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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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TENNESSEE VALLEY AUTHORITY
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CHATTANOOG/ . TENNESSEE 374o1 SN 157B Lookout Place APR 081987 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:
In the Matter of. ) Docket Nos. 50-327 Tennessee Valley Authority ) 50-328 SEQUOYAH NUCLEAR PLANT - ALTERNATE ANALYSIS PROGRAM PHASE II Pursuant to our meeting of March 3, 1987, on Sequoyah Small Bore Piping and Supports Interim Acceptance Criteria Phase II, TVA submits the following information:
4 Enclosure A - Sequoyah Nuclear Plant (SQN) Units 1 and 2, Alternate Analysis Review Program, Phase 2 Scope, Approach, and Schedule Enclosure B - Sequoyah Nuclear Plant (SQN) Alternate Analysis Review Program, Justification for Limited Walkdowns for Unit 2, Phase 1 Enclosure C - Sequoyah Nuclear Plant Units 1 and 2 Instrument Lines
-and Supports Issues, Current Activities, and Planned' Post-Restart Activities Enclosure D - Sheet 9 of the March 3, 1987 handout, revised to i incorporate NRC comments Enclosure E - A list of specific commitments made in this submittal
.This submittal completes commitments made during our meeting of March.3, ,
1987. If additional information is required, please telephone 1 M. R. Harding at (615) 870-6422.
- Very truly yours, j TENNESSEE VALLEY AUTHORITY z j R. ridleyfirector Nuclear Safety and Licensing Enclosures cc
- see page 2 OhhPDR An Equal Opportunity Employer-
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l U.S. Nuclear Regulatory Commission
/ R 081987 cc (Enclosures):
Mr. G. G. Zech, Assistant Director Re51 onal Inspections Division of TVA Projects Office of Special' Projects U.S. Nuclear Regulatory Commission
-Region II 101 Marietta Street, NW, Suite 2900
. Atlanta, Georgia 30323 Sequoyah Resident Inspector Sequoyah Nuclear Plant 2600 18 0u Ferry Road Soddy Daisy, Tennessee 37379 r
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U.S. Nuclear Regulatory Commission.
/\Pil 081987 t
l cc (Enclosures):
L Mr. G. G. Zech,' Assistant Director Regional Inspections Division 'of TVA Projects Office of Special Projects U.S. Nuclear Regulatory Commission Region II
^
101 Marietta Street, NW,LSuite 2900 Atlanta, Ceorgia 30323 Sequoyah Resident Inspector 3
Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy, Tennessee 37379-MRH:MJB:WAM:KLH cc (Et closures):
RIMS, MR 4N 72A-C 4 H. L. Abercrombie, O&PS-4, Sequoyah j E. S. Christenbury, E11 B33 C-K l C. H. Crowell, LP 6N 38A-C W. H. Hannum..:BR IN 76B-C M. R. Harding, O&PS-4, Sequoyah T. A. Ippolito, LP 5N 121B-C N. C. Kazanas, LP AN 45A-C J. A. Kirkebo, W12 A12 C-K D. R. Nichols, LP SN 302B-C
! L. M. Nobles, POB-2, Sequoyah R. K. Seiberling, 716C EB-C j M. B. Whitaker, LP 5N 303B-C j D. L. Williams, W10 B85 C-K f
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Enclosure A Sequoy:2h Nuclear Plant (SQN)-
Units 1 and 2 Alternate Analysis Review Program l Phase 2 Scope, Approach, and Schedule i
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CONTENTS i
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l.0 Scope. . . . . . . . . . . . . . . ..... ........... Al 2.0 Approach . . . . . . . . . . . . . ................. A2 2.1 Process Piping. . . . . . . . ................. A2 l 2.2 Instrument Lines. . . . . . . ................. A3 i l
3.0 Schedule . . . . . . . . . . . . . ................. A3
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1.0 Scope (a) Pipe Classes TVA Classes B, C, D, and M (equivalent to ASME III, classes 2 and 3).
(b) Process Piping The following systems include alternately analyzed piping within the scope of this program. The systems are listed in the order of the priority established for performing the Phase 2 walkdowns and evaluations.
Priority Syst_em Description 1 68 Reactor Coolant 2 3 Main and Auxiliary Feedwater 3 62 Chemical and Volume Control 4 63 Safety Injection 5 74 Residual Heat Removal 6 87 Upper Head Injection 7 61 Ice Condenser 8 72 Containment Spray 9 70 Component Cooling 10 67 Essential Raw Cooling Water 11 82 Diesel Starting Air 12 18 Diesel Fuel Oil 13 1 Main Steam 14 31 HVAC Refrigeration 15 26 Fire Protection 16 77 Waste Disposal 17 78 Fuel Pool Cooling 18 84 Flood Mode Boration (c) Instrumentation The following systems include alternately analyzed instrument lines within the scope of this program. _
Instrument Sensing Sampling and Radiation Monitoring Essential Control Air Al DNE1 - 1655M
s (d) Approximate Quantities Unit 1 Unit 2 Process Piping (feet) 10,000 30,000 Instrument Lines (feet) 300,000 300,000 2.0 Approach Completion of Phase 2 of the Alternate Analysis Review Program will provide confidence that all alternately analyzed process piping, instrument lines, and supports are in accordance with design basis requirements. A general description of the different approaches to be used for process piping and instrument lines is provided below. (Note:
Alternately analyzed non-ANS safety class process piping, which performs a secondary safety function and is required for unit 2 restart has been walked down and evaluated by EQE Incorporated. Similar walkdowns and evaluations are being performed for piping required for unit 1 restart.
This piping is not within the scope of the program described in this document.)
2.1 process Piping (a) Walkdowns 100 Percent walkdowns will be performed for all process piping and supports within the scope of this program. Walkdown drawings will be developed for all piping and associated supports.
(b) Pipe Stress Evaluations Using the information collected by the walkdowns, all process piping within the scope of this program will be evaluated for compliance to the design basis requirements. Evaluation methods will include using a now or revised Alternate Analysis Criteria (i.e., "ecokbook"), performing hand calculations, and performing computer analysis. Support design loads will be _
tabulated for all supports. Support modifications will be specified where necessary.
(c) Support Evaluations Using the information collected by the walkdowns and the loads derived by the piping analysis, all process piping supports will be evaluated for compliance to design basis requirements.
Evaluation methods will include hand calculations and computer analysis. New support designs will be prepared and existing' support designs will be medified as necessary.
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.l 2.2 Instrument Lines (a) Walkdowns )
Walkdowns will be performed for all instrument lines and ,
supports within the scope of this program. Screening criteria !
I will be used by the walkdown teams to identify instrument lines with overspans, and with thermal expansion and anchor movement flexibility concerns. Sketches will be prepared for portions of these lines which do not pass the screening criteria. Also, problem supports (i.e., supports with marginal designs) and supports that have been significantly varied from standard designs will be identified by the walkdowns and sketched.
(b) Instrument Line Stress Evaluations Using the sketches and information generated for the walkdowns, overspans and areas with flexibility concerns will be evaluated for compliance to design basis requirements. Evaluation methods will include using a new or revised Alternate Analysis Criteria (i.e. , " cookbook"), performing hand calculations and performing computer analysis. Support design loads will be screened against an accepted nominal load and design loads will be tabulated for supports with loads exceeding the nominal value. Support modifications will be specified where necessary.
(c) Support Evaluations Using the sketches and information generated by the walkdowns plus the nominal loads and loads derived by the instrument line analysis, instrument line problem supports, supports with significant variances, and supports with loads greater than the accepted value will be evaluated for compliance to design basis requirements. Evaluation methods will include hand calculations and computer analysis. New support designs will be prepared and existing designs will be modified as necessary.
In addition, a sampling program will be conducted for supports -
outside the areas specifically evaluated to establish confidence that variances to standard support designs have not resulted in unacceptable support configurations.
3.0 Schedule Phase 2 of the Alternate Analysis Review Program (including modifications), will be complete by the end of the cycle 4 refueling outages for both units. The cycle 4 refueling outages will be extended as necessary to permit installation of all Phase 2 modifications.
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TVA/ SON . ALTERNATE ANALYSIS REVIEW PROGRAM .
SCHEDULE YEAR 86 87 88 89 90 MONTH 10 ll 12 1 2 34 5 6 7 8 9 10 11 12 8 2 3 4 5 6 7 8 9 10 11 12 1 2 3 4 5 6 7 8 9 10 11 12 1 l
Q WALKDOWNS p
ESTABLISH b-h(M)
ANALYSIS / SUPPORT EVALUATIONS MODIFICATIONS IMPLEMENTED -
l pTRACOR l STAFFING l
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UNIT I ANALYSIS / SUPPORT EVALUATIONS "
PHASE 2 g uCO FicATIONS IMPLEMENTED I .,
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ANALYSl5 / SUPPORT EVALUATIONS "
UNIT 2 G i ?
PHASE 2 j MOOlFICATIONS IMPLEMENTED i INSTR EVALUATIONS . I l
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' l' il q, STANT Ei40 START END UNIT 2 U T C E
UNIT 2 C4 RESTART l OUTAGE U ( TARGET)
(TARGET) START END (TARGET) -l UNIT I C4 UNIT 4 OUTAGE
. HESTART tTARGET)
(TARGET)
" EVALUATIONS PRIORITIZED AS A FUNCTION OF
, SYSTEM SAFETY SIGNIFICANCE AND ACCESSIBILITY t
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l Enclosure B Sequoyah Nuclear Plant (SQN)
Alternate Analysis Review Program Justification for Limited Walkdowns for Unit 2. Phase 1 I
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j Justification for Limited Walkdowns for Unit 2, Phase 1 For Unit 2, Phase 1, 100 percent walkdowns were performed for piping with operating temperatures above 200 0 F, but walkdowns were generally limited for lower temperature piping to the areas local to other Phase 1 items (i.e.,
motor operated and pneumatic valves, branch /run interfaces, and interfaces with deadweight supported piping larger than 2 inches nominal size). Phase 1 piping evaluations were performed based on the walkdown information.
Evaluation of the Phase 1 items and resolution of associated deficiencies is adequate to ensure that all piping and supports are acceptable for restart.
The selection of the specific Phase 1 items that were addressed has NRC concurrence as noted in the draft Safety Evaluetion Report (Reference 6.2).
Since the Phase 1 items have been determined to be those critical for restart, it follows that limiting walkdowns to piping local to those items is sufficient for restart.
In addition to the evaluation of the Phase 1 items, piping outside the local areas was reviewed for obvious discrepancies using existing drawings.
Walkdowns were performed where drawings could not be located or if information on the drawings was unclear. The existing drawings are sufficiently accurate for the review of piping outside the local areas. Observed discrepancies between the existing drawings and the "as-constructed" piping configuration are generally limited to dimensional differences that are within construction tolerances. Therefore, a review of drawings, rather than performing walkdowns for piping outside the local areas, was justified within Phase 1.
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Enclosure C i SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 l
Instrument Lines and Supports i Issues, Current Activities, and Planned Post-Restart Activities
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CONTENTS P_!yL, 1.0 Purpose. . . . . . . . . . . . . . J. . . . . . . . . . . ..... Cl 2.0 Scope. . . . . . . . . . . . . . . . . . . . . . . . . . ... . .. C1 3.0 Issues and Current Activities. . . . . . . . . . . . . . . .. ... C1 3.1 Instrument Lines. . . . . . . . . . . . . . . . . . ...... C1 3.2 Instrument Line Supports. . . . . . . . . . . . . . ...... C5 3.3 Documentation . . . . . . . . . . . . . . . . . . . . ..... C5 4.0 Planned Post-Restart Activities. . . . . . . . . . . . . . .. ... C6 4.1 Walkdowns . . . . . . . . . . . . . . . . . . . . . . .. ... C6 4.2 Evaluations . . . . . . . . . . . . . . . . . . . . . .. . .. C7 l
4.3 Documentation . . . . . . . . . . . . . . . . . . . . .. . .. C7 '
l 5.0 Actions to Prevent Recurrence. . . . . . . . . . . . . . . ... .. C8 l l
i 6.0 References . . . . . . . . . . . . . . . . . . . . . . . . .. . .. C8
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1.0 Purpose The purpose of this document is to identify issues related to the support of instrument lines, provide information on any current activities related to those items and to define planned post-restart activities.
2.0 Scope This document is applicable to all safety-related instrument lines and supports at Sequoyah Nuclear Plant (SQN). Systems and approximate quantities are listed below:
Approximate Maximum Quantities (per unit)
System Size / Material Temperature Lineal Feet No. Supports Sensing 1/2" Sch 80 Ambient (1) 250,000 35,000 and 160 SS Piping Sampling 3/8", 1/2", 6500 F(2) 50,000 13,000 and 1" SS Tubing Radiation Monitoring Essential 1/2" SS Tubing, Ambient 4,000 750 Control Air 2", 3", 4" Schedule 40 SS Piping (1) Stagnant lines.
(2) Approximately 15,000 feet per unit has operating temperatures above 1200F.
3.0 Issues Following is a summary of generic issues related to instrument lines and supports and information on any current activities related to those issues. Planned post-restart activities are described in section 4.0. .
3.1 Instrument Lines (a) Instrument sensing line slope.
Instrument sensing line slope issues will be addressed in a separate tran.smittal from TVA to NRC.
(b) Overspans for instrument lines with no concentrated weights or with small (less than 15 pounds) concentrated weights.
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A random sample of 60 supports were walked down and evaluated for Employee Concern Task Group (ECTG) Report 173.03, " Missing ,
Instrument Line Clamps," (see issue 3.3(a)). Overspans !
identified by these walkdowns were evaluated and each case was ,
found to be acceptable (i.e., code allowables were not l exceeded). These evaluations, coupled with experience data and !
test data which indicate that pipe failures do not occur in similarly-supported piping, provide a sufficient basis for ;
deferral of additional activities on this issue to post-restart. l (c) Flexibility of instrument lines at interfaces with process linec to accommodate process line thermal and seismic movements.
In Phase I of the Alternate Analysis Review Program, 256 (total scope) branch /run interfaces were evaluated to determine if the branch line had adequate flexibility to absorb run line movements without overstressing the pipe or supports.
Modifications (i.e., support deletions) were made to 70 cases to ensure compliance with design requirements.
Additional evaluations were performed to determine if failures would have occurred had the modifications not been performed.
It was concluded that for some cases support failures were possible, but in no case would the run line movements or support failures have resulted in a pipe failure.
Instrument lines at SQN ere supported similar to the process line branch connections. However, sine, these lines are either 1/2-inch piping or 1 inch and smaller tubing, instrument lines are more flexible than most process line branches. (Nete:
Essential Control Air includes some 2" 3", and 4" piping but this piping does not interface with process piping run lines.)
Therefore, it is concluded that run line anchor movements will not produce failures in instrument lines and no further evaluation is necessary prior to restart.
(d) Flexibility of instrument lines for thermal expansion.
(Sampling and radiation monitoring lines only. Sensing and
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essential control air lines are not subject to significant .
thermal expansion.)
Thermal expansion flexibility was evaluated in the initial design of sampling and radiation monitoring lines at SQN.
Field inspections show that most supports on these lines are guides (2-way supports) and no cases of multiple, in-line, 3-way supports have been observed on lines with operating temperatures greater than 1200F. The first support around a bond is spaced to provide flexibility for thermal expansion and flexible loops are used in long runs. It is believed that the installation is acceptable (i.e., in compliance with design basis requirements), but due to generic problems in the application of alternate analysis methods, it has been decided to verify the adequacy of the original design in the ,
post-restart portion of the alternate analysis review program.
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A very few cases of constrained thermal expansion in sampling and radiation monitoring lines do exist. During field inspection, four 3/8-inch OD sampling lines (2 on each unit) were observed to be bowed in a sine wave shape that is characteristic of constrained expansion of small diameter j tubing. For those lines, the constraint is the result of an
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interference with a fixed object at the end of a long straight ,
run.
Based on calculations and on previous laboratory tests conducted by TVA, it is concluded that bowing is an acceptcble mechanism for relieving thermal expansion in small diameter tubing for which slope maintenance is not essential to functional capability. During heatup, constrained thermal expansion is relieved by approximately equal lateral displacements (bowing) in each tubing span. This has been confirmed in laboratory tests and by inspection of actual ,
installations which are bowed. During cooldown, bowing of the line will be recueed, but for high temperatures (generally greater than 5000 F), it will not be entirely eliminated. In a subsequent heatup, this new equilibrium condition is the starting point of the next thermal cycle.
For the worst-case of a maximum operating temperature of 6500F, the maximum strain (based on test data) is approximately 0.15 percent for the first cycle. This corresponds to an intensified stress (conservatively using i = 1.3) of approximately 49,000 lb/in2, which is within the 3Se allowable of 56,400 lb/in2 permitted by ASHE III for 316 SS subject to a single nonropeated anchor movement. For all subsequent cycles originating from the already-bowed condition, the maximum strain is approximately 0.1 percent, which corresponds to an intensified stress of approximately 33,000 lb/in2 Combining with stress due to precsure and deadweight yields an equation 11 stress of approximately 36,000 lb/in2 , which is within the equation 11 allowable limit for 316 SS at 6500F, even if conservatively assuming 45,000 cycles. (Fewer than 1/3 that many cycles are expected to occur before corrections are made to relieve the constrained -
thermal expansion.) Therefora, allowable stress levels are not exceeded as a result of the constrained thermal expansion and resultant bowing. In addition, slope is not critical for sampling and radiation monitoring lines and, therefore, bowing does not result in slope-related problems.
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The known quality of the initial design, combined with the acceptable bowing of the very few lines with constrained thermal expansion, plus the fact that no failures attributed to inadequate thermal expansion flexibility have occurred during unit operation, provides a sufficient basis for deferral of further evaluations on this issue to post-restart.
(e) Bent or bowed instrument lines.
Some sections of instrument tubing are bent or bowed due to personnel loads and, in rare instances, due to constrained thermal expansion (see item (d)]. This issue applies primarily to relatively light 3/8-inch and 1/2-inch OD tubing (sampling, radiation monitoring, and control air lines), and in some cases to 1/2-inch Schedule 80 and 160 piping (sensing lines). (Since sensing lines are stagnant and not subject to thermal ,
expansion, bends in those lines are due to personnel loads.)
Bends and bowing in sampling, radiation monitoring, and control air lines do not prevent. normal function of those lines (slope is not critical for these systems) and are not a safety concern. However, bends in sensing lines could result in slope problems. To alleviate these concerns, instrument sensing lines that are required for restart and that may be subject to outgassing or air entrapment will be walked down and evaluated before restart. Corrections will be made as necessary to ensure proper slope for these lines.
(f) Instrument line clearances.
Instrument line clearance concerns were identified during recent walkdowns. This issue applies primarily to sampling and radiation monitoring lines which have displacements due to thermal expansion.
Testing of instrument line interferences with adjacent tubes performed by TVA indicate that the moving tube will easily slip by an adjacent tube without resulting in any overstress for either tube. Also, the testing shows that support loads - ,
resulting from tube contact are generally very small and will i not result in clamp failures. Further [as noted in item (d)),
instrument line interferences with fixed objects are relieved by bowing of the tubing without resulting in any overstress.
The results of the interference testing, and the bowing testing and evaluations noted in item (d), provide a sufficient basis for deferral of additional activities on this issue to post-restart.
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3.2 Instrument line supports (a) Missing, loose, damaged, and improperly installed clamps.
In response to ECTG Report 173.03, a random sample of 60 clamps was reviewed to establish confidence that the instrument line and support installations are acceptable. Those 60 clamps, plus 2 clamps on either side of the 60, were walked down to determine the "as-installed" configuration. A total of approximately 300 clamps were actually included in the sample and only 3 clamps were found to be missing. The instrument lines and supports affected by the missing clamps were evaluated and both the instrument lines and clamps were found to be well within allowable stress or 1.ad limits. Also,
" loose, damageo, and improperly installed" clamps were found to be functional and have more than adequate capacity.
When instrument line clamp maintenance items have been identified during the various waltdowns for instrumentation and other issues, appropriate msintenance work has been performed.
Also, maintenance on instrument line clamps has been performed in accordance with Special Maintenance Instruction SMI-0-317-25 (reference 6.1). The sampling program described above indicates that instrument line clamp issues either are not significant or have been corrected through this maintenance process.
In summary, it has been concluded that the ample margin in both the instrument lines and the clamps, combined with the low occurrence of missing clamps, and the insignificance of the
" loose, damaged, and improperly installed" clamps, provides a sufficient basis for deferral of additional activities on these items to post-restart.
(b) Instrument mounting plates, brackets, and bolts incorrect.
Various instrument mounting concerns were identified during recent walkdowns. Additional walkdowns of instruments required for restart will be performed to identify and correct -
instrument mounting inadequacies.
3.3 Documentation Various concerns have been identified regarding documentation for the installation and inspection of instrument lines and supports.
These concerns do not represent a safety problem and will be addressed after restart.
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4.0 Planned Post-Restart Activities The following describes planned post-restart activities to fully resolve the issues identified in section 3.0 and ensure that all instrument lines and instrument line supports are in compliance with design basis requirements.
4.1 Walkdowns Walkdowns will be performed for all safety-related instrument lines and supports. These walkdowns, along with the evaluations described in section 4.2, will provide confidence that all instrument lines and supports are in accordance with design basis requirements.
These waltdowns will develop the "as-constructed" configuration for instrument lines and supports which do not pass screening criteria developed for the walkdowns.
4.1.1 Walkdowns for instrument line and support stress issues.
A walkdown program will be developed to identify situations that may be significant for instrument line and support stress. Walkdowns utilizing a screening criteria will be used to identify overspans, and to identify potential problems with seismic and thermal anchor movement flexibility and thermal expansion flexibility. For supports, a screening criteria will be used to identify problem supports (e.g.,
marginal designs) and supports that have been varied significantly from the standard design. Skotches will be prepared for those instrument lines and supports that need additional analysis to assure compliance with design basis commitments.
4.1.2 Walkdowns for installation issues.
A walkdown program will be developed to address instrument line and clamp installation issues. These walkdowns will identify and correct damaged and improperly installed instrument lines and clamps (including torqueing clamp bolts, as required) to ensure compliance with design basis -
commitments.
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4.2 Evaluations Using the sketches generated by the walkdowns, evaluations will be performed (see below for criteria) for instrument lines and supports that do not pass the walkdown screening criteria. Modifications will be performed as necessary to ensure compliance with design basis requirements.
In addition, a sampling program will be conducted to establish confidence that variances to standard support designs have not resulted in unacceptable support configurations.
Criteria for Evaluation of Instrument Lines and Supports Instrument Lines - SQN-DC-V-13.3, "SQN - Design Criteria for Detailed Analysis of Category I Piping Systems."
In addition, a strain based criteria similar to that proposed in ASME Code Case N-196 may be proposed for the evaluation of instrument lines. If so, detailed information and justification for that critoria will be provided in a future transmittal.
Instrument Line Supports - SQN-DC-V-24.1, "SQN - Design Criteria for Location and Design of Piping Supports and Supplemental Steel in Category I Structures."
4.3 Documentation Documentation for the qualification of instrument lines and supports will consist of the following:
(a) Walkdown packages with data sheets for all instrument lines and supports, and with sketches for all instrument line and support configurations which do not pass the walkdown screening criteria.
(b) Evaluations of instrument line and support configurations which do not pass the walkdown screening criteria, based on the sketches developed by the walkdowns.
(c) Issued drawings for supports that are required to be modified.
(d) Sampling program showing that variances to standard support designs are acceptable.
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5.0 Actions to Prevent Recurrence Existing procedures will be revised and new procedures will be developed to address all instrument line and support issues and prevent recurrence of similar problems in future installations. Procedure revisions and new procedures will include:
(a) Procedures will be revised to eliminate the future use of variances on instrument line supports. All future changes will be made using issued drawings.
(b) An " Engineering Requirements Specification" for instrument line installation and inspection will be issued for SQN. Existing installation and inspection procedures will be revised and new procedures will be issued to implement this specification.
(c) The instrumentation drawings and typical support drawings will be revised to upgrade and clarify design requirements.
6.0 References 6.1 SMI-0-317-25, SQN - Special Maintenance Instruction " Inspection and Corrective Maintenance of Small Diameter Instrument and Sensing Lines."
6.2 B. J. Youngblood's letter to S. A. White dated December 19, 1986 (A02 861222 032).
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ENCLOSURE D SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 l
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s TVA/ SON ALTERNATE ANALYSIS REVIEW PROGRAM PROGRAM STRUCTURE UNIT 2 : PHASE I y7PROGRAM ,,,,/SCOPE '
PIPING WITH /
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PIPING DESIGNED BY ALT. ANALYSIS WITHIN DVBP SCOPE PIPING DESIGNED BY ALTERNATE ANALYSIS TOTAL PLANT PIPING SYSTEMS IN UNIT 2 AND COMMON u------- . - - - - - - - - - - - - _ - - - - - - - - - - - - - - - _ _ _ _ _ -__
ENCLOSURE E' .
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SEQUOYAH NUCLEAR PLANT ALTERNATE ANALYSIS PHASE II PROGRAM COMMITMENTS
- 1. Phase II of the Alternate Analysis Program will be complete by the end of the cycle 4 refueling outage for each unit. The program scope will include walkdowns, pipe stress evaluations, and support evaluations for process piping and instrument lines.
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