ML100321309

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Attachment 4 - NRC Regulatory Issue Summary 2002-03 Cross-Reference
ML100321309
Person / Time
Site: LaSalle  Constellation icon.png
Issue date: 01/27/2010
From:
Exelon Generation Co, Exelon Nuclear
To:
Office of Nuclear Reactor Regulation
Shared Package
ML100321303 List:
References
RIS-02-003, RS-10-001
Download: ML100321309 (12)
v • d • e


See also: RIS 2002-03

Text

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 ATTACHMENTSECTION 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 Attachment 1 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 Attachment 1 3.2.1 LEFM Flow Measurement I.1.B. Reference to NRCs approval of proposed measurement technique Attachment 1 3.2.1 LEFM Flow Measurement I.1.C. Plant Implementation Attachment 1 3.2.2 Plant Implementation I.1.D. Disposition of NRC criteria Attachment 1 3.2.4 Disposition of NRC Criteria for Use of Topical Reports I.1.E. Total power measurement uncertainty calculation for the plant

Attachment 1

Attachment 11 3.2.3 LEFM and Core Thermal Power Measurement Uncertainty and Methodology

Core Thermal Power Uncertainty Calculation I.1.F. Calibration and maintenance Attachment 1 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 Attachment 1 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 Attachment 1 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 Attachment 6 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 ATTACHMENTSECTION 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 Attachment 6 9.0 Reactor Safety Performance Evaluations IV. Mechanical/Structural/Material Component Integrity and Design IV.1.A.i Reactor vessel, nozzles, supports Attachment 6 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

Attachment 6

Attachment 1 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 Attachment 6 2.5 Reactivity Control

ATTACHMENT 4 NRC Regulatory Issue Summary 2002-03 Cross-Reference

NRC RIS 2002-03 LAR DOCUMENT Item No. DESCRIPTION ATTACHMENTSECTION and TITLE

Page 3 of 11 IV.1.A.iv Nuclear Steam Supply System (NSSS) piping, pipe supports, branch nozzles Attachment 6 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) Attachment 6 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 ATTACHMENTSECTION and TITLE

Page 4 of 11 IV.1.A.ix Safety-related valves Attachment 6 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 Attachment 6 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 Attachment 6 3.2.2 Reactor Vessel Structural Evaluation IV.1.B.iii Flow induced vibration Attachment 6

Attachment 1 3.4 Flow-Induced Vibration 3.4.2 Adverse Flow Effects IV.1.B.iv Changes in temperature (pre- and post- uprate) Attachment 6 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 ATTACHMENTSECTION and TITLE

Page 5 of 11 IV.1.B.v Changes in pressure (pre- and post- uprate) Attachment 6 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) Attachment 6 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 Attachment 6 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 Attachment 6 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

Attachment 6 3.1 Nuclear System Pressure Relief/Overpressure Protection IV.1.C.ii Reactor vessel fluence evaluation Attachment 6 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 ATTACHMENTSECTION and TITLE

Page 6 of 11 IV.1.C.iii Reactor vessel heatup and cooldown pressure-

temperature limit curves Attachment 6 3.2 Reactor Vessel 3.2.1 Fracture Toughness IV.1.C.iv Reactor vessel low temperature overpressure

protection Attachment 6 3.2 Reactor Vessel 3.2.1 Fracture Toughness IV.1.C.v Reactor vessel upper shelf energy Attachment 6 3.2 Reactor Vessel 3.2.1 Fracture Toughness IV.1.C.vi Reactor vessel surveillance capsule withdrawal

schedule Attachment 6 3.2 Reactor Vessel 3.2.1 Fracture Toughness IV.1.D Code of record Attachment 6 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 Attachment 6 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 Attachment 6 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 ATTACHMENTSECTION and TITLE

Page 7 of 11 V.1.B Station blackout equipment Attachment 6 9.3.2 Station Blackout V.1.C Environmental qualification of electrical

equipment Attachment 6 10.3 Environmental Qualification 10.3.1 Electrical Equipment V.1.D Grid stability Attachment 6

Attachment 12 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) Attachment 6 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 Attachment 6 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 ATTACHMENTSECTION and TITLE

Page 8 of 11 VI.1.C Safety-related cooling water systems Attachment 6 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 Attachment 6 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 Attachment 6 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 ATTACHMENTSECTION and TITLE

Page 9 of 11 VI.1.F Engineered safety features (ESF) heating, ventilation, and air conditioning systems Attachment 6 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 Attachment 6 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 Attachment 6 10.9 Emergency Operating Procedures VII.2.B Control room controls, displays (including the safety parameter display system) and alarms

Attachment 1

Attachment 6 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 Attachment 6 10.5 Operator Training and Human Factors VII.2.D The operator training program Attachment 6

10.5 Operator Training and Human Factors VII.3 Modifications completion Attachment 1

Attachment 5 3.4.3 Plant Modifications List of Regulatory Commitments VII.4 Procedure Revisions - Licensed Power Level Attachment 1 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 ATTACHMENTSECTION 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

Attachment 1

Attachment 6 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

Attachment 1

Attachment 6 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

Attachment 1

Attachment 2

2.0 Detailed Description All Markup of Proposed Operating License and Technical Specifications Pages VIII.1.A Description of the change Attachment 1

Attachment 2 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

Attachment 1

Attachment 6 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 ATTACHMENTSECTION 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 Attachment 1

Attachment 6 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

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