Safety of Nuclear Power Plants Generation I to Generation IV

Transcrição

Safety of Nuclear Power Plants
EPR,, GIF,, HTR
Generation I to Generation IV
Spring 2011 / Prof. W. Kröger
2
1
EPR Overall Safety Philosophy
 These objectives, motivated by the continuous search for a
hi h safety
higher
f t level,
l
l involve
i
l reinforced
i f
d application
li ti off th
the
defense in depth concept:
 by improving the preventive measures in order to further reduce the
probability of core melt, and
 by simultaneously incorporating, right from the design stage,
measures for limiting the consequences of a severe accident.
3
EPR Plant Parameter
 Thermal power
4250/4500 MW
 Electrical power
1600 MW
 Efficiency
36%
 No. of primary loops
4
 No of fuel assemblies
241
 Burnup
> 60 GWd/t
 Secondary pressure
78 bar
 Seismic level
0.25 g
 Service live
60 years
4
2
Risk Reduction – EPR Approach
Reinforced protection in the unlikely event of core meltdown
 Preventive features include safetyy devices which further reduce the probability
p
y
of a severe accident
 enlarged water inventory of the main primary system and of the steam
generators; increased reliability of safety systems through
 4-fold 100% redundancy (4-train concept);
 use of diversified technologies for each train of these systems.
 Features to mitigate the consequences of such an event:
 the extremely robust, leaktight containment around the reactor is designed
to prevent radioactivity from spreading outside;
 the arrangement of the blockhouses inside the containment and hydrogen
catalytic recombiners (passive devices) prevent the accumulation of
hydrogen and the risk of deflagration;
 molten core escaping from the reactor vessel would be passively collected
and retained, then cooled in a specific area inside the reactor containment
building (“core catcher”)
Enhanced protection against external hazards
5
EPR – Safety Systems
6
3
EPR Safety Injection / Residual Heat Removal (SIS/RHRS)
7
Risk Reduction - EPR Containment
The unique building housing the reactor is extremely robust. It rests on a 6
m thick concrete base mat and is enclosed by a double shell: the inner is
made of leaktight, prestressed concrete and the outer one of reinforced
concrete, each 1.30 m thick. This total to 2.60 m concrete thickness,
capable of withstanding external hazards such as an aircraft crash.
8
4
EPR Core Catcher
Even in the event of the core melting, and piercing then escaping from
the steel reactor vessel in which it is housed, it would be contained in a
dedicated spreading compartment. This compartment is then cooled to
remove the residual heat.
9
EPR Containment Heat Removal System
10
5
Generation IV International Forum (GIF)
 To meet future energy needs, ten countries have agreed
on a framework
f
k for
f international
i t
ti
l cooperation
ti in
i research
h ffor
an advanced generation of nuclear energy systems, known
as Generation IV.
 Euratome joined as 11th full member;
Switzerland joined in February of 2002
11
Generation IV – Overall Goals
 Development of one or more nuclear energy systems
 Deployable by 2030
 With significant advances in:




Sustainability
Safety and reliability
Proliferation and physical protection
Economics
 Competitive in various markets
 Designed for different applications:
Electricity, Hydrogen, Clean water, Heat
12
6
Generation IV Initiative – Aims (1/2)
Nuclear energy-systems including fuel cycles of the 4th
generation
ti should
h ld ...
SR-1 ... be excellent regarding safety and reliability while in
operation.
SR-2 ... have very low core damage frequency and little
consequences.
SR-3 ... eliminate the need for emergency
g
y planning
g outside of
the plant.
EC-1 ... clear lifecycle-cost advantages over other energy
sources.
EC-2 ... comparable financial risk to other energy projects.
13
Generation IV Initiative – Aims (2/2)
Nuclear energy-systems including fuel cycles of the 4th generation
should ...
SU-l ... produce sustainable energy, following regulations for air
pollution prevention and enhancing the long term availability
of the system and efficient usage of fuel.
SU-2
... minimise the nuclear waste and disposing it, especially they
g term scale and
will reduce administration efforts on a long
hence improve the protection of health and environment.
SU-3
... increase the certainty that they are an undesirable and
difficult source to obtain dangerous materials for usage in
weapons.
14
7
Selected Generation IV Concepts (out of 21)
GEN IV Concepts
Acronym
Pressure
Spectrum Fuel cycle
Temperature
Sodium Cooled Fast
Lead Alloy-Cooled
Gas-Cooled Fast
Very High Temperature
Supercritical Water Cooled
Molten Salt
SFR
LFR
GFR
VHTR
SCWR
MSR
Fast
Fast
Fast
Thermal
Th.&Fast
Thermal
[°C]
530-550
550-800
490-850
640-1000
280-510
565-850
Closed
Closed
Closed
Once-Through
Once/Closed
Closed
[bar]
1
1
90
70
250
<5
15
The Roadmap Addresses Viability and
Performance R&D Phases
 Viability
 Key
K ffeasibility
ibilit and
d proof-of-principle
f f i i l d
decisions
i i
 Performance
 Engineering-scale demonstration and optimization to desired levels
of performance
 Demonstration
 MidMid to large-scale
large scale system demonstration
 [Commercialization]
16
8
VHTR - Technology gaps
Process-specific R&D gaps exist to adapt the chemical process and the nuclear
heat source Qualification of high-temperature
high temperature alloys and coatings
coatings.
Producing hydrogen using the I-S process
Performance issues include
development of a highperformance helium turbine.
Modularization of the reactor
and heat utilization systems is
another challenge for
commercial deployment of
the VHTR.
17
Concept of a modular HTR (1/3)
Fuel element and restraining the fission products up to
1600°C Brennelement
Kr - Freisetzung
10 -1
SiC
Beschichtung
85
Graphit
10
0
10000 bis
20000 coated particles
Graphit
SiC
10 -7
1300 1500 1700 1900 2100 2300 2500
Temperatur / °C
-2
10 -4
Sr 90
Cs 137
Kr 85
10 -5
10 -6
10 -7
0
100
200
300
400
500
Heizdauer bei 1600 °C / h
Graphit
10 -5
10 -3
Freisetzungsanteil
Kern
10 -3
18
9
 Thermal power 300MW
 Circular core
 Pre stressed primary loop as an
burst prove inclusion
 Interior concrete cell with a
limitation on air amount
 Underground arrangement of the
reactor building
 Use of the primary loop to steam
production, gas turbines and
process heat applications (with inbetween heat exchanger)
15
11
feedwater
12
5
steam
10
9
2
3
8
1
7
to the
filter
G
13
16
4
M
14
6
19
Underground arrangement of the reactor
18
20
15
0
0
19
12
11
5
10
Speisewasser
9
3
Frischdampf
p
8
2
7
17
1
G
M
16
13
4
14
6
20
10
Accident behaviour of a modular HTR (1/10)
Accident assumptions
internal
Causes
external
Causes
external
causes
(beyond
licensing)
• Crash of a plane
(phantom)
• Terrorist attacks with
planes (Booing 747)
• Gas cloud explosion
• Sabotage
• Massive water ingress into the
primary system
• Earthquake (b < 0,3
g)
• Impacts
pacts of
o war
a (missiles)
( ss es)
• Massive air ingress into the primary
system
• Fire
• Full loss of coolant
• Full loss of the active residual heat
removal
• Extreme reactivity disturbances
• Tornados, hurricanes
and floods
• Extreme earthquakes
(b > 0,3 g)
• Meteorites
• Massive damage of reactor
components
21
Stabilitätsprinzip
i i
nichtschmelzbare Brennelemente und
C
Corestrukturen
t kt
Wirkung
nukleare
Stabilität
selbsttätige Begrenzung von nuklearer
Leistung und Brennstofftemperaturen
thermische
Stabilität
selbsttätige Nachwärmeabfuhr aus
dem Reaktorsystem
chemische
Stabilität
selbsttätiger Schutz
gegen Korrosion
mechanische
Stabilität
Requirements for a non melting
inherent safe modular HTR
 Fulfilment of the principles for all
accidents due to internal and
foreseeable external cause
(requirements from licensing)
 Fulfilment of the principles for all
accidents including
unforeseeable occurrences such
as terrorist attacks or extreme
earthquakes
selbsttätiger Schutz
gegen mechanische
Corezerstörung
11
Automatic removal of the
residual heat in the case of
failing of all active heat removal
(for example: PBMR, additionally
the total failure of the first shut
down system was assumed)
max
T Brennst.
(°C)
1600
1400
1200
1000
800
max
600
TBrennst.
400
TBrennst.
200
0
Zeit (h)
0
20
40
60
80
120
100
 Fuel elements can never melt
 Maximum fuel temperature
remains below 1600°C
 Most fuel elements stay far
below this temperature
 Total fission product release
stays below 10-5 of the inventory
Biological analogue for the automatic residual heat removal and
p
limitation of accident temperature
Tmax < 1600 °C
QN
Tw
Tmax < 38 °C
< 400 °C
QN  a  A ( Tw - Tu)
Q
Tmax - To  QN /H
4c
Tmax
QN  600 kW
To
Tw
Tu
< 37 °C
Tmax - To  Q/H
4 i
Tmax
(entspricht der Nachwärme
eines Cores mit Pth = 300 MW
nach etwa 30 h )
To
Q  a  A ( To - Tu)
To
Tu
Q  60 W
(entspricht der Wärmeabgabe
eines Menschen unter
“normalen” Bedingungen)
24
12
New safety requirements for
nuclear technology
105 t
K (t)
4
1,1 10 t
zukünftige
Annahmen
Flammentemperatur
(°C)
max. 1200 °C
zukünftige
Annahmen
heutige
Annahmen
h ti
heutige
Annahmen
800 °C
t
0
Schutz der Gebäude gegen
Penetration und Zerstörung
30 min
2h
t
Schutz gegen langandauernde
Flammeneinwirkungen mit hohen
Temperaturen bei Kerosinbrand
25
Reaktor
Bauschutt
(mit Isolationswirkung)
T (°C)
1500
Core (maximum)
1000
Reaktordruckbehälter
500
0
Behälter zu 60% isoliert
Behälter zu 99% isoliert
0
20 40 60 80 100 120 140 160 180 200 220 240
Zeit (h)
Automatic removal of the residual
heat after the destruction of the
reactor
t b
building
ildi
(for
(f example
l
caused by an extreme earthquake
or a plane crash caused by
terrorists)
 Reactor totally covered with
rumble
 Residual
es dua heat
eat is
s st
still removed
e o ed
automatically, but slowed down
 Maximum fuel temperature stays
below 1600°C
 Total fission product release
stays below 10-5 of the inventory
26
13
P/PO 6000
(%)
Automatic limitation of nuclear
power and fuel temperature
when faced with extreme
reactivity transient (total loss of
the first shut down system: ρ =
1,2 % in a short period of time)
0.1 s
1.0 s
4000
Leistung des
Reaktors
10.0 s
2000
t/s
0
0
10
20
 Strongly negative temperature
coefficient ends the excursion
 Transient
T
i t heat
h t is
i quickly
i kl and
d ffor
the most part stored in the fuel
element graphite
 Maximum fuel temperature stays
below 1600°C
TB 1300
(°C)
1200
1100
0.1 s
1000
1.0 s
10.0 s
900
Brennelementtemperatur
800
700
t/s
10
0
20
27
Concepts for minimising the
consequences of an air ingress
accident:
acc
de t
Vluft  5000m3
Granulat
6
8

1

2
4
5
3
7
1)
Internal concrete cell (sealed against outside air)
2)
Reactor building
3)
Draining channel
4)
Sealing flaps (gravity)
5)
Sealing plug
6)
Granulate silo
7)
Filter
8)
Flue


Draining the primary Helium
through pipes over a filter
Automatic limitation on the amount
of air within the concrete cell (V <
5000 m3) by automatic closing
mechanisms (flaps and pouring
granulate)
Limitation of possible graphite
corrosion to < 500 kg C and
therefore no radiological
consequences
Limitation of possible graphite
corrosion to < 100 kg by simple
intervention methods (protection of
investment)
28
14
Verhinderung der
K
Korrosion
i
von Brennelementen
Begrenzung
der Mengen
Mengenrate an
Luft
Begrenzung
g
g
der Menge
an Luft
 Volumen der inneren
Betonzelle ist begrenzt
(<5000m³)
 Vorgespannter Reaktordruckbehälter mit
kleinen Öffnungen
 Korrosionsbeständige
SiC-beschichtete
Brennelemente
 nach
h erfolgter
f l t D
Druckk
entlastung schließt eine
Klappe oder ein Granulatvorrat den Kanal ab
 Berst-Schutz für
die Behälter
 innere
i
B
Betonzelle
t
ll
ist mit Inergas
gefüllt
 Intervention in
innerer Betonzelle
 innere Betonzelle wird
mit Inertgas gefüllt
29
Comparison of the fuel behaviour after loss of the active residual heat removal.
PWR
5
~
HTR
Brennstofftemperatur
TB
(°C)
Tmelt =
2850 °C
10 Ci
Brennstab
~<
ZrCanning
0
S
TB
Brennstofftemperatur
<1600 °C
UO2-Kern
TC ~
600°C
UO2Tablette
1Ci
Partikel
C
1h
SiC
t
Schäden an den
Brennelementen
30 h
t
C
C Matrix
S
Schäden an den
Brennelementen
1
-5
10
1h
t
30 h
t
30
15
Total Assessment of Safety (1/2)
Integral proof of safety behaviour 
elektrische
l kt i h
Widerstandsheizung
Simulation of the residual heat in a
graphite-sphere-pile by electrical
resistance
i t
h
heating.
ti
((~ 2 MW ffor a 300
MWth - Core)
 Analysis of all assumable accidents
(total loss of coolant, air ingress,
water ingress, pressure discharge,
mechanical impacts on the elements
of the shut down system, component
damage when pressure relief)
f)
 Validation of all calculating programs
for extreme accidents (heat transfer
throughout all structures, flow
phenomena, acts of pressure
discharge).
31
Total Assessment of Safety (2/2)







Fuel elements can never melt; there is no heating up of the fuel elements to a
temperature above 1600°C
The principal of automatic residual heat removal can never fail, even after the
destruction of the reactor building it stays intact.
The reactor would even stay resistant against extreme reactivity transient.
There is no temperature raise of the fuel elements to above 1600°C
The pre-stressed reactor containment can not burst; It is impossible that an
unacceptable amount of air enters the primary loop, graphite corrosion is
minimal.
The ingress of larger amounts of water into the primary loop will not lead to
unacceptable states of the reactor; in the case of gas turbines operating this
accident is not applicable.
Against extreme, in future even more severe external impacts, the
underground way of building with covering mound, and fast removal of
spherical fuel elements as protection
There is no accident in which case a significant amount of radiation is released
32
16

Safety of Nuclear Power Plants Generation I to Generation IV

Transcrição

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