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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217P2061999-10-26026 October 1999 Forwards for First Energy Nuclear Operating Co Insp Rept 50-346/99-17 on 990928-1001.Insp Was Exam of Activities Conducted Under License Re Implementation of Physical Security Program.No Violations Identified ML20217N3851999-10-20020 October 1999 Forwards RAI Re Licensee 990521 Request for License Amend to Allow Irradiated Fuel to Be Stored in Cask Pit at Davis-Besse,Unit 1.Response Requested within 60 Days from Receipt of Ltr ML20217N2321999-10-15015 October 1999 Requests NRC Approval to Use Alternative to Requirements of 10CFR50.55a(f)(4)(ii).Licensee Requests Extension to Specified Schedule for Implementing Updates to IST Program ML20217G9201999-10-14014 October 1999 Discusses Utils Request for Approval of Quality Assurance Program Changes PY-CEI-NRR-2438, Informs That DBNPS & Pnpp Staffs Have Modified or Withdrawn Several of Positions Proposed within Re Request for Approval of Qap.Revised Positions Encl1999-10-14014 October 1999 Informs That DBNPS & Pnpp Staffs Have Modified or Withdrawn Several of Positions Proposed within Re Request for Approval of Qap.Revised Positions Encl ML20217F8371999-10-0808 October 1999 Forwards Insp Rept 50-346/99-10 on 990802-0913.One Violation Occurred Being Treated as NCV ML20217A5641999-09-30030 September 1999 Informs of Completion of mid-cycle PPR of Davis-Besse on 990901.Informs That NRC Plans to Conduct Addl Insps to Address Questions Raised by Issues Re Operator Errors & Failure to Commit to JOG Topical Rept on MOV Verification ML20212L0691999-09-30030 September 1999 Forwards,For Review & Comment,Copy of Preliminary ASP Analysis of Operational Condition Discovered at Unit 1 on 981014,as Reported in LER 346/98-011 ML20216J6701999-09-24024 September 1999 Forwards Post Examination Documentation for Written Operator Initial License Examination Administered at Davis-Besse Nuclear Power Station on 990920.Without Encls ML20212D3501999-09-21021 September 1999 Forward Copy of Final Accident Sequence Precursor Analysis of Operational Event at Plant,Unit 1 on 980624,reported in LER 346/98-006 05000346/LER-1998-001, Forwards Rev 1 for LER 1998-001,which Updates Corrective Actions & Revises Completion Date Re Implementation of Changes to Plant Emergency Operating Procedure.List of Commitments Attached1999-09-0909 September 1999 Forwards Rev 1 for LER 1998-001,which Updates Corrective Actions & Revises Completion Date Re Implementation of Changes to Plant Emergency Operating Procedure.List of Commitments Attached ML20216E5961999-09-0707 September 1999 Forwards Application for Amend to License NPF-3,revising Tech Specs 3/4.3.2.1,safety Features Actuation Sys Instrumentation & Associated Bases 3/4.3.1 & 3/4.3.2,reactor Protection Sys & Safety Sys Instrumentation ML20211P3001999-09-0707 September 1999 Forwards FEMA Transmitting FEMA Evaluation Rept for 990504 Emergency Preparedness Exercise at Davis-Besse Nuclear Power Plant.No Deficiencies Identified.One Area Requiring C/A & Two Planning Issues Identified ML20211K6681999-08-30030 August 1999 Forwards Copies of Certified Personal Qualification Statement - Licensee (NRC Form 398) for Operator Candidates Listed Below.Without Encls ML20211K6611999-08-30030 August 1999 Forwards Copies of Operator License Renewal Applications for Individuals Listed.Operators Have Successfully Completed Appropriate Operator Requalification Training Program at Dbnps.Without Encls ML20211K0951999-08-30030 August 1999 Forwards Request for Addl Info Re Fire & Seismic Analyses of IPEEE for Davis-Besse Nuclear Power Station,Unit 1. Response Requested within 60 Days ML20211H0201999-08-25025 August 1999 Forwards semi-annual FFD Rept for 990101-0630 for DBNPS, Unit 1,IAW 10CFR26.71(d) ML20211D1171999-08-20020 August 1999 Forwards Insp Rept 50-346/99-09 on 990623-0802.Violations Identified & Being Treated as Noncited Violations ML20211G3911999-08-20020 August 1999 Forwards Update to Estimated Info for Licensing Action Requests Through 010930,re Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates PY-CEI-NRR-2411, Informs That Firstenergy Nuclear Operating Co Has Developed Corporate QA Program Manual for Davis-Besse Nuclear Power Station & Perry Nuclear Power Plant,As Discussed on 990318 Between Util & Nrc.Revised USAR Pages,Encl1999-08-19019 August 1999 Informs That Firstenergy Nuclear Operating Co Has Developed Corporate QA Program Manual for Davis-Besse Nuclear Power Station & Perry Nuclear Power Plant,As Discussed on 990318 Between Util & Nrc.Revised USAR Pages,Encl ML20211J9201999-08-13013 August 1999 Urges NRC to Find Funds for Stockpiling Radiation Pills for Residents Living Near Plant ML20211B0161999-08-13013 August 1999 Forwards SE Accepting Evaluation of Second 10-year Interval Inservice Insp Program Request for Relief Numbers RR-A16, RR-A17 & RR-B9 for Plant,Unit 1 ML20210T1061999-08-12012 August 1999 Forwards Preliminary NRC Forms 398 & 396 for Listed Candidates,Per Operator License Exam Scheduled for Week of 990913.Encl Withheld ML20210S6071999-08-11011 August 1999 Provides Final Response to NRC RAI Re GL 98-01, Y2K Readiness of Computer Systems at Npps ML20210P8051999-08-0909 August 1999 Forwards Insp Rept 50-346/99-15 on 990712-16.No Violations Noted.However,Several Deficiencies Were Identified with Implementation of Remp,Which Collectively Indicated Need for Improved Oversight of Program IR 05000346/19980211999-08-0606 August 1999 Refers to NRC Insp Rept 50-346/98-21 Conducted on 980901- 990513 & Forwards Nov.Two Violations Identified Involving Failure to Maintain Design of Valve & Inadequate C/A for Degraded Condition Cited in Encl NOV 05000346/LER-1998-009, Forwards LER 98-009-01,IAW 10CFR50.73(a)(2)(ii)(B). Commitments Made by Util Are Encl1999-08-0606 August 1999 Forwards LER 98-009-01,IAW 10CFR50.73(a)(2)(ii)(B). Commitments Made by Util Are Encl ML20210H6101999-07-30030 July 1999 Informs That Region III Received Rev 21 to Various Portions of Davis-Besse Nuclear Power Station Emergency Plan.Revision Was Submitted Under Provisions of 10CFR50.54(q) in Apr 1999 ML20210H0491999-07-28028 July 1999 Forwards Application for Amend to License NPF-3,revising TS 3/4.7.5.1, Ultimate Heat Sink, to Allow Plant Operation in Modes 1-4 with Water Temp Less than or Equal to 90 F ML20210G5521999-07-28028 July 1999 Provides Addl Response to 980923 OL Licensing Exam Rept 50-346/98-301 Re OL Exam Administered in Aug 1998.Results of Root Cause Investigation & Corrective Actions,Discussed ML20210G3831999-07-27027 July 1999 Forwards Application for Amend to NPF-3,changing TSs 6.4, Training, 6.5.2.8, Audits, 6.10, Record Retention, 6.14, Process Control Program & 6.15, Odcm ML20211P3071999-07-26026 July 1999 Forwards Final Rept for 990504 Biennial Radiological Emergency Preparedness Exercise for David-Besse Power Station.No Deficiencies Identified for Any Jurisdiction During Exercise ML20210G4391999-07-26026 July 1999 Forwards Application for Amend to License NPF-3,revising Tech Specs 3/4.3.2.1, Safety Features Actuation Sys Instrumentation, & Associated Bases 3/4.3.1 & 3/4.3.2, Reactor Protection Sys & Safety Sys Instrumentation ML20210G7151999-07-26026 July 1999 Forwards Application for Amend to License NPF-3,revising TSs 3/4.3.3.1, Radiation Monitoring Instrumentation, 3/4.3.3.2, Instrument - Incore Detectors & 3/4.3.3.9, Instrumentation - Waste Gas Sys Oxygen Monitor ML20210G5151999-07-26026 July 1999 Forwards Application for Amend to License NPF-3,revising Tech Specs for Implementation of 10CFR50,App J,Option B for Type B & C Containment Leakage Rate Testing ML20210G3211999-07-26026 July 1999 Forwards Written OL Exam & Supporting Matl for Exam to Be Administered at DBNPS During Week of 990913.Listed Encls Withheld from Public Disclosure Until After Exam Complete ML20210C4381999-07-20020 July 1999 Forwards Insp Rept 50-346/99-08 on 990513-0622.Unidentified RCS Leak Approached TS Limit of 1 Gallon Per Minute Prior to Recently Completed Maint Outage.Three Violations of NRC Requirements Identified & Being Treated as NCVs ML20209G3681999-07-15015 July 1999 Advises That Info Submitted in & 990519 Affidavit Re Design & Licensing Rept,Davis-Besse,Unit 1 Cask Pit Rack Installation Project,Holtec Intl, HI-981933,marked Proprietary,Will Be Withheld from Public Disclosure ML20207H6401999-07-0909 July 1999 Discusses Closure of TAC MA0540 Re Util Responses to RAI on GL 92-01,rev 1,suppl 1, Rv Structural Integrity. Staff Has Revised Info in Rvid & Releasing It as Rvid Version 2 ML20209D1341999-07-0808 July 1999 Forwards Notice of Withdrawal of Application for Amend to Operating License.Proposed Change Would Have Modified Facility TSs Pertaining to Allowable as-found Pressure Lift Setting Tolerance of Two Pressurizer Code Safety Valves 05000346/LER-1998-012, Forwards LER 98-012-01,which Is Being Submitted to Provide Addl Info Re 981018 Occurrence.Commitment List Attached1999-07-0707 July 1999 Forwards LER 98-012-01,which Is Being Submitted to Provide Addl Info Re 981018 Occurrence.Commitment List Attached ML20209C3981999-07-0101 July 1999 Responds to NRC Re Violations Noted in Insp Rept 50-346/98-21.Corrective Actions:Developed Rev to Boric Acid Control Program & Work Process Guideline on Plant Leakage ML20209B5821999-06-24024 June 1999 Provides Justification for Rev to Completion Date for One of Insp follow-up Items Cited in Insp Rept 50-346/98-03, Designated as Inspector follow-up Item 50-346/97-201-10 ML20196G1251999-06-23023 June 1999 Responds to NRC RAI Re GL 98-01, Y2K Readiness of Computer Sys at Nuclear Power Plants ML20196E5321999-06-17017 June 1999 Forwards Addl Info Re Relief Request RR-A16 to Support NRC Approval of Relief Request ML20195K2751999-06-16016 June 1999 Forwards Safety Evaluation Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves ML20196A6601999-06-16016 June 1999 Forwards Master Decommissioning Trust Agreements Revised After 1990 for Ohio Edison Co,Cleveland Electric Illuminating Co,Toledo Edison Co & Pennsylvania Power Co Re Bvnps,Units 1 & 2,DBNPS,Unit 1 & Perry Unit 1 ML20195F9071999-06-10010 June 1999 Forwards Application for Amend to NPF-3,changing Tech Specs 3/4.6.4.4, Hydrogen Purge Sys, TS 3/4.6.5.1, Shield Bldg Emergency Ventilation Sys & TS 3/4.7.6.1, Crevs ML20195F8851999-06-0707 June 1999 Withdraws 950929 License Amend Application,Proposing Mod to Allowable as-found Pressure Lift Setting Tolerance of Two Pressurizer Code Safety Valves ML20207G0751999-06-0707 June 1999 Forwards Insp Rept 50-346/99-04 on 990323-0513.Violations Occurred & Being Treated as non-cited Violations,Consistent with App C of Enforcement Policy 1999-09-09
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217N2321999-10-15015 October 1999 Requests NRC Approval to Use Alternative to Requirements of 10CFR50.55a(f)(4)(ii).Licensee Requests Extension to Specified Schedule for Implementing Updates to IST Program PY-CEI-NRR-2438, Informs That DBNPS & Pnpp Staffs Have Modified or Withdrawn Several of Positions Proposed within Re Request for Approval of Qap.Revised Positions Encl1999-10-14014 October 1999 Informs That DBNPS & Pnpp Staffs Have Modified or Withdrawn Several of Positions Proposed within Re Request for Approval of Qap.Revised Positions Encl ML20216J6701999-09-24024 September 1999 Forwards Post Examination Documentation for Written Operator Initial License Examination Administered at Davis-Besse Nuclear Power Station on 990920.Without Encls 05000346/LER-1998-001, Forwards Rev 1 for LER 1998-001,which Updates Corrective Actions & Revises Completion Date Re Implementation of Changes to Plant Emergency Operating Procedure.List of Commitments Attached1999-09-0909 September 1999 Forwards Rev 1 for LER 1998-001,which Updates Corrective Actions & Revises Completion Date Re Implementation of Changes to Plant Emergency Operating Procedure.List of Commitments Attached ML20216E5961999-09-0707 September 1999 Forwards Application for Amend to License NPF-3,revising Tech Specs 3/4.3.2.1,safety Features Actuation Sys Instrumentation & Associated Bases 3/4.3.1 & 3/4.3.2,reactor Protection Sys & Safety Sys Instrumentation ML20211K6611999-08-30030 August 1999 Forwards Copies of Operator License Renewal Applications for Individuals Listed.Operators Have Successfully Completed Appropriate Operator Requalification Training Program at Dbnps.Without Encls ML20211K6681999-08-30030 August 1999 Forwards Copies of Certified Personal Qualification Statement - Licensee (NRC Form 398) for Operator Candidates Listed Below.Without Encls ML20211H0201999-08-25025 August 1999 Forwards semi-annual FFD Rept for 990101-0630 for DBNPS, Unit 1,IAW 10CFR26.71(d) ML20211G3911999-08-20020 August 1999 Forwards Update to Estimated Info for Licensing Action Requests Through 010930,re Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates PY-CEI-NRR-2411, Informs That Firstenergy Nuclear Operating Co Has Developed Corporate QA Program Manual for Davis-Besse Nuclear Power Station & Perry Nuclear Power Plant,As Discussed on 990318 Between Util & Nrc.Revised USAR Pages,Encl1999-08-19019 August 1999 Informs That Firstenergy Nuclear Operating Co Has Developed Corporate QA Program Manual for Davis-Besse Nuclear Power Station & Perry Nuclear Power Plant,As Discussed on 990318 Between Util & Nrc.Revised USAR Pages,Encl ML20211J9201999-08-13013 August 1999 Urges NRC to Find Funds for Stockpiling Radiation Pills for Residents Living Near Plant ML20210T1061999-08-12012 August 1999 Forwards Preliminary NRC Forms 398 & 396 for Listed Candidates,Per Operator License Exam Scheduled for Week of 990913.Encl Withheld ML20210S6071999-08-11011 August 1999 Provides Final Response to NRC RAI Re GL 98-01, Y2K Readiness of Computer Systems at Npps 05000346/LER-1998-009, Forwards LER 98-009-01,IAW 10CFR50.73(a)(2)(ii)(B). Commitments Made by Util Are Encl1999-08-0606 August 1999 Forwards LER 98-009-01,IAW 10CFR50.73(a)(2)(ii)(B). Commitments Made by Util Are Encl ML20210G5521999-07-28028 July 1999 Provides Addl Response to 980923 OL Licensing Exam Rept 50-346/98-301 Re OL Exam Administered in Aug 1998.Results of Root Cause Investigation & Corrective Actions,Discussed ML20210H0491999-07-28028 July 1999 Forwards Application for Amend to License NPF-3,revising TS 3/4.7.5.1, Ultimate Heat Sink, to Allow Plant Operation in Modes 1-4 with Water Temp Less than or Equal to 90 F ML20210G3831999-07-27027 July 1999 Forwards Application for Amend to NPF-3,changing TSs 6.4, Training, 6.5.2.8, Audits, 6.10, Record Retention, 6.14, Process Control Program & 6.15, Odcm ML20210G4391999-07-26026 July 1999 Forwards Application for Amend to License NPF-3,revising Tech Specs 3/4.3.2.1, Safety Features Actuation Sys Instrumentation, & Associated Bases 3/4.3.1 & 3/4.3.2, Reactor Protection Sys & Safety Sys Instrumentation ML20210G7151999-07-26026 July 1999 Forwards Application for Amend to License NPF-3,revising TSs 3/4.3.3.1, Radiation Monitoring Instrumentation, 3/4.3.3.2, Instrument - Incore Detectors & 3/4.3.3.9, Instrumentation - Waste Gas Sys Oxygen Monitor ML20210G5151999-07-26026 July 1999 Forwards Application for Amend to License NPF-3,revising Tech Specs for Implementation of 10CFR50,App J,Option B for Type B & C Containment Leakage Rate Testing ML20211P3071999-07-26026 July 1999 Forwards Final Rept for 990504 Biennial Radiological Emergency Preparedness Exercise for David-Besse Power Station.No Deficiencies Identified for Any Jurisdiction During Exercise ML20210G3211999-07-26026 July 1999 Forwards Written OL Exam & Supporting Matl for Exam to Be Administered at DBNPS During Week of 990913.Listed Encls Withheld from Public Disclosure Until After Exam Complete 05000346/LER-1998-012, Forwards LER 98-012-01,which Is Being Submitted to Provide Addl Info Re 981018 Occurrence.Commitment List Attached1999-07-0707 July 1999 Forwards LER 98-012-01,which Is Being Submitted to Provide Addl Info Re 981018 Occurrence.Commitment List Attached ML20209C3981999-07-0101 July 1999 Responds to NRC Re Violations Noted in Insp Rept 50-346/98-21.Corrective Actions:Developed Rev to Boric Acid Control Program & Work Process Guideline on Plant Leakage ML20209B5821999-06-24024 June 1999 Provides Justification for Rev to Completion Date for One of Insp follow-up Items Cited in Insp Rept 50-346/98-03, Designated as Inspector follow-up Item 50-346/97-201-10 ML20196G1251999-06-23023 June 1999 Responds to NRC RAI Re GL 98-01, Y2K Readiness of Computer Sys at Nuclear Power Plants ML20196E5321999-06-17017 June 1999 Forwards Addl Info Re Relief Request RR-A16 to Support NRC Approval of Relief Request ML20196A6601999-06-16016 June 1999 Forwards Master Decommissioning Trust Agreements Revised After 1990 for Ohio Edison Co,Cleveland Electric Illuminating Co,Toledo Edison Co & Pennsylvania Power Co Re Bvnps,Units 1 & 2,DBNPS,Unit 1 & Perry Unit 1 ML20195F9071999-06-10010 June 1999 Forwards Application for Amend to NPF-3,changing Tech Specs 3/4.6.4.4, Hydrogen Purge Sys, TS 3/4.6.5.1, Shield Bldg Emergency Ventilation Sys & TS 3/4.7.6.1, Crevs ML20195F8851999-06-0707 June 1999 Withdraws 950929 License Amend Application,Proposing Mod to Allowable as-found Pressure Lift Setting Tolerance of Two Pressurizer Code Safety Valves ML20207F4231999-06-0202 June 1999 Forwards Rev 1 to DBNPS Emergency Preparedness Evaluated Exercise Manual 990504, IAW 10CFR50.4.NRC Evaluated Exercise Has Been Rescheduled for 991208,since NRC Did Not Evaluate 990504 Exercise ML20207E9561999-05-28028 May 1999 Forwards Update to NRC AL 98-03,re Estimated Info for Licensing Activities Through Sept 30,2000 ML20207E2521999-05-28028 May 1999 Forwards Rev 18,App A,Change 1 to Davis-Besse Nuclear Power Station,Unit 1,industrial Security Plan IAW Provisions of 10CFR50.54(p).Encl Withheld,Per 10CFR73.21 ML20207E7801999-05-21021 May 1999 Forwards Application for Amend to License NPF-3,allowing Use of Expanded Spent Fuel Storage Capacity.Proprietary & non- Proprietary Version of Rev 2 to HI-981933 Re Cask Pit Rack Installation Project,Encl.Proprietary Info Withheld ML20206N0231999-05-0606 May 1999 Forwards License Renewal Applications for Davis-Besse Nuclear Power Station,Unit 1 for ML Klein,Cn Steenbergen & CS Strumsky.Without Encls ML20206D2421999-04-28028 April 1999 Forwards Combined Annual Radiological Environ Operating Rept & Radiological Effluent Release Rept for 1998. Rev 11, Change 1 to ODCM & 1998 Radiological Environ Monitoring Program Sample Analysis Results Also Encl PY-CEI-NRR-2382, Forwards 1998 Annual Rept for Firstenergy Corp, for Perry Nuclear Power Plant & Davis-Besse Nuclear Power Station.Form 10-K Annual Rept to Us Securities & Exchange Commission for FY98 Also Encl1999-04-21021 April 1999 Forwards 1998 Annual Rept for Firstenergy Corp, for Perry Nuclear Power Plant & Davis-Besse Nuclear Power Station.Form 10-K Annual Rept to Us Securities & Exchange Commission for FY98 Also Encl ML20206B8831999-04-17017 April 1999 Forwards 1634 Repts Re Results of Monitoring Provided to Individuals at Davis-Besse Nuclear Power Station During 1998,per 10CFR20.2206.Without Repts ML20205K5641999-04-0707 April 1999 Forwards Response to NRC 980415 RAI Re GL 96-06, Assurance of Equipment Operability & Ci During Design-Basis Accident Conditions. Rept FAI/98-126, Waterhammer Phenomena in Containment Air Cooler Swss, Encl ML20205K3871999-04-0707 April 1999 Forwards Copy of Application of Ceic,Oec,Ppc & Teco to FERC, Proposing to Transfer Jurisdictional Transmission Facilities of Firstenergy Operating Companies to American Transmission Sys,Inc.With One Oversize Drawing ML20205J1171999-03-29029 March 1999 Forwards Rev 1 to BAW-2325, Response to RAI Re RPV Integrity, Per GL 92-01,Rev 1,Suppl 1, Reactor Vessel Structural Integrity. Rev Includes Corrected Values in Calculations PY-CEI-NRR-2377, Submits Decommissioning Repts for Bvps,Units 1 & 2,Davis- Besse Nuclear Power Station,Unit 1 & Perry Nuclear Power Plant,Unit 1,per 10CFR50.75(f)(1)1999-03-29029 March 1999 Submits Decommissioning Repts for Bvps,Units 1 & 2,Davis- Besse Nuclear Power Station,Unit 1 & Perry Nuclear Power Plant,Unit 1,per 10CFR50.75(f)(1) ML20205F5961999-03-27027 March 1999 Forwards Comments on Preliminary Accident Sequence Percursor (ASP) Analysis of 980624 Operational Event at Dbnps,Unit 1, as Transmitted by NRC Ltr ML20205D4791999-03-26026 March 1999 Forwards Rept Submitting Results of SG Tube ISI Conducted in Apr 1998.Rept Includes Description of Number & Extent of Tubes Inspected,Location & Percent wall-thickness Penetration for Each Indication of Imperfection ML20205L2031999-03-26026 March 1999 Submits Correction to Dose History of Tj Chambers.Dose Records During 1980-1997 Were Incorrectly Recorded Using Wrong Social Security Number.Nrc Form 5 Not Encl ML20205C7371999-03-25025 March 1999 Certifies That Dbnps,Unit 1,plant-referenced Simulator Continues to Meet Requirements of 10CFR55.45(b) for Simulator Facility Consisting Solely of plant-referenced Simulator.Acceptance Test Program & Test Schedule,Encl ML20205E3551999-03-19019 March 1999 Requests That Proposed Changes to TS 6.8.4.d.2 & TS 6.8.4.d.7 Be Withdrawn from LAR Previously Submitted to NRC ML20204J6361999-03-17017 March 1999 Forwards Firstenergy Corp Annual Rept for 1998 & 1999 Internal Cash Flow Projection as Evidence of Util Guarantee of Retrospective Premiums Which May Be Served Against Facilities PY-CEI-NRR-2375, Forwards List That Details Util Insurers,Policy Numbers & Coverage Limits for Two Power Plants,Per Requirements of 10CFR50.54(w)(3) Re Reporting of Property Insurance Coverage1999-03-15015 March 1999 Forwards List That Details Util Insurers,Policy Numbers & Coverage Limits for Two Power Plants,Per Requirements of 10CFR50.54(w)(3) Re Reporting of Property Insurance Coverage ML20204E6821999-03-12012 March 1999 Requests That Listed Changes Be Made to NRC Document Svc List for Davis-Besse Nuclear Power Station,Unit 1 1999-09-09
[Table view] Category:UTILITY TO NRC
MONTHYEARML20065D0491990-09-14014 September 1990 Forwards Operator & Senior Operator Licensing Exam Ref Matl for Exam Scheduled for Wk of 901112,per 900607 Request ML20065D4951990-09-14014 September 1990 Forwards Updated Exam Schedule for Facility,In Response to Generic Ltr 90-07, Operator Licensing Natl Exam Schedule ML20059K4681990-09-14014 September 1990 Provides Supplemental Info Re Emergency Response Data Sys (Erds).Data Transmitted by Util ERDS Will Have Quality Tag of 4 & Point Identification for ERDS Renamed ML20059G2341990-09-10010 September 1990 Provides Response to Request for Addl Info Re Interpretation of Tech Spec 3/4.7.10, Fire Barriers. Interpretation Is Implemented & Unnecessary Compensatory Measures Removed.List of Fire Barriers Inspected on One Side Only Encl ML20059G4961990-09-0606 September 1990 Submits Voluntary Rept of Svc Water HX Testing During Sixth Refueling Outage.Expected Flow Rates Not Achieved.Periodic Tests Developed to Check Efficiency of Containment Air Coolers ML20064A6271990-09-0606 September 1990 Requests That Requirement Date for Installation & Testing of Alternate Ac Power Source & Compliance w/10CFR50.63 Be Deferred Until Completion of Eighth Refueling Outage ML20028G8611990-08-28028 August 1990 Forwards Davis Besse Nuclear Power Station Semiannual Rept: Effluent & Waste Disposal,Jan-June 1990. ML20059D4121990-08-28028 August 1990 Forwards Second 10-Yr Interval Pump & Valve Inservice Testing Program ML20059D5521990-08-24024 August 1990 Forwards Semiannual Fitness for Duty Rept for Jan-June 1990 ML20059B5291990-08-23023 August 1990 Forwards Updated Fracture Mechanics Analysis of Hpi/Makeup Nozzle,Per 900510 Meeting W/Nrc.Util Believes That Addl Analysis to Assess Structural Integrity of Nozzle Using More Conservative Fracture Model Supports Previous Analysis ML20058Q3911990-08-16016 August 1990 Requests NRC Concurrence on Encl Interpretation & Technical Justification of Tech Spec 3/4.7.10, Fire Barriers ML20058P7801990-08-10010 August 1990 Advises of Intentions to Revise Testing Requirements for Fire Protection Portable Detection Sys at Plant & Functional Testing of auto-dialer & Telephone Line Subsys from Daily to Weekly Testing ML20063P9981990-08-0909 August 1990 Submits Supplemental Response to Insp Rept 50-346/89-21. Util Rescinds Denial & Accepts Alleged Violation ML20056A5341990-08-0303 August 1990 Confirms Electronic Transfer of Payment of Invoice I0942 Covering Annual Fee for FY90,per 10CFR171 ML20058M7791990-08-0303 August 1990 Forwards Rev 10 to Industrial Security Plan & Rev 6 to Security Training & Qualification Plan.Revs Withheld ML20058L1821990-08-0101 August 1990 Forwards Davis-Besse Dcrdr Human Engineering Discrepancy Repts 1988 Summary Addendum 1,Vol 1, Per NRC Audit Team Request ML20056A8341990-07-23023 July 1990 Forwards Revised Monthly Operating Rept for June 1990 for Davis-Besse Nuclear Power Station Unit 1 ML20055H4601990-07-20020 July 1990 Discusses Resolution of Draft SER Open Item on Voluntary Loss of Offsite Power.Util Preparing License Amend Request Per Generic Ltrs 86-10 & 88-12 to Relocate Fire Protection Tech Specs & Update Fire Protection License Condition ML20055F9681990-07-17017 July 1990 Forwards Application for Amend to License NPF-3,adding Centerior Svc Company as Licensee in Facility Ol.Change Allows for Improved Mgt Oversight,Control & Uniformity of Nuclear Operations ML20055F8561990-07-17017 July 1990 Discusses Util Planned Activities Re Instrumented Insp Technique Testing Performed at Facility in View of to Hafa Intl.Relief Requests Being Prepared by Util for Sys on Conventional Hydrostatic Testing ML20044B3001990-07-12012 July 1990 Provides Written Confirmation of Util Electronic Transfer of Funds to NRC on 900711 in Payment of Invoice Number I1050 ML20044B1841990-07-10010 July 1990 Requests Approval of Temporary non-code Repair & Augmented Insp of Svc Water Piping,Per 900626 Telcon ML20055D9701990-06-29029 June 1990 Provides Written Confirmation of Util Electronic Transfer of Funds for Payment of Invoice 0111 Covering Insp Fees for 890326-0617 ML20043H5291990-06-14014 June 1990 Forwards Plans Re Reorganization & Combining of Engineering Assurance & Svc Program Sections ML20055C7521990-06-14014 June 1990 Responds to NRC Bulletin 89-002, Potential Stress Corrosion Cracking of Internal Preloaded Bolting in Swing Check Valves & Justification for Alternate Insp Schedule for One Valve. No Anchor Darling Swing Check Valves Installed at Plant ML20055F2261990-06-14014 June 1990 Forwards 1990 Evaluated Emergency Exercise Objectives for Exercise Scheduled for 900919 ML20043G5661990-06-14014 June 1990 Forwards Rev 9 to Industrial Security Plan.Rev Withheld (Ref 10CFR73.21) ML20043G7811990-06-12012 June 1990 Forwards Info Re Implementation of NUREG-0737,Item II.F.2, Inadequate Core Cooling Instrumentation, Per NRC 900214 Safety Evaluation.Item II.B.1 Issue Re Reactor Vessel Head Vent Also Considered to Be Closed ML20043F6091990-06-11011 June 1990 Forwards Util Comments on NRC Insp Rept 50-346/90-12, Per 900601 Enforcement Conference Re Core Support Assembly Movement & Refueling Canal Draindown.Refueling Canal Draindown Procedure Provides Specific Draining Instructions ML20043E1301990-06-0101 June 1990 Withdraws 870831 & 890613 Applications to Amend License NPF-3.Changes Requested Addressed by Issuance of Amend 147 or Can Now Be Made as Change to Updated SAR Under 10CFR50.59 ML20043D5601990-05-31031 May 1990 Forwards Application for Amend to License NPF-3,revising Tech Spec 3/4.6.4.1, Combustible Gas Control - Hydrogen Analyzers. Request Consistent W/Nrc Guidance,Generic Ltr 83-37,dtd 831101,NUREG-0737 Tech Specs & Item II.F.1.6 ML20043D5691990-05-31031 May 1990 Forwards Application for Amend to License NPF-3,requesting Extension of Expiration Date of Section 2.H to Allow Plant Operation to Continue Approx 6 Yrs Beyond Current Expiration Date ML20043D1451990-05-31031 May 1990 Forwards Rev 11 to Updated SAR for Unit 1.Rev Updates Table 6.2-23 Re Containment Vessel Isolation Valve Arrangements ML20043D1621990-05-29029 May 1990 Documents Util Understanding of NRC Interpretation of Plant Tech Spec 3.7.9.1,Action b.2 Re Fire Suppression Water Sys, Per 891206 Telcon.Nrc Considered Electric Fire Pump Operable Provided Operator Stationed to Open Closed Discharge Valve ML20043C2331990-05-25025 May 1990 Forwards Summary of 900510 Meeting W/B&W & NRC in Rockville, MD Re Hpi/Makeup Nozzle & Thermal Sleeve Program.List of Attendees & Meeting Handout Encl ML20043B1701990-05-18018 May 1990 Forwards Revised Exemption Request from 10CFR50,Section III.G.2,App R for Fire Areas a & B,Adding Description of Specific Limited Combustibles That Exist Between Redundant Safe Shutdown Components in Fire Area a ML20043A5441990-05-16016 May 1990 Discusses Status of Safety & Performance Improvement Program Portion of B&W Owners Group EOP Review Project ML20043A2311990-05-11011 May 1990 Responds to Violation Noted in Insp Rept 50-346/90-08. Corrective Actions:Results of Analysis of Radiological Environ Samples & Radiation Measurements Included in 1989 Annual Radiological Environ Operating Rept ML20043A4901990-05-10010 May 1990 Forwards Summary of Differences Between Rev 5 to Compliance Assessment Rept & Rev 1 to Fire Area Optimization,Fire Hazards Safe Shutdown Evaluation, Vols 1-3.Rept Demonstrates Compliance W/Kaowool Wrap Removal ML20042F9801990-05-0404 May 1990 Provides Written Confirmation of Util Electronic Transfer of Payment of Invoice Number 10716 to Cover Third Quarterly Installment of Annual Fee for FY90 ML20042F5781990-05-0303 May 1990 Provides Status of Hpi/Makeup Nozzle & Thermal Sleeve Program.Nrc Approval Requested for Operation of Cycle 7 & Beyond Based on Program Results.Visual Insp of Thermal Sleeve Identified No Thermal Fatique Indications ML20042F0951990-04-30030 April 1990 Responds to Violations Noted in Insp Rept 50-346/90-02. Corrective Actions:Maint Technician Involved in Tagging Violation Counseled on Importance of Procedure Adherence W/ Regard to Personnel Safety ML20042F0841990-04-27027 April 1990 Responds to Violations Noted in Insp Rept 50-346/89-201 for Interfacing Sys LOCA Audit on 891030-1130.Corrective Actions:Plant Startup Procedure Will Be Revised Prior to Restart from Sixth Refueling Outage ML20042E7311990-04-27027 April 1990 Forwards Application for Amend to License NPF-3,deleting 800305 Order Requiring Implementation of Specific Training Requirements Which Have Since Been Superseded by INPO Accredited Training Program ML20042F1961990-04-27027 April 1990 Informs of Adoption of Reorganization Plan Re Plants on 900424.Reorganization Will Make No Changes in Technical or Financial Qualifications for Plants.Application for Amends to Licenses Adding Company as Licensee Will Be Submitted ML20043F7261990-04-20020 April 1990 Requests Exemption from 10CFR55.59(a)(2) to Permit one-time Extension of 6 Months for Reactor Operators & Senior Reactor Operators to Take NRC 1990 Requalification Exam. Operators Will Continue to Attend Training Courses ML20042E7091990-04-17017 April 1990 Forwards Annual Environ Operating Rept 1989 & Table 1 Providing Listing of Specific Requirements,Per Tech Spec 6.9.1.10 ML20012F5091990-04-0303 April 1990 Forwards Completed NRC Regulatory Impact Survey Questionnaire Sheets,Per Generic Ltr 90-01 ML20012F6001990-04-0202 April 1990 Submits Supplemental Response to Station Blackout Issues,Per NUMARC 900104 Request.Util Revises Schedule for Compliance W/Station Blackout Rule (10CFR50.63) to within 2 Yrs of SER Issuance Date ML20012E0181990-03-22022 March 1990 Forwards Application for Amend to License NPF-3,changing License Condition 2.C(4) Re Fire Protection Mods to Fire Extinguishers,Fire Doors,Fire Barriers,Fire Proofing,Fire Detection/Suppression & Emergency Lighting 1990-09-06
[Table view] |
Text
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TOLEDO EDISON Docket No. 50-346 dce License No. NPF-3 m ai g39 3 2, Serial No. 569 December 28, 1979 Director of Nuclear Reactor Regulation Attention: Mr. Robert N. Reid, Chief Operating Reactors Branch No. 4 Division of Operating Reactors
- United States Nuclear Regulatory Commission Washington, D. C. 20555
Dear Mr. Reid:
This transmits additional responses to your letter of August 21, 1979 for Davis-Besse Nuclear Power Station, Unit 1, as defined in the joint NRC/B&W Owners Group meeting of September 13, 1979. Specifically the following items are addressed:
Attachment A - Noncondensible Cas Discussion - in response to item 3 of Attachment A to Enclosure 1 of your letter.
Attachment B - Small Break and PORV Stuck Open Discussion -
in response to item 2B of Attachment A to Enclosure 1 of your letter.
It is noted that these have been reviewed to be specifically applicable to the Davis-Besse Unit I design.
"ary truly yours, ffhrzu RPC:TJM: cts cc: Mr. Robert Capra Bulletins & Orders Task Force United States Nuclear Regulatory Cocmission Washington, D. C. 20555 kDYl 1685 031 essa
//
THE TCLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO, OHIO 43852 8 0 010 40 d3,
Docket tw . w-wu l
Am o -
License No. NPF-3
- w o . 3 . December 28, 1979 Attachment A.
Question 3 - Honcondensible Gases Regarding the presence of noncondensible gases within the reactor coolant system following a small break LOCA:
A. Provide the sources of noncondensible gases in the primary system.
B. Discuss the effect of noncondensible gases on:
(1) condensation heat transfer, (2) system pressure calculations and (3) natural circulation flow.
C. Describe any operator actions and/or emergency procedures
, necessary to preclude introduction'of sinnificant quantities '
of noncondensible gases into the primary 'ystem. s D. Describe' operator actions to be taken in the event of a significant accumulation of noncondensible gas;s in the primary system.
Resoonse A. Sources of Noncondensible Gases in the Primary System Table 1 lists the potential sources and amounts of noncondensible gases for a 177 fuel assembly plant. However, most of these gases would not be rele.ased for small break transients. Appendix K evaluations performed for the 177FA plants demonstrate that cladding temperatures remain low cnd no cladding rupture nor metal water reaction occur. Thus, these sources can be neglected. The core flooding tasks d only for breaks large enough to depressurize the RCSjscharce into Also, the the RCS steam generator (SG) is a heat sink only if primary system pressure is above that which corresponds to the sec,ondary system safety valve setpoint (% 1050 psia).
Therefore, gases present in the core flooding tank can be neglected in .
addressing the effect of noncondensibles. The only sources of noncondon-sibles which might separate in the RCS are the gases dissolved in the coolant, the gases in the pressurizer, gases in the makeup and borated water storage tank and gases released from an allowed 1% failed fuel in the core.
B. Effects of Noncondensible Gases on the Primary System Response followino a Small Break LOCA .
There are two possible ways in which the release of noncondensible gases in the primary system could interfere with the condensation heat transfer processes which occur in the steam generator during s, mall loss of coolant accidents. If noncondensible gases filled the U bend at the top of the hot 109, then water vapor would have to diffuse through the noncondensible gases before they could be condensed in the steam generator. This would be a very slow proces's and would effectively inhibi t natural circulation. Lesser amaunts of noncondensibles would reduce the heat transfer by condensation because the vapor would have to diffuse thitugh the noncondensibles to get to the condensate on the tubes.
1685 032
As discussed in response to Part A of this question, the
'only sources of noncondensibles which might separate in thr ?.CS are the gases dissolved in the coolant, the gasc.; in the pressurizer, gases in the makeup and borated water storage tank and gases released from an allowed 1% failed fuel in the core. Thus, the maximum amount of noncondensible gases in the system, assuming all gas comes out of solution, no noncondensibles are lost through the break flow, that there was one percent failed fuel, and the injection of 6.4 x 10 4 lbm from the makeup tank and BWST (typical of % 1500 sec of HPI),
would be:
Dissolved in coolant 563 scf In pressurizer 166 Fission gas 2 -
Fuel rod fill gas 11 MU tank 24 BWST 14 Total ,
780 scf 3
This gas would occupy a volume of 22.4 ft at a pressure of 1050 psia, the lowest pressure condition in the primary system for which condensation heat removal will occur. It should be noted that the assumed integrated injection flow does not have a significant effect on the total volume of noncondensibles which might be present in the primary system. Since the volume required to complMely fill the U-bend in the hot leg is 125 ft , the 3 noncondensible gases will not impede the flow of vapor to the steam generator.
The heat transfer during condensation is made up oV'the sensible heat trans-ferred through the diffusion layer and the latent heat released due to conden-sation of thq vapor reaching the interface (see Figure 1). The model of Colburn and Hougenlli gives the following equation for the heat transfer in the vapor phase:
4 = hg( gg - Tg j) + Kg Mg hfg (Pgg - Pgg ) (1) where 2 4 = condensation heat flux, btu /hr~ ft ~
hg = heat transfer coefficient for vapor layer, Btu /hr ft2 of Tgg = bulk temperature, OF Tgg = temperature of interface, O F I
- Kg = mass transfer coefficient, Mg = molecular weight, ibm /lb mole hfg = latent heat of vaporization, Btu /lbm lb Pgo = partial pressure of vapor at bulk conditions, Pgj = partial pressure of vapor at the interface, lb f D ft2 1685 033 Q
r <.373 Kg = 1.02 0 9* P - I) p/pam z RT s pD ,
2 D = diffusion coefficient, f t /hr z = height Ib ft f
R = gas constant,1545 lb mole O R T = absol'ute temperature at bulk conditions OR g = acceleration of gravity, ft/hr 2 p = density, lbm/f t 3 po = density at bulk conditions, lbm/ft 3 pi = density at interface cpnditions, lbm/ft3 ,
u = viscosity, lbm/hr ft pam = pai - gao in P#I pao pai = partial pressure of gas at interface, h pao = partial pressure of gas at bulk conditions, h
~
For the application to OTSG condensing heat transfer during small break tran-sients, the term hg(Tgo - Tgi) can conservatively be neglected since the vapor velocities would be very low. Thus,
& = Kg Mg hfg(Pgo - Pgi). (2) ,
The heat transfer with noncondensible gases present is obtained by iteration.
An interface temperature Tgj is assumed, which fixes Pgt, and the heat transfer across the liquid condensate film is computed from
& = hf (Tgj - Tg) (3)
- where 3', 1/4 P(Pf-P)9hfg f g k f hf = .943 Uf z(Igj-T)g .
of = density of fluid, lbm/ft3 pg = density of vapor, ibm /ft 3 kf = thermal conductivity of fluid, Btu /hr ft OF Tg = wall temperature, 'F A
'1685 034
10f r D 'P D* n "
O k .
lhe partial pressure of the gas at the bulk conditions can be calculated from the nole fraction of noncondensible gases. When the heat flux con.puted from anuatinn 2 matches that computed by equation 3, the proper interface temperature i as been found.
The impact of noncondensibles on the condensation heat transfer process during l a small break was examined for the 0.04 ft2 and 0.01 ft2 cold leg breaks analy-
- zed for the 177-FA plants. TM breaks utilize the SG for heat removal for a significant portion of the transient. Hand calculations were performed, using l the theory presented above, to ascertain the effect of non-condensibles on the ,
transient. l t
The amunt of noncondensible gases, assuning that all gases come out of solu- i tien, would be 2.61 moles. The effect of these gases is to raise the pressure '
and primary temperature to obtain the same heat transfer. Assuming that the i noncondensibles accumulated only within the steam generator upper plenums and i the steam generator tubes, the system pressure increase, due to ncncondensibles, ;
vould only be 25 osi, for a 0.04 ft2 break, and 40 psi, for a 0.01 ft2 break. .
It shculd be noted that this effect is predominantly due to the inclusion of *!
the partial pressure of the noncondensibles, which is 24 psi for the 0.04 ft2 l hreak and 34 psi for the 0.01 ft2 break, in the total system pressure. These ,
calculations represent the maximum impact as they were computed at the time -
of c.:ximum condensation heat flux for the respective cases. ;
- s shm n, the influence of noncondensibles does not significantly effect the condensation heat transfer process. The estinates made are conservative in '
tnat they assumed all the gas is located in the steam generators (none is in the too of the reactor vessel or Dressurizer) and no cases escape through the t.r ea k . Thus, it is shown that the presence of noncondensible gases in the system considering the effect on condensation heat transfer, system pressure and natural circulation should not significantly affect the small break transient.
htions to Precl'ude Introduction of Noncondensible Gases into the P.ir.ary System introduction of significant quantities of noncondensible gases into the primary system following a small break LOCA is prevented if the core is not uncovered during a small break. The sma il break quidelines which have been developed by B&W, are designed to ,
prevent core uncovery by assuring continued ECC injection. Thus, t.he arount of loncondensibles which might separate in the RCS is small and would not significantly effect the small break transient (See Part B above).
D. Operator Actions During Accumulation of Noncondensible Gases in
. t,te Primary Systen A significant accumulation ^of rronromtens-tble gases within the primary system during a small break is not expected. This position is confirmed by small break transient predictions, using conservative Appendix K assumptions, which show that little core uncovery occurs.
(2,3,4) As a result of the small core uncovery fuel clad temperature excursions are limited to 1100f; .and, fuel clad failure or H2 gas format. ion due to metal water reaction' will not occur.
1685 035
Small amounts of noncondensible gases can be released into the primary system during a small break. For the break size range where noncondensible gases could have a detrimental effect (i.e., breaks where natural circulation is required for energy removal) the quantities of gases that are predicted to exist within the primary system are not significant. For larger quantities of noncondensible gases to exist, a core transient that is not predicted must occur. The probability for such an occurrence is believed to be small because of the detailed emergency procedures for post-LOCA conditions that have been developed and the extensive operator training that has been conducted in their use.
Emergency procedures have been developed to accommodate nonconden-sible gases, to maintain plant control, and to achieve a stable .
long term cool _ing condition. Provided below is a brief summary of plant control measures contained in present emergency procedures which will counteract the effects of noncondensible gases and additional guidance for operator action developed for an inadequate core cooling condition, which will be incorporated into emergency procedures in the near future. To upgrade the RCS venting and/or degassing capabilities, remote operated hot leg vents will be designed and installed by 1981. Small break-emergency procedures will also be revised a.t,that time to include use of the hot leg high point vents to aid the re-establish-ment of natural circulation and to vont noncondensible gases which '
may evolve during small break transient.
CURRENT PROCEDURAL ACTIONS During a small break, the principle effect of noncondensible gases is to minimize the performance of the steam generators during natural circulation (either single phase water flow- or reflux boiling). Table 2 lists the primary symptoms and the corresponding operator actions identified in current emergency procedures. As indicated in Table 2, a restart of the RC pumps (one per loop) is -
the optimum action. A return to forced circulation will aid in condensation of existing steam and removal of noncondensible cas (if present) within the hot leg piping. Noncondensible gases, originally within the loop pipinq, would then tend to be suspended within the coolant stream and collect within the upper regions of the reactor vessel (RV). A substantial quantity (s 1000 ft.3) of gas can be accommodated within the upper region of the RV; therefore, there is good assurance that natural circulation can be maintained if RC pump operation must be terminated. If the RC pumps cannot be started and/or no secondary side heat sink is available, the operator will utilize the PORV, HPI, Makeup Pumps and Startup feedwater pump for core cooling and RC pressure control until the RC pumps can be restarted and/or normal secondary cooling is re-established.
1685 036
D""D,*DlT os # J S N The above actions are suffic.ient to enable the operator to bring the unit to a stable, long term cooling condition based on expected plant performance using Appendix K evaluation methods. Although a large accumulation of noncondensible gases is not expected under these assumptions, the above actions are believed to be sufficient if the anticipated amounts of non-condensible gases are increased by an order of magnitude because of the larce volume available for gases in the upper head of the RV and the loss of noncondensible gases out the break.
Once stable long term cooling conditions are established, RCS venting and/or degassing procedures can be initiated. If the RC pump (s) are operative and pressurizer spray is available, the reactor coolant can be degassed within the pressurizer where the stean-gas space can be vented to the Quench Tank inside containment. If letdown is available, the reactor coolant can also be degassed utilizing the makeup tank. The reduction of the .
cmount of gases dissolved in the RC will encourage remaining gas pockets within the RCS to redissolve in the water. The operator can monitor the progress of dogassing activities via analysis of pressurizer fluid and/or letdown water samples.
SMALL JREAK - INADEQUATE CORE COOLING CONDITIONS An inadequate core cooling condition is not expected for B&W g 177 FA plants. However, guidelines which identify the symptoms and operator actions for several circumstances, including a small break, have been prepared by B&H. This information is discussed in detail in Reference 3.
The operator actions discussed in Reference 5 are aimed at restoration of core cooling (restart an RC pump) followed by an increased rate of plant cooldown and depressurization (via SG cooling and PORV operation) to acquire use of the high volumetric flow capability of the CFT and LPI system to maintain core cooling.
From a noncondensible gas standpoint, the actions accomplished the following: .,
- 1. Prevention: By initiating corrective action when cladding temperatures are below those for which metal water reaction is significant, gas accumulat4on is minimized. RC pump operation (if possible) to restore core cooling and to increase the plants cooldown/depressurization capability is the preferred action.
- 2. Venting: For the core to be inadequately cooled, the RCS must be in a highly void condition. Therefore, PORV operation in combination with the break should provide a vent mechanism for the noncondensible gas that do exist.
As discussed in the previeus section, normal venting and degassing procedures can also be undertaken once s. table long term cooling is established.
b i685 037
REFERENCES
- 1. Colburn, A. P. and flougen, D. A. , " Design of Cooler Condensers for liixtures of Vapors with Noncondensing Gases", Ind. Ena.
Chem. 26(11), 1934
- 2. P. C. Jones, J. R. Biller, and B. M. Dunn, "ECCS Analysis of B&W's 177 FA Lowered Loop NSS", BAW-10103, Rev. 3, Babcock & Wilcox, J u l y , 19"
- 3. Letter from s. H. Tayl v.. B& to S. A. Varga, NRC, July 18, 1978
- 4. "Evaestion of T aasier,t Behavior and Small Reactor Coolant System Dreat:s in the li/ " A Assembly Plant", Volume 1, 2, and 3, Babcock &
Wilco.. May 7, 1979 . . . May 16, 1979.
- 5. R. B. Davis to B&W 177 Owners Group, "Small Break Operating Gui,delines", ,
November 7, 1979.
e e
e.
o e
O.< .
e e
e 4 %
1685 038
- TA5LE 1 SOUFCF.S OF ECNCON;U S!fl.ES - 177 FA FIR;T Total a n 11abl ]" FAIIID } VEL ,
I t'IfAL-WATER REACTIO*3 Individual Individual Individual Individual Total cas Total Cas Total Total Cas Total Total Cas Volu:n Vclures thss !!a nses Vol. Ind. Ces rucs tus<ca Vol. Ind. Cas 'tiss H-sres Sverce cas scf scf lb. Ib. nef Vol. ret th. Ib. ecf Vol sef 15. 15.
Lt.toived in reactor Hy&N2 563 H2 - 305 14 11 - I'#
2 coolant g2 - 158 N2 - I2* 3 Frc uurtter steam I12 I "2 136 H - 65 5.9 11 - 0.4 2
'Y*** Ng - 5.5 Ng - 11 Fressurizer water Hy&Ny 30 fl y - 20 0.91 11 - 0.11 2
'P*** N - 0.8 Ny - 10 2 Fission gases in Er & xe 1 86 Kr - 20 65.5 K r - 4. 8 1.9 Kr - 0.2 0.66 Kr - 0.05 1.9 . .
core le - 166 Xe - 60.7 Xe - 1.7 Ze - 0.61 Fuel roJ fill gas He & sore 1133 Ile - 1092 14.8 I!a - 11.5 11.3 Ile - 10.9 0.16 Ha - 0.12 11.3 N2 &02 N2 - 32 N2-28 N2 - 0.3 N2 - 0.02 02-9 02 - 0. 8 02 - 0.1 02 - 0.01 P:ctal water reaction Hy 416,500 - 2320 - 4165 - .* 3.2 -
(1002) 8:::23 1:U tank gas space H2 ' "2 726 H2 - 421 26.1 H2 - 2.3 H2 - 305 N2 - 23.8 .
- 3 t*J tank water space H2 ' "2 24 H2 - 16 0.71 H2 - 0.09 N2-8 N2 - 0.62 BWST Air (N2 1383 N2 - 902 121.2 N2 - 70.3
- 50) 2 02 - 481 02 - 50.9 p
CF tack gas space N2 26,248 - 2047 -
(two tanks)
CF tank water space H 964 -
75 -
g 2
(two tanks) g M suntions .
- 1. RCS contains 40 std. cc H2/Kg water & 20 std. cc N /Kg 2 water, with water volume = 10,690 f t at 583F and 2200 psia.
- 2. 'Fressurizer water contains 40 std. cc H /Kg 2 water & 20 std. cc N2/Kg water with Henry's Law relation between water space and steam space at 650F.
. *'ater
. volu=e = 825 ft3 and steam volume - 716 f t3
& 3. Fission gases based on inventory in core at 292 EFFD.
CO &. Fuct rod gas based on each rod containing 0.0375, geol Ite. 0.0011 gmol N2 and 0.00029 smol/02 .
W 5. Metal-water reaction based on 52,000 lb. Zr cladding.
- 6. MU t.uk values bascJ on taak containing 200 ft3 gas space and I.00 ft3 water space at 120F with the water containing 40 std. cc H2/Kg and 20 std.
C cc N 2 /K;; with Henry's Lae relationship betwecn cases in water and in gas space.
U EWST contains 450,000 Callons of water saturated with air, t.e.,15 std. cc N /Kg
- 7. 2 and 8 std. cc 0 /K8*
2 3
- 8. Each CF tank contains 1040 ft waterand370f[gasspacewith600pst C N2 at 120F with tienry's 1.aw relation between water and gas.
- 9. Values for 11 f atte3_ fuel based on Xe and Kr fis :lon product inventory acJ fuct rsd fill gas (!!e) in 12 of fuct rods being released to coolant.
- 10. Valws for 1% retal-witer reaction ba:,eJ on gases in Item 9 above and I'2 released f rom 12 of Zr cladding (520 lb.) reacting with coolant.
J 6
s.
TABLE 2: Symptoms and Corrective Actions for a Loss of flatural Circulation During a Small Break (Current Procedures)
SYMPTOMS
- 1. Saturated coolant conditions
- 2. Increasing primary system pressure and temperature with stable or decreasing secondary pressure. -
CORRECTIVE ACTION
- l. Maximize liPI (control HPI if a subcooled margin is re-established)
- 2. E,nsure secondary cooling (i .e'. , auxiliary feedwater available with proper steam generator level control), ,
- 3. Restore RCP flow (one per loop) when possible per the instructions below. If RC pumps cannot be operated and pressure is increasing go to Step 3.5.
3.1 If pressure is increasing, starting a pump is permissible at RC pressure greater than 1600 psig.
3.2 If reactor coolant system pressure exceeds steam generator secondary pressure by 600 psig or more " bump" one reactor coolant pump for a period of approximately 10 seconds (preferably in operable steam generator loop). Allow reactor coolant system pressure to stabilize. Continue cooldown.
If reactor coolant system pressure again exceeds secondary pressure by 600 psi, wait at least 15 minutes and repeat the pump " bump". Bump alternate pumps so that no pump is .
bumped more than once in an hour. This may be repeated, with an interval of 15 minutes, up to 5 times. After the fifth " bump", allow the reactor coolant pump to continue in
~
operation. -
3.3 If pressure has stabilized for greater than one hour, secondary pressure is less than 100 psig and primary pressure is greater than 250 psig, bump a pump, wait 30 minutes, and start an alternate pump.
3.4 If forced flow is established, continue plant cooldown at 100F/
hr. to achieve long term cooling with the LPI/DHR systems.
3.5 If a reactor coolant pump cannot be operated and reactor coolant system pressure reaches 2300 psig, open pressurizer PORV to reduce reactor coolant system pressure. Reclose PORV when RCS pressure falls to 100 psi above the secondary pressure. Repeat if necessary. If PORY is not operable, pressurizer safety valves will reliev overpressure.
2 1685 040
e 3.6 liaintain RC pressure as indicated in 3.5 if pressure increases.
Maintain this cooling mode until an RC pump is started or steam generator cooling is established.
3.7 If SG cooling is established, initiate plant cooldown at 100F/hr. to achieve long term cooling with the LPI/DHR systems. .
4 O
e
=
0 g
0 e
e e
4
, '-c . .
5 1685 041
FIGURE 1 T
L j =~cd DIFFUSION LAYEN
/ l p
/ Pgo
/ i j j g, _p 0
VAPOR AND j f l , NON-CONDENSIBLES
/ O21 L -T
} , !
/ pTgi
. / '
/(' T, l
/
/
/ # CONCENSATE
/ -
P = TOTAL PRESSURE P
PARTIAL PRESSURE OF GAS AT INTERFACE, Ibr/ft2 aj =
Pa = 2
, a PARTI AL PRESSURE OF GAS AT BULK CONDITl0NS,1bg/f t Pgj = 2 PARTIAL PRESSURE OF VAPOR AT' INTERFACE, Ibg/ft
=
Pgo PARTIAL PRESSURE OF VAPOR AT BULK CONDITIONS, Ibg/ft2' T, = WALL TEMPERATURE, F Tgj = TE!'PERATURE AT INTERFACE, *F
= BULK TEUPERATURE, Tgo F
REFERENCE:
- 1) COLBURN, A.P. AND HOUGEN, D.A., "DESICN OF COOLER CONDENSERS FOR MIXTURES OF YAPORS WITH NONCONDENSING GASES", IND. ENG. CHEM 26(11), 1934.
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ovv.sa ..v . au s,v License No. NPF-3 Serial No. 569
, December 28, 1979 Attachment B D
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.t duestion23: Provide the reactor coolant system response to a stuck open PORV for the case of a small break which causes the reactor coolant system to pressurize to the PORV setpoint.
. i, Response
- The resultant system response for a case of.a small break which !
causes the reactor coolant system to pressurize to the PORV setpoint {-
and result in a stuck open PORV can be qualitatively assessed based on previous analyses and is provided below. As is demonstrated .
the,small break op, crating guidelines which have been developed are
Numerous small break calculations have been performed for the {3 operating 177 FA plants. These calculations are provided in .
References 1 through 6. As is demonstrated by these studies, ;,
repressurization to the FORV following a small break is ppssible j only if the break is extremely small (<0.01 ft2) and if there is no feedwater available to the steam generators. For plants with a j,
i safety grade, redundant auxiliary feedwater system, sue.h as the Davis- .
Besse plant , the probability of a small break which pressurizes :
to the PORV setpoint is considered extremely unlikely.. -
The system response of a very small break (<0.01 ft2) with a concurrent :
loss of all feedwater is presented in Reference .4 .
The system will initially undergo a subcooled depressurization. ;
During this period of the transient, the reactor trips, the pressurizer
- drains, and the initial SG inventory boils off. For these smaller -
sized breaks (<0.01 ft2), the SG initial inventory boils off prior .
to system depressurization to th,e ESFAS signal ~. Following the loss of the SG heat sink, the fluid *in the RCS increases in temperature ,. l, and becomes saturated. Since the volumetric flowrate out the ,,
break, following the establishment of saturation conditions in the ..
RCS, is less than the volumetric steam production caused by decay
heat removal, the RCS repressurizes the pressur'izer starts to !'
refill. Thus, for these breaks, no ECCS equipment is automatically
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actuated prior to system repressurization. .
l*
. System repressurization would continue until the PORV setpoint is '
reached if no operator action is taken to-prevent it. The earliest time that the PORV setpoint would be reached is >I 4 minutes, for a zero break case and % 20 minutes for the 0.01 ft break. It should be noted that actuation of the AFW system prior to these times would prevent opening of the PORV.
While analys'is of this break combination has not presently been performed, the present operator guidelines for small breaks were constructed to mitigate the consequences of such an event. The
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o'perator is instructed to maintain maximum HPI flow and're-establish feedwate.'
to the SG as soon as possible if AFW is not automatically initiated. Also, the guidelines require manual initiation of the startup feedwater pump upon loss of the SG heat sink. Should auxiliary feedwater continue to remain unavailable and the primary system pressure starts to increase, the operator is instructed to initiate the makeup system, open the PORV and leave it open in order to maintain the RCS pressure as low as possible and maximize the flows. These operator guidelines thus will minimize the consequences of a small break which repressurizes the RCS.
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References
- 1. BAW-10103A, Rev. 2, "ECCS Analysis of B&W's 177-FA Lowered-Loop NSS," July 1977.
- 2. Letter from J. H. Taylor to S. A. Varga of July 18, 1978, concerning 177 FA plants small break analysis.
- 3. BAW-10075A, Rev.1 "Multinode Analysis of Small Breaks for B&W's 177-Euel Assembly Nuclear Plants with Raised Loop Arrangement and Internals Vent Valves," March 1976.
- 4. Letter from J. H. Taylor to R. Mattson of May 'i,1979,
- " Evaluation of Transient Behavior and Small Reactoic Coolant System Breaks in the 177-Fuel Assembly Plant," Volume I,
- Section 6; and Volume III, Revision'1, by Toledo Edison letter dated May 22, 1979.
- 5. Letter from J. H. Taylor to R. J. Mattson of May 12, 1979, "Small Break in the Pressurizer (PORV) with No Auxiliary Feedwater and One HPI Pump."
- 6. Letter from R. B. Davis to B&W 177 Owners Group, Technical Subcommittee on TMI-2 Incident Related Tasks,
Subject:
Tasponses to IE Bulletin 79-05C Action Items, August 21, 1979.
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