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Category:CONTRACTED REPORT - RTA
MONTHYEARML20203K4121998-02-28028 February 1998 H. B. ROBINSON-2 Pressure Vessel Benchmark ML20236B8051989-03-13013 March 1989 Monthly Progress Rept for Contract NRC-02-87-004 for Period 890204-890303.Financial Data Available in Central Files ML20237J1411987-08-17017 August 1987 Facility Sys Analysis Support, Progress Rept 23 for 870713 -0807 ML20211N5031987-01-0909 January 1987 Reactor Trip Sys Reliability Conformance to Item 4.5.2 of Generic Ltr 83-28,HB Robinson Steam Electric Plant,Unit 2, Salem Generating Station Units 1 & 2,Shearon Harris Nuclear Power Plant Unit 1..., Technical Evaluation Rept ML20210M4571986-09-30030 September 1986 Lepricon Analysis of Pressure Vessel Surveillance Dosimetry Inserted Into H.B. ROBINSON-2 During Cycle 9 ML20215M2351986-08-21021 August 1986 Errata to NUREG/CR-4439, Lepricon Analysis of Pressure Vessel Surveillance Dosimetry Inserted Into Hb Robinson-2 During Cycle 2, Correcting Pages ML20140G6501985-12-31031 December 1985 Review of RELAP5 Calculations for H.B. Robinson Unit 2 Pressurized Thermal Shock Study ML20151U3481985-10-15015 October 1985 Comparison of Plant-Specific Analyses of Pressurized Thermal Shock (Based on Oconee,Calvert Cliffs & Hb Robinson), Technical Ltr Rept ML20137M6621985-09-30030 September 1985 Pressurized Thermal Shock Evaluation of the H.B. Robinson Unit 2 Nuclear Power Plant ML20138J9491985-09-30030 September 1985 Conformance to Generic Ltr 83-28,Items 3.1.3 & 3.2.3, Farley Units 1 & 2,Turkey Point Units 3 & 4,Robinson Unit 2 & Summer Unit 1 ML20140C5621985-09-30030 September 1985 Pressurized Thermal Shock Evaluation of the H.B. Robinson Unit 2 Nuclear Power Plant ML20128M0201985-07-31031 July 1985 THERMAL-HYDRAULIC Analyses of Overcooling Sequences for the H.B. Robinson Unit 2 Pressurized Thermal Shock Study ML20126D0541985-04-30030 April 1985 RELAP5 THERMAL-HYDRAULIC Analyses of Pressurized Thermal Shock Sequences for H.B. Robinson Unit 2 Pressurized Water Reactor ML20116C4531985-04-15015 April 1985 Evaluation of Detailed Control Room Design Review Program Plan for Hb Robinson Steam Electric Plant,Unit 2 ML20112H7951984-12-26026 December 1984 Draft Chapter 5, Pressurized Thermal Shock Evaluation of Hb Robinson Unit 2 Nuclear Power Plant, Re Probabilistic Fracture Mechanics Analysis of Potential Overcooling Sequences ML20101Q3211984-11-21021 November 1984 Draft Chapter 3, Development of Overcooling Sequences for Hb Robinson Unit 2 Nuclear Power Plant. to Pressurized Thermal Shock Evaluation of Hb Robinson Unit 2 Nuclear Power Plant ML20101Q3081984-11-20020 November 1984 Draft App B, Pressurized Thermal Shock (PTS) Initiating Event Frequency & Branch Probability Screening Estimates for Hb Robinson Unit 2 Nuclear Power Station, to PTS Evaluation of Hb Robinson Unit 2 Nuclear Power Plant ML20106A6571984-07-31031 July 1984 Inservice Insp Program Technical Evaluation Rept ML20091J5591984-05-0707 May 1984 Control of Heavy Loads (C-10),CP&L,HB Robinson Unit 2, Technical Evaluation Rept ML20077R5531983-08-31031 August 1983 Concentration and Distribution of Copper in Effluents from the H.B. Robinson Steam Electric Plant ML20080E0391983-05-31031 May 1983 Generator Overfill Transient Failure Mode & Effects Analysis & Rejected Sys Justification Rept, Preliminary Rept ML20080E0601983-05-31031 May 1983 Overcooling Transient Failure Mode & Effects Analysis & Rejected Sys Justification Rept, Preliminary Rept ML20126E9941982-12-0707 December 1982 Containment Leak Rate Testing Investigations, Monthly Progress Rept for Nov 1982 ML20070D1221982-11-30030 November 1982 Concentrations of COPPER-BINDING Proteins in Livers of Bluegills from the Cooling Lake at the H.B. Robinson Nuclear Power Station ML20066G3171982-11-18018 November 1982 PWR Main Steam Line Break W/Continued Feedwater Addition (B-69),HB Robinson Unit 2, Technical Evaluation Rept ML20077K0291982-09-30030 September 1982 Inservice Insp Program, Technical Evaluation Rept ML20023B0131982-08-31031 August 1982 Tech Specs for Redundant DHR Capability,Hb Robinson Steam Electric Plant,Unit 2. ML20028C6191982-07-31031 July 1982 Evaluation of Failed Reactor Coolant Pump Internal Bolts from Hb Robinson Nuclear Power Station, Informal Rept ML20079G5031982-07-0808 July 1982 Review of Licensee Resolution of Outstanding Issues from NRC Equipment Environ Qualification Safety Evaluation Repts (F-11 & B-60), Technical Evaluation Rept ML20079F2921982-03-31031 March 1982 Override & Reset of Control Ciruitry in Ventilation/Purge Isolation & Other ESF Sys,Hb Robinson Unit 2, Technical Evaluation Rept ML20052D0181982-02-28028 February 1982 Tech Specs for Redundant Decay Heat Removal Capability,Hb Robinson Steam Electric Plant,Unit 2, Preliminary Technical Evaluation Rept ML14190A6291981-07-28028 July 1981 Containment Leakage Rate Testing. ML19309C6161980-02-29029 February 1980 Electrical,Instrumentation & Control Aspects of Proposed Tech Specs Change Re Blocking of Safety Injection Signals When Pressure Is Below 2000 Psig, Technical Evaluation Rept ML20086C3771971-06-23023 June 1971 Rept on Insp of Leaks & Cracks in (a) Steam Generator at Hb Robinson II ML20087B9581971-06-23023 June 1971 Rept on Insp of Leaks & Cracks in (a) Steam Generator at Hb Robinson II ML20086B5241970-07-10010 July 1970 Stress Analysis of Modified Main Steam Safety Valve Nozzle & Adjacent Steamline ML20086B5731970-05-23023 May 1970 Rept on Design & Stress Evaluation of Failed Main Steam Safety Valve Nozzle 1998-02-28
[Table view] Category:QUICK LOOK
MONTHYEARML20203K4121998-02-28028 February 1998 H. B. ROBINSON-2 Pressure Vessel Benchmark ML20236B8051989-03-13013 March 1989 Monthly Progress Rept for Contract NRC-02-87-004 for Period 890204-890303.Financial Data Available in Central Files ML20237J1411987-08-17017 August 1987 Facility Sys Analysis Support, Progress Rept 23 for 870713 -0807 ML20211N5031987-01-0909 January 1987 Reactor Trip Sys Reliability Conformance to Item 4.5.2 of Generic Ltr 83-28,HB Robinson Steam Electric Plant,Unit 2, Salem Generating Station Units 1 & 2,Shearon Harris Nuclear Power Plant Unit 1..., Technical Evaluation Rept ML20210M4571986-09-30030 September 1986 Lepricon Analysis of Pressure Vessel Surveillance Dosimetry Inserted Into H.B. ROBINSON-2 During Cycle 9 ML20215M2351986-08-21021 August 1986 Errata to NUREG/CR-4439, Lepricon Analysis of Pressure Vessel Surveillance Dosimetry Inserted Into Hb Robinson-2 During Cycle 2, Correcting Pages ML20140G6501985-12-31031 December 1985 Review of RELAP5 Calculations for H.B. Robinson Unit 2 Pressurized Thermal Shock Study ML20151U3481985-10-15015 October 1985 Comparison of Plant-Specific Analyses of Pressurized Thermal Shock (Based on Oconee,Calvert Cliffs & Hb Robinson), Technical Ltr Rept ML20137M6621985-09-30030 September 1985 Pressurized Thermal Shock Evaluation of the H.B. Robinson Unit 2 Nuclear Power Plant ML20138J9491985-09-30030 September 1985 Conformance to Generic Ltr 83-28,Items 3.1.3 & 3.2.3, Farley Units 1 & 2,Turkey Point Units 3 & 4,Robinson Unit 2 & Summer Unit 1 ML20140C5621985-09-30030 September 1985 Pressurized Thermal Shock Evaluation of the H.B. Robinson Unit 2 Nuclear Power Plant ML20128M0201985-07-31031 July 1985 THERMAL-HYDRAULIC Analyses of Overcooling Sequences for the H.B. Robinson Unit 2 Pressurized Thermal Shock Study ML20126D0541985-04-30030 April 1985 RELAP5 THERMAL-HYDRAULIC Analyses of Pressurized Thermal Shock Sequences for H.B. Robinson Unit 2 Pressurized Water Reactor ML20116C4531985-04-15015 April 1985 Evaluation of Detailed Control Room Design Review Program Plan for Hb Robinson Steam Electric Plant,Unit 2 ML20112H7951984-12-26026 December 1984 Draft Chapter 5, Pressurized Thermal Shock Evaluation of Hb Robinson Unit 2 Nuclear Power Plant, Re Probabilistic Fracture Mechanics Analysis of Potential Overcooling Sequences ML20101Q3211984-11-21021 November 1984 Draft Chapter 3, Development of Overcooling Sequences for Hb Robinson Unit 2 Nuclear Power Plant. to Pressurized Thermal Shock Evaluation of Hb Robinson Unit 2 Nuclear Power Plant ML20101Q3081984-11-20020 November 1984 Draft App B, Pressurized Thermal Shock (PTS) Initiating Event Frequency & Branch Probability Screening Estimates for Hb Robinson Unit 2 Nuclear Power Station, to PTS Evaluation of Hb Robinson Unit 2 Nuclear Power Plant ML20106A6571984-07-31031 July 1984 Inservice Insp Program Technical Evaluation Rept ML20091J5591984-05-0707 May 1984 Control of Heavy Loads (C-10),CP&L,HB Robinson Unit 2, Technical Evaluation Rept ML20077R5531983-08-31031 August 1983 Concentration and Distribution of Copper in Effluents from the H.B. Robinson Steam Electric Plant ML20080E0391983-05-31031 May 1983 Generator Overfill Transient Failure Mode & Effects Analysis & Rejected Sys Justification Rept, Preliminary Rept ML20080E0601983-05-31031 May 1983 Overcooling Transient Failure Mode & Effects Analysis & Rejected Sys Justification Rept, Preliminary Rept ML20126E9941982-12-0707 December 1982 Containment Leak Rate Testing Investigations, Monthly Progress Rept for Nov 1982 ML20070D1221982-11-30030 November 1982 Concentrations of COPPER-BINDING Proteins in Livers of Bluegills from the Cooling Lake at the H.B. Robinson Nuclear Power Station ML20066G3171982-11-18018 November 1982 PWR Main Steam Line Break W/Continued Feedwater Addition (B-69),HB Robinson Unit 2, Technical Evaluation Rept ML20077K0291982-09-30030 September 1982 Inservice Insp Program, Technical Evaluation Rept ML20023B0131982-08-31031 August 1982 Tech Specs for Redundant DHR Capability,Hb Robinson Steam Electric Plant,Unit 2. ML20028C6191982-07-31031 July 1982 Evaluation of Failed Reactor Coolant Pump Internal Bolts from Hb Robinson Nuclear Power Station, Informal Rept ML20079G5031982-07-0808 July 1982 Review of Licensee Resolution of Outstanding Issues from NRC Equipment Environ Qualification Safety Evaluation Repts (F-11 & B-60), Technical Evaluation Rept ML20079F2921982-03-31031 March 1982 Override & Reset of Control Ciruitry in Ventilation/Purge Isolation & Other ESF Sys,Hb Robinson Unit 2, Technical Evaluation Rept ML20052D0181982-02-28028 February 1982 Tech Specs for Redundant Decay Heat Removal Capability,Hb Robinson Steam Electric Plant,Unit 2, Preliminary Technical Evaluation Rept ML14190A6291981-07-28028 July 1981 Containment Leakage Rate Testing. ML19309C6161980-02-29029 February 1980 Electrical,Instrumentation & Control Aspects of Proposed Tech Specs Change Re Blocking of Safety Injection Signals When Pressure Is Below 2000 Psig, Technical Evaluation Rept ML20086C3771971-06-23023 June 1971 Rept on Insp of Leaks & Cracks in (a) Steam Generator at Hb Robinson II ML20087B9581971-06-23023 June 1971 Rept on Insp of Leaks & Cracks in (a) Steam Generator at Hb Robinson II ML20086B5241970-07-10010 July 1970 Stress Analysis of Modified Main Steam Safety Valve Nozzle & Adjacent Steamline ML20086B5731970-05-23023 May 1970 Rept on Design & Stress Evaluation of Failed Main Steam Safety Valve Nozzle 1998-02-28
[Table view] Category:ETC. (PERIODIC
MONTHYEARML20203K4121998-02-28028 February 1998 H. B. ROBINSON-2 Pressure Vessel Benchmark ML20236B8051989-03-13013 March 1989 Monthly Progress Rept for Contract NRC-02-87-004 for Period 890204-890303.Financial Data Available in Central Files ML20237J1411987-08-17017 August 1987 Facility Sys Analysis Support, Progress Rept 23 for 870713 -0807 ML20211N5031987-01-0909 January 1987 Reactor Trip Sys Reliability Conformance to Item 4.5.2 of Generic Ltr 83-28,HB Robinson Steam Electric Plant,Unit 2, Salem Generating Station Units 1 & 2,Shearon Harris Nuclear Power Plant Unit 1..., Technical Evaluation Rept ML20210M4571986-09-30030 September 1986 Lepricon Analysis of Pressure Vessel Surveillance Dosimetry Inserted Into H.B. ROBINSON-2 During Cycle 9 ML20215M2351986-08-21021 August 1986 Errata to NUREG/CR-4439, Lepricon Analysis of Pressure Vessel Surveillance Dosimetry Inserted Into Hb Robinson-2 During Cycle 2, Correcting Pages ML20140G6501985-12-31031 December 1985 Review of RELAP5 Calculations for H.B. Robinson Unit 2 Pressurized Thermal Shock Study ML20151U3481985-10-15015 October 1985 Comparison of Plant-Specific Analyses of Pressurized Thermal Shock (Based on Oconee,Calvert Cliffs & Hb Robinson), Technical Ltr Rept ML20137M6621985-09-30030 September 1985 Pressurized Thermal Shock Evaluation of the H.B. Robinson Unit 2 Nuclear Power Plant ML20138J9491985-09-30030 September 1985 Conformance to Generic Ltr 83-28,Items 3.1.3 & 3.2.3, Farley Units 1 & 2,Turkey Point Units 3 & 4,Robinson Unit 2 & Summer Unit 1 ML20140C5621985-09-30030 September 1985 Pressurized Thermal Shock Evaluation of the H.B. Robinson Unit 2 Nuclear Power Plant ML20128M0201985-07-31031 July 1985 THERMAL-HYDRAULIC Analyses of Overcooling Sequences for the H.B. Robinson Unit 2 Pressurized Thermal Shock Study ML20126D0541985-04-30030 April 1985 RELAP5 THERMAL-HYDRAULIC Analyses of Pressurized Thermal Shock Sequences for H.B. Robinson Unit 2 Pressurized Water Reactor ML20116C4531985-04-15015 April 1985 Evaluation of Detailed Control Room Design Review Program Plan for Hb Robinson Steam Electric Plant,Unit 2 ML20112H7951984-12-26026 December 1984 Draft Chapter 5, Pressurized Thermal Shock Evaluation of Hb Robinson Unit 2 Nuclear Power Plant, Re Probabilistic Fracture Mechanics Analysis of Potential Overcooling Sequences ML20101Q3211984-11-21021 November 1984 Draft Chapter 3, Development of Overcooling Sequences for Hb Robinson Unit 2 Nuclear Power Plant. to Pressurized Thermal Shock Evaluation of Hb Robinson Unit 2 Nuclear Power Plant ML20101Q3081984-11-20020 November 1984 Draft App B, Pressurized Thermal Shock (PTS) Initiating Event Frequency & Branch Probability Screening Estimates for Hb Robinson Unit 2 Nuclear Power Station, to PTS Evaluation of Hb Robinson Unit 2 Nuclear Power Plant ML20106A6571984-07-31031 July 1984 Inservice Insp Program Technical Evaluation Rept ML20091J5591984-05-0707 May 1984 Control of Heavy Loads (C-10),CP&L,HB Robinson Unit 2, Technical Evaluation Rept ML20077R5531983-08-31031 August 1983 Concentration and Distribution of Copper in Effluents from the H.B. Robinson Steam Electric Plant ML20080E0391983-05-31031 May 1983 Generator Overfill Transient Failure Mode & Effects Analysis & Rejected Sys Justification Rept, Preliminary Rept ML20080E0601983-05-31031 May 1983 Overcooling Transient Failure Mode & Effects Analysis & Rejected Sys Justification Rept, Preliminary Rept ML20126E9941982-12-0707 December 1982 Containment Leak Rate Testing Investigations, Monthly Progress Rept for Nov 1982 ML20070D1221982-11-30030 November 1982 Concentrations of COPPER-BINDING Proteins in Livers of Bluegills from the Cooling Lake at the H.B. Robinson Nuclear Power Station ML20066G3171982-11-18018 November 1982 PWR Main Steam Line Break W/Continued Feedwater Addition (B-69),HB Robinson Unit 2, Technical Evaluation Rept ML20077K0291982-09-30030 September 1982 Inservice Insp Program, Technical Evaluation Rept ML20023B0131982-08-31031 August 1982 Tech Specs for Redundant DHR Capability,Hb Robinson Steam Electric Plant,Unit 2. ML20028C6191982-07-31031 July 1982 Evaluation of Failed Reactor Coolant Pump Internal Bolts from Hb Robinson Nuclear Power Station, Informal Rept ML20079G5031982-07-0808 July 1982 Review of Licensee Resolution of Outstanding Issues from NRC Equipment Environ Qualification Safety Evaluation Repts (F-11 & B-60), Technical Evaluation Rept ML20079F2921982-03-31031 March 1982 Override & Reset of Control Ciruitry in Ventilation/Purge Isolation & Other ESF Sys,Hb Robinson Unit 2, Technical Evaluation Rept ML20052D0181982-02-28028 February 1982 Tech Specs for Redundant Decay Heat Removal Capability,Hb Robinson Steam Electric Plant,Unit 2, Preliminary Technical Evaluation Rept ML14190A6291981-07-28028 July 1981 Containment Leakage Rate Testing. ML19309C6161980-02-29029 February 1980 Electrical,Instrumentation & Control Aspects of Proposed Tech Specs Change Re Blocking of Safety Injection Signals When Pressure Is Below 2000 Psig, Technical Evaluation Rept ML20086C3771971-06-23023 June 1971 Rept on Insp of Leaks & Cracks in (a) Steam Generator at Hb Robinson II ML20087B9581971-06-23023 June 1971 Rept on Insp of Leaks & Cracks in (a) Steam Generator at Hb Robinson II ML20086B5241970-07-10010 July 1970 Stress Analysis of Modified Main Steam Safety Valve Nozzle & Adjacent Steamline ML20086B5731970-05-23023 May 1970 Rept on Design & Stress Evaluation of Failed Main Steam Safety Valve Nozzle 1998-02-28
[Table view] Category:RESEARCH AND TECHNICAL ASSISTANCE REPORTS
MONTHYEARML20023B0131982-08-31031 August 1982 Tech Specs for Redundant DHR Capability,Hb Robinson Steam Electric Plant,Unit 2. ML20028C6191982-07-31031 July 1982 Evaluation of Failed Reactor Coolant Pump Internal Bolts from Hb Robinson Nuclear Power Station, Informal Rept ML20052D0181982-02-28028 February 1982 Tech Specs for Redundant Decay Heat Removal Capability,Hb Robinson Steam Electric Plant,Unit 2, Preliminary Technical Evaluation Rept ML19309C6161980-02-29029 February 1980 Electrical,Instrumentation & Control Aspects of Proposed Tech Specs Change Re Blocking of Safety Injection Signals When Pressure Is Below 2000 Psig, Technical Evaluation Rept 1982-08-31
[Table view] Category:TEXT-PROCUREMENT & CONTRACTS
MONTHYEARML20203K4121998-02-28028 February 1998 H. B. ROBINSON-2 Pressure Vessel Benchmark ML20236B8051989-03-13013 March 1989 Monthly Progress Rept for Contract NRC-02-87-004 for Period 890204-890303.Financial Data Available in Central Files ML20237J1411987-08-17017 August 1987 Facility Sys Analysis Support, Progress Rept 23 for 870713 -0807 ML20211N5031987-01-0909 January 1987 Reactor Trip Sys Reliability Conformance to Item 4.5.2 of Generic Ltr 83-28,HB Robinson Steam Electric Plant,Unit 2, Salem Generating Station Units 1 & 2,Shearon Harris Nuclear Power Plant Unit 1..., Technical Evaluation Rept ML20210M4571986-09-30030 September 1986 Lepricon Analysis of Pressure Vessel Surveillance Dosimetry Inserted Into H.B. ROBINSON-2 During Cycle 9 ML20215M2351986-08-21021 August 1986 Errata to NUREG/CR-4439, Lepricon Analysis of Pressure Vessel Surveillance Dosimetry Inserted Into Hb Robinson-2 During Cycle 2, Correcting Pages ML20140G6501985-12-31031 December 1985 Review of RELAP5 Calculations for H.B. Robinson Unit 2 Pressurized Thermal Shock Study ML20151U3481985-10-15015 October 1985 Comparison of Plant-Specific Analyses of Pressurized Thermal Shock (Based on Oconee,Calvert Cliffs & Hb Robinson), Technical Ltr Rept ML20137M6621985-09-30030 September 1985 Pressurized Thermal Shock Evaluation of the H.B. Robinson Unit 2 Nuclear Power Plant ML20138J9491985-09-30030 September 1985 Conformance to Generic Ltr 83-28,Items 3.1.3 & 3.2.3, Farley Units 1 & 2,Turkey Point Units 3 & 4,Robinson Unit 2 & Summer Unit 1 ML20140C5621985-09-30030 September 1985 Pressurized Thermal Shock Evaluation of the H.B. Robinson Unit 2 Nuclear Power Plant ML20128M0201985-07-31031 July 1985 THERMAL-HYDRAULIC Analyses of Overcooling Sequences for the H.B. Robinson Unit 2 Pressurized Thermal Shock Study ML20126D0541985-04-30030 April 1985 RELAP5 THERMAL-HYDRAULIC Analyses of Pressurized Thermal Shock Sequences for H.B. Robinson Unit 2 Pressurized Water Reactor ML20116C4531985-04-15015 April 1985 Evaluation of Detailed Control Room Design Review Program Plan for Hb Robinson Steam Electric Plant,Unit 2 ML20112H7951984-12-26026 December 1984 Draft Chapter 5, Pressurized Thermal Shock Evaluation of Hb Robinson Unit 2 Nuclear Power Plant, Re Probabilistic Fracture Mechanics Analysis of Potential Overcooling Sequences ML20101Q3211984-11-21021 November 1984 Draft Chapter 3, Development of Overcooling Sequences for Hb Robinson Unit 2 Nuclear Power Plant. to Pressurized Thermal Shock Evaluation of Hb Robinson Unit 2 Nuclear Power Plant ML20101Q3081984-11-20020 November 1984 Draft App B, Pressurized Thermal Shock (PTS) Initiating Event Frequency & Branch Probability Screening Estimates for Hb Robinson Unit 2 Nuclear Power Station, to PTS Evaluation of Hb Robinson Unit 2 Nuclear Power Plant ML20106A6571984-07-31031 July 1984 Inservice Insp Program Technical Evaluation Rept ML20091J5591984-05-0707 May 1984 Control of Heavy Loads (C-10),CP&L,HB Robinson Unit 2, Technical Evaluation Rept ML20077R5531983-08-31031 August 1983 Concentration and Distribution of Copper in Effluents from the H.B. Robinson Steam Electric Plant ML20080E0391983-05-31031 May 1983 Generator Overfill Transient Failure Mode & Effects Analysis & Rejected Sys Justification Rept, Preliminary Rept ML20080E0601983-05-31031 May 1983 Overcooling Transient Failure Mode & Effects Analysis & Rejected Sys Justification Rept, Preliminary Rept ML20126E9941982-12-0707 December 1982 Containment Leak Rate Testing Investigations, Monthly Progress Rept for Nov 1982 ML20070D1221982-11-30030 November 1982 Concentrations of COPPER-BINDING Proteins in Livers of Bluegills from the Cooling Lake at the H.B. Robinson Nuclear Power Station ML20066G3171982-11-18018 November 1982 PWR Main Steam Line Break W/Continued Feedwater Addition (B-69),HB Robinson Unit 2, Technical Evaluation Rept ML20077K0291982-09-30030 September 1982 Inservice Insp Program, Technical Evaluation Rept ML20023B0131982-08-31031 August 1982 Tech Specs for Redundant DHR Capability,Hb Robinson Steam Electric Plant,Unit 2. ML20028C6191982-07-31031 July 1982 Evaluation of Failed Reactor Coolant Pump Internal Bolts from Hb Robinson Nuclear Power Station, Informal Rept ML20079G5031982-07-0808 July 1982 Review of Licensee Resolution of Outstanding Issues from NRC Equipment Environ Qualification Safety Evaluation Repts (F-11 & B-60), Technical Evaluation Rept ML20079F2921982-03-31031 March 1982 Override & Reset of Control Ciruitry in Ventilation/Purge Isolation & Other ESF Sys,Hb Robinson Unit 2, Technical Evaluation Rept ML20052D0181982-02-28028 February 1982 Tech Specs for Redundant Decay Heat Removal Capability,Hb Robinson Steam Electric Plant,Unit 2, Preliminary Technical Evaluation Rept ML14190A6291981-07-28028 July 1981 Containment Leakage Rate Testing. ML19309C6161980-02-29029 February 1980 Electrical,Instrumentation & Control Aspects of Proposed Tech Specs Change Re Blocking of Safety Injection Signals When Pressure Is Below 2000 Psig, Technical Evaluation Rept ML20086C3771971-06-23023 June 1971 Rept on Insp of Leaks & Cracks in (a) Steam Generator at Hb Robinson II ML20087B9581971-06-23023 June 1971 Rept on Insp of Leaks & Cracks in (a) Steam Generator at Hb Robinson II ML20086B5241970-07-10010 July 1970 Stress Analysis of Modified Main Steam Safety Valve Nozzle & Adjacent Steamline ML20086B5731970-05-23023 May 1970 Rept on Design & Stress Evaluation of Failed Main Steam Safety Valve Nozzle 1998-02-28
[Table view] |
Text
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EGG-EA-5981
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TECHNICAL SPECIFICATIONS FOR REDUNCANT DECAY HEAT A.t'oA REMOVAL CAPABILITY, H. B. ROBINSON STEAM ELECTRIC ggf g PLANT, UNIT NO. 2 M
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J. A. Steverson '
L. J. Vogel .
idaho National Engineering Laboratory l Operated by the U.S. Department of Energy
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This is an informal report intended for uso as a prelirninary or working document ._
Prepared for the 3 U.S. Nuclear Regulatory Comission "
Under DOE Contract No. DE-AC07-76ID01570 - 9 FIN No. A6429 I y 's ,
8211010470 820831 pg b b E b Idaho l PDR
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FORM EG&G 398 -
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s INTERIM. REPORT Accession No.
Report No. EGG-EA-5981 Contract Program or Project
Title:
Selected Operating Reactors Issues-Subject of this Document:
Technical Specifications for Redundant Decay Heat Removal Capability, H. B. Robinson Steam Electric Plant, Unit No. 2 Type of Documer.t:
Inforre t Author (s):
J. A. Steverson L. J. Vogel Date of Document:
August 1982 ,
Responsible NRC Individual and NRC Office or Division:
J. N. Donohew, Division of Licensing This document was prepared primarily lor preliminary or internal use. it has not received
~
full review and approval. Since there may be substantive changes, this document should not be considered final.
EG&G Idaho, Inc.
Idaho Falls, Idaho 83415
- s 1 -N Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.
Under DOE Contract No. DE-AC07-761D01570 NRC FIN No. A6429 ,
INTERIM REP, ORT 1
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. TECHNICAL SPECIFICAT.JONS FOR REDUNDANT DECAY HEAT H. B. R03kHSON STEAM ELECTRIC PLANT, UNIT N0. 2
e' i August 1982 J. A. Steverson L. J. Vogel Reliability and Statistics Branch Engineering Analysis Division ,
, EG&G Idaho, Inc.
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TAC No. 42099 Docket No. 50-261
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e ABSTRACT This report reviews the H.;B. Robinson Steam Electric Plant Unit No. 2 proposed technical' specifications for redundancy in decay heat removal cap-ability in all modes of operation. .
FOREWORD This report is supplied as part of the " Selected Operating Reactors Issues" being conducted for the U.S. Nuclear Regulatory Comission, Office of Nuclear Reactor Regulation, Division of Licensing by EG&G Idaho, Inc.,
Reliability and Statistics Branch.
The U.S. Nuclear Regulatory Commission funded the work under the authorization, B&R 20 19 01.06, FIN No. A6429.
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a e
i Selected Operating Reactors Issues j . FIN No. A6429-
<- ii 1
4 CONTENTS
. ABSTRACT ............................................................. 11 FOREWORD ............................................................. 11
1.0 INTRODUCTION
.................................................... 1
, 2.0 DISCUSSION ...................................................... 1 i
3.0 CONCLUSION
S ..................................................... 3
4.0 REFERENCES
...................................................... 3 APPENDIX A--MODEL TECHNICAL SPECIFICATIONS FOR REDUNDANT DECAY HEAT REMOVAL FOR WESTINGHOUSE PRESSURIZED WATER REACT 0RS......................................................... 5 Y
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.4; I
iii
TECHNICAL SPECIFICATIONS FOR REDUNDANT DECAY HEAT REMOVAL CAPABILITY H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 3 INTRODUCTION A number of events have occurred at operating PWR facilities where e decay heat removal capability has been seriously degraded due to inadequate administrative controls during shutdown modes of One of these events,describedinIEInformationNotice80-20,gperation. occurred at t Besse, Unit No. I plant on April 19, 1980. InIEBulletin80-12geDavis-dated May 9, 1980, licensees were requested to immediately implement administra-tive controls which would ensure that proper means are available to provide redundant methods of decay heat removal. While the function of the bulletin was to effect immediate action with regard to this problem, the NRC con-sidered it necessary that an amendment of each license be made to provide forpermanentlong-termassurancethatredundancyindecayheagremoval capability will be maintained. B 11, 1980, all PWR licensees were requested to: 1) yproposeletter dated Junespecification changes technical that provide for redundancy in decay heat removal capability in all modes of operation, 2) use the NRC model technical specifications (MTS) which provide an acceptable solution of the concern and include an appropriate safety analysis as basis, and 3) submit the proposed technical specifica-tions (TS) with the bases by October 11, 1980.
On September 26, 1980, Carolina Power and Light Company (CP&L) submit-ted propose UnitNo.2.gchangestotheirtechnicalspecificationsforH.B.
The following discussion evaluates the proposed TS Robinson, and notes any differences existing between them and the model TS provided by the NRC (Appendix A). The requirements are compared for equivalent modes of operation.
2.0 DISCUSSION H. B. Robinson, Unit No. 2 is a three-loop Westinghouse PWR plant.
ThefgllowingsectionsoftheWestinghouseStandardTechnicalSpecifica-tions apply to this task:
3/4.4 Reactor Coolant System ,
3/4.4.1 Reactor Coolant Loops and Coolant Circulation 3/4.9 Refueling Operations 3/4.9.8 Residual Heat Removal and Coolant Circulation p H. B. Robinson's proposed TS varied from the NRC MTS in the following requirements:
, 1. Limiting Condition 3.4.1.1. The proposed Robinson TS require that power operation with less than three loops in service is prohibited, which meets the NRC MTS. However, no action is given if the above requirement is failed to be maintained.
1
- 2. Limiting Condition 3.4.1.2. Robinson's TS require that at least two steam generators shall be operable whenever the average pri-mary coolant temperature is above 350*F; also, whenever a reduc-tion of the boron concentration is made, at least one reactor coolant pump or Residual Heat Removal (RHR) system shall be in operation. Robinson requires that the plant shall be placed in cold shutdown if any one component in a reactor coolant loop is inoperable for a specified time period while the plant is in hot shutdown (as defined by Robinson's TS). The time period ranges e from-4 to 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> for the specific component plus 48 hours2 days <br />0.286 weeks <br />0.0658 months <br />, after which the specified. action is required. . There is no specified time period given in which the plant has to be in cold i shutdown after the action is initialized.
The NRC MTS require that at least two coolant loops shall be ooerable and at least one loop be in operation; with less than the above, restore the required loops to operable status within 72 hours3 days <br />0.429 weeks <br />0.0986 months <br /> or have Tavg below 350*F within the next 72 hours3 days <br />0.429 weeks <br />0.0986 months <br />,
- 3. Mode 4. H. B. Robinson's proposed TS define hot shutdown as when the reactor is subcritical and T av is > 540*F. The NRC MTS equivalentmode,hotstandby,iswkenT- is > 350*F. The NRCMTShotshutdown(Mode 4)isdefinehlhenT is between 200* and 350 F. RobinsondidnothaveadefinehyOonditionthat corresponds to the NRC MTS hot shutdown and so did not include coverage of Limiting Condition 3.4.1.3 for hot shutdown (NRC Mode 4).
- 4. Limiting Condition 3.4.1.3. In cold shutdown botn RHR loops must be operable. If an RHR loop becomes inoperable, the proposed Robinson TS states: "within 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> verify the existance of a method to add make-up water to the reactor coolant system."4 The NRC MTS also requires at least one loop to be in operation.
The action requirement given by the NRC is if no coolant loop is in operation suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.
- 5. Limiting Condition 3.9.8.2. When the water level is less than 23 feet above the top of the irradiated fuel assemblies, the pro-posed Robinson TS require one RHR loop to be in operation. The NRC MTS require two independent RHR loops to be operable. If all of the specified limiting conditions are not met, Robinson requires refueling of the reactor to cease, no operations shall
- be made to increase the reactivity of the core and work shall be initiated to correct the conditions. The NRC MTS action require-ment is, "if less than two RHR loops are operable, immediately -
initiate corrective action to restore the loops to cperable status."5 2
- 6. Surveillance Requirements. The proposed H. B. Robinson TS do not include any surveillance r'equirements in their TS. The justifi- '
cation is given in their September 26, 1980, letter that NRC MTS l do not conform with the format of Robinson TS and surveillance requirements are considered routine operator functions which have been and will continue to be controlled and documented at the Robinson plant through administrative methods (Operator's Logs, Checklists, ghift Turnover Checklists and Operating Work
., Procedures).
3.0 CONCLUSION
S The proposed Robinson TS varied from the NRC MTS in almost all areas.
During startup and power operation no action was given if the required con-ditions were not met. During hot standby (as defined by NRC) action was different between the NRC MTS and Robinson's TS. Robinson did not have a classification that corresponded to the NRC hot sh'itdown so these limiting conditions were not covered. In cold shutdown the limiting conditions, and
'therefore the action requirements, are different. During refueling when the water level is below 23 feet above the top _ of the irradiated fuel assemblies within the reactor pressure vessel, Robinson had different limiting condi-tions. - All surveillance requirements in the NRC MTS were not covered as Robinson's TS do not include them according to their format.
4.0 REFERENCES
- 1. NRC IE Information Notice 80-20, May 8, 1980.
- 2. NRC IE Bulletin 80-12, May 9, 1980.
- 3. NRC letter, D. G. Eisenhut, To All Operating Pressurized Water Reactors (PWRs) dated June 11, 1980.
- 4. CP&L letter, E. E. Utley to NRC, Steven A. Varga, dated September 26, 1980.
- 5. . Standard Technical Specifications for Westinghouse Pressurized Water Reactors, NUREG-0452, Rev. 3, Fall 1980.
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C APPENDIX A MODEL TECHNICAL SPECIFICATIONS FOR REDUNDANT DECAY HEAT REMOVAL FOR WESTINGHOUSE PRESSURIZED WATER REACTORS (PWRs) r 5
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION D
3.4.1.1 All' react'or coolant . loops shall be in operation. -
APPLICABILITY: MODES I and 2.* *
-ACTION:
With less than' the above required reactor coolant loops in operation, be in.
at least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. '
SURVEILLANCE REQUIREMENT 4.4.1.1 The above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br />.
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See Special Test Exception 3.10.4.
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- REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 a. Atlleast two of the reactor coolant loops listed below shall be OPERABLE: i
- 1. Reactor Coolant Loop (A) ar.d its associated steam generator.and reactor coolant pump, .
- 2. Reactor Coolant Loop (8) and its associated steam generator and reactor coolant pump,
- 3. Reactor Coolant Loop (C) and its associated steam generator and reactor coolant pump,
- 4. Reactor Coolant Loop (D) and its associated steam generator and reactor coolant pump.
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! b. At least one of the above coolant loops shall be .in l
operation.*
APPLICABILITY: MODE 3
- ACTION
- a. With less than the above required r'eactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours3 days <br />0.429 weeks <br />0.0986 months <br /> or
, be in HOT SHUT 00WN within the next 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br />.
(
i o All reactor coolant pumps may be de-energized for up to I hour provided (1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (2) core outlet temperature is maintained at least 10 0F below saturation temperature.
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REACTOR COOLANT SYSTEM'
-b. With no. reactor coolant: loop in operation, suspend all operations involving a reduction in boron concentration of.the Reactor Coolant System and immediately' initiate corrective action to
, . return the required coolant loop to operation.
SURVEILLANCE REQUIREMENT 4.4.1.2.1 At least the above. required reactor coolant pumps, if not in
- - operation,.shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability..
4.4.'l.2.2 At least one cooling loop shall be_ verified'to be in operation
, - and circulating: reactor coolant at least once per 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br />.
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. REACTOR COOLANT SYSTEM SHUTDOWN '
-LIMITING CONDITION FOR OPERATION 3.4.1.3 .a.- ' At least two of the coolant loops listed below sha',1 be '
~0PERABLE: (.
- 1. Reactor Coolant Loop (A) and its associated steam gen-
. erator and reactor coolant pump,* .
- 2. -Reactor Coolant Loop (B) and its associated steam gen-erator'and reactor coolant pump,*
- 3. Reactor Coolant Loop (C) and.its associated-steam gen-erator and reactor coolant pump,*
- 4. Reactor Coolant Loop (D) and its associated steam gen-erator and reactor coolant pump,*
- 5. Residual Heat Removal Loop (A),**
- 6. Residual Heat Removal Loop (B).** -
- b. At least one of the.above coolant loops shall be in operation.***
-* A reactor coolant pump shall not be started with one or more of the
~RCS cold leg temperatures.less than or equal to (275)0F unless 1) the pressurizer water volume is less than cubic feet or 2) the secondary
- water temperature of each steam generator is less than OF above each of the RCS cold. leg temperatures. 5,
- .The normal or emergency power source may be inoperable in MODE 5.
C** -All reactor coolant pumps and decay heat removal pumps may be de-energized for up to I hour provided .1) no operations are permitted that
. would cause dilution of the reactor coolant system boron concentration, and
-2)' core ~ outlet temperature is maintained at least 100F below saturation temperature.
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REACTOR COOLANT SYSTEM APPLICABILITY: MODES 4 and 5.
ACTION:
- a. With less than_the'above required. loops OPERABLE, immediately initiate corrective action to~ return the required loops to OPERABLE status as soon as'possible; be in COLD SHUTDOWN within 20 hours0.833 days <br />0.119 weeks <br />0.0274 months <br />.
- b. With no coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.
SURVEILLANCE REQUIREMENT 4.4.1.3.1 The required residual heat removal loop (s) shall be determined OPERABLE per Specification 4.0.5.
4.4.1.3.2 The required reactor coolant pump (s), if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.
4.4.1.3.3 The required steam generator (s) shall be determined OPERABLE by verifying secondary side level to be greater tha'n or equal to ( )% at least once per 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br />.
4.4.1.3.4 At least one coolant loop shall be verified to tue in operation and circulating reactor coolant at least once per 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br />.
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REFUELING OPERATIONS -
3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION ALL WATER LEVELS LIMITING CONDITION FOR OPERATION t
3.9.8.1 At least one residual heat removal (RHR) loop shall be in operation. .
APPLICABILITY: MODE 6 ACTION:
- a. With less than one residual heat removal loop in operation, except as provided in b. below, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System. Close all containment penetrations providing direct access from the con-tainment atmosphere to the outside atmosphere within 4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br />.
- b. The residual heat removal loop may be removed from operation for up to I hour per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel (hot) legs.
- c. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENT 4.9.8.1 At least one residual heat removal loop shall be verified to be in operation and circulating reactor coolant at a flow rate of greater than or equal to (2800) gpm at least once per 4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br />.
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3 REFUELING OPERATIONS -
- LOW WATER LEVEL -
LIMITING CONDITION FOR OPERATION
- 3.9.8.2 Two independent Residual Heat Removal -(RHR) . loops shall be' OPERABLE.* .
, APPLICABILITY: MODE 6 when the. water' level above the top of the irradiated fuel assemblies-seated within the reactor pressure vessel!is-less than 23 feet.
, ACTION:
- a. With'less than the required RHR loops'0PERABLE,-immediately initiate corrective action to return the required RHR loops to.
0PERABLE status as soon as possible.
- b. The. provisions of Specification 3.0.3 are not applicable.
j SURVEILLANCE REQUIREMENT f
- . 4.9.8.2 The required Residual Heat Removal-loops shall be determined OPERABLE per Specification 4.0.5.
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.'.The. normal or emergency power source may be inoperable for each RHR loop. '
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- _ . . . . . . . _ - . .z. , , ._ . . . _ . _ . . . . . _ , , , _ . . _ .
r 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in g operation, and maintain DNBR above 1.30 during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation this specification requires that the plant be ir. at least HOT ,
STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure con-siderations require that two loops be OPERABLE.
In MODES 4 and 5, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE. Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.
The operation of one Reactor Coolant Pump or one RHR pump provides .
adequate flow to ensure mixing, prevent stratification and produce gradual reactivity. changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.
-The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less than or equal to (275)0F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into, or (2) by restricting starting of the RCPs to when the secon-dary water temperature of each steam generator is less than ( )0F above each of the RCS cold leg temperatures.
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~'
REFUELING OPERATIONS-
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-BASES L
~
3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION j' The requirement that at least-one residual heat removal (RHR) loop be in operation ensures that (1) sufficient cooling capacity is available'to
~
remove decay heat and maintain the water in the reactor pressure. vessel
, below 140 F as required during'the REFUELING MODE,.and (2) sufficient cool-ant circulation is maintained through the reactor core.to minimize the
- effect of.a boron dilution incident and prevent boron stratification.
The requirement to have two RHR loops OPERABLE when there is.less than 23 feet of water above the core ensures that a single failure of the oper-ating RHR loop will not result 1n a complete loss of residual heat removal
~
capability. With the reactor vessel head removed and 23 feet of water above.the core, a large heat sink is available for core cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided
- to initiate emergency procedures to cool the core.
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