IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
IAEA Activities on Nuclear Fuel Cycle - Programme B & INPRO
Chaitanyamoy GANGULYHead , Nuclear Fuel Cycle & Materials Section
Division of Nuclear Fuel Cycle & Waste TechnologyDepartment of Nuclear Energy
International Atomic Energy Agency
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
NUCLEAR FUEL CYCLENUCLEAR FUEL CYCLE
Natural Uranium & Thorium
For nuclear energy to be sustainable as a global source of emission – free energy, the reactor fuel cycle must also remain sustainable (DG-IAEA Scientific Forum 2004)
Natural Uranium & Natural Thorium are the basic raw materials for nuclear fuels.
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
General Features of Uranium and Thorium Fuel CycleGeneral Features of Uranium and Thorium Fuel Cycle
Pu 239 : best fissile material forU238-Pu239 cycle in fast reactor
U233: best fissile material forTh232-U233 cycle in thermal reactor
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
IAEA Major Programme 1.2 (Prog. B) –Nuclear Fuel Cycle & Materials Technology
Natural Uranium & Thorium
Natural Uranium & Thorium
To facilitate development of ‘nuclear fuels’ and ‘fuel cycle technologies’ that:i) are economically viable;ii) make efficient use of uranium and thorium natural resources; iii) are safe; iv) generate minimum quantities of high level wastes; v) are proliferation-resistant; vi) are sustainable.
Subprog. 1.2.1 (B1) Uranium Resources & Production and Databases for Nuclear Fuel Cycle Working Group for B1: OECDNEA – IAEA Uranium Group(~34 members including EU), Databases iNFCIS [NFCSS, UDEPO, NFCIS, PIE, MADB]
Subprog. 1.2.2 (B2) Nuclear Power Reactor Fuel Engineering Working Group for B2: Technical Working Group on Nuclear Fuel Performance and Technology (TWGFPT)(28 Members including OECD/NEA & EC)
Subprog. 1.2.3 (B3) Management of Spent Fuel from Nuclear Power ReactorsSubprog. 1.2.4 (B4) Topical Issues of Nuclear Fuels and Fuel Cycle for Advanced and Innovative Reactors
Working Groups for B3&B4: Technical Working Group on Nuclear Fuel Cycle Options and Spent Fuel Management (TWGNFCO) (~41 members including ISTC, OECD/NEA & EC)
International Working Group
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Identified Resources (8 000>2 500Fast reactors with closed fuel cycle and recycling.
> 675300100Current technology
Using Total Conventional and Unconventional
Phosphate Resources
Using Total Conventional
Resources
Using only Identified Resources
Reactor/Fuel cycle
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Canada 8%
USA6%
Brazil6%
Niger5%
Namibia5%
Australia22%
South Africa8%
Russian Fed.10%
Kazakhstan15%
Uzbekistan2%
China1%
Ukraine4%
India1%
Distribution of Identified Uranium resources Worldwide Distribution of Identified Uranium resources Worldwide (Is the supply secure?) (Is the supply secure?)
Total Identified Resources: 5.55 Mt (2007) Total Identified Resources: 5.55 Mt (2007)
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Important data on Uranium resources, production, front-end of Uranium Fuel Cycle and Nuclear Power worldwide
(as on Nov. 2008)Uranium Production
(Mines & Mills) UF6
Conversion Enrichment Uranium
Fuel Fabrication
Reactors in Operation Reprocessing
MOX Fabrication
Identified Reserves
(
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
REVIVAL OF IAEA’s URANIUM PRODUCTION APPRAISAL TEAM (UPSAT) in 2008[UPSAT Guidelines : Ref: IAEA-TECDOC-878(1996)]
The IAEA Uranium Production Site Appraisal Team (UPSAT) programme is designed to assist Member States to improve the Operational and Safety performance of Uranium Production Facilities by peer review by a team of selected International Experts having direct experience in the technical areas specific to that operation.Member States should make a formal request to
IAEA for UPSAT service.
The major areas to be considered in developing the programme for an UPSAT mission are•Organization and management;•Mining •Ore milling and processing engineering;•Staff training;•Waste management;•General safety and Radiation protection;•Monitoring systems;•Environmental impact assessment;•Security;•Decommissioning/environmental remediation planning
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
TYPICAL FUELS FOR OPERATING NUCLEAR POWER REACTORS IN THE WORLDTYPICAL FUELS FOR OPERATING NUCLEAR POWER REACTORS IN THE WORLD
ZrZr--1%Nb clad1%Nb clad(
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
FUEL FORMSFUEL FORMS
Pebble Bedcoated particle fuels embedded in spherical shapeGermany, South Africa, China
Pebble fuel elementPebble has diameter of 60 mm
TRISO Coated fuel particles (left) are formed into fuel rods (center) and inserted into graphite fuel elements (right)
COATED RTICLES COMPACTS FUEL ELEMENTS
Pyrolytic CarbonSilicon CarbidePorous Carbon BufferUCO Kernel
Prismatic block
US, Japan, Russia and France
Trisocoated particle
B: Coated Fuel Particles1. Coated fuel particles for HTGR
2. Fuel particles (dry or wet route) for vibratory compacted fuel pins
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Coordinated Research Projects:• Fuel element performance modelling
(Fuel Modelling at Extended Burnup, FUMEX-II 2002-2007, FUMEX-III started in Dec.2008, RCM, Vienna)
• Optimisation of water chemistry(Optimisation of coolant water chemistry to ensure reliable fuel performance at high burnup and in ageing plants FUWAC 2007-2010)
• Delayed hydride cracking of Zr alloys(DHC-II, fuel cladding materials, 2005-2009)
• Simulation & Modelling of Radiation Effects(SMoRE) started in November 2008(RCM, Vienna)
Natural
Natural Uranium & Thorium
IAEA Technical Working Group on Fuel Performance & Technology (TWGFPT)
Data Bases:• IAEA PIE Data Base• Joint IAEA-OECD/NEA IFPE Data BaseExpert Reviews:• Handbook on Zirconium & its Alloys for Nuclear Applications (final draft in 2008)• Fuel Failures Review (final draft in 2008)
Subprogramme 1.B.2Nuclear Power Reactor Fuel Engineering
(focus on LWRs and PHWRs)
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
NFCIS NFCIS (Nuclear Fuel Cycle Information System)• Directory of Civilian Nuclear Fuel
Cycle Facilities Worldwide• Facilities from uranium milling to
reprocessing, spent fuel storage and heavy water production
• Includes facilities at planning stage and decommissioned
• Available online since 2001
NFCSSNFCSS(Nuclear Fuel Cycle Simulation System)• Scenario based simulation system• Estimates nuclear fuel cycle
material and service requirements• Calculates spent fuel arisings and
actinide contents• The simple web version is
expected to be online as of end of 2005
UDEPO UDEPO (World Distribution of Uranium Deposits)• Technical and geological
information on uranium deposits• Country level maps of the deposits
will be displayed on the web site• Deposits containing ≥0.03%U3O8
included • More than 800 deposits stored in
the database• Available online since 2004
MADB MADB (Minor Actinide Property Database)• Bibliographical database on
physico-chemical properties of minor actinides bearing materials
• Carbides, Nitrides, Alloys, Oxides, Halides, Elements and other forms are covered
• More than 750 data records from 164 publications
• Under development
PIEPIE(Post Irradiation Examination) • Catalogue of PIE facilities
worldwide• General information about the
facilities• Technical capabilities of the
facilities• Available online since 2004
www-nfcis.iaea.org
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Input and output information flow in Nuclear Fuel Cycle Simulation System(NFCSS) code
Input Output
NFCSS
Strategy Parameters• Nuclear power projections• Reprocessing-recycling
strategies• Reactor mixtures• Load factors
Fuel Parameters• Avg. discharge burnup• Avg. initial enrichment• Avg. tails assay
Control Parameters• Share of MOX fuel in core• Lead and lag times for
different processes• # of reprocessing cycles
• Natural uranium requirements
• Conversion requirements
• Enrichment requirements
• Fresh fuel requirements
• Spent fuel arisings
• Plutonium accumulation
• Minor Actinide accumulation
• Reprocessing requirements
• MOX fuel fabrication requirements
CAINCalculation of Actinide
Inventory
Input Output
NFCSS
Strategy Parameters• Nuclear power projections• Reprocessing-recycling
strategies• Reactor mixtures• Load factors
Fuel Parameters• Avg. discharge burnup• Avg. initial enrichment• Avg. tails assay
Control Parameters• Share of MOX fuel in core• Lead and lag times for
different processes• # of reprocessing cycles
• Natural uranium requirements
• Conversion requirements
• Enrichment requirements
• Fresh fuel requirements
• Spent fuel arisings
• Plutonium accumulation
• Minor Actinide accumulation
• Reprocessing requirements
• MOX fuel fabrication requirements
CAINCalculation of Actinide
Inventory
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Nuclear Power Capacity Projection
0
100
200
300
400
500
600
700
1950 1970 1990 2010 2030 2050
Year
Nuc
lear
Pow
er (G
We)
Cumulative Reprocessed Uranium Amount(PWR, BWR, WWER, GCR and AGR)
0
20 000
40 000
60 000
80 000
100 000
120 000
140 000
160 000
180 000
200 000
1950 1960 1970 1980 1990 2000 2010 2020 2030 2040 2050
Year
Rep
U (t
)
Cumulative Pu Figures
-1000.00
0.00
1000.00
2000.00
3000.00
4000.00
5000.00
6000.00
7000.00
8000.00
1950 1960 1970 1980 1990 2000 2010 2020 2030 2040 2050
Year
Pu (t
)
Discharged
Separated Used
Stock
Total MA Accumulation
0
200
400
600
800
1000
1200
1400
1990 2000 2010 2020 2030 2040 2050
Year
Tota
l MA
Am
ount
(t H
M)
Total
Am
Np
Cm
Projected power profile, cumulative Pu, RepUand MAs inventories up to 2050
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Increasing interest in back-end – reprocessing of spent fuel and recycling of fissile and fertile materials
• International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) - 2001
• Generation IV International Forum (GIF) – 2001
• IAEA initiated Multilateral Approaches to the Nuclear Fuel Cycle (MNA) - 2005
• Russian proposal for an International Fuel Cycle Centre –January 2006
• US Global Nuclear Energy Partnership (GNEP) - February2006
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
NUCLEAR FUEL CYCLENUCLEAR FUEL CYCLE
Natural Uranium & Thorium
The management of spent fuel and disposal of high level radioactive waste remain a challenge for the nuclear power industry. Public opinion will likely remain sceptical, and nuclear waste disposal will likely remain a topic of controversy, until the first geological repositories are operational and the disposal technologies fully demonstrated. ( DG – IAEA, 30 November 2006, JAEA, Tokyo, Japan)
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Global Statistics on Spent Fuel
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Developments and Challenges in Spent Fuel StorageDevelopments and Challenges in Spent Fuel Storage
• Currently, annual spent fuel discharge from power reactors:10,500tHM/y.Likely to rise to 11,500tHM/y by 2010. Annual spent fuel reprocessing capacity: 5000tHM/y . Cumulative spent fuel discharge/reprocessing( Sept. 2006): ~280,000/100,000 tHM’ Storage quantities and durations continue to grow…“long term storage is becoming a progressive reality.”
• More incentives for efficiencies, license extensions, possible multi-national cooperation…“major current issue – provide additional storage space.”
• As durations grow, “new challenges arise in the institutional as well as technical area…management of liabilities and knowledge…longevity of spent fuel packages and behavior of structural materials of storage facilities.”
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
MAJOR CHALLENGE IN NUCLEAR FUEL CYCLE
Develop fuel cycle options that make efficient use of natural urnium & thorium
resources, are economically viable, environmentally benign, proliferation-
resistant, safe and sustainable
Partitioning & Transmutation will make this possible
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Liquid Metal-cooled Fast Reactor Fuel Cycle with multiple recycling of U, Pu and Minor Actinides
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Fuel Cycle Flexibility
• CANDU technology provides the flexibility to use, and to optimise, a variety of fuels.
• NU, LEU, MOX, Thorium, DUPIC (Direct Use of Spent PWR Fuel in CANDU reactors)
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
1. Liquid Metal-cooled Fast Reactor (LMFR) Fuels & Fuel Cycle Options –multiple recycling of plutonium & burning minor actinides
2. High Temperature Gas-cooled Reactor (HTGR) Fuels & Fuel Cycle Options3. Small & Medium Size Reactor (SMR) Fuel with long core life4. Proliferation-Resistant Fuels and Fuel Cycles5. Thorium Fuel Cycle Options 6. Re-use option of reprocessed uranium
Major Activities UnderwayPreparation of TecDoc on LMFR Fuels TechnologyPreparation of TecDoc on Structural Materials for LMFR Fuel Assembly Preparation of TecDoc on LMFR Fuel Cycle TechnologyPreparation of TecDoc on Proliferation Resistant Fuel CyclePreparation of TecDoc on Protected Plutonium Production (PPP)Organization of IAEA- Technical Meetings on above topics Conducting Co-ordinated Research Projects (CRPs) on: i) Simulation and Modelling of Radiation Effects (SMoRE) , ii) Advanced structural materials for LMFR fuel assembly (2008-2011) & iii) Thorium fuel cycle options
IAEA Subprogramme 1.2..4(B 4)Topical Nuclear Fuel Cycle Issues
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Fissile and Fertile Materials are of dual use: Peaceful Application & Weapons
Why the need for “Proliferation Resistance” of nuclear materials (fissile & fertile) in Nuclear Fuel Cycle ?
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Dr, El-Baradei’s Statement in the “Special Event”, Sept. 19-22, 2006,IAEA General Conference.
"It is time to limit the processing of weapons-usable material (separated plutonium and high-enriched uranium) in civilian nuclear programmes, as well as the production of new material through reprocessing and enrichment, by agreeing to restrict these operations exclusively to facilities under multinational control,"."These limitations would need to be accompanied by proper rules of transparency and, above all, by an assurance that legitimate users could get their supplies."
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
eutecticsgood
goodgood
Carburisationgood
averageaverage
Compatibility - cladcoolant
Inert atmosInert atmospyrophoricEasyHandling
limitedvery littlelimited LargeGood
Fabrication/Irradiation experience
Pyro-reprocessing
risk of C14
DemonstratedGoodDissolution & reprocessing amenability
HighHigh (?)HighModerateSwelling 1.35 - 1.41.2 - 1.251.2 – 1.251.1 - 1.15Breeding ratio
γNaClNaClFluoriteCrystal structure
4015.820.1
18.821.2
2.6 2.4
Thermal conductivity(W/m ºK) 1000 K
2000 K
1400307027503083Melting point ºK
15.7314.3213.5811.04Theoretical Density g/cc
U-19Pu-10Zr(U0.8Pu0.2)N(U0.8 Pu0.2)C(U0.8Pu0.2)O2Properties
FUTURE CHALLENGES FUTURE CHALLENGES –– FUEL MATERIALSFUEL MATERIALS
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Increasing demands on structural materials
250
500
750
1000
0.1 1 10 100 1000
Radiation dose (dpa)
Ope
ratin
g te
mpe
ratu
re (°
C)
HTR Materials
Fast Reactor Materials
Fusion Materials
Thermal Reactor Materials
ITER
DEMO
• Growing operational requirements
• New fuel for the next generation of NPP
• Simulation and modelling of radiation effects (with the use of accelerators and research reactors)
• Synergies related to fission and future fusion power plants
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Cladding TubeCladding Tube
Handling HeadHandling Head
Duct TubeDuct Tube
Entrance NozzleEntrance NozzleLower End PlugLower End Plug
Upper End PlugUpper End Plug
Spacer WireSpacer Wire
Blanket Fuel Pellets (Blanket Fuel Pellets (UOUO22) )
Core Fuel Pellets (Core Fuel Pellets (MOXMOX))
WeldedWelded
WeldedWelded
Basic Structure of Core Fuel Pin and Subassembly in SFRBasic Structure of Core Fuel Pin and Subassembly in SFR
Fuel PinFuel Pin
SubassemblySubassembly
Core Support StructureCore Support Structure
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
PrePre--alloyed Powderalloyed Powderoror
Elemental PowderElemental Powder
YY22OO3 3 Powder Powder MA powder
Mild steel capsule:Φ67mm
MMechanical echanical AAlloying (MA)lloying (MA)(Ball Milling)(Ball Milling)
Hot Extrusion@ 1,423 K(1,150Hot Extrusion@ 1,423 K(1,150℃℃))
Raw MaterialRaw MaterialPowderPowder
CanningCanningDegassing@673K(400Degassing@673K(400℃℃))
DispersoidDispersoid Size ControlSize Control
Mother Tube Mother Tube
OD18mmOD18mm××IDID12mm12mm××LL180mm:200g180mm:200g
AnnealingAnnealing((ΦΦ25mm)25mm)DrillingDrilling
Manufacturing Process:1Manufacturing Process:1PPowder owder MMetallurgy (PM) Process for Mother Tubesetallurgy (PM) Process for Mother Tubes
Oxygen pickOxygen pick--upup
Two Candidates of ODS Steels for Further R&D in Two Candidates of ODS Steels for Further R&D in FaCTFaCT
1. Primary Candidate: 1. Primary Candidate: 9Cr9Cr--ODSODS-- FeFe--0.13C0.13C--9Cr9Cr--2W2W--0.2Ti0.2Ti--0.35Y0.35Y22OO33-- Normalizing@1,323 K(1,050Normalizing@1,323 K(1,050℃℃))××6060minmin→→TemperingTempering@1,053@1,053--1,073 K1,073 K××6060minmin-- Tempered Martensitic Matrix Tempered Martensitic Matrix -- Alpha to Gamma TransformationAlpha to Gamma Transformation-- Cooling Rate ControlCooling Rate Control
2. Secondary Candidate: 2. Secondary Candidate: 12Cr12Cr--ODSODS-- FeFe--0.03C0.03C--12Cr12Cr--2W2W--0.26Ti0.26Ti--0.23Y0.23Y22OO33--Annealed@~1,423K(1,150Annealed@~1,423K(1,150℃℃))××~~6060minmin-- RecrystallizationRecrystallization-- TwoTwo--step Annealingstep Annealing
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Partitioning processes
Aqueous based partitioning methods:
►PUREX: Plutonium Uranium Recovery by Extraction (F.Ps + MAs) in Raffinate + pure (U as well as Pu) streams ►TRUEX:- TRU (Transuranic) Extraction: Pu and minor actinide recovery from Raffinate►Supercritical fluid extraction
Dry partitioning methods
۩ Voloxidation, AIROX {USA} ۩ DUPIC (Direct Use of spent PWR fuel In
CANDU) [ROK]۩ Volatilization (Fluoride volatility
process)۩ Melt-refining, skull reclamation
methods۩ Halide-slagging {EBR-II,USA}۩ Electro-refining process (metal, oxide
& nitride) [USA, EC, Japan, India, ROK, RF]
۩ Oxide reduction [USA, EC, Japan, India, ROK, RF]
۩ Electro-winning process - DOVITA (Dry reprocessing, Oxide fuel, Vibropac, Integral, Transmutation of Actinides) [RF]
Processes primarily adaptations of TRUEX with co-recovery of actinides
• DIAMEX (Diamide Extraction process) to extract Am+Cm from raffinate
• SANEX: Selective Actinide Extraction (removal of actinidesfrom Raffinate viz., separating actinides from lanthanides)
• SESAME process for Am/Cm separation• NEXT (New Extraction System for TRU Recovery) for co-
recovery of actinides which includes uranium crystallization and micro-wave de-nitration (Japan)
• GANEX (Actinide Group Separation method) (France)• UREX process (Uranium Extraction) (USA) • CCD-PEG for Cs and Sr extraction from raffinate with
chlorinated cobalt dicarbollide/polyethylene glycol
Synergistic combinations of aqueous and pyro-methods
☼ UREX+PYRO-A [USA]
Manufacturing of MOX Fuels
Pu + U
FP + MAs
U+Pu co-conversion
Dissolution Partitioning
Actinides+
FP
U+Pu
Fuel processing
Used fuel COEXTM
The Proposed “Advanced Co-Extraction” Process for MOX Fuel in FranceEx: hydrometallurgy, co-processing of uranium and plutonium
Ref. S.Grandjean, et al., ATALANTE 2008, France
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Sol-gel Microsphere Pelletization (SGMP) Process for Manufacturing of (U,Pu)O2, (U,Pu)C or (U,Pu)N fuel pellets, using mixed uranium plutonium nitrate
solution as feed material
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
U Pu C
External Gelation
Calcination
Am Infiltration
Calcination
Carbothermal Reduction
Sintering
Pressing
ConventionalGloveboxesConventionalGloveboxes
ConventionalGloveboxes withPurified Atmospheres
MALABMALAB
(U,Pu)O2 + AmO2 + C
(U,Pu)O2 + C
U0,805Pu0,175Am0,02N
First Gen IV Nitride
Fuel at ITU
Nitride fuel fabrication
Mixed Nitride with MA by combined Infiltration – SGMP Process Ref. A. Fernandez, et al., ATALANTE 2008, France
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Electrorefining of Metallic Fuels
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Injection Casting Sphere and Vibro-Packing
Molten Metal Fuel
U
Reprocessing
Metal Electro-refining Oxide Electro-winningAdvanced Aqueous
Fuel Fabrication
Simplified Pelletizing
Mold Sphere Fuels
覆覆覆覆覆覆Cladding
UO2/PuO2/FPs
UO22+
UO2
UO2/PuO2/FPs
UO22+
UO2
UO2/PuO2/FPs
UO22+
UO2
UO2/PuO2/FPs
UO22+
UO2
UO2/PuO2/FPs
UO22+
UO2
UO2/PuO2/FPs
UO22+
UO2U
- Abundant Experiences- Rationalize to 1/3 Processes
[ Special Features ]- Simple Process- Small Fabrication Equipment
[ Special Features ]- Simple Process- Possibility of High Economics
[ Special Features ]
- Abundant Experiences- Higher Technical Realization
[ Special Features ]- Possibility of Higher Economics- Accomplishments in the U.S.
[ Special Features ]- Many Fundamental Subjects- Accomplishments in Russia
[ Special Features ]
14
CANDIDATES OF FUEL CYCLE SYSTEMSCANDIDATES OF FUEL CYCLE SYSTEMS
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
IAEA Minor Actinide Property Database (MADB)
• Bibliographic database on thermodynamic and thermophysical properties of minor actinide (Np, Am, Cm) metals, alloys and compounds
• Access on the internet with some search and filter capabilities (www-nfcis.iaea.org)
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
• Protected Plutonium Production (PPP) and utilization is a collaboration between Tokyo Institute of Technology and IAEA
on “intrinsic proliferation resistance” of growing Plutonium inventories (~1900 tons) through utilisation of Minor Actinides (MA) inventories (~200tons)[MA: Np, Am & Cm].
• PPP addresses the challenges of introducing low enriched uranium (
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Conclusions - continued3. Long-term Sustainability of Nuclear Energy
• Efficient utilization of natural resources, namely Uranium and Thorium through commercial introduction of ‘Fast Reactor’ and ‘Closed Fuel Cycle’,
• Addressing issues related to ‘proliferation resistance’ and ‘environmental protection’ through proper management of high level waste;
• “PPP” could be one of the attractive options – however, the economics has to be worked out to check its viability,
• Multilateral mechanisms for ‘Assurance of Fuel Supply’ and ‘Proliferation Resistance’ for peaceful use of nuclear energy for generation of electricity and process heat.
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) – established in 2001
Objectives• To help to ensure that nuclear energy is available to contribute, in a sustainable manner, to
meeting the energy needs of the 21st century • To provide a forum where experts and policy makers from countries that are technology
holders and technology users can consider jointly the international and national actions required for achieving desired innovations in nuclear reactors and fuel cycles
Missions • To provide a forum for discussion for experts and policy makers from industrialized and
developing countries on all aspects of nuclear energy planning as well as on the development and deployment of innovative nuclear energy systems in the 21st century;
• To develop the methodology to assess innovative nuclear systems on a global, regional and national basis and to establish it as an Agency recommendation;
• To facilitate coordination and cooperation among Member States for planning of innovative nuclear system development and deployment; and
• To pay particular attention to the needs of developing countries interested in innovative nuclear systems.
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
INPRO has 28 members: Argentina, Armenia, Belarus, Belgium, Brazil, Bulgaria, Canada, Chile, China, Czech Republic, France, Germany, India, Indonesia, Japan, Republic of Korea, Morocco, Netherlands, Pakistan, Russian Federation, Slovakia, South Africa, Spain, Switzerland, Turkey, Ukraine, USA and the European Commission (EC)
IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) – Membership as on August 2008
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
• Economics• Safety• Waste Management• Environment• Proliferation Resistance• Physical Protection and • Infrastructure.
IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO takes a holistic approach to assess innovative nuclear systems in 7 area)
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Assessment on additional nuclear energy capacity, for the period 2010-2030, using the INPRO Methodology for the Evaluation of the Nuclear Fuel Cycle (Argentina) National assessment study in Armenia using the INPRO methodology for an innovative nuclear energy system in a country with small grids (Armenia)Assessment of two small innovative reactors for electricity generation in Brazil using INPRO methodology (Brazil)Assessment of advanced high temperature gas cooled reactor (AHTGR) (China)Assessment using the INPRO methodology for hydrogen generating innovative nuclear system in the national energy mix (India)Korean assessment of the proliferation resistance on the whole fuel cycle of DUPIC (covering proliferation resistance issues) (Republic of Korea).Assessment of the innovative nuclear system of the Ukraine based on the INPRO methodology (Ukraine)Joint Assessment Study based on “Closed Fuel Cycle with Fast Reactors (Canada, China, France, India, Japan, the Republic of Korea, Russian Federation and Ukraine)
Application and development of the INPRO methodology (INPRO methodology is being or has been used in the following studies)
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Infrastructure and institutional innovation INPRO seeks to facilitate coordination for planning the development and deployment of innovative nuclear systems also by addressing issues such as a regional approach to smooth deployment of innovative nuclear systems, licensing and financing for developing countries, innovative options for nuclear fuel cycles and non-stationary small and medium sized reactors.
Common User Considerations in prospective new user countries regarding nuclear systems. Through the CUC activity, INPRO reached out to a total of twenty-six countries that are not members of INPRO itself. These are: Bangladesh, Cameroon, Croatia, Dominican Republic, Egypt, Estonia, Ethiopia, Georgia, Ghana, Jordan, Kenya, Lithuania, Malaysia, Mexico, Moldova, Mongolia, Namibia, Nigeria, Poland, Romania, Sudan, Syria, Tunisia, Uruguay, Venezuela, and Vietnam.A report summarizing the considerations of developing countries regarding future nuclear energy systems to be deployed by those countries was compiled in 2008.
INPRO’s Programme on Infrastructure & Common User Consideration
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Coordinated Research Project (CRP); Technical Cooperation Project (TCP); or Joint Initiatives (JI).As a key element of Phase 2 of INPRO, 12 Collaborative Projects are being implemented Scenarios for nuclear energy development• Global architecture of innovative nuclear systems based on thermal and fast reactors including closed fuel
cycles (GAINS). The objective of GAINS is to develop a standard framework (methodological platform, assumptions and boundary conditions) for assessing nuclear energy systems regarding sustainable development, covering from the existing systems to the innovative ones potentially deployed in the 21st Century, and validate its results via sample analyses.
• Meeting energy needs in the period of raw materials insufficiency during the 21st century (RMI)Nuclear safety • Performance assessment of passive gaseous provisions (PGAP)• Safety issues for advanced high temperature reactors and their combined operation with hydrogen
producing plants (HTR H2)Proliferation resistance• Proliferation resistance: acquisition/diversion pathway analysis (PRADA)
Collaborative Projects underway in INPRO
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Collaborative Projects underway in INPRO
Technical challenges in reactor technologies• Investigation of technological challenges related to the removal of heat by liquid metal and molten salt
coolants from reactor cores operating at high temperatures (COOL)• Advanced water cooled reactors (AWR) • Decay heat removal for liquid metal cooled reactors (DHR)
Environment, nuclear fuel cycle and infrastructure• Implementation issues for the use of nuclear power in smaller countries (SMALL)• Investigations of the 233Uranium/Thorium Fuel Cycle (THORIUM)• Fuel cycles for innovative nuclear systems through integration of technologies (FINITE)• Environmental impact benchmarking relative to an innovative nuclear system component (ENV)
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
The differences between INPRO and GIF include the following:
• Mission and activities: GIF is primarily focused on R&D of nuclear technology to meet global needs. INPRO has a broad variety of missions and activities, including providing a forum for experts on necessary innovation in nuclear energy, developing methodology to assess innovative nuclear systems, providing common user considerations for the deployment of nuclear power in developing countries, and facilitating international cooperation on technological issues.
• Membership: GIF membership is limited to those countries that can bring substantial resources and expertise to its R&D programs, whereas INPRO members include both developed and developing countries. INPRO is open to all IAEA Member States. Currently, all members of GIF are also members of INPRO.
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
Coordination with GIF
IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles1-2 December 2008, Fukui, Japan
…Thank you for your attention
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