Chaitanyamoy GANGULY - JAEA...IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles 1-2 December 2008, Fukui, Japan IAEA Activities on Nuclear Fuel Cycle - Programme B & INPRO

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


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.


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 Major Programme 1.2 (Prog. B) –Nuclear Fuel Cycle & Materials Technology



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


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


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)


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

(


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


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(


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


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


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)


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


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


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


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


NUCLEAR FUEL CYCLENUCLEAR FUEL CYCLE


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)


Global Statistics on Spent Fuel


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.”


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


Liquid Metal-cooled Fast Reactor Fuel Cycle with multiple recycling of U, Pu and Minor Actinides


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)


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


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 ?


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."


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


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


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


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


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


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

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


Electrorefining of Metallic Fuels


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 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)


• 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 (


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 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.


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


• 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)


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)


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


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


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)


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.


Coordination with GIF


…Thank you for your attention

Documents

Chaitanyamoy GANGULY - JAEA...IAEA IAEA Technical Meeting on Nuclear Fuel Cycle Profiles 1-2 December 2008, Fukui, Japan IAEA Activities on Nuclear Fuel Cycle - Programme B & INPRO