10.1 Boiling water reactors

Boiling water reactors use light water as both coolant and moderator, generating steam directly in the reactor core. This design allows for lower operating pressures and simpler overall systems compared to pressurized water reactors, though it introduces unique challenges.

BWRs rely on natural circulation driven by density differences between the two-phase mixture in the core and single-phase liquid in the downcomer. Understanding the complex two-phase flow regimes and heat transfer mechanisms in the core is crucial for safe and efficient BWR operation.

Boiling water reactor fundamentals

• Boiling water reactors (BWRs) are a type of light water reactor that uses light water as both coolant and moderator
• BWRs generate steam directly in the reactor core, which is used to drive a turbine and generate electricity
• BWRs operate at lower pressure compared to pressurized water reactors (PWRs), typically around 7 MPa

Light water as coolant and moderator

• Light water (H2O) serves as both coolant and moderator in BWRs
• As a coolant, light water removes heat generated by nuclear fission in the fuel rods
• As a moderator, light water slows down fast neutrons to thermal energies, increasing the probability of fission reactions
• Light water's properties, such as high heat capacity and neutron moderating ability, make it suitable for use in BWRs

Natural circulation in BWRs

Density differences and void fraction

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• Natural circulation in BWRs is driven by density differences between the two-phase mixture in the core and the single-phase liquid in the downcomer
• As water boils in the core, steam bubbles form, reducing the density of the two-phase mixture
• The void fraction, which is the volume fraction of steam in the two-phase mixture, increases with height in the core
• The density difference between the two-phase mixture and the single-phase liquid creates a driving force for natural circulation

Chimney effect for circulation

• BWRs utilize the chimney effect to enhance natural circulation
• The chimney is a tall, cylindrical structure located above the core
• As the two-phase mixture rises through the core and enters the chimney, it expands due to the lower pressure
• The expansion of the two-phase mixture in the chimney creates a buoyancy-driven flow, promoting natural circulation
• The chimney effect helps to maintain a stable and efficient circulation of coolant in the reactor

BWR pressure vessel design

Major internal components

• The BWR pressure vessel contains several major internal components:
  1. Reactor core: Houses the fuel assemblies where nuclear fission occurs
  2. Core shroud: Surrounds the core and directs the coolant flow
  3. Steam separators: Separate steam from the two-phase mixture exiting the core
  4. Steam dryers: Remove moisture from the steam before it enters the main steam lines
  5. Jet pumps: Provide forced circulation during startup and low-power operation

Steam separation and drying

• Efficient steam separation and drying are crucial for BWR operation
• Steam separators use centrifugal force to separate steam from the two-phase mixture
  • The two-phase mixture enters the steam separator tangentially
  • The heavier water droplets are forced to the outer wall and drain back to the downcomer
  • The steam rises through the center of the separator and enters the steam dryers
• Steam dryers remove remaining moisture from the steam
  • Dryers consist of chevron-shaped vanes that cause steam to change direction
  • Moisture droplets impinge on the vanes and drain back to the downcomer
  • Dry steam exits the top of the dryers and enters the main steam lines

Fuel assembly design for BWRs

Fuel rod arrangement and spacing

• BWR fuel assemblies consist of an array of fuel rods arranged in a square lattice
• Typical BWR fuel assemblies have a 7x7, 8x8, 9x9, or 10x10 array of fuel rods
• Fuel rods are spaced using spacer grids, which maintain the proper geometry and prevent rod-to-rod contact
• The spacing between fuel rods is optimized to promote efficient heat transfer and maintain adequate coolant flow

Fuel channel boxes

• Each BWR fuel assembly is enclosed in a fuel channel box
• The fuel channel box is a square, zirconium alloy tube that surrounds the fuel rods
• Functions of the fuel channel box:
  1. Provides structural support for the fuel assembly
  2. Directs coolant flow through the assembly
  3. Separates the coolant flow in the assembly from the bypass flow outside the channel
  4. Helps maintain the proper geometry of the fuel rods
• The fuel channel box also serves as a barrier to prevent cross-flow between adjacent assemblies

Two-phase flow in BWR core

Bubbly and slug flow regimes

• In the lower part of the BWR core, bubbly flow is the dominant flow regime
  • Bubbly flow is characterized by dispersed steam bubbles in a continuous liquid phase
  • As the coolant heats up and more steam is generated, the bubble size and void fraction increase
• As the void fraction increases, the flow transitions to the slug flow regime
  • Slug flow is characterized by large, bullet-shaped steam bubbles (Taylor bubbles) separated by liquid slugs
  • Taylor bubbles occupy a significant portion of the flow channel cross-section
  • Liquid slugs contain smaller bubbles entrained in the liquid phase

Annular flow and dryout

• In the upper part of the BWR core, annular flow becomes the dominant flow regime
  • Annular flow is characterized by a continuous steam core surrounded by a liquid film on the fuel rod surface
  • The liquid film is maintained by the balance between entrainment and deposition of droplets
• As the heat flux increases, the liquid film in the annular flow regime may become depleted, leading to dryout
  • Dryout occurs when the liquid film on the fuel rod surface evaporates completely
  • Dryout can result in a significant decrease in heat transfer efficiency and an increase in fuel rod temperature
  • Predicting and avoiding dryout is crucial for the safe operation of BWRs

Heat transfer in BWR fuel assemblies

Nucleate boiling and critical heat flux

• Nucleate boiling is the primary heat transfer mechanism in BWR fuel assemblies
  • Nucleate boiling occurs when steam bubbles form and detach from the heated fuel rod surface
  • The formation and detachment of bubbles enhance heat transfer by agitating the boundary layer and promoting mixing
• As the heat flux increases, nucleate boiling becomes more vigorous, and the heat transfer coefficient increases
• The critical heat flux (CHF) is the maximum heat flux that can be achieved before the transition to film boiling
  • At CHF, the steam generation rate is so high that the liquid cannot rewet the fuel rod surface
  • The transition to film boiling results in a sudden decrease in heat transfer efficiency and a rapid increase in fuel rod temperature

Post-dryout heat transfer

• Post-dryout heat transfer occurs when the heat flux exceeds the critical heat flux, and the liquid film on the fuel rod surface has evaporated
• In the post-dryout regime, heat transfer is dominated by convection to the steam phase and radiation
• Post-dryout heat transfer is less efficient than nucleate boiling, leading to higher fuel rod temperatures
• Accurate prediction of post-dryout heat transfer is essential for determining the thermal limits of BWR fuel assemblies

BWR thermal-hydraulic limits

Minimum critical power ratio (MCPR)

• The minimum critical power ratio (MCPR) is a thermal limit that ensures the fuel rods operate below the critical heat flux
• MCPR is defined as the ratio of the critical power (power at which CHF occurs) to the actual operating power of the fuel assembly
• Maintaining MCPR above a specified limit prevents fuel damage due to dryout and excessive fuel temperatures
• The MCPR limit is determined by considering uncertainties in power distribution, coolant flow, and other operational parameters

Maximum average planar linear heat generation rate (MAPLHGR)

• The maximum average planar linear heat generation rate (MAPLHGR) is a thermal limit that restricts the average heat flux in the plane of the fuel assembly
• MAPLHGR is expressed in units of power per unit length (e.g., kW/ft)
• The MAPLHGR limit ensures that the fuel operates within acceptable temperature and strain limits during normal operation and anticipated operational occurrences
• Compliance with MAPLHGR limits prevents fuel rod failure due to excessive thermal expansion and cladding strain

BWR core power distribution

Axial and radial power profiles

• The power distribution in a BWR core is non-uniform, with variations in both the axial and radial directions
• The axial power profile is influenced by factors such as:
  1. Control rod positions
  2. Void distribution
  3. Fuel burnup
• The radial power profile is affected by:
  1. Fuel assembly design and enrichment
  2. Core loading pattern
  3. Burnable poison distribution
• Accurate prediction of the axial and radial power profiles is essential for ensuring that thermal limits are not exceeded

Power peaking factors

• Power peaking factors quantify the non-uniformity of the power distribution in the core
• The local peaking factor is the ratio of the maximum local power density to the average power density in the core
• The radial peaking factor is the ratio of the maximum assembly power to the average assembly power
• The axial peaking factor is the ratio of the maximum local power density to the average power density in the same horizontal plane
• Power peaking factors are used to determine the thermal margins and ensure compliance with thermal limits

BWR instability phenomena

Density wave oscillations

• Density wave oscillations (DWOs) are a type of thermal-hydraulic instability that can occur in BWRs
• DWOs are caused by the feedback between the flow rate, void fraction, and pressure drop in the core
• The mechanism of DWOs:
  1. A perturbation in the flow rate leads to a change in the void fraction
  2. The change in void fraction affects the pressure drop across the core
  3. The pressure drop change induces a flow rate change, which amplifies the initial perturbation
• DWOs can result in sustained oscillations of flow rate, power, and other parameters, potentially leading to fuel damage

Coupled neutronic-thermal-hydraulic instabilities

• Coupled neutronic-thermal-hydraulic instabilities involve the interaction between neutron kinetics and thermal-hydraulics in the core
• The mechanism of coupled instabilities:
  1. A perturbation in the void fraction changes the moderator density and affects the neutron moderation
  2. The change in neutron moderation alters the power distribution and heat generation rate
  3. The heat generation rate change affects the void fraction, creating a feedback loop
• Coupled instabilities can lead to regional power oscillations and challenge the thermal limits of the fuel
• Predicting and mitigating coupled instabilities is crucial for the safe and stable operation of BWRs

BWR safety systems

Emergency core cooling system (ECCS)

• The emergency core cooling system (ECCS) is designed to provide cooling to the core in the event of a loss-of-coolant accident (LOCA)
• ECCS consists of several subsystems:
  1. High-pressure coolant injection (HPCI): Provides high-pressure coolant injection during small-break LOCAs
  2. Low-pressure coolant injection (LPCI): Provides low-pressure coolant injection during large-break LOCAs
  3. Core spray system: Sprays water onto the top of the core to provide cooling during LOCAs
• ECCS is activated automatically when certain emergency conditions are detected, such as low reactor water level or high drywell pressure

Reactor core isolation cooling (RCIC)

• The reactor core isolation cooling (RCIC) system is designed to provide cooling to the core during a reactor isolation event
• RCIC is a steam-driven system that operates independently of the main steam system and does not require AC power
• The RCIC system consists of a steam-driven turbine and a pump that injects water into the reactor vessel
• RCIC is activated automatically when the reactor water level drops below a predetermined setpoint
• The system can provide cooling to the core for several hours, allowing time for operators to restore normal cooling systems

BWR vs PWR design comparison

Advantages and disadvantages

• BWRs have several advantages compared to PWRs:
  1. Simpler design due to the absence of steam generators and pressurizer
  2. Lower operating pressure, reducing the thickness of the reactor pressure vessel and piping
  3. Better load-following capabilities due to the direct steam cycle
• However, BWRs also have some disadvantages:
  1. Higher radioactivity in the steam cycle due to direct steam generation in the core
  2. More complex water chemistry control due to the presence of two-phase flow in the core
  3. Larger reactor pressure vessel to accommodate steam separation equipment

Safety and operational considerations

• Both BWRs and PWRs have robust safety systems and multiple barriers to prevent the release of radioactivity
• BWRs rely on natural circulation for core cooling during normal operation, while PWRs use forced circulation with reactor coolant pumps
• PWRs have a secondary steam cycle, which provides an additional barrier between the radioactive primary coolant and the turbine
• BWRs require strict water chemistry control to minimize corrosion and radiation fields in the steam cycle components
• The choice between BWR and PWR depends on factors such as plant size, load-following requirements, and utility preferences