping", and found this area acceptable in SSER14.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 3 .9 .3 15 S 02 3.9.3.1: OUTSTANDING ISSUE involving use of experience data to qualify category I(L) piping The 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 valves The 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 loads The 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 supports The 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 3.9.3.4: OUTSTANDING ISSUE, staff was awaiting TVA concurrence on their position with respect to margin for critical buckling of pipe supports In 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.

3 .9 .4 0 C Approved for both units in SER.

3 .9 .5 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 3 .9 .6 20 S 02 LICENSE CONDITION on inservice testing of pumps and valves The 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.

3 .9 .7 NA Area not addressed in 1981 Standard Review Plan.

3 .9 .8 NA Area not addressed in 1981 Standard Review Plan.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 3 . 10 . 0 21 CI 02 In 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).

TVA's September 26, 2008, letter proposed the use of the Unit 1 approach to resolve the Equipment Seismic Qualification CAP .

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION In SSER21, the Equipment Seismic Qualification CAP was resolved.

Completion of the Equipment Seismic Qualification CAP is tracked under 23.2.6.

3 . 11 . 0 21 CI 02 OUTSTANDING ISSUE - TVA program not submitted at time of SER The 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.

3 . 12 . 0 NA Addressed in 3.9.1 through 3.9.3.

3 . 12 . 1 NA Addressed in 3.9.1 through 3.9.3.

3 . 12 . 2 NA Addressed in 3.9.1 through 3.9.3.

3 . 12 . 3 NA Addressed in 3.9.1 through 3.9.3.

3 . 12 . 4 NA Addressed in 3.9.1 through 3.9.3.

3 . 12 . 5 NA Addressed in 3.9.1 through 3.9.3.

3 . 12 . 6 NA Addressed in 3.9.1 through 3.9.3.

3 . 13 . 0 NA Area not addressed in 1981 Standard Review Plan.

4 .0 .0 0 C Approved for both units in SER.

4 .1 .0 0 C Approved for both units in SER.

4 .2 .0 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 4 .2 .1 13 S 02 In 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) 4 .2 .2 2 S 02 CONFIRMATORY ISSUE on cladding collapse calculations The 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.

4 .2 .3 13 S 02 CONFIRMATORY 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.

4 .2 .4 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 4 .2 .5 0 S 02 "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.

4 .3 .0 0 C Approved for both units in SER.

4 .3 .1 13 S 02 In 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.

4 .3 .2 15 S 02 In 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 4 .3 .3 13 S 02 In 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.

4 .3 .4 13 S 02 In 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.

4 .4 .0 0 C Approved for both units in SER.

4 .4 .1 0 C Approved for both units in SER.

4 .4 .2 12 S 02 In 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 4 .4 .3 16 S 02 OUTSTANDING ISSUE concerning removal of RTD bypass system This 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.

4 .4 .4 13 S 02 In 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 4 .4 .5 16 O 01 CONFIRMATORY ISSUE / LICENSE CONDITION on review of Loose Parts Monitoring System (LPMS) startup report and inclusion of limiting conditions for LPMS in Technical Specifications TVA 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.

4 .4 .6 0 C Approved for both units in SER.

4 .4 .7 0 S 02 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.

4 .4 .8 10 O 01 LICENSE 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.

4 .4 .9 0 O 01 "CONCLUSION" left open until all items in subsection are closed.

4 .5 .0 0 C Approved for both units in SER.

4 .5 .1 0 C Approved for both units in SER.

4 .5 .2 0 C Approved for both units in SER.

4 .6 .0 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 5 .0 .0 0 C Approved for both units in SER.

5 .1 .0 6 S 02 The 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.

5 .2 .0 0 C Approved for both units in SER.

5 .2 .1 0 C Approved for both units in SER.

5 .2 .2 15 C 01 OUTSTANDING ISSUE on staff review of sensitivity study of required safety valve flow rate versus trip parameter TVA 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.

5 .2 .3 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 5 .2 .4 16 O 03 LICENSE CONDITION - Inservice inspection (ISI) program The 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).

5 .2 .5 21 O 02 In 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).

5 .2 .6 16 C 01 In SSER16, the staff reviewed the analysis of the RPV and internal components and found the use of the WECAN computer code acceptable.

5 .3 .0 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 5 .3 .1 21 O 02 The 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).

5 .3 .2 16 S 02 OUTSTANDING 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.

5 .3 .3 0 S 02 OUTSTANDING ISSUE for staff to complete evaluation of Unit 2 after receipt of P-T limits 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.

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.

5 .4 .0 0 C Approved for both units in SER.

5 .4 .1 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 5 .4 .2 4 C 01 5.4.2.2: OUTSTANDING ISSUE for staff to evaluate TVAs proposed resolution to concerns about flow induced vibrations in Model D-3 SGs pre-heat region In 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. TVAs 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.

5 .4 .3 21 O 02 CONFIRMATORY ISSUE to verify installation of an RHR flow alarm and proper function of dump valves when actuated manually In the SER, staff accepted TVAs 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 refueling Branch 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).

5 .4 .4 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 5 .4 .5 21 CI 02 LICENSE CONDITION - NUREG-0737, II.B.1, "Reactor Coolant System Vents" - In the original SER, the NRC found TVAs 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).

6 .0 .0 0 C Approved for both units in SER.

6 .1 .0 0 C Approved for both units in SER.

6 .1 .1 0 C Approved for both units in SER.

6 .1 .2 0 C Approved for both units in SER.

6 .1 .3 0 C Approved for both units in SER.

6 .2 .0 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 6 .2 .1 21 S 02 6.2.1.1: CONFIRMATORY ISSUE involves reviewing analysis that ensures that containment external pressure will not exceed design value of 2.0 psi In 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 6 .2 .2 21 CI 02 In 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.

6 .2 .3 16 C 01 In 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 6 .2 .4 21 S 02 CONFIRMATORY 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 Operation LICENSE CONDITION - Containment isolation dependability In the original 1982 SER, NRC concluded that WBN met all the requirements of NUREG-0737, item II.E.4.2 except subsection (6) concerning containment purging during normal operation. In SSER3, the outstanding issue was closed and the LICENSE CONDITION was left open. NRC completed the review and issued a 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."

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 6 .2 .5 21 S 04 OUTSTANDING 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 LOCA In 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 instrumentation II.F.1 - containment hydrogen In 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 6 .2 .6 21 O 02 In 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).

6 .2 .7 4 C CONFIRMATORY 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 staffs analysis of fracture toughness requirements for load bearing component of the containment system In 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.

6 .3 .0 0 C 01 Approved for both units in SER.

6 .3 .1 11 S 02 OUTSTANDING ISSUE - involving removal of upper head injection system The 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 6 .3 .2 5 S 02 In 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.

6 .3 .3 9 S 02 OUTSTANDING ISSUE - involving containment sump screen design In 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-LOCA In 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 staffs 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.

6 .3 .4 0 C Approved for both units in SER.

6 .3 .5 0 O 01 Closure based on 6.3.1 to 6.3.3.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 6 .4 .0 21 O 02 In 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).

6 .5 .0 0 C Approved for both units in SER.

6 .5 .1 21 O 02 In SSER5, the staff found the Reactor Building Purge Ventilation System acceptable.

REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

6 .5 .2 0 C Approved for both units in SER.

6 .5 .3 0 C Approved for both units in SER.

6 .5 .4 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 6 .6 .0 15 O 03 OUTSTANDING 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 justification NRC 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).

7 .0 .0 0 C Approved for both units in SER.

7 .1 .0 0 C Approved for both units in SER.

7 .1 .1 16 S 02 In 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.

7 .1 .2 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 7 .1 .3 15 S 02 In 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.

7 .2 .0 0 C Approved for both units in SER.

7 .2 .1 15 S 02 In 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.

7 .2 .2 0 C Approved for both units in SER.

7 .2 .3 0 C Approved for both units in SER.

7 .2 .4 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 7 .2 .5 21 O 02 CONFIRMATORY ISSUE - address IEB 79-21 to alleviate temperature dependence problem associated with measuring SG water level In 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).

7 .2 .6 13 S 02 In 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.

7 .3 .0 13 S 02 In 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 7 .3 .1 14 S 02 In 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 section 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.

7 .3 .2 2 C CONFIRMATORY ISSUE is commitment to make a design change to provide protection that prevents debris from entering containment sump level sensors In 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.

7 .3 .3 0 C Approved for both units in SER.

7 .3 .4 0 C Approved for both units in SER.

7 .3 .5 21 CI 02 CONFIRMATORY 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 reset In 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).

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 7 .3 .6 13 S 02 In 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.

7 .4 .0 0 C Approved for both units in SER.

7 .4 .1 0 C Approved for both units in SER.

7 .4 .2 21 O 02 By 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).

7 .4 .3 0 C Approved for both units in SER.

7 .5 .0 0 C Approved for both units in SER.

7 .5 .1 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 7 .5 .2 21 O 02 OUTSTANDING ISSUE involving RG 1.97 instruments following course of an accident In 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).

7 .5 .3 21 CI 02 B 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).

7 .5 .4 21 CI 02 "CONCLUSIONS" left CI until all items in subsection are closed.

REVISION 02 UPDATE:

The status in SSER21 is Open (Inspection).

7 .6 .0 0 C Approved for both units in SER.

7 .6 .1 0 C Approved for both units in SER.

7 .6 .2 0 C Approved for both units in SER.

7 .6 .3 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 7 .6 .4 0 C Approved for both units in SER.

7 .6 .5 4 C CONFIRMATORY 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.

7 .6 .6 0 C Approved for both units in SER.

7 .6 .7 0 C Approved for both units in SER.

7 .6 .8 0 C Approved for both units in SER.

7 .6 .9 4 C Approved for both units SER subject to completion of Confirmatory Issue in 7.6.5.

7 .7 .0 0 C Approved for both units in SER.

7 .7 .1 0 C Approved for both units in SER.

7 .7 .2 13 C 01 LICENSE 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.

7 .7 .3 0 C Approved for both units in SER.

7 .7 .4 0 C Approved for both units in SER.

7 .7 .5 0 C Approved for both units in SER.

7 .7 .6 0 C Approved for both units in SER.

7 .7 .7 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 7 .7 .8 21 S 02 ATWS 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).

7 .8 .0 0 C Approved for both units in SER.

7 .8 .1 21 O 02 NUREG-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).

7 .8 .2 21 CI 02 NUREG-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).

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 7 .8 .3 21 O 02 NUREG-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).

7 .8 .4 21 S 02 NUREG-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.

7 .8 .5 0 C 01 NUREG-0737, II.K.3.12, "Confirm Existence of Anticipatory Reactor Trip Upon Turbine Trip" Approved for both units in the SER 7 .9 .0 NA Area not addressed in 1981 Standard Review Plan.

8 .0 .0 0 C Approved for both units in SER.

8 .1 .0 0 C Approved for both units in SER.

8 .2 .0 0 C Approved for both units in SER.

8 .2 .1 13 C 01 Approved 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 8 .2 .2 21 O 02 8.2.2.1 CONFIRMATORY ISSUE - document additional information in FSAR on control power supplies and distribution system for the Watts Bar Hydro Plant Switchyard In 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 Contingencies 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 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 Trip 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 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).

8 .2 .3 0 C Approved for both units in SER.

8 .2 .4 0 C Approved for both units in SER.

8 .3 .0 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 8 .3 .1 20 S 02 8.3 Fifth Diesel Generator In 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 FSAR In 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 report In 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 temperature In the original 1982 SER, NRC required that the capability of the DGs to start at normal temperature be demonstrated. TVAs 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 By letter dated June 20, 1991, for both units, NRC requested additional Page 46 of 88 * = See last page for status code definition.

SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 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 Testing 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 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.

REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 8 .3 .2 14 C 01 8.3.2.2: LICENSE CONDITION - DC monitoring and annunciation system In 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 FSAR In 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 Inverters 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.

8.3.2.5.1 Transfer of Loads Between Power Supplies Associated with the Same Load Group but Different Units 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 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 Power 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 8 .3 .3 21 S 02 8.3.3.1.1: CONFIRMATORY ISSUE involving submergence of electrical equipment as result of a LOCA In 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 Submergence 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 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 device In 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 Runs 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 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 Conductors In 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 8.3.3.2: CONFIRMATORY ISSUE - revise FSAR to reflect requirements of shared safety systems In 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 2 In 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 units In 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 switchgear In 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.

8.3.3.3 Physical Independence (Compliance with GDC 17)

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 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 time In 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 Testing In 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 device In the original 1982 SER, staff required a reevaluation of the penetrations 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 breakers Page 52 of 88 * = See last page for status code definition.

SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION In 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.

8 .4 .0 CI Station Blackout (SBO) - SE for both units - March 18, 1993; SSE for both units - September 9, 1993.

Unit 2 Action: Implement SBO requirements.

8 .5 .0 NA Area not addressed in 1981 Standard Review Plan.

8 .5 .1 NA Area not addressed in 1981 Standard Review Plan.

9 .0 .0 10 C 01 In 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.

9 .1 .0 5 C 01 In 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.

9 .1 .1 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 9 .1 .2 21 O 02 In 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).

9 .1 .3 21 O 02 In 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).

9 .1 .4 13 OV 01 LICENSE 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 TVAs response to NUREG-0612 guidelines for Phase I in TVA letter dated July 28, 1993.

Unit 2 Action: Implement NEI guidance on heavy loads.

9 .1 .5 NA Addressed in 9.1.4.

9 .2 .0 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 9 .2 .1 18 O 01 In 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.

9 .2 .2 5 CI 01 CONFIRMATORY ISSUE - relocate component cooling thermal barrier booster pumps above probable maximum flood (PMF) level before receipt of an OL TVA 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.

9 .2 .3 0 C Approved for both units in SER.

9 .2 .4 9 C 01 In 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.

9 .2 .5 0 C Approved for both units in SER.

9 .2 .6 12 C 01 In 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.

9 .3 .0 0 C Approved for both units in SER.

9 .3 .1 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 9 .3 .2 21 S 02 LICENSE CONDITION - Post-Accident Sampling System In 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.

9 .3 .3 0 C Approved for both units in SER.

9 .3 .4 0 C Approved for both units in SER.

9 .4 .0 0 C Approved for both units in SER.

9 .4 .1 9 C 01 In 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.

9 .4 .2 0 C Approved for both units in SER.

9 .4 .3 0 C Approved for both units in SER.

9 .4 .4 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 9 .4 .5 21 O 02 In 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).

9 .5 .0 10 C 01 In 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 9 .5 .1 19 C 01 9.5.1.2: OUTSTANDING ISSUE for Fire Protection Program 9.5.1.3: CONFIRMATORY ISSUE - Electrical penetrations documentation 9.5.1.3: 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 TVAs 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 TVAs revised position on emergency lighting, which was found to be acceptable.

9 .5 .2 21 O 02 LICENSE CONDITION - Performance testing of communications system The staff resolved this license condition in SSER5 based on TVAs 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).

9 .5 .3 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 9 .5 .4 12 C 01 9.5.4.1: 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 load In 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 with NUREG/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 3 The 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 lines The staff found TVAs 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.

In SSER11, the staff noted the revised capacity of the 7-day fuel oil storage Page 59 of 88 * = See last page for status code definition.

SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 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).

9 .5 .5 11 C 01 OUTSTANDING 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 3 The 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 9 .5 .6 21 O 02 OUTSTANDING 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 3 The 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).

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 9 .5 .7 21 O 02 OUTSTANDING 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 features In 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 3 The 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).

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 9 .5 .8 21 O 02 OUTSTANDING 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 1.26 The 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).

10 . 0 .0 0 C Approved for both units in SER.

10 . 1 .0 0 C Approved for both units in SER.

10 . 2 .0 21 O 02 In 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).

10 . 2 .1 12 C 01 In 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.

10 . 2 .2 0 C Approved for both units in SER.

10 . 3 .0 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 10 . 3 .1 21 O 02 In 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).

10 . 3 .2 0 C Approved for both units in SER.

10 . 3 .3 0 C Approved for both units in SER.

10 . 3 .4 5 S 02 LICENSE CONDITION - Secondary water chemistry monitoring and control program The 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.

10 . 4 .0 0 C Approved for both units in SER.

10 . 4 .1 9 C 01 In SSER9, the staff clarified the description of the main condenser and stated that this clarification did not affect the conclusion reached in the SER.

10 . 4 .2 0 C Approved for both units in SER.

10 . 4 .3 0 C Approved for both units in SER.

10 . 4 .4 21 O 02 In 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).

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 10 . 4 .5 0 C Approved for both units in SER.

10 . 4 .6 0 C Approved for both units in SER.

10 . 4 .7 14 C 01 In 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.

10 . 4 .8 0 C Approved for both units in SER.

10 . 4 .9 14 C 01 In 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.

11 . 0 .0 0 C Approved for both units in SER.

11 . 1 .0 16 OV 01 This item remains open pending closure of 11.4.0 and 11.5.0 11 . 2 .0 16 C 01 In 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.

11 . 3 .0 16 C 01 In 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 TVAs February 17, 1995, letter (submitted on both dockets), the staff accepted the WBNs 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 11 . 4 .0 16 OV 01 On 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.

11 . 5 .0 20 OV 01 In 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 TVAs 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 11 . 6 .0 21 O 02 In 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).

11 . 7 .0 0 OT 01 This item will remain open pending resolution of Item 11.7.2.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 11 . 7 .1 21 CI 02 LICENSE CONDITION (6a) - Accident monitoring instrumentation II.F.1 -

Noble Gas monitor In 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 staffs 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 staffs concern and eliminated the need for License Condition 6a.

LICENSE CONDITION (6b) - Accident monitoring instrumentation II.F.1 - Iodine particulate sampling See 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 of high-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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION REVISION 02 UPDATE:

The status in SSER21 is Open (Inspection).

11 . 7 .2 16 S 02 NUREG-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 TVAs 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.

12 . 0 .0 14 C Approved for both units in SER.

12 . 1 .0 21 O 02 In 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).

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 12 . 2 .0 21 O 02 In 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).

12 . 3 .0 21 O 02 In 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).

12 . 4 .0 21 O 02 In SSER10, the staff reviewed revised operational test frequency of area radiation monitors from monthly to quarterly and found that TVAs 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).

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 12 . 5 .0 21 O 02 In 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).

12 . 6 .0 21 O 02 OUTSTANDING ISSUE involving Health Physics Program The staff reviewed TVAs 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 staffs 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).

12 . 7 .0 0 C Approved for both units in SER.

12 . 7 .1 21 O 02 NUREG-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).

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 12 . 7 .2 21 O 02 NUREG-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 TVAs 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).

12 . 7 .3 21 O 02 NUREG-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).

13 . 0 .0 0 C Approved for both units in SER.

13 . 1 .0 16 C 01 In 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.

13 . 1 .1 0 C Approved for both units in SER.

13 . 1 .2 0 C Approved for both units in SER.

13 . 1 .3 8 O 01 LICENSE CONDITION - Use of experienced personnel during startup In 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 TVAs 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.

13 . 2 .0 0 C Approved for both units in SER.

13 . 2 .1 10 C 01 In 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 13 . 2 .2 0 C Approved for both units in SER.

13 . 3 .0 13 O 01 In 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.

13 . 3 .1 20 O 01 In 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.

13 . 3 .2 20 O 01 In 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.

13 . 3 .3 20 O 01 LICENSE 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.

13 . 4 .0 8 OV 01 LICENSE 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.

13 . 5 .0 0 C Approved for both units in SER.

13 . 5 .1 21 O 02 Approved for both units in SER.

REVISION 02 UPDATE:

The status in SSER21 is Open (NRR).

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 13 . 5 .2 21 CI 02 OUTSTANDING ISSUE involving operating, maintenance and emergency procedures In the original 1982 SER, this issue was used to track the staffs 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).

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SER SECTION SSER #

In the original 1982 SER, the NRC accepted TVAs 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 13 . 6 .0 21 O 02 OUTSTANDING ISSUE to file appropriate revision to the Physical Security Plan In the original 1982 SER, the staff identified certain outstanding issues with TVAs 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).

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 14 . 0 .0 21 S 02 LICENSE CONDITION - Report changes to Initial Test Program In 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 TVAs 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 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 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 21, 2010.

Table 14.2-1 was revised to clarify the testing requirement.

15 . 0 .0 0 C Approved for both units in SER.

15 . 0 .1 NA Area not addressed in 1981 Standard Review Plan.

15 . 0 .2 NA Area not addressed in 1981 Standard Review Plan.

15 . 1 .0 0 C Approved for both units in SER.

15 . 1 .1 NA Addressed in 15.2.1 15 . 1 .2 NA Addressed in 15.2.1 15 . 1 .3 NA Addressed in 15.2.1 15 . 1 .4 NA Addressed in 15.2.1 15 . 1 .5 NA Addressed in 15.2.1 and 15.4.2.

15 . 2 .0 0 S 02 Approved 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 15 . 2 .1 14 S 02 In 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.

15 . 2 .2 0 C Approved for both units in SER.

15 . 2 .3 18 S 02 In 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 15 . 2 .4 14 S 02 15.2.4.1 Uncontrolled Rod Cluster Assembly Bank Withdrawal from Zero-Power Condition In 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 criteria In 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 Ejection In 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.

15 . 2 .5 4 C Approved for both units in SER subject to completion of Outstanding Issue in 15.2.4.4.

15 . 2 .6 NA Addressed in 15.2.1.

15 . 2 .7 NA Addressed in 15.2.1.

15 . 3 .0 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 15 . 3 .1 15 S 02 In 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.

15 . 3 .2 14 S 02 In 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..

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 15 . 3 .3 14 S 02 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.

15 . 3 .4 14 S 02 In 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.

15 . 3 .5 14 S 02 In 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.

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SER SECTION SSER #

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 TVAs 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).

15 . 3 .7 0 C Approved for both units in SER.

15 . 4 .0 0 C Approved for both units in SER.

15 . 4 .1 18 S 02 In 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 15 . 4 .2 15 S 02 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.

15 . 4 .3 15 S 02 LICENSE CONDITION - Steam Generator tube rupture In 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 involving

1) 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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 15 . 4 .4 15 S 02 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.

15 . 4 .5 15 S 02 In 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.

15 . 4 .6 0 S 02 Approved 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.

15 . 4 .7 0 S 02 Approved 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.

15 . 5 .0 0 C Approved for both units in SER.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 15 . 5 .1 4 C LICENSE 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.

15 . 5 .2 4 C LICENSE 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.

15 . 5 .3 5 C LICENSE 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.

15 . 5 .4 21 CI 02 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).

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 15 . 5 .5 21 S 02 NUREG-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.

15 . 6 .0 0 C Approved for both units in SER.

15 . 6 .1 0 C Approved for both units in SER.

16 . 0 .0 S 02 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.

16 . 1 .0 NA Area not addressed in 1981 Standard Review Plan.

17 . 0 .0 0 C Approved for both units in SER.

17 . 1 .0 0 C 01 Approved for both units in SER. See 17.3.

17 . 2 .0 0 C 01 Approved for both units in SER. See 17.3.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 17 . 3 .0 15 C 01 OUTSTANDING ISSUE - QA program The 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.

17 . 4 .0 0 C 01 Approved for both units in SER. See 17.3.

17 . 5 .0 NA Area not addressed in 1981 Standard Review Plan.

17 . 6 .0 OV 10 CFR 50.65- Maintenance Rule Unit 2 Action: Implement Maintenance Rule for Unit 2 systems 1 month prior to fuel load 18 . 0 .0 0 NA See 18.1.

18 . 1 .0 21 CI 02 NUREG-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.

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SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 18 . 2 .0 21 CI 02 "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.

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

OV: 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.

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Enclosure 2 SER and Supplements Review Matrix - Revision 4 Changes

SAFETY EVALUATION REPORT AND SUPPLEMENTS (NUREG-0847) REVIEW MATRIX:

REVISION 4 CHANGES SER SECTION SSER #

  • REV. ADDITIONAL INFORMATION 6 .2 .5 21 S 04 OUTSTANDING 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 LOCA In 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 instrumentation II.F.1 - containment hydrogen In 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.

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SER SECTION SSER #

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

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

CT: CLOSED/TECHNICAL SPECIFICATIONS: Item has been approved either for both units at WBN or explicitly for WBN Unit 2; however, a change to the original approval requires submittal of the Technical Specifications and staff review.

NA: NOT APPLICABLE: Justification as to why a section / subsection is not 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: OPEN/TECHNICAL SPECIFICATIONS: No action or documentation is provided that shows the staff has reviewed the item for WBN Unit 2, and the resolution is through submittal of a Technical Specification.

OV: 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.

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Enclosure 3 Generic Communications - Master Table

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

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

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

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

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

B 73-001 Faulty Overcurrent Trip Delay Device in C TVA: letter dated April 4, 1973 Circuit Breakers for Engineered Safety Systems NRC: IR 390/391 75-5 B 73-002 Malfunction of Containment Purge Supply C TVA: letter dated August 22, 1973 Valve Switch NRC: IR 390/391 75-5 B 73-003 Defective Hydraulic Snubbers and C TVA: letter dated February 7, 1985 Restraints NRC: IR 390/391 85-08 B 73-004 Defective Bergen-Patterson Hydraulic C TVA: memo dated February 7, 1985 Shock Absorbers NRC: IR 390/391 85-08 B 73-005 Manufacturing Defect in BWR Control NA Boiling Water Reactor Rods B 73-006 Inadvertent Criticality in a BWR NA Boiling Water Reactor B 74-001 Valve Deficiencies C TVA: letter dated April 15, 1974 NRC: IR 390/391 75-5 B 74-002 Truck Strike Possibility NA Info Page 1 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 74-003 Failure of Structural or Seismic Support CI TVA: memo dated January 22, 1985 Bolts on Class I Components NRC: IR 390/391 85-08 Approach accepted in IR 50-390/85-08 and 50-391/85-08 (March 29, 1985).

Unit 2 Action: Implement per NUREG-0577 as was done for Unit 1.

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

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

B 74-008 Deficiency in the ITE Molded Case Circuit C TVA: letter dated August 21, 1974 Breakers, Type HE-3 NRC: IR 390/391 75-5 B 74-009 Deficiency in GE Model 4KV Magne-Blast C TVA: letter dated September 20, 1974 Circuit Breakers NRC: IR 390/391 76-6 B 74-010 Failures in 4-Inch Bypass Pipe at NA Boiling Water Reactor Dresden 2 B 74-011 Improper Wiring of Safety Injection Logic C NRC: IR 390/391 75-6 at Zion 1 & 2 B 74-012 Incorrect Coils in Westinghouse Type SG C NRC: IR 390/391 75-5 Relays at Trojan B 74-013 Improper Factory Wiring on GE Motor C TVA: letter dated December 24, 1974 Control Centers at Fort Calhoun NRC: IR 390/391 75-5 B 74-014 BWR Relief Valve Discharge to NA Boiling Water Reactor Suppression Pool Page 2 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 74-015 Misapplication of Cutler-Hammer Three CI TVA: letter dated May 5, 1975 Position Maintained Switch Model No.

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

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

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

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

Unit 2 Action: Modify valves not modified at factory.

B 75-004 Cable Fire at BFNPP CI NRC: IR 390/391 85-08 Closed to Fire Protection CAP Part of Fire Protection CAP B 75-005 Operability of Category I Hydraulic Shock CI TVA: letter dated June 16, 1975 and Sway Suppressors NRC: IR 390/391 75-6 NRC accepted in IR 50-390/75-6 and 50-391/75-6 (August 21, 1975).

Unit 2 Action: Install proper suppressors.

B 75-006 Defective Westinghouse Type OT-2 CI TVA: letter dated July 31, 1975 Control Switches NRC: IR 390/85-25 and 391/85-20 Unit 2 Action: Inspect Westinghouse Type OT-2 control switches.

[WAS "NOTE 3."]

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ITEM TITLE REV ADDITIONAL INFORMATION B 75-007 Exothermic Reaction in Radwaste NA Does not apply to power reactor.

Shipment B 75-008 PWR Pressure Instrumentation S NRC: IR 390/391 85-08 02 Unit 2 Action: Ensure that Technical Specifications and Site Operating Instructions address importance of maintaining temperature and pressure within prescribed limits.

REVISION 02 UPDATE:

Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.

Adherence to Pressure and Temperature limits is required by the following portions of the Unit 2 TS: 1.1 [definition of PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)]; 3.4.3 [RCS Pressure and Temperature (P/T) Limits]; 3.4.12 [Cold Overpressure Mitigation System (COMS)]; and 5.9.6 [Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)].

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

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

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ITEM TITLE REV ADDITIONAL INFORMATION B 77-001 Pneumatic Time Delay Relay Setpoint C TVA: letter dated July 1, 1977 Drift NRC: IR 390/391 85-08 B 77-002 Potential Failure Mechanism in Certain C TVA: letter dated November 11, 1977 Westinghouse AR Relays with Latch Attachments NRC: IR 390/391 85-08 B 77-003 On-Line Testing of the Westinghouse CI Unit 2 Action: Include necessary periodic testing in test Solid State Protection System procedures.

B 77-004 Calculation Error Affecting The Design S TVA: letter dated January 23, 1978 Performance of a System for Controlling pH of Containment Sump Water 02 NRC: IR 390/78-11 and 391/78-09 Following a LOCA ------------------------

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

REVISION 02 UPDATE:

Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.

TS Surveillance Requirement 3.6.11.5 requires verification that the boron concentration is within a specified range.

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

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

TVA: letter dated March 1, 1978 NRC: IR 390/78-11 and 391/78-09 B 78-001 Flammable Contact - Arm Retainers in C TVA: letter dated March 20, 1978 GE CR120A Relays NRC: IR 390/78-11 and 391/78-09 Page 5 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 78-002 Terminal Block Qualification C TVA: letter dated March 1, 1978 NRC: IR 390/78-11 and 391/78-09 B 78-003 Potential Explosive Gas Mixture NA Boiling Water Reactor Accumulations Associated with BWR Offgas System Operations B 78-004 Environmental Qualification of Certain CI TVA: letter dated December 19, 1978 Stem Mounted Limit Switches Inside Reactor Containment NRC: IR 390/82-13 and 391/82-10 Closed to EQ Program IR 50-390/82-13 and 50-391/82-10 (April 22, 1982) accepted approach.

Unit 2 Action: Ensure NAMCO switches have been replaced.

B 78-005 Malfunctioning of Circuit Breaker Auxiliary C TVA: letter dated June 12, 1978 Contact Mechanism - GE Model CR105X NRC: IR 390/78-17 and 391/78-15 B 78-006 Defective Cutler-Hammer Type M Relays C NRC: IR 390/78-22 and 391/78-19 With DC Coils B 78-007 Protection Afforded by Air-Line NA Item was applicable only to units with operating license at the Respirators and Supplied-Air Hoods time the item was issued.

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

NRC: IR 390/391 85-08 B 78-009 BWR Drywell Leakage Paths Associated NA Boiling Water Reactor with Inadequate Drywell Closures B 78-010 Bergen-Patterson Hydraulic Shock C TVA: letter dated August 14, 1978 Suppressor Accumulator Spring Coils NRC: IR 390/78-22 and 391/78-19 B 78-011 Examination of Mark I Containment Torus NA Boiling Water Reactor Welds B 78-012 Atypical Weld Material in Reactor C TVA: Westinghouse letter dated October 29, 1979 Pressure Vessel Welds NRC: IR 390/391 81-04 Page 6 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 78-013 Failures in Source Heads Kay Ray, Inc. NA Does not apply to power reactor.

Gauges Models 7050, 7050B, 7051, 7051B, 7060, 7060B, 7061 and 7061B B 78-014 Deterioration of Buna-N Components in NA Boiling Water Reactor ASCO Solenoids B 79-001 Environmental Qualification of Class 1E C NRC: IR 390/80-06 and 391/80-05 Equipment B 79-002 Pipe Support Base Plate Designs Using CI NRC review of HAAUP Program in NUREG-1232, SSER6, and Concrete Expansion Anchor Bolts SSER8.

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

B 79-003 Longitudinal Weld Defects in ASME C TVA: letter dated July 16, 1981 SA-312 Type 304 SS Pipe Spools Manufactured by Youngstown Welding & NRC: IRs 390/82-21 and 391/82-17; 390/84-35 and Engineering 391/84-33 B 79-004 Incorrect Weights for Swing Check C TVA: letter dated October 20, 1980 Valves Manufactured by Velan Engineering Corporation NRC: IR 390/83-15 and 391/83-11 B 79-005 Nuclear Incident at TMI NA Applies only to Babcock and Wilcox designed plants B 79-006 Review of Operational Errors and System C NRC: IR 390/80-06 and 391/80-05 Misalignments Identified During the Three Mile Island Incident B 79-007 Seismic Stress Analysis of Safety-Related C TVA: letter dated May 31, 1979 Piping NRC: IR 390/79-30 and 391/79-25 B 79-008 Events Relevant to BWRs Identified NA Boiling Water Reactor During TMI Incident B 79-009 Failure of GE Type AK-2 Circuit Breaker CI TVA: letter dated June 20, 1979 in Safety Related Systems Unit 2 Action: Complete preservice preventive maintenance on AK-2 Circuit Breakers.

[WAS "NOTE 3."]

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

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ITEM TITLE REV ADDITIONAL INFORMATION B 79-011 Faulty Overcurrent Trip Device in Circuit C TVA: letter dated July 20, 1979 Breakers for Engineering Safety Systems NRC: IR 390/79-30 and 391/79-25 B 79-012 Short Period Scrams at BWR Facilities NA Boiling Water Reactor B 79-013 Cracking in Feedwater Piping C Item was applicable only to units with operating license at the time the item was issued.

TVA: letter dated December 1, 1983 NRC: IR 390/391 85-08 B 79-014 Seismic Analysis for As-Built CI NRC review of HAAUP Program in NUREG-1232, SSER6, and Safety-Related Piping Systems SSER8.

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

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

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

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

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

NRC: IR 390/80-06 and 391/80-05 Page 8 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 79-020 Packaging, Transport and Burial of NA Item was applicable only to units with operating license at the Low-Level Radioactive Waste time the item was issued.

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

Measurements Unit 2 Action: Update accident calculation.

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

Unit 2 Action: Update accident calculation.

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

Used in Time Pieces for Luminosity NRC: IR 390/80-06 and 391/80-05 B 79-023 Potential Failure of Emergency Diesel C TVA: letter dated October 29, 1979 Generator Field Exciter Transformer NRC: IR 390/80-06 and 391/80-05 B 79-024 Frozen Lines CI Unit 2 Actions: Insulate the section of piping in the containment spray full-flow test line that is exposed to outside air. Confirm installation of heat tracing on the sensing lines off the feedwater flow elements.

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

Unit 2 Action: Issue appropriate emergency procedures.

B 79-028 Possible Malfunction of NAMCO Model C TVA: letter dated April 1, 1993 EA180 Limit Switches at Elevated Temperatures NRC: IR 390/391 93-32 B 80-001 Operability of ADS Valve Pneumatic NA Boiling Water Reactor Supply Page 9 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 80-002 Inadequate QA for Nuclear Supplied NA Boiling Water Reactor Equipment B 80-003 Loss of Charcoal from Standard Type II, 2 C TVA: letter dated March 21, 1980 Inch, Tray Adsorber Cells NRC: IR 390/80-15 and 391/80-12 B 80-004 Analysis of a PWR Main Steam Line CI IR 50-390/85-60 and 50-391/85-49 (December 6, 1985) required Break with Continued Feedwater Addition completion of actions that included determination of temperature profiles inside and outside of containment following a MSLB for Unit 1.

Unit 2 Action: Complete analysis for Unit 2.

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

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

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

Unit 2 Action: Perform verification during the preoperational testing.

B 80-007 BWR Jet Pump Assembly Failure NA Boiling Water Reactor B 80-008 Examination of Containment Liner C TVA: letter dated July 8, 1980 Penetration Welds NRC: IR 390/391 81-19 B 80-009 Hydramotor Actuator Deficiencies C TVA: letter dated January 15, 1981 NRC: NUREG/ CR 5291; IR 390/391 85-08; IR 390/85-60 and 391/85-49 B 80-010 Contamination of Nonradioactive System CI Unit 2 Actions: 2) Include proper monitoring of non-radioactive and Resulting Potential for Unmonitored, systems in procedures.

Uncontrolled Release of Radioactivity to Environment B 80-010 Contamination of Nonradioactive System CI Unit 2 Actions: 1) Correct deficiencies involving monitoring of and Resulting Potential for Unmonitored, systems.

Uncontrolled Release of Radioactivity to Environment B 80-011 Masonry Wall Design CI NRC accepted all but completion of corrective actions in IR 50-390/93-01 and 50-391/93-01(February 25, 1993) and closed for Unit 1 in IR 50-390/95-46 (August 1, 1995).

Unit 2 Action: Complete implementation for Unit 2.

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

[WAS "NOTE 3."]

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

NRC: IR 390/391 85-08 B 80-016 Potential Misapplication of Rosemount, C TVA: letter dated August 29, 1980 Inc. Models 1151 and 1152 Pressure Transmitters With Either A or D Output NRC: IR 390/391 81-17 Codes B 80-017 Failure of 76 of 185 Control Rods to Fully NA Boiling Water Reactor Insert During a Scram at a BWR B 80-018 Maintenance of Adequate Minimum Flow CI IR 50-390/85-60 and 50-391/85-49 (Unit 1)

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

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

Spring Return to Neutral Control Switches B 80-021 Valve Yokes Supplied by Malcolm C TVA: letter dated May 6, 1981 Foundry Co., Inc.

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

008 Sealed-Source Connectors B 80-023 Failures of Solenoid Valves Manufactured C TVA: letter dated March 31, 1981 by Valcor Engineering Corporation NRC: IR 390/391 81-17; NUREG/CR 5292 Page 11 of 95 * = See last page for status code definition.

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

B 80-025 Operating Problems with Target Rock NA Boiling Water Reactor Safety-Relief Valves at BWRs B 81-001 Surveillance of Mechanical Snubbers NA NRC: IR 390/391 81-17 B 81-002 Failure of Gate Type Valves to Close C TVA: letter dated September 30, 1983 Against Differential Pressure NRC: IR 390/391 84-03 B 81-003 Flow Blockage of Cooling Water to Safety C TVA: letters dated July 21, 1981 and March 21, 1983 System Components by Asiatic Clams and Mussels NRC: IR 390/391 81-17 B 82-001 Alteration of Radiographs of Welds in C NRC: IR 390/391 85-08 Piping Subassemblies B 82-002 Degradation of Threaded Fasteners in the CI TVA: memo dated February 6, 1985 Reactor Coolant Pressure Boundary of PWR Plants NRC: IR 390/391 85-08 Approach accepted in IR 50-390/85-08 and 50-391/85-08 (March 29, 1985).

Unit 2 Action: Implement same approach as Unit 1.

B 82-003 Stress Corrosion Cracking in Thick-Wall, NA Boiling Water Reactor Large Diameter, Stainless Steel, Recirculation System Piping at BWR Plants B 82-004 Deficiencies in Primary Containment C TVA: letter dated January 24, 1983 Electrical Penetration Assemblies NRC: IR 390/83-10 and 391/83-08 B 83-001 Failure of Trip Breakers (Westinghouse C NRC: IRs 390/391 85-08 and 390/391 92-13 DB-50) to Open on Automatic Trip Signal B 83-002 Stress Corrosion Cracking in Large- NA Boiling Water Reactor Diameter Stainless Steel Recirculation System Piping at BWR Plants B 83-003 Check Valve Failures in Raw Water NA Addressed by Inservice Testing for Construction Permit holders Cooling Systems of Diesel Generators Page 12 of 95 * = See last page for status code definition.

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

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

Unit 2 Action: Implement as necessary.

REVISION 04 UPDATE:

NRC Inspection Report Nos. 50-390/90-02 and 50-391/90-02 found the proposed alternative to ASME code paragraph NA-3451 (a) to be acceptable. It noted that TVA must revise the FSAR to document this deviation from ASME Section III requirements.

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

Final action was to incorporate the exemption in the Unit 2 FSAR.

This exemption is documented in Unit 2 FSAR Section 3.2 in paragraph 3.2.3.2 and Table 3.2-2a as explained in Note 4. of the table.

B 83-007 Apparently Fraudulent Products Sold by C TVA: letter dated March 22, 1984 Ray Miller, Inc.

NRC: IR 390/85-03 and 391/85-04 B 83-008 Electrical Circuit Breakers With an C TVA: letter dated March 29, 1984 Undervoltage Trip Feature in Safety-Related Applications Other Than NRC: IR 390/84-35 and 391/84-33 the Reactor Trip System B 84-001 Cracks in BWR Mark 1 Containment Vent NA Boiling Water Reactor Headers B 84-002 Failure of GE Type HFA Relays In Use In C TVA: letter dated July 10, 1984 Class 1E Safety Systems NRC: IR 390/391 84-42 and IR 390/84-77 and 391/84-54 Page 13 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 84-003 Refueling Cavity Water Seal CI Reviewed in IR 390/93-11.

Unit 2 Action: Ensure appropriate abnormal operating instructions (AOIs) are used for Unit 2.

B 85-001 Steam Binding of Auxiliary Feedwater CI TVA: letter dated January 27, 1986 Pumps NRC: IR 390/391 90-20 NRC accepted approach in letter dated July 20, 1988, and reviewed response in Appendix EE of SSER16.

Unit 2 Action: Procedures and hardware will be in place to ensure recognition of indications of steam binding and maintenance of system operability until check valves are repaired and back leakage stopped.

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

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

Not Terminate Treatment Dose B 87-001 Thinning of Pipe Walls in Nuclear Power C TVA: letter dated September 18, 1987 Plants NRC: NUREG/CR 5287 Closed to GL 89-08 Page 14 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 87-002 Fastener Testing to Determine CI TVA: letters dated April 15, 1988, July 6, 1988, Conformance with Applicable Material September 12, 1988, and January 27, 1989 Specifications 03 NRC: letter dated August 18, 1989 NRC closed in letter dated August 18, 1989.

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

REVISION 03 UPDATE:

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

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

Steam Generator Tubes Unit 2 Actions: Evaluate E/C data to determine anti-vibration bar penetration depth; perform T/H analysis to identify susceptible tubes; modify, if necessary.

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

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

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

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

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

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

Unit 2 Action: Implement program to prevent thermal stratification.

Page 15 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 88-009 Thimble Tube Thinning in Westinghouse CI Reviewed in Appendix EE of SSER16.

Reactors Unit 2 Action: TVA letter dated March 11, 1994, for both units committed to establish a program and inspect the thimble tubes during the first refueling outage.

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

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

Unit 2 Action: Complete modifications to accommodate Surge Line thermal movements and incorporate a temperature limitation during heatup and cooldown operations into Unit 2 procedures.

B 89-001 Failure of Westinghouse Steam CI NRC acceptance letter dated September 26, 1991 for both Generator Tube Mechanical Plugs units.

Unit 2 Action: Remove SG tube plugs.

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

High-Hardness Type 410 Stainless Steel Internal Preloaded Bolting in Anchor Unit 2 Actions: Replace the flapper assembly hold-down bolts Darling Model S350W Swing Check fabricated on the 14 (12 valves are installed) Atwood and Morrell Valves or Valves of Similar Nature 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 89-003 Potential Loss of Required Shutdown CI TVA: letter dated June 19, 1990 Margin During Refueling Operations NRC: IR 390/391 94-04 and letter dated June 22, 1990 NRC acceptance letter dated June 22, 1990.

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

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

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

Controls Page 16 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 92-001 Failure of Thermo-Lag 330 Fire Barrier NA ----------------------------------------------------------------------------------------

System to Maintain Cabling in Wide ----------------------------------------------------------------------------------------

Cable Trays and Small Conduits Free 02 From Fire Damage REVISION 02 UPDATE:

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

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

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

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

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

REVISION 02 UPDATE:

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

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

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

Transportation of Radioactive Material B 95-002 Unexpected Clogging of a Residual Heat NA Boiling Water Reactor Removal Pump Strainer While Operating in Suppression Pool Cooling Mode Page 17 of 95 * = See last page for status code definition.

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

04 Unit 2 Action: Issue Emergency Operating Procedure.

REVISION 02 UPDATE:

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

REVISION 03 UPDATE:

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

REVISION 04 UPDATE:

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

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

04 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.

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

Fuel, Over Fuel in the Reactor, or Over Safety-Related Equipment 02 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.

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

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

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

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

Nozzles 04 Unit 2 Action: Perform baseline inspection.

REVISION 02 UPDATE:

Unit 2 Action: 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.

Page 19 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION B 02-001 RPV Head Degradation and Reactor CI NRC review of Unit 1's 15 day response in letter dated Coolant Pressure Boundary Integrity May 20, 2002 - Initial response for Unit 2 on 04 September 7, 2007.

Unit 2 Action: Perform baseline inspection.

REVISION 02 UPDATE:

Unit 2 Action: 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.

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

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

04 Unit 2 Action: Perform baseline inspection.

REVISION 02 UPDATE:

Unit 2 Action: 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.

B 03-001 Potential Impact of Debris Blockage on NA TVA: letter dated September 7, 2007 Emergency Sump Recirculation at PWRs Page 20 of 95 * = See last page for status code definition.

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

Pressure Boundary Integrity (PWRs) 02 Unit 2 Action: Perform baseline inspection.

REVISION 02 UPDATE:

NRC issued the Safety Evaluation for Bulletin 2003-002 on January 21, 2010.

Unit 2 Action: Perform baseline inspection. Evaluate or repair as necessary.

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

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

Page 21 of 95 * = See last page for status code definition.

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

Used in the Fabrication of Pressurizer Penetrations and Steam Space Piping 04 Unit 2 Actions: Provide details of pressurizer and penetrations Connections at PWRs and apply Material Stress Improvement Process.

REVISION 02 UPDATE:

TVA provided details of the pressurizer and penetrations on September 29, 2008. This letter committed to:

Prior to placing the pressurizer in service, TVA will apply the Material Stress Improvement Process (MSIP) to the Pressurizer Power Operated Relief Valve connections, the safety relief valve connections, the spray line nozzle and surge line nozzle connections.

TVA will perform a bare metal visual (BMV) inspection of the upper pressurizer Alloy 600 locations at the first refueling outage.

REVISION 03 UPDATE:

April 1, 2010, letter committed to:

TVA will perform NDE prior to and after performance of the MSIP. If circumferential cracking is observed in either pressure boundary or non-pressure boundary portions of any locations covered under the scope of the bulletin, TVA will develop plans to perform an adequate extent-of-condition evaluation, and TVA will discuss those plans with cognizant NRC technical staff prior to starting Unit 2.

After performing the BMV inspection during the first refueling outage, if any evidence of apparent reactor coolant pressure boundary leakage is discovered, then NDE capable of determining crack orientation will be performed in order to accurately characterize the flaw, the orientation, and extent. TVA will develop plans to perform an adequate extent of condition evaluation, and plans to possibly expand the scope of NDE to other components in the pressurizer will be discussed with NRC technical staff prior to restarting of Unit 2.

REVISION 04 UPDATE:

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

B 05-001 Material Control and Accounting at C TVA: letters dated March 21, 2005 and May 11, 2005 Reactors and Wet Spent Fuel Storage Facilities --------------------

Item concerns a multi-unit issue that was completed for both units.

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ITEM TITLE REV ADDITIONAL INFORMATION B 05-002 Emergency Preparedness and Response C TVA: letters dated January 20, 2006 and August 16, 2006.

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

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

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

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

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

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

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

C 77-001 Malfunctions of Limitorque Valve NA Info Operators C 77-002a Potential Heavy Spring Flooding (CP) NA Item was applicable only to units with operating license at the time the item was issued.

C 77-003 Fire Inside a Motor Control Center NA Info Page 23 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION C 77-004 Inadequate Lock Assemblies NA Info C 77-005 Fluid Entrapment in Valve Bonnets NA Info C 77-006 Effects of Hydraulic Fluid on Electrical NA Info Cables C 77-007 Short Period During Reactor Startup NA Boiling Water Reactor C 77-008 Failure of Feedwater Sample Probe NA Item was applicable only to units with operating license at the time the item was issued.

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

C 77-011 Leakage of Containment Isolation Valves NA Info with Resilient Seats C 77-012 Dropped Fuel Assemblies at BWR NA Boiling Water Reactor Facilities C 77-013 Reactor Safety Signals Negated During NA Info Testing C 77-014 Separation of Contaminated Water NA Info Systems from Noncontaminated Plant Systems C 77-015 Degradation of Fuel Oil Flow to the NA Info Emergency Diesel Generator C 77-016 Emergency Diesel Generator Electrical NA Info Trip Lock-Out Features Page 24 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION C 78-001 Loss of Well Logging Source NA Does not apply to power reactor.

C 78-002 Proper Lubricating Oil for Terry Turbines NA Info C 78-003 Packaging Greater Than Type A NA Info Quantities of Low Specific Activity Radioactive Material for Transport C 78-004 Installation Errors That Could Prevent NA Info Closing of Fire Doors C 78-005 Inadvertent Safety Injection During NA Info Cooldown C 78-006 Potential Common Mode Flooding of NA Info ECCS Equipment Rooms at BWR Facilities C 78-007 Damaged Components of a NA Info Bergen-Paterson Series 25000 Hydraulic Test Stand C 78-008 Environmental Qualification of NA Info Safety-Related Electrical Equipment at Nuclear Power Plants C 78-009 Arcing of General Electric Company NA Info Size 2 Contactors C 78-010 Control of Sealed Sources in Radiation NA Does not apply to power reactor.

Therapy C 78-011 Recirculation MG Set Overspeed Stops NA Boiling Water Reactor C 78-012 HPCI Turbine Control Valve Lift Rod NA Boiling Water Reactor Bending C 78-013 Inoperability of Service Water Pumps NA Info Page 25 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION C 78-014 HPCI Turbine Reversing Chamber Hold NA Boiling Water Reactor Down Bolting C 78-015 Tilting Disc Check Valves Fail to Close NA Info with Gravity in Vertical Position C 78-016 Limitorque Valve Actuators NA Info C 78-017 Inadequate Guard Training/Qualification NA Info and Falsified Training Records C 78-018 UL Fire Test NA Info C 78-019 Manual Override (Bypass) of Safety NA Info System Actuation Signals C 79-001 Administration of Unauthorized Byproduct NA Does not apply to power reactor.

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

in Nuclear Medicine C 79-007 Unexpected Speed Increase of Reactor NA Boiling Water Reactor Recirculation MG Set Resulted in Reactor Power Increase Page 26 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION C 79-008 Attempted Extortion - Low Enriched NA Fuel facilities and operating reactors at the time the item was Uranium issued C 79-009 Occurrences of Split or Punctured NA Info Regulator Diaphragms in Certain Self Contained Breathing Apparatus C 79-010 Pipefittings Manufactured from NA Info Unacceptable Material C 79-011 Design/Construction Interface Problem NA Info C 79-012 Potential Diesel Generator Turbocharger NA Info Problem C 79-013 Replacement of Diesel Fire Pump NA Info Starting Contactors C 79-014 Unauthorized Procurement and NA Does not apply to power reactor.

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

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

Members of the General Public and a Radiographer C 79-017 Contact Problem in SB-12 Switches on NA Info General Electric Company Metalclad Circuit Breakers C 79-018 Proper Installation of Target Rock NA Boiling Water Reactor Safety-Relief Valves C 79-019 Loose Locking Devices on Ingersoll-Rand NA Info Pumps C 79-020 Failure of GTE Sylvania Relay Type PM NA Info Bulletin 7305 Catalog 5U12-11-AC with a 120V AC Coil Page 27 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION C 79-021 Prevention of Unplanned Releases of NA Info Radioactivity C 79-022 Stroke Times for Power Operated Relief NA Info Valves C 79-023 Motor Starters and Contactors Failed to C The Circular did not require a response.

Operate 01 TVA reported a nonconformance under 10 CFR 50.55e on January 17, 1980, that four motor starters of this type had been located in the 480V control and auxiliary vent boards at WBN.

Gould factory representatives supervised the replacement of the carrier assemblies in accordance with the Gould instructions.

The starters with replaced carriers were acceptable.

NRC IR 50-390/80-03 and 50-391/80-02 reviewed and closed the associated nonconformance reports.

C 79-024 Proper Installation and Calibration of Core NA Boiling Water Reactor Spray Pipe Break Detection Equipment on BWRs C 79-025 Shock Arrestor Strut Assembly C The Circular did not require a response.

Interference 01 TVA reported a nonconformance under 10 CFR 50.55e on March 6, 1980, that a review had determined that nine installed supports had brackets with the potential of hindering full function of the support. Additional supports that were not installed had the same potential problem. TVA initially determined that the supports would be modified in accordance with a vendor approved drawing. TVA subsequently determined that no actual problem existed and no field work was required.

NRC IR 50-390/83-15 and 50-391/83-11 reviewed and closed the associated nonconformance reports.

C 80-001 Service Advice for GE Induction Disc NA Info Relays C 80-002 Nuclear Power Plant Staff Work Hours NA Info C 80-003 Protection from Toxic Gas Hazards NA Info C 80-004 Securing of Threaded Locking Devices on NA Info Safety-Related Equipment C 80-005 Emergency Diesel-Generator Lubricating NA Info Oil Addition and Onsite Supply Page 28 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION C 80-006 Control and Accountability Systems for NA Does not apply to power reactor.

Implant Therapy Sources C 80-007 Problems with HPCI Turbine Oil System NA Boiling Water Reactor C 80-008 BWR Technical Specification NA Boiling Water Reactor Inconsistency - RPS Response Time C 80-009 Problems with Plant Internal NA Info Communications Systems C 80-010 Failure to Maintain Environmental NA Info Qualification of Equipment C 80-011 Emergency Diesel Generator Lube Oil NA Info Cooler Failures C 80-012 Valve-Shaft-to-Actuator Key May Fall Out NA Info of Place when Mounted Below Horizontal Axis C 80-013 Grid Strap Damage in Westinghouse Fuel NA Info Assemblies C 80-014 Radioactive Contamination of Plant NA Info Demineralized Water System and Resultant Internal Contamination of Personnel C 80-015 Loss of Reactor Coolant Pump Cooling NA Info and Natural Circulation Cooldown C 80-016 Operational Deficiencies in Rosemount NA Info Model 510DU Trip Units and Model 1152 Pressure Transmitters C 80-017 Fuel Pin Damage Due to Water Jet from NA Info Baffle Plate Corner C 80-018 10 CFR 50.59 Safety Evaluations for NA Info Changes to Radioactive Waste Treatment Systems Page 29 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION C 80-019 Noncompliance with License NA Does not apply to power reactor.

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

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

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

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

C 81-003 Inoperable Seismic Monitoring NA Info Instrumentation C 81-004 The Role of Shift Technical Advisors and NA Info Importance of Reporting Operational Events C 81-005 Self-Aligning Rod End Bushings for Pipe NA Info Supports C 81-006 Potential Deficiency Affecting Certain NA Info Foxboro 10 to 50 Milliampere Transmitters Page 30 of 95 * = See last page for status code definition.

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

Cooldown C 81-011 Inadequate Decay Heat Removal During NA Boiling Water Reactor Reactor Shutdown C 81-012 Inadequate Periodic Test Procedure of NA Info PWR Reactor Protection System C 81-013 Torque Switch Electrical Bypass Circuit C The Circular did not require a response.

for Safeguard Service Valve Motors 01 TVA reported a nonconformance under 10 CFR 50.55e on April 4, 1986 (NCR W367-P), that required closing torque switches were 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.

C 81-014 Main Steam Isolation Valve Failures to NA Info Close C 81-015 Unnecessary Radiation Exposures to the NA Info Public and Workers During Events Involving Thickness and Level Measuring Devices GL 77-001 Intrusion Detection Systems Handbook NA Info GL 77-002 Fire Protection Functional Responsibilities NA Info GL 77-003 Transmittal of NUREG-0321, A Study of NA Info the Nuclear Regulatory Commission Quality Assurance Program Page 31 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 77-004 Shipments of Contaminated Components NA Info From NRC Licensed Power Facilities to Vendors & Service Companies GL 77-005 Nonconformity of Addressees of Items NA Info Directed to the Office of Nuclear Reactor Regulation GL 77-006 Enclosing Questionnaire Related to NA Item was applicable only to units with operating license at the Steam Generators time the item was issued.

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

GL 77-008 Revised Intrusion Detection Handbook NA Info and Entry Control Systems Handbook GL 78-001 Correction to Letter of 12/15/77 [GL 77-07] NA Item was applicable only to units with operating license at the time the item was issued.

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

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

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

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

GL 78-005 Internal Distribution of Correspondence NA Info

- Asking for Comments on Mass Mailing System GL 78-006 This GL was never issued. NA GL 78-007 This GL was never issued. NA GL 78-008 Enclosing NUREG-0408 Re Mark I NA Boiling Water Reactor Containments, and Granting Exemption from GDC 50 and Enclosing Sample Notice Page 32 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 78-009 Multiple-Subsequent Actuations of NA Boiling Water Reactor Safety/Relief Valves Following an Isolation Event GL 78-010 Guidance on Radiological Environmental NA Info Monitoring GL 78-011 Guidance on Spent Fuel Pool NA Info Modifications GL 78-012 Notice of Meeting Regarding NA Info Implementation of 10 CFR 73.55 Requirements and Status of Research GL 78-013 Forwarding of NUREG-0219 NA Info GL 78-014 Transmittal of Draft NUREG-0219 for NA Info Comment GL 78-015 Request for Information on Control of NA See GL 81-007.

Heavy Loads Near Spent Fuel GL 78-016 Request for Information on Control of NA Info Heavy Loads Near Spent Fuel Pools GL 78-017 Corrected Letter on Heavy Loads Over NA Info Spent Fuel GL 78-018 Corrected Letter on Heavy Loads Over NA Duplicate of GL 81-007 Spent Fuel GL 78-019 Enclosing Sandia Report SAND 77-0777, NA Info Barrier Technology Handbook GL 78-020 Enclosing - A Systematic Approach to NA Info the Conceptual Design of Physical Protection Systems for Nuclear Facilities GL 78-021 Transmitting NUREG/CR-0181, NA Info Concerning Barrier and Penetration Data Needed for Physical Security System Assessment Page 33 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 78-022 Revision to Intrusion Detection Systems NA Info and Entry Control Systems Handbooks and Nuclear Safeguards Technology Handbook GL 78-023 Manpower Requirements for Operating NA Info Reactors GL 78-024 Model Appendix I Technical NA Boiling Water Reactor Specifications and Submittal Schedule For BWRs GL 78-025 This GL was never issued. NA GL 78-026 Excessive Control Rod Guide Tube Wear NA Applies only to Babcock and Wilcox designed plants GL 78-027 Forwarding of NUREG-0181 NA Info GL 78-028 Forwarding pages omitted from 07/11/78 NA Boiling Water Reactor letter [GL 78-24]

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

Page 34 of 95 * = See last page for status code definition.

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

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

Specifications, PWR, and NUREG-0133 GL 78-039 Forwarding of 2 Tables of Appendix I, NA Boiling Water Reactor Draft Radiological Effluent Technical Specifications, BWR, and NUREG-0133 GL 78-040 Training & Qualification Program NA Info Workshops GL 78-041 Mark II Generic Acceptance Criteria For NA Boiling Water Reactor Lead Plants GL 78-042 Training and Qualification Program NA Info Workshops GL 79-001 Interservice Procedures for Instructional NA Info Systems Development - TRADOC GL 79-002 Transmitting Rev. to Entry Control NA Info Systems Handbook (SAND 77-1033),

Intrusion Detection Handbook (SAND 76-0554), and Barrier Penetration Database GL 79-003 Offsite Dose Calculation Manual NA Info GL 79-004 Referencing 4/14/78 Letter - NA Info Modifications to NRC Guidance "Review and Acceptance of Spent Fuel Pool Storage and Handling" GL 79-005 Information Relating to Categorization of NA Info Recent Regulatory Guides by the Regulatory Requirements Review Committee Page 35 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 79-006 Contents of the Offsite Dose Calculation NA Info Manual GL 79-007 Seismic (SSE) and LOCA Responses NA Info (NUREG-0484)

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

Comment GL 79-011 Transmitting "Summary of Operating NA Info Experience with Recalculating Steam Generators, January 1979," NUREG-0523 GL 79-012 ATWS - Enclosing Letter to GE, with NA Info NUREG-0460, Vol. 3 GL 79-013 Schedule for Implementation and NA Info Resolution of Mark I Containment Long Term Program GL 79-014 Pipe Crack Study Group - Enclosing NA Info NUREG-0531 and Notice GL 79-015 Steam Generators - Enclosing Summary NA Info of Operating Experience with Recirculating Steam Generators, NUREG-0523 GL 79-016 Meeting Re Implementation of Physical NA Info Security Requirements GL 79-017 Reliability of Onsite Diesel Generators at NA Info Light Water Reactors GL 79-018 Westinghouse Two-Loop NSSS NA Addressed to specific plant(s).

Page 36 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 79-019 NRC Staff Review of Responses to NA Addressed to specific plant(s).

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

GL 79-021 Enclosing NUREG/CR-0660, NA Info Enhancement of on Site Emergency Diesel Generator Reliability" GL 79-022 Enclosing NUREG-0560, "Staff Report on NA Applies only to Babcock and Wilcox designed plants the Generic Assessment of Feedwater Transients in PWRs Designed by B&W GL 79-023 NRC Staff Review of Responses to B 79- NA Boiling Water Reactor 08 GL 79-024 Multiple Equipment Failures in NA GL 79-24 provided a discussion of an inadvertent reactor scram Safety-Related Systems and safety injection during monthly surveillance tests of the 01 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.

GL 79-025 Information Required to Review NA Info Corporate Capabilities GL 79-026 Upgraded Standard Technical NA Info Specification Bases Program GL 79-027 Operability Testing of Relief and Safety NA Boiling Water Reactor Relief Valves GL 79-028 Evaluation of Semi-Scale Small Break NA Info Experiment GL 79-029 Transmitting NUREG-0473, Revision 2, NA Info Draft Radiological Effluent Technical Specifications Page 37 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 79-030 Transmitting NUREG-0472, Revision 2, NA Info Draft Radiological Technical Specifications GL 79-031 Submittal of Copies of Response to NA Info 6/29/79 NRC Request [79-25]

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

(FR 43280-285)

GL 79-035 Regional Meetings to Discuss Impacts on NA Info Emergency Planning GL 79-036 Adequacy of Station Electric Distribution CI This GL tracked compliance with BTP PSB-1, Adequacy of Systems Voltages Station Electric Distribution System Voltages.

Unit 2 Action: Perform verification during the preoperational testing.

GL 79-037 Amendment to 10 CFR 73.55 Deferral NA Info from 8/1/79 to 11/1/79 GL 79-038 BWR Off-Gas Systems - Enclosing NA Boiling Water Reactor NUREG/CR-0727 GL 79-039 Transmitting Division 5 Draft Regulatory NA Does not apply to power reactor.

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

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

Systems and Safety Grade Systems Page 38 of 95 * = See last page for status code definition.

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

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

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

Operability GL 79-047 Radiation Training NA Info GL 79-048 Confirmatory Requirements Relating to NA Boiling Water Reactor Condensation Oscillation Loads for the Mark I Containment Long Term Program GL 79-049 Summary of Meetings Held on 9/18-20/79 NA Info to Discuss Potential Unreviewed Safety Question on Systems Interaction for B&W Pl GL 79-050 Emergency Plans Submittal Dates NA Info GL 79-051 Follow-up Actions Resulting from the NA GL 79-51 provided follow-up actions resulting from the Three NRC Staff Reviews Regarding the TMI-2 Mile Island Unit 2 accident. GL 79-51 was provided for planning Accident 01 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.

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

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

Normal Operation Page 39 of 95 * = See last page for status code definition.

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

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

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

Duplicate of GL 79-058 GL 79-063 Upgraded Emergency Plans C GL 79-63 advised applicants for licenses of proposed rulemaking that NRC concurrence in State and local emergency plans would 01 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.

GL 79-064 Suspension of All Operating Licenses NA Info (PWRs)

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

GL 79-067 Estimates for Evacuation of Various NA Info Areas Around Nuclear Power Reactors Page 40 of 95 * = See last page for status code definition.

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

GL 80-010 Issuance of NUREG-0588, "Interim Staff NA Info Position On Equipment Qualifications of Safety-Related Electrical Equipment Page 41 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 80-011 B 80-03, Loss of Charcoal From C GL 80-11 transmitted Bulletin 80-03. TVA responded to B 80-03 Standard Type II, 2 Inch, Tray Absorber on March 21, 1980. See B 80-03 for further information.

Cells 01 GL 80-012 B 80-04, Analysis of a PWR Main Steam NA Info Line Break With Continued Feedwater Addition GL 80-013 Qualification of Safety Related Electrical NA Item was applicable only to units with operating license at the Equipment time the item was issued.

GL 80-014 LWR Primary Coolant System Pressure S TVA: FSAR 5.2.7.4 Isolation Valves 02 NRC: 1.14.2 of SSER 6 NRC reviewed in 1.14.2 of SSER6.

Unit 2 Action: Incorporate guidance into Technical Specifications.

REVISION 02 UPDATE:

Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.

TS Surveillance Requirement 3.4.13.1 verifies RCS operational leakage by performance of an RCS water inventory balance.

GL 80-015 Request for Additional Management and NA Info Technical Resources Information GL 80-016 B 79-01b, Environmental Qualification of NA Info Class 1E Equipment GL 80-017 Modifications to BWR Control Rod Drive NA Boiling Water Reactor Systems GL 80-018 Crystal River 3 Reactor Trip From NA Applies only to Babcock and Wilcox designed plants Approximately 100% Full Power GL 80-019 Resolution of Enhanced Fission Gas NA Info Release Concern GL 80-020 Actions Required From OL Applicants of NA Info NSSS Designs by W and CE Resulting From NRC B&O Task Force Review of TMI2 Accident Page 42 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 80-021 B 80-05, Vacuum Condition Resulting in CI Closed in IR 50-390/84-59 and 50-391/84-45.

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

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

01 GL 80-029 Modifications to Boiling Water Reactor NA Boiling Water Reactor Control Rod Drive Systems GL 80-030 Clarification of The Term "Operable" As It NA Item was applicable only to units with operating license at the Applies to Single Failure Criterion For time the item was issued.

Safety Systems Required by TS GL 80-031 B 80-09, Hydramotor Actuator NA Info Deficiencies GL 80-032 Information Request on Category I C GL 80-32 transmitted NRC questions on masonry walls.

Masonry Walls Employed by Plants TVA provided the information requested by letters dated Under CP and OL Review 01 February 12, 1981, for reinforced walls and August 20, 1981, for nonreinforced walls. TVA provided a final response on January 22, 1982. See B 80-11 for further information.

GL 80-033 Actions Required From OL Applicants of NA Applies only to Babcock and Wilcox designed plants B&W Designed NSSS Resulting From NRC B&O Task Force Review of TMI2 Accident Page 43 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 80-034 Clarification of NRC Requirements for NA Info Emergency Response Facilities at Each Site GL 80-035 Effect of a DC Power Supply Failure on NA Boiling Water Reactor ECCS Performances GL 80-036 B 80-10, Contamination of NA Info Non-Radioactive System and Resulting Potential For Unmonitored, Uncontrolled Release to Environment GL 80-037 Five Additional TMI-2 Related NA Item was applicable only to units with operating license at the Requirements to Operating Reactors time the item was issued.

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

Physical Protection Requirements GL 80-039 B 80-11, Masonry Wall Design NA Info GL 80-040 Transmittal of NUREG-0654, "Report of NA Info the B&O Task Force and Appropriate NUREG-0626, "Generic Evaluation of FW Transient and Small Break LOCA GL 80-041 Summary of Meetings Held on NA Info April 22 &23, 1980 With Representatives of the Mark I Owners Group GL 80-042 B 80-12, Decay Heat Removal System NA Info Operability GL 80-043 B 80-13, Cracking In Core Spray NA Boiling Water Reactor Spargers GL 80-044 Reorganization of Functions and NA Info Assignments Within ONRR/SSPB GL 80-045 Fire Protection Rule NA Item was applicable only to units with operating license at the time the item was issued.

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

Generator Coolant Pump Supports" Page 44 of 95 * = See last page for status code definition.

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

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

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

Letter [GL 80-37]

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

GL 80-054 B 80-14, Degradation of Scram NA Boiling Water Reactor Discharge Volume Capability GL 80-055 B 80-15, Possible Loss of Hotline With NA Info Loss of off-Site Power GL 80-056 Commission Memorandum and Order on NA Info Equipment Qualification GL 80-057 Further Commission Guidance For Power NA Info Reactor Operating Licenses NUREG-0660 and NUREG-0694 GL 80-058 B 80-16, Potential Misapplication of NA Info Rosemount Inc. Models 1151/1152 Pressure Transmitters With "A" Or "D" Output Codes GL 80-059 Transmittal of Federal Register Notice RE NA Info Regional Meetings to Discuss Environmental Qualification of Electrical Equipment GL 80-060 Request for Information Regarding NA Info Evacuation Times Page 45 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 80-061 TMI-2 Lessons Learned NA Info GL 80-062 TMI-2 Lessons Learned NA Boiling Water Reactor GL 80-063 B 80-17, Failure of Control Rods to Insert NA Boiling Water Reactor During a Scram at a BWR GL 80-064 Scram Discharge Volume Designs NA Boiling Water Reactor GL 80-065 Request for Estimated Construction NA Info Completion and Fuel Load Schedules GL 80-066 B 80-17, Supplement 1, Failure of NA Boiling Water Reactor Control Rods to Insert During a Scram at a BWR GL 80-067 Scram Discharge Volume NA Boiling Water Reactor GL 80-068 B 80-17, Supplement 2, Failures NA Boiling Water Reactor Revealed by Testing Subsequent to Failure of Control Rods to Insert During a Scram at a BWR GL 80-069 B 80-18, Maintenance of Adequate NA Info Minimum Flow Through Centrifugal Charging Pumps Following Secondary Side HELB GL 80-070 B 80-19, Failures of Mercury-Wetted NA Info Matrix Relays in RPS of Operating Nuclear Power Plants Designed by GE GL 80-071 B 80-20, Failures of Westinghouse Type NA Info W-2 Spring Return to Neutral Control Switches GL 80-072 Interim Criteria For Shift Staffing NA Info GL 80-073 "Functional Criteria For Emergency NA Info Response Facilities, NUREG-0696 Page 46 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 80-074 Notice of Forthcoming Meeting With NA Info Representatives of EPRI to Discuss Program For Resolution of USI A-12, Fracture Toughness Issue GL 80-075 Lessons Learned Tech. Specs. NA Item was applicable only to units with operating license at the time the item was issued.

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

02 REVISION 02 UPDATE:

Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.

TS LCO 3.9.7 requires the refueling cavity water level to be maintained greater than or equal to 23 feet above the top of the reactor vessel flange during movement of irradiated fuel assemblies within containment.

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

GL 80-082 B 79-01b, Supplement 2, Environmental NA Info Qualification of Class 1E Equipment GL 80-083 Environmental Qualification of NA Info Safety-Related Equipment GL 80-084 BWR Scram System NA Boiling Water Reactor Page 47 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 80-085 Implementation of Guidance From NA Info USI A-12, Potential For LOW Fracture Toughness and Lamellar Tearing On Component Support GL 80-086 Notice of Meeting to Discuss Final NA Info Resolution of USI A-12 GL 80-087 Notice of Meeting to Discuss Status of NA Info EPRI-Proposed Resolution of the USI A-12 Fracture Toughness Issue GL 80-088 Seismic Qualification of Auxiliary NA Item was applicable only to units with operating license at the Feedwater Systems time the item was issued.

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

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

GL 80-091 ODYN Code Calculation NA Boiling Water Reactor GL 80-092 B 80-21, Valve Yokes Supplied by C GL 80-92 transmitted Bulletin 80-21. TVA responded to Malcolm Foundry Company, Inc. B 80-21 on May 6, 1981. See B 80-21 for further information.

01 GL 80-093 Emergency Preparedness NA Does not apply to power reactor.

GL 80-094 Emergency Plan NA Info GL 80-095 Generic Technical Activity A-10, NA Boiling Water Reactor NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking GL 80-096 Fire Protection NA Addressed to specific plant(s).

GL 80-097 B 80-23, Failures of Solenoid Valves NA Info Manufactured by Valcor Engineering Corporation Page 48 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 80-098 B 80-24, Prevention of Damage Due to NA Info Water Leakage Inside Containment GL 80-099 Technical Specifications Revisions For NA Info Snubber Surveillance GL 80-100 Appendix R to 10 CFR 50 Regarding Fire NA Item was applicable only to units with operating license at the Protection - Federal Register Notice time the item was issued.

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

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

GL 80-103 Fire Protection - Revised Federal NA Info Register Notice GL 80-104 Orders On Environmental Qualification of NA Info Safety Related Electrical Equipment GL 80-105 Implementation of Guidance For USI NA Info A-12, Potential For Low Fracture toughness and Lamellar Tearing On Component Supports GL 80-106 Report On ECCS Cladding Models, NA Info NUREG-0630 GL 80-107 BWR Scram Discharge System NA Boiling Water Reactor GL 80-108 Emergency Planning NA Info GL 80-109 Guidelines For SEP Soil Structure NA Info Interaction Reviews GL 80-110 Periodic Updating of FSARS NA Item was applicable only to units with operating license at the time the item was issued.

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

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

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

GL 81-007 Control of Heavy Loads 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 TVAs response to NUREG-0612 guidelines for Phase I in TVA letter dated July 28, 1993.

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

GL 81-008 ODYN Code NA Boiling Water Reactor Page 50 of 95 * = See last page for status code definition.

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

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

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

[WAS "OL."]

GL 81-015 Environmental Qualification of Class 1E NA Info Electrical Equipment - Clarification of Staffs Handling of Proprietary Information GL 81-016 NUREG-0737, Item I.C.1 SER on NA Applies only to Babcock and Wilcox designed plants Abnormal Transient Operating Guidelines (ATOG)

GL 81-017 Functional Criteria for Emergency NA Info Response Facilities GL 81-018 BWR Scram Discharge System - NA Boiling Water Reactor Clarification of Diverse Instrumentation Requirements GL 81-019 Thermal Shock to Reactor Pressure NA Item was applicable only to units with operating license at the Vessels time the item was issued.

GL 81-020 Safety Concerns Associated With Pipe NA Boiling Water Reactor Breaks in the BWR Scram System Page 51 of 95 * = See last page for status code definition.

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

Unit 2 Action: Issue operating procedures.

GL 81-022 Engineering Evaluation of the NA Info H. B. Robinson Reactor Coolant System Leak on 1/29/81 GL 81-023 INPO Plant Specific Evaluation Reports NA Info GL 81-024 Multi-Plant Issue B-56, Control Rods Fail NA Boiling Water Reactor to Fully Insert GL 81-025 Change in Implementing Schedule For NA Info Submission and Evaluation of Upgraded Emergency Plans GL 81-026 Licensing Requirements for Pending NA Applicants with pending Construction Permits Construction Permit and Manufacturing License Applications GL 81-027 Privacy and Proprietary Material in NA Info Emergency Plans GL 81-028 Steam Generator Overfill NA Info GL 81-029 Simulator Examinations NA Info GL 81-030 Safety Concerns Associated With Pipe NA Boiling Water Reactor Breaks in the BWR Scram System GL 81-031 This GL was never issued. NA GL 81-032 NUREG-0737, Item II.K.3.44, Evaluation NA Boiling Water Reactor of Anticipated Transients Combined With Single Failure GL 81-033 This GL was never issued. NA Page 52 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 81-034 Safety Concerns Associated With Pipe NA Boiling Water Reactor Breaks in the BWR Scram System GL 81-035 Safety Concerns Associated With Pipe NA Boiling Water Reactor Breaks in the BWR Scram System GL 81-036 Revised Schedule for Completion of TMI NA Info Action Plan Item II.D.1, Relief and Safety Valve Testing GL 81-037 ODYN Code Reanalysis Requirements NA Boiling Water Reactor GL 81-038 Storage of Low Level Radioactive Wastes NA Info at Power Reactor Sites GL 81-039 NRC Volume Reduction Policy NA Info GL 81-040 Qualifications of Reactor Operators NA Info GL 82-001 New Applications Survey NA Info GL 82-002 Commission Policy on Overtime NA Info GL 82-003 High Burnup MAPLHGR Limits NA Boiling Water Reactor GL 82-004 Use of INPO See-in Program NA Info GL 82-005 Post-TMI Requirements NA Item was applicable only to units with operating license at the time the item was issued.

GL 82-006 This GL was never issued. NA Page 53 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 82-007 Transmittal of NUREG-0909 Relative to NA Boiling Water Reactor the Ginna Tube Rupture GL 82-008 Transmittal of NUREG-0909 Relative to NA Info the Ginna Tube Rupture GL 82-009 Environmental Qualification of Safety NA Info Related Electrical Equipment GL 82-010 Post-TMI Requirements NA Item was applicable only to units with operating license at the time the item was issued.

GL 82-011 Transmittal of NUREG-0916 Relative to NA Info the Restart of R. E. Ginna Nuclear Power Plant GL 82-012 Nuclear Power Plant Staff Working Hours NA Info GL 82-013 Reactor Operator and Senior Reactor NA Info Operator Examinations GL 82-014 Submittal of Documents to the NRC NA Info GL 82-015 This GL was never issued. NA GL 82-016 NUREG-0737 Technical Specifications NA Item was applicable only to units with operating license at the time the item was issued.

GL 82-017 Inconsistency of Requirements Between NA Info 50.54(T) and 50.15 GL 82-018 Reactor Operator and Senior Reactor NA Info Operator Requalification Examinations GL 82-019 Submittal of Copies of Documentation to NA Info NRC - Copy Requirements for Emergency Plans and Physical Security Plans Page 54 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 82-020 Guidance for Implementing the Standard NA Info Review Plan Rule GL 82-021 Fire Protection Audits NA Info GL 82-022 Congressional Request for Information NA Item was applicable only to units with operating license at the Concerning Steam Generator Tube time the item was issued.

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

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

GL 82-027 Transmittal of NUREG-0763, Guidelines NA Boiling Water Reactor For Confirmatory In-Plant Tests of Safety-Relief Valve Discharge for BWR Plants GL 82-028 Inadequate Core Cooling Instrumentation O LICENSE CONDITION - Detectors for Inadequate core cooling System (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 82-029 This GL was never issued. NA GL 82-030 Filings Related to 10 CFR 50 Production NA Info and Utilization Facilities Page 55 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 82-031 This GL was never issued. NA GL 82-032 Draft Steam Generator Report (SAI) NA Item was applicable only to units with operating license at the time the item was issued.

GL 82-033 Supplement to NUREG-0737, CI Safety Parameter Display System (SPDS) / Requirements for Requirements for Emergency Response Emergency Response Capability - NRC reviewed in SSER5, Capability SSER6, and 18.2.2 of SSER15.

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

GL 82-034 This GL was never issued. NA GL 82-035 This GL was never issued. NA GL 82-036 This GL was never issued. NA GL 82-037 This GL was never issued. NA GL 82-038 Meeting to Discuss Developments for NA Info Operator Licensing Examinations GL 82-039 Problems With Submittals of Subsequent NA Info Information of CURT 73.21 For Licensing Reviews GL 83-001 Operator Licensing Examination Site Visit NA Info GL 83-002 NUREG-0737 Technical Specifications NA Boiling Water Reactor GL 83-003 This GL was never issued. NA GL 83-004 Regional Workshops Regarding NA Info Supplement 1 to NUREG-0737, Requirements For Emergency Response Capability Page 56 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 83-005 Safety Evaluation of "Emergency NA Boiling Water Reactor Procedure Guidelines, Revision 2,"

June 1982 GL 83-006 Certificates and Revised Format For NA Info Reactor Operator and Senior Reactor Operator Licenses GL 83-007 The Nuclear Waste Policy Act of 1982 NA Info GL 83-008 Modification of Vacuum Breakers on Mark NA Boiling Water Reactor I Containments GL 83-009 Review of Combustion Engineering NA Applies only to Combustion Engineering designed plants Owners' Group Emergency Procedures Guideline Program GL 83-010a Resolution of TMI Action Item II.K.3.5., NA Applies only to Combustion Engineering designed plants "Automatic Trip of Reactor Coolant Pumps" GL 83-010b Resolution of TMI Action Item II.K.3.5., NA Applies only to Combustion Engineering designed plants "Automatic Trip of Reactor Coolant Pumps" GL 83-010c Resolution of TMI Action Item II.K.3.5., CI TVA: letters dated January 5, 1984 and June 25, 1984 "Automatic Trip of Reactor Coolant Pumps" NRC: letter dated June 8, 1990.

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

[WAS "NOTE 3."]

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

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

Actions Page 57 of 95 * = See last page for status code definition.

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

GL 83-015 Implementation of Regulatory Guide NA Info 1.150, "Ultrasonic Testing of Reactor Vessel Welds During Preservice &

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

Amendments to Operating Licenses GL 83-020 Integrated Scheduling for Implementation NA Info of Plant Modifications GL 83-021 Clarification of Access Control NA Info Procedures for Law Enforcement Visits GL 83-022 Safety Evaluation of "Emergency NA Info Response Guidelines" GL 83-023 Safety Evaluation of "Emergency NA Applies only to Combustion Engineering designed plants Procedure Guidelines" Page 58 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 83-024 TMI Task Action Plan Item I.G.1, NA Boiling Water Reactor "Special Low Power Testing and Training," Recommendations for BWRs GL 83-025 This GL was never issued. NA GL 83-026 Clarification Of Surveillance NA Info Requirements For Diesel Fuel Impurity Level Tests GL 83-027 Surveillance Intervals in Standard NA Info Technical Specifications GL 83-028 "Required Actions Based on Generic C TVA: letters dated November 7, 1983 and Implications of Salem ATWS Events: December 4, 1987 1.2 - Post Trip Review Data and NRC: IR 50-390, 391/86-04 Information Capability GL 83-028 "Required Actions Based on Generic CI TVA: letters dated November 7, 1983 and August 24, 1990 Implications of Salem ATWS Events:

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

System Components)

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

[WAS "NOTE 3."]

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

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

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ITEM TITLE REV ADDITIONAL INFORMATION GL 83-028 "Required Actions Based on Generic S TVA: letters dated November 7, 1983, January 17, 1986 and Implications of Salem ATWS Events: November 1, 1993 02 3.1 - Post-Maintenance Testing NRC: letters dated December 10, 1985, October 27, 1986, and (Reactor Trip System Components) July 2, 1990; IR 390, 391/86-04 Unit 2 Action: Test and maintenance procedures and Technical Specifications will include post-maintenance operability testing of safety-related components of the reactor trip system.

REVISION 02 UPDATE:

Developmental Revision A of the Unit 2 TS (including the TS Bases) was submitted on March 4, 2009.

The Bases for TS Surveillance Requirement 3.0.1 states, in part, Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2.

GL 83-028 "Required Actions Based on Generic S TVA: letters dated November 7, 1983, January 17, 1986 and Implications of Salem ATWS Events: November 1, 1993 02 3.2 - Post-Maintenance Testing (All SR NRC: letters dated December 10, 1985, October 27, 1986, and Components) 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.

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

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

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ITEM TITLE REV ADDITIONAL INFORMATION GL 83-028 "Required Actions Based on Generic S TVA: letters dated November 7, 1983, February 10, 1986, and Implications of Salem ATWS Events: May 19, 1986 02 4.2 - Reactor Trip System Reliability NRC: letters dated July 26, 1985 and June 18, 1992; SSER 16 (Preventive Maintenance and Surveillance Program for Reactor --------------------

Trip Breakers)

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

REVISION 02 UPDATE:

Developmental Revision B of the Unit 2 TS was submitted on February 2, 2010.

Item 17. (Reactor Trip Breakers) of TS Table 3.3.1-1 states the requirement for the reactor trip breakers.

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

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

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

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

Attachment)

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

REVISION 02 UPDATE:

Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.

Item 18. (Reactor Trip Breaker Undervoltage and Shunt Trip Mechanisms) of TS Table 3.3.1-1 states the requirement for the shunt trip attachment.

GL 83-029 This GL was never issued. NA GL 83-030 Deletion of Standard Technical NA Info Specifications Surveillance Requirement 4.8.1.1.2.d.6 For Diesel Generator Testing Page 61 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 83-031 Safety Evaluation of "Abnormal Transient NA Applies only to Babcock and Wilcox designed plants Operating Guidelines" GL 83-032 NRC Staff Recommendations Regarding NA Info Operator Action for Reactor Trip and ATWS GL 83-033 NRC Positions on Certain Requirements NA Info of Appendix R to 10 CFR 50 GL 83-034 This GL was never issued. NA GL 83-035 Clarification of TMI Action Plan Item NA Info II.K.3.31 GL 83-036 NUREG-0737 Technical Specifications NA Boiling Water Reactor GL 83-037 NUREG-0737 Technical Specifications NA Item was applicable only to units with operating license at the time the item was issued.

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

GL 83-042 Clarification to GL 81-07 Regarding NA Info Response to NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants" GL 83-043 Reporting Requirements of 10 CFR 50, NA Info Sections 50.72 and 50.73, and Standard Technical Specifications Page 62 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 83-044 Availability of NUREG-1021, "Operator NA Info Licensing Examiner Standards GL 84-001 NRC Use Of The Terms "Important To NA Info Safety" and "Safety Related" GL 84-002 Notice of Meeting Regarding Facility NA Info Staffing GL 84-003 Availability of NUREG-0933, "A NA Info Prioritization of Generic Safety Issues" GL 84-004 Safety Evaluation of Westinghouse NA Info Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops GL 84-005 Change to NUREG-1021, "Operator NA Info Licensing Examiner Standards" GL 84-006 Operator and Senior Operator License NA Does not apply to power reactor.

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

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

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

Amendments GL 84-019 Availability of Supplement 1 to NA Info NUREG-0933, "A Prioritization of Generic Safety Issues" GL 84-020 Scheduling Guidance for Licensee NA Info Submittals of Reloads That Involve Unreviewed Safety Questions GL 84-021 Long Term Low Power Operation in NA Info Pressurized Water Reactors GL 84-022 This GL was never issued. NA GL 84-023 Reactor Vessel Water Level NA Boiling Water Reactor Instrumentation in BWRs GL 84-024 Certification of Compliance to CI See Special Program for Environmental Qualification.

10 CFR 50.49, Environmental Qualification of Electric Equipment Important To Safety For Nuclear Power Plants GL 85-001 Fire Protection Policy Steering Committee NA Only issued as draft Report Page 64 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 85-002 Recommended Actions Stemming From CI TVA responded to the GL on June 17, 1985.

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

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

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

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

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

Unit 2 Action: Implement modifications as required.

GL 85-013 Transmittal Of NUREG-1154 Regarding NA Info The Davis-Besse Loss Of Main And Auxiliary Feedwater Event Page 65 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 85-014 Commercial Storage At Power Reactor NA Item was applicable only to units with operating license at the Sites Of Low Level Radioactive Waste time the item was issued.

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

Qualification Of Electric Equipment Important To Safety At Nuclear Power Plants" GL 85-016 High Boron Concentrations NA Info GL 85-017 Availability Of Supplements 2 and 3 To NA Info NUREG-0933, "A Prioritization Of Generic Safety Issues" GL 85-018 Operator Licensing Examinations NA Info GL 85-019 Reporting Requirements On Primary NA Info Coolant Iodine Spikes GL 85-020 Resolution Of Generic Issue 69: High NA Applies only to Babcock and Wilcox designed plants Pressure Injection/Make-up Nozzle Cracking In Babcock And Wilcox Plants GL 85-021 This GL was never issued. NA GL 85-022 Potential For Loss Of Post-LOCA NA Info Recirculation Capability Due To Insulation Debris Blockage GL 86-001 Safety Concerns Associated With Pipe NA Boiling Water Reactor Breaks In The BWR Scram System GL 86-002 Technical Resolution of Generic Issue NA Boiling Water Reactor B Thermal Hydraulic Stability GL 86-003 Applications For License Amendments NA Info GL 86-004 Policy Statement On Engineering C TVA responded to GL 86-04 on May 29, 1986. TVA provides Expertise On Shift engineering expertise on shift in the form of a dedicated Shift 01 Technical Advisor (STA) or an STA qualified Senior Reactor Operator.

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ITEM TITLE REV ADDITIONAL INFORMATION GL 86-005 Implementation Of TMI Action Item NA Applies only to Babcock and Wilcox designed plants II.K.3.5, "Automatic Trip Of Reactor Coolant Pumps" GL 86-006 Implementation Of TMI Action Item NA Applies only to Combustion Engineering designed plants II.K.3.5, "Automatic Trip of Reactor Coolant Pumps" GL 86-007 Transmittal of NUREG-1190 Regarding NA Info The San Onofre Unit 1 Loss of Power and Water Hammer Event GL 86-008 Availability of Supplement 4 to NA Info NUREG-0933, "A Prioritization of Generic Safety Issues" GL 86-009 Technical Resolution of Generic Issue S N-1 Loop operation was addressed in original 1982 SER (4.4.7).

B-59, (N-1) Loop Operation in BWRs and PWRs 02 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.

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

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

From Conversion From The Use of Heu Fuel GL 86-013 Potential Inconsistency Between Plant NA Applies only to Babcock and Wilcox and Combustion Safety Analyses and Technical Engineering designed plants Specifications GL 86-014 Operator Licensing Examinations NA Info Page 67 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 86-015 Information Relating To Compliance With NA Info 10 CFR 50.49, "Environmental Qualification of Electric Equipment Important To Safety For Nuclear Power Plants" GL 86-016 Westinghouse ECCS Evaluation Models NA Info GL 86-017 Availability of NUREG-1169, "Technical NA Boiling Water Reactor Findings Related to Generic Issue C-8, BWR MSIC Leakage And Treatment Methods" GL 87-001 Public Availability Of The NRC Operator NA Info Licensing Examination Question Bank GL 87-002 Verification of Seismic Adequacy of NA Item was applicable only to units with operating license at the and Mechanical and Electrical Equipment in time the item was issued.

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

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

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

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

Checks GL 87-011 Relaxation in Arbitrary Intermediate Pipe NA Info Rupture Requirements Page 68 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 87-012 Loss of Residual Heat Removal While C This GL was superseded by GL 88-17.

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

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

GL 88-003 Resolution of GSI 93, Steam Binding of CI TVA: letter June 3, 1988. NRC letters dated Auxiliary Feedwater Pumps February 17, 1988 and July 20, 1988 NRC: 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 88-004 Distribution of Gems Irradiated in NA Does not apply to power reactor.

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

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

GL 88-006 Removal of Organization Charts from NA Info Technical Specification Administrative Control Requirements Page 69 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 88-007 Modified Enforcement Policy Relating to CI See Special Program for Environmental Qualification.

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

of Reactor Vessel Material and its Impact on Plant Operations 02 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.

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

Affecting Safety-Related Equipment 04 Unit 2 Action: Complete Unit 2 implementation.

REVISION 04 UPDATE:

The compressed air system is a common system at Watts Bar; therefore, the requirements for this GL have been satisfied for Unit 2.

Watts Bar revised the response in a letter dated July 14, 1995.

NRC letter dated July 27, 1995, stated that their conclusion as stated on July 26,1990, had not changed and that their effort was complete.

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ITEM TITLE REV ADDITIONAL INFORMATION GL 88-015 Electric Power Systems - Inadequate NA Info Control Over Design Process GL 88-016 Removal of Cycle-Specific Parameter NA Info Limits from Technical Specifications GL 88-017 Loss of Decay Heat Removal CI NRC acceptance letter dated March 8, 1995 (Unit 1).

Unit 2 Action: Implement modifications to provide RCS temperature, RV level and RHR system performance.

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

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

Accident Vulnerabilities 04 ----------------------------------------------------------------------------------------

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.

GL 89-001 Implementation of Programmatic and NA Info Procedural Controls for Radiological Effluent Technical Specifications GL 89-002 Actions to Improve the Detection of C GL 89-02 did not require a response.

Counterfeit and Fraudulently Marketed Products 01 WBN Unit 2 program for procurement and dedication of materials is based in part on and complies with the guidance of GL 89-02. The program is implemented through project procedures.

GL 89-003 Operator Licensing Examination Schedule NA Info GL 89-004 Guidelines on Developing Acceptable OV NRC reviewed in 3.9.6 of SSER14 (Unit 1).

Inservice Testing Programs Unit 2 Action: Submit an ASME Section XI Inservice Test Program for the first ten year interval six months before receiving an Operating License.

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ITEM TITLE REV ADDITIONAL INFORMATION GL 89-005 Pilot Testing of the Fundamentals NA Info Examination GL 89-006 Task Action Plan Item I.D.2 - Safety CI Safety Parameter Display System (SPDS) / Requirements for Parameter Display System - 10 CFR Emergency Response Capability - NRC reviewed in SSER5, 50.54(f) SSER6, and 18.2.2 of SSER15.

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

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

Unit 2 Actions: Prepare procedure and perform baseline inspections.

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

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

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

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

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

GL 89-014 Line-Item Improvements in Technical NA Info Specifications - Removal of 3.25 Limit on Extending Surveillance Intervals Page 72 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 89-015 Emergency Response Data System NA Info GL 89-016 Installation of a Hardened Wetwell Vent NA Boiling Water Reactor GL 89-017 Planned Administrative Changes to the NA Info NRC Operator Licensing Written Examination Process GL 89-018 Resolution of Unresolved Safety Issues NA Info A-17, "Systems Interactions in Nuclear Power Plants" GL 89-019 Request for Actions Related to Resolution CI TVA responded by letter dated March 22, 1990. NRC of Unresolved Safety Issue A-47, "Safety acceptance letter dated October 24, 1990, for both units.

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

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

GL 89-021 Request for Information Concerning S TVA responded to GL 89-21 with the status of USIs for both units Status of Implementation of Unresolved on November 29, 1989. NRC provided an assessment of WBN Safety Issue (USI) Requirements 02 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.

GL 89-022 Potential For Increased Roof Loads and C TVA: letter dated December 16, 1981 Plant Area Flood Runoff Depth At Licensed Nuclear Power Plants Due To -------------------

Recent Change In Probable Maximum Precipitation Criteria Developed by the Answer to informal question provided in TVA letter dated National Weather Service December 16, 1981, and subsequently included in FSAR. GL did not require a response. No further action required.

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ITEM TITLE REV ADDITIONAL INFORMATION GL 89-023 NRC Staff Responses to Questions NA Info Pertaining to Implementation of 10 CFR Part 26 GL 90-001 Request for Voluntary Participation in NA Info NRC Regulatory Impact Survey GL 90-002 Alternative Requirements for Fuel NA Info Assemblies in the Design Features Section of Technical Specifications GL 90-003 Relaxation of Staff Position in Generic NA Info Letter 83-28, Item 2.2 Part 2 "Vendor Interface for Safety-Related Components" GL 90-004 Request for Information on the Status of C TVA responded on June 23, 1990 Licensee Implementation of GSIs Resolved with Imposition of Requirements or CAs GL 90-005 Guidance for Performing Temporary NA Info Non-Code Repair of ASME Code Class 1, 2, and 3 Piping GL 90-006 Resolution of Generic Issues 70, "PORV S NRC letter dated January 9, 1991, accepted TVAs response for and Block Valve Reliability," and 94, both units.

"Additional LTOP Protection for PWRs" 02 Unit 2 Actions: 1) Revise operating instruction and surveillance procedure; and 2) Incorporate testing requirements in the Technical Specifications.

REVISION 02 UPDATE:

Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.

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

GL 90-007 Operator Licensing National Examination NA Info Schedule GL 90-008 Simulation Facility Exemptions NA Info GL 90-009 Alternative Requirements for Snubber NA Info Visual Inspection Intervals and Corrective Actions Page 74 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 91-001 Removal of the Schedule for the NA Info Withdrawal of Reactor Vessel Material Specimens from Technical Specifications GL 91-002 Reporting Mishaps Involving LLW Forms NA Item was applicable only to units with operating license at the Prepared for Disposal time the item was issued.

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

Supplies," Pursuant to 10 CFR 50.54(f)

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

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

49, Interlocks and LCOs for Class 1E Tie Breakers," Pursuant to 10 CFR 50.54 GL 91-012 Operator Licensing National Examination NA Info Schedule GL 91-013 Request for Information Related to NA Addressed to specific (non-TVA) plants.

Resolution of Generic Issue 130, Essential Service Water System Failures

@ Multi-Unit Sites Page 75 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 91-014 Emergency Telecommunications NA Info GL 91-015 Operating Experience Feedback Report, NA Info Solenoid-Operated Valve Problems at U.S. Reactors GL 91-016 Licensed Operators' and Other Nuclear NA Info Facility Personnel Fitness for Duty GL 91-017 Generic Safety Issue 29, "Bolting NA Info Degradation or Failure in Nuclear Power Plants" GL 91-018 Information to Licensees Regarding Two NA GL 91-18 has been superseded by RIS 2005-20.

NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability GL 91-019 Information to Addressees Regarding NA Info New Telephone Numbers for NRC Offices Located in One White Flint North GL 92-001 Reactor Vessel Structural Integrity C By letter dated May 11, 1994, for both units NRC confirmed TVA had provided the information requested in GL 92-01. NRC issued GL 92-01 revision 1, supplement 1 on May 19, 1995.

By letter dated July 26, 1996, NRC closed GL 92-01, Revision 1, Supplement 1 for both Watts Bar units.

GL 92-002 Resolution of Generic Issue 79, NA Info "Unanalyzed Reactor Vessel (PWR)

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

GL 92-005 NRC Workshop on the Systematic NA Info Assessment of Licensee Performance (SALP) Program GL 92-006 Operator Licensing National Examination NA Info Schedule GL 92-007 Office of Nuclear Reactor Regulation NA Info Reorganization Page 76 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 92-008 Thermo-Lag 330-1 Fire Barriers OV TVA configurations for Thermo-Lag 330-1 were reviewed in SSER18 and accepted in NRC letter dated January 6, 1998 (includes a supplemental SE).

Unit 2 Actions: 1) Review Watts Bar design and installation requirements for Thermolag 330-1 fire barrier system and evaluate the Thermolag currently installed in Unit 2.

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

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

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

Assemblies, 10 CFR 50.54(f)

Unit 2 Action: Implement modifications and testing.

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

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

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

Emergency Diesel Generators GL 94-002 Long-Term Solutions and Upgrade of NA Boiling Water Reactor Interim Operating Recommendations for Thermal-Hydraulic Instabilities in BWRs Page 77 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 94-003 IGSCC of Core Shrouds in BWRs NA Boiling Water Reactor GL 94-004 Voluntary Reporting of Additional NA Info Occupational Radiation Exposure Data GL 95-001 NRC Staff Technical Position on Fire NA Does not apply to power reactor.

Protection for Fuel Cycle Facilities GL 95-002 Use of NUMARC/EPRI Report TR- NA Info 102348, "Guideline on Licensing Digital Upgrades," in Determining the Acceptability of Performing Analog-to-Digital Replacements under 10 CFR 50.59 GL 95-003 Circumferential Cracking of Steam CI NRC acceptance letter dated May 16, 1997 (Unit 1) - Initial Generator Tubes response for Unit 2 on September 7, 2007. TVA responded to a 02 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.

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

Affected by Outside Diameter Stress 02 Corrosion Cracking ----------------------------------------------------------------------------------------

REVISION 02 UPDATE:

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

GL 95-006 Changes in the Operator Licensing NA Info Program Page 78 of 95 * = See last page for status code definition.

ITEM TITLE REV ADDITIONAL INFORMATION GL 95-007 Pressure Locking and Thermal Binding of CI Unit 1 SER for GL 95-07 dated Sept 15, 1999 Safety-Related Power-Operated Gate Valves 04 Unit 2 Action: Perform evaluation for pressure locking and thermal binding of safety related power-operated gate valves and take corrective actions for those valves identified as being susceptible.

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.

GL 95-008 10 CFR 50.54(p) Process for Changes to NA Info Security Plans Without Prior NRC Approval GL 95-009 Monitoring and Training of Shippers and NA Info Carriers of Radioactive Materials GL 95-010 Relocation of Selected Technical NA Info Specifications Requirements Related to Instrumentation GL 96-001 Testing of Safety-Related Circuits CI TVA responded for both units on April 18, 1996.

Unit 2 Action: Implement Recommendations.

GL 96-002 Reconsideration of Nuclear Power Plant NA Info Security Requirements Associated with an Internal Threat Page 79 of 95 * = See last page for status code definition.

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

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

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

Capability of Safety-Related Motor-Operated Valves Unit 2 Action: Implement the Joint Owners Group recommended GL 96-05 MOV PV program, as described in Topical Report No. OG-97-018, and begin testing during the first refueling outage after startup.

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

Design-Basis Accident Conditions 02 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.

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

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

Mechanism Nozzle and Other Vessel Closure Head Penetrations 04 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.

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

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ITEM TITLE REV ADDITIONAL INFORMATION GL 97-003 Annual Financial Update of Surety NA Does not apply to power reactor.

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

Cooling and Containment Heat Removal 02 Pumps Unit 2 Actions: Install new sump strainers, and perform other modification-related activities identical to Unit 1.

REVISION 02 UPDATE:

NRC issued the Safety Evaluation for Generic Letter 1997-004 on February 18, 2010.

GL 97-005 Steam Generator Tube Inspection CI NRC acceptance letter dated September 22, 1998 (Unit 1) -

Techniques Initial response for Unit 2 on September 7, 2007.

02 Unit 2 Action: Employ the same approach used on the original Unit 1 SGs. TVA responded to a request for additional information on December 17, 2007.

REVISION 02 UPDATE:

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

GL 97-006 Degradation of Steam Generator Internals CI NRC acceptance letter dated October 19, 1999 (Unit 1) - Initial response for Unit 2 on September 7, 2007. TVA responded to a 02 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.

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

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ITEM TITLE REV ADDITIONAL INFORMATION GL 98-002 Loss of Reactor Coolant Inventory and CI Initial response for Unit 2 on September 7, 2007.

Associated Potential for Loss of Emergency Mitigation Functions While in 03 Unit 2 Actions: 1) Review the ECCS designs to ensure they do a Shutdown Condition 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.

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

Year 2000 Readiness Programs GL 98-004 Potential for Degradation of the ECCS OV NRC closure letter dated November 24, 1999 (Unit 1). - Initial and the Containment Spray System After response for Unit 2 on September 7, 2007.

a LOCA Because of Construction and 02 Protective Coating Deficiencies and Unit 2 Actions: Install new sump strainers, and perform other Foreign Material in Containment modification-related activities identical to Unit 1.

REVISION 02 UPDATE:

NRC issued the Safety Evaluation for Generic Letter 1998-004 on February 1, 2010.

GL 98-005 Boiling Water Reactor Licensees Use of NA Boiling Water Reactor the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds GL 99-001 Recent Nuclear Material Safety and NA Info Safeguards Decision on Bundling Exempt Quantities Page 82 of 95 * = See last page for status code definition.

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

GL 03-001 Control Room Habitability S Initial response for Unit 2 on September 7, 2007 02 Unit 2 Action: Incorporate TSTF-448 into Technical Specifications.

REVISION 02 UPDATE:

NRC issued the Safety Evaluation for Generic Letter 2003-01 on February 1, 2010.

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).

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

02 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.

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

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

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ITEM TITLE REV ADDITIONAL INFORMATION GL 06-001 Steam Generator Tube Integrity and S Initial response for Unit 2 on September 7, 2007.

Associated Technical Specifications 02 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).

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

Risk and the Operability of Offsite Power 02 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.

GL 06-003 Potentially Nonconforming Hemyc and CI TVA does not rely on Hemyc or MT materials to protect electrical MT Fire Barrier Configurations and instrumentation cables or equipment that provide safe 02 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.

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

Cable Failures That Disable Accident Mitigation Systems or Cause Plant 02 Unit 2 Action: Complete testing of four additional cables.

Transients REVISION 02 UPDATE:

NRC issued the Safety Evaluation for Generic Letter 2007-001 on January 26, 2010.

REVISION 04 UPDATE:

NRC Inspection Report 391/2010-603 closed GL 2007-001.

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

Emergency Core Cooling, Decay Heat Removal, and Containment Spray 02 ----------------------------------------------------------------------------------------

Systems ----------------------------------------------------------------------------------------

REVISION 02 UPDATE:

Unit 2 Actions:

- TVA will provide a submittal within 45 days of completion of the engineering for the ECCS, RHR, and CSS systems.

- WBN Unit 2 will complete the required modifications and provide a submittal consistent with the information requested in the GL 90 days prior to fuel load.

NUREG- Shift Technical Advisor NA Not applicable to WBN per SSER16.

0737, I.A.1.1 NUREG- Shift Supervisor Responsibilities NA Not applicable to WBN per SSER16.
0737, I.A.1.2 NUREG- Shift Manning C Closed in SSER16.
0737, I.A.1.3 NUREG- Immediate Upgrade of RO and SRO C Closed in SSER16.

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

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

0737, Exams I.A.3.1 Page 85 of 95 * = See last page for status code definition.

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

I.B.1.2 Resolved for Unit 1 only in SSER8.

Unit 2 action: Implement the alternate ISEG that was approved for the rest of the TVA units including WBN Unit 1 by NRC on August 26, 1999. The function will be performed by the site engineering organizations.

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

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

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

0737, I.C.2 NUREG- Shift Supervisor Responsibility C Closed in SSER16.
0737, I.C.3 NUREG- Control Room Access C Closed in SSER16.
0737, I.C.4 NUREG- Feedback of Operating Experience C Closed in SSER16.
0737, I.C.5 NUREG- Verify Correct Performance of Operating C Closed in SSER16.

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

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

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

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

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

I.D.1 Unit 2 Actions: Complete the CRDR process. Perform rewiring in accordance with ECN 5982. Take advantage of the completed Human Engineering reviews to ensure appropriate configuration for Unit 2 control panels. See CRDR Special Program.

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

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

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ITEM TITLE REV ADDITIONAL INFORMATION NUREG- Training During Low-Power Testing C Closed in SSER16.

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

Unit 2 Action: Verify installation of reactor coolant vents.

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

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

NUREG- Post-Accident Sampling S NRC reviewed in 9.3.2 of SSER16. TVA submitted a TS 0737, improvement to eliminate requirements for the Post Accident II.B.3 02 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.

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

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

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

2) Reanalysis of fluid transient loads for pressurizer relief and safety valve supports and any required modifications;
3) Modifications to pressurizer safety valves, PORVs, PORV block valves and associated piping; and 4) Change motor operated block valves.

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

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

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

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

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

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

Unit 2 Action: Implement procedures and testing.

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

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ITEM TITLE REV ADDITIONAL INFORMATION NUREG- Containment Isolation Dependability S TVA: letters dated October 29, 1981, and 0737, February 25, 1985 II.E.4.2 02 NRC: letters dated March 29, 1985, July 12, 1990 and October 31, 1995; SSER 16.

OUTSTANDING ISSUE for NRC to complete review of information provided by TVA to address Containment Purging During Normal Plant Operation LICENSE CONDITION - Containment isolation dependability In the original 1982 SER, NRC concluded that WBN met all the requirements of NUREG-0737, item II.E.4.2 except subsection (6) concerning containment purging during normal operation. In SSER3, the outstanding issue was closed and the LICENSE CONDITION was left open.

NRC completed the review and issued a Technical Evaluation Report for both units on July 12, 1990. NRC concluded that the isolation valves can close against the buildup of pressure in the event of a design basis accident if the lower containment isolation valves are physically blocked to an opening angle of 50 degrees or less. (SSER5)

Unit 2 Action: Reflect valve opening restriction in the Technical Specifications.

REVISION 02 UPDATE:

Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.

TS Surveillance Requirement 3.6.3.7 requires verification that the valves are "blocked to restrict the valve from opening

> 50 degrees."

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

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

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

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

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

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

Unit 2 Action: Install high range in-containment monitor for Unit 2.

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

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

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

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

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

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

NUREG- Instrumentation For Detection of O LICENSE CONDITION - Detectors for Inadequate core cooling 0737, Inadequate Core-Cooling (II.F.2)

II.F.2 In the original SER, the review of the ICC instrumentation was incomplete. The January 24, 1992, letter superseded the previous responses on this issue. TVA letter for Units 1 and 2 dated January 24, 1992, committed to install Westinghouse ICCM-86 and associated hardware. NRC completed the review for Units 1 and 2 in SSER10. For Unit 2 due to obsolescence of the ICCM-86 system, TVA intends to install the Westinghouse Common Q Post-Accident Monitoring System.

Unit 2 Action: Install Westinghouse Common Q PAM system.

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

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

NUREG- Review ESF Valves C NRC: letter dated March 29, 1985; SSER 16

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

II.K.1.10 NUREG- Trip Per Low-Level B/S C NRC: letter dated March 29, 1985; SSER 16

0737, II.K.1.17 NUREG- Effect of High Pressure Injection for Small C LICENSE CONDITION - Effect of high pressure injection for 0737, Break LOCA With No Auxiliary Feedwater small break LOCA with no auxiliary feedwater II.K.2.13 (NUREG-0737, II.K.2.13)

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

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ITEM TITLE REV ADDITIONAL INFORMATION NUREG- Voiding in the Reactor Coolant System C LICENSE CONDITION - Voiding in the reactor coolant system 0737, (NUREG-0737, II.K.2.17)

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

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

NUREG- Reporting SV/RV Failures/Challenges S (Action from GL 82-16) - NRC reviewed in Appendix EE of 0737, SSER16.

II.K.3.3 02 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.

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

Unit 2 Action: Implement modifications as required.

NUREG- PID Controller CI Reviewed in original 1982 SER.

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

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ITEM TITLE REV ADDITIONAL INFORMATION NUREG- Anticipatory Trip at High Power S NRC: letter dated October 31, 1995; SSER 16

0737, II.K.3.10 02 --------------------

Unit 2 Action: Unit 2 Technical Specifications and surveillance procedures will address this issue.

REVISION 02 UPDATE:

Developmental Revision A of the Unit 2 Technical Specifications (TS) was submitted on March 04, 2009.

Items 14.a. (Turbine Trip - Low Fluid Oil Pressure) and 14.b.

(Turbine Trip - Turbine Stop Valve Closure) of TS Table 3.3.1-1 are the trips of interest. The table and the Bases for these items state that below the P-9 setpoint, these trips do not actuate a reactor trip.

Per item 16.d. (Power Range Neutron Flux, P-9) of TS Table 3.3.1-1, the Nominal Trip Setpoint for P-9 is 50% RTP and the Allowable Value is < 52.4% RTP.

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

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

II.K.3.17 In the original 1982 SER, the NRC accepted TVAs commitment to develop and implement a plan to collect emergency core cooling system outage information. In SSER3, the staff accepted a revised commitment from an October 28, 1983, letter to participate in the nuclear power reliability data system and comply with the requirements of 10 CFR 50.73 NUREG- Power On Pump Seals CI NRC reviewed and closed in IR 390/84-35 based on Diesel 0737, Generator (DG) power to pump sealing cooling system.

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

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

0737, II.K.3.30 NRC: letters dated March 29, 1985, and July 24, 1986; SSER 16 The staff determined in SSER4 that their review of Items 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.

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ITEM TITLE REV ADDITIONAL INFORMATION NUREG- Plant Specific Analysis CI The staff determined in SSER4 that their review of Items II.K.3.30 0737, and II.K.3.31 did not have to be completed to support the II.K.3.31 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- Emergency Preparedness, Short Term C LICENSE CONDITION - Emergency Preparedness (NUREG-0737, 0737, III.A.1, III.A.2)

III.A.1.1 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- Upgrade Emergency Support Facilities C LICENSE CONDITION - Emergency Preparedness (NUREG-0737, 0737, III.A.1, III.A.2)

III.A.1.2 The NRC review of Emergency Preparedness in SSER13 superseded the review in the original 1982 SER. In SSER13, the staff concluded that the WBN Radiological Emergency Plan (REP) provided an adequate planning basis for an acceptable state of onsite emergency preparedness, and the LICENSE CONDITION was deleted. The NRC completed the review of the REP in SSER20.

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

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

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ITEM TITLE REV ADDITIONAL INFORMATION NUREG- Primary Coolant Outside Containment S Resolved for Unit 1 only in SSER10; reviewed in Appendix EE of 0737, SSER16.

III.D.1.1 02 Unit 2 Actions: Include the waste gas disposal system in the leakage reduction program and incorporate in Unit 2 Technical Specifications.

REVISION 02 UPDATE:

Developmental Revision B of the Unit 2 Technical Specifications (TS) was submitted on February 2, 2010.

TS 5.7.2.4 is the Primary Coolant Sources Outside Containment program. This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. This program includes the "Waste Gas" system.

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

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

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

0737, III.D.3.4 NRC: SSER 16 NRC reviewed in SER and in Appendix EE of SSER16.

Unit 2 Action: Complete with CRDR completion.

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

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

CT: CLOSED/TECHNICAL SPECIFICATIONS: Item has been approved either for both units at WBN or explicitly for WBN Unit 2; however, a change to the original approval requires submittal of the Technical Specifications and staff review.

NA: NOT APPLICABLE: Justification as to why a section / subsection is not applicable is provided in the ADDITIONAL INFORMATION column.

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

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

OV: 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.

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Enclosure 4 Generic Communications - Revision 4 Changes

GENERIC COMMUNICATIONS: REVISION 4 CHANGES ITEM TITLE REV ADDITIONAL INFORMATION B 83-006 Nonconforming Material Supplied by CI TVA: letter dated February 2, 1984 Tube-Line Facilities 04 NRC: IR 390/391 84-03; NUREG/CR 4934 NRC SER for both units dated September 23, 1991, provided an alternate acceptance for fittings supplied by Tube-Line.

Unit 2 Action: Implement as necessary.

REVISION 04 UPDATE:

NRC Inspection Report Nos. 50-390/90-02 and 50-391/90-02 found the proposed alternative to ASME code paragraph NA-3451 (a) to be acceptable. It noted that TVA must revise the FSAR to document this deviation from ASME Section III requirements.

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

Final action was to incorporate the exemption in the Unit 2 FSAR.

This exemption is documented in Unit 2 FSAR Section 3.2 in paragraph 3.2.3.2 and Table 3.2-2a as explained in Note 4. of the table.

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

04 Unit 2 Action: Issue Emergency Operating Procedure.

REVISION 02 UPDATE:

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

REVISION 03 UPDATE:

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

REVISION 04 UPDATE:

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

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

04 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.

B 01-001 Circumferential Cracking of Reactor CI NRC acceptance letter dated November 20, 2001 (Unit 1) -

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

Nozzles 04 Unit 2 Action: Perform baseline inspection.

REVISION 02 UPDATE:

Unit 2 Action: 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.

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ITEM TITLE REV ADDITIONAL INFORMATION B 02-001 RPV Head Degradation and Reactor CI NRC review of Unit 1's 15 day response in letter dated Coolant Pressure Boundary Integrity May 20, 2002 - Initial response for Unit 2 on 04 September 7, 2007.

Unit 2 Action: Perform baseline inspection.

REVISION 02 UPDATE:

Unit 2 Action: 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.

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

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

04 Unit 2 Action: Perform baseline inspection.

REVISION 02 UPDATE:

Unit 2 Action: 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.

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

Used in the Fabrication of Pressurizer Penetrations and Steam Space Piping 04 Unit 2 Actions: Provide details of pressurizer and penetrations Connections at PWRs and apply Material Stress Improvement Process.

REVISION 02 UPDATE:

TVA provided details of the pressurizer and penetrations on September 29, 2008. This letter committed to:

Prior to placing the pressurizer in service, TVA will apply the Material Stress Improvement Process (MSIP) to the Pressurizer Power Operated Relief Valve connections, the safety relief valve connections, the spray line nozzle and surge line nozzle connections.

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.

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

Affecting Safety-Related Equipment 04 Unit 2 Action: Complete Unit 2 implementation.

REVISION 04 UPDATE:

The compressed air system is a common system at Watts Bar; therefore, the requirements for this GL have been satisfied for Unit 2.

Watts Bar revised the response in a letter dated July 14, 1995.

NRC letter dated July 27, 1995, stated that their conclusion as stated on July 26,1990, had not changed and that their effort was complete.

GL 88-020 Individual Plant Examination for Severe S Unit 2 Action: Complete evaluation for Unit 2.

Accident Vulnerabilities 04 ----------------------------------------------------------------------------------------

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.

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ITEM TITLE REV ADDITIONAL INFORMATION GL 95-007 Pressure Locking and Thermal Binding of CI Unit 1 SER for GL 95-07 dated Sept 15, 1999 Safety-Related Power-Operated Gate Valves 04 Unit 2 Action: Perform evaluation for pressure locking and thermal binding of safety related power-operated gate valves and take corrective actions for those valves identified as being susceptible.

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.

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

Mechanism Nozzle and Other Vessel Closure Head Penetrations 04 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.

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

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

CT: CLOSED/TECHNICAL SPECIFICATIONS: Item has been approved either for both units at WBN or explicitly for WBN Unit 2; however, a change to the original approval requires submittal of the Technical Specifications and staff review.

NA: NOT APPLICABLE: Justification as to why a section / subsection is not applicable is provided in the ADDITIONAL INFORMATION column.

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

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

OV: 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.

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