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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
[Table view] |
Text
s 9
,o-TENNESSEE VALLEY ' AUTHORITY CHATTANOOGA. TENNESSEE 374ot 400 Chestnut Street Tower II March 1, 1985 Director of Nuclear Reactor Regulation Attention: Ms. E. Adensam, Chief Licensing Branch No. 4 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Ms. Adensam:
In the Matter of ) Docket Nos. 50-327 Tennessee Valley Authority ) 50-328 Enclosed is our response to your October 1, 1984 letter to H. G. Parris regarding additional information on the Sequoyah Nuclear Plant safety parameter display system (SPDS). The delay in submittal of this response was discussed with Carl Stahle of your staff. Previous information regarding the SPDS was provided to you by a January 4, 1984 letter from L. M. Mills.
If you have any questions concerning this matter, please get in touch with Jerry Wills at FTS 858-2683 Very truly yours, TENNESSEE VALLEY AUTHORITY R..H. Shell
. Nuclear Engineer Sworn to and subscribed before me this ./sf day of /7//> vM 1985 '
/ 4t Y $ /)
Notary Pubit6 My Commission Expires 6,/Je//fh Enclosure
, cc: 'U.S. Nuclear Regulatory Commission (Enclosure)
Region II Attn: Dr. J. Nelson Grace, Regional Administrator 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 8503120291 850301 PDR ADOCK 05000327 F. PDR _
fe\
I l An Equal Opportunity Employer
ENCLOSURE TVA RESPONSE TO NRC LETTER DATED OCTOBER 1,1984 FROM E. ADENSAM T3 H. G. PARRIS REQUEST FOR INFORMATION ON THE SEQUOYAH NUCLEAR PLANT SAFETY PARAMETER DISPLAY SYSTEM (SPDS)
The following attachments provide information in the areas identified by NRC: ,
Attachment '. - SPDS Isolation Devices
.- Attachment 2 - SPDS Description Attachment 3 - SPDS Verification and Validation Program
~
Attachment 4 - Unreviewed Safety Questions Attachment 5 - Implementation Schedules a
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SEQUOYAH NUCLEAR PLANT SPDS ISOLATION DEVICES.
- The' Safety Parameter Display System (SPDS) computer included points run 4 'friom!the instruments and inputs wired from the input of the plant process computer to the SPDS computer. Since isolation of inputs to the plant process computer was part of_the design considerations prior.to the installation and startup of that system ind jhave not been affected by. ,
i the SPDS installation, the information on isolation devices is supplied t
- for the 28 devices installed solely for the SPDS computer.
\
. The following information on the isolation devices is given in the same
- order as paragraphs a through f in reference 1.
- Paragraph a. Testing performed to demonstrate that the device is acceptable for its application.-
The electrical isolator associated with this particular request is a 0 to 100 MV range (input to output), E-Max Instruments, Inc. ,
model No.175C304. The application calls for a device capable of, '
! providing electrical isolation of a low-level 0-100 MV load. The j device must perform this isolation at.all times--before,.during, l and after the design basis event; it does not need to maintain i
signal continuity during or af ter these events. Verification that this device is acceptable and qualified for the installed application
] is demonstrated by tests in the area of seismic, electrical, environmental, and EMI parameters.
1
'ht internal schematic diagram for the model 175C304 isolator, E-Max
, drawing 175C304, and the test procedure ~are attached for reference.
i This drawing shows the electrical input is completely isolated
{ electrically from the output and that coupling to maintain continuity j of the information signal is accomplished by means_of optical isolation. The optical isolation provides a 1/4-inch physical
[ clearance between the input and output electrical circuitry.
i.
E-Max Test procedure covering electridal tests of_the_175D304 device is attached .to show integration of factory test- fixture and j: module being tested.
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- Na TEST PROCEDURE ANALOG OPTICAL'ISOIATOR, VOLTAGE g .... ,
P/N 175D304 ,
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A. Test Equipment Required: '- '
l.ea. E-MAX Test Fixture j l'ea. Scope - Tektronix T932A or equivalent l i
l 1,ea. Counter - HP5232A or equivalent -
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. ,. I ea. Signal Generator - Wavetek 180 or evuivalent l ea. DVM - Keithly 177 or evuivalent 1 ea. Hypot Testdr 0-3 kV AC 1 ea. 100 mV DC Source B. Procedure
- 1. Set up the test fixture as shown:
100 MV Signal
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Output Meter Test Fixture -
py e Module -
- 4 Set signal generator to output a 10 Hz square wave",
1 a.
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10 Volts peak-to-peak from OV to 10V.
- b. Set Voltage / Current switch to voltage.
Soarce select to external c.
- d. .
sw ee Range select-to maximum. y
- e. Common mode switch to common mode position.-
- 2. Apply power to the module to be tested. *
- 3. Measure the i 12 Volt supplies for the input and. output sections. They should be 12V + .5V.
- 4. Adjust R2 to minimize the signal seen on the scope at TP2 using TPl as scope ground.
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Tact Procsdura - P/N 175D304 Paga Two
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- 5. Set the test fixture switches as below: ~
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Turn cc von mode switch to normal. -
- b. Turn source select to +V.
- c. Turn range select to #V.' l l
- 6. Observe TP7 with DVM. Set TP7 to zero Volts + .lv MVDC
~
using R25. Use TP8 as ground.
- 7. Observe the output on the oscilloscope. The noise should
, be less than .5MV P-P.
- 8. '
Set the range select switch to 100MV and measure the input at test points A and C. Set the input level to +100HV 6 sing
- 9. Measure the output at TP7 and adjust the voltage to be the same as the input + .10 MV using R21.
- 10. Switch source select to external, set signal generator to sine wave and measure input at points A and C. Adjust level to 100MV peak. .
- 11. Observe the output on the oscillograph. Observable distor-tion or excessive noise are reasons to fail the unit. *
- 12. Recheck zero and gain adjustments and redo steps 6 through 11 if necessary.
C. Hypot Test
- 1. Connect together Al through A 6 and Bl through B6. .
- 2. Hypot between Al-6 and B1-6 at 2500V AC for one minute.
D. Aging
! 1. Install the module in a cabinet and run under load a i
i minimum of 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> at 140 F. The unit should have a 100MV input and the output loaded with a 1 meg. resistor.
- 2. Retest for zero and full, scale gain as in steps 6 through 11 of the procedure. Zero should be within + .2MV, gain within + .2MV. If adjustments are necessary for the unit to be within specifications, the unit has failed and must begin all tests over again.
/g,- 3 W Analog Optical M ator, Voltago Test Proceduro -- P/N 175D304 Pago Three
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' C' E. Acceptance
- 1. - The module'shall pass all the above tests prior to
~ acceptance. Should it fail at any time, it must begin .
all tests at'the beginning.
- 2. All failures shall be logged by serial number in the log.
An MIR form shall be used to reject all non-conforming materials. -
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Paragraphs b and c. Data to verify maximum credible faults.
The concerns addressed in the reference, paragraphs b and c, involve possible electrical faults on the output terminals of the device. The maximum credible fault that the device could see would be 120 volt ac supply power in the SPDS and isolator module cabinets.
Since our design provided for all signal leads to be routed through conduit with only signal leads and to be terminated on dedicated terminal blocks in the cabinets, a fault involving both a power source hot and common lead being applied directly to the output signal and return lead is virtually impossible. Therefore, no requirement or test for this event was specified in our qualification requirements. Our requirements were for electrical isolation between input-output circuitry, and' test results verified there is a minimum of 2,500 V RMS isolation between the input and output terminals. This test further demonstrated that a short or open circuit on the output signal terminals could not affect the input
- side since each side of the isolator is supplied operating power from separate sources. This precludes a fault on the output source affecting the input source.
Paragraph d. Define the pass / fail acceptance criteria for each type of device.
J Acceptance criteria dictated that the devices provide electrical isolation for an electrical signal having a range of 0-100 FN dc.
The accuracy of the signal must be maintained within 0.5% full scale from input to output during normal operating conditions.
Electrical isolation between input and output circuitry was to be demonstrated by applying 2,500 V RMS 60 Hz for one minute. The criteria required that the device not degrade or affect any lE device associated with the input signal source during normal operation or during any design basis event referenced in paragraphs e and f.
There is no requirement for the device to maintain signal continuity and accuracy during or af ter a design basis event. If the device did not maintain signal accuracy requirements and electrical capability during the test for normal operating conditions, it would be determined that the device had failed. Had the input side of the isolator been affected by any normal or abnormal event in such a way as to degrade the input source devices, the test would have been determined a failure.
Paragraph e. Environmental and seismic qualifications.
The equipment is located in a mild environment and certification by the supplier documents qualification of the devices for the following environmental conditons.
The abnormal conditons could exist for up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per excursion and will occur less than 1% of the plant life.
Environmental Conditions:
Normal-Temperature: Maximum 75, minimum 75, average 75 (F)
-Pressure: . At.m (+)
, Maximum 60, minimum 40, average 50 (%)
-RelativeHumiditg(:
-Radiation: 5x10 RADS), TID 40 year
, ' " -Vibration: SeismicLCategory'I (Active) ' "' - "
Abnormal'-Temperature: ' Maximum 1040F,' ' minimum '600F ' '
-Pressure: Atm (+)
-Relative Humidity: Maximum 60, minimum -10 (%) -
-Radiationi N/A .
Vibration; Seismic Category (Active)
Accident-Temperature
-Pressure: ~
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-Relative Hu$1dity: N/A
-Radiat' ion:
-Caustic Spray:
Verification that the Sequoyah E-Max electrical isolator assembly complies with seismic qualification, which was the basis for plant licensing, is demonstrats.d by these seismic qualification tests. One test was performed J
i on the typt of isolator installed at Sequoyah and the other test was performed on cabinet assemblies for TVA's Bellefonte Nuclear Plant. The test reports are applicable to the Sequoyah installation since the-cabinets are generic in nature. Test response spectra generated during j the scismic test of the isolator and the cabinet assembly show the response spectra of the isolator to envelope that of the cabinet assembly. .
, The response spectra of the cabinet envelopes that of the required floor response spectra of the installation. Therefore, the two seismic qualification t
test reports referenced herein demonstrate seismic qualification of the isolator assembly as installed at Sequoyah.
Seismic qualification for the isolator module is documented in Engineering Dynamics Seismic Qualification Report for E-Max Analog Voltage Isolator P/N 175C304 dated October 27, 1982. Seismic qualification for the cabinet assembly is documented in Wyle Laboratories' Report No. 58430 for E-Max isolator cabinets P/N 17502020-200 and P/N 17502020-300 dated October 9,1979. .Both tests were conducted in accordance with IEEE Std 344-1975 entitled " Seismic Qualification of Class lE Equipment for Nuclcar Power Generating Stations."
Testing of the module consisted of mounting the article on the shaker table at the test facility and subjecting the article to a resonance scarch and random excitation tests along three mutually perpendicular axes. The article was functionally energized during the tests.
4 ,
The resonance search tests consisted of subjecting the test article to sinusoidal swept frequency excitation in the frequency range from 1 to 50 Hz. Below 5 Hz, the acceleration input level was limited by the table displacement and ranged from 0.1 g at 1 Hz to 2 g at 5 Hz. From 5 to 50 Hz, the table was maintained at a constant magnitude of 2 g. The relative magnitudes of the table input motion and test article excitation were recorded as the frequency of excitation swept from 1 to 50 to 1 Ha in each of three perpendicular axes.
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The random excitation. tests consisted'of subjecting the test article to random table motion which is amplitude controlled in one-third octave increments from 1 to 40 Hz. The input spectrum was shaped such that the response spectra for OBE and SSE tests enveloped the required response spectra as specified in the Seismic Qualification Test Plan.
i Random excitation tests consisted of five OBE tests and two SSE tests in each of three mutually perpendicular.. principle axes.
- The test article was functionally. monitored prior to, during, anc!ufter completion of the tests. No interruption or. change.in the output voltage occurred as a result of resonance search or random excitation tests.
The test article was visually inspected during testing and at thd completion of tests, and no nonconformances were observed.
r [esting of the isolator cabinets consisted of mounting the' specimen on 2
the shaker table and subjecting the specimen to a resonance search and
, random excitation tests along three mutually perpendicular axes.
A steady state sinusoidal resonance search was performed in'each of the I three mutually perpendicular axes., The resonance search was performed in the frequency range of 1 to 35 and back to 1 Hz with an input of 0.2 l g. A frequency sweep rate of one octave per minute was used. One control and 10 response accelerometers were used to determine the resonance '
frequencies of the -test specimens. The output of each accelerometer was recorded on a direct readout recorder.
The test specimens were subjected to a seismic random motion which was e
amplitude-controlled in one-third octave increments from 1.25 to 35 Hz.-
! A 30-second recording of random signals was used as the input source.
The input signal was tuned with a bank of parallel one-third octave
- filters with individual output attenuators to meet the required response t- spectra.
Independent signal sources were used_for the horizontal and' vertical axes so that input motion phasing was random.
l Visual inspection of the test specimen upon completion of each test revealed no structural damage had occurred. '
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. Provide a description of.the measures taken to protect ~
the. safety sy'5tems from electrical 'inerference. .
t AEcept'able' pddt'icis' oY r'Ndtiiig' cSNie diid' grounding of cliEdits have been utilized in the design to minimize effects of radiated '
or' coupled ~ signals,on,theC, Input l7eads to the device.
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Electromagnetic compatibility esting of E-Max power supply.a,nd isolator cabinets were performed on equipment supplied,for TVA's Bellefonte Nuclear Plant.' ' The reports 'ar'e" applicable to the Sequo'yah installation since the equipment is generic in nature and has been certified as such by the manufacturer. , , , ,,,__ . , , , , , ,,
~
The test results are documented in Teledyne Ryan Aeronastical Environmental l
Laboratory test report 'No'. ' EL-80-Ol' dated ' January 30, 1980. The isolator cabinets were tested in accordance with requirements specified in TVA contract 79KJ2-822988 with E-Max Instruments, Inc. Two cabinet assemblies of different size were tested.
The large'and medium'isol'ator cabinets were' set up on a ground plane which was 3 ft. by 8 f t. by .043 in' ch sheet of copper. Signal ground and
' equipment ground were isolated within each cabinet and tied together at
, earth ground. Testing was performed on Class lE and non-Class lE power.
i The isolators, as applicable, were loaded as follows:
f DC isolator 250V de at 100 mA
=
DC isolator 120V de at 10 mA i DC isolator 48V de at 20 mA AC isolator 120V ac at 200 mA Analog Current Isolator .15 mA Input Analog Voltage Isolator .5V Input i
Each cab'inet was tested for electromagnetic compatibility as follows:
i j
Conducted EMI Transient Susceptibility Conducted RF EMI Susceptibility Radiated Transient EMI Field Susceptibility Radiated RF EMI Field Susceptibility Conducted Emissions Surge Withstand Capability Test Test data verifies that the devices passed all specified requirements.
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{ ATTACHNENT 2 i
SEQUOYAH NUCLEAR PLANT
- SAFETY PARAMETER DISPLAY SYSTEM
- .The Safety Parameter Display System (SPDS) consists of the block type j critical safety function status trees from the upgraded Westinghouse Owners Group (WOG) Emergency Response Guidelines-(ERGS). Documentation
- for these status trees " Emergency Response Guidelines Revision 1" were transmitted to Hugh L. Thompson, Jr.,' Director, Division of Human Factors Safety, by the Westinghouse Owners Group on May 4,1984. i i
j Sach tree uses several blocks containing questions with a yes or no
- output which leads.to a status. When a status tree branch is not satisfied, l- 'it directs the operator to an appropriate function restoration guideline. 1 1 .
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! Six generic status trees from the WOG ERGS'are attached. These trees i
! will be converted to plant-specific trees for Sequoyah. The different ,.
l- branches are color coded to show the operator _how serious any challenge }
i is to a critical safety function. The ordering of the trees also defines priorities. The colors in order of priority are: red (solid line), -
, magenta (dashed line), yellow (short dashed line), and green (double
- j. line).
- When any status tree is displayed, colors are shown in a designated j area, giving the status of the other five trees.
The critical safety function status trees have been developed using human factors principles. When the SPDS system is operational, the
- control room design review team will make a human factors. review on the j status tree displays.-
i j In addition to the critical safety function status trees, a radiation
! monitoring display is included. This display provides readings for 1mportant I radiation monitor points (including shield building, auxiliary building, steam i
! generator blowdown, and condenser vacuum exhaust) to supplement the-containment j critical safety function status trees. The critical safety function status-
- tre=a alanr with this additional radiation monitoring display fulfill the five j
- SPDS functions (reactivity control.. reactor core cooling and heat' removal . i from primary system, reactor coolant system integrity, radioactivity control, I and containment) as identified in Supplement l'to NUREG-0737. j i
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Number: Titi:: Rsv. Issut/Drts:
F-0.1 SUBCRITICALITY HP/LP, REV.1 1 Sept.,1983 8
GO TO FR-S.1 GO TO E' FR-S.1
,g NO N l
- PO.WER RANGE g LESS THAN 5% ,
E GO TO O FR-S.2 YES
. O O
E INTERMEDIATE NO NO RANGESUR INTERMEDIATE MORE
_ NEGATIVE ZERO OR THAN -0.2 DPM YES i NEGATIVE YES i
CSF i
l SAT NO SOURCE RANGE ;
ENERGlZED '
YES !
NO SOURCE I
RANGE SUR ZERO OR NEGATIVE YES r CSF SAT
Number: Titi : Rsv. Istur/DIts:
F-0.2 CORE COOLING HP/LP, REV.1 1 Sept.,1983 GO TO .
FR-C.1 R C.
CORE EXIT RVLIS NO
+ TCs LESS FULL RANGE THAN12000F GREATER YES THAN (2)
YES
CORE EXIT GO TO !
TCs LESS FR C.2 :
THAN7000F YES GO TO FR-C.2 NO AT LEAST ONERCP NO RVLIS RUNNING FULL RANGE YES GREATER THAN (2)
YES GO TO COOLING NO FR C.3 BASED ON CORE EXIT TCs i
GREATER THAN YES (1)or GO TO l , FR C.2 i
RVLIS DYNAMIC HEAD RANGE NO GREATER THAN (3)-4 RCP (4)-3 RCP YES (5)-2 RCP (6)-1 RCP
, CSF SAT l
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l Number: Titi2: R;v. Issu:/D:tI:
F-0.3 HEAT SINK HP/LP, REV.1 l
1 Sept.,1983 I RH
'~ ^
TOTAL ~ NO FEEDWATER ., l
_ FLOW TO y, n SGs GREATER l
THAN (2) GPM YES -
eeeeeeeeees GO TO e FR-H.2 f
NAR OW NO RANGE NO PRESSUREIN 4 I" ^
ALL SGs LESS ONE SG ATER THAN YES YES
- '*******@?eJi NARROW NO RANGE LEVELIN ALL SGs LESS THAN (4)% YES
- @?eJi NO PRESSURE IN ALL SGs LESS THAN (5) PSIG YES gee.e. $pg;;
NARROW NO RANGE LEVEL IN ALL SGs GREATER THAN YES (1)%
7 CSF SAT
Number: Tills: Rev. Issue /Date:
HP/LP, REV.1 F-0.4 INTEGRITY 1 Sept.,1983 E
a w '
E E
8 0
c:
- T1 T2 COLD LEG TEMPERATURE :
FR-P.
k
- ALL RCS PRESSURE NO m n 335 clllI sing GO TO FR P.1
- COLD LEG TEMPERATURE
[
POINTS TO E
RIGHT OF YES LIMIT A ALL RCS NO GOTO i COLD LEG e9 FR-P.2 1 TEMPERATURES g GREATER THAN g
, (1) F YES ALL RCS NO TEMPERATURE COLD LEG DECREASElN NO TEMPERATURES ALL RCS COLD GREATER THAN LEGS LESS (2)0F YES THAN 1000 FIN THELASTSO YES MINUTES -
CSF SAT R-P.
ALL RCS NO COLD LEG TEMPERATURES I
GREATER THAN (1)0F YES RCS PRESSURE NO LESS THAN p COLD '3 88 GOTO OVERPRESSURE FR P.2 LIMIT YES "O
RCS
_ TEMPERATURE j GREATER THAN p) F YES I CSF SAT I
, CSF SAT
Numbir: Titis: Rev. Issun/ Dita:
F-D.5 CONTAINMENT "
GO TO FR-Z.1 CONTAINMENT
. PRESSURE LESS THAN (1) PSlG YES GO TO MMMMMM FR-Z.1 s
CONTAINMENT NO PRESSURE
~~
LESS THAN (2) PSIG GO TO YES gm m m FR-Z.2 I
i 0
CONTAINMENT SUMP LEVEL LESS THAN
() YES GOTO
- ..t FR.z.3
{
NO CONTAINMENT RADIATION LESS THAN (4)
YES CSF SAT
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Number: Titis: Rsv. Issus/ Dita:
HP/LP, REV.1 F-0.6 INVENTORY 1 Sept.,1983 l l
ggg GO TO g FR l.3 RVLIS NO INDICATES UPPER HEAD FULL (3) YES
... GO TO FR l.1 NO
- [& Snag"a . . . . . . . . . . . . @ a Tp THAN (1)%
YES .
PRESSURIZER LEVEL GREATER THAN (2)% YES ... GOTO FR-\.3 g
9 RVLIS NO INDICATES UPPER HEAD FULL (3) YES l
, CSF SAT W -
= =
v-w--m * - ~ - - _ _ _ - - -- _ - _
Numbir: Titis: gg) Rav. issus/ Data:
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F-0 SAFETY FUNCTION HP/LP, REV.1 STATUS TREES 1 Sept.,1983 C
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-FOOTNOTES F-0.2 CORE COOLING - _.
(1) Enter sum of temperature and pressure measurement system errors, including i allowances for normal channel accuracles and post accident transmitter
- errors, translated into temperature using saturation tables.
(2) Enter plant specific value which is 3-1/2 feet above the bottom of active fuelin
. core with zero void fraction, plus uncertainties.
(3) Enter plant specific value corresponding to an average system void fraction of 50 percent with 4 RCPs running, plus uncertainties.
(4) Enter plant specific value corresponding to an average system void fraction of 50 percent with 3 RCPs running, plus uncertainties.
(5) Enter plant specific value corresponding to an average system void fraction of 50 percent with 2 RCPs running, plus uncertainties.
(6) Enter plant specific value corresponding to an average system void fraction of 50 percent with.1 RCP running, plus uncertainties.
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F-0.3 HEAT SINK (1) Enter plant specific value showing SG level just in the narrow range, including 1 allowances f or normal channel accuracy, post-accident transmitter errors, and I reference leg process errors, not to exceed 50%
, (2) Enter the minimum safeguards AFW flow requirement for heat removal, plus allowances for normal channel accuracy (typically one MD AFW pump capac-ity at SG design pressure).
(3) Enter plant specific pressure for highest steamline safety valve setpoint.
(4) Enter plant specific value for SG high-high level feedwater isolation setpoint.
(5) Enter plant specific pressure for lowest steamline safety valve setpoint.
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_ _ _ _ _ . . __ _ _ _ .__.~._,_._. ..__ _ ._ ___ f .__
Number: Titis: CRITICAL Rsv. Istut/D:ts:
r .
F-0 SAFETY FUNCTION HP/LP, REV.1 STATUS TREES 1 Sept.,1983 i
I FOOTNOTES (Continued) l F-0.4 INTEGRITY - l (1) Enter plant specific value corresponding to temperature T1. Refer to back-ground document for status tree F-0.4.
(2) Enter plant specific value corresponding to temperature T2. Refer to back-ground document for status tree F-0.4.
(3) Enter plant specific temperature setpoint below which cold overpressure pro-tection system is in service.
F-0.5 CONTAINMENT e (1) Enter plant specific containment design pressure.
(2) Enter plant specific containment high-2 pressure setpoint.
(3) Enter plant specific containment water level just below design flood level minus allowances for normal channel accuracy.
(4) Enter plant specific value corresponding to radiation level alarm setpoint for post accident containment radiation monitor.
F-0.6 INVENTORY (1) Enter plant specific pressurizer high level reactor trip setpoint. i (2) Enter plant specific pressurizer low level letdown isolation setpoint.
(3) Enter plant specific instrument channel and setpoint which indicates upper head is full.
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.: : w .+ c . ... . , " . . . . '
t ATTACIDIENT 3 -
SEQUOYAH NUCLEAR PLANT
, SPDS VERIFICATION AND VALIDATION PROGRAM The-letter from L. M. Mills to E. Adensam, dated January 4, 1984, gave
-details on the.TVA V&V Programi ' Additional requested information is as follows. -
L With the' block 4iype t status tree displays, computer points' arc displayed below a block, where applicable. _The point and the yes/no outputs will , , .
be shown as bad or suspect when internal sof tware checks show the data 4
to be questionable. There are four quality classifications:
a'. Good data.
. b. Sensor data inconsistent with the majority of redundant sensor I
values. 2
- c. Data evaluated as bad because it is outside the process sensor or data acquisition system span, or because hardware checks indicate a malfunctioning input device.
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- d. Data which is operator entered.
Further validation of data is accomplished by field verification tests which are performed after system installation. This verifies that the
! system will properly display the input signals and that the inputs are j connected correctly.
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ATTACHMENT 4 SPDS SEQUOYAH NUCLEAR PLANT UNREVIEWED SAFETY QUESTIONS-A 10 CFR 50.59 evaluation has been performed, and .TVA does not consider the SPDS an unreviewed safety question.
On Technical Specification Improvement, NUREG 1024, NRC referenced statements by the Atomic Safety and Licensing Appeals Board (ALAB-531 in the matter of Portland General Electric, ET AL Trojan Nuclear Plant). In part, the Appeal Board stated:
Technical Specifications are to be reserved for those matters as to
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which the imposition of rigid conditions or limitations upon reactor operation is deemed necessary to obviate the possibility of an event giving rise to an immediate threat to the public health-and safety.
Inoperability of the SPDS would not pose an immediate threat to the health and safety of the public. TVA does not plan to submit technical d-specifications for the SPDS. This decision will enhance regulatory performance in regards to compliance with existing technical specifications. . -
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ATTACHMENT 5 SEQUOYAH NUCLEAR PLANT
. . ns ,..s., , 3-y . , . . ,
SAFETY PARAMETER DISPLAY SYSTEM t e, t r u i v u , . , :. r p, ; s n o...
IMPLEMENTATION SCHEDULES
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a TVA has installed'the SPDS/ Technical Support Center (TSC) computer system hardware before startup following the second refueling outage for each unit. The-SPDS, including computer systems software, will be operable with procedures and users manuals verified an'd' validated with
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operators trained no later than -September' 1985 for unit 1 and October 1,985 for unit 2.
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