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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217M4461999-10-20020 October 1999 Forwards Rev 8 to Sequoyah Nuclear Plant Physical Security/ Contingency Plan, IAW 10CFR50.54(p).Encl Withheld,Per 10CFR73.21 ML20217J4151999-10-15015 October 1999 Forwards Request for Addl Info Re Util 990624 Application for Amend of TSs That Would Revise TS for Weighing of Ice Condenser Ice Baskets 05000327/LER-1999-002, Forwards LER 99-002-00 Re Start of Units 1 & 2 EDGs as Result of Cable Being Damaged During Installation of Thermo- Lag for Kaowool Upgrade Project1999-10-15015 October 1999 Forwards LER 99-002-00 Re Start of Units 1 & 2 EDGs as Result of Cable Being Damaged During Installation of Thermo- Lag for Kaowool Upgrade Project ML20217G1141999-10-0707 October 1999 Responds to from P Salas,Providing Response to NRC Risk Determination Associated with 990630 Flooding Event at Sequoyah Facility.Meeting to Discuss Risk Determination Issues Scheduled for 991021 in Atlanta,Ga ML20217B2981999-10-0606 October 1999 Discusses Closeout of GL 92-01,rev 1,suppl 1, Reactor Vessel Integrity, for Sequoyah Nuclear Plant,Units 1 & 2. NRC Also Hereby Solicits Any Written Comments That TVA May Have on Current Rvid Data by 991101 ML20217B8431999-10-0505 October 1999 Requests NRC Review & Approval of ASME Code Relief Requests That Were Identified in Plant Second 10-yr ISI Interval for Both Units.Encl 3 Provides Util Procedure for Calculation of ASME Code Coverage for Section XI Nondestructive Exams IR 05000327/19990041999-10-0101 October 1999 Ack Receipt of Providing Comments on Insp Repts 50-327/99-04 & 50-328/99-04.NRC Considered Comments for Apparent Violation Involving 10CFR50.59 Issue ML20217C7101999-10-0101 October 1999 Forwards Response to NRC 990910 RAI Re Sequoyah Nuclear Plant,Units 1 & 2 URI 50-327/98-04-02 & 50-328/98-04-02 Re Ice Weight Representative Sample ML20212J5981999-10-0101 October 1999 Forwards SE Accepting Request for Relief from ASME Boiler & Pressure Vessel Code,Section Xi,Requirements for Certain Inservice Insp at Plnat,Unit 1 ML20212M1081999-09-29029 September 1999 Confirms Intent to Meet with Utils on 991025 in Atlanta,Ga to Discuss Pilot Plants,Shearon Harris & Sequoyah Any Observations & Lessons Learned & Recommendations Re Implementation of Pilot Program ML20217A9451999-09-27027 September 1999 Forwards Insp Repts 50-327/99-05 & 50-328/99-05 on 990718- 0828.One Violation Identified & Being Treated as Non-Cited Violation ML20216J9351999-09-27027 September 1999 Responds to NRC Re Violations Noted in Insp Repts 50-327/99-04 & 50-328/99-04.Corrective Actions:Risk Determination Evaluation Was Performed & Licensee Concluded That Event Is in Green Regulatory Response Band ML20212F0751999-09-23023 September 1999 Forwards SER Granting Util 981021 Request for Relief from ASME Code,Section XI Requirements from Certain Inservice Insp at Sequoyah Nuclear Power Plant,Units 1 & 2 Pursuant to 10CFR50.55a(a)(3)(ii) ML20212F4501999-09-23023 September 1999 Forwards Amends 246 & 237 to Licenses DPR-77 & DPR-79, Respectively & Ser.Amends Approve Request to Revise TSs to Allow Use of Fully Qualified & Tested Spare Inverter in Place of Any of Eight Required Inverters ML20212M1911999-09-21021 September 1999 Discusses Exercise of Enforcement Discretion Re Apparent Violation Noted in Insp Repts 50-327/99-04 & 50-328/99-04 Associated with Implementation of Procedural Changes Which Resulted in Three Containment Penetrations Being Left Open ML20211Q0311999-09-10010 September 1999 Requests Written Documentation from TVA to Provide Technical Assistance to Region II Re TS Compliance & Ice Condenser Maint Practices at Plant ML20216F5441999-09-0707 September 1999 Provides Results of Risk Evaluation of 990630,flooding Event at Sequoyah 1 & 2 Reactor Facilities.Event Was Documented in Insp Rept 50-327/99-04 & 50-328/99-04 & Transmitted in Ltr, ML20211N5681999-09-0101 September 1999 Submits Clarification of Two Issues Raised in Insp Repts 50-327/99-04 & 50-328/99-04,dtd 990813,which Was First Insp Rept Issued for Plant Under NRC Power Reactor Oversight Process Pilot Plant Study ML20211G5881999-08-27027 August 1999 Submits Summary of 990820 Management Meeting Re Plant Performance.List of Attendees & Matl Used in Presentation Enclosed ML20211F8891999-08-25025 August 1999 Forwards Sequoyah Nuclear Plant Unit 1 Cycle 9 Refueling Outage, Re Completed SG Activities,Per TSs 4.4.5.5.b & 4.4.5.5.c ML20211A1851999-08-16016 August 1999 Forwards Proprietary TR WCAP-15128 & non-proprietary Rept WCAP-15129 for NRC Review.Repts Are Provided in Advance of TS Change That Is Being Prepared to Support Cycle 10 Rfo. Proprietary TR Withheld,Per 10CFR2.790 ML20210V1471999-08-13013 August 1999 Forwards Insp Repts 50-327/99-04 & 50-328/99-04 on 990601- 0717.One Potentially Safety Significant Issue Identified.On 990630,inadequate Performance of Storm Drain Sys Caused Water from Heavy Rainfall to Backup & Flood Turbine Bldg ML20211A1921999-08-12012 August 1999 Requests Proprietary TR WCAP-15128, Depth-Based SG Tube Repair Criteria for Axial PWSCC at Dented TSP Intersections, Be Withheld from Public Disclosure Per 10CFR2.790 ML20210Q5011999-08-0505 August 1999 Informs That NRC Plans to Administer Gfes of Written Operator Licensing Exam on 991006 at Sequoyah Nuclear Plant. Sample Registration Ltr Encl ML20210L4291999-08-0202 August 1999 Forwards Sequoyah Nuclear Plant Unit 2 Cycle 9 12-Month SG Insp Rept & SG-99-07-009, Sequoyah Unit-2 Cycle 10 Voltage-Based Repair Criteria 90-Day Rept. Repts Submitted IAW TS 4.4.5.5.b & TS 4.4.5.5.c ML20210L1611999-07-30030 July 1999 Forwards Request for Relief RV-4 Re ASME Class 1,2 & 3 Prvs, Per First ten-year Inservice Test Time Interval.Review & Approval of RV-4 Is Requested to Support Unit 1 Cycle 10 Refueling Outage,Scheduled to Start 000213 ML20210G5301999-07-28028 July 1999 Forwards Sequoyah Nuclear Plant Unit 2 ISI Summary Rept That Contains Historical Record of Repairs,Replacement & ISI & Augmented Examinations That Were Performed on ASME Code Class 1 & 2 Components from 971104-990511 ML20211B9661999-07-26026 July 1999 Informs That Sequoyah Nuclear Plant Sewage Treatment Plant, NPDES 0026450 Outfall 112,is in Standby Status.Flow Has Been Diverted from Sys Since Jan 1998 ML20210B2521999-07-14014 July 1999 Confirms 990712 Telcon Between J Smith of Licensee Staff & M Shannon of NRC Re semi-annual Mgt Meeting Schedule for 990820 in Atlanta,Ga to Discuss Recent Sequoyah Nuclear Plant Performance ML20210J1091999-07-10010 July 1999 Submits Suggestions & Concerns Re Y2K & Nuclear Power Plants ML20196K0381999-06-30030 June 1999 Provides Written Confirmation of Completed Commitment for Final Implementation of Thermo-Lag 330-1 Fire Barrier Corrective Actions at Snp,Per GL 92-08 ML20209E4071999-06-30030 June 1999 Forwards Insp Repts 50-327/99-03 & 50-328/99-03 on 990328- 0531.Violations Being Treated as Noncited Violations ML20196J8261999-06-28028 June 1999 Forwards Safety Evaluation Authorizing Request for Relief from ASME Boiler & Pressure Vessel Code,Section XI Requirements for Certain Inservice Inspections at Sequoyah Nuclear Plant,Units 1 & 2 ML20196G7881999-06-22022 June 1999 Informs NRC of Changes That Util Incorporated Into TS Bases Sections & Trm.Encl Provides Revised TS Bases Pages & TRM Affected by Listed Revs ML20196G1801999-06-21021 June 1999 Requests Termination of SRO License SOP-20751-1,for Lf Hardin,Effective 990611.Subject Individual Resigned from Position at TVA ML20195G1821999-06-0808 June 1999 Requests NRC Review & Approval of ASME Code Relief for ISI Program.Encl 1 Provides Relief Request 1-ISI-14 That Includes Two Attachments.Encl 2 Provides Copy of Related ASME Code Page ML20195E9521999-06-0707 June 1999 Requests Relief from Specific Requirements of ASME Section Xi,Subsection IWE of 1992 Edition,1992 Addenda.Util Has Determined That Proposed Alternatives Would Provide Acceptable Level of Quality & Safety ML20195E9311999-05-28028 May 1999 Informs of Planned Insp Activities for Licensee to Have Opportunity to Prepare for Insps & Provide NRC with Feedback on Any Planned Insps Which May Conflict with Plant Activities ML20195B3631999-05-21021 May 1999 Requests Termination of SRO License for Tj Van Huis,Per 10CFR50.74(a).TJ Van Huis Retired from Util,Effective 990514 ML20207A5721999-05-20020 May 1999 Forwards Correction to Previously Issued Amend 163 to License DPR-79 Re SR 4.1.1.1.1.d Inadvertently Omitted from Pp 3/4 1-1 of Unit 2 TS ML20206Q8791999-05-13013 May 1999 Forwards L36 9990415 802, COLR for Sequoyah Nuclear Plant Unit 2,Cycle 10, IAW Plant TS 6.9.1.14.c 05000327/LER-1999-001, Forwards LER 99-001-00 Re Condition That Resulted in Granting of Enforcement Discretion,Per Failure of Centrifugal Charging Pump.Condition Being Reported IAW 10CFR50.73(a)(2)(i)(B) & (a)(2)(iv)1999-05-11011 May 1999 Forwards LER 99-001-00 Re Condition That Resulted in Granting of Enforcement Discretion,Per Failure of Centrifugal Charging Pump.Condition Being Reported IAW 10CFR50.73(a)(2)(i)(B) & (a)(2)(iv) ML20206M9341999-05-10010 May 1999 Forwards Rept of SG Tube Plugging During Unit 2 Cycle 9 Refueling Outage,As Required by TS 4.4.5.5.a.ISI of Unit 2 SG Tubes Was Completed on 990503 ML20206K6271999-05-0606 May 1999 Requests Termination of SRO License for MR Taggart,License SOP-21336 Due to Resignation on 990430 ML20206J2061999-05-0404 May 1999 Requests Relief from Specified ISI Requirements in Section XI of ASME B&PV Code.Tva Requests Approval to Use Wire Type Penetrameters in Lieu of Plaque Type Penetrameters for Performing Radiographic Insps.Specific Relief Request,Encl ML20209J0391999-04-27027 April 1999 Forwards Annual Radioactive Effluent Release Rept, Radiological Impact Assessment Rept & Rev 41 to ODCM, for Period of Jan-Dec 1998 ML20206C6541999-04-23023 April 1999 Forwards Response to NRC 990127 RAI Re GL 96-05 for Sequoyah Nuclear Plant,Units 1 & 2 ML20206C0841999-04-23023 April 1999 Forwards Insp Repts 50-327/99-02 & 50-328/99-02 on 990214-0327.No Violations Noted ML20206B9591999-04-20020 April 1999 Responds to 990417 Request That NRC Exercise Discretion Not to Enforce Compliance with Actions Required in Unit 1 TS 3.1.2.2,3.1.2.4 & 3.5.2 & Documents 990417 Telephone Conversation When NRC Orally Issued NOED ML20205S5891999-04-17017 April 1999 Documents Request for Discretionary Enforcement for Unit 1 TS LCOs 3.1.2.2,3.1.2.4 & 3.5.2 to Support Completion of Repairs & Testing for 1B-B Centrifugal Charging Pump (CCP) 1999-09-07
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217M4461999-10-20020 October 1999 Forwards Rev 8 to Sequoyah Nuclear Plant Physical Security/ Contingency Plan, IAW 10CFR50.54(p).Encl Withheld,Per 10CFR73.21 05000327/LER-1999-002, Forwards LER 99-002-00 Re Start of Units 1 & 2 EDGs as Result of Cable Being Damaged During Installation of Thermo- Lag for Kaowool Upgrade Project1999-10-15015 October 1999 Forwards LER 99-002-00 Re Start of Units 1 & 2 EDGs as Result of Cable Being Damaged During Installation of Thermo- Lag for Kaowool Upgrade Project ML20217B8431999-10-0505 October 1999 Requests NRC Review & Approval of ASME Code Relief Requests That Were Identified in Plant Second 10-yr ISI Interval for Both Units.Encl 3 Provides Util Procedure for Calculation of ASME Code Coverage for Section XI Nondestructive Exams ML20217C7101999-10-0101 October 1999 Forwards Response to NRC 990910 RAI Re Sequoyah Nuclear Plant,Units 1 & 2 URI 50-327/98-04-02 & 50-328/98-04-02 Re Ice Weight Representative Sample ML20216J9351999-09-27027 September 1999 Responds to NRC Re Violations Noted in Insp Repts 50-327/99-04 & 50-328/99-04.Corrective Actions:Risk Determination Evaluation Was Performed & Licensee Concluded That Event Is in Green Regulatory Response Band ML20211N5681999-09-0101 September 1999 Submits Clarification of Two Issues Raised in Insp Repts 50-327/99-04 & 50-328/99-04,dtd 990813,which Was First Insp Rept Issued for Plant Under NRC Power Reactor Oversight Process Pilot Plant Study ML20211F8891999-08-25025 August 1999 Forwards Sequoyah Nuclear Plant Unit 1 Cycle 9 Refueling Outage, Re Completed SG Activities,Per TSs 4.4.5.5.b & 4.4.5.5.c ML20211A1851999-08-16016 August 1999 Forwards Proprietary TR WCAP-15128 & non-proprietary Rept WCAP-15129 for NRC Review.Repts Are Provided in Advance of TS Change That Is Being Prepared to Support Cycle 10 Rfo. Proprietary TR Withheld,Per 10CFR2.790 ML20211A1921999-08-12012 August 1999 Requests Proprietary TR WCAP-15128, Depth-Based SG Tube Repair Criteria for Axial PWSCC at Dented TSP Intersections, Be Withheld from Public Disclosure Per 10CFR2.790 ML20210L4291999-08-0202 August 1999 Forwards Sequoyah Nuclear Plant Unit 2 Cycle 9 12-Month SG Insp Rept & SG-99-07-009, Sequoyah Unit-2 Cycle 10 Voltage-Based Repair Criteria 90-Day Rept. Repts Submitted IAW TS 4.4.5.5.b & TS 4.4.5.5.c ML20210L1611999-07-30030 July 1999 Forwards Request for Relief RV-4 Re ASME Class 1,2 & 3 Prvs, Per First ten-year Inservice Test Time Interval.Review & Approval of RV-4 Is Requested to Support Unit 1 Cycle 10 Refueling Outage,Scheduled to Start 000213 ML20210G5301999-07-28028 July 1999 Forwards Sequoyah Nuclear Plant Unit 2 ISI Summary Rept That Contains Historical Record of Repairs,Replacement & ISI & Augmented Examinations That Were Performed on ASME Code Class 1 & 2 Components from 971104-990511 ML20210J1091999-07-10010 July 1999 Submits Suggestions & Concerns Re Y2K & Nuclear Power Plants ML20196K0381999-06-30030 June 1999 Provides Written Confirmation of Completed Commitment for Final Implementation of Thermo-Lag 330-1 Fire Barrier Corrective Actions at Snp,Per GL 92-08 ML20196G7881999-06-22022 June 1999 Informs NRC of Changes That Util Incorporated Into TS Bases Sections & Trm.Encl Provides Revised TS Bases Pages & TRM Affected by Listed Revs ML20196G1801999-06-21021 June 1999 Requests Termination of SRO License SOP-20751-1,for Lf Hardin,Effective 990611.Subject Individual Resigned from Position at TVA ML20195G1821999-06-0808 June 1999 Requests NRC Review & Approval of ASME Code Relief for ISI Program.Encl 1 Provides Relief Request 1-ISI-14 That Includes Two Attachments.Encl 2 Provides Copy of Related ASME Code Page ML20195E9521999-06-0707 June 1999 Requests Relief from Specific Requirements of ASME Section Xi,Subsection IWE of 1992 Edition,1992 Addenda.Util Has Determined That Proposed Alternatives Would Provide Acceptable Level of Quality & Safety ML20195B3631999-05-21021 May 1999 Requests Termination of SRO License for Tj Van Huis,Per 10CFR50.74(a).TJ Van Huis Retired from Util,Effective 990514 ML20206Q8791999-05-13013 May 1999 Forwards L36 9990415 802, COLR for Sequoyah Nuclear Plant Unit 2,Cycle 10, IAW Plant TS 6.9.1.14.c 05000327/LER-1999-001, Forwards LER 99-001-00 Re Condition That Resulted in Granting of Enforcement Discretion,Per Failure of Centrifugal Charging Pump.Condition Being Reported IAW 10CFR50.73(a)(2)(i)(B) & (a)(2)(iv)1999-05-11011 May 1999 Forwards LER 99-001-00 Re Condition That Resulted in Granting of Enforcement Discretion,Per Failure of Centrifugal Charging Pump.Condition Being Reported IAW 10CFR50.73(a)(2)(i)(B) & (a)(2)(iv) ML20206M9341999-05-10010 May 1999 Forwards Rept of SG Tube Plugging During Unit 2 Cycle 9 Refueling Outage,As Required by TS 4.4.5.5.a.ISI of Unit 2 SG Tubes Was Completed on 990503 ML20206K6271999-05-0606 May 1999 Requests Termination of SRO License for MR Taggart,License SOP-21336 Due to Resignation on 990430 ML20206J2061999-05-0404 May 1999 Requests Relief from Specified ISI Requirements in Section XI of ASME B&PV Code.Tva Requests Approval to Use Wire Type Penetrameters in Lieu of Plaque Type Penetrameters for Performing Radiographic Insps.Specific Relief Request,Encl ML20209J0391999-04-27027 April 1999 Forwards Annual Radioactive Effluent Release Rept, Radiological Impact Assessment Rept & Rev 41 to ODCM, for Period of Jan-Dec 1998 ML20206C6541999-04-23023 April 1999 Forwards Response to NRC 990127 RAI Re GL 96-05 for Sequoyah Nuclear Plant,Units 1 & 2 ML20205S5891999-04-17017 April 1999 Documents Request for Discretionary Enforcement for Unit 1 TS LCOs 3.1.2.2,3.1.2.4 & 3.5.2 to Support Completion of Repairs & Testing for 1B-B Centrifugal Charging Pump (CCP) ML20205B1091999-03-19019 March 1999 Submits Response to NRC Questions Concerning Lead Test Assembly Matl History,Per Request ML20204H0161999-03-19019 March 1999 Resubmits Util 990302 Response to Violations Noted in Insp Repts 50-327/98-11 & 50-328/98-11.Corrective Actions:Lessons Learned from Event Have Been Provided to Operating Crews ML20204E8251999-03-0505 March 1999 Forwards Sequoyah Nuclear Plant,Four Yr Simulator Test Rept for Period Ending 990321, in Accordance with Requirements of 10CFR55.45 ML20207E6851999-03-0202 March 1999 Responds to NRC Re Violations Noted in Insp Repts 50-327/98-11 & 50-328/98-11.Corrective Actions:Lessons Learned from Event Have Been Provided to Operating Crews ML20207J1171999-01-29029 January 1999 Forwards Copy of Final Exercise Rept for Full Participation Ingestion Pathway Exercise of Offsite Radiological Emergency Response Plans site-specific to Sequoyah NPP ML20202A7141999-01-20020 January 1999 Provides Request for Relief for Delaying Repair on Section of ASME Code Class 3 Piping within Essential Raw Cooling Water Sys ML20198S7141998-12-29029 December 1998 Forwards Cycle 10 Voltage-Based Repair Criteria 90-Day Rept, Per GL 95-05.Rept Is Submitted IAW License Condition 2.C.(9)(d) 05000327/LER-1998-004, Forwards LER 98-004-00,providing Details Concerning Inability to Complete Surveillance within Required Time Interval1998-12-21021 December 1998 Forwards LER 98-004-00,providing Details Concerning Inability to Complete Surveillance within Required Time Interval ML20198D5471998-12-14014 December 1998 Requests That License OP-20313-2 for Je Wright,Be Terminated IAW 10CFR50.74(a).Individual Retiring ML20197J5541998-12-10010 December 1998 Forwards Unit 1 Cycle 9 90-Day ISI Summary Rept IAW IWA-6220 & IWA-6230 of ASME Code,Section Xi.Request for Relief Will Be Submitted to NRC Timeframe to Support Second 10-year Insp Interval,Per 10CFR50.55a 05000327/LER-1998-003, Forwards LER 98-003-00 Re Automatic Reactor Trip with FW Isolation & Auxiliary FW Start as Result of Failure of Vital Inverter & Second Inverter Failure.Event Is Being Reported IAW 10CFR50.73(a)(2)(iv)1998-12-0909 December 1998 Forwards LER 98-003-00 Re Automatic Reactor Trip with FW Isolation & Auxiliary FW Start as Result of Failure of Vital Inverter & Second Inverter Failure.Event Is Being Reported IAW 10CFR50.73(a)(2)(iv) ML20196F9841998-11-25025 November 1998 Provides Changes to Calculated Peak Fuel Cladding Temp, Resulting from Recent Changes to Plant ECCS Evaluation Model ML20195H7891998-11-17017 November 1998 Requests NRC Review & Approval of Five ASME Code Relief Requests Identified in Snp Second ten-year ISI Interval for Units 1 & 2 ML20195E4991998-11-12012 November 1998 Forwards Rev 7 to Physical Security/Contingency Plan.Rev Adds Requirement That Security Personnel Will Assess Search Equipment Alarms & Add Definition of Major Maint.Rev Withheld (Ref 10CFR2.790(d)(1)) 05000328/LER-1998-002, Forwards LER 98-002-00 Re Automatic Turbine & Reactor Trip, Resulting from Failure of Sudden Pressure Relay on 'B' Phase Main Transformer1998-11-10010 November 1998 Forwards LER 98-002-00 Re Automatic Turbine & Reactor Trip, Resulting from Failure of Sudden Pressure Relay on 'B' Phase Main Transformer ML20195G5701998-11-10010 November 1998 Documents Util Basis for 981110 Telcon Request for Discretionary Enforcement for Plant TS 3.8.2.1,Action B,For 120-VAC Vital Instrument Power Board 1-IV.Licensee Determined That Inverter Failed Due to Component Failure ML20155J4031998-11-0505 November 1998 Provides Clarification of Topical Rept Associated with Insertion of Limited Number of Lead Test Assemblies Beginning with Unit 2 Operating Cycle 10 Core ML20154R9581998-10-21021 October 1998 Requests Approval of Encl Request for Relief ISI-3 from ASME Code Requirements Re Integrally Welded Attachments of Supports & Restraints for AFW Piping ML20155B1481998-10-21021 October 1998 Informs That as Result of Discussion of Issues Re Recent Events in Ice Condenser Industry,Ice Condenser Mini-Group (Icmg),Decided to Focus Efforts on Review & Potential Rev of Ice condenser-related TS in Order to Clarify Issues ML20154K1581998-10-13013 October 1998 Forwards Rept Re SG Tube Plugging Which Occurred During Unit 1 Cycle 9 Refueling Outage,Per TS 4.4.5.5.a.ISI of Unit 1 SG Was Completed on 980930 ML20154H6191998-10-0808 October 1998 Forwards Rev 0 to Sequoyah Nuclear Plant Unit 1 Cycle 10 COLR, IAW TS 6.9.1.14.c 05000328/LER-1998-001, Forwards LER 98-001-00 Providing Details Re Automatic Turbine & Reactor Trip Due to Failure of Sudden Pressure Relay on 'B' Phase Main Transformer1998-09-28028 September 1998 Forwards LER 98-001-00 Providing Details Re Automatic Turbine & Reactor Trip Due to Failure of Sudden Pressure Relay on 'B' Phase Main Transformer ML20151W4901998-09-0303 September 1998 Responds to NRC Re Violations Noted in Insp Repts 50-327/98-07 & 50-328/98-07.Corrective Actions:Revised Per SQ971279PER to Address Hardware Issues of Hysteresis, Pressure Shift & Abnormal Popping Noise 1999-09-27
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059K6661990-09-17017 September 1990 Forwards Evaluation That Provides Details of Plug Cracks & Justification for Continued Operation Until 1993 ML20059H4031990-09-10010 September 1990 Discusses Plant Design Baseline & Verification Program Deficiency D.4.3-3 Noted in Insp Repts 50-327/86-27 & 50-328/86-27.Evaluation Concluded That pre-restart Walkdown Data,Loops 1 & 2 Yielded Adequate Design Input ML20059E1851990-08-31031 August 1990 Responds to NRC Re Violations Noted in Insp Repts 50-327/90-22 & 50-328/90-22.Corrective Actions:Extensive Mgt Focus Being Applied to Improve Overtime Use Controls ML20059E2881990-08-31031 August 1990 Forwards Addl Info Re Alternate Testing of Reactor Vessel Head & Internals Lifting Rigs,Per NUREG-0612.Based on Listed Hardships,Util Did Not Choose 150% Load Test Option ML20059H1831990-08-31031 August 1990 Forwards Nonproprietary PFE-F26NP & Proprietary PFE-F26, Sequoyah Nuclear Plan Unit 1,Cycle 5 Restart Physics Test Summary, Re Testing Following Vantage 5H Fuel Assembly installation.PFE-F26 Withheld (Ref 10CFR2.790(b)(4)) ML18033B5031990-08-31031 August 1990 Forwards Financial Info Required to Assure Retrospective Premiums,Per 10CFR140 & 771209 Ltr ML20028G8341990-08-28028 August 1990 Forwards Calculation SCG1S361, Foundation Investigation of ERCW Pumping Station Foundation Cells. ML20063Q2471990-08-20020 August 1990 Submits Implementation Schedule for Cable Tray Support Program.Util Proposes Deferral of Portion of Remaining Activities Until After Current Unit 2 Cycle 4 Refueling Outage,Per 900817 Meeting.Tva Presentation Matl Encl ML20056B5181990-08-20020 August 1990 Responds to NRC Re Order Imposing Civil Monetary Penalty & Violations Noted in Insp Repts 50-327/90-01 & 50-328/90-01.Corrective Actions:Organizational Capabilities Reviewed.Payment of Civil Penalty Wired to NRC ML20063Q2461990-08-17017 August 1990 Forwards Cable Test Program Resolution Plan to Resolve Issues Re Pullbys,Jamming & Vertical Supported Cable & TVA- Identified Cable Damage.Tva Commits to Take Actions Prior to Startup to Verify Integrity of safety-related Cables ML20059A5121990-08-15015 August 1990 Provides Clarification of Implementation of Replacement Items Project at Plant for Previously Procured Warehouse Inventory.Util Committed to 100% Dedication of Commercial Grade,Qa,Level Ii,Previous Procurement Warehouse Spare ML20058M2321990-08-0707 August 1990 Forwards Rept of 900709 Fishkill,Per Requirements in App B, Environ Tech Spec,Subsections 4.1.1 & 5.4.2.Sudden Water Temp Increase Killed Approximately 150 Fish in Plant Diffuser Pond ML20058N2361990-08-0707 August 1990 Confirms That Requalification Program Evaluation Ref Matl Delivered to Rd Mcwhorter on 900801.Ref Matl Needed to Support NRC Preparation for Administering Licensed Operator Requalification Exams in Sept 1990 ML20058M4471990-07-27027 July 1990 Responds to Unresolved Items Which Remain Open from Insp Repts 50-327/90-18 & 50-328/90-18.TVA in Agreement W/Nrc on Scope of Work Required to Address Concerns W/Exception of Design Basis Accident & Zero Period Accelaration Effects ML20058M0111990-07-27027 July 1990 Forwards Addl Info Re Plant Condition Adverse to Quality Rept Concerning Operability Determination.Probability of Cable Damage During Installation Low.No Programmatic Cable Installation Problems Exist ML20055J3531990-07-27027 July 1990 Forwards Revised Commitment to Resolve EOP Step Deviation Document Review Comments ML20055J0771990-07-26026 July 1990 Requests Termination of Senior Reactor Operator License SOP-20830 for Jh Sullivan Due to Resignation from Util ML20055G6611990-07-17017 July 1990 Forwards Justification for Continued Operation for safety- Related Cables Installed at Plant,Per 900717 Telcon.No Operability Concern Exists at Plant & No Programmatic Problems Have Been Identified.Summary of Commitments Encl ML20058L7001990-07-16016 July 1990 Forwards Response to SALP Repts 50-327/90-09 & 50-328/90-09 for 890204 - 900305,including Corrective Actions & Improvements Being Implemented ML20055F6151990-07-13013 July 1990 Provides Addl Bases for Util 900320 Proposal to Discontinue Review to Identify Maint Direct Charge molded-case Circuit Breakers Procured Between Aug 1983 & Dec 1984,per NRC Bulletin 88-010.No Significant Assurance Would Be Expected ML20044B2211990-07-12012 July 1990 Forwards Addl Info Clarifying Certain Conclusions & Recommendation in SER Re First 10-yr Interval Inservice Insp Program ML20055D2531990-07-0202 July 1990 Provides Status of Q-list Development at Plant & Revises Completion Date for Effort.Implementation of Q-list Would Cause Unnecessary & Costly Delays in Replanning Maint,Mod, outage-related Activities & Associated Procedure Revs ML20043H9061990-06-21021 June 1990 Responds to Generic Ltr 90-04, Request for Info on Status of Licensee Implementaion of Generic Safety Issues Resolved W/Imposition or Requirements or Corrective Actions. No Commitments Contained in Submittal ML20043H2281990-06-18018 June 1990 Informs of Issue Recently Identified During Startup of Facility from Cycle 4 Refueling Outage & How Issue Addressed to Support Continued Escalation to 100% Power,Per 900613 & 14 Telcons ML20043G4901990-06-14014 June 1990 Forwards Tabs for Apps a & B to Be Inserted Into Util Consolidated Nuclear Power Radiological Emergency Plan ML20043F9261990-06-13013 June 1990 Responds to NRC Bulletin 89-002, Stress Corrosion Cracking of High-Hardness Type 410 Stainless Steel Internal Preloaded Bolting in Anchor/Darling Model S3502 Swing Check Valves or Valves of Similar Design. ML20043F9301990-06-13013 June 1990 Responds to NRC 900516 Ltr Re Violations Noted in Insp Repts 50-327/90-17 & 50-328/90-17.Corrective Action:Test Director & Supervisor Involved Given Appropriate Level of Disciplinary Action ML20043H0361990-06-11011 June 1990 Forwards Supplemental Info Re Unresolved Item 88-12-04 Addressing Concern W/Double Differentiation Technique Used to Generate Containment Design Basis Accident Spectra,Per 900412 Request ML20043D9921990-06-0505 June 1990 Responds to NRC 900507 Ltr Re Violations Noted in Insp Repts 50-327/90-14 & 50-328/90-14.Corrective Actions:Util Reviewed Issue & Determined That Trains a & B Demonstrated Operable in Jan & Apr,Respectively of 1989 ML20043C2821990-05-29029 May 1990 Requests Relief from ASME Section XI Re Hydrostatic Pressure Test Requirements Involving RCS & Small Section of Connected ECCS Piping for Plant.Replacement & Testing of Check Valve 1-VLV-63-551 Presently Scheduled for Completion on 900530 ML20043C0581990-05-29029 May 1990 Forwards Response to NRC 900426 Ltr Re Violations Noted in Insp Repts 50-327/90-15 & 50-328/90-15.Response Withheld (Ref 10CFR73.21) ML20043B3051990-05-22022 May 1990 Forwards Detailed Scenario for 900711 Radiological Emergency Plan Exercise.W/O Encl ML20043B1201990-05-18018 May 1990 Forwards, Diesel Generator Voltage Response Improvement Rept. Combined Effect of Resetting Exciter Current Transformers to Achieve flat-compounding & Installing Electronic Load Sequence Timers Produced Acceptable Voltage ML20043A6101990-05-15015 May 1990 Forwards Rev 16 to Security Personnel Training & Qualification Plan.Rev Withheld (Ref 10CFR2.790) ML20043A2391990-05-15015 May 1990 Forwards Revised Tech Spec Pages to Support Tech Spec Change 89-27 Re Steam Generator Water Level Adverse Trip Setpoints for Reactor Trip Sys Instrumentation & Esfas. Encl Reflects Ref Leg Heatup Environ Allowance ML20043A0581990-05-11011 May 1990 Forwards Cycle 5 Redesign Peaking Factor Limit Rept for Facility.Unit Redesigned During Refueling Outage Due to Removal & Replacement of Several Fuel Assemblies Found to Contain Leaking Fuel Rods ML20043A0571990-05-10010 May 1990 Forwards List of Commitments to Support NRC Review of Eagle 21 Reactor Protection Sys Function Upgrade,Per 900510 Telcon ML20042G9771990-05-0909 May 1990 Responds to NRC 900412 Ltr Re Violations Noted in Insp Repts 50-327/90-01 & 50-328/90-01 & Proposed Imposition of Civil Penalty.Corrective Actions:Rhr Pump 1B-B Handswitch in pull- to-lock Position to Ensure One Train of ECCS Operable ML20042G4651990-05-0909 May 1990 Provides Addl Info Re Plant Steam Generator Low Water Level Trip Time Delay & Function of P-8 Reactor Trip Interlock,Per 900430 Telcon.Trip Time Delay Does Not Utilize P-8 Interlock in Any Manner ML20042G4541990-05-0909 May 1990 Provides Notification of Steam Generator Tube Plugging During Unit 1 Cycle 4 Refueling Outage,Per Tech Specs 4.4.5.5.a.Rept of Results of Inservice Insp to Be Submitted by 910427.Summary of Tubes Plugged in Unit 1 Encl ML20042G0441990-05-0808 May 1990 Forwards Nonproprietary WCAP-11896 & WCAP-8587,Suppl 1 & Proprietary WCAP-8687,Suppls 2-E69A & 2-E69B & WCAP-11733 Re Westinghouse Eagle 21 Process Protection Sys Components Equipment Qualification Test Rept.Proprietary Rept Withheld ML20042G1431990-05-0808 May 1990 Forwards WCAP-12588, Sequoyah Eagle 21 Process Protection Sys Replacement Hardware Verification & Validation Final Rept. Info Submitted in Support of Tech Spec Change 89-27 Dtd 900124 ML20042G1001990-05-0808 May 1990 Forwards Proprietary WCAP 12504 & Nonproprietary WCAP 12548, Summary Rept Process Protection Sys Eagle 21 Upgrade,Rtd Bypass Elimination,New Steam Line Break Sys,Medical Signal Selector .... Proprietary Rept Withheld (Ref 10CFR2.790) ML20042G1701990-05-0808 May 1990 Provides Addl Info Re Eagle 21 Upgrade to Plant Reactor Protection Sys,Per 900418-20 Audit Meeting.Partial Trip Output Board Design & Operation Proven by Noise,Fault,Surge & Radio Frequency Interference Testing Noted in WCAP-11733 ML20042G1231990-05-0707 May 1990 Forwards Detailed Discussion of Util Program & Methodology Used at Plant to Satisfy Intent of Reg Guide 1.97,Rev 2 Re Licensing Position on post-accident Monitoring ML20042F7741990-05-0404 May 1990 Informs of Completion of Eagle 21 Verification & Validation Activities Re Plant Process Protection Sys Upgrade.No Significant Disturbances Noted from NRC Completion Date of 900420 ML20042F1691990-05-0303 May 1990 Responds to NRC Bulletin 88-009, Thimble Tube Thinning in Westinghouse Electric Corporation Reactors. Wear Acceptance Criteria Established & Appropriate Corrective Actions Noted. Criteria & Corresponding Disposition Listed ML20042G1381990-04-26026 April 1990 Forwards Westinghouse 900426 Ltr to Util Providing Supplemental Info to Address Questions Raised by NRC Re Eagle-21 Process Protection Channels Required for Mode 5 Operation at Facilities ML20042E9641990-04-26026 April 1990 Forwards Rev 24 to Physical Security/Contingency Plan.Rev Withheld (Ref 10CFR73.21) ML20012E6181990-03-28028 March 1990 Discusses Reevaluation of Cable Pullby Issue at Plant in Light of Damage Discovered at Watts Bar Nuclear Plant. Previous Conclusions Drawn Re Integrity of Class 1E Cable Sys Continue to Be Valid.Details of Reevaluation Encl 1990-09-17
[Table view] |
Text
TENNESSEE VALLEY AUTHORITY CH ATTANOOGA. TENNESSEE 37401 400 Chestnut Street Tower II Dece=ber 19, 1979 Director of Nuclear Reactor Regulation Attention: Mr. L. S. Rubenstein, Acting Chief Light Water Reactors Branch No. 4 Division of Proj ect Manage =ent U.S. Nuclear Regulatory Co==ission Washington, DC 20555
Dear Mr. Rubenstein:
In the hatter of the Application of ) Docket Sos. 50-327 Tennessee Valley Authority )50-32S Enclosed are proposed revisions to the Sequoyah Nuclear Plant Final Safety Analysis Report (FSAR). These revisions are required to res.lve inconsistencies between the draft technical spec 2Neatieno and the FSAR. Anend ent c4 will incorporate these revisions in the Sequoyah FSAR.
Very truly yours, TEhTESSEE VALLEY AUTHORITY
\
L. M. Mills, Olanager Nuclear Regulation and Safety Enclosure 1645 043 @s i 7912'270533 i 4 , z u c: =
on an eight grid assembly is thus less than 5 percent greater than the lif t j force on a seven grid assembly for the same flow rate. Reference 5 shows that ,
lift off of an eight grid fuel assembly (5 percent greater than the seven grid ,
assembly shown) is not predicted during a postulated pump overspeed transient even though it is not necessary to preclude lif t of f. Additionally, the flow ,
29 rates used in the test are based on a plant with flow ratese ,.' 3 mme-greater than Sequoyah. This results in additional margin for Sequoyah since the lift force is reduced.
The hydraulic loads during normal operation can be obtained from Reference 5 by adjustin: the loads fcr the Sequcyah pressure drop and flow rate. The effect II of startup and shutdown transients are shewn to be inconsequential in Ref erence ,
5.
4.4.2.8 Correlation and Physical Data .m 4.4.2.8.1 Surface Heat Transfer Coefficients Forced convection heat transfer coefficients are obtained from the familiar ,
Dittus-Boelter correlation (Reference 53), with the properties evaluated at bulk '
fluid conditions: -
hD DG 0.8 Cu 0.4 ,
K
= 0.023 ( ) ( f,A ) (4. 4-15 )
u where: !
9
~
h = heat transfer coefficient, BTU /hr-f t *F De = equivalent diameter, ft K = thermal conductivity, BTU /hr-ft *F '
G = mass velocity, lb/hr-ft 2 u = dynamic viscosity, lb/ft-hr C = heat capacity, BTU /lb *F F This correlation has been shown to be conservative (Reference 54) for rod bundle geometries with pitch to diameter ratioe in the range used by P'a~ds. 6 The onset of nucleate boiling occurs when the clad wall temperature reaches f the amount of superheat predicted by Thom's (Reference 55) correlation. After ,
this occurrence the outer clad wall temperature is determined by:
O 4.4-20 November 27, 1974 ,
2 1645 046
f i
SNP-23 through 27), and by measurements of the fuel and clad dimensions during fabrication. The resulting uncertainties are then used in all evaluations t involving the fuel temperature. The ef fect of densification on fuel temperature uncertainties is presented in Reference 6.
f In additica to the temperature uncertainty described above, the measure- ,
ment uncertainty in determining the local power, and the effect of density and enrichment variations on the local power are considered in establishing the heat flux hot channel factor. These uncertainties are described in Subparagraph 4.3.2.2.1.
Reacter trip setpoints as specified in the Technical Specifications Subsection i 16.2.3 include allowance for instrument and measurement uncertainties such b as calorimetric error, instrument drif t and channel reproducibility, temperature measurement uncertainties, noise, and heat capacity variations.
Uncertainty in determining the cladding temperature results f rem uncertainties in the crud and oxide thicknesses. Because of the excellent heat transfer between the surf ace of the rod and the coolant , the film temperature drop ,
does not appreciably centribute to the uncertainty. ,
F 4.4.2.10.2 Uncertainties in Pressure Drops .
Core and vessel pressure drops based on the Best Estimate Flow, as described in Section 3.1, are quoted in Table 4.4-1. The uncertainties quoted are ,
based on the uncertainties in both the test results and the analytical extension of these values to the reactor applicatien. The-cagnitude-ofCt hel-uncertaintics_.
Will- be-cenfirmed-when-the c .g. i.. ental-datarenethe prototype-fuel asser.blA._ -
Bett.ica--triFinobtained . ,
1 A majnr ase of the core and vessel pressure drops is to determine the primary ,
system coolant flow rates as discussed in Section 5.1. In addition, as ..
discussed in Paragraph 4.4.4.1, tests on the primarv system prior to initial r criticality will be made to verify that a conservative primary system coolant flow rate has been used in the design and analyses of the plant.
p 4.4.2.10.3 Uncertainties Due to Inlet Flow Ma1 distribution .
The ef f ects of uncertainties in the inlet flow maldistributica criteria '
used in the core thermal analyses is discussed in Subparagraph 4.4.3.1.2.
4.4.2.10.4 Uncertainty in DNB Correlatica The uncertainty in the DN3 correlation (Paragraph 4.4.2.3) can be written
.as a statement on the probability of not being in DSB based on the statistics -
of the DNB data. This is discussed in Subparagraph 4.4.2.3.2. g 4.4.2.10.5 Uncertainties in DNER Calculations [
The uncertainties in the DNBR's calculated by THINC analysis (seeSubparagraph4.4.3.4.1)f
[~ due to uncertainties in the nuclear peaking factors are accounted for by applying 4.4-23 July 1 1974 -
" 1
- i 6-4-5-041 -
i SNP-23 conservatively high values of the nuclear peaking factors and including measurement error allowances. In addition, conservative values for the -
engineering hot channel factors are used as discussed in Subparagraph 4.4.2.3.4. The results of a sensitivity study (Reference 52) with THINC-IV show that the minimum DNBR in the hot channel is relatively insensitive to variations in the core-wide radial power distribution (for thesamevalueofFjg). ,
The ability of the THIMC-IV computer code to accurately predict flow and L enthalpy distributinns in rod bmidlea is discussed in Subparagraph 4.4.3.4.1 I and in Reference 63. Studies have been performed (Reference 52) to determine the sensitivity of the minimum DNBR in the hot channel to the void fraction [
correlation (see also Subparagraph 4.4.2.8.3); toe inlet velocity and exit pressure distributions assumed as boundary conditions for the analysis; ,
and the grid pressure loss coef ficients. The results of these studies show that the minimum DN3R in the hot channel is relatively insensitive to variations in these parameters. The range of variations considered in these studies ,
covered the range of possible variations in these parameters. ,
4.4.2.10.6 Uncertainties in Flow Rates The uncertainties associated with loop flew rates are discussed in Section s 5.1. For core thermal performance evaluations , a Thermal Desian Loop Flow is used which is less than the Best Estimate Loop Flow (approximately 4% for g,the four-loop plant and 57: for the three-loop plant) . In addition another
(, ,giv5Y of the Thermal Design Flow is assumed to be ineffective for core heat removal capability because it bypasses the core through the various available -
vessel flow paths described in Subparagraph 4.4.3.1.1. E 4.4.2.10.7 Uncertainties in Hydraulic Loads -
I As discussed in Subparagraph 4.4.2.7.2, hydraulic loads on the fuel assembly ,
are evaluated for a pump overspeed transient whidi create flow rates 207 greater than the Mechanical Design Flow. The Mechanical Design Flew as stated in i Section 5.1 is greater than the Best Estimate or most likely flow rate value F for the actual plant operating condition (by approximately 4.5%).
4.4.2.10.8 Uncertainty in Mixing Coefficient The value of the mixing coef ficient, TDC, used in THINC analyses for this application is 0.033. The mean value of TDC obtained in the "R" grid mixing p tests described in Subparagraph 4.4.2.3.1 was 0.042 (for 26 inch grid spacing) .
The value of 0.038 is one standard deviation below the mean value; and s90% of
- the data gives values of TDC greater than 0.038 (Reference 46) . s The results of the mixing tests done on 17 x 17 geometry, as discussed in k Subparagraph 4.4.2.3.3, had a mean value of TDC of 0.059 and standard deviation ,
of a = 0.007. Hence the current desig.n value of TDC is almos t 3 standard #"h deviations below the mean for 26 inch grid spacing.
1645 048 4.4-24 July 1, 1974 -
t
. ~ * * -
e .m---
SNP-29
- 3. Leakage flow from the vessel inlet nozzle directly to the vessel outlet nozzle through the gap between the vessel and the barrel. Q<
%4 Flow entering into the core f rom.the baffle-barrel region through the gaps between the baffle plates, y, Vhe above contributions are evaluated to confirm that the desi,gn value of . core b' ass flow is cet. The design value of core bypass flow fo Sequoyah is equal t 9 f the total vessel flow. Of the total allowance, .2 M is associated 29 with the internals (Items 1, 3, and 4 above) and 2.0% for the core. Calculati ons have been performed using drawing tolerances on a worst case basis and accounting .
for uncertainties in pressure losses. Based on these calculations, the core bypass [
flow for Sequovah is < s.K. This design bypass value is also used in the evaluation of the core pressure drops uoted in Table. 4.4-1, and the determinatica of reactor l flow rates in Section 5.1.
7 5~ % 7 Flow model test results for the flow path through the reactor are discussed in Section 4.4.2.8.2.
4.4.3.1.2 Inlet Flow Distributions s
Data has been considered from several 1/7 scale hydraulic reactor codel tests .
(References 56, 57, and 64) in arriving at the core inlet ficw maldistribution !
criteria to be used in the THINC analyses (See Subparagraph I. 4. 3.4.1) . THINC -
I ana13 3es :ade tuin;; this data ha te indicated that a conservative design basis is to consider a 5 percent reduction in tae flew to the hot as s emb ly . Reference
- 65. The same design basis of 52 reduction to the hot assembly inlet is used in n THINC IV analyses.
The experimental error esticated in the inlet velocity distribution has 1 een considered as outlined in Reference 52 where the sensitivity of changes in inlet velocity distributions to hot channel thermal performance is shown to e be small. Studies (Reference 52) made with the improved THINC model (THINC-IV) -
show that it is adequate to use the 57. reduction ~ in inlet flow to the hot assembly .
for a loop out of service based on the experimental data in References 56 and 57. r The effect of the total flow rate on the inlet velocity distribution was studied -
in the experiments of Reference 56. As was expected, on the basis of the theo- F retical analysis, no significant variation could be found in inlet velocity distribution with reduced flow rate.
No relative ef fects on the core inlet velocity distribution caused by the change f rom a 15 x 15 to 17 .17 fuel assembly array are expected since the icwer internals design will remain unchanged. The flow impedance of the lower core plate and fuel assembly noz:les is equal at all locations. #
4.4.3.1.3 Empirical Friction Factor Correlations ,'
Two empirical f riction factor correlaticas are used in the THINC-IV computer code (described in Subparagraph 4.4.3.4.1) . I P
The f riction f actor in the axial direction, parallel to the fuel rod axis, is eval-usted using the Novendstern-Sandberg correlation (Reference 66). This correlation Q consists of the following:
4.4-26 November 27, 1974 e
1645-049
Added by Amendment 23, July 1, 1974
, TABLE 4.4-1 (Continued)
REACTOR DESIGN COMPARISON TABLE L
Sequoyah Units Reference Plant 1 & 2 17 x 17 17 x 17 With Thermal and Hydraulic Design Parameters With Densification Densification Average in Core, *F fC1
,5S18 585.9
~
Average in Vessel, 'F 578.2 584.7 r
Heat Transfer .
Active Heat Trancier, Surfacc Area, Ft 59,700 59,700 Average Heat Flux, BTU /hr-ft' 139,300 189,300 ,
I Maximum deat Flux, for normal
'74,500 bl l ~
operation BTU /hr-ft '74,500 Average Thermal Cutput, k'i/ f t 5.44 5.44
~ f Maximum Thermal Output, for normal operation kW/f: 13.6E "l 13.6I ^l -
Peak Linear Pcwer for Determination .
of protection setpcints, kJ/f t 13.0[c] 13.0 Fuel Central Temperature !
Peak at 100" Power, 'F 3400 3400 Peak at Thermal Output Maximum for Maximum Overpower Trip Point, 'F 4150 4150 Pressure Drop , z 3 g y, ej g, ,j ./ g , (,
Across Core, psi J 2/crf + dr0' .2K O. %
Across Vessel, including no::le psi ~L 424 6-i_E.4--
._SY.G h.GS y, j 3 g, g -
[a] This limit is associated with the value of FQ = 2.50 I
[b] Based on best estimate reactor flow rate as discussed in Section 5.1.
[c] See Subparagraph 4.3.2.2.6.
4.,-49 1645 050 -
I
N Three reactor design coolant flow rates are identified for the various plant considerations. O The definitions of these flowc are presented in the following paragraphs, and the application of the definitions 3 is illustrated by the system and pu=p hydraulic characteristics on Figure 5.1-11.
Eest Estimate Flow The best estimate flow is the most likely value for the actual plant operating condition.
This flow is based on the best estinate of the reactor vessel, stean generator and piping flow resistance, and on the best estimate of the reactor coolant pu=p head, with no uncer-tainties head. assigned to either the system flow resistance or the pump System pressure losses based on best estimate flow are pre-sented in Table 5.1-1. Although the best estimate flow is the most likely value to be expected in operation, more conservative flow rates are applied in the thermal and =echanical designs.
Thernal Desi;n Flew Thermal ance, design flow is the basis for the reactor core thermal perfor -
the steam generator thermal performance, and the naminal plant parameters used throughout the design.
To provide the required cargin, the thermal design flow accounts for the uncertainties in reactor
~
vessel, steam generator and piping flow resistances. The cc:binc. tion cf these uncertainties, which includes a conservative estimate of the punp discharge weir flow resistance, is equivalent to increasing q the best esticate reactor coolant systen flow resistance by approxi-cately 18 percent. The intersection of this conservative flou resist-ance with the best estimate pump curve, as shown in Figure 5.1-11, establishes the thernal design flow. This p margin for thermal design of approximately !.qocedure provides
.5/ percent. a ficw The ther:al design flow will be confirmed when the plant , s placed in operation.
Tabulations of i=portant design parameters based on the theraal design flow are provided in Table 5.1-1.
Mechanical Design Flow s.51 Mechanical design flow is the conservatively high flow used in the mechanical To assure that design of the reactor vessel internals and fuel assemblies.
a conservatively high flow is specified, the mechanical design best flow is based on a reduced systes resistance (90 percent of (105.5 estinate) percentand on the maxi =un uncertainty on pung head capability of best estimate for machined pump impellers). The intersection of this flow resistance with the higher pump curve , as shown on Figure 5.1-2, establishes the mechanical design flow. The rasulting flow is approxi=ately 4 percent greater than the best estimate flow % 'psJ.'
/ 0 / 70 0 o p m Pu=p overspeed, due to a turbine generatcr overspeed at zu percent, rasults flow.
dasign in a peak reactor coolant flow of 120 percent of the mechanical The overspeed condition is applicable only to operating conditions when the reactor and turbine generator are at power.
O 5.1-4 1645 051
TABLE 5.1-1 SYSTDI DESIGN A!iD OPERATTiG PARA!ETERS Plant Design Life, years 40 tiominal Operating Pressure, psig 2235 .
.~.
Total System Volume, including pressurizer and surge line, ft3 12,612 System Liquid Volume. including pressurizer water at maximum guaranteed power, ft3 11,892 04 tbet~ Power, M We 34 // 5 NSSS Power , .T>tch:r- MW( , Z ,.?
M23 System Thermal and Hydraulic Data Temperatures (Eased on Thermal Design Flow)
Thermal Design Flow, gpm/ loop -4&,-500-9/ 400 Total Reactor Coolant Flow, lb/hr /3g,/ h x 10 Reactor Vessel Inle Temperature, 'F -MM C/[. '7 Reactor Vessel Outlet Tempera *ure. *F 4&~n+ dC4 7 Steam Generatar Out'.e Temperature, 'F l,.:. M 6M.. p Steam Pressure at Full Power, psia 857 Steam Generator Steam Temperature, ? 516 Steam Flow at Full Power, lb/hr (total) / t/,92 '.l M x 10 Feedwater Inlet Temperature, *F <r39 4 3 Y. [
Pressurizer Spray Rate, max., gpm 800 Pressurizer lieat Capacity, kW 18C0 Pressurizer Relief Tank Vclume, ft 1800 Flows and Pressure Drops (Eased on Best Estimate Flow)
Best Estimate Flow, gpa/ loop -Gh 500--- */ 7 80 0 Pump ilead ft. 26o Reactor Vessel AP, psi 46.2 Steam Generator AP, psi 34.6 h Piping AP, psi 6.4 1645 052 5.1-9
. 5.5.7.3.3 Overpressurization Protection
' The inlet line to the Residual Heat Re= oval System is equipped with e
a pressure relief valve sized to relieve the combined flow of all s
the charging pumps at the relief valve set pressure.
Each discharge line to the Reactor Coolant System is equipped with
' a pressure relief valve to relieve the maximum possible back-leakage through the valves separating the Residual Heat Removal Syste= froci the Reactor Coolant System. These relief valves are located in the Frergency Core Cooling System (see Figure 6.3-1).
The design of the Residual Heat ".enoval System in-ludes two isolation valves in series on the inlet line between the high pressure Reactor Coolant System and the lower pressure Residual Heat Removal System.
Each isolation valve is interlocked with one of the two independent Reactor Coolant System pressure signals. The interlocks prevent
@gp the valves from being opened when Reactor Coolant System pressure is greater than approximately 425 psig. If the valves are in the
, \\[ hf
. f /
open position, the interlocks cause the valves to automatically close when the Reactor Coolant System pressure increases to ap,. -
p n ic;-; p sig. These interlocks are described in more detail
,)[,y in 3ubsection 7.6.2.
5.5.7.3.4 Shared Function The safety function performed by the Residual Heat Removal System is not compromised by its normal function which is normal plant cooldown. The valves associated with the Residual Heat Removal d,
Syste are nornally aligned to allow i=ediate use of this system in its safeguard code of operation. The system has been designed in such a canner that two redundant flow circuits are available, assuring the availability of at least ene train for safety purposes.
The nor al plant cooldown function of the Residual Heat Removal System is accomplished through a suction line arrange =ent which is independent of any safeguards f uncticn. The normal cooldown return lines are arranged in parallel redundant circuits and are utilized also as the low head safguards injection lines to the Reactor Coolant System. Utilization of the same return circuits for safeguards as well as for normal cooldown lends assurance to the proper functioning of these lines for cafeguards purposes.
5.5.7.3.5 Radiological Considerations The highest radiation levels experienced by the Residual Heat Removal System are those which would result f rom a loss of coolant accident.
Following a loss of coolant accident, the Residual Heat Removal System is used as part of the Emergency Core Cooling System. During the recirculation phase of emergency core cooling, the Residual Heat Removal System is designed to operate for up to a year pumping water from the containment sump, cooling it, and returning it to the containment to cool the core.
5.5-28 1645 053
5.5.7.3.3 Overpressurization Protection The inlet line to the Residual Heat Removal System is equipped with a pressure relief valve sized to relieve the combined flow of all the charging pumps at the relief valve set pressure.
Each discharge line to the Reactor Coolant System is equipped with a pressure relief valve to relieve the maximum possible back-leakage through the valves separating the Residual Heat Removal System from the Reactor Coolant System. These relief valves are located in the Emergency Core Cooling System (see Figure 6.3-1).
The design of the Residual Heat Removal System includes two isolation valves in series on the inlet line between the high pressure Reactor Ccolant System and the lower pressure Residual Heat Removal System.
Each isolation valve is interlocked with one of the two independent
, Reactor Coolant System pressure signals. The interloc's e prevent h[ y the valves from being opened when Reactor Coolant System pressure y is greater than approximately 425 psig. If the valves are in the
,, f j open position, the interlocks cause the valves to automatically V
))f p) -
/ close when the Reacter Coolant System pressure increases to apprc E-
% eat ef-60 W g. These interlocks are described in more detail in Subsectiot. ,' . 6 . 2 .
5.5.7.3.1 Shared Function The safety function performed bv the Residual Heat Recoval System
~
is not ccepremised by its nor=al function *.hich is ner=al plhnt cooldown. The valves associated with the Resi::ual Heat Renoval Syste= are normally aligned to allow innediate use of this systen in its safeguard code of operation. The system has been designed in such a nanner that two redundant flow circuits are availabla, assuring the availability of at least one train for safety purposes.
The normal plant cooldown function of the Residual Heat Renoval Systen is accomplished through a suc'; ion line arrange =ent which is independent of any safeguards f unction. The normal cooldown recurn lines are arranged in parallel redundant circuits and are utilized also as the low head safguards injection lines to tne Reactor Coolant System. Utilization of the same return circuits for safeguards
, as well as for normal ecoldown lends assurance to the proper functioning of these lines for safeguards purposes.
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5.5.7.3.5 Radiological Considerations The highest radiation levels experienced by the Residual Heat Remeval System are those which would result f rom a loss of coolant accident.
Following a loss of coolant accident, the Residual Heat Removal System is used as part of the Emergency Core Cooling System. During the recirculation phase of emergency core cooling, the Residual Heat Removal System is designed to operate for up to a year pu= ping water from the containment sump, cooling it, and returning it to the containment to cool the core.
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