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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217M4461999-10-20020 October 1999 Forwards Rev 8 to Sequoyah Nuclear Plant Physical Security/ Contingency Plan, IAW 10CFR50.54(p).Encl Withheld,Per 10CFR73.21 ML20217J4151999-10-15015 October 1999 Forwards Request for Addl Info Re Util 990624 Application for Amend of TSs That Would Revise TS for Weighing of Ice Condenser Ice Baskets 05000327/LER-1999-002, Forwards LER 99-002-00 Re Start of Units 1 & 2 EDGs as Result of Cable Being Damaged During Installation of Thermo- Lag for Kaowool Upgrade Project1999-10-15015 October 1999 Forwards LER 99-002-00 Re Start of Units 1 & 2 EDGs as Result of Cable Being Damaged During Installation of Thermo- Lag for Kaowool Upgrade Project ML20217G1141999-10-0707 October 1999 Responds to from P Salas,Providing Response to NRC Risk Determination Associated with 990630 Flooding Event at Sequoyah Facility.Meeting to Discuss Risk Determination Issues Scheduled for 991021 in Atlanta,Ga ML20217B2981999-10-0606 October 1999 Discusses Closeout of GL 92-01,rev 1,suppl 1, Reactor Vessel Integrity, for Sequoyah Nuclear Plant,Units 1 & 2. NRC Also Hereby Solicits Any Written Comments That TVA May Have on Current Rvid Data by 991101 ML20217B8431999-10-0505 October 1999 Requests NRC Review & Approval of ASME Code Relief Requests That Were Identified in Plant Second 10-yr ISI Interval for Both Units.Encl 3 Provides Util Procedure for Calculation of ASME Code Coverage for Section XI Nondestructive Exams IR 05000327/19990041999-10-0101 October 1999 Ack Receipt of Providing Comments on Insp Repts 50-327/99-04 & 50-328/99-04.NRC Considered Comments for Apparent Violation Involving 10CFR50.59 Issue ML20217C7101999-10-0101 October 1999 Forwards Response to NRC 990910 RAI Re Sequoyah Nuclear Plant,Units 1 & 2 URI 50-327/98-04-02 & 50-328/98-04-02 Re Ice Weight Representative Sample ML20212J5981999-10-0101 October 1999 Forwards SE Accepting Request for Relief from ASME Boiler & Pressure Vessel Code,Section Xi,Requirements for Certain Inservice Insp at Plnat,Unit 1 ML20212M1081999-09-29029 September 1999 Confirms Intent to Meet with Utils on 991025 in Atlanta,Ga to Discuss Pilot Plants,Shearon Harris & Sequoyah Any Observations & Lessons Learned & Recommendations Re Implementation of Pilot Program ML20217A9451999-09-27027 September 1999 Forwards Insp Repts 50-327/99-05 & 50-328/99-05 on 990718- 0828.One Violation Identified & Being Treated as Non-Cited Violation ML20216J9351999-09-27027 September 1999 Responds to NRC Re Violations Noted in Insp Repts 50-327/99-04 & 50-328/99-04.Corrective Actions:Risk Determination Evaluation Was Performed & Licensee Concluded That Event Is in Green Regulatory Response Band ML20212F0751999-09-23023 September 1999 Forwards SER Granting Util 981021 Request for Relief from ASME Code,Section XI Requirements from Certain Inservice Insp at Sequoyah Nuclear Power Plant,Units 1 & 2 Pursuant to 10CFR50.55a(a)(3)(ii) ML20212F4501999-09-23023 September 1999 Forwards Amends 246 & 237 to Licenses DPR-77 & DPR-79, Respectively & Ser.Amends Approve Request to Revise TSs to Allow Use of Fully Qualified & Tested Spare Inverter in Place of Any of Eight Required Inverters ML20212M1911999-09-21021 September 1999 Discusses Exercise of Enforcement Discretion Re Apparent Violation Noted in Insp Repts 50-327/99-04 & 50-328/99-04 Associated with Implementation of Procedural Changes Which Resulted in Three Containment Penetrations Being Left Open ML20211Q0311999-09-10010 September 1999 Requests Written Documentation from TVA to Provide Technical Assistance to Region II Re TS Compliance & Ice Condenser Maint Practices at Plant ML20216F5441999-09-0707 September 1999 Provides Results of Risk Evaluation of 990630,flooding Event at Sequoyah 1 & 2 Reactor Facilities.Event Was Documented in Insp Rept 50-327/99-04 & 50-328/99-04 & Transmitted in Ltr, ML20211N5681999-09-0101 September 1999 Submits Clarification of Two Issues Raised in Insp Repts 50-327/99-04 & 50-328/99-04,dtd 990813,which Was First Insp Rept Issued for Plant Under NRC Power Reactor Oversight Process Pilot Plant Study ML20211G5881999-08-27027 August 1999 Submits Summary of 990820 Management Meeting Re Plant Performance.List of Attendees & Matl Used in Presentation Enclosed ML20211F8891999-08-25025 August 1999 Forwards Sequoyah Nuclear Plant Unit 1 Cycle 9 Refueling Outage, Re Completed SG Activities,Per TSs 4.4.5.5.b & 4.4.5.5.c ML20211A1851999-08-16016 August 1999 Forwards Proprietary TR WCAP-15128 & non-proprietary Rept WCAP-15129 for NRC Review.Repts Are Provided in Advance of TS Change That Is Being Prepared to Support Cycle 10 Rfo. Proprietary TR Withheld,Per 10CFR2.790 ML20210V1471999-08-13013 August 1999 Forwards Insp Repts 50-327/99-04 & 50-328/99-04 on 990601- 0717.One Potentially Safety Significant Issue Identified.On 990630,inadequate Performance of Storm Drain Sys Caused Water from Heavy Rainfall to Backup & Flood Turbine Bldg ML20211A1921999-08-12012 August 1999 Requests Proprietary TR WCAP-15128, Depth-Based SG Tube Repair Criteria for Axial PWSCC at Dented TSP Intersections, Be Withheld from Public Disclosure Per 10CFR2.790 ML20210Q5011999-08-0505 August 1999 Informs That NRC Plans to Administer Gfes of Written Operator Licensing Exam on 991006 at Sequoyah Nuclear Plant. Sample Registration Ltr Encl ML20210L4291999-08-0202 August 1999 Forwards Sequoyah Nuclear Plant Unit 2 Cycle 9 12-Month SG Insp Rept & SG-99-07-009, Sequoyah Unit-2 Cycle 10 Voltage-Based Repair Criteria 90-Day Rept. Repts Submitted IAW TS 4.4.5.5.b & TS 4.4.5.5.c ML20210L1611999-07-30030 July 1999 Forwards Request for Relief RV-4 Re ASME Class 1,2 & 3 Prvs, Per First ten-year Inservice Test Time Interval.Review & Approval of RV-4 Is Requested to Support Unit 1 Cycle 10 Refueling Outage,Scheduled to Start 000213 ML20210G5301999-07-28028 July 1999 Forwards Sequoyah Nuclear Plant Unit 2 ISI Summary Rept That Contains Historical Record of Repairs,Replacement & ISI & Augmented Examinations That Were Performed on ASME Code Class 1 & 2 Components from 971104-990511 ML20211B9661999-07-26026 July 1999 Informs That Sequoyah Nuclear Plant Sewage Treatment Plant, NPDES 0026450 Outfall 112,is in Standby Status.Flow Has Been Diverted from Sys Since Jan 1998 ML20210B2521999-07-14014 July 1999 Confirms 990712 Telcon Between J Smith of Licensee Staff & M Shannon of NRC Re semi-annual Mgt Meeting Schedule for 990820 in Atlanta,Ga to Discuss Recent Sequoyah Nuclear Plant Performance ML20210J1091999-07-10010 July 1999 Submits Suggestions & Concerns Re Y2K & Nuclear Power Plants ML20196K0381999-06-30030 June 1999 Provides Written Confirmation of Completed Commitment for Final Implementation of Thermo-Lag 330-1 Fire Barrier Corrective Actions at Snp,Per GL 92-08 ML20209E4071999-06-30030 June 1999 Forwards Insp Repts 50-327/99-03 & 50-328/99-03 on 990328- 0531.Violations Being Treated as Noncited Violations ML20196J8261999-06-28028 June 1999 Forwards Safety Evaluation Authorizing Request for Relief from ASME Boiler & Pressure Vessel Code,Section XI Requirements for Certain Inservice Inspections at Sequoyah Nuclear Plant,Units 1 & 2 ML20196G7881999-06-22022 June 1999 Informs NRC of Changes That Util Incorporated Into TS Bases Sections & Trm.Encl Provides Revised TS Bases Pages & TRM Affected by Listed Revs ML20196G1801999-06-21021 June 1999 Requests Termination of SRO License SOP-20751-1,for Lf Hardin,Effective 990611.Subject Individual Resigned from Position at TVA ML20195G1821999-06-0808 June 1999 Requests NRC Review & Approval of ASME Code Relief for ISI Program.Encl 1 Provides Relief Request 1-ISI-14 That Includes Two Attachments.Encl 2 Provides Copy of Related ASME Code Page ML20195E9521999-06-0707 June 1999 Requests Relief from Specific Requirements of ASME Section Xi,Subsection IWE of 1992 Edition,1992 Addenda.Util Has Determined That Proposed Alternatives Would Provide Acceptable Level of Quality & Safety ML20195E9311999-05-28028 May 1999 Informs of Planned Insp Activities for Licensee to Have Opportunity to Prepare for Insps & Provide NRC with Feedback on Any Planned Insps Which May Conflict with Plant Activities ML20195B3631999-05-21021 May 1999 Requests Termination of SRO License for Tj Van Huis,Per 10CFR50.74(a).TJ Van Huis Retired from Util,Effective 990514 ML20207A5721999-05-20020 May 1999 Forwards Correction to Previously Issued Amend 163 to License DPR-79 Re SR 4.1.1.1.1.d Inadvertently Omitted from Pp 3/4 1-1 of Unit 2 TS ML20206Q8791999-05-13013 May 1999 Forwards L36 9990415 802, COLR for Sequoyah Nuclear Plant Unit 2,Cycle 10, IAW Plant TS 6.9.1.14.c 05000327/LER-1999-001, Forwards LER 99-001-00 Re Condition That Resulted in Granting of Enforcement Discretion,Per Failure of Centrifugal Charging Pump.Condition Being Reported IAW 10CFR50.73(a)(2)(i)(B) & (a)(2)(iv)1999-05-11011 May 1999 Forwards LER 99-001-00 Re Condition That Resulted in Granting of Enforcement Discretion,Per Failure of Centrifugal Charging Pump.Condition Being Reported IAW 10CFR50.73(a)(2)(i)(B) & (a)(2)(iv) ML20206M9341999-05-10010 May 1999 Forwards Rept of SG Tube Plugging During Unit 2 Cycle 9 Refueling Outage,As Required by TS 4.4.5.5.a.ISI of Unit 2 SG Tubes Was Completed on 990503 ML20206K6271999-05-0606 May 1999 Requests Termination of SRO License for MR Taggart,License SOP-21336 Due to Resignation on 990430 ML20206J2061999-05-0404 May 1999 Requests Relief from Specified ISI Requirements in Section XI of ASME B&PV Code.Tva Requests Approval to Use Wire Type Penetrameters in Lieu of Plaque Type Penetrameters for Performing Radiographic Insps.Specific Relief Request,Encl ML20209J0391999-04-27027 April 1999 Forwards Annual Radioactive Effluent Release Rept, Radiological Impact Assessment Rept & Rev 41 to ODCM, for Period of Jan-Dec 1998 ML20206C6541999-04-23023 April 1999 Forwards Response to NRC 990127 RAI Re GL 96-05 for Sequoyah Nuclear Plant,Units 1 & 2 ML20206C0841999-04-23023 April 1999 Forwards Insp Repts 50-327/99-02 & 50-328/99-02 on 990214-0327.No Violations Noted ML20206B9591999-04-20020 April 1999 Responds to 990417 Request That NRC Exercise Discretion Not to Enforce Compliance with Actions Required in Unit 1 TS 3.1.2.2,3.1.2.4 & 3.5.2 & Documents 990417 Telephone Conversation When NRC Orally Issued NOED ML20205S5891999-04-17017 April 1999 Documents Request for Discretionary Enforcement for Unit 1 TS LCOs 3.1.2.2,3.1.2.4 & 3.5.2 to Support Completion of Repairs & Testing for 1B-B Centrifugal Charging Pump (CCP) 1999-09-07
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217M4461999-10-20020 October 1999 Forwards Rev 8 to Sequoyah Nuclear Plant Physical Security/ Contingency Plan, IAW 10CFR50.54(p).Encl Withheld,Per 10CFR73.21 05000327/LER-1999-002, Forwards LER 99-002-00 Re Start of Units 1 & 2 EDGs as Result of Cable Being Damaged During Installation of Thermo- Lag for Kaowool Upgrade Project1999-10-15015 October 1999 Forwards LER 99-002-00 Re Start of Units 1 & 2 EDGs as Result of Cable Being Damaged During Installation of Thermo- Lag for Kaowool Upgrade Project ML20217B8431999-10-0505 October 1999 Requests NRC Review & Approval of ASME Code Relief Requests That Were Identified in Plant Second 10-yr ISI Interval for Both Units.Encl 3 Provides Util Procedure for Calculation of ASME Code Coverage for Section XI Nondestructive Exams ML20217C7101999-10-0101 October 1999 Forwards Response to NRC 990910 RAI Re Sequoyah Nuclear Plant,Units 1 & 2 URI 50-327/98-04-02 & 50-328/98-04-02 Re Ice Weight Representative Sample ML20216J9351999-09-27027 September 1999 Responds to NRC Re Violations Noted in Insp Repts 50-327/99-04 & 50-328/99-04.Corrective Actions:Risk Determination Evaluation Was Performed & Licensee Concluded That Event Is in Green Regulatory Response Band ML20211N5681999-09-0101 September 1999 Submits Clarification of Two Issues Raised in Insp Repts 50-327/99-04 & 50-328/99-04,dtd 990813,which Was First Insp Rept Issued for Plant Under NRC Power Reactor Oversight Process Pilot Plant Study ML20211F8891999-08-25025 August 1999 Forwards Sequoyah Nuclear Plant Unit 1 Cycle 9 Refueling Outage, Re Completed SG Activities,Per TSs 4.4.5.5.b & 4.4.5.5.c ML20211A1851999-08-16016 August 1999 Forwards Proprietary TR WCAP-15128 & non-proprietary Rept WCAP-15129 for NRC Review.Repts Are Provided in Advance of TS Change That Is Being Prepared to Support Cycle 10 Rfo. Proprietary TR Withheld,Per 10CFR2.790 ML20211A1921999-08-12012 August 1999 Requests Proprietary TR WCAP-15128, Depth-Based SG Tube Repair Criteria for Axial PWSCC at Dented TSP Intersections, Be Withheld from Public Disclosure Per 10CFR2.790 ML20210L4291999-08-0202 August 1999 Forwards Sequoyah Nuclear Plant Unit 2 Cycle 9 12-Month SG Insp Rept & SG-99-07-009, Sequoyah Unit-2 Cycle 10 Voltage-Based Repair Criteria 90-Day Rept. Repts Submitted IAW TS 4.4.5.5.b & TS 4.4.5.5.c ML20210L1611999-07-30030 July 1999 Forwards Request for Relief RV-4 Re ASME Class 1,2 & 3 Prvs, Per First ten-year Inservice Test Time Interval.Review & Approval of RV-4 Is Requested to Support Unit 1 Cycle 10 Refueling Outage,Scheduled to Start 000213 ML20210G5301999-07-28028 July 1999 Forwards Sequoyah Nuclear Plant Unit 2 ISI Summary Rept That Contains Historical Record of Repairs,Replacement & ISI & Augmented Examinations That Were Performed on ASME Code Class 1 & 2 Components from 971104-990511 ML20210J1091999-07-10010 July 1999 Submits Suggestions & Concerns Re Y2K & Nuclear Power Plants ML20196K0381999-06-30030 June 1999 Provides Written Confirmation of Completed Commitment for Final Implementation of Thermo-Lag 330-1 Fire Barrier Corrective Actions at Snp,Per GL 92-08 ML20196G7881999-06-22022 June 1999 Informs NRC of Changes That Util Incorporated Into TS Bases Sections & Trm.Encl Provides Revised TS Bases Pages & TRM Affected by Listed Revs ML20196G1801999-06-21021 June 1999 Requests Termination of SRO License SOP-20751-1,for Lf Hardin,Effective 990611.Subject Individual Resigned from Position at TVA ML20195G1821999-06-0808 June 1999 Requests NRC Review & Approval of ASME Code Relief for ISI Program.Encl 1 Provides Relief Request 1-ISI-14 That Includes Two Attachments.Encl 2 Provides Copy of Related ASME Code Page ML20195E9521999-06-0707 June 1999 Requests Relief from Specific Requirements of ASME Section Xi,Subsection IWE of 1992 Edition,1992 Addenda.Util Has Determined That Proposed Alternatives Would Provide Acceptable Level of Quality & Safety ML20195B3631999-05-21021 May 1999 Requests Termination of SRO License for Tj Van Huis,Per 10CFR50.74(a).TJ Van Huis Retired from Util,Effective 990514 ML20206Q8791999-05-13013 May 1999 Forwards L36 9990415 802, COLR for Sequoyah Nuclear Plant Unit 2,Cycle 10, IAW Plant TS 6.9.1.14.c 05000327/LER-1999-001, Forwards LER 99-001-00 Re Condition That Resulted in Granting of Enforcement Discretion,Per Failure of Centrifugal Charging Pump.Condition Being Reported IAW 10CFR50.73(a)(2)(i)(B) & (a)(2)(iv)1999-05-11011 May 1999 Forwards LER 99-001-00 Re Condition That Resulted in Granting of Enforcement Discretion,Per Failure of Centrifugal Charging Pump.Condition Being Reported IAW 10CFR50.73(a)(2)(i)(B) & (a)(2)(iv) ML20206M9341999-05-10010 May 1999 Forwards Rept of SG Tube Plugging During Unit 2 Cycle 9 Refueling Outage,As Required by TS 4.4.5.5.a.ISI of Unit 2 SG Tubes Was Completed on 990503 ML20206K6271999-05-0606 May 1999 Requests Termination of SRO License for MR Taggart,License SOP-21336 Due to Resignation on 990430 ML20206J2061999-05-0404 May 1999 Requests Relief from Specified ISI Requirements in Section XI of ASME B&PV Code.Tva Requests Approval to Use Wire Type Penetrameters in Lieu of Plaque Type Penetrameters for Performing Radiographic Insps.Specific Relief Request,Encl ML20209J0391999-04-27027 April 1999 Forwards Annual Radioactive Effluent Release Rept, Radiological Impact Assessment Rept & Rev 41 to ODCM, for Period of Jan-Dec 1998 ML20206C6541999-04-23023 April 1999 Forwards Response to NRC 990127 RAI Re GL 96-05 for Sequoyah Nuclear Plant,Units 1 & 2 ML20205S5891999-04-17017 April 1999 Documents Request for Discretionary Enforcement for Unit 1 TS LCOs 3.1.2.2,3.1.2.4 & 3.5.2 to Support Completion of Repairs & Testing for 1B-B Centrifugal Charging Pump (CCP) ML20205B1091999-03-19019 March 1999 Submits Response to NRC Questions Concerning Lead Test Assembly Matl History,Per Request ML20204H0161999-03-19019 March 1999 Resubmits Util 990302 Response to Violations Noted in Insp Repts 50-327/98-11 & 50-328/98-11.Corrective Actions:Lessons Learned from Event Have Been Provided to Operating Crews ML20204E8251999-03-0505 March 1999 Forwards Sequoyah Nuclear Plant,Four Yr Simulator Test Rept for Period Ending 990321, in Accordance with Requirements of 10CFR55.45 ML20207E6851999-03-0202 March 1999 Responds to NRC Re Violations Noted in Insp Repts 50-327/98-11 & 50-328/98-11.Corrective Actions:Lessons Learned from Event Have Been Provided to Operating Crews ML20207J1171999-01-29029 January 1999 Forwards Copy of Final Exercise Rept for Full Participation Ingestion Pathway Exercise of Offsite Radiological Emergency Response Plans site-specific to Sequoyah NPP ML20202A7141999-01-20020 January 1999 Provides Request for Relief for Delaying Repair on Section of ASME Code Class 3 Piping within Essential Raw Cooling Water Sys ML20198S7141998-12-29029 December 1998 Forwards Cycle 10 Voltage-Based Repair Criteria 90-Day Rept, Per GL 95-05.Rept Is Submitted IAW License Condition 2.C.(9)(d) 05000327/LER-1998-004, Forwards LER 98-004-00,providing Details Concerning Inability to Complete Surveillance within Required Time Interval1998-12-21021 December 1998 Forwards LER 98-004-00,providing Details Concerning Inability to Complete Surveillance within Required Time Interval ML20198D5471998-12-14014 December 1998 Requests That License OP-20313-2 for Je Wright,Be Terminated IAW 10CFR50.74(a).Individual Retiring ML20197J5541998-12-10010 December 1998 Forwards Unit 1 Cycle 9 90-Day ISI Summary Rept IAW IWA-6220 & IWA-6230 of ASME Code,Section Xi.Request for Relief Will Be Submitted to NRC Timeframe to Support Second 10-year Insp Interval,Per 10CFR50.55a 05000327/LER-1998-003, Forwards LER 98-003-00 Re Automatic Reactor Trip with FW Isolation & Auxiliary FW Start as Result of Failure of Vital Inverter & Second Inverter Failure.Event Is Being Reported IAW 10CFR50.73(a)(2)(iv)1998-12-0909 December 1998 Forwards LER 98-003-00 Re Automatic Reactor Trip with FW Isolation & Auxiliary FW Start as Result of Failure of Vital Inverter & Second Inverter Failure.Event Is Being Reported IAW 10CFR50.73(a)(2)(iv) ML20196F9841998-11-25025 November 1998 Provides Changes to Calculated Peak Fuel Cladding Temp, Resulting from Recent Changes to Plant ECCS Evaluation Model ML20195H7891998-11-17017 November 1998 Requests NRC Review & Approval of Five ASME Code Relief Requests Identified in Snp Second ten-year ISI Interval for Units 1 & 2 ML20195E4991998-11-12012 November 1998 Forwards Rev 7 to Physical Security/Contingency Plan.Rev Adds Requirement That Security Personnel Will Assess Search Equipment Alarms & Add Definition of Major Maint.Rev Withheld (Ref 10CFR2.790(d)(1)) 05000328/LER-1998-002, Forwards LER 98-002-00 Re Automatic Turbine & Reactor Trip, Resulting from Failure of Sudden Pressure Relay on 'B' Phase Main Transformer1998-11-10010 November 1998 Forwards LER 98-002-00 Re Automatic Turbine & Reactor Trip, Resulting from Failure of Sudden Pressure Relay on 'B' Phase Main Transformer ML20195G5701998-11-10010 November 1998 Documents Util Basis for 981110 Telcon Request for Discretionary Enforcement for Plant TS 3.8.2.1,Action B,For 120-VAC Vital Instrument Power Board 1-IV.Licensee Determined That Inverter Failed Due to Component Failure ML20155J4031998-11-0505 November 1998 Provides Clarification of Topical Rept Associated with Insertion of Limited Number of Lead Test Assemblies Beginning with Unit 2 Operating Cycle 10 Core ML20154R9581998-10-21021 October 1998 Requests Approval of Encl Request for Relief ISI-3 from ASME Code Requirements Re Integrally Welded Attachments of Supports & Restraints for AFW Piping ML20155B1481998-10-21021 October 1998 Informs That as Result of Discussion of Issues Re Recent Events in Ice Condenser Industry,Ice Condenser Mini-Group (Icmg),Decided to Focus Efforts on Review & Potential Rev of Ice condenser-related TS in Order to Clarify Issues ML20154K1581998-10-13013 October 1998 Forwards Rept Re SG Tube Plugging Which Occurred During Unit 1 Cycle 9 Refueling Outage,Per TS 4.4.5.5.a.ISI of Unit 1 SG Was Completed on 980930 ML20154H6191998-10-0808 October 1998 Forwards Rev 0 to Sequoyah Nuclear Plant Unit 1 Cycle 10 COLR, IAW TS 6.9.1.14.c 05000328/LER-1998-001, Forwards LER 98-001-00 Providing Details Re Automatic Turbine & Reactor Trip Due to Failure of Sudden Pressure Relay on 'B' Phase Main Transformer1998-09-28028 September 1998 Forwards LER 98-001-00 Providing Details Re Automatic Turbine & Reactor Trip Due to Failure of Sudden Pressure Relay on 'B' Phase Main Transformer ML20151W4901998-09-0303 September 1998 Responds to NRC Re Violations Noted in Insp Repts 50-327/98-07 & 50-328/98-07.Corrective Actions:Revised Per SQ971279PER to Address Hardware Issues of Hysteresis, Pressure Shift & Abnormal Popping Noise 1999-09-27
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059K6661990-09-17017 September 1990 Forwards Evaluation That Provides Details of Plug Cracks & Justification for Continued Operation Until 1993 ML20059H4031990-09-10010 September 1990 Discusses Plant Design Baseline & Verification Program Deficiency D.4.3-3 Noted in Insp Repts 50-327/86-27 & 50-328/86-27.Evaluation Concluded That pre-restart Walkdown Data,Loops 1 & 2 Yielded Adequate Design Input ML20059E1851990-08-31031 August 1990 Responds to NRC Re Violations Noted in Insp Repts 50-327/90-22 & 50-328/90-22.Corrective Actions:Extensive Mgt Focus Being Applied to Improve Overtime Use Controls ML20059E2881990-08-31031 August 1990 Forwards Addl Info Re Alternate Testing of Reactor Vessel Head & Internals Lifting Rigs,Per NUREG-0612.Based on Listed Hardships,Util Did Not Choose 150% Load Test Option ML20059H1831990-08-31031 August 1990 Forwards Nonproprietary PFE-F26NP & Proprietary PFE-F26, Sequoyah Nuclear Plan Unit 1,Cycle 5 Restart Physics Test Summary, Re Testing Following Vantage 5H Fuel Assembly installation.PFE-F26 Withheld (Ref 10CFR2.790(b)(4)) ML18033B5031990-08-31031 August 1990 Forwards Financial Info Required to Assure Retrospective Premiums,Per 10CFR140 & 771209 Ltr ML20028G8341990-08-28028 August 1990 Forwards Calculation SCG1S361, Foundation Investigation of ERCW Pumping Station Foundation Cells. ML20063Q2471990-08-20020 August 1990 Submits Implementation Schedule for Cable Tray Support Program.Util Proposes Deferral of Portion of Remaining Activities Until After Current Unit 2 Cycle 4 Refueling Outage,Per 900817 Meeting.Tva Presentation Matl Encl ML20056B5181990-08-20020 August 1990 Responds to NRC Re Order Imposing Civil Monetary Penalty & Violations Noted in Insp Repts 50-327/90-01 & 50-328/90-01.Corrective Actions:Organizational Capabilities Reviewed.Payment of Civil Penalty Wired to NRC ML20063Q2461990-08-17017 August 1990 Forwards Cable Test Program Resolution Plan to Resolve Issues Re Pullbys,Jamming & Vertical Supported Cable & TVA- Identified Cable Damage.Tva Commits to Take Actions Prior to Startup to Verify Integrity of safety-related Cables ML20059A5121990-08-15015 August 1990 Provides Clarification of Implementation of Replacement Items Project at Plant for Previously Procured Warehouse Inventory.Util Committed to 100% Dedication of Commercial Grade,Qa,Level Ii,Previous Procurement Warehouse Spare ML20058M2321990-08-0707 August 1990 Forwards Rept of 900709 Fishkill,Per Requirements in App B, Environ Tech Spec,Subsections 4.1.1 & 5.4.2.Sudden Water Temp Increase Killed Approximately 150 Fish in Plant Diffuser Pond ML20058N2361990-08-0707 August 1990 Confirms That Requalification Program Evaluation Ref Matl Delivered to Rd Mcwhorter on 900801.Ref Matl Needed to Support NRC Preparation for Administering Licensed Operator Requalification Exams in Sept 1990 ML20058M4471990-07-27027 July 1990 Responds to Unresolved Items Which Remain Open from Insp Repts 50-327/90-18 & 50-328/90-18.TVA in Agreement W/Nrc on Scope of Work Required to Address Concerns W/Exception of Design Basis Accident & Zero Period Accelaration Effects ML20058M0111990-07-27027 July 1990 Forwards Addl Info Re Plant Condition Adverse to Quality Rept Concerning Operability Determination.Probability of Cable Damage During Installation Low.No Programmatic Cable Installation Problems Exist ML20055J3531990-07-27027 July 1990 Forwards Revised Commitment to Resolve EOP Step Deviation Document Review Comments ML20055J0771990-07-26026 July 1990 Requests Termination of Senior Reactor Operator License SOP-20830 for Jh Sullivan Due to Resignation from Util ML20055G6611990-07-17017 July 1990 Forwards Justification for Continued Operation for safety- Related Cables Installed at Plant,Per 900717 Telcon.No Operability Concern Exists at Plant & No Programmatic Problems Have Been Identified.Summary of Commitments Encl ML20058L7001990-07-16016 July 1990 Forwards Response to SALP Repts 50-327/90-09 & 50-328/90-09 for 890204 - 900305,including Corrective Actions & Improvements Being Implemented ML20055F6151990-07-13013 July 1990 Provides Addl Bases for Util 900320 Proposal to Discontinue Review to Identify Maint Direct Charge molded-case Circuit Breakers Procured Between Aug 1983 & Dec 1984,per NRC Bulletin 88-010.No Significant Assurance Would Be Expected ML20044B2211990-07-12012 July 1990 Forwards Addl Info Clarifying Certain Conclusions & Recommendation in SER Re First 10-yr Interval Inservice Insp Program ML20055D2531990-07-0202 July 1990 Provides Status of Q-list Development at Plant & Revises Completion Date for Effort.Implementation of Q-list Would Cause Unnecessary & Costly Delays in Replanning Maint,Mod, outage-related Activities & Associated Procedure Revs ML20043H9061990-06-21021 June 1990 Responds to Generic Ltr 90-04, Request for Info on Status of Licensee Implementaion of Generic Safety Issues Resolved W/Imposition or Requirements or Corrective Actions. No Commitments Contained in Submittal ML20043H2281990-06-18018 June 1990 Informs of Issue Recently Identified During Startup of Facility from Cycle 4 Refueling Outage & How Issue Addressed to Support Continued Escalation to 100% Power,Per 900613 & 14 Telcons ML20043G4901990-06-14014 June 1990 Forwards Tabs for Apps a & B to Be Inserted Into Util Consolidated Nuclear Power Radiological Emergency Plan ML20043F9261990-06-13013 June 1990 Responds to NRC Bulletin 89-002, Stress Corrosion Cracking of High-Hardness Type 410 Stainless Steel Internal Preloaded Bolting in Anchor/Darling Model S3502 Swing Check Valves or Valves of Similar Design. ML20043F9301990-06-13013 June 1990 Responds to NRC 900516 Ltr Re Violations Noted in Insp Repts 50-327/90-17 & 50-328/90-17.Corrective Action:Test Director & Supervisor Involved Given Appropriate Level of Disciplinary Action ML20043H0361990-06-11011 June 1990 Forwards Supplemental Info Re Unresolved Item 88-12-04 Addressing Concern W/Double Differentiation Technique Used to Generate Containment Design Basis Accident Spectra,Per 900412 Request ML20043D9921990-06-0505 June 1990 Responds to NRC 900507 Ltr Re Violations Noted in Insp Repts 50-327/90-14 & 50-328/90-14.Corrective Actions:Util Reviewed Issue & Determined That Trains a & B Demonstrated Operable in Jan & Apr,Respectively of 1989 ML20043C2821990-05-29029 May 1990 Requests Relief from ASME Section XI Re Hydrostatic Pressure Test Requirements Involving RCS & Small Section of Connected ECCS Piping for Plant.Replacement & Testing of Check Valve 1-VLV-63-551 Presently Scheduled for Completion on 900530 ML20043C0581990-05-29029 May 1990 Forwards Response to NRC 900426 Ltr Re Violations Noted in Insp Repts 50-327/90-15 & 50-328/90-15.Response Withheld (Ref 10CFR73.21) ML20043B3051990-05-22022 May 1990 Forwards Detailed Scenario for 900711 Radiological Emergency Plan Exercise.W/O Encl ML20043B1201990-05-18018 May 1990 Forwards, Diesel Generator Voltage Response Improvement Rept. Combined Effect of Resetting Exciter Current Transformers to Achieve flat-compounding & Installing Electronic Load Sequence Timers Produced Acceptable Voltage ML20043A6101990-05-15015 May 1990 Forwards Rev 16 to Security Personnel Training & Qualification Plan.Rev Withheld (Ref 10CFR2.790) ML20043A2391990-05-15015 May 1990 Forwards Revised Tech Spec Pages to Support Tech Spec Change 89-27 Re Steam Generator Water Level Adverse Trip Setpoints for Reactor Trip Sys Instrumentation & Esfas. Encl Reflects Ref Leg Heatup Environ Allowance ML20043A0581990-05-11011 May 1990 Forwards Cycle 5 Redesign Peaking Factor Limit Rept for Facility.Unit Redesigned During Refueling Outage Due to Removal & Replacement of Several Fuel Assemblies Found to Contain Leaking Fuel Rods ML20043A0571990-05-10010 May 1990 Forwards List of Commitments to Support NRC Review of Eagle 21 Reactor Protection Sys Function Upgrade,Per 900510 Telcon ML20042G9771990-05-0909 May 1990 Responds to NRC 900412 Ltr Re Violations Noted in Insp Repts 50-327/90-01 & 50-328/90-01 & Proposed Imposition of Civil Penalty.Corrective Actions:Rhr Pump 1B-B Handswitch in pull- to-lock Position to Ensure One Train of ECCS Operable ML20042G4651990-05-0909 May 1990 Provides Addl Info Re Plant Steam Generator Low Water Level Trip Time Delay & Function of P-8 Reactor Trip Interlock,Per 900430 Telcon.Trip Time Delay Does Not Utilize P-8 Interlock in Any Manner ML20042G4541990-05-0909 May 1990 Provides Notification of Steam Generator Tube Plugging During Unit 1 Cycle 4 Refueling Outage,Per Tech Specs 4.4.5.5.a.Rept of Results of Inservice Insp to Be Submitted by 910427.Summary of Tubes Plugged in Unit 1 Encl ML20042G0441990-05-0808 May 1990 Forwards Nonproprietary WCAP-11896 & WCAP-8587,Suppl 1 & Proprietary WCAP-8687,Suppls 2-E69A & 2-E69B & WCAP-11733 Re Westinghouse Eagle 21 Process Protection Sys Components Equipment Qualification Test Rept.Proprietary Rept Withheld ML20042G1431990-05-0808 May 1990 Forwards WCAP-12588, Sequoyah Eagle 21 Process Protection Sys Replacement Hardware Verification & Validation Final Rept. Info Submitted in Support of Tech Spec Change 89-27 Dtd 900124 ML20042G1001990-05-0808 May 1990 Forwards Proprietary WCAP 12504 & Nonproprietary WCAP 12548, Summary Rept Process Protection Sys Eagle 21 Upgrade,Rtd Bypass Elimination,New Steam Line Break Sys,Medical Signal Selector .... Proprietary Rept Withheld (Ref 10CFR2.790) ML20042G1701990-05-0808 May 1990 Provides Addl Info Re Eagle 21 Upgrade to Plant Reactor Protection Sys,Per 900418-20 Audit Meeting.Partial Trip Output Board Design & Operation Proven by Noise,Fault,Surge & Radio Frequency Interference Testing Noted in WCAP-11733 ML20042G1231990-05-0707 May 1990 Forwards Detailed Discussion of Util Program & Methodology Used at Plant to Satisfy Intent of Reg Guide 1.97,Rev 2 Re Licensing Position on post-accident Monitoring ML20042F7741990-05-0404 May 1990 Informs of Completion of Eagle 21 Verification & Validation Activities Re Plant Process Protection Sys Upgrade.No Significant Disturbances Noted from NRC Completion Date of 900420 ML20042F1691990-05-0303 May 1990 Responds to NRC Bulletin 88-009, Thimble Tube Thinning in Westinghouse Electric Corporation Reactors. Wear Acceptance Criteria Established & Appropriate Corrective Actions Noted. Criteria & Corresponding Disposition Listed ML20042G1381990-04-26026 April 1990 Forwards Westinghouse 900426 Ltr to Util Providing Supplemental Info to Address Questions Raised by NRC Re Eagle-21 Process Protection Channels Required for Mode 5 Operation at Facilities ML20042E9641990-04-26026 April 1990 Forwards Rev 24 to Physical Security/Contingency Plan.Rev Withheld (Ref 10CFR73.21) ML20012E6181990-03-28028 March 1990 Discusses Reevaluation of Cable Pullby Issue at Plant in Light of Damage Discovered at Watts Bar Nuclear Plant. Previous Conclusions Drawn Re Integrity of Class 1E Cable Sys Continue to Be Valid.Details of Reevaluation Encl 1990-09-17
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TENNESSEE VALLEY AUTHORITY g
CH ATTANOOGA. TENNESSEE 374ot SN 157B Lookout Place FEB 061987 U.S. Nuclear Regulatory Consission Attn: Document Control Desk Office of Nuclear Reactor Regulation Washington, D.C. 20555 Attention: Mr. B. J. Youngblood In the Matter of ) Docket Nos. 50-327 Tennessee Valley Authority ) 50-328 SEQUOYAH NUCLEAR PLANT - CONCRETE EVALUATION REPORT On November 13, 1966 TVA submitted a report on the evaluation of concrete quality at Watts Bar Nuclear Plant. Exhibit F to that report addressed.
generic applicability of the Watts Bar concrete issues to Sequoyah and other TVA nuclear facilities. Upon review, NRC requested that a more extensive evaluation be performed for closeout of this issue at Sequoyah units 1 and 2.
TVA subsequently established and implemented a more thorough review program for Sequoyah concrete.
Enclosure 1 documents the results of this review effort. Revision 1 to Exhibit F discusses in detail the program conclusions and the basis for acceptability of concrete at Sequoyah. It supersedes and should replace the original Exhibit F to the Watts Bar Concrete Evaluation Report (CEB-86-19-C).
As enclosure 2 TVA submits CEB Report No. 87-02, which provides the results of our Mondestructive Testing Program at Sequoyah. Data serves to confirm the acceptability of in-place concrete at Sequoyah and is offered for your information.
If any questions exist regarding the enclosed, please call M. R. Harding at (615) 870-6422.
Very truly yours, TENNESSEE VALLEY AUTHORITY R. ridley, D: rector Nuclear Safet# and Licensing Enclosures cc: See Page 2 0j 0
h2]
P 6 870206 K 05000327 PCR
..m An Equal Opportunity Employer
U.S. Nuclear. Regulatory Conmission B 06 M7 cc (Enclosures):
U.S. Nuclear Regulatory Commission Region II Attn: Dr. J. Nelson Grace, Regional Administrator 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Mr. J. J. Holonich Sequoyah Project Manager U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, Maryland 20814 Mr. G. G. Zech, Director TVA Projects U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Sequoyah Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy, Tennessee 37319
ENCLOSURE 1 Revision 1 EXHIBIT F GENERIC REVIEWS Generic Review of Concrete Quality for Sequoyah (SQN),
Browns Ferry (BFN), and Bellefonte (BLN) Nuclear Plants 1.0 GENERAL 1.1 Background -
The evaluation of concrete quality for Watts Bar Nuclear Plant (WBN) has identified several deficiencies which are being cddressed in the following nonconformances:
NCR 6719 - Low Concrete Compressive Strength Results NCR 6720 - Use of Bedding Mortar in Structural Concrete Placements NCR 6721 - Frequency of Sampling Concrete 1.2 General Discussion of Generic Evaluations 1.2.1 WBN NCR 6719 This NCR identified periods of time when the percentage of low concrete compressive strength results exceeded the G-2 specification requirements.
A complete review of the concrete strength test records was performed for Sequoyah Nuclear Plant (SQN). Some deviations from the specification requirements were identified and a Problem Identification Report (PIR) PIRSQNCEB8691 was issued. A detailed evaluation of the deviations has been performed for SQN and is discussed in Section 2.0.
Reviews of the concrete records for Browns Ferry Nuclear Plant (BFN) and Bellefonte Nuclear Plant (BLN) were performed for NCR 6719. These reviews, which are discussed in Sections 3.0 and 4.0, indicate that the deficiency does not exist at BFN and BLN. However, detailed reviews of the concrete records will be performed for the generic evaluation of PIRSQNCEB8691. The results of these reviews will be reported independently for each of the plants as part of the PIR resolution.
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1.2.2 WBN NCR 6720 This NCR identified the use of a special mix, lack of proper sampling, and low strength test results for bedding mortar.
ACI 318-71 recognizes the use of bedding mortar but requires that it be the mortar fraction of the concrete being placed.
The application and the use of bedding mortar were not specifically noted in ,TVA General Construction Specification G-2 (G-2), but were acceptable practices in accordance with ACI 318-71. ACI 318 is the licensing basis for WBN as well as the other plants.
A generic evaluation for the other TVA plants was performed to determine if the use of bedding mortar was properly controlled in structural concrete placements at BFN, SQN, and BLN. As discussed in Sections 3.0 and 4.0, it has been determined that bedding mortar was not routinely used for bedding mortar at BFN or BLN, respectively.
At SQN, mortar was frequently used for concrete pump lubrication which is an accepted industry practice. Mortar was also used in some congested areas but mortar was not used as a standard practice.for bedding on joints. The results of the SQN evaluation are discussed in Section 2.3.
1.2.3 WBN NCR 6721 This NCR identified a failure to always follow the procedure specified in G-2 and WBN QCP 2.02 for Frequency of Sampling Concrete Mixes. The investigation for this NCR showed that the sampling frequencies used at WBN substantially complied with G-2. In view of the relative small percentage of samples deviating from specification requirements (60 out of approximately 6100 total samples for less than 1 percent),
the overall offect is negligible and this NCR was not considered to be significant.
Reviews of the records cad audits for BFN and BLN indicate that requirements for sampling were met. For SQN, a detailed review of the concrete quantities and test data identified only 10 time periods from a total of 497 when the sampling frequency requirements were not met. The 10 occurrences were evaluated and determined to be acceptable. Due to the small number of deviations, the WBN deficiencies were determined to not be generic to SQN.
1.3 Variation of In-Place Concrete Between TVA Plants TVA General Construction Specification G-2 is applicable to all TVA nuclear plants. However, the implementation differed between the nuclear sites. This section describes the background for these differences.
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The physical properties and variability of the concrete at each of the nuclear plants are inherently different because even though the controlling specifications are essentially the same (some differences in the detailed specification provisions exist for the different plants), the materials, and the implementation procedures contain differences. Program results are directly related to the personnel and management implementing its provisions.
Interpretation of specific provisions of control documents will vary. As pointed out in Volume I of TVA's Nuclear Performance Manual, decentralization of program activities contributed to inconsistant application of requirements. Concrete specifications for all nuclear plants state that no more than 10 percent of the strength test results are to be less than the specified strength (20 percent for less than 3000 psi specifie<1 strength). This permits an individual materials engineer to use any percent between 0 and 10 as a target. All nuclear plants changed materials engineers during the course of construction, and at some plants this changed control efforts. Changes also occurred in the design personnel reviewing the control efforts. Our investigations to date indicate that while differences among the plants do exist, the in-place concrete will adequately perform its intended function.
2.0 SEQUOYAH NUCLEAR PLANT
2.1 Background
A generic evaluation was performed for SQN for the three deficiencies listed in Section 1.1 in April 1986. This review determined that the WBN deficiencies were not generic to SQN. This review was based on interviews with cognizant personnel and on a review of approximately a third of the SQN concrete records. In October 1986, the NRC performed an audit to investigate the generic implications of the WBN deficiencies and, as the result of the audit. TVA was requested to perform a more comprehensive review of the concrete records for SQN. The following sections provide the results of the expanded review.
2.2 Sampling Frequency Evaluations 2.2.1 Procedure A detailed review of the concrete sampling frequency was performed using the monthly concrete reports and the concrete testing ledger. The monthly reports include the quantity of concrete placed for several mixes placed during the calendar month covered by the report. The ledger provides a tabulation of all sampling and testing for each concrete mix. The construction procedures were also reviewed for compliance with the requirements of the specification.
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Compliance with the specification requirements was evaluated by correlating the number of tests performed on a mix during the reported time period to the total yardage placed. A
, total of 497 time perio_ds were evaluated.
2.2.2 Results The review of the procedures showed that the requirements of G-2 were correctly incorporated into the procedures and revisions to the requirements were incorporated in a timely manner.
The results of this evaluation show that, for the major mixes, only 10 sample periods out of 497 did not meet the requirements of G-2. Eight of the ten periods were early in the plant construction period and involved fill concrete. An evaluation of these mixes indicated that the strength levels were sufficiently above the specified strength to offset any decrease in strength not identified as the result of inadequate sampling.
The other two periods are in October 1971 and June 1973 for the 300.75AFW. The October 1971 period was reviewed and determined that only about 54 yards of the reported 624 yards was placed in QA structures and that the sampling was
, acceptable. The strength levels for the mix in 1973 were significantly above the required strength.
2.2.3 Conclusion Only minor deviations from the sampling requirements of TVA General Construction Specification G-2. occurred.at SQN and all deviations were evaluated and determined to be acceptable.
2.3 Mortar Usage Evaluations The initial generic evaluation for use of bedding mortar was based on interviews with cognizant engineers involved in the concrete placements at SQN. These reviews indicated that mortar had been used in a few isolated areas but that the use of bedding mortar was not a standard practice as it was at WBN.
Additional data on mortar usage became available as the result of the strength evaluations discussed in Section 2.4, and specifically in development of the data base for the concrete pour cards discussed in Section 2.4.5. The pour cards provide the volumes of all mixes used in a pour including the mortar mix.
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.The information on the pour cards shows that mortar was frequently used'in small quantities for priming and lubrication of the concrete pump. .The pour cards also frequently state that the mortar was wasted. . Occasionally, the pour cards noted that mortar was used in
' congested areas. A total of.about 800 pour cards noted the use of mortar (5500 total pour cards for Category I structures).
Approximately 680 of these pours noted a total mortar volume of 1 cubic yard or less.which was generally for pump lubrication. Only about 50 pours noted mortar volumes exceeding 20 percent of the total volume of the pour. In many of these cases, part of the mortar was likely wasted.
Since the use of mortar for pump lubrication is an accepted and necessary practice, and the use of mortar in congested areas was limited, the deficiency occurring at WBN is not considered to be generic to SQN. However, for the structural evaluations discussed in Section 2.5, if the mortar strength was less than the concrete strength the following adjustments were made: the estimated inplace strength of the pour was assumed to equal the mortar strength if the mortar volume exceeded 20 percent of the total volume of the pour.
For mortar volumes between 5' percent and 20 percent, the average strength of the mortar and concrete was used.
2.4 Compressive Strength Evaluations 2.4.1 Background and Purpose The design drawings for SQN specify the concrete mix to be used in the structures. The requirements for obtaining these required strengths were given in G-2. G-2 specified that no more than 10 percent of the strength test results could fall below the specified strength.
The purpose of the compressive strength evaluation was to identify and evaluate any mixes and time periods when the G-2 requirements for strength were not met.
2.4.2 Data Base Development A computerized data base was developed that consisted of all concrete strength test results obtained for structural concrete mixes which were produced during the construction period and which were placed in Category I structures. This data base consists of approximately 3800 tests on a total of 40 mixes. The data was obtained from the compressive strength test report (TVA Form 331) and the concrete ledger, which was a laboratory worksheet summarizing all strength test results. Both documents were obtained from QA record storage.
The data base showed that approximately 5 percent of the tests did not meet the specified strength requirement at the specified age. This compares to 13 percent for WBN.
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2.4.2 Strength Evaluation Procedure Compliance with the requirements for compressive strength was evaluated by determination of an equivalent specified compressive strength for each six for all periods of time that mix was used. The equivalent specified strength is the strength level below which a statistical calculation indicates no more than 10 percent of the strength tests would be expected to fall. This calculation was based on methods provided in ACI 214 and used the mean and standard deviation for the last 30 tests. The equivalent strength by the ACI 214 method is equal to the mean minus 1.28 times the standard deviation.
The method discussed above is not generally applicable to less than 30 tests. Therefore, for the first 30 tests of a mix, the equivalent specified strengths were based on the ACI 318 criteria. The ACI 318 criteria is that the average of three consecutive tests must exceed the specified strength for the mix and no individual test is more than 500 psi less than the specified strength.
2.4.4 Strength Evaluation Results The strength evaluation identified 7 mixes which had one or more time periods when the G-2 requirements for strength were not met. As the result of this determination, the deficiency identified at WBN was determined to be generic to SQN and PIRSQNCEB8691 was written.
The following mixes were identified as having " low strength" periods: 300.75AFW, 301,5AFW (2 periods), 401,5AFW, 401.5AFWR1, 500.75AFW (3 periods), 500.75AFWG, and 800.75BFW (2 periods). Two of the mixes had previously been evaluated for lower strength as documented in the Section 3.8 of the FSAR (mixes 401.5AFWR1 and 800.75BFW).
Equivalent specified strengths were assigned to each of the mixes for " low strength" time period. The value assigned was the lowest equivalent strength calculated during the time period that the G-2 requirements were not met. The equivalent specified strengths for the 7 mixes (and each time period) are provided in Attachment 1 (ESS28 for class A mixes and ESS90 for all mixes),
f i
l i
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4
2.4.5 Id:ntific:tien cf Pcurs Affceted by "L:w Stesngth" Time' Periods An additional data base was developed that consisted of the data from all pour cards for each structural concrete pour in Category I structures. The pour card is the QA record of the placement and, in addition to providing documentation of the inspections, it identifies the concrete mix used and the date of placement. A total of 5500 QA concrete pours were identified with 420 having mixes with " low strength" time periods.
The pour card data base was sorted by concrete mix and placement date to allow determination of all concrete pours made during " low strength" time periods. A computer summary was developed that listed all concrete pours with concrete having an equivalent specified strength that did not meet the G-2 strength requirement at the specified age. This sununary along with the construction pour drawings was used for identification of concrete members that required additional evaluation.
- 2. 5' Structural Evaluations j 2.5.1 Concrete Structures i All Category I concrete placements were identified where the equivalent specified strength was less than that specified on
- the drawings. Based on this information, average estimated
- inplace strengths were established using the method described in Section 2.4 of this report for all such placements. (The rtrengths used were verified as detailed in Section 2.6.)
I Approximately 9,900 cubic yards of concrete, representing 6
- percent of the concrete placed in Category I structures was
- identified as having an equivalent specified strength less than the specified strength at 28 or 90 days.
{
l All design calculations for the affected Category I j structures were reviewed. If the inplace strength was less
! than the strength used for the existing design, an evaluation l_ using the estimated in-place concrete strength was done to
! determine if the structure or feature meets the design requirements. The estimated in-place strengths satisfied structural requirements since either the originally specified l strengths were higher than required, the original loads are
, larger than up-to-date loads, or more reinforcement was provided than was originally required. This evaluation confirms that all concrete placements with estimated in-place i strengths less than specified on the design drawings or used in the design calculations are structurally adequate to perform their intended function.
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All design calculations have been checked and documented.
Design criteria revisions and FSAR revisions to incorporate the results of this evaluation are being developed. All concrete design drawings for Category I structures where the estimated in-place strength is less than the specified strength are being revised to incorporate a reference to ensure the use of proper concrete strengths in future design evaluations. .
2.5.2 Anchorages to Hardoned Concrete The design of concreto expansion anchors is based on a specified compressive strength of 3000 psi. Exceptions to this occur where increased concrete strength was used in a reevaluation for higher loads. Only two cases were identified. Calculations for these supports were reviewed using the estimated in-place strength and were found to be acceptable.
In summary, it is concluded that the designs for anchorages to hardened concrete are adequate for the in-place strength of the concrete.
2.5.3 Embedments All design calculations for embedded plates and bolt assemblies (including some grouted bolts) in the reactor buildings and steam valve room were reviewed. Where estimated in-place strengths were less that those used in the design calculations, the embedment calculation was revised.
All revised calculations demonstrated that the embedments are adequate to carry their design loads.
In addition to the review of all embedded plate calculations completed as discussed above, all Sequoyah Reactor Building and East Steam Valve Room pours with estimated in-place concrete strengths less than the estimated in-place strengths for corresponding Watts Bar pours were identified. All embedments in these lower strength pours were identified, reviewed, and found to be acceptable.
For all other Category I structures (auxiliary, control, diesel generator, and intake pumping station) estimated in-place strengths in this report were compared to specified strengths. Only a few nonstructural elements in the auxiliary building had estimated in-place strengths less than the specified strength. Embedments in these areas were reviewed and found to be acceptable.
In summary, the designs for embedded plates and anchorages are adequate for the in-place strength of the concrete.
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.l 1
2.5.4 Emb:dddd Pleto Rsvitw Prcgrams ,
Three review programs using independent. calculations and unrelated to this investigation have been performed at SQN.
Each of.these programs was reviewed to' determine if in-place strengths were less than the strength used for the calculations. Several cases were found, and revised calculations were generated.- All cases were found to be acceptable. These programs were found to be satisfactory.
-with all designs being' adequate for the estimated in-place strengths.
2.6 Verificationlof Compressive Strenath Adequaev from Later Aae Standard Cylinder Tests 2.6.1 Background Section 4.2 of.the main report describes the method used for application of concrete strength gains with age to the equivalent specified compressive strength. The strength gains are based on tests on cores and cylinders from test blocks. These blocks, although of the same nominal proportions as the mixes evaluated, did not use the same materials as the concrete which did not meet the specification strength requirements.
To further verify the adequacy of the compressive strengths utilized in the structural evaluations, an additional investigation was performed. The purpose of the investigation was to determine the equivalent specified strengths based on the 90-day compressive strength for the standard cured cylinders, and to compare these values to the concrete strength used.for the. structural evaluations. The 90-day compressive strength is sometimes used in the concrete industry. The investigation was performed on the 6 class A mixes that had " low strength" periods. The class B mir (800.75BFW) used the 90-day equivalent specified strength as the inplace strength with no additional strength increase with age.
2.6.2 90-Day compressive Strength A review of the 90-day compressive strength test results for the " low strength" mixes and time periods was performed. The review showed that 9.0-day standard cured cylinders were tested for over 90 percent of the samples made for class A concrete during these periods.
The 90-day compressive strengths for the standard-cured cylinders from the concrete made during periods of low strength document the strength gain with age expected to occur except for very thin members. Using 90-day results, an equivalent specified strength was calculated.
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~
2.6.3 Equivelant Cancreto Strangth at 90 Days Theequivalentspecifiedstrengthat90dayswascalculahd
- using the same methods used previously for the 28-day ~ " ,, ' '
strengths (calculated from the average and standard _ deviation ,
for the last 30 tests or using the ACI criteria for less.than .
30 tests). The calculated values are tabulated on Attach 5ent i 1 (ESS90 for class A mixes). '['
Attachment 1 shows the*6 class A mixes that had " low strength" timo periods and with the equivalent specified strength at 28 and 90 days. The summary shows that for each period the specified strength is achieved or exceeded at 90 days. A review of the structural evaluations shows that the'- f estimated inplace strength used for the evaluation wrs always~
less than or equal the equivalent specified strength at 90 ,
days.
2.6.4 Identification of " Thin" Structural Members As discussed previously, the strength gain indicated by 90-day cylinders for concrete with specified 28-day strength is appropriate if an evaluation for thin members is performed. Therefore, an additional review of the e calculations was performed to verify that the strength gairis used for thin members was conservative.
All concrete pours placed during " low strength" time. periods -
have previously been identified for the evaluations discussed-in Section 2.5.1 and the volume-to-surface ratio (V/S) for' '
each pour calculated. Also, all pours were previously ?
identified which used later age strength in the analysis as '
specified in the design criteria.
1 For the previous evaluation using estimated inplace t , ,
strengths, no strength gain with age was used for members ,
with V/S less than or equal to 0.25 feet (6 inch wall for,, '
example). The full strength gain with age was added iffene, 4 3 V/S was 1 foot (2-foot-thick wall for example). A lincat' i ,
interpolation was used for V/S between 0.25 and 1.0 feet /.' 'r '
6 7 To provide a consistent comparison and verification of the strengths used for the structural analyses, a similar -
approach was used except that the interpolation was between '
the equivalent strength at 28 days and the equivaluat strength at 90 days. < ,;.c
~
Allstructuralmembersplacedduring"lowstrengt[* time periods and all structural members with long-terni strength specified on the drawings were ident'.fied which have V/S less than 1 foot. The smallest V/S was 0.31 feet.., '
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.'?
% .,, . u . ,
t' y Fce occh cf thzsa p:urs en equivslant strength wzs calculatsd based on interpolation between the equivalent strengths at 28 and 90 days. These values were generally higher than the
. strength used for the structural evaluation and were never more than 100 psi less. Therefore, the structural m evaluations based on the estimated in-place strength are
. ; adequate.
2.7 Conclusions , i T!ie generic evaluation has determined that the WBN deficiencies for j sampling frequency and bedding mortar usage do not apply to SQN.
i Some deviations from specification requirements dealing with the
'.* strength level of some concrete mixes were identified and PIRSQNCEB8691 has been issued. The evaluation of the deficiency has
- been completed and all structures were determined to be acceptable.
3.0' BNOWNS FERRY NUCLEAR PLANT
' Discussions were held with H. S. Sheppard and J. Davenport, who were e -
directly involved in the concrete placement during the construction of
[' BFN as materials engineers, and S. Carr, who was assigned to SME.
[g' In addition, a partial review of concrete records was conducted at BFN (Attachment 2). Cylinder tests results show acceptable strengths at the specified age. These revealed the following:
0 '/ There were fewer mixes used on BFN in comparison to WBN. The maximum strength mix was 401.5 AFW. The average strength of this
(/ mix was 5170 psi with less than 2-percent low tests.
0 fe'r a 300.75 AFW . * :. here were no low tests.
0 ' Based on the above and other records, cylinder test results show
~
acceptable strengths at the specified age.
O Concrete mix designs were prepared by SME.
1 0,
'~
Congested areas were encountered in the fuel pool and in columns for the turbine and reactor buildings.
0 4 A high mortar volume, 3/4-inch-maximum aggregate size concrete mix
, was developed by SME and used in congested areas and around water
,. seals.
,/ q 0
[, t here was no knowledge of the use of grout as bedding mortar.
/ r 0 The frequency requirement for test sampling specified in G-2 were s
- adhered to. Since this was TVA's first nuclear plant, the NRC
, performed an extensive audit on TVA concrete procedures at the time
- , , + 'of construction.
l Basad upon the above, the concerns identified for WBN do not apply to BFN.
Howher, a more detailed evaluation of strength test results will be performed and documented for the generic review of PIRSQNCEB8691.
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- ," I
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4.0 3ELLEFONTE NUCLEAR PLANT Discussions were held with D. Nixon and R. Norris, who are assigned to BLN as materials / civil engineers, and other personnel knowledgeable of concrete design, placement, and testing for BLN and a partial review of BLN concrete records was conducted (Attachment 3). This assessment revealed the following:
i 0
An effective working relations, hip existed between the DNC materials
, engineers and DNE concreting specialists along with very good experience levels at both positions.
O The required monthly reports were submitted by DNC and evaluated by e
DNE.
O Concrete mix adjustments were frequently made by DNC to maintain strength levels and major adjustments were reviewed by DNE.
O Concrete compressive strength test records show strengths in excess of requirements.
O Concrete mix designs were prepared by SME or by BLN DNC with DNE g approval.
O Congested areas were encountered.
O High mortar volume, 3/8-inch-maximum aggregate size concrete mix was developed. This was used in congested areas.
- 0 Mortar mixes were developed to correspond to each class of concrete. These specialized mortar mixes were used to lubricate concrete pump lines.
O Informal reviews of the WBN NCRs were performed by BLN DNC personnel under the direction of the onsite nuclear licensing unit. In addition, an internal audit was performed in 1985 on the concrete unit.
Sampling frequencies for concrete were reviewed during the audit and no major deficiencies were noted.
Based upon the above, the deficiencies identified for WBN do not apply to BLN. However, a more detailed evaluation of strength test results will be performed and documented for the generic review of PIRSQNCEB8691.
5.0
SUMMARY
As a result of these discussions and reviews, it is concluded that the deficiencies discovered at WBN in regard to bedding mortar and sampling frequency are plant-specific to WBN and that similar deficiencies do not exist at BFN, SQN, and BLN. The WBN deviations with respect to compressive strength also occurred at SQN but to a lesser degree.
Reviews of compressive strength results at BFN and BLN indicate compliance with specification requirements, however, additional reviews will be performed for the generic review of PIRSQNCEB8691.
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ATTACHMENT 1 Sequoyah Nuclear Plant Equivalent Specified Strengths The following table summarizes the strengths for the 7 mixes with " low strength" time periods. ,.
MIX f'c ESS28 ESS90 psi 28 days 300.75AFW 3000 2600 4000 301.5AFW 3000 2700 4300 2500 4000 401.5AFW 4000 3300 4800 401.5AFWR1 4000 3500 4900 500.75AFW 5000 4800 6800 4600 6200 4800 6500 500.75AFWG 5000 4890 6500 800.75BFN 8000* N/A 7600 N/A 7000 N/A 6800 ESS28 = Equivalent specified strength at 28 days.
ESS90 = Equivalent specified strength at 90 days.
- Specified strength at 90 days.
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