properties and behaviour of advanced nuclear fuels · 2016-06-03 · reprocessing plant u mining...
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ESOF 2010, Turin, 3rd July 2010 1
Properties and Behaviour of Advanced Nuclear FuelsJ. Somers
Joint Research Centre (JRC)
ITU - Institute for Transuranium ElementsKarlsruhe - Germany
http://itu.jrc.ec.europa.eu/http://www.jrc.ec.europa.eu/
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A European vision of nuclear energy development
First First ReactorsReactors
Current Current ReactorsReactors
Future Future SystemsSystems
Generation I
Generation II
1950 1970 1990 2010 2030 2050 2070 2090
Generation III
Dismantling& clean-up
Generation IV
Produces 31% of Europe’s electricity
New build in Finland and France (EPR), other countries… Start of industrial deployment in 2040-
2050
Advanced Advanced ReactorsReactors
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The Sustainable Nuclear Energy Technology Platform (SNE-TP)
• Research to improve sustainability of nuclear energy● Conservation / recycling of resources● minimisation of radioactive waste (volume, lifetime, radio-toxicity)● maintaining high level of safety and favourable economics
• Gather all EU stakeholders involved in the nuclear sector and structure and finance nuclear fission R&D
• Develop knowledge and competence through education and training linked with a network of research infrastructures
• Maintain Europe’s leadership in nuclear technology in the longer term by building demonstrators and prototypes for a future generation of reactors
• Develop international cooperation on a basis of mutual interest and benefit
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Strategic Research Agenda: 3 pillars
Over 180 researchers, scientists and engineers have contributed to the SRA
www.snetp.eu
Vision document
Strategic Research Agenda (SRA)
Deployment StrategySustainable FuelsMinimising Waste Issues
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The Nuclear Fuel Cycle
EnrichmentEnrichment
FuelFuelFabricationFabrication
Reactor
SNFSNFStorageStorage
ProcessingProcessing
High LevelHigh LevelWasteWaste
SpentSpentNuclearNuclear
FuelFuel
RepositoryRepository
UraniumUraniumStorageStorage
DepletedDepletedUraniumUranium
NaturalNaturalUraniumUranium
reprocessing plant U mining
nuclear reactor
spent fuel storage
repository
Fissile andFissile andFertileFertile((U,Pu,MAU,Pu,MA))
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Evolution of radiotoxicity as a function of time:Evolution of radiotoxicity as a function of time:evaluation by CEA, FZK and ITUevaluation by CEA, FZK and ITU
Waste Radiotoxicity EvolutionWaste Radiotoxicity Evolution
time (y)10 1 10 2 10 3 10 4 10 5 1010 2
10 3
10 4
10 5
10 6
10 7
10 8
10 9
with P&T
1000 y [99% Pu, 98% MA removal]
270 y
500 y [99.5% Pu, 99% MA removal]
130,000 y
results based on ICRP72
Inge
stio
n ra
diot
oxic
ity (S
v pe
r ton
spe
nt fu
el)
Totalactinides
fission products
ref. 7.83 t U in equilibriumwith P&T
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ActinideActinide--RecyclingRecycling--SchemesSchemes
Homogeneous recycling
R T
U
FP
U Pu MA
R T
U
MAU Pu
FPFP&MA
Heterogeneous MArecycling
R T
U
UU & Pu recycling,
PUREX
Pu
R T
U
U Pu
U & Pu recycling, COEX
FP&MA R T
U
U Pu
FP
TADS
« double strata »
MA (Pu)
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Fuel requirements for MA Fuel Recycling
• Neutronics and core physics– Safe operation, Conversion ratio =1
• Material properties– Fabrication feasibility– Margin to melt (Tf, λ, Cp)– Mechanical properties– Interaction with coolant– Interaction with cladding (chemical and mechanical)
• Irradiation Performance– High burnup– Swelling behaviour– Relocation/vaporisation behaviour– FP retention– Reprocessability (or not)
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Candidate fuels for transmutation
Fuel properties (U,Pu)
Metal Oxide Nitride Carbide
Heavy metal density (g.cm-3) 14.1 9.3 13.1 12.4
Melting point (K) 1350 3000 3035 2575
λ (W.m-1K-1) 16 2.3 26 20
T centreline (K) 1050 2350 1000 1000
Cp (J.g-1.K-1) 17 34 26 26
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Fabrication Feasibility
Additional Shielding (γ,n) compared to UO2 or MOX FacilitiesMinor Actinide Laboratory at JRC-ITU
(a) Shielded installations → remote handling(b) Automation → use of robots(c) dust free(c) process simplification : minimises the (active) fabrication steps
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Uranyl Nitrate Solution Plutonium Nitrate Solution
Ammonia Precipitation Oxalate Precipitation
Drying/ Calcination Drying/ Calcination
Powder Blending / Milling
Compaction
Sintering
Extension of Traditional Fabrication Technology
MA Nitrate Solution
Oxalate Precipitation
Drying/ Calcination
Advantages: Blend for individuals fuel pins or assemblies Commercially use
Disadvantages: Dust with particle sizes 2-3µm or less
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Liquid to solid conversion
U, Pu, Np , Am,( Cm)nitrate solution
Addition of polymers
Atomisation
Gelation in ammonia bath
Drying/ Calcination
Compaction
Sintering
U depO2 (NO3 )2
Sol gel route for group conversion of Gen V fuels
No fine powder but BEADS(Important for plant operation)
Beads diameter : 20 -600µm(depending droplet dispersion device)
Solid solution
However,
MA in all production steps
Extensive shielding, remote operation
and automation
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(U0.74Pu0.24Np0.02)O2
1mm 1mm
100µmm
Sol-gel beads
Fabrication ⇒sol gel method
Oxide fuel: SUPERFACT
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• Fuel Restructuring similar to standard MOX • Pore migration and central hole formation• Oxide layer (10 µm) on cladding• Reprocessing demonstrated
Typical observations for(U0.74Pu0.24Am00.2)O2 Fuel
SUPERFACT – a milestone irradiation test (CEA/ITU)Homogeneous MA Recycle in FR
Post Irradiation Examination (PIE)
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Before irradiation After irradiation
50 µm 50 µm
AmAlO3
HELIUM is an issue in MA fuelsRelease or swelling
241Am242mAm
242Cm
239Pu238Pu
242Am
243Cm
242Pu
10%
90%17%
83%
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matrix
Inclusion
iv
i
iii ii
i. The actinide particle with very high displacement damageii. Matrix damaged by recoil of fission products and energetic alpha
particles (5 µm) iii. Matrix damaged alpha particles (13 µm)iv. Matrix (neutron damage only) TAILOR PROPERTIES of the FUEL
Composite fuels for Accelerator Driven Systems
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CERMET Fuels
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CONCLUSIONS AND OUTLOOK
• Capitalise on past programmes
• Dedicated sample fabrication and property measurement
• Separate effect studies (in and out of pile)
• Integral irradiation testing
•Improve modelling at all time and distance scalesA NEW APPROACH NOW BECOMING POSSIBLE
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Thank you for your attention