final exam Flashcards
1
Q
Describe what a power plant is and what they do
A
a power station where fuel is exploited to produce energy/electricity economically
they use fuel in unique systems and processes to convert heat energy produced into mech energy, which then operates an elec generator
2
Q
what define the type of power station/plant
A
type of fuel
3
Q
purpose of nuclear plant is to
A
allow the safe, efficient, reliable, cost effective and economic conversion of the fission energy of fuel to electricity, with minimum impact on the environment
4
Q
3 main methods of generating conventional electricity
A
thermal power plants
hydroelectric power plants
nuclear power plants
5
Q
differentiate between the 3 types of power plants
A
nuclear - fission in core produces heat to generate steam
thermal - burning coal/natural gas/oil/fossil fuels in combustion chamber to boil water in generating steam
hydro- hydro-electric dams generate potential energy- hydraulic head power to run turbines
6
Q
how do you make electricity
A
in each plant - turbine turns a giant magnet inside the generator
7
Q
write Nth eq (carnot efficiency)
A
Slide 10
8
Q
the steam temp is ___ in current nuclear reactors than in fossil fuel power plants
A
lower
9
Q
most efficient fossil-fuel plants have thermal efficiencies of __
A
40%
10
Q
thermal efficiencies of reactors are ___ than fossil-fuel power plants
A
lower
11
Q
thermal efficiency of PWR and BWRS
A
32%
12
Q
thermal efficiency of HTGR and fast breeders
A
40%
13
Q
thermal efficiency of CANDU
A
30%
14
Q
All equations on slide 11
A
15
Q
candu uses what as a moderator. what is it and what does it do
A
heavy water
- slows down neutrons
- is deuterium water = D2O
- Deuterium has one more neutron then hydrogen (in conventional H2O)
16
Q
difference in light vs heavy water
A
heavy water has more mass (since 2 neutrons)
17
Q
of CANDU what is the dimensions of the core
diamater
length
walls (and material)
ends
A
7.6m
7.6m
2.5cm (1inch) (stainless steel)
5cm (2inch) thick
18
Q
candu
explain what the pressure tubes do and where they lead to
A
primary coolant flows through hundreds of pressure tubes, each have a feeder at either end leading to headers and steam generators
19
Q
CANDU
why is there no need for a large pressure vessel
A
moderator surrounds the pressure tubes at atm pressure
20
Q
CANDU
how is fuelilng done
A
on-line
21
Q
Length of Candu fuel bundle
A
50 cm
22
Q
how is fuel fabricated
A
UO2 pressed into small pellets and placed in zirconium
tubes to form fuel rods
fuel rods are arranged in bundles and inserted into a nuclear reactor
23
Q
Nuclear Fission is useful because we can control it by:
A
adding neutron scavengers
keeping below critical mass
moderating the chain reaction
‘delayed neutrons’
24
Q
nuclear reactor will not operate without _
A
neutrons
25
purpose of neutrons in reactors
induce the fission reaction, which produces heat in nuclear power reactors
26
why is fission important
creates more neutrons that are used to sustain the chain reaction
27
why does fission occur
high potential energy to low potential energy state
28
typical candu uses __ between calandria and pressure tube
CO2
29
productive vs non-productive absoptions
end in fission
do not end in fission (leakage)
30
define neutron economy
very delicate balance between fission reaction, neutron capture and neutron leakage
31
flux and cross section eq slide 26
32
what is neutron flux
of balls per sec per area
33
neutron flux
cross sectional area
catchers glove
34
flux and cross section
n catchers,, then rate they will catch the ball is proportional to _
of catchers and number of balls that hit the glove
35
flux and cross section. why is there a negative sign
catchers is going down (disappearing)
36
how do reactions begin and what is involved in this process
starts with controlled reaction called fission
where the splitting or uranium nucleus generates heat
37
types of fission
spontaneous
induced
38
spontaneous fission
- what type of nucleus, what happens, what are products
isolated nucleus undergoes fission - splitting into 2 smaller nuclei - typically accompanied by emission of one to a few neutrons
39
spontaneous fission
describe fission fragment
typically unequal in mass and highly radio active
40
spontaneous fission
type of energy released
in form of kinetic energy of the products and as excitation energy of the (radio active) fission fragments
41
induced fission
describe the process and fission fragments
capture of neutron causes a nucleus to become unstable and undergo fission
fragment similar to spontaneous (typically unequal in mass and highly radio active)
42
what is nuclear fission
Splitting of the uranium nucleus to generate heat
43
what was the first nuclear fission reaction identified.
fission of U-235
44
who was involved in the first nuclear fission reaction identified and when
Otto Hahn, Lise Meitner, Fritz Strassmann
1938.
45
where does energy come from in fission
binding energy aka mass defect
46
what does mass defect refer to
difference in mass between an atom and sum of the masses of protons neutrons and electrons of atom
47
mass in mass defect is typically associated with ?
binding energy between nucleons
48
Where does the energy come from in fission?
missing mass is ?
the energy released by the formation of the atomic nucleus
49
binding energy eq (slide 31)
50
nuclear fission occurs when
U-235 captures an additional neutron and it fissions into 2 main components
51
U-235 fissions in to what - write eq
U235 + n -->Cs143 + Sr90 + 3n
52
see slide 32 in binding energy calc
53
slide 33 calculating mass defect
54
where does fission energy go
energy released by fission goes into kinetic energy of the fission products (168 MeV) and neutrons (168 MeV) and gammas (5 MeV) and betas (19 MeV), which ultimately becomes heat- remaining goes to neutrino energy - which doesnt become heat
55
electron volts are useful for and represent
atomic and nuclear processes
the kinetic energy gained by an electron passing through
an electric potential difference of 1 Volt (in vacuum).
56
we get useful energy from
burning hydrogen and oxygen because the product (H2O) is ar a lower gibbs free energy
57
why is combustion of fossil fuel complicated then neuclear
large numbers of different radicals - general chain reaction is the same
58
slide 37?
59
how to distinguish types of reactors
fuel
neutron moderation
cooling
control and safety systems
balance of plant systems
60
3 main types of reactors
PWR, BWR, CANDU
61
PWR
coolant/moderator is pressurized to __ Mpa
Tinlet__C
Toutlet__C
1.8
289
325
62
BWR
coolant/moderator is pressurized to __ Mpa
Tinlet__C
Toutlet__C
Exit steam quality (wt% steam)__%
7.5
216
288
14.7
63
CANDU
coolant is pressurized to __ Mpa
Tinlet__C
Toutlet__C
moderator is at __C and pressurized to ___
11
270
205
70
atmospheric pressure
64
what is unique about safety of nuclear power plants
list 4
tremendous amount of energy produced
a lot of heat produced even after the reaction is terminated
core contains large inventory of radioactive material that is hazardous to humans and the environment
things can happen very fast and there are considerations over long periods of time
65
characteristics of a reactor core
control
cool
66
control
greater than ___ watts at full power
equivalent to ___ 747's at full power
2 billion
30
67
cool - in relation to 747 engines
* 5 min. after shutdown - ___
* 5 hours after shutdown - ___
* 5 days after shutdown - ___
* 5 min. after shutdown - 1 747 (all engines)
* 5 hours after shutdown - 1 747 engine
* 5 days after shutdown - 1 747 engine at
half throttle
68
safety 3 types
common sense
rigorous application of rules
blind obedience
69
safety approach in canada
name all 5
government regulations (CNSC)
REGDOC 2.5.2 deign:-safety of plant, environment-workers
REGDOC 2.41,2.3.2 - Management of Risk: -deterministic, probablilistic assessment, safe operating envelope
operating procedures,trainings,philosophy, culture
license/license condition handbook
70
golden riles of reactor safety and what they mean
control
- design should enable nuclear fission to proceed in a controlled manner
cool
- design should be capable to remove the heat produced and convert the heat produced to useful work (power)
contain
- design should have the means of ensuring that its function can continue over the expected plant life in a safe and reliable manner
71
control purpose
keep fission products within the fuel
control reactor power - reactivity additions and shutdown reliably
72
cool purpose
keep fission products within the fuel
fool the reactor and spent fuel
maintain coolant inventory, coolant flow, coolant heat sinks
73
contain purpose
keep radioactive material within reactor
maintain containment integrity - prevent: over pressurization, overheating, containment bypass
capture material within the containment:-scrubbing deposition, chemical capture
keep out of biosphere
74
safety design what are the 3 things we are trying to do and why (goals)
contain fission products and other radioactive species under all operating conditions, normal, abnormal
-goal of public protection activities
protect operating staff from harm
- goal of radiological and conventional safety activities
prevent damage to plant equipment
- goal of financial loss control activities
75
elements of design process
4 different elements
nuclear design
thermal and hydraulic analysis
reactor control and kinetic analysis
mech design
76
elements of design process
nuclear design
basic design of the reactor core and required sheilding
77
elements of design process
thermal and hydraulic analysis
thermal analysis of reactor core and fuel, design of the primary coolant system
78
elements of design process
reactor control and kinetic analysis
reactor control system
79
elements of design process
mech design
design of fuel elements in conjunction with nuclear and thermal analysis
design of primary containment system
80
main components of nuclear reactor
control rods, core, water from condenser, steam to turbine, reflector
81
features of nuclear reactor (try and name a few)
neutron source
reactor core
fuel elements
coolant
moderator (depends on design)
reflector
pressurizer (depends on design)
primary containment
containment
heat exchanger (depends on design)
safety systems
82
other systems in nuclear reactors
safety systems
balance of plant systems: - auxiliary, nuclear steam supply, power generating system (turbines and generator)
83
REACTOR CORE
relation to nuclear system
heart of any nuclear system
84
reactor core
fuel requirements and geometry requirements
fuel must be ideal geometry to allow fission
core geometry must allow heat generated to be readily and economically removed by the reactor coolant system
85
reactor core
provides what for other components
shielding of other components from core radiation
86
fuel elements
what type of fuel has been used in all major design concepts
solid fuel elements
87
fuel elements
fuel assembly must ____
and should retain___
must be able to contain the fuel (most of fission products formed) in a configuration that can be properly cooled and safely handled
should retain its integrity under the expected operating/ design conditions
88
fuel elements must be compatible with __
should be capable of _____operation
should provide ____parasitic neutron absoption
core design
high temp
minimum
89
fuel elements
what is the basic fuel and how is it arranged
uranium
pellets of uranium oxide (UO2) are arranged in tubes to from fuel rods
90
where are fuel elements contained
in reactor core
91
what is the only natural occurring fissile nuclide and what happens to it
U-235
during operation - other fissile nuclides (Pu-239, U-233) produced from fertile material
92
2 types of fuel elements and the difference
fissile
- nuclide that is capable of being split by an interaction with a thermal neutron (U-233. U-235, Pu-239)
Fertile
- nuclide that may capture a neutron to form a product that eventually decays to become a fissile nucleus (TH-232 , U-238)
93
initiating neutron source
In a new reactor with new fuel, a neutron source is needed to get the chain reaction going - this is usually___
Be mixed with Po an Ra or other alpha emitter
94
Initiating Neutron Source
what do the alpha particles from deacy do
cause a release of neutrons from the Be as it turn to Carbon-12
95
primary coolant system
coolant must be capable of sustaining ___
and be compatible with___
high temp
core design
96
primary coolant system
provides what?
sufficient coolant circulation to remove the heat generated within the core and transport the energy to a prime mover or secondary system which also transports it to a prime mover
97
typical coolants
H2O, D2O, liquid Na (or Na-K alloy), liquid organic compounds, air, CO2, He, boiling H2O
98
in some design the coolant may also provide
for necessary neutron moderation
99
In light water reactors the water moderator functions also as the___
primary coolant.
100
how many coolants in PWRs
there is secondary coolant circuit where water becomes steam
it can have 2 to 4 primary coolant loops with pumps driven by steam or electricity
101
how many coolants in BWRs
has no secondary coolant circuit
water boils to steam directly
102
reflector
what is it?
purpose?
surrounds the core
to reduce the loss of neutrons from the core
103
reflector
material?
material determined by energy distribution of neutrons in the core
Be and graphite make good reflectors as they can also act as a moderator. steel and lead also work but have less effect on neutron energy
depleted uranium oxide can be used as U238 can absorb neutrons and become new fissile material
104
reflector
benefits?
lower critical mass (mass of fuel needed to sustain chain reaction)
act as thermal and radiation shield
105
what is a moderator
material in core that slows down neutrons released from fission process so that they cause mroe fission
106
best moderators are:
elements of low mass numbers with small neutron capture cross sections
H20 , D20, Be, Be oxide, graphite
107
do fast reactors need a moderator?
no
108
Why not mix the fuel and moderator together like a nuclear smoothie?
238U captures a neutron and forms 239U, which eventually decays to 239Pu after about 3 days.
239Pu is fissile, and contributes to CANDU reactor energy output, but we need 235U to undergo fission to keep the reactor operating, especially during start up or after refueling, where 239Pu is not yet present.
239Pu fission accounts for about half of the heat produced by a CANDU reactor during critical operation!
109
primary containment
for PWR, BWR and CANDU
pressure vessel - PWR,BWR
pressure tubes -CANDU
110
primary containment
requirements
reactor core (including fuel and primary coolant) must be contained in leak tight system
The containment must also serve as a reliable barrier to the release of radioactivity:
✓From failed fuel
✓From coolant activation to the environment
Must be designed to withstand shock loading which could result from pipe severance or earthquake/seismic activity.
111
primary containment examples
ie steel vessel, tubes etc
✓robust steel vessel containing the reactor core and moderator/coolant in some concepts
✓In others, e.g. CANDU, it may be a series of tubes holding the fuel and conveying the coolant through the surrounding moderator.
112
Secondary External Containment Structure
what is it and what does it do
structure around the reactor and associated steam generators,
designed to protect from outside intrusion, and to protect those outside from the effects of radiation in case of any serious malfunction inside.
113
Steam Generator (Boiler)
what is the role of the coolant
relate to car
* High-pressure primary coolant brings heat from the reactor to a secondary circuit to make steam for the turbine.
* From high-pressure primary circuits in some concepts, to a secondary circuit where water turns to steam.
* Essentially a heat exchanger like a car’s radiator. Reactors have multiple “loops', each with a steam generator
114
Steam Generator (Boilers)/Heat Exchangers
__ water must flow through the support structures for the tubes.
___ build up impede the flow of steam and must be avoided.
* Chemically maintained to avoid __
* Tubes which fail and leak are__
* Detection of leaks by monitoring ___ levels in the steam as it exits the steam generator.
secondary
* Deposits
corrosion.
plugged
N-16
115
reactivity control system
reactor must be capable of____
and control system must repsond to ___
of safely bringing the reactor to power, maintaining it there and shutting the reactor down
unexpected load variations and rapid reactor shutdown (scram) during an emergency
116
how do control rods work
✓ Neutron-absorbing material (poison) inserted or withdrawn from the core to control fission
✓ Control is achieved by varying neutron density.
✓ Poison materials: Boron (B), Gadolinium (Gd), Cadmium
(Cd), Dysprosium (Dy).
117
what does neutron poisoning do
and what does it do when injected or removed
Allows control of reactor
Insertion of poison results in a decrease in reactivity (or neutron multiplication) of the core, i.e., decreases neutron density. Hence, reactor power level is reduced.
✓Withdrawing the poison increases the neutron density and power level.
✓In some design concepts, special control rods are used to enable the core to sustain a low level of power efficiently. (Secondary control systems involve other neutron absorbers, usually boron in the coolant).
118
pressurizer
what is it used for
what is it connected to
Used for Heat Transport System inventory (liquid state)
control and pressure control
The pressurizer is connected to the primary loop through a surge nozzle at the bottom.
✓ Heaters are provided at the bottom of the pressurizer internals, and
✓ a spray nozzle, relief nozzle, and safety nozzle are installed at the top of the pressurizer head.
119
pressurizer
positive vs negative surge
* A “positive surge” of water from the primary loop due to
increasing loop pressure is compensated for by injecting
cold water from the top of the pressurizer to condense
steam
* A “negative surge” of water empties the pressurizer,
reducing steam pressure at the top of the pressurizer and
thus loop pressure.
In this situation, the electrical heaters at the bottom of
the pressurizer are automatically activated, converting a
portion of the water into steam, resulting in a loop
pressure increase
120
auxiliary system is responsible for
Removal of radioactive material and other contaminants
from the primary coolant.
The refuelling system
Removal of radioactive waste from discharged air and
water streams
121
Feed water systems does __
The feedwater system supplies demineralized and
preheated light water to the steam generators
122
main system responsible for
The steam from the boilers is fed by separate steam
mains to the turbine steam chest via the turbine stop
valves, and its flow is controlled by the governor
valves;
✓Excess steam can be discharged to the atmosphere or
bypass the turbine by flowing directly to the
condenser;
✓Over-pressure protection is as safety relief valves on
each steam source
123
Power Generating Systems responsible for
✓The energy transferred from the primary coolant
must be transferred to a prime mover, which is
always the turbine.
124
Safety Systems responsible for
✓Demonstration that no accident situation can
significantly endanger the health and safety of
workers and the public.
✓Systems to provide heat sink to cool under
accident conditions.
✓Enclosures to retain any radioactive releases must
also be provided.
125
to operate the reactor
we want keff=? and why
1 to keep power steady
126
to reduce power in reactor/ shut down
what does keff=? and what is this process called and what does it do?
k less than 1
done by inserting neutron absorbers (water, cadium,boron,gadolinium)
process called negative reactivity insertion
127
to increase power in reactor
keff=?
how is this done and what is it called
keff greater than 1 (slightly for a short time)
achieved by removing bit of absorber
called positive reactivity insertion
128
why do we not want to make keff much greater than 1 for long times
power could increase to high values, potentially with
undesirable consequences, e.g. melting of the fuel.
129
do we need reactivity devices to counteract perturbations to chain reactions when we want to keep keff=1
yes
130
what does the movement of reactivity devices (i.e., control rods) do
allows neutron absorbers to be added or removed in order to manipulate keff
131
every nuclear reactor contains regulating and shutdown systems to ____
to increase, decrease, or keep keff steady, as desired.
132
Any imbalance between neutron production and loss causes the neutron population to ____
to increase or decrease, from one generation to the next.
133
The rate at which the neutron population, and consequently the power, changes depends on the ____
what kind of neutrons is this
‘mean-generation’ time T, which is the average time between successive fission generations.
✓Prompt and delayed neutrons
134
The minimum quantity of fissile material that is capable of sustaining a fission chain:
✓depends upon what? list 4
its nuclear material and properties
▪ nuclear fission cross-section
▪ its density, its shape,
▪ its enrichment, its purity,
▪ its temperature, and its surroundings
135
Origin of issue
✓__ or __neutrons are liberated per fission, but __ is required to maintain the fission chain.
do all neutrons resulting from fission carry on the fission chain?
2 or 3
1
no - there are losses
136
what are neutron losses
non-fission reactions with other nuclei, but also with fissile materials; and escape from the system through its physical boundaries (leakage)
137
what is leakage and how is it controlled (in terms of critical mass)
can be controlled/reduced by increasing the size, i.e., mass of the fissile material.
137
at critical mass, what happens to chain reaction
it becomes self sustaining
138
Critical mass depends on many factors, examples are
✓physical form of fissile material and moderator type
139
Examples of critical mass of 235U:
✓<1 kg for a homogeneous solution in water of uranium salt containing about 90% of fissile isotope
✓200 kg present in 30,000 kg of natural uranium embedded in a matrix of graphite
140
what happens to leakage of neutrons out of reactor when reactor size decreases
what is required of the reactor in this case
it increases
reactor must have a minimum size to work.
141
what happens to keff when the reactor is below the minimum size (critical mass)
leakage is too high and keff cannot possibly be equal to 1.
✓Less sustaining nuclear fission
142
critical mass depends on (3 things)
✓shape of the reactor
✓composition of the fuel
✓other materials in the reactor
143
Shape with lowest relative leakage, i.e. for which critical mass is least, is the shape with the smallest surface-to-volume ratio is a _
: a sphere
144
In CANDU reactors, fission neutrons travel about __cm before being absorbed by the fuel.
The ___ the reactor, the ___ the chance that the neutron will leak out before it is absorbed
50
bigger
lower
145
are spherical reactors practical? why? what do we use?
Spherical reactors are not practical, so we use a cylindrical core with a diameter slightly bigger than the length.
146
The larger the____ is relative to the ____, the larger the leakage.
surface area
volume
147
Small reactors have ____ leakage and usually require ___ enrichments
larger
higher
148
reactor size
PWR (__ MWTh)
vessel height and diameter
core size
3400 MWTh)
✓ The vessel is ~12m high with a 4 m diameter,
✓ Core size ( 3.7 m high x 3.4 diameter)
149
reactor size
BWR (__ MWTh)
vessel height and diameter
core size
(3300 MWTh)
✓ The vessel is ~22m high with a 6 m diameter, with
✓ Core size ( 3.7 m high x 4.8 diameter)
150
reactor size
CANDU (__ MWTh)
core size
* CANDU-6 (2060 MWTh)
✓ The core is relatively “small”: ~ 7.6 m diameter, 6m
length
✓ But notice the diameter
151
why are BWRS length so long?
the steam separators and dryers, which are necessary to separate steam from water in the primary circuit, are located above the reactor core within the vessel
152
what happens to the heat (energy) from nuclear decay
after the originating fission reaction
how long does this last
Many fission products are still decaying long after
Energy (heat) from this nuclear decay is produced in the reactor for many hours, days, even months after the chain reaction is stopped.
153
is this decay heat is not negligible? why or why not?
what percent of total heat generated is decay heat in steady operation
no
7% of the total heat generated.
✓shutting down the fission process brings core heat down to about 7 per cent of its
running temperature.
✓Even after reactor shutdown, decay heat must be dissipated safely, otherwise the
fuel and reactor core can seriously overheat.
154
explain the percentage split of fission heat
* 5% of fission heat goes to moderator
✓ 20% delayed
✓ 80% prompt
* 95% of fission heat goes to fuel
✓ 6% delayed
✓ 93-94% prompt
* 1% additional from pump heat
155
where does this type of energy source take place and how is it dissipated
Fission fragments
Fuel pellets
(fragments slowed in a few tenths of a mm)
156
where does this type of energy source take place and how is it dissipated
Kinetic energy of fission neutrons
Mostly in moderator,
(collisions with moderator molecules/atoms)
157
where does this type of energy source take place and how is it dissipated
Beta particles from fission products decay
Fuel pellets and cladding
158
where does this type of energy source take place and how is it dissipated
Prompt gammas, gammas from (n,γ) reactions and
from fission product decay
Throughout reactor and shielding
(About 1/3 in moderator and shielding
159
CANDU
Turbine, Steam Generator, Condensate and Feed
heating Systems
it is the main system involved in converting __?
✓heat energy of the steam in the turbine (Primary mover)
✓ Flow of steam to turbine
✓Rotational energy (Rotor and stator)
▪ which turn and drives the generator
✓Converts the mechanical energy to electrical energy
159
Decay heat must be considered in two scenarios during design
explain each
1.When used fuel is removed from the reactor
* Remedy is the fuel must be safely stored, to cool it and to contain its radioactivity.
* Design of spent fuel bays as part of the Balance of Plant
2. During a transient when the reactor shuts down
Remedy is to obtain adequate core cooling for safe shutdown to prevent
✓ melt down of reactor and
✓ other safety issues
▪ release of radioactivity into the environment, hydrogen/steam explosions
▪ molten-concrete core interactions
* Design of emergency core cooling, emergency mitigation equipment and other associated safety and auxiliary systems for safe shutdown
160
CANDU
All CANDU generating stations have turbine assemblies
consisting of
One double flow high pressure cylinder
* three double flow low pressure cylinders
with external moisture separators
* live steam reheaters between the high- and
low-pressure stages
* Called a tandem compound
* Note the size
✓ HP and LP Turbines
161
CANDU
what does the governing system do
The governing system controls the turbine's speed of rotation when the generator is not synchronized to the grid,
* Governor valve on a steam engine * the governing system determines turbine/generator power
162
CANDU
design features of governing system
Reliable and fast acting
governor valves
* Have emergency trip systems
which can detect and prevent
turbine overspeed and safely
unload the turbine.
163
CANDU
Separator and Reheater
✓Steam exiting the high pressure turbine has about __% moisture content, which must be removed prior to
admitting the steam to the low pressure stages.
10
164
CANDU
what does the separator do?
uses mechanical means to remove much of the moisture content,
165
CANDU
what does the reheater do?
live steam raises the steam to superheated conditions.
166
CANDU
how many phases does the generator have
what is the normal output voltage
what cools the rotor
✓It is a three-phase four-pole machine directly coupled to the turbine.
✓The output voltage is typically 24,000 volts, and is connected via forced air cooled,
▪ isolated phase bus duct to the step-up Main Output
Transformer.
✓Cooling of the rotor winding and stator core is by hydrogen
167
CANDU
how do you extract maximum energy from the steam
where does this take place
it needs to be condensed to a pressure and temperature that is as low as practicable.
condenser
167
generator eq?? slide 107
168
CANDU
condenser consists of how many shells
what happens when turbine not available
* Consists of three separate shells
✓ one for each low-pressure turbine cylinder;
* can handle full steam bypass flow when the
turbine is not available.
169
CANDU
Feed Heating
uses what to preheat?
consists of what?
where does it return feedwater to?
* uses extraction steam to preheat the feedwater in order to optimize thermodynamic efficiency;
* consists of three low pressure (LP), a deaerator, and two HP heaters;
* feed pumps return the feedwater to the boilers.
170
CANDU
Groups of Power for Entire Operation
✓Normal (Group 1) and Emergency (Group 2)
171
CANDU
The Group 2 power supplies is ___ in Group 1 (Class __)
what is it for?
more reliable than the highest (class 1)
✓Seismic qualified – Can survive a seismic event
✓Emergency Power Supply
171
CANDU
* The Group 1 power supplies are classified in terms
of their ____
list classes
✓Class I Power
✓Class II Power
✓Class III Power
✓Class IV Power
level of reliability (Non seismic qualified)
✓Class I Power
✓Class II Power
✓Class III Power
✓Class IV Power
172
CANDU
group 1 power
The ___ the number the more reliable is the power
lower
173
CANDU
class1 power
what kind of power
give ex
how strong
dc or ac?
Uninterruptible direct current (DC) supplies for essential auxiliaries are obtained from the Class 1 power supply,
✓ Three independent DC instrument buses,
✓ These buses are each supplied from a Class III bus via a rectifier in parallel with a battery
✓ Buses provide power for DC motors, switchgear operation and for the Class II AC buses via inverters.
✓ typically 50V to 250 V DC.
174
CANDU
class 2 power
what kind of power
give ex
how strong
dc or ac?
Uninterruptible AC supplies for essential auxiliaries are obtained from Class II power supply, which comprises:
✓ Two low voltage AC three phase buses which supply critical motor loads and emergency lighting. These
buses are each supplied through an inverter from a Class III bus via a rectifier in parallel with a battery.
✓ If a disruption or loss of Class III power occurs, the battery in the applicable circuit will provide the
necessary power without interruption.
✓ supplies equipment and instrumentation essential to safe station operation;
✓ typically 50 V to 250 V AC.
175
CANDU
class 3 power
what kind of power
give ex
how strong
dc or ac?
Supplied by on-site Standby Generators if cannot be supplied from Class IV;
✓ AC supplies to auxiliaries that are necessary for the safe shutdown of the reactor and turbine are
obtained from the Class III power supply with a standby diesel generator
✓ These auxiliaries can tolerate short interruptions in their power supplies (may be unavailable for 3
minutes);
✓ Essentially to ensure reactor cooling is maintained
✓ typically 400 V to 5 kV AC
176
CANDU
class 4 power
what kind of power
give ex
how strong
dc or ac?
* Supplies all major loads directly and all station equipment under normal operating conditions via the other
classes
✓ may be unavailable for extended periods (hrs);
✓ Essentially ensure heat sinks are operating
✓ Complete loss of Class IV power will initiate a reactor shutdown.
✓ highest distribution voltage within station(4 kV to 14 kV AC)
177
CANDU
what does EPOS stand for
what does it consist of
what is it the link between and what does it do with the electrical power produced
* The Electrical Power Output Systems (EPOS) consists of :
✓Main Output Transformer
✓The Switchyard
✓The Station Service and Unit Service transformers
* The EPOS - link between electrical power generated by the station and the Power Grid of the Utility that utilizes the energy.
✓It also produces the electrical power that operates the station
178
CANDU
Main Transformer
what does it do
what is it rated to meet and why
The main transformer steps up the generator output voltage to the same level as the switchyard transmission voltage.
✓ The transformer is rated to meet the generator output requirements and site
environment. It is equipped with all standard accessories and the necessary protective equipment
179
CANDU
Switchyard
where is it located
what does it do/ what does it contain
HOW MANY INCOMING LINES
✓ The switchyard, located near the turbine hall,
contains the automatic switching mechanism, including the breakers and disconnects, which is the interface between the station and the power grid transmission lines.
✓ There are at least two incoming lines which are synchronized under normal
conditions.
▪ However, the switchyard electrical equipment allows
transmission of full station power through any one of the
incoming lines
180
CANDU
Unit Service Transformer
✓ During normal station operation the station services power is supplied by __?
why is this
where is the transformer fed from
does its operation need to be connected to the grid?
both the unit service transformer and the system service transformer.
either transformer can provide the total service load
in the event of a failure of one supply. The transformer is fed from the output system of the turbine generator.
✓ Its operation does not require that the unit be connected to the grid
181
CANDU
System Service Transformer
what is it similar to
where is the transformer fed from? via what?
what does it supply under normal conditions?
what does it supply during start-up
✓ The system service transformer is similar to the unit service transformer.
Fed from the electrical grid via the switchyard
✓ It supplies half of the plant services power requirements under normal operating conditions and is able to provide the total service load when necessary.
This transformer is fed from the switchyard and supplies all plant loads during the start-up of the plant, or when the turbine generator is unavailable
182
CANDU
Computerized Reactor Control
✓Digital computers are used for:
how many computers are used to control the unit? why? what kind?
▪ station control,
▪ alarm annunciation,
▪ graphical data display and logging;
✓Two independent computers, both normally running,
but each capable of controlling the unit;
▪ only the 'controlling' computer's outputs are
connected to the field devices;
▪ a fault in any essential part of one computer
results in automatic transfer of control to the
other computer
183
Because of the complex interdependence of the
control systems in a CANDU unit, Digital Control
Computers (DCC) perform all 5 major control
functions
they are :
unit power regulator (UPR)
reactor regulating system (RRS)
heat transport pressure and inventory control (HTP&I)
steam generator pressure control (SGPC)
steam generator level control (BLC)
184
CANDU
special Safety Systems
so they take part in everyday activity? yes or no? why?
✓ These systems do not take any part in normal power plant operations, but are “poised” to act
✓ In other words, they are waiting and watching in case the processes and their control systems cannot keep key operating parameters within prescribed limits
✓ In such cases, when there is the potential for fuel failure to occur with a risk of radioactivity release, these special safety systems spring into action.
184
Each CANDU has four special safety systems.
what are they and why do they exist
* The two independent shutdown systems
* Shut down system1 (SDS#1) and SDS#2
* The emergency coolant injection system (ECIS) and
* the containment system
The plant design includes these four special safety systems to protect the public from a harmful radiation release in the event of failure, transient or accident
185
4 Special Safety Systems in a CANDU
✓ If the control of reactor power is not assured what happends?
what differentiates sds1 and sds2
▪ SDS1 and SDS2 based on different physics
a. SDS 1- Shut off Rods to absorb neutrons
b. SDS 2 - liquid Poison (Gadolinium nitrate solution) to absorb neutrons
186
CANDU
If cooling of the fuel is judged to be insufficient,
what is implemented?
using what?
Emergency Core Cooling will be implemented;
using light water to cool the core
187
CANDU
If there is a risk, or perhaps an actual release of
radioactivity from any of the plant systems, then what system ensures that no unsafe level of radiation is released to areas outside the plant’s boundary.
what is this called
Containment System
▪ Called boxed-up
▪ Containment is a leak tight structure
188
CANDUReactor Shutdown Systems
describe each and how they work together and separately; SDS1 and SDS2
✓ functionally and physically independent of each other, and each able to shut down the reactor;
]
* functional independence is provided by using different methods of shutdown
✓ dropping solid neutron absorbing rods into the core for SDS#1,
✓ injecting liquid poison into the moderator for SDS#2;
* physical independence is achieved by positioning the shutdown rods vertically through the top of the reactor, and the poison injection tubes horizontally through the sides of the reactor;
✓ The two shutdown systems (SDS1 and SDS2) respond
automatically to both neutronic and process signals
189
If cooling of the fuel is judged to be insufficient - What is implemented
Emergency Core Cooling
190
CANDU
* ECC is initiated when the Heat Transport System pressure has dropped below -- Mpa
what else happens
5.5 Mpa
✓Indication that Loss of coolant from the HTS has taken place
✓High reactor building pressure in case of break leaking coolant into the reactor building
✓High moderator level in case of a fuel channel break leaking into the moderator
191
CANDU
Containment System
Provides what
a sealed envelope around the nuclear steam
supply systems if an accidental release of radioactivity is detected:
✓ plastic lined pre-stressed post tensioned pressure-retaining concrete containment structure;
✓ Structure and supporting systems which provide the
final barrier to limit radioactive releases to the environment to acceptable levels
192
CANDU
Containment System
* Designed to withstand ___
the maximum pressure which could occur following
the largest possible LOCA
193
CANDU
* Containment is subdivided into:
✓ Containment envelope R/B including extensions and
penetrations
✓ Containment penetrations and isolation
✓ Atmospheric Control
194
CANDU
containment system
* Atmosphere Control
✓ automatic dousing system-
▪ Dousing system for pressure suppression
✓ R/B Air Coolers for heat removal
▪ provide a long-term containment atmosphere heat sink
✓ Filtered Air Discharge (in the long term if required)
✓ access airlocks;
✓ automatic containment isolation system that closes all reactor building penetrations open to the containment atmosphere when an increase in containment pressure or radioactivity level is detected
195
CANDU
Common Features Relevant to Containment
and Accidents
name 3
Heavy water for coolant and moderator
✓10% mass difference between heavy and lightwater affects thermodynamic properties only slightly
✓Tritium generated when heavy water absorbs a
neutron, requires periodic removal
* Dual reactor shutdown systems; accident analysis
assumes failure incredible.
✓Positive void coefficient causes brief power
increase during LOCA before reactor shutdown
* Fuel bundles in several hundred horizontal pressure
tubes rather than pressure vessel, pressure tubes
enclosed in cold moderator tank (calandria)
196
CANDU
The Fuel Handling and Storage Facilities to support
this operation include
1. receiving, storing, inspecting and loading new fuel into fuelling machines;
2. on-line removal of spent fuel and insertion of fresh fuel;
3. cooling of irradiated fuel during its removal and transfer to storage bays;
4. underwater storage of irradiated fuel (at least 10 years) until it can be transferred to dry storage
197
Section 24(4) of the Nuclear Safety and Control Act
* No licence may be issued, renewed, amended or replaced
unless, in the opinion of the Commission, the applicant:
✓is qualified to carry on the activity that the licence will
authorize the licensee to carry on; and
✓will, in carrying on that activity, make adequate
provision for the protection of the environment, the
health and safety of persons and the maintenance of
national security and measures required to implement
international obligations to which Canada has agreed
198
skide 134
199
nuclear safety and Control Act
Class I Nuclear Facilities Regulations
Provide requirements for site preparation licence applications,
personnel certifications, record-keeping and sets timelines for
regulatory reviews.
Apply to: Class 1A and Class 1B nuclear facilities including
nuclear reactors, large particle accelerators, nuclear processing
plants, fuel fabrication plants and waste disposal facilities
200
CNSC Design Requirements
CNSC expects the applicant to demonstrate their safety case is commensurate with the risks posed by the facility.
* For any size reactor, this allows for the use of a graded approach to safety.
* CNSC have regulated NPPs to small reactors using this approach. This is consistent with IAEA standards
201
difference between Nuclear power plant and small reactor
Nuclear power plant (NPP): a fission
reactor with a thermal power approx.
greater than 200 MWt (about 75 MW
electrical output)
Small reactor: a fission reactor
with a thermal power approx.
less than 200 MWt
202
Defence in Depth Principle
provisions
defence in depth is to ensure overlapping
provisions are applied to all
organizational, behavioural, and designrelated safety and security activities
✓ provisions for all five levels of defence
(later in presentation)
✓ balanced provisions for prevention and
mitigation
▪ imbalance: e.g., prevention with
inadequate mitigation or mitigation
with little prevention
203
Fundamental Safety Functions of a
Nuclear Reactor
Control of reactivity
Removal of heat from the fuel
Confinement of radioactive material
Shielding against radiation
Control of operational discharges and
hazards substances, as well as
limitation of accidental releases
Monitoring of safety-critical
parameters to guide operator actions
204
REGDOC 2.5.2
Document present the requirements in a hierarchy manner
Allows alternative approach especially for small reactors
* Safety objectives
* Safety criteria
* Safety management
* Design and engineering principles
* General system requirements
* System specific requirements
* Graded approach
205
REGDOC 2.5.2
* Provide the requirements for various aspects of the design including:
✓Environment protection
✓Radiation protection
✓Ageing
✓Human factors
✓Security
✓Safeguards
✓Transportation
✓Accident and emergency response planning
✓Safety analysis
206
REGDOC 2.5.2
* Design requirements have been extended from past CNSC practice, e.g.,
✓Safety classification of SSCs
✓Classification of events (next slide)
✓Adoption of quantitative safety goals
✓Design for severe accidents
✓AOOs to be handled by the control systems
✓Design for reliability (safety systems and safety support systems)
✓Technical Specifications (Operating Limits and Conditions)
✓Consideration of malevolent acts in the design
207
Classification of events
Normal operation
* Anticipated Operational Occurrence (AOO)
✓Frequency of occurrence > 10-2/y
* Design basis accident (DBA)
✓Frequency of occurrence, 10-2/y to 10-5/y
* Design extension conditions (DEC) including severe accidents
✓Frequency of occurrence < 10-5/y
✓We shall not consider this classification
208
Safety Criteria
* Dose acceptance criteria
* ≤ 0.5 mSv for any AOO
* ≤ 20 mSv for any DBA
209
Safety Criteria
* Quantitative safety goals
* Core-Damage Frequency: the sum of frequencies of all event sequences that can lead to significant core degradation < 10-5 /y
* Small-Release Frequency: the sum of frequencies of all event sequences
that can lead to a release to the environment of more than 1015 Bq of I131 <10-5 /y
* Large-Release Frequency: the sum of frequencies of all event sequences
that can lead to a release to the environment of more than 1014 Bq of Cs137 <10-6 /y
210
To design for common cause failures, the
principles of ____ are applied to achieve the
necessary reliability
diversity, separation, and
independence
211
The safety systems and their support systems
shall be designed to ensure that the probability
of system failure on demand from all causes is
lower than --
10^-3
212
All safety systems and their safety support
systems meet the __
single failure criterion.
213
Regulatory Challenges
New designs that are non-water water cooled
reactors
214
Examples of Challenges
Two means of shutdown
systems;
Containment;
* Independent shutdown systems
* Inherent designed shutdown systems
Definition of containment
* Means of containment
* Containment function
* Containment system
215
Pre-Licensing Vendor Design Review – What is it?
how many phases
The reviews take place in three
phases, each of which is conducted
against related CNSC regulatory
documents and Canadian codes &
standards:
✓basically, use REGDOC 2.5.2
216
Pre-Licensing
Vendor Design
Review
phase 1
Phase 1: Pre-Licensing Assessment of Compliance
with Regulatory Requirements:
✓This phase involves an overall assessment of
the vendor's nuclear power plant design against
the most recent CNSC design requirements for
new nuclear power plants in Canada
✓REGDOC-2.5.2,
✓And all other related CNSC regulatory documents and
Canadian codes & standards.
217
Pre-Licensing
Vendor Design
Review
phase 2
Phase 2: Pre-Licensing Assessment for Any
Potential Fundamental Barriers to Licensing
✓This phase goes into further details with
a focus on
✓identifying any potential fundamental barriers
to licensing the vendor's nuclear power plant
design in Canada.
218
Pre-Licensing
Vendor Design
Review
phase 3
Phase 3: Follow-up
✓This phase allows the vendor to follow-up on certain
aspects of Phase 2 findings by:
▪ seeking more information from the CNSC about a
Phase 2 topic; and/or
▪ asking the CNSC to review activities taken by the
vendor towards the reactor's design readiness,
following the completion of Phase 2.
219
R&D Challenges
Due to conceptual
FOAK Designs
list 6
* Materials
✓Higher temperatures
✓Irradiation damage
* Fuel
✓Modified designs need to be tested and qualified
* Hydrogen production
✓Methods
✓Coupling to the plant
* Thermal hydraulics and safety
* Fuel cycles
* Cross-cutting groups (fuel cycles, hydrogen,
economic evaluation, safety, proliferation
resistance) were formed within GenIV
220
Importance of SMRs
give 3
SMRs are flexible systems that can be adapted for non-electric applications, according to their key operating parameters,
i.e., exit working fluid temperature.
They can be integrated effectively into the energy systems supporting communities from largest cities to most remote locations
Opportunity to support applications beyond solely electricity generation
221
How is each SMR different from CANDU?
1. Design?
i. Core design
ii. Fuel
2. Nuclear Steam Supply
i. Steam Generator
ii. Heat transport system
3. Main Reactor Systems
i. Control systems
ii. Moderator
iii. Coolant
iv. Pressuriser
4. Main Reactor Safety Systems
i. Shut off Rods, Poisons
ii. Emergency Core Cooling
5. Containment
6. Balance of Plant
i. Turbine Hall
ii. Generator
iii. Transformers
7. Plant layout?
8. Economics?
222
BWRX-300
WATER COOLED SMALL MODULAR REACTORS (LAND BASED)
is the simplest and most innovative BWR design
* Built by General Electric-Hitachi
* 300 MW(e)
* Water-cooled
* Natural circulation SMR
* Utilizing simple, natural
phenomena driven safety
systems.
* 10th generation of the Boiling
Water Reactor (BWR)
* The design has been developed
with a strict adherence to a
philosophy which follows the
IAEA Defense-in-Depth guidelines.
223
Xe-100 (X Energy, LLC, United States of
America)
* HIGH TEMPERATURE GAS COOLED
SMALL MODULAR REACTORS
The Xe-100 is a pebble bed reactor
* It features a continuous refueling system
with low enriched fuel spheres or
pebbles of approximately 15.5 wt.%
* Enters the top of the reactor and passing
through the core six (6) times to achieve
a final average burnup of 165 000 MWd /
tHM
* The reactor pressure vessel (RPV) and
internal structures are designed for a 60-
year life.
224
* ARC-100 (ARC Nuclear Canada, Inc., Canada)
FAST NEUTRON SPECTRUM SMALL
MODULAR REACTORs
Sodium-cooled, fast flux, pooltype reactor with metallic fuel
* ARC-100 addresses the four
challenges which limits public
acceptance/expansion of nuclear
✓ Small size with affordable upfront cost
✓ liquid sodium coolant, its ‘fast’ neutrons have
much more energy, giving it the capacity to
be fueled with and recycle its own used fuel
✓ utilizes a metallic alloy of uranium instead of
uranium oxide, foundation for its inherent,
walk away safety.
✓ refuels once every 20 years. Replacement of
the entire 20-year fuel cartridge for recycling
reduces the risk of nuclear proliferation.
225
* Integral Molten Salt Reactor
(Terrestrial Energy Inc., Canada)
MOLTEN SALT SMALL MODULAR
REACTORS
440 megawatts-thermal molten salt
fueled reactor
* It features a completely sealed
reactor vessel with integrated pumps,
heat exchangers and shutdown rods
all mounted inside a single vessel;
✓ The sealed core-unit is replaced
completely at the end of its useful
service life (nominally 7 years)
* highest levels of inherent safety
✓ no dependence on operator intervention,
powered mechanical components, coolant
injection or their support systems such as
electricity supply or instrument air in
dealing with upset conditions
226
* E-Vinci heat Pipe Micro Reactor
(GE-Hitachi Nuclear Energy, USA and
Hitachi-GE Nuclear Energy, Japan
MICRO SMALL MODULAR REACTORS
Very Low Pressure
* Passive safety
* Single failure tolerant
* Isothermal
* Self-Regulating
* Demonstration planned
for 2024
* Design Philosophy
* Proven heat pipe technology developed by the Los Alamos National Laboratory for space
application.
* Does not use a bulk primary coolant
* heat is removed from its core using passive heat pipes, limiting the number of its moving parts and
providing overall plant simplicity.
* The design utilizes the inherent safety features to enhance safety and self-regulation
capability in the
* fuel, moderator and
* heat pipes technology
227
General Approach and Philosophy
Assumptions Inherent in the perspective
Defence in Depth Safety Philosophy and
Approach, assumes the following:
Reactor Safety is the ultimate goal
Definition of Risk
✓Nuclear station design will have some flaws
✓Equipment will occasionally fail and
✓Operating personnel will occasionally make
mistakes
✓The key is to ensure sufficient DiD that flaws,
failures and mistakes can be accommodated
without increasing the risk or consequences
of an accident.
✓Chance of injury, damage or loss
✓The frequency of an undesired event
multiplied by its consequences
227
Defence-in-Depth: Concept
Applied throughout the design
process and operation of the plant
✓Provide a series of levels of defence
* aimed at preventing accidents, and
* ensuring appropriate protection in the
event that prevention fails.
✓allows failure to be detected and
compensated for or corrected
✓considers organizational and human
performance
* The levels of defence-in-depth shall
be independent to the extent
practicable and subject to
overlapping provision
228
Defence-in-Depth: Concept
Provide a series of levels of defence aimed at
* aimed at preventing accidents, and
* ensuring appropriate protection in the event that prevention fails
229
Reactor Safety
Nuclear Power Plant Risk
Undesired events at Nuclear stations could lead to the
following consequences
✓Severe Core Damage
✓Large Off-Site Release
✓Public Fatalities (immediate and delayed)
230
Reactor Safety
Potential consequences are given and severe, thus
frequencies must be very low
* Very low frequencies require
___
very deep defence in depth
✓Reduces Risk Nuclear Power Plant Risk
▪ The frequency of an undesired event multiplied by its consequences
231
Cool - Heatsinks
Primary heatsink
breakup heat sink
emergency heatsink
– normal means of cooling the fuel at
power or shutdown
a designated alternate to maintain
normal fuel cooling in case of certain failures in the
primary heatsink when shutdown
cools the fuel to ensure fission
products are contained in the event of an accident (LOCA,
LOFW, earthquake etc.)[
232
defence in depth
does it translate precisely
do it does not
233
Reliability is concerned with the ____
overall
operation of nuclear generating
station
234
In addition to trained and motivated
staff, overall station reliability is a
_____
function of the reliability of systems
and equipment.
235
Reliability is critical piece of the ____
DiD
model and is therefore critical to the
safe operation of our reactors.
236
Reliability is defined as
what do 0 and 1 represent
as the probability that a
device will work adequately for the period intended
(system will perform its design function for a
specified mission time ) under the operating
conditions encountered.
Reliability is a probability with a numerical
value ranging from
✓0 (totally unreliable) to
✓1 (always operates for the time intended).
* If a pump is judged to have a reliability of 0.99
for its first year of operation (based on
historical data for this type of pump),
✓this means that for 1000 hours of
operation the pump will be unavailable for
no more than 10 hours.
237
slide 197 , 198, 199, 120, 121, 122
238
Design Methods for Improving System
Reliability/Availability
Redundancy
* more than one way
to do a job
Independence
* channelization
✓safety system
trip channels
✓odd and even
power
* different supplies
(e.g. fuel tanks,
suction supply,
etc.)
Elimination of
common cause
effects
* environmental
qualification
* seismic
qualification
* group 1 and 2
systems
* diversity
Fail-safe
* failure results in
function performed
239
High reliability and availability can be
achieved by attention to a number reliability
of principles during design and operation of
a station. They are:
✓Redundancy
✓Independence
✓Diversity
✓Periodic testing
✓Fail-safe operation
✓Single-failure criterion
✓Operational Surveillance
✓Preventative maintenance and
✓Predictive maintenance
239
Redundancy
✓ If only one component exists to perform a certain function, when it fails, the system fails.
✓ This problem can be reduced by installing additional components, so that if one fails, there is another to do the job.
✓ In other words, higher reliability can be attained by providing a backup (or redundant) component.
✓ It is important to understand that this redundancy is provided primarily to ensure reliable operation, not to
allow more convenient maintenance.
* In Nuclear Plant Design - Redundancy is the provision of components or capacity in excess of 100% of system
requirements, such that failures of excess components or capacity do not disable the system function.
✓ e.g. two 100% capacity pumps placed in parallel
* Redundancy does not by itself protect against system failures; e.g. bulk electrical failure, loss of pump suction etc.
✓ Redundant equipment must be independent and protected from external common cause failures.
✓ Taking redundant equipment out of service for maintenance will lower the reliability of the system.
✓ Balance of Plant Systems
▪ SST and UST
▪ Digital Computer
o DCCX AND DCCY
✓ Space shuttle program
▪ The computer control system in each shuttle contains more than one computer.
▪ Redundancy is provided by running the same software control program on more than one
computer. If one computer fails, another is immediately available to assume control.
240
* Redundancy and _____Criterion
✓“Redundant” means
Single Failure Criterion
having a second source of power or piece of
equipment that acts as a backup in case the first fails to operate properly.
* The second piece of equipment replicates the function of the first. This is
where designing to the “single-failure criterion” comes into play.
✓Designing to fulfil single failures is an aspect of the defense-in-depth design
mentality
▪ Equipment will fail (One of DiD principles)
✓“Single failure” is defined in REGDOC 2.5.2 and Appendix A of 10CFR50, of
the USNRC
241
✓Single Failure Criterion is --
and example
a requirement that a system which is designed to
carry out a defined safety function must be capable of carrying out its mission
in spite of the failure of any single component within the system or in an associated system which supports its operation
▪ Example, Emergency Core Cooling System.
o Its electrical power system can be performed by an “Electrical
isolation” achieved “by the use of separation distance, isolation
devices, shielding and wiring techniques, or combinations thereof.”
242
Single-failure criterion
in relation to safety groups
requires that each safety group can ___
✓All safety groups shall function in the presence of a single failure.
perform all
safety functions required for a PIE in the presence of any single component
failure, as well as:
▪ all failures caused by that single failure
▪ all identifiable but non-detectable failures, including those in the nontested components
▪ all failures and spurious system actions that cause (or are caused by) the
PIE
243
Common
Mode/Cause
Failures
A concurrent failure of two or more structures,
systems or components due to a single specific
event or cause, such as natural phenomena design
deficiency, manufacturing flaws, operation and
maintenance errors, and human-induced
destructive events.
244
Design Methods to Counter failures
✓Independence
✓Odd/Even equipment
✓Diversity
✓Separation
✓Channelization
245
Odd and
Even
Equipment
what does it mean
* Electrical power supplies are designated as odd or
even
✓odd and even supplies are independent
✓this is another example of channelization
* Typically, half the equipment providing a function
is supplied by an even source, and half by an odd
source, so that the effect of one power supply
failure is limited to either odd or even equipment.
✓This eliminates some power failures as a
common mode failure for odd and even
equipment.
✓If the odd and even equipment provides at
least 100% redundancy, then system failure is
prevented for these power failures (e.g. SDC
pumps)
246
Independence
what is it
example in candu
physical separation of systems or components so that a fault in one system will not affect the others.
✓ In dependence is achieved by having no shared components or common services (functional separation), and by physical separation.
✓ Using the space shuttle, an example of independence is separate power supplies for each of the computers. This way failure of the power supply to a computer does not at the same time disable the other computers.
* In CANDU use of two independent shutdown systems SDS1 and SDS2
➢ Two systems are independent if the failure of one, or the failure of any system or structure necessary to support it, cannot cause the other to fail.
➢ SDS1 failure of power it works as rods
➢ SDS2 actuation of helium gas injection of poison
SDS 1 and 2 have no shared
components or services, such as
electrical power, and components
are physically separated North and
South of the reactor.
2
247
Reliability principles - Diversity
def
in candu
in balance of plant systems
It is variety in design, manufacture, operation and maintenance of redundant components or systems for the
purpose of reducing unavailability due to common cause effects, such as design or manufacturing flaws,
unforeseen failure modes, and Operating and Maintenance (O&M) errors.
✓ Diversity is an attempt to ensure that there is more than one way of doing a job.
✓ The presence of two or more redundant systems or components to perform an identified function, where
the different systems or components have different attributes so as to reduce the possibility of common cause failure.
* In CANDU use of two diverse shutdown systems SDS1 and SDS2
✓ SDS1 achieves emergency shutdown by dropping shutoff rods into core under gravity, whereas SDS2 injects
liquid absorber under pressure.
✓ SDS1 and SDS2 components, and SGs and EPGs are from different manufacturers
* Balance of Plant Systems
✓ Four Classes of Power
✓ Digital Computers DCCX and DCCY
248
Diversity avoids Common Cause Failure - how?
✓Using the space shuttle, diversity is provided by running entirely
different software control programs on different computers to achieve
the same purpose.
✓The software is even created by a different design team. This ensures
that a bug in one piece of software is not duplicated in the other so
that one mistake cannot disable more than one computer (common
cause failures)
249
what avoids common cause failure
Diversity
250
A system or component is called fail safe if ____
give ex
after failing it leaves the remainder of the system in a safer state.
✓ Failure does not contribute to system unavailability.
✓ Failures are not eliminated, but the failures are safe:
Examples
* SDS1; Power failure releases the rods into the core
✓ Rods are energized during operation
* Train locomotives are equipped with a deadman brake.
✓ It must be depressed by the engineer to allow the locomotive to move.
✓ If the engineer falls over dead, his foot will come off the brake and the locomotive will come to a
halt
251
Reliability principles- Separation
* Cables routed in different cable trays
* Separate rooms for odd and even Class III switchgear
* Separate control areas. (MCR, SCAs - secondary control areas)
252
Reliability principles – Periodic Testing
what is it
When a component in a process system fails the effects are
immediately apparent.
Failure of a poised system, on the other hand, is not readily
apparent and can only be determined by testing.
✓ Since it is not possible to determine at what point the failure
occurred, unavailability is considered to be half the time since the
system was last tested (plus however long it takes to make the
repairs). Unavailability can be kept low by more frequent testing.
✓ The frequency of testing must, however, be balanced against:
▪ Wear and tear on the system and components caused by
testing,
▪ Unavailability due to removing components from service for
the duration of the test
▪ The risk (by human error) of leaving the system in a degraded
state after a test, and
▪ The danger of activating the system during the testing process.
253
Reliability principles – Operational Surveillance
✓Process of continual monitoring and trending of process parameters
and equipment with the intent of spotting potential problems before
they become real problems
✓Thus, corrective action can be taken before a major problem occurs.
✓An example is vibration monitoring of rotating equipment. If unusual
vibrations are detected, the equipment can be stopped and repaired
before the vibration causes serious damage.
254
Preventive Maintenance
✓Reliability data on different types of equipment offers a means of estimating
when failures are likely to occur.
✓By planning replacement or maintenance before any appreciable
deterioration occurs that can contribute to the predicted failure, it is possible
to reduce the number of unscheduled outages and consequent loss of
production.
✓This sometimes has the appearance of throwing away good equipment, but
the reliability statistics indicate that the equipment is likely to fail shortly and
probably inconveniently (remember Murphy’s Law).
* The best form of preventive maintenance is predictive
maintenance
* Based on equipment condition.
* Maintenance or replacement is only done when diagnostic test results
(such as vibration monitoring) indicate equipment degradation.
255
Multiple Physical Barriers
2 kinds
Multiple fission product barriers
* Multiple radioactivity barriers
256
The multiple barrier approach that
has been built into station design:
✓is intended to
prevent or impede
the release of radioactivity from
the fuel to the public
257
* There are __ passive radioactivity
barriers continuously available
they are :
5
1. The uranium fuel is molded into ceramic fuel pellets
which have a high melting point and lock in most of the
fission products
2. The fuel sheath which is made of high integrity welded
metal (zircaloy) and contains the ceramic fuel
3. The heat transport system which is constructed of high
strength pressure tubes, piping and vessels and contains
the fuel bundles
4. The containment system which provides a relatively
leak tight envelope maintained slightly below
atmospheric pressure (Except CANDU-6). This partial
vacuum encourages air to leak in instead of out
thereby helping to prevent release of radioactivity
that escapes from the heat transport system, and
5. The exclusion zone of at least one kilometre radius
around the reactor that ensures any radioactive
releases from the station are well diluted by the
time they reach the boundary
258
Process systems are ___
ex
those systems
performing a continuous function in the
normal operation of the plant.
✓For example, the primary heat transport
system is a process system that is continuously
active in the removal of heat from the fuel
259
✓The reactor regulating system is a process
system that ____
is continuously active in the
normal control of reactor power.
260
Safety systems are poised systems that
___
how do they do this
what does reliability mean in this context.
operate only to compensate for the failure of
process systems.
* They can do this by shutting down the reactor to
regain control (shutdown systems),
✓by providing additional cooling to the fuel (emergency
coolant injection system), and
✓by containing radioactivity, which has escaped from
the fuel (containment system).
Reliability in this context means that in the
rare event these systems are called upon to
act, they will be available to perform their
intended function.
260
* Reliable process systems ensure that:
✓heat is produced and electricity generated
✓while maintaining control, cooling and
containing.
261
* The safety systems are designed to operate
automatically and the five passive barriers
are always in place, but the DiD concept
does not allow reliance on equipment and
systems to prevent accidents.
✓It is important that operating and maintenance
staff are knowledgeable about system
conditions, alert for any evidence that systems
or equipment may be on the verge of failure,
and act promptly to prevent or minimize the
consequences of such failures.
✓To achieve a high level of competence, the
qualification criteria for each job family are
clearly defined.
✓Considerable effort goes into performancebased training of staff to meet those criteria and
maintain their qualification.
262
Adequate detection and correction of failures requires
not just competent staff but also processes and
procedures for the staff to carry out in a systematic
fashion.
✓For example, a routine testing program for safety
systems helps meet the availability targets
✓An operational surveillance program in conjunction
with a planned preventive maintenance program
helps to ensure that equipment and systems are
monitored, inspected and repaired before they fail.
✓Failures, when they do occur, are thoroughly
investigated and solutions applied through a
rigorous change approval process.
✓Elaborate work control processes exist, allowing the
quick reporting, prioritizing and repair of
deficiencies
263
REGDOC 2.5.2
level 1
shall include conservative design and high-quality construction to provide confidence that plant failures and
deviations from normal operations are minimized and accidents are prevented. This shall entail careful attention to selection of
appropriate design codes and materials, design procedures, equipment qualification, control of component fabrication and
plant construction, and use of operational experience
264
REGDOC 2.5.2
level 2
* Level Two -shall be achieved by controlling plant behaviour during and following a postulated initiating event (PIE) using both
inherent and engineered design features to minimize or exclude uncontrolled transients to the extent possible
265
REGDOC 2.5.2
level 3
* Level Three - shall include the provision of inherent safety features, fail-safe design, engineered design features, and
procedures that minimize the consequences of DBAs. These provisions shall be capable of leading the plant first to a controlled
state, and then to a safe shutdown state, and maintaining at least one barrier for the confinement of radioactive material.
Automatic activation of the engineered design features shall minimize the need for operator actions in the early phase of a
DBA.
266
REGDOC 2.5.2
level 4
* Level Four - shall be achieved by providing equipment and procedures to manage accidents and mitigate their consequences
as far as practicable. Most importantly, adequate protection shall be provided for the confinement function by way of a robust
containment design. This includes the use of complementary design features to prevent accident progression and to mitigate
the consequences of DECs. The confinement function shall be further protected by severe accident management procedures
267
REGDOC 2.5.2
level 5
* Level Five - The design shall provide adequately equipped emergency support facilities, and plans for onsite and offsite
emergency response
268
DiD levels match (approximately) to plant states:
list 1-5
1. Normal operation (NO)
2. Anticipated Operational Occurrences (AOO)
3. Design Basis Accidents (DBA)
4. Design Extension Conditions (DEC)
5. Events more severe than DEC
269
Defence-in-Depth Framework
level 1
implementation
To prevent deviations from normal operation, and to prevent
failures of structures, systems and components important to
safety
* Conservative design
* High-quality construction (e.g., appropriate design codes and materials,
design procedures, equipment qualification, control of component
fabrication and plant construction, operational experience) (NO)
270
Defence-in-Depth Framework
level 2
implementation
2. To detect and intercept deviations from normal operation, to
prevent anticipated operational occurrences from escalating
to accident conditions and to return the plant to a state of
normal operation.
* Inherent and engineered design features to minimize or exclude
uncontrolled transients to the extent possible
(AAOs)
271
Defence-in-Depth Framework
level 3
implementation
3. To minimize the consequences of accidents, and prevent
escalation to beyond-design-basis accidents
* Inherent safety features
* Fail-safe design
* Engineered design features, and procedures that minimize consequences
of DBAs (DBA)
272
Defence-in-Depth Framework
level 4
implementation
4. To ensure that radioactive releases caused by severe
accidents or design-extension conditions are kept as low as
practicable.
* Equipment and procedures to manage accidents and mitigate their
consequences as far as practicable
* Robust containment design
* Complementary design features to prevent accident progression and to
mitigate the consequences of design-extension conditions
273
Defence-in-Depth Framework
level 5
implementation
5. To mitigate the radiological consequences of potential
releases of radioactive materials that may result from
accident conditions.
* Emergency support facilities * Onsite and offsite emergency response
plan (Beyond DEC
274
Qualitative Safety Objectives for New Nuclear Power Plants
* General nuclear safety objective:
✓ Design and operate nuclear power plants (NPPs) in a manner that will protect individuals, society and the
environment from harm.
✓ This objective relies on the establishment and maintenance of effective defences against radiological hazards in NPPs.
275
Qualitative Safety Objectives for New Nuclear Power Plants
Technical safety objective
✓ Provide all reasonably practicable measures to prevent accidents in the NPP, and mitigate the consequences of
accidents if they do occur.
✓ This takes into account all possible accidents considered in the design, including those of very low probability. Any
radiological consequences will be below prescribed limits, and the likelihood of accidents with serious radiological
consequences will be extremely low
276
Qualitative Safety Objectives for New Nuclear Power Plants
A limit is placed on the ____ shall be provided a level of
protection from the consequences of NPP operation, such that there is no significant additional risk to the life and health of
individuals.
societal risks posed by NPP operation Individual members of the public
✓ Societal risks to life and health from NPP operation shall be comparable to or less than the risks of generating electricity by
viable competing technologies, and shall not significantly add to other societal risks.
277
Qualitative Safety Objectives for New Nuclear Power Plants
* The design shall be such that plant states that could lead to ___
significant radioactive releases are practically eliminated.
✓ For plant states that are not practically eliminated, only protective measures that are of limited scope in terms of area and
time shall be necessary for protection of the public, and sufficient time shall be made available to implement these measures
278
Quantitative Safety Goals for New Nuclear
Power Plants
Core damage frequency
* The sum of frequencies of all event sequences that can lead to significant core degradation shall be less than
__ per reactor year.
10^-5
279
Quantitative Safety Goals for New Nuclear
Power Plants
Large release frequency
* The sum of frequencies of all event sequences that can lead to a release to the environment of more than
___ becquerels of __shall be less than ___ per reactor year. A greater release may require ___
10^14
cesium-137
10^-6
A greater release may require long
term relocation of the local population.
279
Quantitative Safety Goals for New Nuclear
Power Plants
Small release frequency
* The sum of frequencies of all event sequences that can lead to a release to the environment of more than
__ becquerels of __ shall be less than __ per reactor year.
* A greater release may require ___
10^15
iodine-131
10^-5
temporary evacuation of the local population.
280
Post-Fukushima Enhancements to Defence in
Depth (1)
recommendations
✓ Strengthening defence in depth
▪ external events and beyond design basis accidents
▪ design and safety analysis
▪ severe accident management
✓ Enhancing emergency preparedness
▪ onsite and offsite emergency response
✓ 5 Levels of DiD
✓ adequate
✓ opportunity for improvement
281
Post-Fukushima Enhancements to Defence in
Depth 2
general req
✓ Improving regulatory framework and
processes
▪ Regulatory framework
▪ Industry and operator oversight
procedures
✓ International collaboration
▪ CANDU countries
▪ Other NPP regulators
282
Post-Fukushima Enhancements to Defence in
Depth
* Level 3: Protecting spent fuel pools
✓ Makeup water capability and instrumentation
283
Post-Fukushima Enhancements to Defence in
Depth
* Level 4: Preventing and mitigating severe accidents
✓ Protecting fuel
✓ Makeup water capability to steam generators / primary heat transport system / emergency core
coolant / dousing spray
✓ Preventing severe core damage
✓ Makeup water capability to moderator system and calandria vessel/vault
✓ Protecting containment
✓ Passive recombiners and containment venting
✓ Severe accident management guidelines validation/exercise
284
Post-Fukushima Enhancements to Defence in
Depth
* Level 5: Protecting the public
▪ Containment filtered venting
▪ Integrated emergency plans and full-scale emergency exercises
285
Reactor Defence in Depth Protect Fuel (1)
* Analyses and reassessments
✓ site-specific magnitudes of external
events * high winds, seismic, tsunami /storm surges,
flooding
✓ station blackout event on spent fuel
bundles inside fueling machine
✓ Multi
-unit events
286
Reactor Defence in
Depth Protect Fuel (2)
* Design improvements
✓ Emergency mitigating equipment ▪ mobile equipment ✓ Water makeup connections to ▪ steam generators ▪ heat transport system ✓ Provision for main steam safety valves after
station blackout
✓ Upgrades of power systems ▪ load shedding to extend battery
availability
▪ power supply for key instrumentation
✓ Protection against flooding (barriers, water
-
tight doors, sealing penetrations
287
Reactor Defence
-in
-Depth
Prevent Severe Core Damage
* Analyses and reassessments
✓ modelling for multi-unit plant events
✓ Reassessment of control room habitability
during emergencies
✓ Instrumentation qualification for severe
accident
288
Reactor Defence
-in
-Depth
Prevent Severe Core Damage
* Design improvements
✓ Water makeup connections to
▪ calandria vessel
▪ calandria vault
▪ relief capability of calandria/vault
▪ Instrumentation upgrade
289
Reactor Defence in Depth Protect
Containment
* Analyses and reassessments
✓ Enhancement of filtered containment venting system
✓ Severe accident management guidelines (SAMGs)
✓ Instrumentation for SA conditions monitoring (qualify
existing or new)
✓ Control facilities’ habitability during SA
✓ Improved modelling of SAs for multi-unit plants
290
Reactor Defence in Depth Protect
Containment
* Design improvements
✓ Filtered containment venting
✓ Passive autocatalytic recombiners (PARS)
291
Reactor Defence in Depth
Protect Spent Fuel Pools
Analyses and reassessments
✓Structural integrity check for
temperatures
* Design improvements ✓Instrumentation for pool parameters ✓Piping and connections for extra
heat sinks
✓Procedure in event of loss of heat
sinks (pool water)
292
Reactor Defence in Depth
Protect Spent Fuel Pools
Analyses and reassessments
✓Structural integrity check for
temperatures
* Design improvements ✓Instrumentation for pool parameters ✓Piping and connections for extra
heat sinks
✓Procedure in event of loss of heat
sinks (pool water)
293
How Safe is the
Safety Case?
✔ the risk from nuclear reactors
✔ radioactive material is in the
reactor fuel/core
✔ risk can be quantified
✔ society can tolerate a level of
risk that is reasonable
✔ Epidemiological studies
✔ Dose limits
294
Design Basis Accidents
DBAs are the set of accidents for which
the designer makes explicit provision (defence),
while remembering that more severe or peculiar accidents can occur and ensuring that
his/her design has some capability to deal with them.
✔There is no way of identifying possible accidents beforehand
✔Technology is replete with unpleasant surprises, especially at the beginning
✔Technologies could have their accidents early on (Boeing 737 MAX)
✔Nuclear Power Plants have had their share too
▪ TMI-3, Chernobyl, Fukushima
295
CNSC Safety Goals and Risk Assessment
Quantifying _-___
Risk, and present probabilistic safety assessment tools used to
determine if plant meets safety goals.
296
CNSC Safety Goals and Risk Assessment
Safety Goals - “how safe is safe enough?”
“safe as possible” mean different when no guidance is given to designer.
* “The reactor must never have a severe accident” - physically impossible and
expectations that cannot be complied with
297
CNSC Safety Goals and Risk Assessment
CANDU 3 safety goals
* Recall: the 3 safety goals (REGDOC 2.5.2)
* (1)) Core damage (2) Large Release Frequencies (3) Small Release Frequencies
298
Mitigating Systems
* Mitigate
✔ shut down
✔ remove decay heat
✔ contain radioactive material
✔ monitor state of the plant
299
Some Accidents by Phenomena
Reactivity accidents
* reactor coolant inventory
✔Typically called LOSS OF COOLANT
ACCIDENTS (LOCA)
* reactor coolant pressure transients
* Pressure control (increase)
* secondary-side heat removal transients
* Moderator and shield-cooling system failures
* Fuel-handling accidents
300
Elements of
Safety
Analysis for a
Safety Case
Main Elements
* Development of Trip Set Points
* Initial Conditions
* Initiating Event
* Event Sequence
* Barrier Protection
The Last 3 is - “Accident Walk Through”
Discussion of the elements next
301
Safety Analysis: Trip Set Points
what are they
Automatic Instruments monitor reactor conditions:
✔heat transport system pressure,
✔reactor power and coolant flow.
302
Safety Analysis: Trip Set Points
Any ___, triggers a reactor shutdown
this is called?
measurement with unsafe operating condition
* A shutdown by a protective system is called a reactor trip.
303
Safety Analysis: Trip Set Points
how do they occur
what are manual trips called
Trip set points
* have conservative safety margins exist.
✔a trip occurs automatically whenever a trip parameter
exceeds its trip set point (a limit).
✔There is manual trips called operator actions
304
Safety Analysis: Initial Conditions
hoe are they chosen
what are they based on
A number of key parameters are chosen in a “conservative” direction for analysis
Based on
✔ Initial plant conditions,
✔ system performance
measures, and
✔ assumptions on
unavailability of
mitigating systems or
portion of it.
305
Safety Analysis: Initiating Event
✔assumed that the pipe break is instantaneous,
✔ bears little relationship to reality
✔selected to ensure conservatism in that it maximizes the predicted coolant voiding rate
and hence the coolant void-reactivity insertion and reactor power pulse.
✔greatest challenge to shutdown-system effectiveness
306
Safety Analysis: Initiating Event
Accident Walk-Through: Large Loss of Coolant Accident (LOCA)
Accident Walk-Through: Large Loss of Coolant Accident (LOCA)
✔Large Heat Transport System pipe break.
✔A large LOCA in a CANDU
✔ break area is larger than twice the cross-sectional area of the largest feeder pipe. (Since being
changed)
✔A large LOCA can be located only in the large piping above the core. There are three
representative locations (think of largest challenge)
▪ reactor inlet header (RIH),
▪ reactor outlet header (ROH), and
▪ pump suction line (PSH).
307
Safety Analysis: Event Sequence
* Estimate the approximate time of the following
-
-
-
* Event sequence is usually best represented by the concept of ___
described by
✔ reactor trip
✔ start of fuel overheating
✔ failure of first channel
✔ core collapse
✔ shield-tank failure (CANDU specific)
✔ containment behaviour
stylized accidents
(design basis accidents- DBA) and run using computer codes/analytical tools
✔ described later in Analytical Models and Computer Tools
308
safety Analysis - Barrier Protection
Safety functions required in a nuclear
reactor:
1. control by shutting down the
reactor in case of transient
2. remove decay heat to cool
3. contain any radioactive
material
4. monitor the state of the plant.
309
the risk from nuclear reactors
comes from accidental ____
✔therefore, one can postulate
accidents (safety analyses)
which might allow ___
Release of radioactive material.
▪ radioactive material is the
reactor fuel.
radioactivity to escape
310
Nuclear
Plant
Safety
Assured
Reliable safety is knowledge based
Supported by
in Candu
* regulations and
* Design engineering and
* computer codes and analytical work (Tools)
CANDU reactors
* safety analysis (computer codes and analytical
tools) to support the safety case
* 4 special safety systems (mitigating systems);
* principles of Defence in Depth
311
* Reactor containment
✔A Special safety systems
✔contains the reactor vessel/calandria
✔Contains radioactive fuel.
* The containment the last barrier left preventing
large releases
* Requirements are formed from CNSC regulations
and
* Fulfilments of requirements must be verified with
safety analyses.
312
Nuclear Plant and
Safety Analysis
Computer Codes/Tools
* Used to model the accident sequence
to ensure safety of plants
Code Validation and
Verification
* The codes are validated against
experiments
✔ demonstrate they can predict the
reality
✔plant can operate as designed (NO)
and
✔respond in the event of abnormal
events (AAOs, DBAs) and beyond its
design
313
CATHENA
Brief code description
Brief code description
✔Canadian Algorithm for THErmal-hydraulics
Network Analysis(CATHENA)
✔one-dimensional, two-fluid system code
▪ Steam (Gas) and Liquid
✔flow regime dependent relations coupled with the
two-phase model
✔full network, user defined by input file
✔modeling of heat transfer in pin bundles
✔heat transfer correlations for entire boiling curve
✔built-in temperature dependent property tables
✔variety of component models available
▪ generalized tank, valves, discharge break,
pump, etc
313
Mathematical/Analytical
Tools and Computer
Codes
Used in safety analysis and assessments
Consists of the basic science underlying safety
analysis utilizing analytical correlations and
models
These tools/codes/models allow analysts to
effectively and efficiently meet the following
objectives:
* Evaluation acceptance criteria
* understanding of the system behaviour
* licensee (vendor/operator) reactor licence
* Covers various disciplines from the upstream analysis of
reactor physics to the downstream analysis of dose
* User’s requirements - fed back to the code developer;
therefore, an excellent/strong interface is established
314
TUBRUPT
Brief code description
✔TUBe RUPTure
✔used to determine the pressure transients
within calandria vessel due to the injection of
fuel channel content during in-core break
accidents
✔phenomena modeled
▪ flashing coolant hydrodynamic transient in
moderator
▪ high temperature channel debris
interaction with water
▪ ruptures channel projectile formation and
impact on the calandria vessel, shutoff
rods guide tubes, and other fuel channels
314
CATHENA
* Code general application
✔multi-purpose reactor cooling system thermalhydraulics and thermo-mechanical analysis
✔full reactor cooling system network transient
analysis
✔reactor fuel channel transient analysis
✔secondary side transient analysis
✔reactor fuel bundle transient analysis
✔fuel channel thermo-mechanical transient
analysis
✔ECCS system operation analysis
✔auxiliary system thermal-hydraulics transient
analysis
✔capability to be coupled with other codes (e.g.
ELOCA, etc.)
315
MODTURC_CLAS
* Brief code description
✔ MODerator TURbulent Circulation Co-Located Advanced
Solution
✔ consists of coupling CANDU moderator related specific
modules
✔ volume-based porosity & distributed hydraulic resistance
✔ two-equation k-epsilon model of turbulence
✔ code capabilities include
▪ calculation of pressure losses in the calandria tube
array
▪ calculation of the volumetric heat load distribution
in the calandria vessel from steady-state and
transient neutronic power and radioactive decay
distributions
▪ simulation of the moderator temperature control
system
▪ modeling of the moderator heat exchangers and
associated control valves
▪ the setting up of transient boundary conditions
(inlet/outlet mass flows, transient poison
concentration and other scalar inlet/outlet
conditions, and restart capability)
316
MODTURC_CLAS
* Code general application
✔used in nuclear safety analysis to predict velocity,
temperature and/or poison concentration
distributions (corresponding to SDS2 activation)
✔this information is used to determine:
▪ moderator subcooling availability for the
following postulated accident scenarios:
▪ large break LOCA’s with and without
ECC(involving PT/CT contact heat loads to the
moderator)
▪ loss of moderator circulation
▪ loss of moderator cooling
▪ the poison distribution corresponding to:
❑ in-core, single-channel breaks in an overpoisoned guaranteed shutdown state
317
TUBRUPT
✔Code general application
✔estimate the extent of in-core damage due to
a single fuel channel rupture caused by either
of the following scenarios
▪ spontaneous pressure tube /calandria tube
rupture
▪ severe flow blockage
▪ feeder stagnation break
318
TUBRUPT
✔code calculates the following parameters
▪ moderator pressure transient
▪ damage mapping of
❑adjacent channels to the broken fuel
channel
❑shut-off rods guide tubes
❑calandria vessel
319
TUBRUPT
Application
Used to determine __
estimate the extent of ___
the peak pressure pulse in the
calandria vessel when a PT/CT ruptures
in-core damage due to a single
fuel channel rupture
320
SMART
* Brief code description
✔Simple Model for Activity Removal and
Transport
✔calculates radionuclide behavior in
containment
✔the code is composed of a set of onedimensional, partial differential equations that
describe the aerosol and fission product
behavior
✔an aerosol general dynamics equation is solved
to calculate aerosol size distribution as a
function of space and time
✔mass conservation equations are solved to
predict fission product concentrations in
containment
321
SMART
* Code general application
✔calculates aerosol-fission product behavior
within containment under accident conditions
and releases to the outside atmosphere
✔provides details about the concentration of
individual isotopes present in various parts of
the containment
✔provides input to the ADDAM code for
calculating dose to public and station staff
322
ADDAM
* Brief code description
✔Atmospheric Dispersion and Dose Analysis
Method
✔it is a Gaussian Dispersion model
✔used for analysis of hypothetical accident
releases of radioactive material to the
atmosphere from CANDU stations
✔conceptual model is based on modeling 15
atmospheric dispersion phenomena (covered
in the ADDAM validation matrix)
✔ADDAM calculates concentrations of
radioactivity in the air and on the ground and
doses to members of the public following an
atmospheric release
323
ADDAM
* Code general application
✔Simulates the dispersion of aerosol-fission
product release into the environment
✔considers release height
✔condisers time and duration of release
✔Considers weather pattern(s)
324
325
✔ADDAM calculates concentrations of
Radioactivity
in the air and on the ground and doses to
members of the public following an atmospheric
release
326
MAAP
* Brief code description
✔Modular Accident Analysis Program
✔a family of integrated computer codes designated
for Severe Accident Analysis in nuclear plants, used
by more than 40 international utilities
✔intended for PSA Level 2 analysis (out of scope for
this course)
✔name of code is MAAP-CANDU in CANDU Reactors
✔MAAP_CANDU has models for horizontal CANDUtype fuel channels and CANDU-specific systems,
such as:
▪ calandria vessel,
▪ reactor vault,
▪ reactor cooling system,
▪ containment systems (dousing), etc.
327
MAAP
* Code general application
✔ MAAP_CANDU calculates severe accident progression starting from
normal operating conditions for a set of plant system faults and
initiating events leading to:
▪ reactor cooling system inventory blow-down or/and boil-off
▪ core heat-up and melting
▪ fuel channel failure and core disassembly
▪ calandria vessel failure
▪ shield tank / reactor vault failure
▪ containment failure
328
MAAP
✔ Physical processes modeled
▪ thermal-hydraulics processes in reactor cooling system,
calandria vessel, reactor vault and shield tank, end-shield, and
containment components
▪ core heat-up, melting and disassembly
▪ zirconium oxidation by steam and hydrogen generation
▪ material creep and possible rupture of reactor cooling system
components, calandria vessel and shield tank walls
▪ ignition of combustible gases
▪ energetic and molten corium-coolant interactions
▪ fission product release, transport and deposition
329
GOTHIC
* Brief code description
✔Generation Of Thermal-Hydraulic Information
for Containment
✔multi-dimensional thermal-hydraulic code
specialized for containment analysis
▪ code allows for hybrid modeling of
containment volumes, ie, combinations of
lumped parameter, 1D, 2D or 3D volumes
▪ conservation equations are solved for 3
fields
❑steam/gas mixture
❑continuous liquid
❑liquid droplet
▪ thermal non-equilibrium is allowed
between phases and unequal phase
velocities
330
GOTHIC
* Code general application
✔treatment of momentum transport in multidimensional models
▪ With optional models for turbulent, mass and
energy diffusion
✔hydrogen combustion
✔engineering model options include
▪ pumps, fans, valves, doors, heat exchangers,
fan coolers
▪ vacuum breakers, spray nozzles
▪ coolers, heaters, volumetric fans
▪ Hydrogen recombiners and ignitors
▪ pressure relief valves
✔modeling of solid structures (thermal conductors)
for flat plate (e.g., walls), cylindrical tube, solid rod
331
Golden Rules
of Reactor
Safety
Design should enable nuclear fission to proceed in a
controlled manner
CONTROL
* Design should be able to remove the heat produced
COOL
* Design should ensure that radioactivity is contained
throughout all operational and postulated accident
conditions
CONTAIN
332
1. Service Water Unexplained Failures
Context:
* Class III powered LPSW pump
* Full overhaul conducted
* Did not pass post-maintenance
testing:
* High vibrations
* Trips on thermal protection
* Out of service for two years
* What is the consequence?
* Potential Causes?
Loss of redundancy on Loss of
Class III power.
* Increase degradation on
other pump since no duty
swaps
Improper bearing installation
* Poor motor windings
* Thermal trip continues to keep
the pump out of service.
* Out of options…we pull the pump
out once again.
333
3.Recirculation
Gate Failure
Context:
*Middle of winter
*Lake temperature low
*Call from Shift Manager
*“Seems like a cable is
broken on the recirc gate”
* No heating of intake, so what?
* Cold lake and ambient air temperature
* Formation of Frazil Ice
* Blocks pump intakes and starves the pumps.
which of 3 c's
COOL
Keep Fission Products Within the
Fuel
✓ Cool the Reactor and Spent
Fuel ▪ Maintain coolant
inventory
▪ Maintain coolant flow ▪ Maintain coolant heat
sinks
334
3.Recirculation
Gate Failure
Context:
*Middle of winter
*Lake temperature low
*Call from Shift Manager
*“Seems like a cable is
broken on the recirc gate”
intake supplies
* Low Pressure Service Water (LPSW)
* Moderator
* Shutdown Cooling
* Maintenance Cooling
* End Shield Cooling
* Vault Coolers
* Condenser Cooling Water (CCW)
* ……yup….the condensers
