Nuclear Reactor Safety - PowerPoint by Y92vtGW


									 Nuclear Reactor Safety

Technology of Accident Analysis


         Winter 2004

●   Reactor Physics Review
●   Fuel Behavior
●   Thermalhydraulics
●   Design-Basis Accidents
●   Severe Accidents in CANDU Reactors
Neutron Cycle
            Criticality Equation

   η.ε.p.f.Pf.Pth = keff = 1
η = fast neutrons emitted per absorption in fuel
ε = neutrons produced by fast fission in U-238
p = resonance escape probability
f = thermal neutrons absorbed in fuel
Pf = fast neutron non-leakage probability
Pth = thermal neutron non-leakage probability
keff = effective multiplication factor
kinf = η.ε.p.f
 Non-Leakage Probabilities

Pf = exp(-B2 . τ)

Pth= 1/(1+B2.L2)

B2 = Buckling, cm-2
  τ = Fermi Age, cm2
  L = thermal diffusion length, cm
Materials Buckling:
 B2m = (υ.Σf – Σa)/D

Geometric Buckling, Cylindrical Geometry
 B2g = (2.405/R)2 + (π/H)2
 υ = fast neutrons produced per fission
 Σf = macroscopic fission cross-section, cm-1
 Σa = macroscopic absorption cross-section, cm-1
 D = neutron diffusion coefficient, cm
 R = reactor core radius, cm
 H = reactor core length, cm
         CANDU Core Parameters

          Fresh Fuel    Equilibrium Fuel

η          1.2433         1.1990
ε          1.0277         1.0276
p          0.9036        0.9039
f          0.9315        0.9375
kinf       1.0755         1.0440
τ, cm2     155           155
L, cm      15.55         14.95
         Neutron Diffusion Equation

  One energy group, one dimension, steady-state

           D.d2Φ/dx2 + [υ.Σf- Σa].Φ = 0

Φ = neutron flux, n/(cm2.s)
Σa = macroscopic absorption cross-section, cm-1
Σf = macroscopic fission cross-section, cm-1
υ = neutrons per fission
D = neutron diffusion coefficient, cm
x = distance, cm
Fuel Critical Mass and Implications
         Coolant Void Reactivity (CVR)

●   CANDU “over-moderated”
●   Voiding of channels in a LOCA reduces
    moderator/uranium ratio
●   Core reactivity goes positive, power increases
●   Requires two independent, reliable, diverse
    shutdown systems
●   LWR “under-moderated”, opposite behavior
●   ACR designed for negative CVR
    Point Kinetics Equations

dN(t)/dt = (ρ-β).N(t)/l* +Σ [λi.Ci(t)]

dCi(t)/dt = βi.N(t)/l* - λ iCi(t)

    β = Σ βi , i = 1 to 6

    ρ = (k - 1)/k
         Importance of Delayed Neutrons

●   Small fraction of neutrons (~ 0.6% in CANDU)
    are emitted by decay of fission products
●   Characterized by six groups with half lives from
    0.2 to 53.7 seconds, compared to prompt neutron
    lifetime of about 10- 4 seconds in CANDU
●   Also, in CANDU, photoneutrons are produced in
    the moderator by core gamma rays, with average
    half life of 16.7 minutes
●   Delayed neutrons make reactor control practical
       Neutron Flux Behavior in CANDU

●   Neutron flux varies spatially in normal
    operation because of leakage, adjuster rods,
    zone controllers, fuelling, xenon

●   Zone controllers needed to prevent xenon
    oscillations caused by on-power fuelling which
    could cause bundle power limits to be exceeded

●   Neutron flux tilts occur in accidents because of
    voiding of one loop and insertion of shutoff rods
    (SDS1) from the top
           Analysis of Neutron Behavior

●   Diffusion Theory: Simplified model, analogous to heat
    conduction. Neutron flux function of position only

●   Transport Theory: Neutron flux function of position,
    energy, direction

●   Diffusion theory breaks down at exterior boundaries,
    near or within strong absorbers and at interfaces
    between dissimilar materials.
    Industry Standard Toolset (IST) Computer
●   Single set of key computer codes for licensing
    and safety analysis

●   Validation, quality assurance, maintenance

●   Meet CNSC quality assurance standards: CNSC
    regulatory guides and CSA standards

●   AECL responsible for certain codes, OPG for
CANDU Reactor Physics IST Codes
●     Transport theory cell codes for fuel bundle,
    pressure tube, calandria tube, control mechanisms

●     Diffusion theory core code; models neutron
    diffusion from cell to cell
                     Decay Heat

●   After reactor trip, power still produced by
    radioactive decay of fission products
●   For trip after long operation, the decay power as a
    fraction of initial power can be approximated by:
    P(t)/Po = 6.6 x 10-2 t-2 , with t in seconds
●   Relatively slow decay results in sizeable heat load
    for a long time after shutdown.
            Heat Flow in Fuel Elements (1)

Conduction in fuel:
  -kf .[δ2T/δr2+1/r.(δT/δr)] = ρ.C.δT/δt + Qv (Transient)
  -kf .[d2T/dr2+1/r.(dT/dr)] = Qv (Steady state)
  To – Tfo = Qv.R2/(4.kf) (S.S., integrated, Qv & kf const.)
Heat transfer across gap (fission product gases plus fill gas
 & contact conductance):
  q = hg.(Tfo – Tsi)
Heat transfer across sheath:
  q = ks .(Tsi – Tso)/w   (Steady state)
        Heat Flow in Fuel Elements (2)

T = temperature, K
Qv = heat generation in fuel,Wm-3
q = heat flux, Wm-2
r = radius, m    w = sheath thickness, m     t = time, s
kf = fuel conductivity, Wm-1K-1
ks = sheath conductivity, Wm-1K-1
hg = gap conductance, Wm-2K-11
ρ = fuel density, kg.m-3   C = fuel specific heat,
        Location of Fission Products

• Fission products
  formed within
  fuel grains
• Diffuse                       <10%   >90%
   – Bound inventory   Fission
     – in grains       move this way
   – Grain boundary    increasing
     inventory         temperature
                       & burnup
   – Gap inventory
       Fuel Behavior under Irradiation

•   Cracking
•   Swelling
•   Dishing/ridging
•   Gas pressure increase
•   Pellet-clad interaction

                              Un-irradiated       Irradiated
                                  Fuel        Fuel (exaggerated)
            Fuel Behavior in Accidents

Key Safety Parameters:
●   Fuel temperature
    * Potential sheath failure
    * Potential pressure tube failure (Creep rupture,
    * Limited effect on physics
●   Fuel sheath integrity
●   Fission product inventory and release
         Fuel Behavior in Power Increase

●   Temperature gradient across fuel pellet causes
    tensile stress in outer region.
●   As power, and thus pellet temperature, increase,
    pellet cracks increase.
●   As power and temperature rise further, central
    melting may begin. (UO2 MP = 2840 oC)
●   Gas and volatile fission products release rates will
       IST Codes for Fuel Behavior
Models micro-structural, mechanical and thermal
behavior of CANDU fuel element under normal
operating conditions. Used to quantify pre-
accident conditions.

Models thermo-mechanical behavior of a
CANDU fuel element under accident conditions.

●   Two-Phase Flow Patterns
●   System Thermalhydraulics Model
●   Critical Heat Flux (DNB, Dryout)
●   Post-Dryout Heat Transfer
●   Two-Phase Critical Flow
●   Two-Phase Flow Instabilities
Two-Phase Flow
Lamari & Rogers Horizontal Flow Regime Map
   Jg*= Gg/[g.d.ρg.(ρf-ρg)]0.5; X= (ρg/ρf)0.5.(μf/μg)0.1.[(1-x)/x]0.9
               Two-Phase Flow Models

●   Homogeneous model: Pseudo-fluid with averaged
    properties of vapor and liquid.
    Void fraction, α = Ag/At= 1/[1+ ρg .(1-x)/(ρf.x)]
●   Slip flow model: allows for relative velocity (slip)
    between phases.
    α = 1/[1+ S.ρg .(1-x)/(ρf.x)] , S = ug/uf
●   Separated flow model or two-fluid model (non-
    equilibrium model): vapor and liquid flow at
    different temperatures and velocities
    CANDU System Thermalhydraulics Model
●   Equations of mass, momentum and energy
    conservation for each phase in transient two-
    phase flow in a one-dimensional network
    + liquid and vapor plus non-condensable gases
    + non-equilibrium between phases
    + parallel and series paths

●   Equations of state for each phase and n-c gases
    CANDU System Thermalhydraulics Model
●   Component models for fuel, fuel channels,
    feeders, headers, piping, valves, pumps, steam
    generators, secondary system, etc.
●   Correlations for pressure drop, void fraction, heat
    transfer, critical heat flux
●   Models of plant controllers
               Conservation of Mass

Liquid phase:
δ[ρf .A.{(1-α).uf}]/δz + δ[ρf.A.(1-α)]/δt = - Γ.A

Vapor phase:
δ[ρg.A.{α.ug}]/δz + δ[ρg.A.α]/δt = Γ.A
           Conservation of Momentum

Liquid phase:
δ[ρf.A.{(1-α).uf2}]/δz + δ[ρf.A.{(1-α).uf}]/δt =
-(1-α).A.δp/δz + ρf.g.sinθ.(1-α).A - τwf.Pwf + τi.Pi -Γ.ufi.A

Vapor phase:
δ[ρg.A.{α.ug2}]/δz + δ[ρg.A.{α.ug}]/δt =
-α.A.δp/δz + ρg.g.sinθ.α.A - τwg.Pwg - τi.Pi + Γ.ugi.A
             Conservation of Energy

Liquid phase:
δ[ρf.A.{(1-α).uf.hf}]/δz + δ[ρf.A.{(1-α).hf}]/δt =
A.(1-α).δp/δt +qf".Phf + qif".Pi – Γ.hfi.A

Vapor phase:
δ[ρg.A.{α.ug.hg}]/δz + δ[ρg.A.{α.hg}]/δt =
A.α.δp/δt + qg".Phg – qig".Pi + Γ.hgi.A
   Evaporation Term and Equation of State

Γ = [(qig" – qif").Pi + τi.Pi.(ug – uf)]/[(hg – hf).A]

ρf , ρg = f(p,T) Steam Tables for water, heavy
               Closure Equations

To solve the conservation equations, data or
correlations are needed for void fraction (α),
wetted and heated wall perimeters for each phase
(Pwf, Pwg, Phf, Phg), interface perimeter (Pi), wall
and interface shear stresses (τwf, τwg, τi), and wall
and interface heat transfer rates (qf", qg", qif", qig").
Empirical correlations are available for void
fraction, shear stresses and heat transfer rates.
         Conservation of Momentum

Steady state, horizontal flow, no area change, no
  heat addition (i.e., no evaporation and thus no
  momentum change)
Liquid phase:
 - (1-α).A.[dp/dz]FTP – τwf.Pwf + τi.Pi = 0
Vapor phase:
  - α.A.[dp/dz]FTP – τwg.Pwg – τi.Pi = 0
Wall and interface shear stresses can be expressed
 in terms of frictional pressure gradient
     Empirical Correlations for [dp/dz]FTP

For convenience, the frictional two-phase pressure
  gradient, [dp/dz]FTP is generally expressed as:
  [dp/dz]FTP = Φfo2.[dp/dz]fo
 where [dp/dz]fo is the frictional pressure gradient for the
 liquid flowing alone at the same mass flow rate as the
 total two-phase flow rate and Φfo2 is the two-phase flow
 Empirical correlations for Φfo2 include those of Lockhart
 & Martinelli (1949), Friedel (1979) and others. Friedel is
     Finite Difference Modelling of System

●   System represented by nodes, containing mass
    and energy, and links joining the nodes

●   Mass and energy conservation equations for

●   Momentum conservation equations for links
            Typical Link-Node Structure
• Break the circuit up into
   – nodes containing mass
   – links joining the nodes
• Mass & energy conservation equations for nodes
• Momentum equation for links
                         Wk, Lk, Ak
                         f k, Dk, k etc.
  Mi, i, Pi, ei, Q i,                      Mj, j, Pj, ej, Q j,
  etc                                       etc
   Thermalhydraulics Codes for CANDU
 One-dimensional, steady-state or transient, two-fluid model of
 two-phase flow in piping networks, including allowance for
 one to four non-condensable gases. Also models heat transfer
 with solid surfaces. Used for design, licensing and safety
 Models steady-state or transient single- or two-phase flows in
 sub-channels of fuel bundles.
 Models thermalhydraulic behavior of CANDU reactors using
 homogeneous model. Used for design and operational studies.
          Heat Transfer in Boiling Flows

●   Effect of flow regime

●   Nucleate boiling [Rohsenow (1952), Forster &
    Zuber (1955)]

●   Forced convection boiling, nucleation suppressed
    [Chen (1963), Steiner & Taborek (1992)]
              Critical Heat Flux (CHF)

●   Surface cooling changes from liquid cooling to
    vapor cooling
●   In pool boiling or flow boiling at low velocities,
    increased vapor flow rate away from surface
    prevents liquid from reaching surface (Departure
    from nucleate boiling, DNB)
●   In flow boiling at higher velocities, thickness of
    annular liquid film on wall decreases until film
    disappears (Dryout)
Dryout in Flow Boiling (1)
            Dry Out in Flow Boiling (2)
●   In annular flow, liquid film keeps surface well-
●   Film thickness controlled by evaporation (-),
    wave entrainment (-), droplet mass transfer (+)
●   Dryout occurs when film thickness goes to zero
    [Hewitt (1970)]
●   Integrated effect over boiling length (BLA
●   Initiation of boiling length (saturation? OSV?).
    (OSV correlation Rogers & Li, 1992)
    Critical Heat Flux in CANDU Fuel Bundles

●   Different power inputs, flow rates, qualities in
    different subchannels
●   Flow and quality distribution also affected by
    mixing between subchannels, bundle end-plates,
    randomly aligned bundles
●   Dryout initiates in one or more subchannels
●   CHF data for CANDU bundles from experiments
    in full-scale electrically heated horizontal bundle
    strings with water or Freon
      Predictions of CHF in CANDU Fuel

●   CHF Look Up tables developed by AECL and
    University of Ottawa. See handout for CHF in
    tubes, Groeneveld et al, 1986

●   CHF Look Up tables for CANDU 37-element and
    CANFLEX fuel channels

●   CHF correlations for CANDU fuel channels. See
    handout for problem assignment.
          CHF in a CANDU Channel

• CHF determined
                                Heat flux
   – no reliable theory for     (overpower)
     needed accuracy
• Local flux shape means                      Heat flux
                                              (normal) Dryout
  dryout is not at the end
• How can we change the                          Quality

  flux shape to improve                Distance along channel
                              Inlet                               Outlet
     Effects of Critical Heat Flux with Water

●   For DNB, sheath temperature jump is very large,
    several thousand degrees. Surface “burn out”

●   For dryout, sheath temperature jump is moderate
    for normal CANDU conditions. Sheath corrosion
    rate increases and sheath may fail.
         Post-Dryout (PDO) Heat Transfer

●   Forced convection heat transfer to vapor

●   Heat transfer by droplet impingement on wall

●   Heat transfer to droplets in boundary layer

●   Groeneveld correlation (1973) (conservative)

●   AECL Look Up tables
PDO Heat
     Critical Discharge in Two-Phase Flows

●   For compressible gas flows, discharge rate
    through an opening is a maximum for sonic
    velocity. Further reduction of downstream
    pressure does not increase flow rate since
    pressure waves cannot penetrate upstream.
●   Similar behavior for two phase flows. Moody
    (1965) diagram for thermodynamic equilibrium
    flows based on slip flow model.
●   For two-phase discharge through very short paths
    non-equilibrium may exist (metastable flow);
    discharge rate will be much higher
Moody Diagram for Two-Phase Critical
            Two-Phase Flow Instability

●   Static Instability (Ledinegg instability)

●   Dynamic Instability (Density wave instability)
             Density Wave Instability (1)

●   Boiling channel with liquid at the inlet and inlet and
    outlet pressures fixed (e.g., many parallel channels).
●   Frictional pressure drop in the liquid region in phase
    with any inlet flow perturbation.
●   Friction and momentum pressure drops in the two-phase
    region not in phase with any inlet flow perturbation and
    impose a feedback effect on the liquid region that may
    enforce or attenuate the inlet flow perturbation.
●   Flow is stabilized by high pressure, increase of liquid
    phase or inlet pressure drop, increase of liquid region
    length, high velocity.
          Density Wave Instability (3)

Guido model: homogeneous flow, simplified
Phase Change Number:
NPCH = P.(ρf/ρg -1)/(w.hfg)
Subcooling Number:
NSUB = (hfs - hi).(ρf/ρg -1)/hfg
Pressure Loss Coefficients:
k = Δp/(ρ.u2)
    Flow Instability in a CANDU Power Plant

●   CATHENA can predict flow instability.

●   Flow instability not a problem in primary circuit
    under normal conditions because of high
    pressure and velocities

●   Flow instability (oscillations) occurred (1986)
    on a Bruce steam generator secondary side
    because excessive fouling partially blocked TSP
    flow passages, thus reducing recirculation flow
Design Basis Accidents in CANDU Reactors

●   Single Failures (Failure of process system)
     + Loss of Regulation Accident (LORA)
     + Loss of Coolant Accident (LOCA)
     + Loss of Electric Power Accident (LOEPA)
     + Others
●   Dual Failures (Failure of process system plus failure of
    Special Safety System)
    Special Safety Systems: Shutdown System 1, Shutdown
    System 2, Emergency Coolant Injection System,
    Containment System
        Loss of Coolant Accident (1)
●   Break in PHTS, coolant undergoes critical discharge and
    local pressure falls rapidly.
●   Rapid decrease in local saturation temperature.
●   Flashing proceeds rapidly from break through fuel
●   Reactor power surges because of positive void reactivity,
    terminated by shutdown system
●   Fuel sheath temperature rises, sheath strains and
    oxidizes, exothermic Zircaloy-steam reaction generates
        Loss of Coolant Accident (2)
●   Worst conditions during blowdown occur for “stagnation
    breaks” (Pump and break opposing).
●   PHTS pumps cavitate when local pressure approaches
●   Emergency coolant injection is triggered by falling
    PHTS pressure.
●   Emergency coolant flow quenches fuel and channel and
    prevents fuel sheath failure, fission product release and
    pressure tube failure.
Analysis Methodology for LOCA plus LOECI
Fuel Channel Behavior, LOCA + LOECI (1)

●   Decay power plus heat from Zr + steam reaction
●   Channel filled with stagnant steam
●   Heat transfer from fuel fuel to pressure tube by
    radiation (and natural circulation)
●   Above about 800oC, pressure tube begins to
    plastically deform.
●   For 1 MPa > p< 6 MPa, PT balloons into contact
    with CT, for p <1 MPa, PT sags onto CT, for p >
    6 MPa, PT bursts
Fuel Channel Behavior, LOCA + LOECI (2)
Fuel Channel Behavior, LOCA + LOECI (3)

●   Fuel bundle collapses into bottom of PT. Heat
    transfer from fuel bundle by radiation plus natural
●   Heat transfer from PT to CT by contact
    (ballooned), by local contact plus radiation and
    natural convection (sagged)
●   Heat transfer to moderator by nucleate boiling as
    long as CHF is not exceeded.
Fuel Channel Behavior, LOCA + LOECI (4)

●   Moderator cooling system removes heat.
    [3-d moderator circulation analyzed by
    MODTURC_CLAS, (IST code)]

●   Fuel does not melt, fission product release low.

●   With containment intact, dose to public is within
    CNSC requirements for dual failures.
Fuel Channel Behavior, LOCA + LOECI (5)
        Heat Transfer by Radiation
 Qr = σ.A1. F12.(T14 – T24) kW
 σ is the Stefan-Boltzmann constant, 5.669x10-8 Wm-2K-

 F12 is the radiation interchange factor, allowing for
   geometry (view factors) and emissivity of surfaces
 For example, for radiation between PT and CT:
 F12 = 1/[(1/ε1+ d1/d2.(1/ε2-1)]
 ε1, ε2 are emissivities of PT and CT
 d1, d2 are diameters of PT and CT
Fuel Channel Behavior, LOCA + LOECI (6)
   Hydrogen from Zr + Steam Reaction
       Zr + 2.H2O -> ZrO2 + 2.H2 + heat
●   Heat released raises fuel and PT temperatures
●   Hydrogen escapes from PHTS to containment and
    may detonate or burn depending on local
    concentration, presence of steam, etc.
●   If ECI is eventually restored, hydrogen may
    impede emergency water flow.
●   If ECI is eventually restored, embrittled sheaths
    and PTs may shatter.
    Containment Behavior, LOCA + LOECI (1)

●   Containment compartmentalized, 3-d flow in
●   Atmosphere consists of air, steam, water droplets,
●   Heat added by steam, hot water, fission product
●   Heat removed by air-coolers, dousing spray,
    condensation on containment walls and
    Containment Behavior, LOCA + LOECI (2)

●   Heat may be added by hydrogen detonation
    (>10% H2, depending on steam fraction) or
    combustion (>4% H2).
●   Hydrogen can be removed by AECL-designed
    passive catalytic recombiners.

●   Containment pressure history established by
    balance of heat addition and removal rates
    Containment Behavior, LOCA + LOECI (3)

●   Pressure affected by vacuum building, if present,
    leaks through cracks and venting through filters
●   Containment codes include GOTHIC (IST code),
    for pressure, heat and flow analysis, and SMART
    (IST code) for fission product behavior.
      Severe Accidents in CANDU Reactors

●   Severe accidents are those whose probability of
    occurrence is so low ( < 10-6 per year) that they
    need not be analyzed for licensing purposes.
●   Examples of severe accidents are:
    + Loss of the moderator cooling system in a LOCA +
      LOECI, implying three coincident independent
    + Failure to shut down a reactor in a design-basis
      accident, requiring failure of both independent,
      diverse shutdown systems
Dual Failure + Loss of Moderator Cooling (1)

+ Moderator heats up, boils and is expelled through
  relief ducts after rupture disks break (Rogers et al.,
+ As moderator expelled, uncovered channels sag onto
  submerged channels until they fail by shear.
+ Fuel in failed channels in solid state; quenched by
  remaining moderator.
+ Core disassembly end-state: debris bed of coarse
  pieces of UO2 , ZrO2 and re-solidified Zr at bottom of
  calandria (@ ~ 5 hours).
Fuel Channel Disassembly (Blahnik et al
Core Disassembly Transient
   (Blahnik et al, 1993)
Dual Failure + Loss of Moderator Cooling (2)

+ Porous debris bed heats up to melting point
+ Melting occurs with accompanying geometry changes
+ Molten region superheats
+ Molten region cools as heat source decays and as heat
  is lost to shield-tank water
+ Molten region re-solidifies
+ Solidified regions cool as heat source decays further
    DEBRIS.MLT Model (Rogers & Lamari, 1997)

●   Transient, 1-d, explicit finite difference model of porous
    debris bed and molten pool with crusts
●   Porosity, pore size are inputs
●   Decay heat (plus Zr + steam reaction as option)
●   Allows for geometry changes on melting
●   Models all heat transfer mechanisms in bed and from bed
    to shield-tank water
●   Models all heat transfer mechanisms in pool and crusts
    and from crusts to shield-tank water
●   Option for shield tank cooling system: on or off
DEBRIS.MLT Heat Transfer Mechanisms
     DEBRIS.MLT Temperature Transients
(Effects of Debris Bed Porosity; Pore Size 3 cm )
         DEBRIS.MLT Additional Results

●   Temperature transients also very insensitive to pore size
    over wide range
●   Minimum crust thickness: ~ 8 cm top, ~ 13 cm bottom
●   Maximum heat flux on calandria wall ~ 115 kW/m2
    (Shield tank water CHF > ~ 6400 kW/m2)
●   Maximum calandria wall temperature ~ 380oC (no
    concern about stress rupture)
●   Results very insensitive to molten corium conductivity
    and viscosity
●   Shield tank water boil off > 24 hours
            DEBRIS.MLT Conclusions

●   While significant fraction of core melts, it is
    retained within crusts in the bottom of the
●   Calandria readily survives the accident as long as
    shield-tank water is available
●   Significant time is available for emergency
●   Fission product releases from intact containment
    would be within regulatory limits for dual failure
    Application of DEBRIS.MLT to RASPLAV
                  Experiments (1)
●   International RASPLAV program, Moscow
    (OECD Nuclear Energy Agency)
●   Experiments on interaction of molten corium on
    lower head of PWR pressure vessels
●   Experimental geometry simulates CANDU core
    debris in calandria
●   DEBRIS.MLT modification for RASPLAV
    conditions: DEBRIS.RAS
Application of DEBRIS.MLT to RASPLAV
              Experiments (2)
    Application of DEBRIS.MLT to RASPLAV
                  Experiments (3)
●   Comparisons of predictions to RASPLAV
    measurements provides confidence in
●   Recommendations for improvements in
     IST Severe Accident Code (OH, 1990)
●   Incorporates mass, momentum and energy
    balances and models all chemical and other
    physical processes.
●   Models all key structures and systems.
●   Models all known severe accident phenomena.
●   Physical models integrated to simulate all
    dynamic feedback effects at each time step.
    MAAP-CANDU Applied to Severe Accident in
          Pickering-A Reactor Unit (1)

●   Accident: LOCA in inlet header, SDS1 and SDS2
    both fail (triple failure)
●   Power surges (+ve CVR), fuel melts, PTs and
    CTs fail, calandria fails, moderator discharges
●   Loss of moderator terminates power surge (~ 4 s)
●   Containment dome weld seal fails, opening =
    0.073 m2 (quad failure) (~4 s)
●   Core debris quenched by ECI flow (20-95 s)
    MAAP-CANDU Applied to Severe Accident in
          Pickering-A Reactor Unit (2)

●   Containment pressures drop below atmospheric,
    due to VB (~ 200 s) Vacuum depleted (~ 260 m)
●   EFADS activated to keep pressures below
    atmospheric (I and Cs fission products filtered)
●   Noble gas fission products released
●   Effective doses to public within CNSC limits for
    dual failure even for adverse weather conditions.
    (MACCS code)
●   Inherent passive safety of CANDU design help
    ensure public and operator safety.
●   IST and other codes, validated by thorough,
    quality-assured experimental programs:
    + demonstrate CANDU safety in design-basis
    + show that even in low- probability severe
    accidents in CANDU reactors, public safety is

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