|
---|
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 ' . A TENNESSEE VALLEY AUTHORITY' CHATTANCOGA. TENNESSEE 374ol 400 Chestnut Street Tower II (
January 6, 19 %
w
\ i' Director of Nuclear Reactor Regulation Attention: Ms. E. Adensam, Chief g y p~
gg Licensing Branch No. 4 Division of Licensing p). Q,L "c:: . ~~,~E.
U.S. Nuclear Regulatory Commission h Typg e [J Washington, DC 20555
Dear Ms. Adensam:
In the Matter of ) Docket No. 50-328 Tennessee Valley Authority )
As required by item 2.C(3).a of the Sequoyah Nuclear Plant unit 2 operating license, TVA must have NRC approval before making a " major modification" to the initial test program. Startup test SU-9.5, " Rod Group Drop and Plant Trip Test," is required for unit 2 at the 50-percent power level as part of our initial test program. In my November 3 and December 9, 1981 letters to you, we requested approval to delete startup test SU-9.5 from our startup
- l. test program. As requested by members of your staff in subsequent tele-phone conversations and as a result of your rejection of our request by the December 28, 1981 letter from R. L. Tedesco to H. G. Parris, we have revised the justification to include additional information for deletion of the test. Please reconsider our request. Based on the enclosed infort:a-tion and other information provided to the NRC, by TVA and Westinghouse, during the review of the " dropped rod" issue, we believe that startup test SU-9.5 can be deleted without any degradation of the startup program for our Sequoyah Nuclear Plant unit 2.
Please provide us a response to our request as soon as possible in order to delete the test at the appropriate point in our startup test' program.
Very truly yours, TENNESSEE VALLEY AUTHORITY
\
L.M. Mills,Ma&h ,
nager Nuclear Regulation and Safety Sworn to and subscribed bef ore me thisb __ - day othugMte _1982 b
Notary Pu511c C)
My Commission Expires ~
-; I I (
Enclosure
~
8201130267 8201Ois P PDR ADOCK 05000327 F P P DR !h An Equal Opportunity Employer
ti c e> .cd ENCLOSURE JUSTIFICATION FOR DELETION OF STARTUP TEST SU-9.5
" ROD GROUP DROP AND PLANT TRIP" SEQUOYAH NUCLEAR PLANT-The follcaing paragraphs present information to demonstrate that the core -
designs for SNP units 1 and 2 are identical. Figures 1 through 4 (attached) show that the core loading and control patterns are identical.
In addition, the rods to be dropped for unit 2 (P-4 and D-2 in Shutdown Bank A) would be the same rods selected ~ for unit 1. These rods are l selected because their proximity to the excore detectors will make detector response more limiting for the 3/4 trip criterion of SU-9.5. Attachment 1 provides a detailed explanation of why the rods chosen represent the most limiting case for verification of the negative rate trip function.
As further evidence that the core designs are identical for SNP units 1 and 2, the design boron end points for the follcwing rod configurations are identical.
(1) All Rods Out (ARG)
(2) Control Bank D In (3) Control Banks C+D In (4) Control Banks B+C+D In (5) Control Banks A+B+C+D In (6) All Control Banks In + Shutdown Bank D In (7) All Control Banks In + Shutdown Banks C+D In (8) All Control Banks In + Shutdown Banks B+C+D In (9) All Control Banks In + Shutdown Banks A+B+C+D In (All Rods In)
Additionally, the design differential boron worths and isothermal temperature coefficients are identical for units 1 and 2 for the following rod configurations.
I (1) All Rods Out (AR0)
(2) Control Bank D In (3) Control Banks C+D In (4) Control Banks B+C+D In (5) Control Banks A+B+C+D In (6) All Rods In The design differential and integral rod worths are also identical. This indicates that the axial flux shapes in the two reactors are the same.
Further, the assembly power predictions for the following rod configurations for units 1 and 2 are identical:
(1) ARO .
(2) Control Bank D In (3) Control Banks C+D In (4) Control Rod D-12 Rejected (from flat zero power and 30% power)
(5) Control Rod D-12 Dropped (50% power) ,
4
- g. s E This provides a strong indication that the radial flux shapes are the same for the two reactor cores.
Since the core designs for the units are identical as previously illustated and the same rods would be dropped for the unit 2 performance of SU-9.5, the test results are expected to be the same.
The objectives of startup test SU-9.5, Rod Group Drop and Plant Trip, i.e.,
to confirm that the negative rate trip circuit will trip the reactor as a result of dropping two rods and to obtain preliminary data for systems response to plant trip before performing the turbine trip and reactor trip from 100 percent, are fulfilled by other testing requirements. Therefore, there is no reason to perform this startup test on Sequoyah unit 2.
l The negative rate trip circuitey receives a channel calibration (see IMI .l PRM-CAL Section 5.2.7, attachment 2) and a channel ibnction test (see IMI-92-PRM-FT, attachment 3) before initial entry into MODE 2 operations. In addition, the channel calibration and channel functional test are repeated i
at 18-month and 1-month intervals respectively. These tests verify the ,
operability of the entire negative rate trip circuitry before the l production of any reactor power.
SU-8.3, " Static RCCA Drop and RCCS Below Bank Position Measurements," l
, obtains each power range detector's response to a single rod below bank j position rod dropped rod configuration. In SU-8.3, a single rod is l inserted into the core, operating at 50% power. Periodically (at the time l the flux deviation alarm is received and every 25 steps further inserted on
- l. the rod) each power range detector's response is recorded. This power
) range detector response check, coupled with circuitry checks made in the l above channel fbnetional and channel calibration tests, gives us confidence l
the negative rate trip circuitry is operable up to and including the power range detectors.
The second objective of startup test SU-9.5 is verified in startup test SU-1.2.A, Shutdown From Outside Control Room. In this test, the plant is at approximately 30-percent power then tripped from outside the control-l room. Subsequent to every plant trip, our operators verify control rod,
[ pressurizer, steam generator, and reactor coolant system status. SU-1.2.A can be performed at a higher power level if NRC still requires additional testing. In addition, verification of the following acceptance criteria of SU-9.5 pertaining to system response can be obtained with SU-1.2. A.
- 1. All fbil length RCCAs shall have released and bottomed.
- 2. The pressurizer safety valves shall not lift.
3 Steam generator safety valve shall not lift.
- 4. Safety injection is not initiated.
A new acceptance criterion to startup test SU-1.2A could formally verify the operability of the entire negative rate circuitry including the power-range detectors. The acceptance criterion would be:
Verify each power range negative rate bistable is tripped.
_m._________ _ _ _ _ _ _ _ _ - - . - - - - - - "
a '.
,0 - , 1 h .s ..
In . conclusion, -the repetition .of startup test SU-9.5.on Sequoyah. unit 2
- should not be required since
- 1. The core designs - of Sequoyah units 1 and 2 are identical. - Therefore, .
the same results.as were obtained from the performance of the test on-unit.1 would be obtained. .. __
e
- 2. All the objectives of startup- test SU-9.5 are fulfilled by .other I testing requirements.
2 t
$8 '
l-i ii .
I 4
2 I
t a
ij -,
F/Gs /:
CORE MAP WlTW CONTROL ROD PDSITI i R P !! - !! L. ji-35 1L
} K G F E D c
.! h A I
N i ll 1809 1-
~
^ ^
~
3 l [E
. . . .}
o S S
. \,- o i..]_],
( S A
3 (D
.__ OD OD A me m
6 A (c) W @ es o
c 8 1 D D I~ 9- .
10 A @ V 3 A 11 !.._o8 S i
!c].
- c. .
12 S A
I S
A E 13 .
' (]_ f, j 11 6
~ A '
15 .
n_lili -
0 N-h2 ygg . . H-32,N-36 FUNCTION UNIT 1 NUMBER OF ROD CLUSTERS SHUTDOWN DANK S
A 8 SHUTDOWN DANK S 'l. I '*P '
p B SffUTDOWN BANK S O S D n *- 4 4 CONTROL BANK A ;l! **
- N)! CONTROL BANK-l , ,, j 4 B
CONTROL BANK 0
.C -
CONTROL DANK D
, D
' 9 i' . ,
SNP-75-Rt3
F/G(2 -
. CORE MAP. WlTH CONTROL ROD POSITIOh _
~ '
A B C D E F G H I X~ L M- N P R N k3 .
O
_. g_hi 15-
8 A O /\
A
/0\ C
'u l
/D\ A 13 i
lso i sg sg
[s))
i...,
12 8 s
4
)
- t. _C ,
5 lig_ _ _ ,
lC .
C i "o
e e v e 3-90 8
3 y
- T -- -
6 A s y @ A 5- 8
' _0 ,i bl
- ,,C ,
~
la _ S A D 0 A '.'
L - .. .
2 A 'l ~
1_
, 9 ljorib N-h2 160 N-kh IJgIrp NUMOER OF-FUNCTION ROD CL USTE R S SHUTDOWN DANK N-32 ,a S^
SHUTDOWN DANK S D
N-36 s SHUTDOWN BANK S C 0 SD
/
CONTROL BANK 4 4 -
A ,
', , ,, 4 CONTROL BANK 8 8
'f CONTROL BANK .C i l 0 CONTROL BANK D '
9 i
12
-SNP-7M-pin
~
.3 TVA Sequoyah U,iit 1, Cycle 1 Figure 3 a P N M L K J H G F E D C B A C33 C2 C47 C27 C1 C7 C12 1 l
- ; <.. . . . . . . .e en m
- .r.w I
C30 C15 m.h.,, 88 7A31. B61 A47 'B20? A12 4B29; C18 C4 2 1
,' w u.v. e
.s w w -vg m :cri n l l
. - i.
- .:s ,i.;O ,. s m 837, A67. 'B5's -A30 ;B6BL C11
.mn ci & wt. W.:$
- C40 C24 B41~ A32 < B 14' .A70 C45 3 mc.c u M. .+, ; r.r: - .pW ww. ev . mi.
i .. ..?. . ~:X aK. . i*: :Va . u. sn' .~?'." tw;n '
:w. t. - J h i CD B35' ,1A55 031l. 'A45 3:q 16 A15 1850; A71 IB10I A72 'B19L C38 4
, ; y < :g : .+.y
. - 4 p -v sy:p :,.po :y.; .. g.74.s
. .,. . : n n. .. ..m v. . . v. . . m:n ;.=A 44w 1
c42.' B34
.. n .,a: . ny
- A69 [B18 A56 ~ B71.. ..All B24 .A41 868! 'A68 pn) B58~; A57] 6B48' C37 5
4 .
- a ;m.a . :::a '2 m :.u: : %'::. t. *c5 vM1 e:& Wm2 %w Te.T
! \ . . , . ro uc .. , . es ,y : :, + .p. ; .co..y me .an.w . n~s, p:% :,.a.
rr l B54 'A17 B4 n.3.g .lA13. lB 5. AS4 ;B72 A60' ;?B43' A7d B27 A29'
, i.p.U; w.w; C13 6 C10 'l. :.A42
. . -:. :ua. p s .u , .m - km . ,r , Sen
. . .:w w i
.r x.t w ac .>.v; . .=. 7, 2::?L w.y :s .um . tm ~
g .- : u v.
a C6 I B'40' A21' B49. A3,:8 18255 .A44 A43 A73 n.er:; .%
B42 . A52 .
m.,; ,p,.'y B .: f f .A39 C28 7
- .>,- .,d , .y wro
- 7s 1w my )B3i1
- s.
l , , . . y .
- v.;. . c.r:a . v +, . t.:. :m. :: n :% y:r =:w; w =y :yw 0: C9 A49 BSS- . A6 B33: A64 A65 A62 A22 A26 iB22; A27- ,B39- .A61 C14 3
- .:. s . : .. V. m ,d wh .n i
. . 7g .,a v u.s - S. .n. . a.. u. .,:p;3 9% es. ; qu exn,1 C29 338: _A14 B91 A30 B61:' A58 A28 A46 B591 'A53 .B2 d A51 ;B32t C34 9
, o., . w # . a:*, s:.s + e. q
.< ~~.- :w.sg :n wCe e : v:+ .-w r w
- /45 . :!W 4.%:_ ... w; pNa
.,.; a a.. ..:.... .. J. ;c:4 .....i ?
?,9:-
3 CAS A34 B69' A59 B44; A19 ;B13 A33 y. 811 ,n; A10 .B60 23 E47 A40 C10 10
..: u a. - .; o. . g + ..y..,y A. n.sm:.{ ,w
,. ,o .. ..
-e :% .r 7.. ., :...>. n..% 1. .- #x ....u ., w. e C46 B6 A8. 7B52 A9 i iB23< A50 BB3; A60 B57/ A2 #28j A16 iB53. C44 11-m .w. - ;;;; ym ;;; - 2;y ;e: sypy .m:La 9 74 :. . . gp 1.p.,p .
..v. .
. m.. , :.s :, n . . . . ;ra.n7 ..c., ...;.. .
.x . .s; ~ ;
C35 B70 A25 B46 ;. A20 B38. A5 J, 1865,. A3 .' 1B64;: A35' 'B12i C19 12 a- .y,( 3;- rpw +- qqp gg g; .c; ;. 9 3,q:;
s.; .
C31 C43 B1- A18- 1 A4( $b25 :
A,k7 .Y5Y.a. Ak N'$k$ C26 C32 13
. :t ; s:,c. ~~< u :.:, :. - n. .M y u ,
.,w re.n, s. *m c ..
C16 C36 855- A03 B45 A24 B30l A48 .B17. C25 C22 14
. c ,. - . . a- : . . - .w C48 C41 C17 C39 C21 C5 C3 15 0
A: Region 1 (2.1 w/o)
?B[ Region 2 (2.6 w/o)
- Removable Fuel Rod Assembly C Region 3 (3.1 w/o)
-w - . . .
1 .
)
[ figure *4 Sequoyah Unit 2, C cle 1 R P N- M L K J H G F E D C B A g ..
t 3
4 l
N33 N02 N47 N27 N01 N07 N12 1 i
-?,. .
N30 N15 M08. L31 MG7 L47 M201 L12 M29. N18 N04 2
, aae yf ; eg.e syy
. w::
. . . ., r q u . . :y: . .. . . , . . . , 7 ;w Na0 N24 M41 L32' M14 L70 M37J L67' M05 L30 M6G: N11 N45~ 3
.m. wg, xm+ ,
, vy.c g
. n. ,
.r ,.y ,:w, i vw. .1,, gn:
N08 M35 L55 M31, L45 -M10 L15 M50 L71 M10' L72 M19 N38 -4 A -
..;p c.ce :wn weg -
wp w.m[.
.c . ,1: r ::nz p.w . ,u :. n m.t..
i N42 M24 L69 .,M18 .L56 M713 L11 ,M2{ L41. M 8 e,.6 LS8 L57 M48,. N37 5
..., , i.;-3 (.ms
. +.,_ - g ~ . , . .,M,58 ,.: . . . .
w:
d, m n.r. : .
r . .n ve- . ~q . r.m. i .
l N20 L42 M54. L17 M04: L13" M15 L54 M72. L6G M LO7. iM27 L29 N13 6 1
e c:;e. w r., e.43.
.q , m;+
t N'o t v.s tt -
..: ,= y. .w ?.< . r e i I NOS 'M40 L21' .M49 L38 L44 L43 L73 .M42 L52 M07
.'M25'. L39 M03 N28 7 i 4
v,,r :-
--i
>:y y .u;-w i
.m i r.v..- -
- p. :> gm it NC9 L40 MS,G LOG. M33 eg.
LG4 .
LG5 LG2 L22 L26 .M22
- w. ; '
L27 iM39
.nu LG1 N14 8 i t . .- - .
a w. ,, .. . .wc..
i N29 M3G L14 MOD L36 .MG1' L58 L28 L46 M59. L53 M02 L51 M32: N34 9 s
-=--
+v ., ca: ;
. :::,o. .
w w i j
,. . .w - . .e ,n . , . .. m.s ,
1 N23 L34 M69 L59 M.,44' L19 .M13 L33 M11. L10 M60: L23 M47 L40 N10 10 1
-, :: ;n w . ~ <.y : .~ - , ,n
~
[bh 2
N46 M06 LO8 .M b2 Ld9 '$f L50 tNb L60 [57' LO2 N44
. c. : w.d5 . rc ;?.. . v . kk.b a .u:
L1G 39 11 N35 M0 L25 MkG L2h hk[b LOS $ik,>,.LO3 h,$b.k L35 N19 12 c;
-> . 3M r ti r - 3.( W o ... .
N31 N43 M01 L18 M21 ~ LO4 MG2' L37 M51 LO1 .M2G N2G N32
.+ J3 me .. m ,
N16 N36 M5[ LG3 N45 L24[5bb' L48 $1kk N25 N22 14 w ... m. -
N48 N41 N17 N39 N21 N05 NO3 15
~
0 L Region 1 (2.1 w/o)
Mi wy Region 2 (2,G w/o)
N Region 3 (3.1 w/o) sA ,
m
.9 - -
Ik'y , ,
ATT.'CHMENT 1-SELECTION OF DROPPED RODS FOR RATE TRIP TEST The purpose of the dynamic dropped rod test-is to verify that a reactor trip will occur via the Negative Flux Rate _ Trip Protection System for two dropped rods. Therefore, the rods selected for the test should be rods for which a reactor trip is not predicted based on conservative assumptions.
In aclecting these rods, four factors must be considereo:
- 1. The trip logic of the Flux Rate Trip
- 2. The arrangement of the rods in groups and banks 3 The location of the rods with respect to the excore detectors, and ,
l; 4. The worth of the rods.
Each of these is discussed belcw.
The Negative Flux Rate Trip function detects changes in power via the four power range excore detectors located on the core diagonals around the reactor vessel. Each power signal feeds a rate-lag unit which indicates any rapid change in nuclear power. If the rate-lag unit output exceeds the setpoint in two out of four channels, a reactor trip is initiated. To meet the single failure criterion, one channel must be assumed to fail, so the I logic is effectively three out of four. Therefore, the rods selected for the test must predict a trip on less than three channels.
\-
Rod Cluster Control Assemblies (rods) are evenly distributed throughout the core as shutdown rods and control rods. The shutdown and control rods at Sequoyah unit 2 are organized into seven banks, each bank consisting of symmetrically distributed rods. A bank may be further subdivided into two groups which also have symmetric core patterns. The dropped rod accident is assumed to be initiated by a single electrical or mechanical failure
{ which causes rods from the same group of a given bank to drop to the bottom of the core. Therefore, the rods selected for the test must be from the same group. Rods from different groups are not considered since it would require more than one single failure to cause them to drop.
The location of the dropped rods with respect to the excore detectors is important both because of the trip logic which limits (to less than three) the number of detectors which can "see" the dropped rods sufficiently I
enough to trip their bistable and because the response of the detector is dependent upon the power in the assemblies nearest to it. The majority of
- the ' signal produced at any excore detector is obtained from the assemblies on the periphery of the core next to the detector. The detector is less sensitive to power generated at the center or on the opposite side of the core. A rod dropped in a peripheral assembly would produce a large signal change at the nearby. detector; however, the remaining three detectors would see a much smaller change. Dropping two peripheral rods would produce '
strong signals from two detectors (the group requirement prohibits rods from being dropped in the same quadrant) while producing smaller signals at the remaining two channels. Therefore, dropping two rods which are next to two excore detectors will cause two bistables to trip, which is less than the three channels required to meet the single failure criterion.
4 v s m== m. _ _- z . w m e rt c r ~:-
/ . .
The final item of consideration is the worth of the dropped rods. Rods which cause a large amount of negative reactivity insertion will produce larger changes in power which will be seen by all the excore detectors.
Internally located rods have the highest worth since the majority of the power is produced in the inner regions of the core. To minimize detection of cropped rods for trip, it is desirable to select a pair of rods with a low worth.
In selecting rods in locations P-4 and D-2 of shutdown bank A, all of the above criteria are met. These rods are from the same group and are located in the outer edge assemblies of the core next to two excore detectors.
Their location serves to limit the response of the third detector required for the single failure and also minimizes worth.
l l
1" i
s ,
- c. .
s .
ATTACHMENT 2 CHANNEL CALIBRATION PROCEDURE FOR NEGATIVE RATE TRIP CIRCUITRY P
s s
s
/ L qi f
f
~ ~ ^ ^ ^ ^ - --- - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
i
{ . , . *, . ~
, SQNP *
,,i. , .IMI,-92-PRM-CAL . -
1
, Page 18 of 25 l; '
Riev. 8 J -l'
. ( .
i t l g 5.0 PERFORMANCE OF WORK - continued .
. t k 5.2.6 (continued) i t i
- y. 37.
e' If XR-92-5001 AFI'indicites the desired value 10.-5% record on data sheet "as left". Otherwise the recorder may require
- y. * , adj ustment. '
- .;,e , 38. Turn TEST SIGNAL pot R301 fully CCW and repeat steps 36 and yt '37, but use TEST SIGNAL pot R302 and TP302.
W} 39. If the remaining sections of the CAL are not to be completed, go to section 7.
- 1 I
,5.2.7 POWER RANGE RATE CIRCUIT NM311 AND BISTABLE RELAY DRIVER j NC301 AND NC303 ADJUSTMENTS. To Adjust Power Range Rate i
circuit NM311 and bistable relay drivers NC301 and NC303 ,
located in Power Range A drawer assembly proceed as outlined in this section.
, 1. Bypass the Power Range channel being adjusted, as.
specified in paragraph 5.2.1. Place OPERATION SELECTOR switch S303 in the NORMAL position.
~
- 2. Deenergize Power Range 3 drawer by removing the INSTR. !
l 3 .
POWER fuses. -
t s
- 3. Disconnect high voltage connector plug P353 from
(
- - Jack J353. ,
. I I
, 4. Disconnect DET A and B SIGNAL plugs P.351 and P352 from
! i Jacks J351 and J352, respectively.
. 1 > >
- 5. Connect a jumper from GND NM311 TP1H to INPUT NH311 TP211.
! 6. Connect a DVM between test points OUTPUT TP3H (+) and GND TP1H (common) on NM311.
! 7. Rotate DELAY ADJ control NM311R5H fully counterclockwise.
j (Initial calibration only.) ..
- 8. Energize Power Range B drawer by replacing the fuses.
i i !
l .
j ,
. i
~
i -
l, j
4 l.
- (
6
.i j.
SQNP
~
^
,, . It!!-92-PRt!-CAL ' *
Page 19 of 25 3 Rev. 5 r
, 5.0 PE,RFORt!ANCE OF WORK - continued 'I i i i
1 5.2.7 (continued) :
\
1
! , 9. j Record the "as.found" value on data sheet N-41. If the DVM does not indicate.0.000 1 0.005 volts d.c. , adjust
(
i g ' ZERO ADJ potentiometer R711 until the DVi! indication is l within tolerance. Record "as left" value on data _ sheet.
N-41.
- 10. Lock ZERO ADJ potentiometer R711.
, 11. Remove the jumper from TP1 and TP2 cn Nt!311.
, . 12. Rotate OPERATION SELECTOR switch S303 to the DET A & B position.
! 13. Adjust DETECTOR A and B TEST SIGNAL potentiometers R301 '
) and R302 for 2.083 i .005V as read by a DVM connected be-j
- tween LEVEL TP301 (+) and GND TP305 (common) and LEVEL i TP302 (+) and GND (common). ? -
l
- 14. Reconnect the DVt! between test ' points OUTPUT TP31! (+) and GND TP11! (common) on NM311 and wait E 30 sec. for the -
1 indication to reach a steady state condition. Record "as
{ found" on data sheet N-41.
- 15. If the DVM indication is not 0.000 1 0.005 volts d.c.
adjust BALANCE ADJ potentiometer Nt!311R311 until the .
! DVM indicati'on is within tolerance. Record "as left" value on data sheet N-41.
- 16. Lock BALANCE ADJ potentiometer R311.
- 17. , Connect the strip chart recorder between test points 'Grn) 1
' TP1 (common) and OUTPUT TP3 (+) on Power Range Rate cir- i
, I cuit module NM311. Operate strip chart as necessary to obtain the information required in step 18 (2"/sec, recorder '
j set on 10 volts).
l' 18. Introduce step inputs as described in steps 21 and 23 ,,
below. Record "as found" value on data sheet N-41. Ro-tate DELAY ADJ potentiometer Nt!311R51I until the time taken for the output of assembly NM311 to decay to 37%
of the peak value equals the desired delay time ,(3 'l.3 sec.). Record "as left" value on data sheet N-41.
i t .
I l' ,
I :
. s t .
, , t.
. ; a L
. . --. .. : _ _.~ --
) s .
,l. ' f' -
SQNP * *
, 1NI-92-PRM-CAL i
6 Page 20 of 25 p Rev. 10 4
3 5.0 PERFORMANCE OF WORK - continued
[ ,
5.2.7 (continued)
- 19. After the DELAY ADJ cont ol is set, lock the pote'ntiometer.
- v. NOTE: If the "As Found" values for the + 25 volt power
- s- supplies and the bistable in question were found in 4 tolerance, Sections 5.2.7.20 through 5.2.7.31 need
.s not be performed. For this case write the "As Found"
+ data in the "As Lcft" data space. ;
, 20. NEGATIVE RATE TRIP setting is.3% of full power. Rotate OPERATION SELECTOR switch to the A & B position.
, 21. With DET A TEST SIGNAL potentiometer R301 fully CCW, ; ,
adjust DET B TEST SIGNAL potentiometer R302 for a voltage '
reading of 4.167 i .005V at TP 306. Adjust DET ' A TEST '
SIGNAL potentiometer _ for a voltage reading of 4.417 i .005V at TP 306.
- 22. For initial adjustment only, rotate TRIP ADJ potentiometer R4E and LOOP ADJ potentiometer R11E on Negative Rate Trip, bistable relcy driver NC301 fully CW. , t NOTE: The. LOOP ADJ potentiometer R11E should be '
fully CW and will remain fully,CW.
j 23. Rotate OPERATION SELECTOR switch S303 from the DET A
- , & B position to the DET B position to introduce a negative atep signal. .
- 24. Adjust TRIP ADJ potentiometer NC301R4E CCW in small .
increments. While holding RATE H0DE switch S304 in
the RESET position, rotate OPERATION SELECTOR switch
. S303 from DET A & B to DET B and back to DET A & B positions after each increment until the bistable l just trips as indicated by NEGATIVE RATE TRIP lamp ,
DS309 momentarily lighting. Allow 15 seconds to clapse each time before switching the OPERATION SELECTOR switch.
Record "as left" value on data sheet N-41. '
- 25. After the trip point is located, lock TRIP ADJ potentio-( , meter NC301R4E.
. 26. POSITIVE RATE TRIP setting is 5% of full power. Rotate -
0FERATION SELECTOR switch to the A & B position.
27.
~
With DET A TEST SIGNAL potentiometer R301 fully CCW, verify the voltage at TP 306 is 4.167 .005V. If not,
required voltage is obtained. Adjust DET A TEST SIGNAL -
potentiometer for a voltage reading of 4.583.1 .005V ,
i ! at TP306. . i-1 m
.i i
- ( jg.,
, i a :. m
~ ,
i *
' SQNP
! * '. ,c . - IMI-92-PRM-CAL Pag'c 21 of 25
=
Rev. 8
)
,- 5.0 PERFORMANCE OF WORK - continued ,
8
, 5.2.7 (continued) 6
- 28. For initial adjustment only, rotate TRIP ADJ potentiometer R4E and LOOP ADJ potentiometer RllE on Positive Rate Trip bistable relay driver NC303 fully t'W.
' NOTE: The LOOP ADJ potentiometer R11E should be fully CW and will remain fully CW. .
- 29. Rotate OPERATION SELECTOR switch S303 from the DET B position to the DET A & B position to introduce a
, positive step signal.
- 30. Adjust TRIP ADJ potentiometer NC303R3E CCW in small increments. While holding RATE MODE switch S304 in the RESET position, rotate OPERATION SELECTOR suitch S303 from DET B to DET A & B and back to DET B po-sitions after eacit increment until the bistable just trips as indicated by POSITIVE RATE TRIP lamp DS308 momentarily lighting. Allow 15 seconds to elspee cach time before switching the OPERATION SELECTOR switch. Record "as Icf t" value on data sheet N-41.
- 31. After the trip point is located, lock TRIP ADJ potentio- '
meter NC303R4E.
u
- 32. - With DET A TEST SIGNAL potentiometer R301 fully CCW, verify i .- i the voltage at TP306 is 4.167 i .005V. If not, adjust DET B TEST SIGNAL potentiometer R302 until the required y voltage is obtained. Adjust DET A TEST SIGNAL potentiometer for a voltage reading of 4.375 i .005V at TP 306.
- 33. Rotate OPERATION SELECTOR switch S303 from the DET A & B position to the DET B position. NEGATIVE RATE TRIP lamp should not light. ,
NOTE: If the lamp lights, the procedure starting from step 12 must be repeated. l t
-7 34. '
Rotate OPERATION SELECTOR switch S303 to the DET A and B
, 9 position and adjust DET A TEST SIGNAL potentiometer for a
- voltage reading of 4.542 i .005V. Rotate OPERATION -
- SELECTOR SWITCH S303 to the DET B position and reset the negative rate trip with RATE MODE switch S304.
O e
0 1
i 1 ,
-l , -
1.
{
~. ,
l .r*
SQNP .- *
, ,IMI-92,-PRM-CAL -
- r, Page 22 of 25
! Rev. 10 5.0 PERFORMANCE OF WORK (Cont.) j.
- l. l
, e 5.2.7 (Cont.) i j -
! 35. After 15 seconds rotate OPERATION SELECTOR switch S303 from the DET B position to the DET A & B position.
POSITIVE RATE TRIP lamp should not light.
NOTE: If the lamp lights, the procedure starting from step 13 must be repeated.
- 36. If the remaining sections of the CAL are not to be
, completed, go to section 7. '
I ff 5.2.8 BISTABLE RELAY DRIVERS NC302, 304, 305, 306, 307, and NC308 ADJUSTMENTS
. . I 4 NOTE: -If the "As Found" values for the + volt power supplies '
3 and the bistables in question were found in tolerance, t .,t- Section 5.2.8 need not be performed. For this case write j
- the "As Found" data in the "As Left" data space. .
This paragraph describes the adjustment of bistable relay 8 drives NC302, 304, 305, 306, 307, and NC308. The adjustment y procedure for relay driver NC308 is described in detail, the other relay drivers are adjusted in accordance with data sheets in the same manner as histable relay driver e NC308. The adjustment procedure is as follows: -
I
, NOTE: Before any adjustments are made to the bistable i relay drivers make sure the adjustment procedures j
for the summing and 1cvel amplifier NH310 have been
' , completed in accordance with paragraph 5.2.5. i
- 1. Bypass-the Power Range channel being aligned as speci-fled in paragraph 5.2.1, steps 1 thru 5. If the plant is at too high a power level, remove the cables as specified in paragraph 5.2.1, steps 6, 7, 8, and 9.
- 2. Rotate OPERATION SELECTOR switch S303 to the DET A & B position. ..
- 3. Connect the DVM between TP306 (+) and TP305 (-).
- 4. For initial setup, rotate TRIP ADJ potentiometer R4E and LOOP ADJ potentiometer R11E on bistable relay driver 4 NC308 fully clockwise. Following the initial, setup the instrument mcch should use their own experience as a guideline.
e l
~
.g
- e. ! N se, s' +
,