Letter Sequence Other |
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Results
Other: ML20082A843, ML20082S949, ML20083F063, ML20084J011, ML20091E302, ML20095A385, ML20125D934, ML20202E952, ML20212G878, ML20214W452, ML20216H730, ML20236U722, ML20237E005
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MONTHYEARML20082A8431983-11-0707 November 1983 RO 83-43:on 831011,util Notified NRC of Various Nonconservative Errors Made in Original Analyses Which Constitute Basis of Tech Spec 4.1.9.GA Technologies Provided Interim Curves Which Corrected Deficiencies Project stage: Other ML20082S9491983-11-30030 November 1983 Forwards Figure for Limiting Condition for Operation 4.1.9-1 Re Circulation Flow/Core Power,To Replace Figure Submitted w/831107 Ltr.Portions of Less Conservative Figure Deleted. Limit Will Be Observed Until Tech Spec Changed Project stage: Other ML20083F0631983-12-15015 December 1983 Proposed Tech Specs Revising Limiting Condition for Operation 4.1.9 Re Core Region Temp Rise Project stage: Other ML20083F0611983-12-15015 December 1983 Application for Amend to License DPR-34,revising Tech Spec Limiting Condition for Operation 4.1.9 Re Core Region Temp Rise,Per Previous Commitments.Class III Amend Fee Encl Project stage: Request ML20086R7281984-02-24024 February 1984 Responds to Request for Addl Info Re Proposed Amend to Limiting Condition for Operation 4.1.9 Re Average Core Temp During Shutdown Project stage: Request ML20084J0111984-04-0606 April 1984 ORNL Assistance in Evaluating Licensing Request - Fsv Limiting Condition for Operation 4.1.9, Monthly Rept for Mar 1984 Project stage: Other ML20091E3021984-05-0909 May 1984 Interim Rept on ORNL Assistance in Evaluating Licensing Request - Amend of Fort St Vrain Reactor Tech Spec Limiting Condition for Operation 4.1.9 Project stage: Other ML20095A3851984-08-14014 August 1984 Forwards Response to ORNL Concerns & Proposal Identified in NRC .Basis for Discussion of Ornl/Util Concerns & Resolution of Issues Delaying Implementation of Revised Limiting Condition for Operation 4.1.9 Discussed Project stage: Other ML20098H1721984-08-21021 August 1984 Provides Replies & Comments to Util Re Tech Spec Limiting Condition for Operation 4.1.9 in Advance of 840823 Meeting W/Util in Arlington,Tx Project stage: Meeting ML20099H7931984-11-20020 November 1984 Submits Addl Info Requested by NRC During 840823 Meeting Re Proposed Amend to Limiting Condition for Operation 4.1.9 Concerning Automated Orifice Valve Adjustments.Work Continuing on Listed Remaining Issues Project stage: Meeting ML20125D9341985-05-20020 May 1985 Submits Current Status of Work on Tech Spec Limiting Condition for Operation 4.1.9 Re Core Region Temp Rise. Calculational Differences Between Thermal/Hydraulic Analyses Being Investigated.Summary of Open Items Provided Project stage: Other ML20133K6751985-10-17017 October 1985 Notifies That Draft Proposed Rev to Tech Spec Limiting Condition for Operation 4.1.9 Re Core Region Temp Rise Will Be Submitted for Review by 851122 Project stage: Draft Other ML20137E4941985-11-22022 November 1985 Forwards Draft Rev to Tech Spec Limiting Condition for Operation 4.1.9 Re Min Helium Flow/Core Region Temp Rise for Review.Formal Request for Amend to License DPR-34 Will Be Submitted Upon Resolution of NRC Comments Project stage: Draft Request ML20141B7451986-03-28028 March 1986 Summary of 860313 Meeting W/Util in Arlington,Tx Re Tech Spec Limiting Condition for Operation 4.1.9 Covering Frequency of Manipulation of Reactor Controls.Supporting Documentation Encl Project stage: Meeting ML20211E9171986-06-13013 June 1986 Advises That Formal Application for Tech Spec Rev Re Limiting Condition for Operation 4.1.9, Min Helium Flow & Max Core Region Temp Rise, Will Be Submitted by 860703. Extension Required for Resolution of NRC 860606 Comments Project stage: Request ML20202E9521986-07-0909 July 1986 Proposed Tech Specs,Incorporating Min Helium Flow & Max Core Region Temp Rise Surveillance Requirements Project stage: Other ML20202E9371986-07-0909 July 1986 Application for Amend to License DPR-34,changing Tech Specs to Incorporate Helium Flow & Max Core Region Temp Rise Surveillance Requirements Project stage: Request ML20202E9261986-07-0909 July 1986 Forwards Application for Amend to License DPR-34,changing Tech Specs to Incorporate Min Helium Flow & Max Core Region Temp Rise Surveillance Requirements.Fee Paid Project stage: Request ML20237E0051986-10-16016 October 1986 Forwards Final Version of Facility Limiting Condition for Operation 4.1.9 Technical Evaluation Rept for Review & Approval.Related Info Encl Project stage: Other ML20214W4521986-10-16016 October 1986 Review of Proposed Tech Spec Change:Core Inlet Orifice Valves/Min Helium Flow & Max Core Region Temp Rise, Technical Evaluation Rept Project stage: Other ML20214W4361986-12-0505 December 1986 Forwards Review of Proposed Tech Spec Change:Core Inlet Orifice Valves/Min Helium Flow & Max Core Region Temp Rise. Requests Submittal Addressing Staff Concerns within 45 Days of Ltr Date Project stage: Approval ML20212G8781987-01-14014 January 1987 Advises That Util Will Submit Revised Proposal for Tech Spec 4.1.9 Re Min Helium Flow & Max Core Region Temp Rise by 870225.Util Finalizing Curves That Determine Allowed Outage Times for DHR Equipment for Inclusion in Tech Specs Project stage: Other ML20216J4731987-06-25025 June 1987 Forwards Application for Amend to License DPR-34,changing Tech Spec Limiting Condition for Operation 4.1.9 to Reduce Potential for Flow Stagnation & Prevent Excessive Fuel Temps in Core & Spec 5.1.8 to Incorporate Requirements.Fee Paid Project stage: Request ML20216H7301987-06-25025 June 1987 Proposed Tech Specs,Adding Definition of Calculated Bulk Core Temp & Core Average Inlet Temp for Determination of Core Temp Project stage: Other ML20216H7221987-06-25025 June 1987 Application for Amend to License DPR-34,changing Tech Spec Limiting Condition for Operation 4.1.9 to Reduce Potential for Flow Stagnation & Prevent Excessive Fuel Temps in Core & Spec 5.1.8 to Incorporate Associated Requirements Project stage: Request ML20236U7221987-08-0606 August 1987 Addendum to Review of Proposed Tech Spec Change:Core Inlet Valves/Min Helium Flow & Max Core Region Temp Rise, Limiting Conditon for Operation 4.1.9, Technical Evaluation Rept Project stage: Other ML20236U5041987-11-23023 November 1987 Forwards Amend 57 to License DPR-34 & Safety Evaluation. Amend Revises Tech Specs to Ensure Sufficient Helium Coolant Flow to Prevent Overheating of Fuel While in Low Power or Shutdown Modes Project stage: Approval 1985-11-22
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Category:CONTRACTED REPORT - RTA
MONTHYEARML20148D6041997-03-31031 March 1997 Confirmatory Survey for Fsv Nuclear Station,Psc, Platteville,Co,Final Rept ML20148D5861997-03-26026 March 1997 Confirmatory Survey Plan for Group E Effluent Discharge Pathway Areas at Fsv Nuclear Station ML20134C8721997-01-29029 January 1997 Draft Confirmatory Survey for Fsv Nuclear Station Public Svc Co of Co,Platteville,Co ML20092G3681995-06-19019 June 1995 Confirmatory Survey for Repower Area Fsv Platteville,Co, Final Rept ML20084F3931995-05-16016 May 1995 Draft Rept, Confirmatory Survey for Repower Area Fsv Platteville,Co ML20245E6601989-01-0606 January 1989 Independent Review of Fire Protection Program Plan,Public Svc Co of Co,Fort St Vrain Nuclear Generating Station ML20245D1951988-10-31031 October 1988 Integrated Technical Evaluation Rept for Review of Fort St Vrain Tech Spec Upgrade Program ML20151B0031988-03-31031 March 1988 Technical Evaluation Rept for Plant Protective Sys Trip Setpoints for Fort St Vrain Nuclear Generating Station, Informal Rept ML20151S6061988-02-29029 February 1988 Rev 1 to Technical Evaluation Rept for Conformance to NRR Generic Ltrs 83-36 & 83-37 by Fort St Vrain Nuclear Generating Station, Informal Rept ML20195G9611987-12-14014 December 1987 Technical Evaluation of Proposed Redundancy Requirements for Helium Circulation Tech Specs to Accommodate Rapid Depressurization Accident, Technical Evaluation Rept ML20234D6291987-12-14014 December 1987 Technical Assistance to Region IV (Fort St.Vrain), Monthly Business Ltr Rept for Oct 1987 ML20236V3191987-10-31031 October 1987 Draft Plant Protective Sys Trip Setpoints for Fort St Vrain Nuclear Generating Station, Technical Evaluation Rept ML20235R4161987-10-31031 October 1987 Draft Eg&G Technical Evaluation Rept for Plant Protective Sys Trip Setpoints for Fort St Vrain Nuclear Generating Station ML20235V8871987-10-31031 October 1987 Draft Plant Protective Sys Trip Setpoints for Fort St Vrain Nuclear Generating Station, Technical Evaluation Rept ML20236C2611987-09-28028 September 1987 Fort St Vrain Safe Shutdown Using Condensate Sys ML20236U7221987-08-0606 August 1987 Addendum to Review of Proposed Tech Spec Change:Core Inlet Valves/Min Helium Flow & Max Core Region Temp Rise, Limiting Conditon for Operation 4.1.9, Technical Evaluation Rept ML20246J3191987-07-31031 July 1987 Evaluation of Operator Roles in Mitigating High Energy Line Break at Fort St Vrain Nuclear Generating Station ML20235R6121987-07-0101 July 1987 Investigating Causes & Consequences of Cracked Graphite Fuel Elements ML20235M8321987-06-30030 June 1987 Conformance to Generic Ltr 83-28,Item 2.2.1 - Equipment Classification for All Other Safety-Related Components:Fort St Vrain, Final Informal Rept ML20215A9031987-06-0505 June 1987 Technical Assistance to Region IV (Fort St Vrain), Monthly Business Ltr Rept for Apr 1987 ML20235G1141987-06-0404 June 1987 Draft Fort St Vrain Safe Shutdown from 82% Power, Technical Evaluation Rept ML20234C6091987-06-0404 June 1987 Fort St Vrain Safe Shutdown from 82% Power, Technical Evaluation Rept ML20215L9331987-05-31031 May 1987 Evaluation of Integrated Sys Study of CRD Mechanism Rod Position Indication Instrumentation for Fort St Vrain Nuclear Generating Station, Informal Technical Evaluation Rept ML20214H2061987-05-15015 May 1987 Technical Assistance to Region IV (Fort St Vrain), Monthly Business Ltr Rept for Mar 1987 ML20215M3901987-04-30030 April 1987 Final Conformance to Generic Ltr 83-28,Item 2.2.2 - Vendor Interface Programs for All Other Safety-Related Components: Fort St Vrain, Informal Rept ML20210A3851987-04-13013 April 1987 Technical Assistance to Region IV (Fort St Vrain), Monthly Progress Rept for Feb 1987 ML20214R8711987-03-31031 March 1987 Technical Evaluation Rept Evaluation of Bldg 10 Electrical Mods Fort St Vrain Nuclear Generating Station, Informal Rept ML20214Q5411987-03-31031 March 1987 Evaluation of Confinement Environ Temp Following High Energy Line Breaks Proposed for Fort St Vrain Environ Qualification Program, Informal Technical Evaluation Rept ML20207T3721987-03-10010 March 1987 Technical Assistance to Region IV (Fort St Vrain), Monthly Business Ltr Rept for Jan 1987 ML20207R9471987-03-0606 March 1987 Rev 1 to Review of Fort St Vrain Onsite AC `Standby' Power Sys Re Compliance to Single Failure Criterion & Ser ML20212F6771987-02-13013 February 1987 Technical Assistance to Region IV (Fort St Vrain), Monthly Progress Rept for Dec 1986 ML20216J1991987-02-0404 February 1987 Evaluation of Long-Term Effects of Moisture Ingress in Fort St Vrain Nuclear Reactor ML20210P6451987-01-31031 January 1987 Draft Evaluation of Confinement Environ Temps Following High Energy Line Breaks Proposed for Fort St Vrain Environ Qualification Program, Technical Evaluation Rept ML20245B3401986-12-31031 December 1986 Technical Assistance for Fort St Vrain Task Ii:Pcrv & Pcrv Tendon Evaluation, Final Technical Evaluation Rept ML20207K4701986-12-31031 December 1986 Final Review of Convection Heat Transfer Coefficients Utilized in Fort St Vrain Main Steam Line Break Analysis, Technical Evaluation Rept ML20212B7321986-11-30030 November 1986 Evaluation of Proposed CRD & Orifice Assembly 300 F Temp Limits,Fort St Vrain Nuclear Generating Station ML20197C5981986-10-31031 October 1986 Review of Convection Heat Transfer Coefficients Utilized in Fort St Vrain Main Steam Line Break Analyses, Technical Evaluation Rept ML20213D8251986-10-31031 October 1986 Technical Assistance to Region IV (Fort St Vrain), Monthly Progress Rept for Sept 1986 ML20214W4521986-10-16016 October 1986 Review of Proposed Tech Spec Change:Core Inlet Orifice Valves/Min Helium Flow & Max Core Region Temp Rise, Technical Evaluation Rept ML20214P4681986-09-30030 September 1986 Conformance to NRR Generic Ltrs 83-36 & 83-37 by Fort St Vrain Nuclear Generating Station, Informal Rept ML20211A6091986-09-26026 September 1986 Technical Assistance to Region IV (Fort St Vrain), Monthly Business Ltr Rept for Aug 1986 ML20236D3811986-09-24024 September 1986 Technical Evaluation of Dcrdr Info Supplementing Summary Rept for Fort St Vrain Nuclear Generating Station ML20209F1021986-09-0505 September 1986 Technical Assistance to Region IV (Fort St Vrain), Monthly Business Ltr for Jul 1986 ML20212Q1011986-08-31031 August 1986 Conformance to Reg Guide 1.97,Fort St Vrain Nuclear Generating Station, Informal Rept ML20213G5811986-08-0707 August 1986 Draft Review of Fort St Vrain Onsite AC 'Standby' Power Sys W/Regards to Compliance to Single Failure Criterion & Ser ML20205E3711986-07-31031 July 1986 Conformance to Generic Ltr 83-28,Item 2.1 (Part 1), Equipment Classification (Reactor Trip Sys Components) Fort Calhoun,Millstone Unit 2,Nine Mile Point 2 & Fort St Vrain, Technical Evaluation Rept ML20244E1901986-07-31031 July 1986 Conformance to Generic Ltr 83-28,Item 2.1 (Part 1) Equipment Classification (Reactor Trip Sys Components), Fort Calhoun,Millstone Unit 2,Nine Mile Point Unit 2 & Fort St Vrain ML20204E6811986-07-24024 July 1986 Technical Assistance to Region IV - Fort St Vrain, Monthly Business Ltr Repts for May & June 1986 ML20207J7931986-07-0909 July 1986 Fort St Vrain Fuel Element Dynamic Response ML20236D3741986-06-20020 June 1986 Technical Evaluation of Dcrdr Info Supplementing Summary Rept for Fort St Vrain Nuclear Generating Station 1997-03-31
[Table view] Category:QUICK LOOK
MONTHYEARML20148D6041997-03-31031 March 1997 Confirmatory Survey for Fsv Nuclear Station,Psc, Platteville,Co,Final Rept ML20148D5861997-03-26026 March 1997 Confirmatory Survey Plan for Group E Effluent Discharge Pathway Areas at Fsv Nuclear Station ML20134C8721997-01-29029 January 1997 Draft Confirmatory Survey for Fsv Nuclear Station Public Svc Co of Co,Platteville,Co ML20092G3681995-06-19019 June 1995 Confirmatory Survey for Repower Area Fsv Platteville,Co, Final Rept ML20084F3931995-05-16016 May 1995 Draft Rept, Confirmatory Survey for Repower Area Fsv Platteville,Co ML20245E6601989-01-0606 January 1989 Independent Review of Fire Protection Program Plan,Public Svc Co of Co,Fort St Vrain Nuclear Generating Station ML20245D1951988-10-31031 October 1988 Integrated Technical Evaluation Rept for Review of Fort St Vrain Tech Spec Upgrade Program ML20151B0031988-03-31031 March 1988 Technical Evaluation Rept for Plant Protective Sys Trip Setpoints for Fort St Vrain Nuclear Generating Station, Informal Rept ML20151S6061988-02-29029 February 1988 Rev 1 to Technical Evaluation Rept for Conformance to NRR Generic Ltrs 83-36 & 83-37 by Fort St Vrain Nuclear Generating Station, Informal Rept ML20195G9611987-12-14014 December 1987 Technical Evaluation of Proposed Redundancy Requirements for Helium Circulation Tech Specs to Accommodate Rapid Depressurization Accident, Technical Evaluation Rept ML20234D6291987-12-14014 December 1987 Technical Assistance to Region IV (Fort St.Vrain), Monthly Business Ltr Rept for Oct 1987 ML20236V3191987-10-31031 October 1987 Draft Plant Protective Sys Trip Setpoints for Fort St Vrain Nuclear Generating Station, Technical Evaluation Rept ML20235R4161987-10-31031 October 1987 Draft Eg&G Technical Evaluation Rept for Plant Protective Sys Trip Setpoints for Fort St Vrain Nuclear Generating Station ML20235V8871987-10-31031 October 1987 Draft Plant Protective Sys Trip Setpoints for Fort St Vrain Nuclear Generating Station, Technical Evaluation Rept ML20236C2611987-09-28028 September 1987 Fort St Vrain Safe Shutdown Using Condensate Sys ML20236U7221987-08-0606 August 1987 Addendum to Review of Proposed Tech Spec Change:Core Inlet Valves/Min Helium Flow & Max Core Region Temp Rise, Limiting Conditon for Operation 4.1.9, Technical Evaluation Rept ML20246J3191987-07-31031 July 1987 Evaluation of Operator Roles in Mitigating High Energy Line Break at Fort St Vrain Nuclear Generating Station ML20235R6121987-07-0101 July 1987 Investigating Causes & Consequences of Cracked Graphite Fuel Elements ML20235M8321987-06-30030 June 1987 Conformance to Generic Ltr 83-28,Item 2.2.1 - Equipment Classification for All Other Safety-Related Components:Fort St Vrain, Final Informal Rept ML20215A9031987-06-0505 June 1987 Technical Assistance to Region IV (Fort St Vrain), Monthly Business Ltr Rept for Apr 1987 ML20235G1141987-06-0404 June 1987 Draft Fort St Vrain Safe Shutdown from 82% Power, Technical Evaluation Rept ML20234C6091987-06-0404 June 1987 Fort St Vrain Safe Shutdown from 82% Power, Technical Evaluation Rept ML20215L9331987-05-31031 May 1987 Evaluation of Integrated Sys Study of CRD Mechanism Rod Position Indication Instrumentation for Fort St Vrain Nuclear Generating Station, Informal Technical Evaluation Rept ML20214H2061987-05-15015 May 1987 Technical Assistance to Region IV (Fort St Vrain), Monthly Business Ltr Rept for Mar 1987 ML20215M3901987-04-30030 April 1987 Final Conformance to Generic Ltr 83-28,Item 2.2.2 - Vendor Interface Programs for All Other Safety-Related Components: Fort St Vrain, Informal Rept ML20210A3851987-04-13013 April 1987 Technical Assistance to Region IV (Fort St Vrain), Monthly Progress Rept for Feb 1987 ML20214R8711987-03-31031 March 1987 Technical Evaluation Rept Evaluation of Bldg 10 Electrical Mods Fort St Vrain Nuclear Generating Station, Informal Rept ML20214Q5411987-03-31031 March 1987 Evaluation of Confinement Environ Temp Following High Energy Line Breaks Proposed for Fort St Vrain Environ Qualification Program, Informal Technical Evaluation Rept ML20207T3721987-03-10010 March 1987 Technical Assistance to Region IV (Fort St Vrain), Monthly Business Ltr Rept for Jan 1987 ML20207R9471987-03-0606 March 1987 Rev 1 to Review of Fort St Vrain Onsite AC `Standby' Power Sys Re Compliance to Single Failure Criterion & Ser ML20212F6771987-02-13013 February 1987 Technical Assistance to Region IV (Fort St Vrain), Monthly Progress Rept for Dec 1986 ML20216J1991987-02-0404 February 1987 Evaluation of Long-Term Effects of Moisture Ingress in Fort St Vrain Nuclear Reactor ML20210P6451987-01-31031 January 1987 Draft Evaluation of Confinement Environ Temps Following High Energy Line Breaks Proposed for Fort St Vrain Environ Qualification Program, Technical Evaluation Rept ML20245B3401986-12-31031 December 1986 Technical Assistance for Fort St Vrain Task Ii:Pcrv & Pcrv Tendon Evaluation, Final Technical Evaluation Rept ML20207K4701986-12-31031 December 1986 Final Review of Convection Heat Transfer Coefficients Utilized in Fort St Vrain Main Steam Line Break Analysis, Technical Evaluation Rept ML20212B7321986-11-30030 November 1986 Evaluation of Proposed CRD & Orifice Assembly 300 F Temp Limits,Fort St Vrain Nuclear Generating Station ML20197C5981986-10-31031 October 1986 Review of Convection Heat Transfer Coefficients Utilized in Fort St Vrain Main Steam Line Break Analyses, Technical Evaluation Rept ML20213D8251986-10-31031 October 1986 Technical Assistance to Region IV (Fort St Vrain), Monthly Progress Rept for Sept 1986 ML20214W4521986-10-16016 October 1986 Review of Proposed Tech Spec Change:Core Inlet Orifice Valves/Min Helium Flow & Max Core Region Temp Rise, Technical Evaluation Rept ML20214P4681986-09-30030 September 1986 Conformance to NRR Generic Ltrs 83-36 & 83-37 by Fort St Vrain Nuclear Generating Station, Informal Rept ML20211A6091986-09-26026 September 1986 Technical Assistance to Region IV (Fort St Vrain), Monthly Business Ltr Rept for Aug 1986 ML20236D3811986-09-24024 September 1986 Technical Evaluation of Dcrdr Info Supplementing Summary Rept for Fort St Vrain Nuclear Generating Station ML20209F1021986-09-0505 September 1986 Technical Assistance to Region IV (Fort St Vrain), Monthly Business Ltr for Jul 1986 ML20212Q1011986-08-31031 August 1986 Conformance to Reg Guide 1.97,Fort St Vrain Nuclear Generating Station, Informal Rept ML20213G5811986-08-0707 August 1986 Draft Review of Fort St Vrain Onsite AC 'Standby' Power Sys W/Regards to Compliance to Single Failure Criterion & Ser ML20205E3711986-07-31031 July 1986 Conformance to Generic Ltr 83-28,Item 2.1 (Part 1), Equipment Classification (Reactor Trip Sys Components) Fort Calhoun,Millstone Unit 2,Nine Mile Point 2 & Fort St Vrain, Technical Evaluation Rept ML20244E1901986-07-31031 July 1986 Conformance to Generic Ltr 83-28,Item 2.1 (Part 1) Equipment Classification (Reactor Trip Sys Components), Fort Calhoun,Millstone Unit 2,Nine Mile Point Unit 2 & Fort St Vrain ML20204E6811986-07-24024 July 1986 Technical Assistance to Region IV - Fort St Vrain, Monthly Business Ltr Repts for May & June 1986 ML20207J7931986-07-0909 July 1986 Fort St Vrain Fuel Element Dynamic Response ML20236D3741986-06-20020 June 1986 Technical Evaluation of Dcrdr Info Supplementing Summary Rept for Fort St Vrain Nuclear Generating Station 1997-03-31
[Table view] Category:ETC. (PERIODIC
MONTHYEARML20148D6041997-03-31031 March 1997 Confirmatory Survey for Fsv Nuclear Station,Psc, Platteville,Co,Final Rept ML20148D5861997-03-26026 March 1997 Confirmatory Survey Plan for Group E Effluent Discharge Pathway Areas at Fsv Nuclear Station ML20134C8721997-01-29029 January 1997 Draft Confirmatory Survey for Fsv Nuclear Station Public Svc Co of Co,Platteville,Co ML20092G3681995-06-19019 June 1995 Confirmatory Survey for Repower Area Fsv Platteville,Co, Final Rept ML20084F3931995-05-16016 May 1995 Draft Rept, Confirmatory Survey for Repower Area Fsv Platteville,Co ML20245E6601989-01-0606 January 1989 Independent Review of Fire Protection Program Plan,Public Svc Co of Co,Fort St Vrain Nuclear Generating Station ML20245D1951988-10-31031 October 1988 Integrated Technical Evaluation Rept for Review of Fort St Vrain Tech Spec Upgrade Program ML20151B0031988-03-31031 March 1988 Technical Evaluation Rept for Plant Protective Sys Trip Setpoints for Fort St Vrain Nuclear Generating Station, Informal Rept ML20151S6061988-02-29029 February 1988 Rev 1 to Technical Evaluation Rept for Conformance to NRR Generic Ltrs 83-36 & 83-37 by Fort St Vrain Nuclear Generating Station, Informal Rept ML20195G9611987-12-14014 December 1987 Technical Evaluation of Proposed Redundancy Requirements for Helium Circulation Tech Specs to Accommodate Rapid Depressurization Accident, Technical Evaluation Rept ML20234D6291987-12-14014 December 1987 Technical Assistance to Region IV (Fort St.Vrain), Monthly Business Ltr Rept for Oct 1987 ML20236V3191987-10-31031 October 1987 Draft Plant Protective Sys Trip Setpoints for Fort St Vrain Nuclear Generating Station, Technical Evaluation Rept ML20235R4161987-10-31031 October 1987 Draft Eg&G Technical Evaluation Rept for Plant Protective Sys Trip Setpoints for Fort St Vrain Nuclear Generating Station ML20235V8871987-10-31031 October 1987 Draft Plant Protective Sys Trip Setpoints for Fort St Vrain Nuclear Generating Station, Technical Evaluation Rept ML20236C2611987-09-28028 September 1987 Fort St Vrain Safe Shutdown Using Condensate Sys ML20236U7221987-08-0606 August 1987 Addendum to Review of Proposed Tech Spec Change:Core Inlet Valves/Min Helium Flow & Max Core Region Temp Rise, Limiting Conditon for Operation 4.1.9, Technical Evaluation Rept ML20246J3191987-07-31031 July 1987 Evaluation of Operator Roles in Mitigating High Energy Line Break at Fort St Vrain Nuclear Generating Station ML20235R6121987-07-0101 July 1987 Investigating Causes & Consequences of Cracked Graphite Fuel Elements ML20235M8321987-06-30030 June 1987 Conformance to Generic Ltr 83-28,Item 2.2.1 - Equipment Classification for All Other Safety-Related Components:Fort St Vrain, Final Informal Rept ML20215A9031987-06-0505 June 1987 Technical Assistance to Region IV (Fort St Vrain), Monthly Business Ltr Rept for Apr 1987 ML20235G1141987-06-0404 June 1987 Draft Fort St Vrain Safe Shutdown from 82% Power, Technical Evaluation Rept ML20234C6091987-06-0404 June 1987 Fort St Vrain Safe Shutdown from 82% Power, Technical Evaluation Rept ML20215L9331987-05-31031 May 1987 Evaluation of Integrated Sys Study of CRD Mechanism Rod Position Indication Instrumentation for Fort St Vrain Nuclear Generating Station, Informal Technical Evaluation Rept ML20214H2061987-05-15015 May 1987 Technical Assistance to Region IV (Fort St Vrain), Monthly Business Ltr Rept for Mar 1987 ML20215M3901987-04-30030 April 1987 Final Conformance to Generic Ltr 83-28,Item 2.2.2 - Vendor Interface Programs for All Other Safety-Related Components: Fort St Vrain, Informal Rept ML20210A3851987-04-13013 April 1987 Technical Assistance to Region IV (Fort St Vrain), Monthly Progress Rept for Feb 1987 ML20214R8711987-03-31031 March 1987 Technical Evaluation Rept Evaluation of Bldg 10 Electrical Mods Fort St Vrain Nuclear Generating Station, Informal Rept ML20214Q5411987-03-31031 March 1987 Evaluation of Confinement Environ Temp Following High Energy Line Breaks Proposed for Fort St Vrain Environ Qualification Program, Informal Technical Evaluation Rept ML20207T3721987-03-10010 March 1987 Technical Assistance to Region IV (Fort St Vrain), Monthly Business Ltr Rept for Jan 1987 ML20207R9471987-03-0606 March 1987 Rev 1 to Review of Fort St Vrain Onsite AC `Standby' Power Sys Re Compliance to Single Failure Criterion & Ser ML20212F6771987-02-13013 February 1987 Technical Assistance to Region IV (Fort St Vrain), Monthly Progress Rept for Dec 1986 ML20216J1991987-02-0404 February 1987 Evaluation of Long-Term Effects of Moisture Ingress in Fort St Vrain Nuclear Reactor ML20210P6451987-01-31031 January 1987 Draft Evaluation of Confinement Environ Temps Following High Energy Line Breaks Proposed for Fort St Vrain Environ Qualification Program, Technical Evaluation Rept ML20245B3401986-12-31031 December 1986 Technical Assistance for Fort St Vrain Task Ii:Pcrv & Pcrv Tendon Evaluation, Final Technical Evaluation Rept ML20207K4701986-12-31031 December 1986 Final Review of Convection Heat Transfer Coefficients Utilized in Fort St Vrain Main Steam Line Break Analysis, Technical Evaluation Rept ML20212B7321986-11-30030 November 1986 Evaluation of Proposed CRD & Orifice Assembly 300 F Temp Limits,Fort St Vrain Nuclear Generating Station ML20197C5981986-10-31031 October 1986 Review of Convection Heat Transfer Coefficients Utilized in Fort St Vrain Main Steam Line Break Analyses, Technical Evaluation Rept ML20213D8251986-10-31031 October 1986 Technical Assistance to Region IV (Fort St Vrain), Monthly Progress Rept for Sept 1986 ML20214W4521986-10-16016 October 1986 Review of Proposed Tech Spec Change:Core Inlet Orifice Valves/Min Helium Flow & Max Core Region Temp Rise, Technical Evaluation Rept ML20214P4681986-09-30030 September 1986 Conformance to NRR Generic Ltrs 83-36 & 83-37 by Fort St Vrain Nuclear Generating Station, Informal Rept ML20211A6091986-09-26026 September 1986 Technical Assistance to Region IV (Fort St Vrain), Monthly Business Ltr Rept for Aug 1986 ML20236D3811986-09-24024 September 1986 Technical Evaluation of Dcrdr Info Supplementing Summary Rept for Fort St Vrain Nuclear Generating Station ML20209F1021986-09-0505 September 1986 Technical Assistance to Region IV (Fort St Vrain), Monthly Business Ltr for Jul 1986 ML20212Q1011986-08-31031 August 1986 Conformance to Reg Guide 1.97,Fort St Vrain Nuclear Generating Station, Informal Rept ML20213G5811986-08-0707 August 1986 Draft Review of Fort St Vrain Onsite AC 'Standby' Power Sys W/Regards to Compliance to Single Failure Criterion & Ser ML20205E3711986-07-31031 July 1986 Conformance to Generic Ltr 83-28,Item 2.1 (Part 1), Equipment Classification (Reactor Trip Sys Components) Fort Calhoun,Millstone Unit 2,Nine Mile Point 2 & Fort St Vrain, Technical Evaluation Rept ML20244E1901986-07-31031 July 1986 Conformance to Generic Ltr 83-28,Item 2.1 (Part 1) Equipment Classification (Reactor Trip Sys Components), Fort Calhoun,Millstone Unit 2,Nine Mile Point Unit 2 & Fort St Vrain ML20204E6811986-07-24024 July 1986 Technical Assistance to Region IV - Fort St Vrain, Monthly Business Ltr Repts for May & June 1986 ML20207J7931986-07-0909 July 1986 Fort St Vrain Fuel Element Dynamic Response ML20236D3741986-06-20020 June 1986 Technical Evaluation of Dcrdr Info Supplementing Summary Rept for Fort St Vrain Nuclear Generating Station 1997-03-31
[Table view] Category:TEXT-PROCUREMENT & CONTRACTS
MONTHYEARML20148D6041997-03-31031 March 1997 Confirmatory Survey for Fsv Nuclear Station,Psc, Platteville,Co,Final Rept ML20148D5861997-03-26026 March 1997 Confirmatory Survey Plan for Group E Effluent Discharge Pathway Areas at Fsv Nuclear Station ML20134C8721997-01-29029 January 1997 Draft Confirmatory Survey for Fsv Nuclear Station Public Svc Co of Co,Platteville,Co ML20092G3681995-06-19019 June 1995 Confirmatory Survey for Repower Area Fsv Platteville,Co, Final Rept ML20084F3931995-05-16016 May 1995 Draft Rept, Confirmatory Survey for Repower Area Fsv Platteville,Co ML20245E6601989-01-0606 January 1989 Independent Review of Fire Protection Program Plan,Public Svc Co of Co,Fort St Vrain Nuclear Generating Station ML20245D1951988-10-31031 October 1988 Integrated Technical Evaluation Rept for Review of Fort St Vrain Tech Spec Upgrade Program ML20154S3461988-08-31031 August 1988 Notification of Contract Execution,Mod 3,to HTGR (Fort St Vrain) Training Course. Contractor:Ga Co ML20154S3511988-08-31031 August 1988 Mod 3,incorporating Change of Name Agreement from Ga Technologies to General Atomics,To HTGR (Fort St Vrain) Training Course ML20151B0031988-03-31031 March 1988 Technical Evaluation Rept for Plant Protective Sys Trip Setpoints for Fort St Vrain Nuclear Generating Station, Informal Rept ML20151S6061988-02-29029 February 1988 Rev 1 to Technical Evaluation Rept for Conformance to NRR Generic Ltrs 83-36 & 83-37 by Fort St Vrain Nuclear Generating Station, Informal Rept ML20195G9611987-12-14014 December 1987 Technical Evaluation of Proposed Redundancy Requirements for Helium Circulation Tech Specs to Accommodate Rapid Depressurization Accident, Technical Evaluation Rept ML20234D6291987-12-14014 December 1987 Technical Assistance to Region IV (Fort St.Vrain), Monthly Business Ltr Rept for Oct 1987 ML20235R4161987-10-31031 October 1987 Draft Eg&G Technical Evaluation Rept for Plant Protective Sys Trip Setpoints for Fort St Vrain Nuclear Generating Station ML20235V8871987-10-31031 October 1987 Draft Plant Protective Sys Trip Setpoints for Fort St Vrain Nuclear Generating Station, Technical Evaluation Rept ML20236V3191987-10-31031 October 1987 Draft Plant Protective Sys Trip Setpoints for Fort St Vrain Nuclear Generating Station, Technical Evaluation Rept ML20236C2611987-09-28028 September 1987 Fort St Vrain Safe Shutdown Using Condensate Sys ML20236U7221987-08-0606 August 1987 Addendum to Review of Proposed Tech Spec Change:Core Inlet Valves/Min Helium Flow & Max Core Region Temp Rise, Limiting Conditon for Operation 4.1.9, Technical Evaluation Rept ML20246J3191987-07-31031 July 1987 Evaluation of Operator Roles in Mitigating High Energy Line Break at Fort St Vrain Nuclear Generating Station ML20235R6121987-07-0101 July 1987 Investigating Causes & Consequences of Cracked Graphite Fuel Elements ML20235M8321987-06-30030 June 1987 Conformance to Generic Ltr 83-28,Item 2.2.1 - Equipment Classification for All Other Safety-Related Components:Fort St Vrain, Final Informal Rept ML20215A9031987-06-0505 June 1987 Technical Assistance to Region IV (Fort St Vrain), Monthly Business Ltr Rept for Apr 1987 ML20235G1141987-06-0404 June 1987 Draft Fort St Vrain Safe Shutdown from 82% Power, Technical Evaluation Rept ML20234C6091987-06-0404 June 1987 Fort St Vrain Safe Shutdown from 82% Power, Technical Evaluation Rept ML20215L9331987-05-31031 May 1987 Evaluation of Integrated Sys Study of CRD Mechanism Rod Position Indication Instrumentation for Fort St Vrain Nuclear Generating Station, Informal Technical Evaluation Rept ML20214H2061987-05-15015 May 1987 Technical Assistance to Region IV (Fort St Vrain), Monthly Business Ltr Rept for Mar 1987 ML20215M3901987-04-30030 April 1987 Final Conformance to Generic Ltr 83-28,Item 2.2.2 - Vendor Interface Programs for All Other Safety-Related Components: Fort St Vrain, Informal Rept ML20210A3851987-04-13013 April 1987 Technical Assistance to Region IV (Fort St Vrain), Monthly Progress Rept for Feb 1987 ML20206H7461987-04-0808 April 1987 Notification of Contract Execution,Mod 1,to HTGR (Fort St Vrain) Training Course. Contractor:Ga Technologies ML20206H7621987-04-0808 April 1987 Mod 1,reflecting Administrative Changes Due to NRC Reorganization,To HTGR (Fort St Vrain) Training Course ML20214R8711987-03-31031 March 1987 Technical Evaluation Rept Evaluation of Bldg 10 Electrical Mods Fort St Vrain Nuclear Generating Station, Informal Rept ML20214Q5411987-03-31031 March 1987 Evaluation of Confinement Environ Temp Following High Energy Line Breaks Proposed for Fort St Vrain Environ Qualification Program, Informal Technical Evaluation Rept ML20207T3721987-03-10010 March 1987 Technical Assistance to Region IV (Fort St Vrain), Monthly Business Ltr Rept for Jan 1987 ML20207R9471987-03-0606 March 1987 Rev 1 to Review of Fort St Vrain Onsite AC `Standby' Power Sys Re Compliance to Single Failure Criterion & Ser ML20212F6771987-02-13013 February 1987 Technical Assistance to Region IV (Fort St Vrain), Monthly Progress Rept for Dec 1986 ML20216J1991987-02-0404 February 1987 Evaluation of Long-Term Effects of Moisture Ingress in Fort St Vrain Nuclear Reactor ML20210P6451987-01-31031 January 1987 Draft Evaluation of Confinement Environ Temps Following High Energy Line Breaks Proposed for Fort St Vrain Environ Qualification Program, Technical Evaluation Rept ML20245B3401986-12-31031 December 1986 Technical Assistance for Fort St Vrain Task Ii:Pcrv & Pcrv Tendon Evaluation, Final Technical Evaluation Rept ML20207K4701986-12-31031 December 1986 Final Review of Convection Heat Transfer Coefficients Utilized in Fort St Vrain Main Steam Line Break Analysis, Technical Evaluation Rept ML20212B7321986-11-30030 November 1986 Evaluation of Proposed CRD & Orifice Assembly 300 F Temp Limits,Fort St Vrain Nuclear Generating Station ML20197C5981986-10-31031 October 1986 Review of Convection Heat Transfer Coefficients Utilized in Fort St Vrain Main Steam Line Break Analyses, Technical Evaluation Rept ML20213D8251986-10-31031 October 1986 Technical Assistance to Region IV (Fort St Vrain), Monthly Progress Rept for Sept 1986 ML20214W4521986-10-16016 October 1986 Review of Proposed Tech Spec Change:Core Inlet Orifice Valves/Min Helium Flow & Max Core Region Temp Rise, Technical Evaluation Rept ML20214P4681986-09-30030 September 1986 Conformance to NRR Generic Ltrs 83-36 & 83-37 by Fort St Vrain Nuclear Generating Station, Informal Rept ML20211A6091986-09-26026 September 1986 Technical Assistance to Region IV (Fort St Vrain), Monthly Business Ltr Rept for Aug 1986 ML20236D3811986-09-24024 September 1986 Technical Evaluation of Dcrdr Info Supplementing Summary Rept for Fort St Vrain Nuclear Generating Station ML20209F1021986-09-0505 September 1986 Technical Assistance to Region IV (Fort St Vrain), Monthly Business Ltr for Jul 1986 ML20212Q1011986-08-31031 August 1986 Conformance to Reg Guide 1.97,Fort St Vrain Nuclear Generating Station, Informal Rept ML20213G5811986-08-0707 August 1986 Draft Review of Fort St Vrain Onsite AC 'Standby' Power Sys W/Regards to Compliance to Single Failure Criterion & Ser ML20205E3711986-07-31031 July 1986 Conformance to Generic Ltr 83-28,Item 2.1 (Part 1), Equipment Classification (Reactor Trip Sys Components) Fort Calhoun,Millstone Unit 2,Nine Mile Point 2 & Fort St Vrain, Technical Evaluation Rept 1997-03-31
[Table view] |
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o ENCLOSURE TECHNICAL EVALUATION REPORT FORT ST. VRAIN NUCLEAR GENERATING STATION
. DOCKET 50-267 .
LICENSEE: PUBLIC SERVICE CO. OF COLORADO REVIEW OF PROPOSED TECHNICAL SPECIFICATION CHANGE:
CORE INLET ORIFICE VALVES / MINIMUM HELIUM FLOW AND MAXIMUM CORE REGION TEMPERATURE RISE (L.C.O. 4.1.9) s PREPARED BY:
S. J. Ball Oak Ridge National Laboratory Oak Ridge, TN. 37831 l .. .
October 16.- 1986 NRC Lead Engineer: R. E. Ireland - RIV Project: ORNL Assistance in Evaluating Licensing Request-PSV LCO 4.1.9 (FIN A9351) l l
8612100231 DR 861205
( ADOCK 05000267 l PDR l
NOTICE This report was prepared as an account of work sponsored by an agency of the United Statec Government. Neither the United States Government nor any agency thereof, or any of their duployees, makes any warranty, expressed or implied, or assumed any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus, product or process disclosed in this report or represents that its use by such third party would not infringe privately owned rights.
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3 REVIEW OF PROPOSED TECHNICAL SPECIFICATION CHANGE:
CORE INLET ORIFICE VALVES / MINIMUM HELIUM FLOW AND MAXIMUM CORE REGION TEMPERATURE RISE (LCO 4.1.9) 2 ENTRODUCTION The objective of this task is to provide NRC Region IV with technical and analytical support in their evaluation of a request by Public Service Co. of Colorado (PSC) to amend the Fort St.
Vrain (FSV) High-Temperature Gas-Cooled Reactor (RTGR) Technical Specification - Limiting Condition for Operation (LCO) 4.1.9.
3 The intent of LCO 4.1.9 is to ensure that during low power and low flow operating conditions (0-25%), core region temperatures will be limited to acceptable maximum values. The major basis for the concern is that at low core flows (and hence low core pressure drops), the effects of higher buoyancy forces of the pressurized helium coolant channels may lead to flow stagnation and reversals-in some channels. The uncertainties of the region heat removal processes under these circumstances make it desirable to ensure that region flow stagnation and reversals do not occur. The objective of the original LCO 4.1.9 is to specify a set of conservative operating limits for both startup and shutdown, hot and cold, and pressurized and not. NRC, PSC, and GA Technologies (GAT) have all identified problems with the consistency, accuracy, and conservatism of the original and interin technical specifications. It was concluded that an independent analysis should be done to provide a basis for the licensing action required to resolve questions about the l
operating limitations. Resolution of acceptable operating limits may result in changes to the FSV Technical Specifications in order to ensure conservative thermal margins.
, APPROACH The approach taken to help resolve the questions raised in determining acceptable operating limits made use of an existing ORNL code (ORECA-FSV), which calculates the dynamic thermal hydraulic behavior of the FSV core (Ref.1). The problems of flow stagnation' and core overheating were explored for a variety of representative and conservative startup and shutdown scenarios, in some cases requiring that special routines be added to the code. The objective was to determine if LCO 4.1.9 and/or accompanying tech specs provided adequate protection for all forseeable circumstances of plant operation.
The FSV version of the ORNL ORECA code has been used extensively in code verification studies, and, in general, has shown good agreement with both FSV data and calculations by the GAT RECA code (Refs.2-3). In simulating typical startups, the i
ORECA calculation begins with n zero -power uni form t emperat ure core and follows user-input time-varying functions of total
T. .-
4 circulator' flow, thermal power, primary system pressure, and core inlet temperature. Guidelines for typical startup scenarios were initially obtained from the FSV DC-5-2 (Issue C) manual both for startup from refueling conditions and for startup with full helium inventory. Subsequently, plant data logger records were obtained from PSC to get representative information on startup and shutdown' operating procedures, to check data consistency, and to-determine how close the operating parameters approach the prescribed. limits (Ref.4). For shutdowns, the code requires inputs specifying the power and flow rundown conditions. In both
- cases, orifice manipulation routines are executed to go from
, approximately equal-flow to equal region temperature rise
- settings, or vice versa, at specified times. Other user inputs include the refueling region peaking factors and orifice positions and the various core and refueling region bypass flow s fractions.
A watchdog routine was added to ORECA to detect violations of the original LCO (both for LCO 4.1.9 Fig. I and 2 conditions),
, noting the beginning and ending times for the violations and, for 4
the Fig. 2 case, the value of the maximum region temperature rise. An additional watchdog routine was added to look for violations of LCO 4.1.7, which governs adjustments of the core inlet orifice valves, as this turned out to be more effective in limiting core temperatures in many cases than did LCO 4.1.9. .
The ORECA code includes a model of the dynamic response of
- the region outlet thermocouples, which have fairly long response times -- especially at the low flows associated with startup and 1
shutdown (Ref.5). Calculations for LCO 4.1. 9 of core thermal power and region temperature rises are made based on these simulated thermocouple measurements rather than " actual" region outlet temperatures, since the measurements are used by the j ,,
, operators to determine compliance.
Typical startup and shutdown runs were studied in some i detail. Magnetic tapes with plant data logger outputs for the
! November 3, 1983, startup and the January 17, 1984, shutdown were l adapted for use on ORNL computers, along with PSO's " HISTORY" i program for reading, deciphering, printing, and plotting the i
data. PSC also supplied calculated region peaking factors (RPFs) at crucial points in the runs so that the ORECA code could be set up to simulate the runs (Ref.4).
The ORECA code was set up and run for major " stopping
, points", and good agreement between the steady-state calculations and data was found. In each case the agreement was optimized by varying the assumed core bypass flow fraction, and the optimized bypass flows were well within expected ranges.
The PSC HISTORY code was modi fi ed to do further investigntions of possible problems with tech spec limitations.
5 The LCO 4.1.9 and 4.1.7 watchdog routines added to ORECA were adapted to HISTORY and run with the startup and shutdown data.
For LCO 4.1.9, flow " margins" (actual core flow / minimum allowable flow) are output when LCO 4.1.9 Fig. 1 (equal-flow orifice settings) is applicable, and region temperature rise " margins"
'(maximum allowable delta-T minus worst-case measured delta-T) are output when LCO 4.1.9 Fig. 2 (orifices anywhere) is applicable.
For LCO 4.1.7, the worst-case margin (maximum allowable region outlet temperature minus the worst-case measured outlet temperature) is calculated both for the startup case (average core Fig. 4.1.7-1.
outlet (950 F) and for the conditions (>950 F) specified by In the latter analysis, all region outlet temperature readings are taken at face value rather than using .
, comparison regions for some, as is done in more recent versions of the tech spec.
A major effort involved model and code development to include intra-region flow dynamics in the ORECA+ code. This i
addition was needed to determine the limits of flow stability and stagnation more precisely than was possible by representing each refueling region _by an average channel. The model that was
' developed simulates the two worst-case regions, using the computed overall core pressure drop, inlet plenum gas temperature, and power as inputs. Each of the two regions is represented by a single high-power density column (where the
" tilt" power factor may be as high as 1.6) in parallel with the average-power-density combination of the other six columns in the '
region. The common " upper plenum" for these two column models is the space just below the inlet flow orifice. As in the ORECA model, each column model is represented by six fuel, three reflector, and one core support axial nodes. Radial conduction is included but axial conduction is conservatively neglected.
The programs required for implementing this model were largely
. drawn from existing ORECA coutines. The three major advantages i ..
of this technique over the GAT steady-state models are the j ability to simulate the dynamic situation (which allows estimation of limits on allowable times out of compliance), the ability to use a dynamic overall core pressure drop driving function that is derived from a detailed 37-region calculation. ,
and the ability to determine if and how stagnation conditions
! could be corrected.
BENCHMARKING CALCULATIONS FOR INTRA-REGION FLOW LIMITING CASES
' Since the limiting factor in many low power operating regions is the calculated stagnation point for intra-region flow, i
and since the ORNL and GAT approaches to the problem were so different, a set of benchmark calculations were set up by Rick Kapernick of GAT. Some apparent discrepancies had appeared, as well, when in many cases the opernting limits dictated by GAT l analyses were more restrictive than ORNL's. It was decided that
(
i for all benchmark runs, there would be two worst-ense refueling
6 regions. The hotter region had a region peaking factor (RPF) of 3.0 and a column tilt factor of 1.17, while the other had an RPF of 1.6 and a tilt of 1.507. The assumed active core bypass fraction was 0.18625, and all region orifices were set at 20%
open. Runs were made at two different power levels, 1% and 5%,
'with a fixed helium inlet density corresponding to 250 F and 367 psia at 1% power, and 350 F and 419 psia at 5% power. Total reactor flows were varied between 1 and 10% for the 1* power case, and 5% and 13% for the 5% power case. The ORECA code, which normally includes intra-region conduction, was run both with and without it. (Conduction is not included in the GAT analyses, and this accounted for some discrepancies).
Several other important features of the analysis besides the yes/no determination of stagnant or reverse flow conditions in a region were noted. These are:
- 1) How hot does the fuel get, with or without stagnation?
- 2) How long does it take for the fuel to heat up and for the regions to stagnate?
- 3) Could the operator tell from region outlet temperature readings if the worst-case regions are in trouble, and are the tech spec limits on outlet temperature mismatches violated?
- 4) What are the effects on flow stagnation of setting the orifices for equal region outlet temperatures (rather than equal positions)? ~
- 5) How readily can stagnation conditions be remedied by
- increasing flow or adjusting orifice positions?
Also important is the fact that there are relatively large errors in measuring power and flow at the very low values of each; therefore allowances need to be made for these when operating limits are set.
The ORECA runs for the benchmark were set up with a high flow initially (10% for the 14 power cases and 13% for the 5%
power cases), and the flow was subsequently reduced to the lower limits. For the 14 power reference benchmark case, there was no problem with either total region or intra-region _f1_ow stagnation with flows as low as 5%. After about 4 hr operation at 4% flow, however, the intra-region flow in the region with the higher tilt factor stagnated. The measured gas outlet temperature of the region with the higher RPF, although not stagnated, exceeded the mismatch temperature limit (average + 400 F) after about 1.5 hr.
With 3% of full mass flow, flow stagnation / reversals occur right away, and the maximum fuel temperatures reached the 1600 C
" damage limit" after about 4 hr. After as long as about 7 hr, by adjusting the orifices using a simple algorithm that attempts to equalize outlet temperatures, the stagnation can be cleared, and the core conditions can be recovered to acceptable temperatures and flows. In another case in which the flow was reduced to 2%
(and stagnation / reversals occurred), a subsequent increase in the
2 i
4 7
flow to 5% cleared the stagnation. Runs in which the intra-region conduction was neglected showed some effect on the stagnation threshold. Typically about 1* nore flow was required to prevent stagnation without intra-region conduction than for cases with normal conduction included.
GAT's calculated minimum flow to prevent stagnation at 1%
power was 6.9%, vs. 5% per ORECA. The GAT analysis was done for 4
a higher density gas (107.5% vs. 90% inventory in the benchmark) and neglected conduction (a 14 flow effect), so these differences 4
could easily account for the discrepancy. A comparison of the predicted benchmark core conditions at 10% flow also showed no d
differences. .
A variation on the 1* power benchmark runs was made using ORECA for what we had judged (based on a GAT design support physics analysis) to be more realistic worst-case estimates for 1 RPFs and tilts, i.e. RPF = 1.80 and tilt = 1.36. These also
- showed intra-region flow reversals when the total flow was 4
reduced to 44. However, if the outlet temperature equalization scheme is used to reposition the orifices, the reversals didn't occur until the flow was reduced to about 24.
For the 5% power benchmark cases, excellent agreement between ORECA and GAT calculations was obtained for 134 flow, although ORECA showed an outlet temperature mismatch exceeding
- LCO 4.1.7 even for those conditions. For subsequent reductions s in flow, ORECA initially showed no reversals for core flows as low as 5% (7% neglecting conduction), although very high core and gas temperatures were calculated. The GAT analysis (at 107% vs.
904 inventory) gave 10.3% as the minimum flow to avoid stagnation. Further investigation, however, showed that ORECA
, probably would have eventually calculated flow stagnation at flows higher than 5-7%. The ORECA calculations had been i - ' .
terminated when outlet gas temperatures exceeded 3000 F and fuel temperatures exceeded 9000 F; gas flows in the critical channels were still' decreasing. Hence the apparent discrepancy between GAT and ORNL calculations at 5* power were judged not be
,' significant. It should be noted, however, that the use of a flow stability limit to prevent"high fuel temperatures under these
- circumstances is clearly not appropriate.
I For cases in which our "more realistic" worst-case RPFs and i
tilts were used. ORECA runs indicated that no stagnation occurred at 5% flow and that orifice manipulation was able to give reasonable core temperatures.
l Further variations on the benchmark runs were studies to investigate potential problems with the use of equal outlet temperature orifice settings at the low power / flow conditions.
At 1% power, for the minimum peaking factor region (RPF = 0.4 and
, column tilt = 1.507), which is limiting for the equal-outlet l
l r i1
-w ~ , - c.--e, err.ve,,ew.-e .,-r,,.,,---.--,.,m,--,r--_,-vw-.,w-4~-.--w-m-.r_,-% ,,,--.-,mvy,--,37-r--,,--, ,-,--n p.------,--
~
8 temperature mode, stagnation occurred at 44 flow. However, by simply limiting the minimum orifice setting to 10% open, flow stagnation did not occur until the flow was reduced to 24.
The conclusions drawn from the benchmarking exercise was
'that the ORNL and GAT predictions were consistent, and that limiting conditions predicted by GAT would be at least as conservative as ORNL's.
SUMMARY
OF ORNL RECOMMENDATIONS. COMMENTS. AND CONCERNS This task (A9351) was initiated in Dec. 1983 at the request of the FSV project manager, P. C. Wagner. During the course of the study, a large number of recommendations were made and concerns noted. Most of these have been resolved or accounted for either by discussions with PSC and NRC, by further analyses, or by eventual modifications to LCO 4.1.9 and supporting surveillance requirements. A list of meetings attended by ORNL is in Attachment 1. The following summary is a chronology of the more pertinent issues and questions raised, where the eventual 4
dispositions are labeled by:
(R) -
resolved satisfactorily; (0) - ORNL was overruled (recommendation not followed); or (F) - some followup work is still recommended. *
- 1. Low-flow and low-power measurements (F) -
The definitions ~for core flow and core power (which are used in the LCOs to determine operating limits) are not adequate at very low power and flow. For example, the tech spec does not specify how core flow is to be derived, and the various means for calculating it from the instrumentation available have given widely-varying estimates at low flow, especially when one or more circulators are shut down. We recommend using circulator speeds in conjunction with air culator performance maps. We also recommend incorporataug calculated afterheat in the power estimate when approaching or following shutdown.
~
- 2. " Gap" in coverage of operating conditions (R) -
In the figure limiting core outlet termpearture mismatches (for average T-out > 950 F), Fig. 3.2.2-1, there were gaps in the coverage for typical startup operations. These occurred when T-out > 950 F and the average temperature rise from the circulator inlet to core outlet was less than GGO F. From the data we observed, all the region outlet temperatures are typically low i enough so that there would not be a real problem with overheating the core.
- - - _ . - _ _ _ , _ _ _ - _ . _ . _ _ _ _ _ . ~ - _ , _ _ _ _ _ . _ , _ _ _ _ _ _ _ _ - - . _ _ _ _ _ _ _ _ _ _
9
- 3. " Mechanized Calculations" of optimum orifice positions (F)
We recommended that PSC should use a simple program to calculate optimum orifice positions for each desired operating condition rather than let the operators find them with a slowly converging iterative process.
- 4. References to " Sister" LCO 4.1.7 (0) -
Apparently the relationship between LCOs 4.1.7 and 4.1.9 has not been universally well-understood. The limits imposed by LCO 4.1.9 are calculated entirely on the basis of avoiding operating conditions in which refueling region (or sub region) flows would stagnate or reverse. The limits of LCO 4.1.7 are somehwhat more direct, as they relate the inferred maximum fuel temperature to measurements of region outlet coolant temperature. Both LCOs are needed to effectively limit maximum fuel temperatures. Simply assuring no stagnation (LCO 4.1.9) doesn't assure that fuel won't overheat. For LCO 4.1.7, if the flow is stagnated, the region outlet thermocouple may not be measuring a temperature that can be related to fuel temperature. Hence, we recommended a closer tie-in of LCO 4.1.7 (it is now referenced in the Basis section).
However, as an operator guide, we feel that cross-references should be spelled out clearly. For example, in Table 4.1.9-1, core temperature rises are not shown as being limited for equal region flow cases where the system pressure is >SO psia. We '
believe it should be noted in the table, or at least in a +
footnote, that these cases are to be limited by restrictions in LCO 4.1.7.
- 5. Surveillance vs. Corrective Action time requirements (H) -
The time limits are required corrective action and the surveillance time intervals are not consistent. For example, with surveillance required only every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, the alternative corrective actions are required in either 15 min or one hour.
(The 15-min limit implies an urgency not carried by the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> surveillance time interval.) PSC noted that operating procedures called for essentially continuous monitoring of the power, flow, * *
- and core outlet temperatures during orifice maneuvers. While this is a satisfactory response, we would prefer that reference to the more frequent monitoring be made in the LCO to assure the operator's understanding of the requirements. SR 4.1.~8 formerly~~
indicated only that flow was to be monitored continuously during power level changes. It has since been revised to require continuous monitoring during orifice maneuvers as well.
G. Questions on procedures for determining " bulk core temperature" limit of 760F (R)
These questions were addressed in detail and reported in our A9351 monthly repori for January 1986 (See attachment 2). We
~
10 recommended that reference be made in the Basis to the procedures used to calculate bulk core temperature.
- 7. Equal-Flow orifice range specification (R) -
The LCO did not specify a limiting range for orifice positions corresponding to the equal flow setting mode. We recommended 10-20% open. The opening size is crucial to the flow stagnation calculations.
- 8. Experimental Verification (0) -
While benchmarking exercises for GAT and ORNL flow-stagnation analyses showed good general agreement, more comfort could be derived from some good experimental confirmation data.
In simulations of numerous startup tests, it was observed that for wide variations around normal operating paths, very little flow redistribution occurred. Redistributions,'which are precursors to stagnation, were shown to be readily observable by measuring changes in outlet temperature dispersions due to changes in total primary loop flow. The tests which were recommended proposed relatively small flow perturbations (within current tech spec limits). These tests would provide data on -
whole-region, not intra-region, flow redistributions.
4
SUMMARY
AND CONCLUSIONS s.
The approach used in the review of the final and previous versions of the revised Tech Spec LCO 4.1.9 " Core inlet orifice valves / Minimum helium flow and maximum core region temperature rise" included the following:
- 1) Revise the ORNL ORECA (3-D FSV core thermal-hydraulics) code
'* as required to include intra-region flow and to simulate startups and shutdowns with both representative and conservative l assumptions. Benchmark calculations using ORECA and the GAT codes used to derive the limits employed in the new LCO showed good agreement.
- 2) Use FSV-supplied startup/ shutdown plant process computer
! data and PSC's " HISTORY" code to study both typical and conservative transients and to note operational problems.
- 3) Confirm that the revised Tech Spec meets its goals.
j 4) Point out problems and suggest alternatives.
As in the case of previous versions of LCO 4.1.9 the limits imposed by the final LCO (P-86451) range from equivalent to conservative as compared to t h a.s e derived by ORNL analyses.
Hence, the major concern of the Tech Spec, that of providing j limits that will in fact prevent core overheating, has been
11 addressed and confirmed satisfactorily. Remaining disagreements and concerns were primarily with details of clarity and style.
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o 12 LCO 4.1.9 (TER)
REFERENCES
- 1. S. J. Ball, ORECA-l? A Digital Computer Code for Simulating the Dynamics of HTGR Cores for Emergency Cooling Analyses, ORNL/TM-5159 (April 1976).
- 2. S. J. Ball, " Dynamic Model Verification Studies for the Thermal Response of the Fort St. Vrain HTGR Core",
Proceedings of the Fourth Power Plant Dynamics. Control and ~
Testing Symposium, Gatlinburg. Tennessee (March 1980). -
- 3. S. J. Ball, et al., High-Temperature Gas-Cooled Reactor Safety Studies for the Division of Reactor Safety Research Quarterly Progress Report. January 1-March 31, 1978, NUREG/CR-0179, ORNL/NUREG/TM-221 (July 1978).
- 4. Letter from D. W. Warenbourg (PSC) to S. J. Ball, " History
.. . Tape Transmittal for LCO 4.1.9 Evaluations", March 2, 1984 (P-84073).
- 5. S. J. Ball, et al., High-Temperature Gas-Cooled Reactor Safety Studies for the Division of Reactor Safety Research Quarterly Progress Report. October 1-December 31, 1978, NUREG/CR-0716 ORNL/NUREG/TM-314 (April 1979).
LCO 4.1.9 (TER)
ATTACHMENT 1 Meetings on LCO 4.1.9 Proposed Changes attended by ORhL s
- 1. August 23, 1984, at NRC-Region 4 Arlington, TX, with NRC, PSC, and GAT, to discuss the status of the review.
- 2. March 13, 1986, at NRC-Region 4 Arlington, TX, with NRC, PSC, and GAT, to address and resolve the oustanding issues on the most recent PSC drafts of LCO 4.1.9.
e ee m .m ee eeeen # o A-1
ATTACliMENT 2 LfJi,1 tJSilHL ( W .',',1 Monthly f<epor t + or Januar , 1*/ tit. PAGt: I REVIEW OF LCO 4.1.9 BASIS RELAllNG lu /69 F AVERAGE CURE IEMPERATURE LIM 11 The basis an the draft LCO 4.1.9 (P-85442) pertaintny to the 760 F mansmum core temperature lamat as as follows:
"The calculated bu l l- core temperature is the calculated average temperature of the core. Including graphite and fuel but not the reflector, that occurs following a loss of all forced circulation of primary coolant flow. The calculation assumes that all decay heat as retained in the core with no heat transfer to the reflector, PCRV anternals or primary coolant. If the decay heat as sufficiently low, with all primary coolant flow terminated, the calculated bulk core temperature will not exceed 760 degrees F, this specification as not applicable. Below this temperature, there is no damage to fuel or PCRV internal components."
The 760 F bulk core temperature limitation as proposed for use in this and other revised LCOs as a means of assuring that no fuel or PCRV-internals damage will occur for periods when no primary coolant forced circulation is available. The 760 F limit was apparently derived from the design value of core inlet temperature, which is 760 Ft hence there is no safety concern af a!! of the components in the reactor vessel are nominally Inmited to this temperature. Several' questions and concerns do arise, however, regarding the implementatiot of this Inmit to specific reactor operation scenarios:
- 1) How is the afterheat calculated?
- 2) How hot do critical PCRV-internal (metal) components get with a core average temperature of 760 F7
- 3) How conservative is the assumption that there as no heat loss to the surroundings during a no-flow heatup?
ITEM 1: The means of determining afterheat is not specified in the LCO, and it should be. Since there are no sensors in the core which can effectively measure the mean temperature, it is important to have an accurate estimate of afterheat. This can be a complex calculation for cases where the power level has undergone maior perturbattons in _ _.
id"2the period before shutdown. (We have developed such an algorithm that '
could be implemented on the plant computer or a programmable calculator.)
. _ . . .- - ~ ~
ITEM 2: A variety of heatup conditions were simulated using the severe-accident verston of the ORECA-FSV code, ranging from relatively rapid heatupu (6 days after 100% power operation or 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after 35%
power operation) to very slow heatups (1 year after 100% or 100 days after 35%). As expected. the nonuntformattes in PCRV temperatures were l arger in the faster heatuput however, at the time when the averaqo core temperature reached 760 F. mantmum fuel temperatures were well below normal operattnq temperatures and PChV metallic component A 1
ENCLOSURE - A9351 Nonthly Report for January 1986 PAGE 2 temperatures were well below 760 F. Hence we would conclude that the 760 F core average temperature limit is sufficiently conservative.
ITEN 3: Also of interest was the extent of the conservatism in the assumption that there is no heat loss from the core during thu heatup.
Again, by use of the ORECA-FSV code, it was shown that this conservatism is strongly dependent upon the heatup rate (the slower the rate the more conservative the assumption). The " actual" computed rates as percentages of the adiabatic heatup rates are shown for three representative cases in Fig. 1. The cases in which there is a relatively short time to restore circulation are also the cases with the least conservatism.
We would also recommend rewording the next-to-last sentence in the " basis". The way it currently reads, it implies that LCO 4.1.9 flow requirements are waived only in those cases where the afterheat is SO low that an adiabatic core would never reach 760 F.
In conclusion, it appears that the 760 F limitation is a sufficiently conservative means of protecting the' core and PCRV internals from damaget however, the means for Con,Juting the adiabatic temperature rises should be specified in the LCO or its basis.
I A 2-2 l
0 e 9
Figure 1 - Comparisons of Actual vs. Adiabatic Core Heatup Rates oRNL DWG 86C 6189 ETO
. l l l l u 100 - *- -
N'
- =E O u <
< 80 yt h
.K -
si y 60 -
= *. -
9 f 40 -
, h -
E .
20 -
t 0 ' I 300 400 500 600 700 800 '
AVERAGE 8ULK CORE TEMPERATURE (*F)
NOTES:
CURVE 1 -
(100% POWER + 6 DAYS OR 35% POWER + 3 HOURS).
HEATUP FROM 300*F TO 760*F IN 6 HOURS.
s.
.- CURVE 2 - *ME01UM HEATUP RATE *:
(100% POWER + 70 DAYS OR 35% POWER + 9 OAYS).
HEATUP FROM 300*F TO 760*F IN 22 HOURS.
CURVE 3 -
(100% POWER + 460 DAYS OR 35% POWER + 104 OAYS).
HEATUP FROM 300*F TO 760*F IN 5.5 DAYS.
A 2-3
T utto in1 b
OAK RIDGE NATIONAL LABORATORY eost orricE som v OPERATED ey taantsN edAl'NETTA ENERGv Sv$tEMS PeC October 16, 1986 Mr. Richard E. Ireland U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011
Dear Dick:
Subject:
Submittal of TER on FSV LCO 4.1.9 (FIN A 9351)
The final version of our Technical Evaluation Report (TER) on the Fort St. Vrain LCO 4.1.9 is enclosed for your review and approval. Please call me if you have any questions. s Yours truly, S. J. Ball, Manager HTGR Safety Studies for
. . . NRC SJB:djs Enclosure cc: .K. L. Heitner - NRC A. P. Malinauskas D. L. Moses
.