RBMK. FBR. Other. N um ber of U nits or C ontribution [%. ] Number of Reactors Regarding Type and Energy Contribution. Number of Units. Energy Contribution ...
Nuclear Reactors and Nuclear Power Plants (NPP) Aleksandar Delja Directorate of Assessment and Analysis Reactor Thermalhydraulics Division
Carleton University January 10, 2014 , Ottawa
Outline 1 1. 2 2. 3 3. 4 4. 5 5. 6. 6 7. 7 8. 8
Population and Energy Reserves Chronology of Development Nuclear Fission and NPP Heat Generation, Thermalhydraulics Nuclear Power Plant (NPP) Concept PHWR(CANDU), PWR and BWR Fuel, Secondary Side, Steam Generators and BoP Conclusions
Canadian Nuclear Safety Commission
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1e+010 9e+009 8e+009 6e+009
Photo: Smithsonian , Washington DC, November 2012 (A.Delja)
World Population
7e+009 5e+009 4e+009 3e+009 2e+009 1e+009
7e+009
0 0
500 1000 1500 2000 2500 Year
6e+009 5e+009
1e+010
4e+009 3e+009 2e+009 1e+009 1800
1850
1900
1950
Year
2000
1e+009 1800
1850
1900
1950
2000
Year
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Energy Reserves The IAEA estimates the remaining uranium resources to be equal to 2500 ZJ.[1]. This assumes the use of breeder reactors, which are able to create more fissile material than they consume. For onethrough fuel cycle uranium reserves are 2 ZJ. Thorium reserve and cycle options should be added to the above number.
Energy Non-Renewable Reserve 45
S1_BP S2_DERA 40 S3_BP_IAEA 35 Energy [ZJ]
30 25 20 15 10 5 0 Coal
Oil
Gas
Uranium
Total [1]
"Global Uranium Resources to Meet Projected Demand: Latest
Edition of "Red Book" Predicts Consistent Supply Up to 2025". International Atomic Energy Agency. 2 June 2006.
Sources: S1: BP Statistical Review of World Energy (June 1990) S2: Energy Study 2012, Reserves, Resources and Availability, DERA -The Federal Institute for Geosciences and Natural Resources (BGR) on behalf of the German Mineral Resources Agency (DERA) S3: BP Statistical Review of World Energy 2010, for Uranium IAEA,June 2006
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zetta [Z] = 10007=1021
4
Population without electricity: 1.32 billon (2009)
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Source: International Energy Agency
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Brief Chronology of Nuclear Power Plant Experimental Development PKL PHEBUS
LOFT
ATLAS
2D/3D UPTF
CRL PSB-VVER LOBI UPTF UN Atoms for Peace
RD-14M
R-9 (AECB, ECCS)
First NPP APS-1 Obninsk
Until this time ECC was not required
RG 1.157
10 CFR 50.46
CSAU
Industry Started Code Uncertainty Now
1950
1955
1960
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1965
1970
1975
1980 TMI
1985
1990
Chernobyl
1995
2000
2005
2010
Fukushima
2015 6
CANDU Evolution Wolsong (Korea) 1,2,3,4(1982,87,98,99)
Point Lepreau 1971-73 Pick A 4x542 MW Pick B 4x540 MW
NRU 1957 200MW
NPD 1962 CANDU 24 MW
COG
Embalse (Argentina) Cernavoda 1 – Romania (1996) Cernavoda 2 (2002)
Douglas Point 1966 CANDU 220 MW
Quinshan 1,2 – China (2003)
G-2
1990-93 Darlington 4x935 MW
NRX 1957 42 MW 1977-78 Bruce A 4x900 MW Bruce B 4x915 MW ZEEP (10MW) 1945
New Design CANDU ACR700 EC CANDU 3 ACR1000 9
First NPP APS-1 Obninsk
Now
1950
1955
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1960
1965
1970
1975
1980 TMI
1985
1990
Chernobyl
1995
2000
2005
2010
Fukushima
2015 7
Canadian Nuclear Safety Commission
Established May 2000, under the Nuclear Safety and Control Act Replaced the AECB of the 1946 Atomic Energy Control Act
Canada’s independent nuclear regulator 68 years of experience
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Nuclear Fission Fission energy: ~ 200 MeV = 3.2 10-11 J (1 eV = 1.602 10-19 J)
Multiplication constant (factor) k: k = number of neutrons in one generation/number of neutrons in previous generation k = 1 , reactor critical, k reactor supercritical Canadian Nuclear Safety Commission
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Fission and Secondary Fission Fuels Uranium 235 Fission
U+ n →
235 92
1 0
(
)
U → Ba + Kr + 3 n
236 92
141 92
92 36
1 0
Plutonium 239 production
U+ n →
238
1
β−
239
U
239
U
→
β−
239
23.5 min
Np →+
239
Pu
+
233
2.3 d
Thorium – Uranium 233 production 232
Th + 1 n →
233
Th
233
β−
Th
→
23 . 5 mun
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β−
233
Pa
→
U
27 . 4 . d
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Nuclear Energy Utilization
Heat Source
Heat Transport
Work
Electricity
Heat
Heat Sink
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Environment
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Reactor Systems Fast Reactors
Thermal Reactors
Molten Salt
Water
CO2 H2O
He H2O H2O H2O D2O
Magnox
BLW
AGR RBMK HTGR
MSBR
Heavy Water
THTR
PWR
CO2
Sodium/ NaK
CANDU OCR Atucha
ACR 700/ 1000 BWR CANDU SGHW
COOLANT
Natural U
Enriched U
KKN
Thorium
LWBR
Fugen
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Hydro Carbon
MODERATOR
LMFBR
F U E L
Graphite
REACTOR NEUTRON ENERGY
Plutonium-U
12
Other Reactor Types
AGR - Advanced Gas Reactor
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Liquid Metal Fast Breeder Reactor
13
Other Reactor Types (cont.) RBMK- Reactor
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Heat Generation Neutron Flux:
z
Neutron Neutron Neutron ∂n = Source + Absorption + Escaped from ∂t Generation Interaction System
Φ = vn
R
r
H
⎛→⎞ ∂n⎜ r , t ⎟ ⎝ ⎠ ∂t
⎛→⎞ ∂ Φ ⎜ r, t ⎟ → 1 ⎞ ⎛→⎞ ⎛→⎞ ⎝ ⎠ 2 ⎛ = D∇ Φ⎜ r , t ⎟ − Σ a Φ⎜ r , t ⎟ + Q⎜ r , t ⎟ = v ∂t ⎝ ⎠ ⎝ ⎠ ⎝ ⎠
q′′′ = Φ fuel Σ fissionγ E fission Φ fuel − neutron flux in fuel Σ fission − macroscopic cross − sec tion for fission
γ − fraction of fission energy absorbed in fuel ⎡ neutrons ⎤ neutron flux ⎢ ⎥ 2 ⎣ m s ⎦ ⎡m⎤ v neutron velocity ⎢ ⎥ ⎣s⎦ ⎡ neutrons ⎤ n neutron density ⎢ 3 ⎥⎦ ⎣ m
Φ
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E fission − energy released in one fission Heat generation
( )
⎛ πz ⎞ ⎛ 2.405 r ⎞ q ''' r , z = q0''' J 0 ⎜ ⎟ cos ⎜ ⎟ ⎝ R ⎠ ⎝H⎠
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Q = w c p (t out − t in ) = w c p Δt
Core Thermalhydraulics Energy balance: PWR: BWR:
Q
=
w c p (t out − t in ) =
(
Q = w h out − hin
)
h – specifis enthalpy h1 – water saturation specific enthalpy h2 – steam saturated specific enthalpy w – mass flow rate Q – heat generated in time – power cp – specific heat t – temperature out x – steam quality A- heat transfer area K – heat transfer coeficient
w c p Δt
out
= wΔh ,
for BWR hout = h2 ( p )
Or general:
Q = w c p (t sat − tin ) + w x r , r = h2 − h1
− latent heat
Heat Transfer from fuel to coolant: Q = K A (ΔT )eff
in
out
in
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in
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Thermodynamic • Laws of thermodynamics • Fluid properties • Thermodynamic cycles L η= Q
Rankine Cycle
T
qR L − cycle work , Q − su pplied heat to cycle
3
4
2 2a
qC
1
5a
5
η = 1− T 0 T
s
η=
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Pturbine − Ppump qR
=
(h4 − h5 ) − (h2 − h1 ) (h4 − h2 )
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NPP Configurations Separator
Turbine
1
3 Pressurizer
Steam Generator
Generator
2
Condenser
4
Reactor
Circulating Pump
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Liquid-metal Pump
Feed Pump
1 2 3 4
-
Single circuit (BWR) Double circuit (CANDU, PWR) Incomplete double circuit Triple- circuit
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CANDU NPP
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Nuclear Power Plant Reviews and New Development
EPR
ACR-1000
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EC-6
ATMEA
AP1000
mPower
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World NPP Number of Reactors Regarding Type and Energy Contribution 300
Number of Units Energy Contribution [%]
Number of Units or Contribution [%]
250
200
150
100
50
0 PWR
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BWR
PHWR
GCR
RBMK
FBR
Other
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COMPARISON BWR, PWR and CANDU BWR
PWR
CANDU
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Pressurized Water Reactor (PWR)
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PWR Fuel
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BWR Reactor
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CANDU6
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CANDU HTS ACR1000
AP1000 (PWR)
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ACR – Advanced CANDU Reactor
27
CANDU - Fuel Channels, Calandria
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CANDU Fuel
Standard CANDU fuel design: No. of elements: 37 Element diameter: 12 mm UO2 weight: 22 kg Zr-4 weight: 2 kg Burnup: 7500 MWd/Mg Max. power: ~ 850 kW Canadian Nuclear Safety Commission
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CANDU Fuel (cont.)
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CANDU Fuel Core Configuration
D2O Primary Coolant
Gas Annulus Fuel Elements Pressure Tube Calandria Tube
Moderator
CANDU CANDU core core power power distribution distribution Power Power [-] [-]
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XX –– ch chaannnnel el ppoossiti itioonn
onn iittiio s s oo llpp e e nn aann h h cc YY––
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Steam Generators T2
Reactor
T2
Super heater
T1
Tsat(p)
Reactor
T2
Reactor
T1
T1 Tsat(p)
Boiler
Tsat(p)
Economizer Tfeed T2
T
Tfeed
T
T1
T
T2 T1
Tsat(p) Tsat(p)
ΔT
T2 T1
Tfeed
ΔT
Reactor
Tsat(p) Steam Generator
ΔT
Tfeed
Tfeed Environment
Q T
T
s Canadian Nuclear Safety Commission
Q
Q T
s
s T – temperature, Q – Heat Energy, s – specific entropy
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CANDU Steam Generators
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SG: Primary/Secondary Heat Transfer
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Balance of Plant (Secondary System)
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h
t=265 oC
Specific Enthalpy
p= 1M
Pa
Steam Reheater
p= 5
Moisture Separator
High Pressure Turbine
M Pa
Secondary Side Turbine Characteristics
re stu tor i o a M par e S
y1 = 87
t=220 oC
m er ea eat t S eh R
0. 0 p=
% y2 = 89
Specific Entropy
0
0.0 4-
0
5M
%
s
Low Pressure Turbines
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Pa
CANDU vs. PWR
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CANDU HTS
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PWR HTS
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PHWR vs. PWR PHWR
PWR
Natural Uranium
Enriched uranium
Simple fuel assembly
Standard full core length assembly
On-power fuelling
Fuelling during shut down
High Uranium fuel utilization
Standard uranium fuel utilization
Digital Control System from the beginning of CANDU
Analog control system, recently transition to digital
D2O (coolant and moderator) is expensive
H2O is coolant and moderator
Tritium must be controlled
Tritium low
Complex piping
Large pressure vessel
Positive coolant void reactivity
Negative coolant void reactivity
2 independent diverse shutdown systems
1 shutdown system
Large neutron lifetime
Short neutron lifetime
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Design Parameters of PHWR and PWR PHWR
PWR
Core power density
11 MW/m3
60 MW/m3
Maximum fuel rating
57.3 kW/m
42 kW/m
Neutron lifetime
9 E-4 sec
5 E-5 sec
Fuel burnup
7.5 MWd/kg
35 MWd/kg
Uranium usage
157 Mg/MWyr
213 Mg/MWyr
Operating Pressure (exit)
10 MPA
15.5 MPa
Core inlet temperature
266 oC
292 oC
Core exit temperature
312 oC
329 oC
Steam temperature
266 oC
283 oC
Heavy water inventory
.75 Mg/MWe
None
Plant life
40-60 years (PT replacement)
40-60 year (SG replacement)
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PWR vs. CANDU philosophy • Traditional PWR based on – 2 to 4 spatially separated identical trains – Little redundancy within each train – All trains fully qualified
• Traditional CANDU based on – Two diverse separated groups – Redundancy within each group – qualified according to safety functions
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CANDU6 - Heat Transport System (HTS) Parameters
ΔT = 44 oC w = 7.7 103 kg/s Canadian Nuclear Safety Commission
ΔT
43
PWR Operating Parameters ΔT = 37 oC w = 9.6 103 kg/s
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Source: A.Prosek, IJS Slovenia
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CANDU Fuel Cycles
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ACR-1000
Recent CANDU Design
Canadian Nuclear Safety Enhanced CANDU 6 Commission
ACR1000
46
CANDU Systems, Strictures and Components (SSC)
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Special Safety Systems Reactor Shutdown System 1 and 2 RSS1 and 2
•CONTROLE •COOL •CONTAIN
ECC – Emergency Core Cooling
Containment
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Research Reactors (Past and Future) NRU reactor (1957) Jules Horowitz Reactor
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Generation IV
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Conclusions • Nuclear energy will have important role in the future energy supply. • Currently, Light Water Reactors (LWR) are dominant NPP for electricity generation. • CANDU has significant advantage for different nuclear fuel cycles strategies. • Safety of nuclear installation will continue to be fundamental requirement for nuclear energy use in electricity production, industrial applications, research and development. Canadian Nuclear Safety Commission
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Thank you Canadian Nuclear Safety Commission
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