An designers overview of nuclear energy
at the start of the 21st century
Tony Roulstone
March 2010
Summary
•
After 25 years of retrenchment, nuclear power is firmly on the agenda, both in UK and
around the world – driven by the issues of:
–
–
Climate Change;
Energy Security.
•
•
UK will replace (at least twice over) the current ~10GWe nuclear capacity;
Global potential for >1000 new large reactors in next 30 years
– replacing the current 400 reactors (~350 GWe) growing the share of global
electricity from 15% to ~35%
•
Design topics covered:
1.
2.
3.
4.
Part 1
Thermal nuclear reactors design essentials;
Some design safety & reactor vessel considerations;
Extending the fuel resource available for nuclear fission power in thermal systems;
Longer term development of Advanced Systems;
Part 2
Reactor Design
Part 1
Design Safety
Fuel Resource
Part 2
Advanced Systems
Issues for the 21st century?
•
•
•
•
Response to World Credit Crunch;
Climate Change;
Nuclear Proliferation;
International terrorism.
Gordon Brown Mansion House Nov 2009
Nuclear is (for a good or ill) linked to at least 3 of these issues:
1. Credit crunch –> UK over reliance on financial services – new manufacturing?
2. Climate Change -> Expanding and de-carbonising electricity supply;
3. Nuclear Proliferation -> New fuel cycles that avoid creating or protect potential
nuclear bomb materials.
Civil Nuclear Power Global Market
Current capacity:
• Nuclear energy currently provides approximately 15% of the world’s electricity.
• Currently around 440 nuclear plants, across 30 countries, with a total capacity of over 370 GW.
Future Capacity:
• There may be a global build rate of up to 12 nuclear reactors per year between 2007-2030, which
expected to rise to 23-54 reactors a year between 2030-2050.
Market value:
• A recent assessment by Rolls-Royce estimated that:
• Global civil nuclear market is currently worth around £30bn a year;
• By 2023 market could be worth around £50bn per year;
• Of this, approximately £20bn pa will be new build, £13bn pa in support to existing nuclear
plant, and £17bn pa in support of new build reactors.
The Road to 2010 Cabinet Office July 2009
Westinghouse
AP1000
Part 1 – Current Reactors
Some Reactor Design Considerations
Reactor Design
5
Simplified Reactor Physics - Fission
Thermal reactor fission:
Uranium 235 is the only
naturally occurring fissile
atom - typical fission reaction
n
+ U235

thermal
Xe140
+
Sr94
+ ~2n
+
193Mev
fast
Fission Product distribution
High thermal n
absorption
low fast n
absorption
Intermediate energy
resonances linked to
quantum states - loss of n
Uranium 235 fission cross section
6
Simplified Reactor Physics – Criticality
(4 factor equation)
In an infinite homogenous reactor :
1. Fission of Uranium produces more neutrons than are absorbed by fuel as
thermal neutrons – η
& 200 MeV of energy
2. Fission neutrons augmented by some fast fission - ε
3. As neutron slowed down, some are lost to process by capture in
fuel resonance (1 - p) & others by moderator absorption (1 - f);
4. Surviving thermal neutrons lead to new fission – less than 1 sub-critical, more than 1 supercritical.
Infinite multiplication factor
Fast neutron energies
1 MeV
Intermediate energies
Thermal neutrons energies
~0.1eV
k∞ = ε * η * p * f
η = #fission/#absorbed
All these are factor are functions
of cross-sections - hence the
probabilities are all functions of
neutron energy
ε* η
where ε = fast fission multiplier
ε * η * (1 – p)
p - resonance capture
in fuel
ε* η*p*f
ε* η *p
ε* η * p (1-f)
f - moderator absorption
7
The First Nuclear Reactor
•
•
•
•
•
•
Enrico Fermi and a team from Metallurgy
Department of University of Chicago built and
controlled the first sustained nuclear reaction;
In a racquets court in Staggs Field athletics stadium
of the University of Chicago on 2 Dec 1942;
Reactor constructed from Graphite blocks and
Uranium metal constructed in ‘pile’ in 30’ * 60’ room
with Cadmium coated control rods;
The team proved that sustained fission or a
multiplication factor k >1 could be achieved ,
and the measurements were made for the first
practical sustained nuclear reactor;
Three types of control rods:
– Electric motor operated controlling rods;
– Emergency ‘zip’ rod driven in by gravity;
– Liquid Cadium salts to be released into a
control tube.
Sustained fission demonstrated by neutron count
growing exponentially i.e. keff >1
CP1 – Chicago Pile 1
8
Simplified Reactor Physics – Designs
Thermal reactor fission:
Infinite multiplication factor
k∞ = ε * η * p * f
(homogenous infinite reactor)
• Most (80%) of the 193 MeV (3.1 * 10-11 J/fission) of energy is released in the fission products as kinetic
energy –dissipated within a few microns as heat - fission products interacting with the surrounding material;
• 8% as beta and gamma radiation from the decay of fission products – decaying to more stable isotopes;
• Other energy in neutrons, prompt gamma rays and neutrinos.
Practical systems: - different effects of: enrichment, neutron leakage/reflection, moderator/fuel volume ratio,
structural materials, control elements, burnable poisons & re-fuelling strategy etc.
Fast fission
Fissions per absorption
Fuel resonance capture
Moderator capture
ε
ηT
1-p
1-f
Infinite multiplication
k∞
Thermal diffusion length
Core Power rating
Enrichment
PWR (light water - H20)
1.27
1.89
0.37
0.06
1.41
2.8 cm
100MW/m3
4.2%
CANDU (heavy water - D20)
1.0
1.31
0.16
0.03
1.12
1.7m
8MW/m3
0.7% (natural U)
9
Light Water Reactors are Dominant
Pressurised Water & Boiling Water Reactors
PWR
• Derived from submarine propulsion
reactors & widely installed around the
world ~ half of world capacity;
• Low thermal efficiency ~33%;
• Major problem was Three Mile Island in
1981 where minor fault led to confusing
signals & operators damaged reactor;
• Initial materials problems led to low
reliability - since rectified
• Now preferred in EU, Russia & China,
sharing market in US with BWR
BWR
• Simpler plant with integrated core cooling
and power cycle, high radiation dose from
operation;
• Core and steam separation integrated in
one vessel;
• Activated Nitrogen16 limits access to
turbine during operation;
• Some doubts about safety containment;
• More complex coolant chemistry;
• Popular in US, Sweden and Japan.
10
Pressurised Water Reactor
PWR overview:
• Operating conditions :
• Pressure:
16 MPa
• Average temperature: 280-290oC
• Reactor core consist of bundles of enriched
~3% Uranium Oxide or (MoX) in open bundles
with Zircalloy fuel clad tubes;
• Vertical control rods operated from the top of
the reactors;
• Multiple loops (3 or 4) carrying sub-cooled
water, flowing upwards through core;
• Refuelling at 3 years intervals from top of core
by means of removable vessel head;
• Complex coolant injection and decay heat
removal systems – issue addressed by latest
Westinghouse design AP1000;
• Able to separate chemistry strategies of
primary and secondary/condenser water;
• Most popular reactor design ~50% of installed
capacity: US, France, Germany, Spain & now
Russia & China – high availability, long fuel
cycles & low operator dose.
.
Boiling Water Reactor
BWR overview
• Operating conditions:
• Pressure (saturated)
~7.3 MPa
• Average temperature
310oC
• Reactor core consist of bundles of enriched ~3%
Uranium Oxide or (MoX) in shrouded bundles with
Zircalloy fuel clad tubes;
• Vertical control rods operated from bottom of core – do
not drop into core to shut-down reactor;
• Reactor cooling flow by augmented natural convection;
• Steam separators above the core with direct feed to wet
steam turbines – some carry over of contamination;
• Complex power control and multi-level emergency core
cooling systems;
• Refuelling at 2 year intervals from top of core by means of
removable vessel head;
• Coolant chemistry has to cover contamination from
condenser water as well as core requirements – not
preferred for coast sites because of potential chloride
contamination?
• Defence against major accident must take account of
reactor & turbine building – containment, aircraft crash
etc.
12
CANDU Heavy Water Reactor
• Canadian design - using natural uranium metal and pressurised heavy water in
horizontal Zircalloy pressure tubes, surrounded by heavy water moderator/reflector Core & Moderator tank
tank;
• Operating Conditions: Temp:
280oC
Press:
10MPa
• Large dimension core with low power density 8MW/m2;
• Conventional wet steam secondary power cycle with low thermal efficiency – 33%;
• Regular on-load refuelling using built-in horizontal charge machine;
• Complex D2O handling & Tritium is produced & emitted;
• Large evolutionary program in Canada with exports based on good operating
performance in 1980s : China, Korea, Romania, Argentina, India and Pakistan etc.
• More recently, beset with Zircalloy pressure tube delayed hydrogen cracking
problems, requiring wholesale replacement of pressure tubes;
Candu 6 -Fuel
subassemblies
13
Some Design & Safety Considerations
Design Safety
14
Nuclear Safety Philosophy
Developed world safety approach:
• Design base events which are always
protected with a significant margin of safety;
• Large scale events rendered very unlikely by:
o providing defence in depth/multiple
barriers to release;
o including features and best practise in
the design to extent that is ALARP.
Frequency
pa
Complete Protection with
high degree of certainty
10-4
Chernobyl ~2 days after the prompt
criticality followed by a core fire in 1986
Fundamental faults – design: positive void
coefficient & control rod design - plus ,very
poor operating knowledge & controls.
Larger releases - by
design made most
unlikely
Area of
acceptance of
design
10-7
ALARP – best practice continually
challenged
Probabilistic
safety target
Release
Chernobyl – Criticality Accident
•
•
•
•
•
•
•
•
RBMK reactor in the Ukraine on 26 April 1986 at 0123 hrs was
undergoing an experiment – to investigate low power operation;
Automatic shut-down system was switched off because of difficulties
operating at the low power levels required by the test and most of the
control rods withdrawn - all but 6 (safe level ~30);
Experiment led to reduced cooling & boiling in the channels increasing
core reactivity, which was made worse by the shut-down system adding
further reactivity, leading to a very large increase in core power *100 ;
Fuel pellets exploded, damaging cooling the fuel channels – escaping
steam blew off top of reactor – followed by a second explosion from
RBMK vertical Fuel & Cooling
hydrogen produced from either Zr-water or Graphite water reactions;
channels Graphite moderator
Much of core & graphite moderator was ejected and graphite caught fire
continuing the release of radioactivity, which was spread by the weather patterns across Ukraine,
Russia & W Europe;
About 40 operators and fire fighters died within 3 months as a result of exposure to radiation as
they put out the fire and brought the situation under control (plus further ~200 later from effects);
17 mile exclusion zone around the plant with 10,000 people displaced;
Broader concerns about health effects across Europe - but the measured effects is an increased
level of ~4000 cases of treatable thyroid cancer in children;.
Decay Heat & Energy
Decay Heat & Energy
7.00%
Sec
Min
Hour
Day
7000
Week
Month
6.00%
5.00%
6000
P/Po
G/Po
4.00%
5000
4000
P/Po
3.00%
3000
2.00%
2000
1.00%
1000
0.00%
G/Po
MJ/MW
0
Decay Heat (Beta & Gamma):
P(t) = 0.0622P0(t-0.2 – (t+t0)-0.2)
Integrated stored energy
G(t) = 0.0622P0 * 1/0.8 (t0.8 - (t+t0)0.8 + t00.8 )
For: P0 = 3000MW
t0 = 2 years = 2*365*24*3600 = 6.3 * 107 EOL
P (1day) = 0.0046 *3000 ~ 14 MW
- equivalent to Latent Ht ~7 kg/s of water
G (1day) = 534 *3000 ~ 1600 GJ
- equiv to LH of 800m3 of water RPV ~500m3
- energy required to melt large civil core ~200GJ !!
17
Three Mile Island – Loss of Coolant Accident
•
•
•
Loss of cooling accident at Three Mile Island, Pennsylvania on Wed 29 March 1979 which was:
– Exacerbated by operator error and bad human factors design,
– Led to partial melt-down of the core, and
– Gaseous fission product release from the station, and
– Mass evacuation of surrounding area.
Steps:
1. Equipment failure in the steam cycle plant, led to feed-water
s hut-of
loss and automatic reactor shut-down;
2. Back-up feed system was down for maintenance;
3. Decay heat build-up raised primary pressure & relief valve operated;
4. Relief valve did not shut , was not recognised & 120m3 of primary
coolant discharge in 3 hours;
5. Operator did not recognise symptoms, stopped injecting water,
switched off the coolant pumps allowing, core coolant to boil
displacing water into the Pressuriser and uncovering the core;
6. Much of the core melted within about 4 hours , releasing fission
products – 13 million curies of noble gases (Xenon & Kryton) plus
~17 curies of Iodine 131 to the circuit - via the relief valve outside the reactor.
Reactor has been disassembled with the damaged core and the vessel removed – but accident
changed the whole approach to the design and operation of nuclear reactors.
18
Nuclear Safety approach affects the design
Events & Hazards
Plant Faults
External Haz
Opened ended process for
identification of potential hazards
Internal Haz
Normal Operation
Analyse event & accident sequences with frequencies
Consider primary & secondary means of protection
---> Design basis of structures, containment & safety systems, including human factors
Probabilistic Risk Analysis
Design Basis Analysis
Low frequency < 10-4 pa
High frequency >10-4 pa
/high consequence
Demonstrate with high degree
of certainty
Nuclear Safety - some Vessel Design Cases
Design Basis Analysis
High frequency
>10-4
pa
Demonstrate protection with
a high degree of certainty
Probabilistic Risk Analysis
Low frequency < 10-4 pa
/high consequence
Normal Operation:
Cases include:
• Warm-up & Cool down to operating
conditions
• Power transients:
• Normal
• Power & pressure overshoot
• Scram shut-down/cool-down of reactor
• Brittle fracture;
• Low cycle fatigue;
• Max pressure &
protection;
• Cold shock.
Accident Sequences:
Cases include:
• Brittle fast fracture of vessel due to inbuilt defects
• Response to loss of coolant accidents
• Vessel cracking due to in-built defects
• Neutron irradiation
damage;
• Cold shock;
• Defect identification and
eradication by design &
construction.
Design Optimisation Process for Components
Design
• Operation & Performance
requirements
• Configuration & sizing
Materials & Man’f
• Material properties & composition
• Manufacture processes &
Assurance of integrity
Iterative process
1.
2.
3.
Concept
System design
Component design
Safety
• Event sequences
• Degraded modes examined
• Control & protection systems
Stressing
• Modelling of design & loads
• Stress intensity factors &
critical defect sizing
21
Some Pressure Vessel Design Considerations
•
•
Reactor Vessel (RPV) contains the core - fissile & heat generating material:
– Design topics:
• Retain primary circuit pressure up to ~180 bar & temperatures up to 340oC ;
• Able to operate at sub-cooled down to ambient temperatures ~ 20oC;
• Exposed to high level of energetic neutrons from fissile reaction s;
• Maintain core covered with water from the day when it is loaded until/including when
removed after 3 years;
• RPV integrity is absolutely fundamental to plant safety that RPV cracking or other failure
has be ‘incredibly’ unlikely – < 1 in million years - nothing that is feasible of being done to
ensure integrity will not be done;
Issues to be addressed:
– Pressure requirements lead to choice of ferritic steels – low temperature brittle fracture?
– Neutrons embrittle this type of steel – choice of alloy & control of trace elements 7
microstructure;
– Thermal and pressure cycles drive crack growth;
– RPVs constructed from welded plates – absence of inclusions/defect & built-in stresses;
– Measures to ensure that failure is rendered incredible??
Design Safety
PWR Vessel & Materials topics
Control rod motor
support tube –
dis-similar tube to
head welds
Fracture of
neutron embrittled
Reactor Vessel
Issues include: 60 year plant life;
assurance of safety margins; manufacture
& inspection standards, effectiveness of
enhanced material testing.
Vessel nozzle welds
– low cycle fatigue
Idealised Pressure Vessel Design
Purpose:
• Contain pressure to control
cooling at power – sub-cooled
for PWR & saturated for BWR
• Maintain core covered with
water - post shutdown, in all
conditions, including accidents –
decay heat ~1% of full power for
long periods of time.
Core
Neutron
Flux
Profile
Design objectives;
• Lower stress
• Strong material
• High toughness
• Low brittle transition
Two topics:
• Assurance of zero defects
1. Assurance of vessel Integrity
through-life
• Long life ~60 years
2. Ductility & Operating limits
Pressure Vessel Ductility
Irradiation hardening sets plant operating limits
Stress
Intensity
factor
Op Pt
*
290C 15MPa
ASME
Lower bound
reference
temp-toughness
curve
Saturated
water line
ΔT
Temp
Embrittlement processes:
• Degree of embrittlement from irradiation is very dependant on trace elements Ni, Cu & P
• Generation of lattice defects in displacement cascades by high-energy recoil atoms from
neutron scattering;
• Diffusion of primary defects also leading to enhanced solute diffusion and formation of
nano-scale defect-solute cluster complexes, solute clusters, and distinct phases,
primarily copper-rich precipitates (CRPs)
• Dislocation pinning and hardening by these nano-features
• Hardening-induced DTt shifts
Embrittlement of nuclear reactor vessels - Odette &
Lucas JOM 2001
ASME Lower bound reference
toughness-temp curve shift ΔT
Pressure Vessel Design - Integrity
Neutron Embrittlement
• Water is a good moderator -> large
RPV diameter
• Steel is a good reflector
stressed thermal shields
• Vessel made from low alloy high
strength steel of controlled
composition: A533B plates, or
A508 forgings, quenched and
tempered - C(0.05–0.2%), Mn(0.7–
Nozzle
Zone
1.6%), Mo(0.4–0.6%), Ni(0.2– 1.4%),
Si(0.2–0.6%), & Cr (0.05–0.5%).
Design objectives;
• Lower stress
• Strong material
Core
• High toughness
• Low brittle transition
• Assurance of zero defects
• Long life ~60 years
Weld
Zone
–> un-
• Construct from forgings with
minimum welds – avoiding regions
Neutron
of high neutron flux;
Flux
Profile
• Vessels are tempered & stress
relieved, typically at about
620±15°C for about 30 h, resulting
in as-fabricated yield stress values
of ~ 475±50 MPa.
Incredibility of Failure - Vessel Design
Because core safety depends on keeping fuel covered with water, its failure must be ‘incredible’
Three broad strategies – each pursued fully – a practical example of ALARP:
1. Zero defects in manufacture;
– Qualified processes completed by skilled & qualified welders & technicians;
– Multiple & varied inspection of welds by manufacturer;
– Investigation & repair of all detectable defects;
– Independent inspection of welds by qualified third party.
2. Examination of vessel through-out life to ensure that defects are not present/or not
growing:
– Pre-service inspection of vessel provides the base-line;
– Inspection of critical parts of vessel during re-fuelling outages
3. Vessel is tolerant of any defect that might be present.
– Stresses will below critical crack growth levels for postulated cracks based on the
demonstrated inspection capability;
– Evaluate broadest set of operating transients to determine crack growth low cycle
fatigue;
– Take account of irradiation hardening & thermal aging
1. Manufacture for zero defects
Ability to Detect & Repair Weld Defects
Small Voids in Base metal – machine weld
Identified by ultrasonic testing
From: Transactions, SMiRT 19, Toronto, August 2007
Complex Repair Flaw - Lack of Fusion
Collection of small flaws in a weld repair
that make up a larger defect is not fixed.
28
2. NDE Verification – Sizewell B
Inspection verification of manufacturers & pre-service
inspection contractors by AEA Technology
Multiple inspection agencies:
• Creusot- Loire - Forgings;
• Framatome – Main welds
• Babcock Energy – Supplier - Machining & Outfitting
• Rolls-Royce – Pre-service inspection
• OIS – Studs & Nuts
Shows:
• Value of certifying inspection agencies, &
• Inspectors can correctly & repeatedly
identify and size defects.
3. Detailed Analysis all Severe Accidents
Thermal Stressing Analysis LOCA – an example
Small break LOCA
Temperature profile
Konvoi 3 loop PWR
Tangential stress profile
inlet nozzle analysis
Analysis to show that stresses when applied
to critical defect size, does not lead to
catastrophic crack growth or vessel fracture.
30
End of Part 1
www.bracchium.net
31