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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217M4461999-10-20020 October 1999 Forwards Rev 8 to Sequoyah Nuclear Plant Physical Security/ Contingency Plan, IAW 10CFR50.54(p).Encl Withheld,Per 10CFR73.21 ML20217J4151999-10-15015 October 1999 Forwards Request for Addl Info Re Util 990624 Application for Amend of TSs That Would Revise TS for Weighing of Ice Condenser Ice Baskets 05000327/LER-1999-002, Forwards LER 99-002-00 Re Start of Units 1 & 2 EDGs as Result of Cable Being Damaged During Installation of Thermo- Lag for Kaowool Upgrade Project1999-10-15015 October 1999 Forwards LER 99-002-00 Re Start of Units 1 & 2 EDGs as Result of Cable Being Damaged During Installation of Thermo- Lag for Kaowool Upgrade Project ML20217G1141999-10-0707 October 1999 Responds to from P Salas,Providing Response to NRC Risk Determination Associated with 990630 Flooding Event at Sequoyah Facility.Meeting to Discuss Risk Determination Issues Scheduled for 991021 in Atlanta,Ga ML20217B2981999-10-0606 October 1999 Discusses Closeout of GL 92-01,rev 1,suppl 1, Reactor Vessel Integrity, for Sequoyah Nuclear Plant,Units 1 & 2. NRC Also Hereby Solicits Any Written Comments That TVA May Have on Current Rvid Data by 991101 ML20217B8431999-10-0505 October 1999 Requests NRC Review & Approval of ASME Code Relief Requests That Were Identified in Plant Second 10-yr ISI Interval for Both Units.Encl 3 Provides Util Procedure for Calculation of ASME Code Coverage for Section XI Nondestructive Exams IR 05000327/19990041999-10-0101 October 1999 Ack Receipt of Providing Comments on Insp Repts 50-327/99-04 & 50-328/99-04.NRC Considered Comments for Apparent Violation Involving 10CFR50.59 Issue ML20217C7101999-10-0101 October 1999 Forwards Response to NRC 990910 RAI Re Sequoyah Nuclear Plant,Units 1 & 2 URI 50-327/98-04-02 & 50-328/98-04-02 Re Ice Weight Representative Sample ML20212J5981999-10-0101 October 1999 Forwards SE Accepting Request for Relief from ASME Boiler & Pressure Vessel Code,Section Xi,Requirements for Certain Inservice Insp at Plnat,Unit 1 ML20212M1081999-09-29029 September 1999 Confirms Intent to Meet with Utils on 991025 in Atlanta,Ga to Discuss Pilot Plants,Shearon Harris & Sequoyah Any Observations & Lessons Learned & Recommendations Re Implementation of Pilot Program ML20217A9451999-09-27027 September 1999 Forwards Insp Repts 50-327/99-05 & 50-328/99-05 on 990718- 0828.One Violation Identified & Being Treated as Non-Cited Violation ML20216J9351999-09-27027 September 1999 Responds to NRC Re Violations Noted in Insp Repts 50-327/99-04 & 50-328/99-04.Corrective Actions:Risk Determination Evaluation Was Performed & Licensee Concluded That Event Is in Green Regulatory Response Band ML20212F0751999-09-23023 September 1999 Forwards SER Granting Util 981021 Request for Relief from ASME Code,Section XI Requirements from Certain Inservice Insp at Sequoyah Nuclear Power Plant,Units 1 & 2 Pursuant to 10CFR50.55a(a)(3)(ii) ML20212F4501999-09-23023 September 1999 Forwards Amends 246 & 237 to Licenses DPR-77 & DPR-79, Respectively & Ser.Amends Approve Request to Revise TSs to Allow Use of Fully Qualified & Tested Spare Inverter in Place of Any of Eight Required Inverters ML20212M1911999-09-21021 September 1999 Discusses Exercise of Enforcement Discretion Re Apparent Violation Noted in Insp Repts 50-327/99-04 & 50-328/99-04 Associated with Implementation of Procedural Changes Which Resulted in Three Containment Penetrations Being Left Open ML20211Q0311999-09-10010 September 1999 Requests Written Documentation from TVA to Provide Technical Assistance to Region II Re TS Compliance & Ice Condenser Maint Practices at Plant ML20216F5441999-09-0707 September 1999 Provides Results of Risk Evaluation of 990630,flooding Event at Sequoyah 1 & 2 Reactor Facilities.Event Was Documented in Insp Rept 50-327/99-04 & 50-328/99-04 & Transmitted in Ltr, ML20211N5681999-09-0101 September 1999 Submits Clarification of Two Issues Raised in Insp Repts 50-327/99-04 & 50-328/99-04,dtd 990813,which Was First Insp Rept Issued for Plant Under NRC Power Reactor Oversight Process Pilot Plant Study ML20211G5881999-08-27027 August 1999 Submits Summary of 990820 Management Meeting Re Plant Performance.List of Attendees & Matl Used in Presentation Enclosed ML20211F8891999-08-25025 August 1999 Forwards Sequoyah Nuclear Plant Unit 1 Cycle 9 Refueling Outage, Re Completed SG Activities,Per TSs 4.4.5.5.b & 4.4.5.5.c ML20211A1851999-08-16016 August 1999 Forwards Proprietary TR WCAP-15128 & non-proprietary Rept WCAP-15129 for NRC Review.Repts Are Provided in Advance of TS Change That Is Being Prepared to Support Cycle 10 Rfo. Proprietary TR Withheld,Per 10CFR2.790 ML20210V1471999-08-13013 August 1999 Forwards Insp Repts 50-327/99-04 & 50-328/99-04 on 990601- 0717.One Potentially Safety Significant Issue Identified.On 990630,inadequate Performance of Storm Drain Sys Caused Water from Heavy Rainfall to Backup & Flood Turbine Bldg ML20211A1921999-08-12012 August 1999 Requests Proprietary TR WCAP-15128, Depth-Based SG Tube Repair Criteria for Axial PWSCC at Dented TSP Intersections, Be Withheld from Public Disclosure Per 10CFR2.790 ML20210Q5011999-08-0505 August 1999 Informs That NRC Plans to Administer Gfes of Written Operator Licensing Exam on 991006 at Sequoyah Nuclear Plant. Sample Registration Ltr Encl ML20210L4291999-08-0202 August 1999 Forwards Sequoyah Nuclear Plant Unit 2 Cycle 9 12-Month SG Insp Rept & SG-99-07-009, Sequoyah Unit-2 Cycle 10 Voltage-Based Repair Criteria 90-Day Rept. Repts Submitted IAW TS 4.4.5.5.b & TS 4.4.5.5.c ML20210L1611999-07-30030 July 1999 Forwards Request for Relief RV-4 Re ASME Class 1,2 & 3 Prvs, Per First ten-year Inservice Test Time Interval.Review & Approval of RV-4 Is Requested to Support Unit 1 Cycle 10 Refueling Outage,Scheduled to Start 000213 ML20210G5301999-07-28028 July 1999 Forwards Sequoyah Nuclear Plant Unit 2 ISI Summary Rept That Contains Historical Record of Repairs,Replacement & ISI & Augmented Examinations That Were Performed on ASME Code Class 1 & 2 Components from 971104-990511 ML20211B9661999-07-26026 July 1999 Informs That Sequoyah Nuclear Plant Sewage Treatment Plant, NPDES 0026450 Outfall 112,is in Standby Status.Flow Has Been Diverted from Sys Since Jan 1998 ML20210B2521999-07-14014 July 1999 Confirms 990712 Telcon Between J Smith of Licensee Staff & M Shannon of NRC Re semi-annual Mgt Meeting Schedule for 990820 in Atlanta,Ga to Discuss Recent Sequoyah Nuclear Plant Performance ML20210J1091999-07-10010 July 1999 Submits Suggestions & Concerns Re Y2K & Nuclear Power Plants ML20196K0381999-06-30030 June 1999 Provides Written Confirmation of Completed Commitment for Final Implementation of Thermo-Lag 330-1 Fire Barrier Corrective Actions at Snp,Per GL 92-08 ML20209E4071999-06-30030 June 1999 Forwards Insp Repts 50-327/99-03 & 50-328/99-03 on 990328- 0531.Violations Being Treated as Noncited Violations ML20196J8261999-06-28028 June 1999 Forwards Safety Evaluation Authorizing Request for Relief from ASME Boiler & Pressure Vessel Code,Section XI Requirements for Certain Inservice Inspections at Sequoyah Nuclear Plant,Units 1 & 2 ML20196G7881999-06-22022 June 1999 Informs NRC of Changes That Util Incorporated Into TS Bases Sections & Trm.Encl Provides Revised TS Bases Pages & TRM Affected by Listed Revs ML20196G1801999-06-21021 June 1999 Requests Termination of SRO License SOP-20751-1,for Lf Hardin,Effective 990611.Subject Individual Resigned from Position at TVA ML20195G1821999-06-0808 June 1999 Requests NRC Review & Approval of ASME Code Relief for ISI Program.Encl 1 Provides Relief Request 1-ISI-14 That Includes Two Attachments.Encl 2 Provides Copy of Related ASME Code Page ML20195E9521999-06-0707 June 1999 Requests Relief from Specific Requirements of ASME Section Xi,Subsection IWE of 1992 Edition,1992 Addenda.Util Has Determined That Proposed Alternatives Would Provide Acceptable Level of Quality & Safety ML20195E9311999-05-28028 May 1999 Informs of Planned Insp Activities for Licensee to Have Opportunity to Prepare for Insps & Provide NRC with Feedback on Any Planned Insps Which May Conflict with Plant Activities ML20195B3631999-05-21021 May 1999 Requests Termination of SRO License for Tj Van Huis,Per 10CFR50.74(a).TJ Van Huis Retired from Util,Effective 990514 ML20207A5721999-05-20020 May 1999 Forwards Correction to Previously Issued Amend 163 to License DPR-79 Re SR 4.1.1.1.1.d Inadvertently Omitted from Pp 3/4 1-1 of Unit 2 TS ML20206Q8791999-05-13013 May 1999 Forwards L36 9990415 802, COLR for Sequoyah Nuclear Plant Unit 2,Cycle 10, IAW Plant TS 6.9.1.14.c 05000327/LER-1999-001, Forwards LER 99-001-00 Re Condition That Resulted in Granting of Enforcement Discretion,Per Failure of Centrifugal Charging Pump.Condition Being Reported IAW 10CFR50.73(a)(2)(i)(B) & (a)(2)(iv)1999-05-11011 May 1999 Forwards LER 99-001-00 Re Condition That Resulted in Granting of Enforcement Discretion,Per Failure of Centrifugal Charging Pump.Condition Being Reported IAW 10CFR50.73(a)(2)(i)(B) & (a)(2)(iv) ML20206M9341999-05-10010 May 1999 Forwards Rept of SG Tube Plugging During Unit 2 Cycle 9 Refueling Outage,As Required by TS 4.4.5.5.a.ISI of Unit 2 SG Tubes Was Completed on 990503 ML20206K6271999-05-0606 May 1999 Requests Termination of SRO License for MR Taggart,License SOP-21336 Due to Resignation on 990430 ML20206J2061999-05-0404 May 1999 Requests Relief from Specified ISI Requirements in Section XI of ASME B&PV Code.Tva Requests Approval to Use Wire Type Penetrameters in Lieu of Plaque Type Penetrameters for Performing Radiographic Insps.Specific Relief Request,Encl ML20209J0391999-04-27027 April 1999 Forwards Annual Radioactive Effluent Release Rept, Radiological Impact Assessment Rept & Rev 41 to ODCM, for Period of Jan-Dec 1998 ML20206C6541999-04-23023 April 1999 Forwards Response to NRC 990127 RAI Re GL 96-05 for Sequoyah Nuclear Plant,Units 1 & 2 ML20206C0841999-04-23023 April 1999 Forwards Insp Repts 50-327/99-02 & 50-328/99-02 on 990214-0327.No Violations Noted ML20206B9591999-04-20020 April 1999 Responds to 990417 Request That NRC Exercise Discretion Not to Enforce Compliance with Actions Required in Unit 1 TS 3.1.2.2,3.1.2.4 & 3.5.2 & Documents 990417 Telephone Conversation When NRC Orally Issued NOED ML20205S5891999-04-17017 April 1999 Documents Request for Discretionary Enforcement for Unit 1 TS LCOs 3.1.2.2,3.1.2.4 & 3.5.2 to Support Completion of Repairs & Testing for 1B-B Centrifugal Charging Pump (CCP) 1999-09-07
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217M4461999-10-20020 October 1999 Forwards Rev 8 to Sequoyah Nuclear Plant Physical Security/ Contingency Plan, IAW 10CFR50.54(p).Encl Withheld,Per 10CFR73.21 05000327/LER-1999-002, Forwards LER 99-002-00 Re Start of Units 1 & 2 EDGs as Result of Cable Being Damaged During Installation of Thermo- Lag for Kaowool Upgrade Project1999-10-15015 October 1999 Forwards LER 99-002-00 Re Start of Units 1 & 2 EDGs as Result of Cable Being Damaged During Installation of Thermo- Lag for Kaowool Upgrade Project ML20217B8431999-10-0505 October 1999 Requests NRC Review & Approval of ASME Code Relief Requests That Were Identified in Plant Second 10-yr ISI Interval for Both Units.Encl 3 Provides Util Procedure for Calculation of ASME Code Coverage for Section XI Nondestructive Exams ML20217C7101999-10-0101 October 1999 Forwards Response to NRC 990910 RAI Re Sequoyah Nuclear Plant,Units 1 & 2 URI 50-327/98-04-02 & 50-328/98-04-02 Re Ice Weight Representative Sample ML20216J9351999-09-27027 September 1999 Responds to NRC Re Violations Noted in Insp Repts 50-327/99-04 & 50-328/99-04.Corrective Actions:Risk Determination Evaluation Was Performed & Licensee Concluded That Event Is in Green Regulatory Response Band ML20211N5681999-09-0101 September 1999 Submits Clarification of Two Issues Raised in Insp Repts 50-327/99-04 & 50-328/99-04,dtd 990813,which Was First Insp Rept Issued for Plant Under NRC Power Reactor Oversight Process Pilot Plant Study ML20211F8891999-08-25025 August 1999 Forwards Sequoyah Nuclear Plant Unit 1 Cycle 9 Refueling Outage, Re Completed SG Activities,Per TSs 4.4.5.5.b & 4.4.5.5.c ML20211A1851999-08-16016 August 1999 Forwards Proprietary TR WCAP-15128 & non-proprietary Rept WCAP-15129 for NRC Review.Repts Are Provided in Advance of TS Change That Is Being Prepared to Support Cycle 10 Rfo. Proprietary TR Withheld,Per 10CFR2.790 ML20211A1921999-08-12012 August 1999 Requests Proprietary TR WCAP-15128, Depth-Based SG Tube Repair Criteria for Axial PWSCC at Dented TSP Intersections, Be Withheld from Public Disclosure Per 10CFR2.790 ML20210L4291999-08-0202 August 1999 Forwards Sequoyah Nuclear Plant Unit 2 Cycle 9 12-Month SG Insp Rept & SG-99-07-009, Sequoyah Unit-2 Cycle 10 Voltage-Based Repair Criteria 90-Day Rept. Repts Submitted IAW TS 4.4.5.5.b & TS 4.4.5.5.c ML20210L1611999-07-30030 July 1999 Forwards Request for Relief RV-4 Re ASME Class 1,2 & 3 Prvs, Per First ten-year Inservice Test Time Interval.Review & Approval of RV-4 Is Requested to Support Unit 1 Cycle 10 Refueling Outage,Scheduled to Start 000213 ML20210G5301999-07-28028 July 1999 Forwards Sequoyah Nuclear Plant Unit 2 ISI Summary Rept That Contains Historical Record of Repairs,Replacement & ISI & Augmented Examinations That Were Performed on ASME Code Class 1 & 2 Components from 971104-990511 ML20210J1091999-07-10010 July 1999 Submits Suggestions & Concerns Re Y2K & Nuclear Power Plants ML20196K0381999-06-30030 June 1999 Provides Written Confirmation of Completed Commitment for Final Implementation of Thermo-Lag 330-1 Fire Barrier Corrective Actions at Snp,Per GL 92-08 ML20196G7881999-06-22022 June 1999 Informs NRC of Changes That Util Incorporated Into TS Bases Sections & Trm.Encl Provides Revised TS Bases Pages & TRM Affected by Listed Revs ML20196G1801999-06-21021 June 1999 Requests Termination of SRO License SOP-20751-1,for Lf Hardin,Effective 990611.Subject Individual Resigned from Position at TVA ML20195G1821999-06-0808 June 1999 Requests NRC Review & Approval of ASME Code Relief for ISI Program.Encl 1 Provides Relief Request 1-ISI-14 That Includes Two Attachments.Encl 2 Provides Copy of Related ASME Code Page ML20195E9521999-06-0707 June 1999 Requests Relief from Specific Requirements of ASME Section Xi,Subsection IWE of 1992 Edition,1992 Addenda.Util Has Determined That Proposed Alternatives Would Provide Acceptable Level of Quality & Safety ML20195B3631999-05-21021 May 1999 Requests Termination of SRO License for Tj Van Huis,Per 10CFR50.74(a).TJ Van Huis Retired from Util,Effective 990514 ML20206Q8791999-05-13013 May 1999 Forwards L36 9990415 802, COLR for Sequoyah Nuclear Plant Unit 2,Cycle 10, IAW Plant TS 6.9.1.14.c 05000327/LER-1999-001, Forwards LER 99-001-00 Re Condition That Resulted in Granting of Enforcement Discretion,Per Failure of Centrifugal Charging Pump.Condition Being Reported IAW 10CFR50.73(a)(2)(i)(B) & (a)(2)(iv)1999-05-11011 May 1999 Forwards LER 99-001-00 Re Condition That Resulted in Granting of Enforcement Discretion,Per Failure of Centrifugal Charging Pump.Condition Being Reported IAW 10CFR50.73(a)(2)(i)(B) & (a)(2)(iv) ML20206M9341999-05-10010 May 1999 Forwards Rept of SG Tube Plugging During Unit 2 Cycle 9 Refueling Outage,As Required by TS 4.4.5.5.a.ISI of Unit 2 SG Tubes Was Completed on 990503 ML20206K6271999-05-0606 May 1999 Requests Termination of SRO License for MR Taggart,License SOP-21336 Due to Resignation on 990430 ML20206J2061999-05-0404 May 1999 Requests Relief from Specified ISI Requirements in Section XI of ASME B&PV Code.Tva Requests Approval to Use Wire Type Penetrameters in Lieu of Plaque Type Penetrameters for Performing Radiographic Insps.Specific Relief Request,Encl ML20209J0391999-04-27027 April 1999 Forwards Annual Radioactive Effluent Release Rept, Radiological Impact Assessment Rept & Rev 41 to ODCM, for Period of Jan-Dec 1998 ML20206C6541999-04-23023 April 1999 Forwards Response to NRC 990127 RAI Re GL 96-05 for Sequoyah Nuclear Plant,Units 1 & 2 ML20205S5891999-04-17017 April 1999 Documents Request for Discretionary Enforcement for Unit 1 TS LCOs 3.1.2.2,3.1.2.4 & 3.5.2 to Support Completion of Repairs & Testing for 1B-B Centrifugal Charging Pump (CCP) ML20205B1091999-03-19019 March 1999 Submits Response to NRC Questions Concerning Lead Test Assembly Matl History,Per Request ML20204H0161999-03-19019 March 1999 Resubmits Util 990302 Response to Violations Noted in Insp Repts 50-327/98-11 & 50-328/98-11.Corrective Actions:Lessons Learned from Event Have Been Provided to Operating Crews ML20204E8251999-03-0505 March 1999 Forwards Sequoyah Nuclear Plant,Four Yr Simulator Test Rept for Period Ending 990321, in Accordance with Requirements of 10CFR55.45 ML20207E6851999-03-0202 March 1999 Responds to NRC Re Violations Noted in Insp Repts 50-327/98-11 & 50-328/98-11.Corrective Actions:Lessons Learned from Event Have Been Provided to Operating Crews ML20207J1171999-01-29029 January 1999 Forwards Copy of Final Exercise Rept for Full Participation Ingestion Pathway Exercise of Offsite Radiological Emergency Response Plans site-specific to Sequoyah NPP ML20202A7141999-01-20020 January 1999 Provides Request for Relief for Delaying Repair on Section of ASME Code Class 3 Piping within Essential Raw Cooling Water Sys ML20198S7141998-12-29029 December 1998 Forwards Cycle 10 Voltage-Based Repair Criteria 90-Day Rept, Per GL 95-05.Rept Is Submitted IAW License Condition 2.C.(9)(d) 05000327/LER-1998-004, Forwards LER 98-004-00,providing Details Concerning Inability to Complete Surveillance within Required Time Interval1998-12-21021 December 1998 Forwards LER 98-004-00,providing Details Concerning Inability to Complete Surveillance within Required Time Interval ML20198D5471998-12-14014 December 1998 Requests That License OP-20313-2 for Je Wright,Be Terminated IAW 10CFR50.74(a).Individual Retiring ML20197J5541998-12-10010 December 1998 Forwards Unit 1 Cycle 9 90-Day ISI Summary Rept IAW IWA-6220 & IWA-6230 of ASME Code,Section Xi.Request for Relief Will Be Submitted to NRC Timeframe to Support Second 10-year Insp Interval,Per 10CFR50.55a 05000327/LER-1998-003, Forwards LER 98-003-00 Re Automatic Reactor Trip with FW Isolation & Auxiliary FW Start as Result of Failure of Vital Inverter & Second Inverter Failure.Event Is Being Reported IAW 10CFR50.73(a)(2)(iv)1998-12-0909 December 1998 Forwards LER 98-003-00 Re Automatic Reactor Trip with FW Isolation & Auxiliary FW Start as Result of Failure of Vital Inverter & Second Inverter Failure.Event Is Being Reported IAW 10CFR50.73(a)(2)(iv) ML20196F9841998-11-25025 November 1998 Provides Changes to Calculated Peak Fuel Cladding Temp, Resulting from Recent Changes to Plant ECCS Evaluation Model ML20195H7891998-11-17017 November 1998 Requests NRC Review & Approval of Five ASME Code Relief Requests Identified in Snp Second ten-year ISI Interval for Units 1 & 2 ML20195E4991998-11-12012 November 1998 Forwards Rev 7 to Physical Security/Contingency Plan.Rev Adds Requirement That Security Personnel Will Assess Search Equipment Alarms & Add Definition of Major Maint.Rev Withheld (Ref 10CFR2.790(d)(1)) 05000328/LER-1998-002, Forwards LER 98-002-00 Re Automatic Turbine & Reactor Trip, Resulting from Failure of Sudden Pressure Relay on 'B' Phase Main Transformer1998-11-10010 November 1998 Forwards LER 98-002-00 Re Automatic Turbine & Reactor Trip, Resulting from Failure of Sudden Pressure Relay on 'B' Phase Main Transformer ML20195G5701998-11-10010 November 1998 Documents Util Basis for 981110 Telcon Request for Discretionary Enforcement for Plant TS 3.8.2.1,Action B,For 120-VAC Vital Instrument Power Board 1-IV.Licensee Determined That Inverter Failed Due to Component Failure ML20155J4031998-11-0505 November 1998 Provides Clarification of Topical Rept Associated with Insertion of Limited Number of Lead Test Assemblies Beginning with Unit 2 Operating Cycle 10 Core ML20154R9581998-10-21021 October 1998 Requests Approval of Encl Request for Relief ISI-3 from ASME Code Requirements Re Integrally Welded Attachments of Supports & Restraints for AFW Piping ML20155B1481998-10-21021 October 1998 Informs That as Result of Discussion of Issues Re Recent Events in Ice Condenser Industry,Ice Condenser Mini-Group (Icmg),Decided to Focus Efforts on Review & Potential Rev of Ice condenser-related TS in Order to Clarify Issues ML20154K1581998-10-13013 October 1998 Forwards Rept Re SG Tube Plugging Which Occurred During Unit 1 Cycle 9 Refueling Outage,Per TS 4.4.5.5.a.ISI of Unit 1 SG Was Completed on 980930 ML20154H6191998-10-0808 October 1998 Forwards Rev 0 to Sequoyah Nuclear Plant Unit 1 Cycle 10 COLR, IAW TS 6.9.1.14.c 05000328/LER-1998-001, Forwards LER 98-001-00 Providing Details Re Automatic Turbine & Reactor Trip Due to Failure of Sudden Pressure Relay on 'B' Phase Main Transformer1998-09-28028 September 1998 Forwards LER 98-001-00 Providing Details Re Automatic Turbine & Reactor Trip Due to Failure of Sudden Pressure Relay on 'B' Phase Main Transformer ML20151W4901998-09-0303 September 1998 Responds to NRC Re Violations Noted in Insp Repts 50-327/98-07 & 50-328/98-07.Corrective Actions:Revised Per SQ971279PER to Address Hardware Issues of Hysteresis, Pressure Shift & Abnormal Popping Noise 1999-09-27
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059K6661990-09-17017 September 1990 Forwards Evaluation That Provides Details of Plug Cracks & Justification for Continued Operation Until 1993 ML20059H4031990-09-10010 September 1990 Discusses Plant Design Baseline & Verification Program Deficiency D.4.3-3 Noted in Insp Repts 50-327/86-27 & 50-328/86-27.Evaluation Concluded That pre-restart Walkdown Data,Loops 1 & 2 Yielded Adequate Design Input ML20059E1851990-08-31031 August 1990 Responds to NRC Re Violations Noted in Insp Repts 50-327/90-22 & 50-328/90-22.Corrective Actions:Extensive Mgt Focus Being Applied to Improve Overtime Use Controls ML20059E2881990-08-31031 August 1990 Forwards Addl Info Re Alternate Testing of Reactor Vessel Head & Internals Lifting Rigs,Per NUREG-0612.Based on Listed Hardships,Util Did Not Choose 150% Load Test Option ML20059H1831990-08-31031 August 1990 Forwards Nonproprietary PFE-F26NP & Proprietary PFE-F26, Sequoyah Nuclear Plan Unit 1,Cycle 5 Restart Physics Test Summary, Re Testing Following Vantage 5H Fuel Assembly installation.PFE-F26 Withheld (Ref 10CFR2.790(b)(4)) ML18033B5031990-08-31031 August 1990 Forwards Financial Info Required to Assure Retrospective Premiums,Per 10CFR140 & 771209 Ltr ML20028G8341990-08-28028 August 1990 Forwards Calculation SCG1S361, Foundation Investigation of ERCW Pumping Station Foundation Cells. ML20063Q2471990-08-20020 August 1990 Submits Implementation Schedule for Cable Tray Support Program.Util Proposes Deferral of Portion of Remaining Activities Until After Current Unit 2 Cycle 4 Refueling Outage,Per 900817 Meeting.Tva Presentation Matl Encl ML20056B5181990-08-20020 August 1990 Responds to NRC Re Order Imposing Civil Monetary Penalty & Violations Noted in Insp Repts 50-327/90-01 & 50-328/90-01.Corrective Actions:Organizational Capabilities Reviewed.Payment of Civil Penalty Wired to NRC ML20063Q2461990-08-17017 August 1990 Forwards Cable Test Program Resolution Plan to Resolve Issues Re Pullbys,Jamming & Vertical Supported Cable & TVA- Identified Cable Damage.Tva Commits to Take Actions Prior to Startup to Verify Integrity of safety-related Cables ML20059A5121990-08-15015 August 1990 Provides Clarification of Implementation of Replacement Items Project at Plant for Previously Procured Warehouse Inventory.Util Committed to 100% Dedication of Commercial Grade,Qa,Level Ii,Previous Procurement Warehouse Spare ML20058M2321990-08-0707 August 1990 Forwards Rept of 900709 Fishkill,Per Requirements in App B, Environ Tech Spec,Subsections 4.1.1 & 5.4.2.Sudden Water Temp Increase Killed Approximately 150 Fish in Plant Diffuser Pond ML20058N2361990-08-0707 August 1990 Confirms That Requalification Program Evaluation Ref Matl Delivered to Rd Mcwhorter on 900801.Ref Matl Needed to Support NRC Preparation for Administering Licensed Operator Requalification Exams in Sept 1990 ML20058M4471990-07-27027 July 1990 Responds to Unresolved Items Which Remain Open from Insp Repts 50-327/90-18 & 50-328/90-18.TVA in Agreement W/Nrc on Scope of Work Required to Address Concerns W/Exception of Design Basis Accident & Zero Period Accelaration Effects ML20058M0111990-07-27027 July 1990 Forwards Addl Info Re Plant Condition Adverse to Quality Rept Concerning Operability Determination.Probability of Cable Damage During Installation Low.No Programmatic Cable Installation Problems Exist ML20055J3531990-07-27027 July 1990 Forwards Revised Commitment to Resolve EOP Step Deviation Document Review Comments ML20055J0771990-07-26026 July 1990 Requests Termination of Senior Reactor Operator License SOP-20830 for Jh Sullivan Due to Resignation from Util ML20055G6611990-07-17017 July 1990 Forwards Justification for Continued Operation for safety- Related Cables Installed at Plant,Per 900717 Telcon.No Operability Concern Exists at Plant & No Programmatic Problems Have Been Identified.Summary of Commitments Encl ML20058L7001990-07-16016 July 1990 Forwards Response to SALP Repts 50-327/90-09 & 50-328/90-09 for 890204 - 900305,including Corrective Actions & Improvements Being Implemented ML20055F6151990-07-13013 July 1990 Provides Addl Bases for Util 900320 Proposal to Discontinue Review to Identify Maint Direct Charge molded-case Circuit Breakers Procured Between Aug 1983 & Dec 1984,per NRC Bulletin 88-010.No Significant Assurance Would Be Expected ML20044B2211990-07-12012 July 1990 Forwards Addl Info Clarifying Certain Conclusions & Recommendation in SER Re First 10-yr Interval Inservice Insp Program ML20055D2531990-07-0202 July 1990 Provides Status of Q-list Development at Plant & Revises Completion Date for Effort.Implementation of Q-list Would Cause Unnecessary & Costly Delays in Replanning Maint,Mod, outage-related Activities & Associated Procedure Revs ML20043H9061990-06-21021 June 1990 Responds to Generic Ltr 90-04, Request for Info on Status of Licensee Implementaion of Generic Safety Issues Resolved W/Imposition or Requirements or Corrective Actions. No Commitments Contained in Submittal ML20043H2281990-06-18018 June 1990 Informs of Issue Recently Identified During Startup of Facility from Cycle 4 Refueling Outage & How Issue Addressed to Support Continued Escalation to 100% Power,Per 900613 & 14 Telcons ML20043G4901990-06-14014 June 1990 Forwards Tabs for Apps a & B to Be Inserted Into Util Consolidated Nuclear Power Radiological Emergency Plan ML20043F9261990-06-13013 June 1990 Responds to NRC Bulletin 89-002, Stress Corrosion Cracking of High-Hardness Type 410 Stainless Steel Internal Preloaded Bolting in Anchor/Darling Model S3502 Swing Check Valves or Valves of Similar Design. ML20043F9301990-06-13013 June 1990 Responds to NRC 900516 Ltr Re Violations Noted in Insp Repts 50-327/90-17 & 50-328/90-17.Corrective Action:Test Director & Supervisor Involved Given Appropriate Level of Disciplinary Action ML20043H0361990-06-11011 June 1990 Forwards Supplemental Info Re Unresolved Item 88-12-04 Addressing Concern W/Double Differentiation Technique Used to Generate Containment Design Basis Accident Spectra,Per 900412 Request ML20043D9921990-06-0505 June 1990 Responds to NRC 900507 Ltr Re Violations Noted in Insp Repts 50-327/90-14 & 50-328/90-14.Corrective Actions:Util Reviewed Issue & Determined That Trains a & B Demonstrated Operable in Jan & Apr,Respectively of 1989 ML20043C2821990-05-29029 May 1990 Requests Relief from ASME Section XI Re Hydrostatic Pressure Test Requirements Involving RCS & Small Section of Connected ECCS Piping for Plant.Replacement & Testing of Check Valve 1-VLV-63-551 Presently Scheduled for Completion on 900530 ML20043C0581990-05-29029 May 1990 Forwards Response to NRC 900426 Ltr Re Violations Noted in Insp Repts 50-327/90-15 & 50-328/90-15.Response Withheld (Ref 10CFR73.21) ML20043B3051990-05-22022 May 1990 Forwards Detailed Scenario for 900711 Radiological Emergency Plan Exercise.W/O Encl ML20043B1201990-05-18018 May 1990 Forwards, Diesel Generator Voltage Response Improvement Rept. Combined Effect of Resetting Exciter Current Transformers to Achieve flat-compounding & Installing Electronic Load Sequence Timers Produced Acceptable Voltage ML20043A6101990-05-15015 May 1990 Forwards Rev 16 to Security Personnel Training & Qualification Plan.Rev Withheld (Ref 10CFR2.790) ML20043A2391990-05-15015 May 1990 Forwards Revised Tech Spec Pages to Support Tech Spec Change 89-27 Re Steam Generator Water Level Adverse Trip Setpoints for Reactor Trip Sys Instrumentation & Esfas. Encl Reflects Ref Leg Heatup Environ Allowance ML20043A0581990-05-11011 May 1990 Forwards Cycle 5 Redesign Peaking Factor Limit Rept for Facility.Unit Redesigned During Refueling Outage Due to Removal & Replacement of Several Fuel Assemblies Found to Contain Leaking Fuel Rods ML20043A0571990-05-10010 May 1990 Forwards List of Commitments to Support NRC Review of Eagle 21 Reactor Protection Sys Function Upgrade,Per 900510 Telcon ML20042G9771990-05-0909 May 1990 Responds to NRC 900412 Ltr Re Violations Noted in Insp Repts 50-327/90-01 & 50-328/90-01 & Proposed Imposition of Civil Penalty.Corrective Actions:Rhr Pump 1B-B Handswitch in pull- to-lock Position to Ensure One Train of ECCS Operable ML20042G4651990-05-0909 May 1990 Provides Addl Info Re Plant Steam Generator Low Water Level Trip Time Delay & Function of P-8 Reactor Trip Interlock,Per 900430 Telcon.Trip Time Delay Does Not Utilize P-8 Interlock in Any Manner ML20042G4541990-05-0909 May 1990 Provides Notification of Steam Generator Tube Plugging During Unit 1 Cycle 4 Refueling Outage,Per Tech Specs 4.4.5.5.a.Rept of Results of Inservice Insp to Be Submitted by 910427.Summary of Tubes Plugged in Unit 1 Encl ML20042G0441990-05-0808 May 1990 Forwards Nonproprietary WCAP-11896 & WCAP-8587,Suppl 1 & Proprietary WCAP-8687,Suppls 2-E69A & 2-E69B & WCAP-11733 Re Westinghouse Eagle 21 Process Protection Sys Components Equipment Qualification Test Rept.Proprietary Rept Withheld ML20042G1431990-05-0808 May 1990 Forwards WCAP-12588, Sequoyah Eagle 21 Process Protection Sys Replacement Hardware Verification & Validation Final Rept. Info Submitted in Support of Tech Spec Change 89-27 Dtd 900124 ML20042G1001990-05-0808 May 1990 Forwards Proprietary WCAP 12504 & Nonproprietary WCAP 12548, Summary Rept Process Protection Sys Eagle 21 Upgrade,Rtd Bypass Elimination,New Steam Line Break Sys,Medical Signal Selector .... Proprietary Rept Withheld (Ref 10CFR2.790) ML20042G1701990-05-0808 May 1990 Provides Addl Info Re Eagle 21 Upgrade to Plant Reactor Protection Sys,Per 900418-20 Audit Meeting.Partial Trip Output Board Design & Operation Proven by Noise,Fault,Surge & Radio Frequency Interference Testing Noted in WCAP-11733 ML20042G1231990-05-0707 May 1990 Forwards Detailed Discussion of Util Program & Methodology Used at Plant to Satisfy Intent of Reg Guide 1.97,Rev 2 Re Licensing Position on post-accident Monitoring ML20042F7741990-05-0404 May 1990 Informs of Completion of Eagle 21 Verification & Validation Activities Re Plant Process Protection Sys Upgrade.No Significant Disturbances Noted from NRC Completion Date of 900420 ML20042F1691990-05-0303 May 1990 Responds to NRC Bulletin 88-009, Thimble Tube Thinning in Westinghouse Electric Corporation Reactors. Wear Acceptance Criteria Established & Appropriate Corrective Actions Noted. Criteria & Corresponding Disposition Listed ML20042G1381990-04-26026 April 1990 Forwards Westinghouse 900426 Ltr to Util Providing Supplemental Info to Address Questions Raised by NRC Re Eagle-21 Process Protection Channels Required for Mode 5 Operation at Facilities ML20042E9641990-04-26026 April 1990 Forwards Rev 24 to Physical Security/Contingency Plan.Rev Withheld (Ref 10CFR73.21) ML20012E6181990-03-28028 March 1990 Discusses Reevaluation of Cable Pullby Issue at Plant in Light of Damage Discovered at Watts Bar Nuclear Plant. Previous Conclusions Drawn Re Integrity of Class 1E Cable Sys Continue to Be Valid.Details of Reevaluation Encl 1990-09-17
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e TENNESSEE VALLEY AUTHORITY KNOXVILLE. TENNESSEE 379o OFFICE OF THE BOARD OF DIRECTORS December 3, 1979 Joseph M. Hendrie, Chairman U.S. Nuclear Regulatory Commission 1717 H Street, NW.
Washington, DC
Dear Chairman Hendrie:
We believe that tnere are advantages to be gained by pursuing certain limited activities in the case of those power plants where construction has been completed during the Commission's " pause" in issuing new construction permits and operating licenses, particularly where it can be demonstrated that the owner utility has taken the initiative in improving and promoting safety. We believe that the TVA program meets or exceeds the recommendations of the President's Commission and the NRC staff's short term lessons learned requirements. You will recall thac TVA completed a detailed review of our nuclear program in May. TVA has implemented a series of major improvements as a result of that review.
More recently, a special TVA nuclear safety task force has completed a review of the report by the President's Commission. This task force concluded, and we agree, that TVA meets all of the recommendations of the Kemeny commission report.
We are therefore asking that the NRC permit certain activities including fuel loading, =cro power physics testing, "special" testing and operator training to be conducted at the Sequoyah Nuclear Plant unit 1.
We believe that using the Sequoyah unit to conduct tests of the natural circulation cooling phenomena is particularly advantageous at this time.
There are questions about this mode of cooling under normal and degraded conditions which can be resolved by full scale demonstration testing.
Since the fuel in the reactor at Sequoyah would not have been operated at significant power, the inventory of fission products present would be minimal.
We believe that significant testing and operator training can be per-formed which would permit operation of the reactor at no greater than five percent power. A summary description of the type of tests which TVA could perform is included as Enclosure 1.
Construction necessary for fuel loading was completed at Sequoyah unit 1 on November 15, 1979. The NRC staff has completed the review of the operating license application with the exception of items related to Three Mile Island. The TVA response to the NRC Staff Short Term Lessons 1808 137 8001'so242_
2
- t. 4%
, r** }
An Ecual Opportur ity Ernployer h ypg g -
i Learned was submitted September 7, 1979, and your staff has been working with TVA to resolve these issues. Enclosed for your information are the TVA responses to the President's Commission on the Accident at Three Mile Island recommendations.
Our fuel loading and zero power testing would take approximately six weeks. We would then be able to begin special testing in mid-February.
Should events in the interim dictate that modifications to the plant are required, the nuclear fuel could be removed from the reactor vessel and stored in the spent fuel pool with no hazard to the public health and safety.
Additionally, we know you will be interested to know that TVA has ini-tiated a comparative risk analysis of the Sequoyah plant auxiliary feedwater system. This analysis will be complete by the time the pro- i posed low power tests are finished. In addition, we are evaluating other areas of the Sequoyah plant where meaningful risks assessments could be completed before full power operation.
Very truly yours, 7
h YN S. David Freeman Chairman of the Board Enclosures 1808 138
Enclosure 1 SEQUOYAH NUCLEAR PLANT UNIT 1
SUMMARY
OF SPECIAL TESTS Prior to core loading, the plant nuclear instrumentation and temporary nuclear instrumentation will be checked out. Plant systems requiring boration will be borated to the specified concentration.
Following core loading and prior to initial criticality, baseline testing will be performed with the core completely assembled. Major items to be performed are moveable detector system checkout, rod drive mechanism and rod cluster control assembly operation tests, reactor internal vibration measurements, pressurizer system optimization and reactor coolant loop flow coastdown measurements.
Af ter the reactor is brought critical, low power physics testing will begin. Plant baseline parameter measurements will be taken, reactivity measurements conducted, temperature coefficients determined, and boron endpoint measurements made. Reactivity measurements include integral ,
and differential bank worth tests, minimum shutdown margins verification, and determination of the affect of a rod ejection.
These tests are the normal tests performed to verify that integrated system response meets design assumptions, verify the core design basis, and verify that adequate shutdown margin exists throughout cycle 1. ;
i 1808 139
They are described in more detail in the Sequoyah Nuclear Plant Final Safety Analysis Report.
The following special tests conducted prior to exceeding 5 percent power are intended to provide a significant demonstration of reactor operation in the natural circulation mode under both normal and certain degraded conditions. These tests will also provide significant operator training and experience under these conditions. The tests will be repeated such that each operating shif t participates in each test.
To simulate decay heat, the reactor will be operated at less than 5 percent power with the reactor coolant pumps tripped. This mode of operation will closely approximate natural circulation conditions (with subcooling) following a reactor trip from full power after several months of power operation.
Since detailed test procedures and safety evaluations for these tests have not been completed, some modifications in test scope or detail may be required. Test durations and methods of power level control will be provided in the detailed test procedures and evaluation. Once test procedures have been written and corresponding safety evaluations developed for the special tests, they will be submitted to NRC along with appropriate license amendments. We intend to have Westinghouse Electric Corporation review these special test procedures as they are doing with other selected emergency procedures.
1808 140
I. Natural Circulation Verification Purpose Verify establishment of natural circulation in the primary system Initial Conditions Reactor Coolant Pumps operating Steam Generators being fed by normal feedwater supply Pressurizer Heater controlling pressure Reactor Power Z 3%
Normal primary system temperature ar.a pressure Test Description Test will be initiated by tripping of all reactor coolant pumps.
Operator will verify establishment of natural circulation by observing response of the hot leg and colo leg temperature instrumentation in each loop. Core exit thermocouples will be monitored to assess core flow distribution.
1808 141
II. Natural Circulation with Simulated Loss of Offsite Power -
u Purpose Verify that natural circulation cooling can be established and maintained following loss of offsite power.
Initial Conditions Reactor Power 1%.
Reactor Coolant Pumps operating.
Auxiliary Feed System operating on offsite power.
Pressurizer Heaters controlling pressure.
Normal primary system temperature and pressure.
Test Description Test will be initiated by a simulated loss of offsite power.
Reactor coolant pumps will be tripped, auxiliary feed pump and pressurizer heater loads will be transferred to diesel power.
Operator will verify establishment of natural circulation by observing response of hot leg and cold leg temperature instrumentation in each loop. Core exit thermocouples will be monitored to assess the core flow distribution.
1808 142
III. Natural Circulation with Loss of Pressurizer Heaters Purpose Verify establishment of natural circulation and determine the rate of decrease of margin to saturation while in this mode and the ability to reestablish margin through cooldown and makeup.
Initial Conditions Reactor Power Z 3%
Reactor Coolant Pumps operating Cecondary system stea- flow adjusted to maintain constant primary coalant temperature Steam generators being feed by normal feedwater supply Pressurizer heaters controlling pressure Test Description Test will be initiated by tripping pressurizer heaters and reactor coolant pumps. Establishment of natural circulation will be verified by observing response of hot leg and cold leg temperature instrumentation in each loop. Core exit thermocouples will be monitored to assess the core flow distribution. The operator will observe the saturation meter to verify margin. Prior to reaching saturation, secondary side steam flow will be increased to affect cooldown and reestablishment of saturation margin will be verified. In conjunction with cooldown, the operator feeds the primary system to compensate frr shrinkage.
1808 143
IV. Effect of Steam Generator Isolation (Secondary Side) on Natural ;
Circulation Purpose Verify the effects of steam generator isolation (secondary side) on natural circulation.
Initial Conditions Reactor Power 3%
All steam generators fed by normal feedwater supply Reactor coolant pumps on Secondary system steam flow adjusted to maintain constant temperature Test Description Trip reactor coolant pumps and verify establishment of natural circulation. Cooldown using steam dumps to provide sufficient margin to steam gent.ator safeties. Isolate steam generators one at a time until three are isolated or primary system temperature starts to increase. Hot and cold leg temperatures will be monitored to ensure that sufficient heat is being removed by the natural circulation process. The steam generators will be returned to service one at a time and the reestablishment of natural circulation 1808 144
will be verified in each loop. Core exit thermocouples will be monitored to assess core flow distribution.
V. Natural Circulation at Reduced Pressure Purpose
- 1) Verify operation and test accuracy of primary system saturation meter.
- 2) Provide operations personnel vith online experience in using saturation meter to monitor and control margin to saturation.
- 3) Provide operational verification so that changes in saturation margin will not affect natural circulation provided adequate margin to saturation exists.
Initial Conditions Reactor Power Z 3%
Reactor coolant pumps operating Steam generators being fed by normal feedwater supply Pressurizer heaters controlling pressure Reactor coolant system pressure normal Secondary system steam flow adjusted to maintain constant temperature I Test Description _
~
Test is initiated by tripping of reactor coolant pumps and verifying establishment of natural circulation. Primary system pressure will be reduced as' primary system temperature is held constant. Accuracy of saturation meter will be verified during pressure reductions.
The effect of each pressure reduction on natural circulation will 1808 145
be observed. Cr.re exit thermocouples will be monitored to assess core flow distribution.
VI. Determine the cooldown capability of the charging and letdown system Purpose Determine the cooldown capability of the charging and letdown system with the secondary plant isolated.
Initial Conditions Reactor shutdown Pressurizer heaters controlling pressure '
Reactor coolant pumps running All steam generators fed by normal feedwater flow Test Description Trip three reactor coolant pumps. Cooldown using steam dumps to provide margin to steam generator safeties. Isolate all steam generators. Establish charging and letdown for maximum cooling capability. Verify the cooldown capability of the charging and letdown system from the hot and cold leg temperatures in the active loop. This will be accomplished by periodically interrupting feed -
and bleed to permit heatup. Core exit thermocouples will be monitored to assess core flow distribution.
1808 146
VII.J Simulated Loss of All Onsj te and Offsite AC Power Purposa To verify:
1.
Hot standby conditions can be maintained, 2.
Auxiliary feedwater can be controlled by manual means; i.e., with loss of AC power and control air, 3.
Critical plant operations can be performed using emergency lighting, .
- 4. .
Ability AC, and of 125-volt battery to supply 125-volt vital -
5.
Selected equipment areas do not exceed maximum design
=
_ Initial Conditions Reactor critical at N1 percent power.
Reactor Coolant Pumps operating.
Pressurizer heaters controlling primary system pressure.
Test Description Test will be initiated by:
- 1. .
Tripping RCP's and pressurizer heaters, 2.
Tripping auxiliary building and control building lighting boards, 3.
Removing AC power from auxiliary feedwater components and main steam power reliefs, b
- 4. [;;;
Tripping selected space and equipment coolers, E~
5.
Tripping vital battery chargers and AC power to inverter, +=
6.
Isolating main feedwater and main steam lines, 7.
Establishing manual control of auxiliary feedwater, 8.
After two hours, terminating the test by restoring AC power and returning equipment to normal service,
- 9. Shutdown reactor, and 10.
Cooling down primary system crd placing RHR system in service 180LB 147