# Nuclear Reactor Safety - PowerPoint by Y92vtGW

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```									 Nuclear Reactor Safety

Technology of Accident Analysis

J.T.Rogers

Winter 2004

UNENE
Topics
●   Reactor Physics Review
●   Fuel Behavior
●   Thermalhydraulics
●   Design-Basis Accidents
●   Severe Accidents in CANDU Reactors
Neutron Cycle
Diagram
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
where:
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
●   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
Codes
●   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
others
CANDU Reactor Physics IST Codes
●     Transport theory cell codes for fuel bundle,
pressure tube, calandria tube, control mechanisms

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

●   After reactor trip, power still produced by
●   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, J.kg-1
Location of Fission Products

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

•   Cracking
•   Swelling
•   Dishing/ridging
•   Gas pressure increase

Fuel        Fuel (exaggerated)
Fuel Behavior in Accidents

Key Safety Parameters:
●   Fuel temperature
* Potential sheath failure
* Potential pressure tube failure (Creep rupture,
DHC)
* 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
increase.
IST Codes for Fuel Behavior
ELESTRES
Models micro-structural, mechanical and thermal
behavior of CANDU fuel element under normal
operating conditions. Used to quantify pre-
accident conditions.

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

●   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
Patterns
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
(1)
●   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
(2)
●   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
water
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
multiplier.
Empirical correlations for Φfo2 include those of Lockhart
& Martinelli (1949), Friedel (1979) and others. Friedel is
recommended.
Finite Difference Modelling of System

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

●   Mass and energy conservation equations for
nodes

●   Momentum conservation equations for links
• Break the circuit up into
– nodes containing mass
• Mass & energy conservation equations for nodes
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
Reactors
CATHENA
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
analyses
ASSERT (IST code)
Models steady-state or transient single- or two-phase flows in
sub-channels of fuel bundles.
NUCIRC
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)
Departure
from
Nucleate
Boiling
(DNB)
Dryout in Flow Boiling (1)
Dry Out in Flow Boiling (2)
●   In annular flow, liquid film keeps surface well-
cooled
●   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
method)
●   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
experimentally
Heat flux
– no reliable theory for     (overpower)
CHF
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
margins?
Effects of Critical Heat Flux with Water
Coolant

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

●   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
Transfer
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
Discharge
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
assumptions
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
rates
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
channel.
●   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
H2
Loss of Coolant Accident (2)
●   Worst conditions during blowdown occur for “stagnation
breaks” (Pump and break opposing).
●   PHTS pumps cavitate when local pressure approaches
saturation
●   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
●   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
convection
●   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)
Qr = σ.A1. F12.(T14 – T24) kW
σ is the Stefan-Boltzmann constant, 5.669x10-8 Wm-2K-
1

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
compartments
●   Atmosphere consists of air, steam, water droplets,
hydrogen.
●   Heat added by steam, hot water, fission product
decay.
●   Heat removed by air-coolers, dousing spray,
condensation on containment walls and
equipment
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
failures
+ 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)

+ CANDU-6, LOSWA +LOEPA
+ Moderator heats up, boils and is expelled through
relief ducts after rupture disks break (Rogers et al.,
1984).
+ 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
1993)
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 )

●   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
●   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
●   Calandria readily survives the accident as long as
shield-tank water is available
●   Significant time is available for emergency
procedures
●   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)
Conclusions
●   Comparisons of predictions to RASPLAV
measurements provides confidence in
DEBRIS.MLT
●   Recommendations for improvements in
DEBRIS.MLT
MAAP-CANDU
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)
Summary
●   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
accidents.
+ show that even in low- probability severe
accidents in CANDU reactors, public safety is
maintained

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