|
---|
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
' ~
TENNESSEE VALLEY AUTHORITY CH ATTANOOG A, TENNESSEE 37401 400 Chestnut Street Tower II January 31, 1983 Director of Nuclear Reactor Regulation Attention: Ms. E. Adensam, Chief Licensing Branch No. 4 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Ms. Adensam:
In the Matter of ) Docket No. 50-327 Tennessee Valley Authority ) 50-328 As requested by the NRC in November of 1982, TVA performed additional testing on the Tayco igniter for our Permanent Hydrogen Mitigation System (PHMS). Information was subsequently provided to the NRC, on an informal basis, during telephone conversations and during the ACRS meetings of the conclusions of the test results. Enclosure 1 provides documentation of the results of the tests and the conclusions based on the test results.
Enclosure 2 provides formal documentation of information provided to the Nhc to support TVA's belief that the PHMS design was adequate.
If you have any questions concerning this matter, please get in touch with J. E. Wills at FTS 858-2683 Very truly yours, TENNESSEE VALLEY AUTHORITY L. M. Mills, Manager Nuclear Licensing
' Sworn d subs ibed before me thisi day of 1983
/$h l , -
'~'
Notary Pub'lic' My Commission Expires "I Enclosures (2) cc: U.S. Nuclear Regulatory Commission (Enclosures)
Region II Attn: Mr. James P. O'Reilly Administrator 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 8302070321 830131 i l PDR ADOCK 05000327 P PDR An Eaual Opportunity Emptover
. - ENCLOSURE 1 i ADDITIONAL TESTING PERFORMED ON TAYC0 IGNITERS SEQUOYAH NUCLEAR PLANT The obj ective of this submittal is to demonstrate that the Tayco igniter assemblies used in the Permanent Hydrogen Nitigation System at Sequoyah Nuclear Plant would op7 rate eff ectively in the spray environment of the upper compartment of the containment. Effective i operation is considered to be the igniters' ability to ignite mixtures of hydrogen, sir, and water near the lower flammability limit in the expected environment. Ignition depends on the capability of the igniter to transf er enough heat locally into the surrounding mixture to bring the mixture up to the ignition temperature. The approach taken in this submittal will be first to describe the design of the igniter assemblies f or the spray environment. Next, a number of laboratory-scale tests will be described where the igniter was subj ected to various spray environments while the igniter surf ace temperature was measured. Finally. the spray test environment will be compared to the expected upper comparteent conditions and the sustainable igniter surface temperatures will be related to the minimum surface temperature required for ignition in the upper compa r tment.
l Each of the igniter assemblies used in the upper compartment c'onsists of a Tayco igniter mounted through the end plate of a standard NEMA-type enclosure covered by a sheet metal spray shield. The dimensions of the igniter / shield configuration were allowed to vary slightly for ease of field installation. At a minimum, the shfeld extends four inches beyond the tip of the igniter coll;and six inches on both sides. At a maximum, the igniter is located ten inches below the shield. The design of the spray shield would prevent overhead impingement of the containment spray on the igniter. How ev er, intermittent local turbulence postulated to occur in the upper compartment could cause scue. indirect spray impingement. This indirect spray will be addressed in the remainder of this submittal.
Tests were performed by TVA at J ts Singleton'Naterials Engineering Laboratory to determine the effect of spray,on igniter surf ace temperature. Two basic groups of tests were conducted. The first group consisted of spray tests on an igniter assembly with a spray shield in an attempt to simulate the actual upper compartment environmental conditions. These tests used a single nozzle to spray l an igniter protected by a spray shield ten inches above the nozzle but *
! slightly smaller than the ones described previously. The second group consisted of spray tests on an igniter without a spray shield in an effort to determine the sustainable igniter surf ace temperature under various direct spray mass fluxes.
The first test of the first group was reported in a previous TVA submittal of November 1,1982. That test used a hollow cone spray nozzle located three f eet above the igniter with a flow rate of 3.5 gal / min. A large f an placed in front of the igniter was used to simulate a potentially turbulent upper compartment environment causing indirect spray impingement. The initial igniter surf ace temperature in this and the following tests varied between 17 00-173 0 *F. When the l spray and f an were activated, the igniter temperature dropped to L m l.__ m _~ ..~. . . . . - .. . . . _ _ ._ - ,
batvoca 1600-1635'F oithis five olestos and remmisod reintively .
stable for the remaining half hour of the test.
The second test of the first group used the some spray nozzle and flow rate, but located the nozzle only two feet above the igniter to provide the same spray mass flux as calculated for the Sequoyah upper c ompa r tment. Assuming uniform distribution of the 9500 gal / min from both trains of the containment spray system over the 10,400 f t* cross section of the upper compartment would yield a mass flux of 0.915 gal / min ft8 The f an was used again to promote turbulence and was moved to the side of the igniter. When the spray and f an were activated, the igniter temperature dropped to between 1650-1670*F and remained stable.
The third test of the first group used a solid cone spray nozzle located 25 f eet above the igniter with a flow rate of 10.5 gal / min.
The igniter was located on the periphety of the spray cone sach that the edge of the cone impinged on the centerline of the spray shield.
The overall mass flux at the igniter elevation equalled that of the upper compartment. The fan was located on the opposite side of the spray cone from the igniter. When the spray and fan were activated, the igniter temperature dropped to between 1525-1560*F. The sustained igniter surf ace temperatures determined in each of these three tests with a spray shield are well above the minimum required temperature that will be justified later in this submittal.
Since the environmental conditions (especially turbulence) cannot be accurately quantified at all locations in the , upper compartment and since reproducing these conditions exactly in a small-scale test is not possible, the second group of spray tests was conducted without a spray shield in order to obtain a simple correlation between the spray mass flux impinging directly on the ignit,er,and the igniter surface temperature. The solid cone nozzle was used for this group of tests at a number of elevations above the igniter to provide various mass fluxes ranging up to approximately the maximum direct upper compartment mass flux of 0.915 gal / min f t8 No f an was required to provide turbulence since the spray impinged directly on the igniter.
A series of four spray testiwas performed which, when the dry (initial) temp 9rature was included, yielded the curve in Figure 1, which relates the test to upper' compartmenkmass flux ratio and the average igniter surface temperature that could be sustained. The data
-shown in the figure is tabularized below.
Nozzle Heiaht Mass Flux Ratio
- Averaae Sustained Temperature 2.5 ' O.92 12758 F 3' O.64 1337' F 4' O.36 1440' F 6' '*0.16 1517' F
- Where a nozzle height of 2 4' would yield the u perpco pm artment mass flux of 0.915 gal / min /ft8 The sustainable igniter surf ace temperatures in the spray tests may be compared to the minimum igniter surface temperatures required to ignite hydrogen, sir, and water mixtures measured for the Tayco igniter at the Whiteshell Nuclear Research Establishment (WNRE). The results of the WNRE Tayco tests were submitted by TVA on June 14, 1982. One objective of the WNRE tests was to determine the igniter M b4 use 3 t.=eh.e' a-=.w- w ' --%.3
', , .strineo temperntgre required to ylsid a gss temperatere sufficient to ignite the test mixture. These experiments were performed in a 17 liter vessel with the Tayco igniter mounted horizontally near the center. Thermocouples welded to the igniter measured the surface temperature while an ionization probe was used to signal that ignition and flame propagation had occurred. A fan was activated in sane tests i to provide turbulence directly at the igniter. A number of tests were performed a't various concentrations of hydrogen and steam. No sprays were included. The results of the WNRE Tayco test series are shown in l Figure 2.
The WNRE Tayco test data is applicable to the Tayco igniter assemblies in the upper compartment at Sequoyah. Ile phenomenon of ignition is a local one that is based on transferring enough heat f rom the igniter into the immediately surrounding flammable mixture until it reaches a sufficient t empe ra t ur e. Once ignition occurs, the phenomenon of flame propagation may be affected by global conditions. H ow ev e r, since the phenomenon of ignition is local, it is demonstrable in a vessel of smaller scale than the upper conpartment as long as the smaller vessel does not impose unrepresentative boundary conditions. The WNRE vessel size was large enough, the f an flow was turbulent enough, and the time lag between igniter activation and ignition was short enough
]
- that no appreciable bulk heating of the mixture occurred before
! ignition. Since ignition occurred as soon as the igniter warned up i enough to transf er the required amount of heat into the local mixture i to promote ignition, the scale effects of the vessel size did not become important. The WNRE tests utilized a Tayco igniter of 'the same general configuration as the ones used in the. upper compartment a s semblie s. Since the ignition process is dependent on the heat transferred from the igniter to the local gas, it was important to have the same igniter temperature, surf ace area,; and geometrical configuration. The Tayco igniter has a relatively large surf ace area which promotes effective heat transfer. In addition, the Tayco coli configuration tends to create its own local turbulence and increase the residence time of local mixtures which enhances heat transfer.
The WNRE tests were performed with various concentrations of steam instead of spray. The effects of steam on ignition are more pronounced than spray, i.e.,' igniter surface temperatures required to heat a mixture with a high concentra tion of steam would be higher than those required for a typical spray mixtur'e' with more widely-dispersed drops. Accounting for the heat capacity effects of spray would probably yield results closer to the WNEE dry test conditions than q^
those of the high steam f raction tests shown in Figure 2. Note that the surf ace temperature shown in Figure 2 required to ignite the dry mixture was about 1200*F. The WNRE test conditions with f an flow and high st eam fraction should conservatively bound the heat transfer effects of turbulence and spray in the upper com.partment. Therefore. -
based on the WNRE test data f or the Tayco igniter in a turbulent, 40-p erce nt ste am enviro'nment (see Figure 2), we believe that 13258F represents a conservative minimum igniter surf ace temperature. For additional margin,1350*F will be employed as the minimum Tayco igniter surf ace temperature required for ignition in an environment typical of the upper compar tment. We do not believe that higher required surface temperatures that may be determined for other types of igniters are appropriate for use with the Tayco igniter because the Tayco igniter has its own unique combination of surf ace temperature, area, and configuration supported by test data specifically for that combination.
. .-.m . .... m - . ~ . _
,----n----, - - - - , , --
Tho Singleten direct spray test cass fitz s ccy bs. comp;rsd' to t'h o 9 -
expected spray mass fluxes due to local ~ turburence in th~e upper compa rtment. Of course, the mass flux ratio of 1.0'is'the maximum obtainable since it represents the direct overhead spray-impingement that would occur without any spray shield. The actual spray mass' flux ratio that might be expected to. impinge on the ignitors beneath their -
spray shields due to local turbulence would be significantly less.
There are six igniter assemblies located below the containment spray headers in the upper compartment.. Four .of these are mounted on the inside of the crane wall at elevation 784 and one on each of the ice condenser end walls at elevations 760 and 765. The boundary layer that would be associated with these walls in the upper compartment would tend to reduce the turbulence levels at the igniter locations !
below bulk levels. In addition, the lowest igniter assembly is located well of f the operating floor where the spray-induced turbulence would be most pronounced. Furthermore, any local turbulence at the igniter locations would be intermittent and would -
not cause continuous spray impingement on the igniter. Employing the minimum igniter surf ace temperature of 1350'F justified above and referring to Figure 1, the corresponding mass flux ratio that could be tolerated is determined to be 0.6. Because of the presence of the spray shields above each of the igniters, their location in regions of reduced turbulence, and the intermittent nature of any local turbulence, we believe that the highest realistically achievable spray mass flux on the upper compartment igniters with spray shields would be less than this acceptable mass flux ratio of 60 percent of the maximum possible mass flux without the spray shields.
In summary, TVA has designed each of the PHMS " igniter assemblies in the upper compartment with spray shields to prevent overhead impingament by the containment spray. To account
- for any possible eff ects on igniter performance due to spray impingement induced by local turbulence, TVA has established 1350*F as the minimum igniter surf ace temperature required for ignition based on the WNRE test data and has conducted two groups of tests at its Singleton Laboratory to simulate expected environmental conditions in the upper compartment.
The first group of Singleton tests showed directly that the assembly with a spray shield would al'10w the igniter to remain above the minimum required surface temperature for ignition in an upper compartment spray environment. The secon'd' stoup of tests produced data that correlated various directly impinging spray macs fluxes with
-the sustainable igniter surf ace temperature. Based on the established minimum required surf ace temperature, the second group of te st s showed that the igniter could withstand direct spray mass fluxes higher than any spray mass fluxes that could be expected from local turbulence in the upper compartment. Based on our assessment of these test results, TVA has concluded that the Tayco igniter assemblies in the upper compartment would operate eff ectively in that spray environment.
O m , Mwn..es w -. - . . - + . - +
= ..
~....
I l i. I i l l l l. _
l, _' ' _ l , l l,l I .' l I l I ,
I b I I I ! ! ! II I ! Ifpoer Cbenar:m b b pra'p dss: Flux i ! I l' i" -
i ! Li i l i i ! -
l !I I I I I i I I
- -! a i
w I l* I rI *I ,
=1 *:
, PI i i
i I .
i i ! : i i l I i i l I 'i I i i i i i iiI I II I i l i l
II I I I I liI !lI II IlI II I I l im i i ic ! I II III I I I IE l I l .5 i i I i I VI I i I L V l. l -
I l
.- _I I I I lii i i I I f I I Id 7 ! l 1 I I I l l J/[ l i i I l 1 l il P I II' If I 1I l 'l I Ih l I l --! ! I I I I II
- 1 I I fx l l l l l -l I' IE I i i/ !
I i
~
I! I I d 41 b , j fl
) I ,
ll l l
! 5 il P Il/ I -l i I I l l 5, 5 '
I l/ II I I
-1II L II la I/I I I
I IlI I I i .
l l 4m I
-l I l/ l l I l 1 i l il P / I I l- 1 II I i I i I I
i
/l I l l I I si i I I i I ,
l '2 I /! lI ,
I i IIi 1 I I I I II ! l Idb /I I i I l l I I I P xI I I I , ! i i i
II l l l 1I I II l l I II l 1 1 I I I iliiliII ! I IIi l I i i l 1- l I -
! i ! I i i I i l i 1
-! I b ; II I I I I
. I I .a l I I I l1l I i i l I I i 1 -
Il! I I
I I .I II I I I i 1I I I I I I Ii i I l I II I I
!, : I I l! I I I I! I l I I I I I I l I I i I i I i l l I i I i l l 1 I Il ! I I i !
I ! I i I II I I II i i I i l , I III i l I , IiI! l 1 i ! I I 1. i l ! Iil I
- t II I I 1 I I I l t i I I II l 1 I II I i i I I II I I i I I I I I I I! l I l -
1I I I I D Ii' l ,i IIII I i IilI i l l'! l I; I I I I I I II! ! : I II I I i . l l
i .. ..
=.
4 a;;;<
4 .
o '
'900.
i i
. omr.senr .
O ttARGINAl. CO*'M. -
l fl WTT11 FANS t
t U
a' w
. 800 -
e oc . u a - . m e .
- u.
. 2W '
- 1400
a.
a w O c -
x . -
. a o .
.o
-> - . , . . . , y .=. ...
c - =
E.
u a . a
. . 1300 m 700 -
, 3 e
n -
- s. -,. . . * -
)
. 1200 0 . .
=
h -
600 40 50 10 20 30
-Steam Concentration, v/o ,j Tayco Surface Te=perature at Ignition vs. Steam Concentration .
FIGURE 2
~
I 5
h 5 -
1 - .- _ - - . - - - - .
, n.~:>.-:.. .... - . .,,-. _
ENCLOSURE 2 4
ADDITIONAL INFORMATION ON ADEQUACY OF PENS SEQUOYAH NUCLEAR PLANT l TVA submitted a sammary report on. September 17, 1982, that concluded
! that the permanent hydrogen mitigation system (PHMS) was an adequate hydrogen control system for the Sequoyah Nuclear Plant (SQN) that would perform its intended function in a manner that would provide adequate saf e ty margins. TVA continues to believe that the present PEMS design is adequate.
We do not beliwve that modifications to the upper compartment portion of the system are of much benefit because it is highly unlikely that flammable'aistures would ever be present there. Since the potential sources of hydrogen would all be located in the lower compartment, such flammable mixtures in the upper compartment would have had to bypass all 22 lower compartment igniters and al1 16 ice condenser upper plenum igniters without being ignited and burned down to the lower flammability limit. Analyses with the CLASIX containment code sh ow that burns do not occur in the upper compartment unless the lower flammability limit is arbitrarily assumed to be lower there than in any other compartment. .
As stated in a previous informal submittal, TVA believes that data on
~
minimum Tayco igniter surface temperature for ignition obtained at Whiteshell Nuclear Research Establishment (WNRE) is applicable for use in evaluating PHNS igniter performance in the' upper compartment.
These tests showed ignition of dry mixtures et a surf ace temperature of 1200'F, and ignition at high (40 percent) stcan concentrations at a temperature of 1350*F. As stated in the submittal, TVA proposed that 1350*F be used as the minimum Tayco igniter surface temperature required for ignition in the spray environment. The NRC staff has informally proposed a more conservative minimum temperature of 1500'F. Based on the modif red design of the igniter assembly spray shicid and on the recently-conducted spray tests reported in the previous informal submittal, it is TVA's judgment that the PHMS igniters in the upper compartment can be shown to maintain even this conservative temperature of 1500*F.
The bodified spray shield is designed to provide camplete coverage of the igniter surf ace f rom any spray droplets travelling at an angle up to 50' from the vertical (see attached figure). For an igniter located at the maximum distance of 10' from the shield, the, modified shield would be
- 22' deep and 26' wide. This would extend 13' beyond the front tip of the igniter and 121/2' on each side. It is our judgment that long-term spray-induced turbulence levels inside the upper compartment would not be great enough to cause significant prolonged spray impingement at greater than this severe angle (i.e., more horizontal than vertical). Again referring to the previous informal submittal, the igniters are located on walls f ar enough above the floor to avcid local turbulence from turning and f ar enough below the nozzles to avoid any horizontal component f rom the original traj ectory.
Ia cdditica, b2sid ca the rosant sprey tost dassrib:d ecrlier, ths
- Tayco igniter has been demonstrated to maintain surface temperatures in excess of 1500*F even when subj ected to a direct (no spray shield) continuous spray of mass flux equivalent to 20 percent of that average flux calculated to f all throughout the upper compartment.
(Note that the igniter maintained a tamperature above 1350*F for a spray mass flux of 60 percent.) This data demonstrates that margin exists in the igniter temperature even for significant additional spray impingement under the shield from.an angle greater than 50' . Also, the spray tests were conducted at 120-V ac while the upper compartment Igniter should operate at slightly higher voltage levels which would correspondingly increase the tamperature margins.
t In summary, TVA believes that hydrogen combustion is unlikely to occur in the upper compartment. Even if flammable mixtures do occur there, the PENS igniter assemblies would be able to maintain sufficient temperatures for ignition even under severe postulated spray environments. The modified spray shields, supported by recent direct spray data, would allow even the conservative surf ace temperature of 150*F to be maintained. TVA believes that this hardware modification should satisfactorily resolve any spray effect concerns without requiring further analysis or testing.
e
~
.7 - .
'4 *,
4 8 e
0
._d26 M s . - .**.m.w $ h a v -.
a -
4 ,
l '. .-
V .=
s
$0*
i . 8
/ r s.
g e e /
l 5 ,28 = 7 I g $" -
/
. i e 1 (w10 /
/ t
{
J l 7
/
I r' '
GPRAY - /
wisco I f.
(,*- 30,,
} / l ,j
/
I.
@MW r ,
j'
- / T siunT E R N
/
I I [, 20% DIRECT' SPRAY
' 0 T
.- surf >1500 7 60% DIRECT SPRAY WALL NOTE: IGNITER FIELD-LOC TED IN JUNCTION
~
T 0 BOX. MAXURIM SHIELD SIZE FOR IGNITER surf >1350 7 AT LOWEST ELEVATION IS 26"' WIDE X 22" DEEP-
+
- 1. .
- I 6 .- .
A e-6 O
e- .
p
-l i . !_.
[.
F :
r -
gg, r
t
/
/ C4
. , I l
/' /
a 5 22-
= ,
l gg -
/
/
lW /
e I f
(wD l I
/ /
}!
II t .
J f GPRAY_ - /
- ' l statt.D /
f i t,'-id, l .y
/ l /
' N s T6MnTER.
/ .
6" . ; .,1 ,- 20% DIRECT SPRAY
.'- ' T 0
, . surf > 1500 F 60% DIRECT SN
\A/ALL NOTE: IGNITER FIELD-LOC TED IN JUNCT' ION T 0 BOX. MAXIMUM SHIELD SIZE FOR IGNITER surf >1350F AT LOWEST ELEVATION IS 26"' WIDE X 22" DEEP 0
. 4 a
E