Nuclear Power Plant Accidents

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1 Nuclear Power Plant Accidents Design Objectives: Consistent with probability of occurrence Minimize release of radiation to environment. Minimize damage to plant. Initiating Events: Equipment Failure: Examples-Valve opens, electric power loss, pipe break. Human Error: Examples- Operator closes wrong valve, inadvertent control elements withdrawal Act of Nature: Examples- Earthquake, flood, fire. Barriers to Fission Product Migration Fuel Rod (Pellet, Cladding) Primary Pressure Boundary Containment Structure Failure Modes of Fission Product Barriers Fuel o Center-line Melt o LOCA (Peak Clad Temperature, Zr-H2O Reaction, Burst) o CHF (DNB or Dryout) o Mechanical Forces (PCI, Blowdown Load, Seismic Load) Primary Pressure Boundary o Over-pressurization o Mechanical Forces o Missile Containment o Over-pressurization o Mechanical Forces o Missile o H2 Ignition Means of Accomplishing Objectives Prevent: Minimize probability of initiating event. Mitigate: Provide margin between normal operating conditions and limits. Shut down reactor promptly. Protect fission product barriers via Safeguards System. Recover: Allow prompt recovery by isolating damage and contamination.

2 Accident Analysis Purpose: Predict plant performance during accident conditions, resulting in evaluation of Fuel integrity Primary and secondary (for PWR) system integrity Containment integrity to determine whether offsite radioactive dosage limits are satisfied, as well as other NRC imposed limits. Input to Accident Analysis Computer Models 1. Plant Hardware Description Piping volumes, pump head characteristics, etc Engineering Safeguards Systems Core design information 2. Plant Software Description Control Systems Reactor Protection Systems 3. Plant Accident Initiating Event and Single Active Failure Malfunction, e.g., excessive turbine load, LOCA, {Conservative conditions and assumptions made throughout analysis} Computer models contain equations and correlations describing physical processes. Associated Computer Codes: System Modeling: RETRAN, RELAP, TRAC and TRACE Core T-H Modeling: VIPRE and COBRA Containment Modeling: GOTHIC These codes do not model severe accidents, e.g. fuel melting and containment failure.

3 Nuclear Design/Thermal & Hydraulic Design/Plant Simulation Interface Core T/H Behavior Nuclear Behavior Plant Behavior Generally, nuclear and core T/H characteristics, i.e., reactivity, power distribution, fuel temperature and CHF limits, are provided as input to plant simulation model. Nuclear and T/H inputs to safety analysis: General Data Applies to many accidents Example: Reactivity coefficients, geometric info Specific Data Applies to specific accident Example: F T Q (power peaking factor) during ejected rod accident.

4 Accident Classifications Condition Likelyhood Fuel Damage Allowed Site Dosage 1 Normal Operation No (Just Defects) 10CFR20 2 Once/Year Not expected 10CFR20 3 Once/Plant Life Yes Small part of 10CFR100 4 Design Basis Events Yes 10CFR100 Example: PWR Heatup Accidents BROAD ACCIDENT TYPES Cause: Core power output greater than S/G heat removal. Examples: Core Power = 105% T/G Power = 95% Positive Power Mismatch Reactor Coolant Temp Negative Reactivity Addition Core Power Reduced Unless External (Control Elements) Positive Reactivity Addition Without External (+) Reactivity Source: Coolant Temp Until Core Power = S/G Heat Removal. Concerns: CHF and RCS Overpressurization Cooldown Accidents Cause: Core power output less than S/G heat removal Example: Core Power = 95%, T/G Power = 105% Reverse behavior of heatup accidents Concerns: Center-line Melt and CHF Power Distribution Anomalies Cause: Core power distribution anomaly causes excessive local power peaking. Example: Misloaded fuel assembly, dropped control element. Concerns: Center-line Melt and CHF

5 Inadequate Core Cooling Accidents Cause: Coolant insufficient to remove energy from core due to low pressure, flow or inventory. Example: LOCA (loss of coolant accident) and LOFA (loss of flow accident) Concerns: CHF, Zr-H2O Reaction, Clad Burst, PCT Rapid Core Power Excursions Cause: Rapid increase in core power level due to prompt reactivity addition such that excursion over before impact on loops. Example: Control element ejection Concerns: CHF, Clad Burst and Center-line Melt Following are examples of several accident types for PWR. Accident Excessive Turbine Load Increase Loss of Feedwater and Offsite Power LOCA Type Cooldown Heatup and Inadequate Core Cooling Inadequate Core Cooling

6 SAR Standard Chapter 15 Accident Classifications [Listing only for accidents that involve the NSSS] Increase in Heat Removal by the Secondary System (Cooldown) Decrease in Heat Removal by the Secondary System (Heatup) Decrease in RCS Flow Rate (Inadequate Core Cooling) Reactivity and Power Distribution Anomalies (Power Distribution Anomalies, Rapid Core Power Excursion and Heatup) Increase in Reactor Coolant Inventory Decrease in Reactor Coolant Inventory (Inadequate Core Cooling) Anticipated Transients Without Scram (ATWS)

7 Accidents Analyzed in Safety Analysis Report (SAR) [See Chapter 15 of McGuire and Brunswick Updated Final Safety Analysis Reports (UFSAR) for details.] Example: PWR Condition II Events (Frequency: Once per year) 1. Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from a Subcritical Condition 2. Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power. 3. Rod Cluster Control Assembly Misalignment. 4. Uncontrolled Boron Dilution 5. Partial Loss of Forced Reactor Coolant Flow 6. Startup of an Inactive Reactor Coolant Loop 7. Loss of External Electrical Load and/or Turbine Trip 8. Loss of Normal Feedwater 9. Loss of Offsite Power to the Station Auxiliaries (Station Blackout) 10. Excessive Heat Removal Due to Feedwater System Malfunctions 11. Excessive Load Increase Incident 12. Accidental Depressurization of the Reactor Coolant System 13. Accidental Depressurization of the Main Steam System 14. Inadvertent Operation of Emergency Core Cooling System During Power Operation Condition III Events (Frequency: Once per plant lifetime) 1. Complete loss of forced Reactor Coolant Flow (LOFA) 2. Single Rod Cluster Control Assembly Withdrawal at Full Power 3. Small Break SBA 4. Small Break LOCA Condition IV Events (Design Basis) 1. Major Rupture of a Main Steam Line (SBA) 2. Major Rupture of a Main Feedwater Pipe 3. Single Reactor Coolant Pump Locked Rotor 4. Single Reactor Coolant Pump Shaft Break 5. Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection) 6. Large Break LOCA

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12 Time Sequence of Events for Condition II Events Accident Event Time (sec.) Excessive Load Increase 1. Manual Reactor Control (BOL) 10% step load increase 0 Equilibrium conditions reached 150 (approximate times only) 2. Manual Reactor Control (EOL) 10% step load increase 0 (Does not move control rods Equilibrium conditions reached 100 to maintain programmed (approximate times only) primary coolant temp.) 3. Automatic Reactor Control (BOL) 10% step load increase 0 Equilibrium conditions reached Automatic Reactor Control (EOL) 10% step load increase 0 (withdraws control rods to Equilibrium conditions reached 100 maintain programmed primary coolant temperature) BOL- Beginning of Cycle Life EOL- End of Cycle Life

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17 Time Sequence of Events for Condition II Events Accident Event Time (sec.) Loss of Normal feed water Low-low steam generator water And loss of off site power level reactor trip; reactor coolant To the Station Auxiliaries pumps begin to coast down 0 (Station Blackout) Rods begin to drop 2 Emergency Diesels begin to start 10 Two steam generators begin to receive auxiliary feed from one motor-driven auxiliary feedwater 60 pump (Single Failure) Peak water level in pressurizer occurs 1500

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19 Time Sequence of Events for Condition III Events Accident Event Time (sec.) Complete loss of force Reactor coolant Flow Four pumps in operation, All operating pumps lose power all pumps coasting down and begin coasting down 0 Reactor coolant pump undervoltage trip point reached 0 Rods begin to drop 1.5 Minimum DNBR occurs 2.6

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23 Main Steamline Break Accident Analyzed at EOC since the moderator temperature coefficient is most negative at that time. Analyzed starting at hot zero power, since S/G secondary water inventory maximized and decay heat minimized. Analyzed with break location up-stream of S/G line isolation valve. Safety System Actions SIS initiated Low steam line pressure in one loop Low pressurizer pressure High containment pressure SIS signal trips Rx if critical and initiates FW isolation S/G isolation initiated Low steam line pressure High containment pressure High (-) steam line pressure rate signal

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30 Large Break LOCA Based upon best estimate TRAC code prediction (W) 4 loop plant

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35 Sequence of Events for W PWR LOCA Calculation Event Time (s) Transient started: 0.0 Charging pump flow on Power decay initiated (Reactor trip begins) 0.2 Broken-loop (Loop 3) accumulator flow started 3.1 S/G feedwater flow terminated 4.5 Safety injection flow initiated 5.0 Intact loop accumulator flows initiated 13.9 (Loops 1, 2, and 4) Charging pump flow terminated (Runs but realigned to ECCS) 15.0 Initial ECCs entry into lower plenum 25.0 Peak intact loop accumulator flows End of blowdown 26.1 Lower plenum refilled and reflood begins 36.8 Pressurizer empty 37.0 Broken loop (Loop 3) accumulator liquid flow ended and nitrogen flow begins 47.1 Intact loop accumulator liquid flows ended and nitrogen flow initiated 54-59

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47 BWR Accidents Loss of F/W Heating w/ Manual Flow Control Transient: F/W Heating T Voids Core Reactivity Core P M steam In HO 2 Rel Trip: High flux Safety concern: Minimum Critical Power Ratio (MCPR) => OK

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49 Prz Regulator Failure (Open) Transient: Prz Reg Failure (Open) Turbine Valves Open RCS Prz Void Core Reactivity Core PRel Prz < 825 psia MSIVs close Trip: MSIVs close High R/V level Safety Concern: MCPR => OK

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51 Generator Load Rejection- No Bypass Transient: Turbine Valves Close RCS Prz Voids Core Reactivity Core PRel (Prompt Critical - Almost) Trip: Turbine Valve Closes Safety Concern: MCPR => OK R/V Prz < 1375 psig (Relief valves open to discharge to suppression pool.)

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53 Trip of Both Recirc. Pumps Transient: Recirc. Pumps Trip Core Flow Void Core Reactivity Core PRel Trip: May trip on High R/V Level, but no credit taken. Safety Concern: MCPR => OK

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55 LOCA (Break in Recirc. Line Downstream of Pump) Transient: Trip: Recirc Line Break Blowdown of RCS RCS Prz & Level Single Failure: LPCI Lost Other LPCI Train ineffective because of break location Safety Systems Available: HPCI, CS, ADS High Prz in Drywell Low R/V Level } SIS + ADS Safety Concern: PCT< 2200 o F => OK

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