Download - An designers overview of nuclear energy at the start of the 21 st century Tony Roulstone March 2010
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An designers overview of nuclear energy at the start of the 21st century
Tony Roulstone
March 2010
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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. Thermal nuclear reactors design essentials;
2. Some design safety & reactor vessel considerations;
3. Extending the fuel resource available for nuclear fission power in thermal systems;
4. Longer term development of Advanced Systems;
Design SafetyDesign Safety Reactor DesignReactor Design Fuel ResourceFuel Resource Advanced SystemsAdvanced Systems
Part 1
Part 2
Part 1 Part 2
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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.
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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
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Part 1 – Current ReactorsSome Reactor Design Considerations
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Reactor DesignReactor Design
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Simplified Reactor Physics - Fission
Thermal reactor fission:
Uranium 235 is the only
naturally occurring fissile
atom - typical fission reaction
n + U235 Xe140 + Sr94 + ~2n + 193Mev
thermal fast
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High thermal n absorption
low fast n absorption
Intermediate energy resonances linked to
quantum states - loss of n
Fission Product distribution
Uranium 235 fission cross section
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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 k∞ = ε * η * p * f All these are factor are
functions of cross-sections - hence the probabilities are all functions of neutron energy
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Fast neutron energies1 MeV
η = #fission/#absorbed
ε * η where ε = fast fission multiplier
Intermediate energies ε * η * (1 – p)p - resonance capture
in fuel
Thermal neutrons energies~0.1eV
ε * η * p * f ε * η * p ε* η * p (1-f)
f - moderator absorption
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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
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CP1 – Chicago Pile 1
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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.
PWR (light water - H20) CANDU (heavy water - D20)
Fast fission ε 1.27 1.0
Fissions per absorption ηT 1.89 1.31
Fuel resonance capture 1 - p 0.37 0.16
Moderator capture 1 - f 0.06 0.03
Infinite multiplication k∞ 1.41 1.12
Thermal diffusion length 2.8 cm 1.7m
Core Power rating 100MW/m3 8MW/m3
Enrichment 4.2% 0.7% (natural U)
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Light Water Reactors are DominantPressurised Water & Boiling Water Reactors
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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.
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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.
.
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Boiling Water Reactor
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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.
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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 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;
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Core & Moderator tank
Candu 6 -Fuel subassemblies
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Some Design & Safety Considerations
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Design SafetyDesign Safety
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Nuclear Safety Philosophy
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.
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.
Frequencypa
Release
10-4
10-7
Area of acceptance of
design
Complete Protection with high degree of certainty
Larger releases - by design made most
unlikely
Probabilistic safety target
ALARP – best practice continually challenged
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Chernobyl – Criticality Accident
RBMK vertical Fuel & Cooling channels Graphite moderator
• 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 hydrogen produced from either Zr-water or Graphite water reactions;
• 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;.
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Decay Heat & Energy
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 !!
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0
1000
2000
3000
4000
5000
6000
7000
0.00%
1.00%
2.00%
3.00%
4.00%
5.00%
6.00%
7.00%
Sec Min Hour Day Week Month
G/PoMJ/MW
P/Po
Decay Heat & Energy
P/Po
G/Po
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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.
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Nuclear Safety approach affects the design
Events & Hazards
Plant Faults External Haz 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
Low frequency < 10-4 pa
/high consequence
Design Basis Analysis
High frequency >10-4 pa
Demonstrate with high degree
of certainty
Opened ended process for identification of potential hazards
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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:• Warm-up & Cool down to operating
conditions• Power transients:
• Normal• Power & pressure overshoot
• Scram shut-down/cool-down of reactor
Accident Sequences:• Brittle fast fracture of vessel due to in-
built defects• Response to loss of coolant accidents• Vessel cracking due to in-built defects
Cases include:• Brittle fracture;• Low cycle fatigue;• Max pressure &
protection;• Cold shock.
Cases include:• Neutron irradiation
damage;• Cold shock;• Defect identification and
eradication by design & construction.
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Design Optimisation Process for Components
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Design•Operation & Performance
requirements•Configuration & sizing
Safety•Event sequences
• Degraded modes examined•Control & protection systems
Materials & Man’f•Material properties & composition
•Manufacture processes & Assurance of integrity
Stressing•Modelling of design & loads
•Stress intensity factors & critical defect sizing
Iterative process1.Concept2.System design3.Component design
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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 SafetyDesign Safety
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PWR Vessel & Materials topics
Control rod motor support tube – dis-similar tube to head welds
Fracture of neutron embrittled Reactor Vessel
Vessel nozzle welds – low cycle fatigue
Issues include: 60 year plant life; assurance of safety margins; manufacture & inspection standards, effectiveness of enhanced material testing.
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Idealised Pressure Vessel Design
Core
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.
Design objectives;
•Lower stress
•Strong material
•High toughness
•Low brittle transition
•Assurance of zero defects
•Long life ~60 years
Neutron Flux
Profile
Two topics:
1. Assurance of vessel Integrity through-life
2. Ductility & Operating limits
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Pressure Vessel DuctilityIrradiation hardening sets plant operating limits
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
Temp
Saturated water line
ΔT
Op Pt*
290C 15MPaASMELower bound reference temp-toughness curve
Stress Intensity factor
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Pressure Vessel Design - Integrity
CoreNeutron
Flux Profile
Neutron Embrittlement
•Water is a good moderator -> large RPV diameter
•Steel is a good reflector –> un-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(0.7–1.6%), Mo(0.4–0.6%), Ni(0.2– 1.4%), Si(0.2–0.6%), & Cr (0.05–0.5%).
•Construct from forgings with minimum welds – avoiding regions of high neutron flux;
•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.
Design objectives;
•Lower stress
•Strong material
•High toughness
•Low brittle transition
•Assurance of zero defects
•Long life ~60 years
NozzleZone
WeldZone
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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
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1. Manufacture for zero defectsAbility to Detect & Repair Weld Defects
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Small Voids in Base metal – machine weldIdentified by ultrasonic testing
Complex Repair Flaw - Lack of FusionCollection of small flaws in a weld repair
that make up a larger defect is not fixed.
From: Transactions, SMiRT 19, Toronto, August 2007
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2. NDE Verification – Sizewell B
Inspection verification of manufacturers & pre-service inspection contractors by AEA TechnologyMultiple 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.
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3. Detailed Analysis all Severe AccidentsThermal Stressing Analysis LOCA – an example
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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.
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End of Part 1 www.bracchium.net
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