ML21286A252
| ML21286A252 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/04/2021 |
| From: | Tennessee Valley Authority |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML21286A574 | List:
|
| References | |
| Download: ML21286A252 (10) | |
Text
BFN-21 TABLE 1.4-1 BWR SAFETY ENGINEERING CONCEPT FOR CLASSIFICATION OF BWR SYSTEMS, CRITERIA AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION Type of Operation Safety Considerations Power Generation Considerations or Event
- 1. Planned Operation In this category are classified the unacceptable safety results, In this category are classified the unacceptable results for criteria, plant actions, systems, and operational requirements power generation, criteria, plant actions, systems and pertinent to safety during planned operation. This space operational requirements pertinent to the production of represents the aspects of the BWR which must be considered to electrical power during planned operation. Process assure that the BWR operator can operate the plant within systems and normal operational procedures would be specified safety limitations. Certain process indicators, process classified here.
variable limits and limits on the release of radioactive material would be classified here.
- 2. Abnormal Operational In this category are classified the unacceptable safety results, In this category are classified the unacceptable results for Transients criteria, plant actions, systems and operational requirements power generation, criteria, plant actions, systems and pertinent to safety in regard to abnormal operational transients. operational requirements pertinent to the ability to produce Certain protection systems, safety limits, and limiting safety electrical power as that ability is affected by abnormal system settings would be classified here. operational transients. Certain systems not used for planned operation would be classified here.
- 3. Accidents In this category are classified the unacceptable safety results, In this category are classified the unacceptable results for criteria, plant actions, systems and operational requirements power generation, criteria, plant actions, systems and pertinent to safety in regard to accidents. Engineered operational requirements pertinent to the ability to produce safeguards would be classified here. electrical power as that ability is affected by accidents.
Design considerations and post-accident procedures provided to enable the plant to be used for power generation after an accident would be classified here.
- 4. Special Event In this category are classified the unacceptable safety results, In this category are classified the unacceptable results for criteria, plant actions, systems and operational requirements power generation, criteria, plant actions, systems and pertinent to safety in regard to the stated special event. Safety operational requirements pertinent to the ability to produce systems provided especially for the special event would be electrical power as that ability is affected by the stated classified here. special event. Systems and procedures provided to enable the plant to be returned to power operation following the special event would be classified here.
BFN-21 Table 1.4-2A (Sheet 1)
BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION SAFETY CONSIDERATIONS Types of Requirements to Types of Actions Required be Observed in Operation Type of Operation Types of Applicable to Avoid Unacceptable Types of Systems Required of Plant to Avoid or event Unacceptable Safety Results Criteria Results to Carry Out Action Unacceptable Results
- 1. Planned 1-1 The release of Nuclear Safety Design Safety Action- Safety Systems- Operational Nuclear Safety Operation radioactive material Criteria-Type S-1 Type S-1 Type S-1 Requirements-Type S-1 to the environs to such an extent that the Nuclear Safety Process Safety Action Process Safety Systems Operational Nuclear Safety limits of 10CFR20 Operational Criteria- (A Category of Safety (A Category of Safety Limits-Type S-1 are exceeded. Type S-1 Action) Systems)
Process Safety Indication of Process Indicators Technical Specifications-1-2 Fuel failure to such an Design Criteria Variables Type S-1 extent that were the freed fission products Process Safety Rod Worth Monitoring Rod Worth Minimizer Process Safety Limits to the environs via the Operational Criteria Program of Process Computer normal discharge paths Rod Pattern Control Limiting Conditions for for radioactive material, Various Industry Radwaste Systems for Operation for Indicators limits of 10CFR20 Codes Control of Process would be exceeded. Variables Process Radiation Monitors Radioactive Material Radwaste Criteria Release Limits Control Rod Control Control Rod Drive System 1-3 Nuclear System stress Loading Criteria Refueling Block in excess of that (Normal Conditions) Reactor Manual Rod Pattern Limits allowed for planned Control Rod Control Control System operation by applicable industry codes. Refueling Block Limiting Conditions for Refueling Interlocks Operation for Radwaste Systems Core Shutdown Control 1-4 The existence of a plant Reactor Protection System condition not considered Radwaste (Manual Scram) Nuclear System Leakage by plant safety analysis. Limits Isolation Condensate Storage System Neutron Monitoring System
BFN-21 Table 1.4-2A (Sheet 2)
BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION SAFETY CONSIDERATIONS Types of Requirements to Types of Actions Required be Observed in Operation Type of Operation Types of Applicable to Avoid Unacceptable Types of Systems Required of Plant to Avoid or Event Unacceptable Safety Results Criteria Results to Carry Out Action Unacceptable Results
- 2. Abnormal 2.1 The release of radioactive Nuclear Safety Design Safety Action-Type S-2 Safety Systems-Type S-2 Operational Nuclear Operational material to the environs Criteria-Type S-2 Safety Requirements-Transients to such an extent that Scram Protection System (Generic Type S-2 the limits of 10CFR20 Nuclear Safety Term) are exceeded. Operational Criteria- Pressure Relief Type S-2 Nuclear Safety Systems Operational Nuclear Safety Core Cooling (A Category of Protection Limits-Type S-2 Systems) 2.2 Any fuel failure calculated as a result of the transient. Various Industry Codes Containment Reactor Protection System Technical Specifications-Cooling (RHRS) (Scram) Type S-2 IEEE-279 Primary Containment Control Rod Drive System (Scram) Safety Limits Secondary Availability Goals Containment Neutron Monitoring System (IRM, APRM) Limiting Safety System Loading Criteria Settings (Upset Conditions) Pressure Relief System 2.3 Nuclear system stress in Limiting Conditions for excess of that allowed for Single Failure Reactor Vessel Isolation Operation for Protection transients by applicable Criterion Control System Systems industry codes.
Testability Criteria High Pressure Coolant Surveillance Requirements Injection System for Protection Systems D-C Power System Standby A-C Power
BFN-21 Table 1.4-2A (Sheet 3)
BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION SAFETY CONSIDERATIONS Types of Requirements to Types of Actions Required be Observed in Operation Type of Operation Types of Applicable to Avoid Unacceptable Types of Systems Required of Plant to Avoid or Event Unacceptable Safety Results Criteria Results to Carry Out Action Unacceptable Results
- 3. Accidents 3-1 Radioactive Nuclear Safety Design Safety Action-Type S-3 Safety Systems-Type S-3 Operational Nuclear Safety material release Criteria-Type S-3 Requirement-Type S-3 tosuch an extent that the guideline Scram Protection Systems (Generic values of 10CFR50.67 Nuclear Safety Term) Operational Nuclear Safety would be exceeded. Operational Criteria - Core Cooling Limits-Type S-3 Type S-3 Engineered Safeguards 3-2 Fuel cladding temperatures Containment Reactor Protection System Technical Specifications-in excess of 2200F for Control Rod Drive System Type S-3 Pipe Breaks Neutron Monitoring System 3-3 Nuclear system Various Industry Codes Pressure Relief System (Main pressure in excess Containment Cooling Steam Relief Valves) Limiting Safety System of that allowed for IEEE-279 Reactor Vessel Isolation Settings accidents by applicable Control System industry codes. Availability Goals Stop Control Rod Primary Containment Isolation Limiting Conditions for Ejection Control System Operation for Protection Primary Containment Systems Secondary Containment 3-4 Containment stresses Loading Criteria Limit Reactivity Main Steam Line Isolation sufficient to produce (Emergency and Insertion Rate Valves Surveillance Requirements containment failure Faulted Conditions) Main Steam Line Flow Restrictor for Nuclear System when containment is Pressure Relief High Pressure Coolant Injection required. System Automatic Depressurization 3-5 Overexposure to Single Failure Criteria Reactor Vessel System radiation of operating Isolation Low Pressure Coolant Injection personnel in the Testability Criteria Core Spray System control room. Primary Containment RHRS (Containment Cooling)
Isolation Control Rod Velocity Limiter
BFN-21 Table 1.4-2A (Sheet 4)
BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION SAFETY CONSIDERATIONS Types of Requirements to Types of Actions Required be Observed in Operation Type of Operation Types of Applicable to Avoid Unacceptable Types of Systems Required of Plant to Avoid or Event Unacceptable Safety Results Criteria Results to Carry Out Action Unacceptable Results
- 3. Accidents 3-5 Overexposure Testability Criteria Secondary Containment Control Rod Drive Housing Surveillance requirements (Cont.) to radiation Isolation Supports for nuclear systems of operating Standby Gas Treatment System personnel in Standby A-C Power System the control Treatment of Fission D-C Power System room. Products Main Steam Line Radiation Monitoring System Restriction of Coolant Reactor Building Ventilator Loss Rate Radiation Monitoring System RHR Service Water System Control Room Isolation 3-6 Peak enthalpy of fuel in excess of 280 cal/gm for the control rod drop accident.
BFN-21 Table 1.4-2A (Sheet 5)
BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION SAFETY CONSIDERATIONS Types of Requirements to Types of Actions Required be Observed in Operation Type of Operation Types of Applicable to Avoid Unacceptable Types of Systems Required of Plant to Avoid or Event Unacceptable Safety Results Criteria Results to Carry Out Action Unacceptable Results
- 4. Special Event 4-1 The inability to Nuclear Safety Design Safety Action - Type S-4 Safety Systems-Type S-4 Operational Nuclear Safety Loss of bring the reactor Criteria-Type S-4 Requirements-Type S-4 Habitability to the shutdown Special Safety Action Special Safety Systems of the Control condition by manip-Room ulation of the local controls and Nuclear Safety Operational Nuclear Safety equipment which Operational Criteria- Shutdown From Outside Local Controls Outside Limits-Type S-4 are available out- Type S-4 Control Room Control Room side the control room. Special Safety Design Technical Specifications-Criteria Cooldown from Outside Local Indicators Outside Type S-4 Control Room Control Room 4-2 The inability to bring the reactor Special Safety Limiting Conditions for to the cold shut- Operational Criteria Condensate Storage System Operation for Special Safety down condition from Systems outside the control room. Reactor Core Isolation Surveillance Requirements for Cooling System Special Safety Systems Pressure Relief System Reactor Protection System Control Rod Drive System RHR (containment Cooling)
BFN-21 Table 1.4-2A (Sheet 6)
BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION SAFETY CONSIDERATIONS Types of Requirements to Types of Actions Required be Observed in Operation Type of Operation Types of Applicable to Avoid Unacceptable Types of Systems Required of Plant to Avoid or Event Unacceptable Safety Results Criteria Results to Carry Out Action Unacceptable Results
- 5. Special Event- 5-1 The inability to Nuclear Safety Design Safety Action-Type S-5 Safety Systems-Type S-5 Operational Nuclear Safety Inability to shut down the Criteria-Type S-5 Requirements-Type S-5 Shut Down reactor independent Special Safety Action Special Safety Systems Reactor With of control rods Operational Nuclear Safety Control Rods Limits-Type S-5 5-2 The inability to Nuclear Safety maintain the Operational Criteria Shutdown Without Control Standby Liquid Control reactor in the Type S-5 Rods System shutdown condition Technical Specifications-independent of Special Safety Design Maintain Shutdown During RWCU Isolation Type S-5 control rods Criteria Reactor Cooldown Limiting Conditions for Operation for Special Safety Systems Special Safety Operational Criteria Surveillance Requirements for Special Safety Systems
BFN-21 Table 1.4-2B (Sheet 1)
BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION POWER GENERATION CONSIDERATIONS Unacceptable Results for Types of Actions Types of Systems Types of Requirements Power Generation (Where Required to Avoid Required to Avoid to be Observed in More Restrictive Than Unacceptable Results Unacceptable Results Operation of Plant Type of Operation Unacceptable Safety Types of Applicable (Where Not Required (Where Not Required to Avoid Unacceptable or Event Results) Criteria as a Safety Action) as a Safety Action) Results
- 1. Planned Operation 1-1 Inability to generate Power Generator Design Power Generation Action- Power Generator Systems- Operational Power Generator electrical power Criteria-Type PG-1 Type PG-1 Type PG-1 Requirements-Type PG-1 1-2 Fuel Failure Power Generator Operational Power Generator Operational Criteria - Limits-Type PG-1 Type PG-1 1-3 Inability to Perform Process Action Process Systems Routine Maintenance (A Category of Power (A Category of Power with Plant at Power Generation Action) Generator Systems)
Process Design Criteria Normal Operating Procedures 1-4 Inability to Optimize Fuel Performance Process Operational Indications of Process Indicators Maintenance Procedures Criteria Variables 1-5 Inability to Respond Process Operations Process Computer System Calibration Procedures to Changes in Power Demand Fuel Performance Recirculation Flow Control Refueling Procedures Calculations System 1-6 Inability to Shut Down Power Level Control Reactor with Control Reactor Manual Control Rods in the Normal Consideration of Exhaust System Manner Steam Control Rod Drive System Feedwater System Turbine-Generator Main Condenser
BFN-21 Table 1.4-2B (Sheet 2)
BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION POWER GENERATION CONSIDERATIONS Unacceptable Results for Types of Actions Types of Systems Types of Requirements Power Generation (Where Required to Avoid Required to Avoid to be Observed in More Restrictive Than Unacceptable Results Unacceptable Results Operation of Plant Type of Operation Unacceptable Safety Types of Applicable (Where Not Required (Where Not Required to Avoid Unacceptable or Event Results) Criteria as a Safety Action) as a Safety Action) Results
- 2. Abnormal 2-1 Fuel Failure Power Generation Design Power Generation Action Power Generation Systems Operational Power Generation Operational Criteria-Type PG-2 Type PG-2 Type PG-2 Requirements-Type PG-2 Transients 2-2 The Lifting of Main Steam Power Generation Operational Power Generation Relief Valves Operational Criteria Limits-Type PG-2 Type PG-2 2-3 Conditions Requiring the Rod Block Reactor Manual Control Opening of the Reactor System (Rod Block)
Vessel for Inspection or Pressure Relief Repair Pressure Relief System Normal Operating Procedures Refueling Block 2-4 Inability to Return to Scram Refueling Interlocks Post Transient Recovery Power Operation Procedures 2-5 Inadvertent Criticality Core Cooling Reactor Protection Refueling Restrictions During Refueling System (RPS)
Electro Hydraulic Control (EHC) System Reactor Core Isolation Cooling (RCIC)
BFN-21 Table 1.4-2B (Sheet 3)
BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION POWER GENERATION CONSIDERATIONS Unacceptable Results for Types of Actions Types of Systems Power Generation Required to Avoid Required to Avoid Types of Requirements (Where More Restrictive Unacceptable Results Unacceptable Results to be Observed in Operation Type of Operation Than Unacceptable Types of Applicable (Where Not Required (Where Not Required of Plant to Avoid or Event Safety Results) Criteria as a Safety Action) as a Safety Action) Unacceptable Results
- 3. Accidents 3-1 Inability to Return Power Generation Design Power Generation Action - Power Generation Systems - Operational Power Generation to Power Operation Criteria-Type PG-3 Type PG-3 Type PG-3 Requirements-Type PG-3 Power Generation Operational Power Generation Operational Criteria - Limits-Type PG-3 Type PG-3 Post Accident Recovery Procedures
- 4. Special Event 4-1 Inability to Return Power Generation Design Power Generation Action- Power Generation Systems- Operational Power Generation Loss of to Power Operation Criteria-Type PG-4 Type PG-4 Type PG-4 Requirements-Type PG-4 Habitability of the Control Power Generation Operational Power Generation Room Operational Criteria - Limits-Type PG-4 Type PG-4 Post Event Recovery Procedures
- 5. Special Event 5-1 Inability to Return Power Generation Design Power Generation Action - Power Generation Systems- Operational Power Generation Inability to to Power Operation Criteria-Type PG-5 Type PG-5 Type PG-5 Requirements-Type PG-5 Shut Down Reactor With Control Rods Power Generation Operational Power Generation Operational Criteria - Limits-Type PG-5 Type PG-5 Post Event Recovery Procedures