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{{#Wiki_filter:PBAPS UFSAR   APPENDIX J J-i REV. 21, APRIL 2007 APPENDIX J - INDENTIFICATION-RESOLUTION OF AEC-ACRS AND STAFF CONCERNS  TABLE OF CONTENTS
{{#Wiki_filter:PBAPS UFSAR APPENDIX J J-i REV. 21, APRIL 2007 APPENDIX J - INDENTIFICATION-RESOLUTION OF AEC-ACRS AND STAFF CONCERNS  TABLE OF CONTENTS


J.1 SUMMARY DESCRIPTION J.2 AREAS SPECIFIED IN THE PEACH BOTTOM UNITS 2 AND 3--  AEC-ACRS CONSTRUCTION PERMIT LETTER J.2.1 Introduction J.2.2 Browns Ferry ACRS Comments (3/14/67) Applicable  to Peach Bottom J.2.2.1 Effects of Fuel Failure on CSCS Performance J.2.2.2 Effects of Fuel Bundle Flow Blockage J.2.2.3 Verification of Fuel Damage Limit Criterion J.2.2.4 Effects of Cladding Temperatures and Materials on CSCS Performance J.2.2.5 Quality Assurance and Inspection of the Reactor Primary System J.2.2.6 Control Rod Block Monitor Design J.2.2.7 Plant Startup Program J.2.2.8 Main Steam Line Isolation Valve Testing Under Simulated Accident Conditions J.2.2.9 Performance Testing of the Station Standby Diesel Generator System J.2.2.10 Formulation of an In-Service Inspection Program J.2.2.11 Diversification of the CSCS Initiation Signals J.2.2.12 Control Systems for Emergency Power J.2.2.13 Misorientation of Fuel Assemblies J.2.3 Failure of Conowingo Dam - Alternate Heat Removal Capability J.2.4 Ring Header Leakage Protection Capability J.2.5 Station Thermal Effects - Commonwealth of Pennsylvania Limits J.2.6 HPCIS - Depressurization Capability J.2.7 Station Startup Program J.2.8 Conclusions
J.1 SUMMARY DESCRIPTION


J.3 AREAS SPECIFIED IN THE PEACH BOTTOM UNITS 2 AND 3-- AEC-STAFF CONSTRUCTION PERMIT SAFETY EVALUATION REPORT J.3.1 General J.3.2 AEC-Staff SER Section 2.0 Concerns J.3.2.1 Introduction J.3.2.2 RPS - IEEE-279 Design Statement J.3.3 AEC-Staff SER Section 3.0 Concerns J.3.3.1 Introduction J.3.3.2 Station Meteorological Program PBAPS UFSAR  APPENDIX J J-ii REV. 21, APRIL 2007  J.3.3.3 Station-Site Slope Cut Program Studies J.3.3.4 Station-Site Flood Protection Studies J.3.3.5 Station-Site Diffusion and Dispersion Studies - Radiological Effects Determination J.3.3.6 Station Alternative Heat Sink in Event of Dam Failure - Design Capability J.3.4 AEC-Staff SER Section 4.0 Concerns J.3.4.1 Introduction J.3.4.2 Suction Piping System Supply Water to ECCS (CSCS) - Design Aspects J.3.4.3 Adequacy of HPCIS as a Depressurizer J.3.4.4 Engineered Safety Features - Electrical Equipment Inside Primary Containment - Design Capabilities J.3.5 AEC-Staff SER Section 5.0 Concerns J.3.5.1 Introduction J.3.5.2 Steam Line Break Fuel Rod Integrity - Thermal Hydraulic Analytical Justification J.3.6 AEC-Staff SER Section 6.0 Concerns J.3.6.1 Introduction J.3.6.2 Development Program of Significance for All Large Water-Cooled Power Reactors J.3.6.3 Development Program of Significance for BWR's in General J.3.6.4 Areas Requiring Further Technical Information J.3.7 Conclusion
J.2 AREAS SPECIFIED IN THE PEACH BOTTOM UNITS 2 AND 3--
AEC-ACRS CONSTRUCTION PERMIT LETTER J.2.1 Introduction J.2.2 Browns Ferry ACRS Comments (3/14/67) Applicable  to Peach Bottom J.2.2.1 Effects of Fuel Failure on CSCS Performance J.2.2.2 Effects of Fuel Bundle Flow Blockage J.2.2.3 Verification of Fuel Damage Limit Criterion J.2.2.4 Effects of Cladding Temperatures and Materials on CSCS Performance J.2.2.5 Quality Assurance and Inspection of the Reactor Primary System J.2.2.6 Control Rod Block Monitor Design J.2.2.7 Plant Startup Program J.2.2.8 Main Steam Line Isolation Valve Testing Under Simulated Accident Conditions J.2.2.9 Performance Testing of the Station Standby Diesel Generator System J.2.2.10 Formulation of an In-Service Inspection Program J.2.2.11 Diversification of the CSCS Initiation Signals J.2.2.12 Control Systems for Emergency Power J.2.2.13 Misorientation of Fuel Assemblies J.2.3 Failure of Conowingo Dam - Alternate Heat Removal Capability J.2.4 Ring Header Leakage Protection Capability J.2.5 Station Thermal Effects - Commonwealth of Pennsylvania Limits J.2.6 HPCIS - Depressurization Capability J.2.7 Station Startup Program


J.4 AREAS SPECIFIED IN OTHER RELATED AEC-ACRS CONSTRUCTION AND OPERATING PERMIT LETTERS J.4.1 General J.4.2 Instrumentation for Prompt Detection of Gross Fuel Failure J.4.3 AEC General Design Criterion No. 35 - Design Intent and Conformance J.4.4 Scram Reliability Study J.4.5 Design Basis of Engineered Safety Features J.4.6 Hydrogen Generation Study J.4.7 Seismic Design and Analysis Models J.4.8 Automatic Pressure Relief System - Single Component Failure Capability - Manual Operation J.4.9 Flow Reference Scram J.4.10 Main Steam Lines - Standards for Fabrication, Q/C, and Inspection J.4.11 Main Steam Line Isolation Valve Leakage J.4.12 Reactor Startup Vibration Testing Capability J.4.13 Control Rod Drop Accident PBAPS UFSAR  APPENDIX J J-iii REV. 21, APRIL 2007  J.5 AREAS SPECIFIED IN OTHER RELATED AEC-STAFF CONSTRUCTION OR OPERATING PERMIT SAFETY EVALUATION REPORTS J.5.1 General J.5.2 Tornado and Missile Protection - GE BWR - Spent Fuel Storage Pool J.5.3 BWR System Stability Analysis J.5.4 Reactor Pressure Vessel - Stub Tube Design J.5.5 RPS and CSCS Instrumentation - Cable Marking and Identification J.5.6 RPS and CSCS Instrumentation - Design Criteria  (IEEE-279)
J.2.8 Conclusions
PBAPS UFSAR  APPENDIX J J-iv REV. 21, APRIL 2007 APPENDIX J - INDENTIFICATION-RESOLUTION OF AEC-ACRS AND STAFF CONCERNS  LIST OF TABLES TABLE TITLE 


J.2.1 Summary of Design Basis LOCA Calculations With Fuel Loading Errors J.2.2 Peak Cladding Temperatures - Design Basis LOCA
J.3 AREAS SPECIFIED IN THE PEACH BOTTOM UNITS 2 AND 3--
AEC-STAFF CONSTRUCTION PERMIT SAFETY EVALUATION REPORT J.3.1 General J.3.2 AEC-Staff SER Section 2.0 Concerns


J.4.1 Rod Drop Accident Results - Three-Assembly  Gd2O3 Core - 50% Rod Density PBAPS UFSAR  APPENDIX J J-v REV. 21, APRIL 2007 APPENDIX J INDENTIFICATION-RESOLUTION OF AEC-ACRS AND STAFF CONCERNS  LIST OF FIGURES 
J.3.2.1 Introduction J.3.2.2 RPS - IEEE-279 Design Statement J.3.3 AEC-Staff SER Section 3.0 Concerns


FIGURE TITLE J.4.1 Rod Drop Accident Maximum Control Rod Worth with Single Operator Error at BOL with RSCS J.4.2 Rod Drop Accident Maximum Control Rod Worth with Single Operator Error at Most Reactive Point in Core Life (6,500 MWd/T) with RSCS J.4.3a Schematic Block Diagram of RSCS for Core Exposure <6,500 MWd/t J.4.3b RSCS Block Diagram for Core Exposure <6,500 MWd/t
J.3.3.1 Introduction J.3.3.2 Station Meteorological Program PBAPS UFSAR APPENDIX J J-ii REV. 21, APRIL 2007 J.3.3.3 Station-Site Slope Cut Program Studies J.3.3.4 Station-Site Flood Protection Studies J.3.3.5 Station-Site Diffusion and Dispersion Studies -
Radiological Effects Determination J.3.3.6 Station Alternative Heat Sink in Event of Dam Failure
- Design Capability J.3.4 AEC-Staff SER Section 4.0 Concerns


J.4.4a Schematic Block Diagram of RSCS for Core Exposure >6,500 MWd/t J.4.4b RSCS Block Diagram for Core Exposure>6,500 MWd/t
J.3.4.1 Introduction J.3.4.2 Suction Piping System Supply Water to ECCS (CSCS) -
Design Aspects J.3.4.3 Adequacy of HPCIS as a Depressurizer J.3.4.4 Engineered Safety Features - Electrical Equipment Inside Primary Containment - Design Capabilities J.3.5 AEC-Staff SER Section 5.0 Concerns


J.4.5 Fault Tree Model of a Control Rod Stuck in Core Ready for a Drop J.4.6 Reactor States Used in the Analysis of the Control Rod Drop Accident J.4.7 Probability Analysis of Control Rod Drop Accident  
J.3.5.1 Introduction J.3.5.2 Steam Line Break Fuel Rod Integrity - Thermal Hydraulic Analytical Justification J.3.6 AEC-Staff SER Section 6.0 Concerns
 
J.3.6.1 Introduction J.3.6.2 Development Program of Significance for All Large Water-Cooled Power Reactors J.3.6.3 Development Program of Significance for BWR's in General J.3.6.4 Areas Requiring Further Technical Information
 
J.3.7 Conclusion
 
J.4 AREAS SPECIFIED IN OTHER RELATED AEC-ACRS CONSTRUCTION AND OPERATING PERMIT LETTERS J.4.1 General J.4.2 Instrumentation for Prompt Detection of Gross Fuel Failure J.4.3 AEC General Design Criterion No. 35 - Design Intent and Conformance J.4.4 Scram Reliability Study J.4.5 Design Basis of Engineered Safety Features J.4.6 Hydrogen Generation Study J.4.7 Seismic Design and Analysis Models J.4.8 Automatic Pressure Relief System - Single Component Failure Capability - Manual Operation J.4.9 Flow Reference Scram J.4.10 Main Steam Lines - Standards for Fabrication, Q/C, and Inspection J.4.11 Main Steam Line Isolation Valve Leakage J.4.12 Reactor Startup Vibration Testing Capability J.4.13 Control Rod Drop Accident PBAPS UFSAR APPENDIX J J-iii REV. 21, APRIL 2007 J.5 AREAS SPECIFIED IN OTHER RELATED AEC-STAFF CONSTRUCTION OR OPERATING PERMIT SAFETY EVALUATION REPORTS J.5.1 General J.5.2 Tornado and Missile Protection - GE BWR - Spent Fuel Storage Pool J.5.3 BWR System Stability Analysis J.5.4 Reactor Pressure Vessel - Stub Tube Design J.5.5 RPS and CSCS Instrumentation - Cable Marking and Identification J.5.6 RPS and CSCS Instrumentation - Design Criteria (IEEE-279)
 
PBAPS UFSAR APPENDIX J J-iv REV. 21, APRIL 2007 APPENDIX J - INDENTIFICATION-RESOLUTION OF AEC-ACRS AND STAFF CONCERNS  LIST OF TABLES
 
TABLE TITLE 
 
J.2.1 Summary of Design Basis LOCA Calculations With Fuel Loading Errors
 
J.2.2 Peak Cladding Temperatures - Design Basis LOCA
 
J.4.1 Rod Drop Accident Results - Three-Assembly Gd2O3 Core - 50% Rod Density
 
PBAPS UFSAR APPENDIX J J-v REV. 21, APRIL 2007 APPENDIX J INDENTIFICATION-RESOLUTION OF AEC-ACRS AND STAFF CONCERNS  LIST OF FIGURES
 
FIGURE TITLE J.4.1 Rod Drop Accident Maximum Control Rod Worth with Single Operator Error at BOL with RSCS
 
J.4.2 Rod Drop Accident Maximum Control Rod Worth with Single Operator Error at Most Reactive Point in Core
 
Life (6,500 MWd/T) with RSCS
 
J.4.3a Schematic Block Diagram of RSCS for Core Exposure
<6,500 MWd/t
 
J.4.3b RSCS Block Diagram for Core Exposure <6,500 MWd/t
 
J.4.4a Schematic Block Diagram of RSCS for Core Exposure
>6,500 MWd/t
 
J.4.4b RSCS Block Diagram for Core Exposure>6,500 MWd/t
 
J.4.5 Fault Tree Model of a Control Rod Stuck in Core Ready for a Drop  
 
J.4.6 Reactor States Used in the Analysis of the Control Rod Drop Accident  
 
J.4.7 Probability Analysis of Control Rod Drop Accident  


J.4.8 Off-Site Effect of Rod Drop Accident}}
J.4.8 Off-Site Effect of Rod Drop Accident}}

Revision as of 20:40, 29 June 2018

Peach Bottom Atomic Power Station, Units 2 & 3, Revision 26 to Updated Final Safety Analysis Report, Appendix J, Identification-Resolution of AEC-ACR and Staff Concerns
ML17130A306
Person / Time
Site: Peach Bottom  Constellation icon.png
Issue date: 04/06/2017
From:
Exelon Generation Co
To:
Office of Nuclear Reactor Regulation
Shared Package
ML17130A259 List: ... further results
References
Download: ML17130A306 (78)
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Text

PBAPS UFSAR APPENDIX J J-i REV. 21, APRIL 2007 APPENDIX J - INDENTIFICATION-RESOLUTION OF AEC-ACRS AND STAFF CONCERNS TABLE OF CONTENTS

J.1 SUMMARY DESCRIPTION

J.2 AREAS SPECIFIED IN THE PEACH BOTTOM UNITS 2 AND 3--

AEC-ACRS CONSTRUCTION PERMIT LETTER J.2.1 Introduction J.2.2 Browns Ferry ACRS Comments (3/14/67) Applicable to Peach Bottom J.2.2.1 Effects of Fuel Failure on CSCS Performance J.2.2.2 Effects of Fuel Bundle Flow Blockage J.2.2.3 Verification of Fuel Damage Limit Criterion J.2.2.4 Effects of Cladding Temperatures and Materials on CSCS Performance J.2.2.5 Quality Assurance and Inspection of the Reactor Primary System J.2.2.6 Control Rod Block Monitor Design J.2.2.7 Plant Startup Program J.2.2.8 Main Steam Line Isolation Valve Testing Under Simulated Accident Conditions J.2.2.9 Performance Testing of the Station Standby Diesel Generator System J.2.2.10 Formulation of an In-Service Inspection Program J.2.2.11 Diversification of the CSCS Initiation Signals J.2.2.12 Control Systems for Emergency Power J.2.2.13 Misorientation of Fuel Assemblies J.2.3 Failure of Conowingo Dam - Alternate Heat Removal Capability J.2.4 Ring Header Leakage Protection Capability J.2.5 Station Thermal Effects - Commonwealth of Pennsylvania Limits J.2.6 HPCIS - Depressurization Capability J.2.7 Station Startup Program

J.2.8 Conclusions

J.3 AREAS SPECIFIED IN THE PEACH BOTTOM UNITS 2 AND 3--

AEC-STAFF CONSTRUCTION PERMIT SAFETY EVALUATION REPORT J.3.1 General J.3.2 AEC-Staff SER Section 2.0 Concerns

J.3.2.1 Introduction J.3.2.2 RPS - IEEE-279 Design Statement J.3.3 AEC-Staff SER Section 3.0 Concerns

J.3.3.1 Introduction J.3.3.2 Station Meteorological Program PBAPS UFSAR APPENDIX J J-ii REV. 21, APRIL 2007 J.3.3.3 Station-Site Slope Cut Program Studies J.3.3.4 Station-Site Flood Protection Studies J.3.3.5 Station-Site Diffusion and Dispersion Studies -

Radiological Effects Determination J.3.3.6 Station Alternative Heat Sink in Event of Dam Failure

- Design Capability J.3.4 AEC-Staff SER Section 4.0 Concerns

J.3.4.1 Introduction J.3.4.2 Suction Piping System Supply Water to ECCS (CSCS) -

Design Aspects J.3.4.3 Adequacy of HPCIS as a Depressurizer J.3.4.4 Engineered Safety Features - Electrical Equipment Inside Primary Containment - Design Capabilities J.3.5 AEC-Staff SER Section 5.0 Concerns

J.3.5.1 Introduction J.3.5.2 Steam Line Break Fuel Rod Integrity - Thermal Hydraulic Analytical Justification J.3.6 AEC-Staff SER Section 6.0 Concerns

J.3.6.1 Introduction J.3.6.2 Development Program of Significance for All Large Water-Cooled Power Reactors J.3.6.3 Development Program of Significance for BWR's in General J.3.6.4 Areas Requiring Further Technical Information

J.3.7 Conclusion

J.4 AREAS SPECIFIED IN OTHER RELATED AEC-ACRS CONSTRUCTION AND OPERATING PERMIT LETTERS J.4.1 General J.4.2 Instrumentation for Prompt Detection of Gross Fuel Failure J.4.3 AEC General Design Criterion No. 35 - Design Intent and Conformance J.4.4 Scram Reliability Study J.4.5 Design Basis of Engineered Safety Features J.4.6 Hydrogen Generation Study J.4.7 Seismic Design and Analysis Models J.4.8 Automatic Pressure Relief System - Single Component Failure Capability - Manual Operation J.4.9 Flow Reference Scram J.4.10 Main Steam Lines - Standards for Fabrication, Q/C, and Inspection J.4.11 Main Steam Line Isolation Valve Leakage J.4.12 Reactor Startup Vibration Testing Capability J.4.13 Control Rod Drop Accident PBAPS UFSAR APPENDIX J J-iii REV. 21, APRIL 2007 J.5 AREAS SPECIFIED IN OTHER RELATED AEC-STAFF CONSTRUCTION OR OPERATING PERMIT SAFETY EVALUATION REPORTS J.5.1 General J.5.2 Tornado and Missile Protection - GE BWR - Spent Fuel Storage Pool J.5.3 BWR System Stability Analysis J.5.4 Reactor Pressure Vessel - Stub Tube Design J.5.5 RPS and CSCS Instrumentation - Cable Marking and Identification J.5.6 RPS and CSCS Instrumentation - Design Criteria (IEEE-279)

PBAPS UFSAR APPENDIX J J-iv REV. 21, APRIL 2007 APPENDIX J - INDENTIFICATION-RESOLUTION OF AEC-ACRS AND STAFF CONCERNS LIST OF TABLES

TABLE TITLE

J.2.1 Summary of Design Basis LOCA Calculations With Fuel Loading Errors

J.2.2 Peak Cladding Temperatures - Design Basis LOCA

J.4.1 Rod Drop Accident Results - Three-Assembly Gd2O3 Core - 50% Rod Density

PBAPS UFSAR APPENDIX J J-v REV. 21, APRIL 2007 APPENDIX J INDENTIFICATION-RESOLUTION OF AEC-ACRS AND STAFF CONCERNS LIST OF FIGURES

FIGURE TITLE J.4.1 Rod Drop Accident Maximum Control Rod Worth with Single Operator Error at BOL with RSCS

J.4.2 Rod Drop Accident Maximum Control Rod Worth with Single Operator Error at Most Reactive Point in Core

Life (6,500 MWd/T) with RSCS

J.4.3a Schematic Block Diagram of RSCS for Core Exposure

<6,500 MWd/t

J.4.3b RSCS Block Diagram for Core Exposure <6,500 MWd/t

J.4.4a Schematic Block Diagram of RSCS for Core Exposure

>6,500 MWd/t

J.4.4b RSCS Block Diagram for Core Exposure>6,500 MWd/t

J.4.5 Fault Tree Model of a Control Rod Stuck in Core Ready for a Drop

J.4.6 Reactor States Used in the Analysis of the Control Rod Drop Accident

J.4.7 Probability Analysis of Control Rod Drop Accident

J.4.8 Off-Site Effect of Rod Drop Accident

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