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ATTACHMENT 4 NRC Regulatory Issue Summary 2002-03 Cross-Reference
ATTACHMENT 4 NRC Regulatory Issue Summary 2002-03 Cross-Reference NRC RIS 2002-03 LAR DOCUMENT Item No.
DESCRIPTION ATTACHMENT SECTION and TITLE Page 1 of 11 I. Feedwater Flow Measurement Technique and Power Measurement Uncertainty I.1.
Detailed description of plant-specific implementation of feedwater flow measurement technique and power increase gained as a result of implementing technique 3.1 Background and General Approach 3.2 LEFM Ultrasonic Flow Measurement and Core Thermal Power Uncertainty I.1.A.
NRC approval of topical report on flow measurement technique 3.2.1 LEFM Flow Measurement I.1.B.
Reference to NRCs approval of proposed measurement technique 3.2.1 LEFM Flow Measurement I.1.C.
Plant Implementation 3.2.2 Plant Implementation I.1.D.
Disposition of NRC criteria 3.2.4 Disposition of NRC Criteria for Use of Topical Reports I.1.E.
Total power measurement uncertainty calculation for the plant 1
3.2.3 LEFM and Core Thermal Power Measurement Uncertainty and Methodology Core Thermal Power Uncertainty Calculation I.1.F.
Calibration and maintenance 3.2.4 Disposition of NRC Criteria for Use of Topical Reports 3.2.5 Deficiencies and Corrective Actions I.1.G.
Proposed outage time for LEFM and basis for selected time 3.2.4 Disposition of NRC Criteria for Use of Topical Reports I.1.H Proposed actions if outage time is exceeded, and basis for actions 3.2.4 Disposition of NRC Criteria for Use of Topical Reports II.
Accidents and Transients for which the Existing Analyses of Record Bound Plant Operation at the Proposed Uprated Power Level II.1 Matrix for bounded accidents and transients 9.0 Reactor Safety Performance Evaluations
ATTACHMENT 4 NRC Regulatory Issue Summary 2002-03 Cross-Reference NRC RIS 2002-03 LAR DOCUMENT Item No.
DESCRIPTION ATTACHMENT SECTION and TITLE Page 2 of 11 III.
Accidents and Transients for which the Existing Analyses of Record Do Not Bound Plant Operation At the Proposed Uprated Power Level III.1,2, 3 Matrix for unbounded accidents and transients 9.0 Reactor Safety Performance Evaluations IV.
Mechanical/Structural/Material Component Integrity and Design IV.1.A.i Reactor vessel, nozzles, supports 3.2 Reactor Vessel 3.2.1 Fracture Toughness 3.2.2 Reactor Vessel Structural Evaluation IV.1.A.ii Reactor core support structures and vessel internals 3.3 Reactor Internals 3.3.1 Reactor Internal Pressure Difference 3.3.2 Reactor Internals Structural Evaluation 3.3.3 Steam Separator and Dryer Performance 3.4 Flow-Induced Vibration 3.4.2 Adverse Flow Effects IV.1.A.iii Control rod drive mechanisms 2.5 Reactivity Control
ATTACHMENT 4 NRC Regulatory Issue Summary 2002-03 Cross-Reference NRC RIS 2002-03 LAR DOCUMENT Item No.
DESCRIPTION ATTACHMENT SECTION and TITLE Page 3 of 11 IV.1.A.iv Nuclear Steam Supply System (NSSS) piping, pipe supports, branch nozzles 3.4 Flow-Induced Vibration 3.5 Piping Evaluation 3.5.1 Reactor Coolant Pressure Boundary Piping 3.6 Reactor Recirculation System 3.7 Main Steam Line Flow Restrictors 3.8 Main Steam Isolation Valves 3.9 Reactor Core Isolation Cooling 3.10 Residual Heat Removal System 3.11 Reactor Water Cleanup System IV.1.A.v Balance of plant (BOP) piping (NSSS interface systems, safety-related cooling water systems, and containment systems) 3.5 Piping Evaluation 3.5.2 Balance-of-Plant Piping Evaluation 6.4.1 Service Water Systems 6.4.3 Reactor Building Closed Cooling Water System IV.1.A.vi Steam generator tubes, secondary side internal support structures, shell and nozzles NA NA IV.1.A.vii Reactor coolant pumps NA NA IV.1A.viii Pressurizer shell, nozzles, and surge lines NA NA
ATTACHMENT 4 NRC Regulatory Issue Summary 2002-03 Cross-Reference NRC RIS 2002-03 LAR DOCUMENT Item No.
DESCRIPTION ATTACHMENT SECTION and TITLE Page 4 of 11 IV.1.A.ix Safety-related valves 3.1 Nuclear System Pressure Relief/Overpressure Protection 3.8 Main Steam Isolation Valves 4.1 Containment System Performance 4.1.1 Generic Letter 89-10 Program 4.1.2 Generic Letter 95-07 Program 6.5 Standby Liquid Control System IV.1.B.i Stresses 3.2 Reactor Vessel 3.2.2 Reactor Vessel Structural Evaluation 3.4 Flow-Induced Vibration 3.5 Piping Evaluation 3.5.1 Reactor Coolant Pressure Boundary Piping 3.5.2 Balance-of-Plant Piping Evaluation IV.1.B.ii Cumulative usage factors 3.2.2 Reactor Vessel Structural Evaluation IV.1.B.iii Flow induced vibration 3.4 Flow-Induced Vibration 3.4.2 Adverse Flow Effects IV.1.B.iv Changes in temperature (pre-and post-uprate) 1.3 TPO Plant Operating Conditions 1.3.1 Reactor Heat Balance 1.3.2 Reactor Performance Improvement Features Table 1-2 Thermal-Hydraulic Parameters at TPO Uprate Conditions
ATTACHMENT 4 NRC Regulatory Issue Summary 2002-03 Cross-Reference NRC RIS 2002-03 LAR DOCUMENT Item No.
DESCRIPTION ATTACHMENT SECTION and TITLE Page 5 of 11 IV.1.B.v Changes in pressure (pre-and post-uprate) 1.3 TPO Plant Operating Conditions 1.3.1 Reactor Heat Balance 1.3.2 Reactor Performance Improvement Features Table 1-2 Thermal-Hydraulic Parameters at TPO Uprate Conditions IV.1.B.vi Changes in flow rate (pre-and post-uprate) 1.3 TPO Plant Operating Conditions 1.3.1 Reactor Heat Balance 1.3.2 Reactor Performance Improvement Features Table 1-2 Thermal-Hydraulic Parameters at TPO Uprate Conditions IV.1.B.vii High-energy line break locations 10.1 High Energy Line Break 10.1.1 Steam Line Breaks 10.1.2 Liquid Line Breaks IV.1.B.viii Jet impingement and thrust forces 10.1 High Energy Line Break 10.1.1 Steam Line Breaks 10.1.2 Liquid Line Breaks 10.1.2.6 Pipe Whip and Jet Impingement IV.1.C.i Reactor vessel pressurized thermal shock calculations 3.1 Nuclear System Pressure Relief/Overpressure Protection IV.1.C.ii Reactor vessel fluence evaluation 3.2 Reactor Vessel 3.2.1 Fracture Toughness
ATTACHMENT 4 NRC Regulatory Issue Summary 2002-03 Cross-Reference NRC RIS 2002-03 LAR DOCUMENT Item No.
DESCRIPTION ATTACHMENT SECTION and TITLE Page 6 of 11 IV.1.C.iii Reactor vessel heatup and cooldown pressure-temperature limit curves 3.2 Reactor Vessel 3.2.1 Fracture Toughness IV.1.C.iv Reactor vessel low temperature overpressure protection 3.2 Reactor Vessel 3.2.1 Fracture Toughness IV.1.C.v Reactor vessel upper shelf energy 3.2 Reactor Vessel 3.2.1 Fracture Toughness IV.1.C.vi Reactor vessel surveillance capsule withdrawal schedule 3.2 Reactor Vessel 3.2.1 Fracture Toughness IV.1.D Code of record 3.2 Reactor Vessel 3.2.2 Reactor Vessel Structural Evaluation 3.5 Piping Evaluation 3.5.1 Reactor Coolant Pressure Boundary Piping 3.5.2 Balance-of-Plant Piping Evaluation IV.1.E Component inspection/testing programs and erosion/corrosion programs 3.5 Piping Evaluation 3.5.1 Reactor Coolant Pressure Boundary Piping 3.5.2 Balance-of-Plant Piping Evaluation 10.6 Plant Life IV.1.F NRC Bulletin 88-02, Rapidly Propagating Fatigue Cracks in Steam Generator Tubes NA NA V.
Electrical Equipment Design V.1.A Emergency diesel generators 6.1 AC Power 6.1.2 On-Site Power
ATTACHMENT 4 NRC Regulatory Issue Summary 2002-03 Cross-Reference NRC RIS 2002-03 LAR DOCUMENT Item No.
DESCRIPTION ATTACHMENT SECTION and TITLE Page 7 of 11 V.1.B Station blackout equipment 9.3.2 Station Blackout V.1.C Environmental qualification of electrical equipment 10.3 Environmental Qualification 10.3.1 Electrical Equipment V.1.D Grid stability 2
6.1 AC Power 6.1.1 Off-Site Power All Grid Stability Study VI.
System Design
VI.1.A NSSS Interface Systems for BWRs (e.g.,
suppression pool cooling, as applicable) 3.4 Flow-Induced Vibration 3.5 Piping Evaluation 3.5.1 Reactor Coolant Pressure Boundary Piping 3.5.2 Balance-of-Plant Piping Evaluation 3.6 Reactor Recirculation System 3.7 Main Steam Line Flow Restrictors 3.8 Main Steam Isolation Valves 3.9 Reactor Core Isolation Cooling 3.10 Residual Heat Removal System 3.11 Reactor Water Cleanup System VI.1.B Containment Systems 4.1 Containment System Performance 4.1.1 Generic Letter 89-10 Program 4.1.2 Generic Letter 95-07 Program 4.1.3 Generic Letter 96-06
ATTACHMENT 4 NRC Regulatory Issue Summary 2002-03 Cross-Reference NRC RIS 2002-03 LAR DOCUMENT Item No.
DESCRIPTION ATTACHMENT SECTION and TITLE Page 8 of 11 VI.1.C Safety-related cooling water systems 6.4 Water Systems 6.4.1 Service Water Systems 6.4.5 Ultimate Heat Sink VI.1.D Spent fuel pool storage and cooling systems 6.3 Fuel Pool 6.3.1 Fuel Pool Cooling 6.3.2 Crud Activity and Corrosion Products 6.3.3 Radiation Levels 6.3.4 Fuel Racks VI.1.E Radioactive waste systems 4.5 Standby Gas Treatment System 8.1 Liquid and Solid Waste Management 8.2 Gaseous Waste Management 8.3 Radiation Sources in the Reactor Core 8.4 Radiation Sources in Reactor Coolant 8.4.1 Coolant Activation Products 8.4.2 Activated Corrosion Products 8.4.3 Fission Products 8.5 Radiation Levels 8.6 Normal Operation Off-Site Doses
ATTACHMENT 4 NRC Regulatory Issue Summary 2002-03 Cross-Reference NRC RIS 2002-03 LAR DOCUMENT Item No.
DESCRIPTION ATTACHMENT SECTION and TITLE Page 9 of 11 VI.1.F Engineered safety features (ESF) heating, ventilation, and air conditioning systems 4.4 Main Control Room Atmosphere Control System 4.7 Post-LOCA Combustible Gas Control System 6.6 Power Dependent Heating, Ventilation and Air Conditioning VII.
Other VII.1 Operator actions and effects on time available 4.1 Containment System Performance 6.7 Fire Protection 9.3 Special Events 9.3.2 Station Blackout 10.5 Operator Training and Human Factors VII.2.A Emergency and abnormal operating procedures 10.9 Emergency Operating Procedures VII.2.B Control room controls, displays (including the safety parameter display system) and alarms 3.2.4 Disposition of NRC Criteria for Use of LEFM Topical Reports 3.4.3 Plant Modifications 10.5 Operator Training and Human Factors VII.2.C The control room plant reference simulator 10.5 Operator Training and Human Factors VII.2.D The operator training program 10.5 Operator Training and Human Factors VII.3 Modifications completion 3.4.3 Plant Modifications List of Regulatory Commitments VII.4 Procedure Revisions - Licensed Power Level 3.2.6 Reactor Power Monitoring
ATTACHMENT 4 NRC Regulatory Issue Summary 2002-03 Cross-Reference NRC RIS 2002-03 LAR DOCUMENT Item No.
DESCRIPTION ATTACHMENT SECTION and TITLE Page 10 of 11 VII.5.A 10 CFR 51.22, Exclusion of Environmental Review, including discussion of effect of the power uprate on types and amounts of effluents released offsite, and whether bounded by final environmental statement and previous Environmental Assessments for the plant 5.0 Environmental Consideration 8
Radwaste and Radiation Sources VII.5.B 10 CFR 51.22, Exclusion of Environmental Review, including discussion of effect of the power uprate on individual and cumulative occupational radiation exposure 5.0 Environmental Consideration 8.5 Radiation Levels VIII.
Changes to Technical Specifications, Protection System Settings, Emergency System Settings VIII.1 A detailed discussion of each change to the plants Technical Specifications, protection system settings, and/or emergency system settings needed to support the power uprate 2.0 Detailed Description All Markup of Proposed Operating License and Technical Specifications Pages VIII.1.A Description of the change 2.0 Detailed Description All Markup of Proposed Operating License and Technical Specifications Pages VIII.1.B Identification of analyses affected by and/or supporting the change 3.3 Evaluation of Changes to License and Technical Specifications All GEH Nuclear Energy Safety Analysis Report for LaSalle County Station, Units 1 and 2 Thermal Power Optimization, NEDC-33485P
ATTACHMENT 4 NRC Regulatory Issue Summary 2002-03 Cross-Reference NRC RIS 2002-03 LAR DOCUMENT Item No.
DESCRIPTION ATTACHMENT SECTION and TITLE Page 11 of 11 VIII.1.C Justification for the change, including the type of information discussed in Section III, above, for any analyses that support and/or are affected by change 3.3 Evaluation of Changes to License and Technical Specifications All GEH Nuclear Energy Safety Analysis Report for LaSalle County Station, Units 1 and 2 Thermal Power Optimization, NEDC-33485P