Strategies of Fuel Cycle Management
Stéphane Bourg, Dominique Warin
CEA / Nuclear Energy Division Radiochemistry and Processes Department
Marcoule, France [email protected]
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 1 Gen IV Nuclear Reactor, 8-10 February 2010 CEA FACILITIES
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 2 Gen IV Nuclear Reactor, 8-10 February 2010 CEA/Nuclear Energy Directorate missions
A WORLD CLASS LABORATORY FOR SUSTAINABLE NUCLEAR FISSION ENERGY DEVELOPMENT REACTORS ● R and D in support of existing LW Reactors (material ageing, fuels, power plant life time, safety,...) z R and D for innovative future GEN IV Fast Neutron Reactors z Basic nuclear science and technology studies, materials z Thermal and fast neutron Test Reactors z Simulation program
FUEL CYCLE ● R and D in support of existing reprocessing plants (Areva La Hague, Sellafield, Rokkasho Mura,..) and for future recycling plants z R and D for nuclear waste management (1991 and 2006 French Acts) ● Hot radioactive facilities
INTERNATIONAL COLLABORATIONS : Europe (SNE-TP), USA (GNEP), Japan, China,…
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 3 Gen IV Nuclear Reactor, 8-10 February 2010 CEA/ Nuclear Energy Directorate organization
CADARACHE MARCOULE
 Nuclear reactors  Nuclear fuel cycle • Core design, neutronics • Radiochemistry, • Fuels reprocessing • Nuclear technology • Fuel cycle technology • Test reactors for PWRs, • Long term waste (RJH 2014) management • Na fast test reactor (Phénix, ASTRID 2020) SACLAY • Decommissioning DEN
GRENOBLE GRENOBLE SACLAY  Thermo hydraulics MARCOULE  Nuclear materials  Simulation, codes CADARACHE  Nuclear physics and chemistry
4500 staff, 900 M€/year
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 4 Gen IV Nuclear Reactor, 8-10 February 2010 CEA/Atalante: the world largest R&D hot lab for fuel cycle
Industrial support Innovative Co-conversion and processes fabrication
Waste conditioning and long term evolution
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 5 Gen IV Nuclear Reactor, 8-10 February 2010 Marcoule Activities I – Actinides and FP basic chemistry VII – Analyses II – Spent fuel dissolution
III – Extracting molecules
Spent fuel Recycling VI– Atalante facility operation treatment
Waste conditioning
V – Conversion processes IV – Partitioning processes
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 6 Gen IV Nuclear Reactor, 8-10 February 2010 INTRODUCTION
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 7 Gen IV Nuclear Reactor, 8-10 February 2010 The nuclear fuel cycle
Spent Fuel REACTOR
FuelFuel FabricationFabrication
EnrichissementEnrichissementUraniumUranium dedeEnrichmentEnrichment l'uraniuml'uranium UF6
MiningMiningandand MillingMilling
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 8 Gen IV Nuclear Reactor, 8-10 February 2010 The UOX spent nuclear fuel
235U=0.9%
Uranium (95%)
Fission Products (4%)
Plutonium (1%)
. Minor Actinides (0.1%)
Average Content Fissile isotopes = 75%
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 9 Gen IV Nuclear Reactor, 8-10 February 2010 UOX spent nuclear fuel (UOX : 3,5 % 235U ; 33 GWj.t-1) ORIGINE DES ELEMENTS CHIMIQUES DANS LE COMBUSTIBLE IRRADIE -1 1 U : 955 kg.t 2 H Pu : 9,6 kg.t-1 He 3 4 -1 5 6 7 8 9 10 Li Be MA : 0,8 kg.t B C N O F Ne 11 12 FP : 34 kg.t-1 13 14 15 16 17 18 Na Mg Al Si P S Cl A 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 K Ca Sc Ti V Cr Mn Fe Co Ni Cu Zn Ga Ge As Se Br Kr 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 Rb Sr Y Zr Nb Mo Tc Ru Rh Pd Ag Cd In Sn Sb Te I Xe 55 56 72 73 74 75 76 77 78 79 80 81 82 83 84 85 86 Cs Ba Ln Hf Ta W Re Os Ir Pt Au Hg Tl Pb Bi Po At Rn 87 88 104 105 106 107 108 109 110 Fr Ra An Rf Db Sg Bh Hs Mt Uun
57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 LANTHANIDES La Ce Pr Nd Pm Sm Eu Gd Tb Dy Ho Er Tm Yb Lu 89 90 91 92 93 94 95 96 97 98 99 100 101 102 103 ACTINIDES Ac Th Pa U Np Pu Am Cm Bk Cf Es Fm Md No Lr
NOYAUX LOURDS PRODUITS D’ACTIVATION
PRODUITS DE FISSION PRODUITS DE FISSION ET D’ACTIVATION
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 10 Gen IV Nuclear Reactor, 8-10 February 2010 Radiotoxicity of LWR-UOX Spent Fuel
Total UOX Fuel Pu 1010 Minor Actinides Fission Products 109 U 8 10 Activation Products
107
106
105
104
103
102
10 Radiotoxicity Inventory (Sv/TWhe)Radiotoxicity Inventory 10 100 1 000 10 000 100 000 1 000 000
Time (years)
Ingestion Doses Coefficients from ICRP72
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 11 Gen IV Nuclear Reactor, 8-10 February 2010 UOX and MOX spent fuels : typical content
UOX (45 GW/t) MOX (45 GWj/t)
URANIUM ∼ 940 kg ∼ 890 kg
PLUTONIUM ∼ 11 kg ∼ 56 kg
FISSION ∼ 48 kg ∼ 48 kg PRODUCTS MINOR ACTINIDES ∼ 1 kg ∼ 6 kg
(for 1 tonne initial HM)
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 12 Gen IV Nuclear Reactor, 8-10 February 2010 Spent fuel management : what options ?
- Interim storage
- Direct disposal
YUCCA MOUNTAIN, NEVADA
- Process and recycle
MOX fuel
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 13 Gen IV Nuclear Reactor, 8-10 February 2010 Nuclear acceptance and a responsible management of spent fuel and waste A responsible management of nuclear spent fuel :
• Recycles 96% of spent fuel materials • Saves 30% of natural resources • Costs less than 6% of the kWh total cost • Reduces by 5% the amount of wastes • Reduces by 10% the waste radio toxicity • Adapted technologies allow a safe conditioning of wastes to guarantee their long term confinement and stability, during dozens of thousands of years
Recycling, such as implemented today in France, GB or Japan, gives time and opens a large range of options for the satisfactory management of nuclear wastes. more sustainable policy better public acceptance
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 14 Gen IV Nuclear Reactor, 8-10 February 2010 Recycling strategy
• Aim of recycling : to minimize the quantity and radiotoxicity of long lived nuclear waste
• Potential gain : to reduce at hundreds y
years the life time of future waste t
i
c
i
x
e
v
o
i t t Spent fuel
(compared to the level of initial U ) o
a i
nat l d
e (Pu + MA + FP)
a r and to ease the conditions of a deep R Natural uranium ore MA + FP geological repository FP
Time (years) • Main radiotoxicity contributors : 1) Pu 2) MA 3) LLFPs • Fundamental hypothesis : closed cycle, GEN IV fast neutron reactor implemented at mid or long term (whatever the cooling : Na, gaz,…) • Reference strategy : – PWRs: Pu management by multi recycling – GEN IV SFR or GFR: full actinides recycling, beginning 2040 – Option of the « double strata » PWRs – Accelerator Driven Systems
• Demonstrations of scientific and technical feasibilities,
before pre industrial development of recycling
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 15 Gen IV Nuclear Reactor, 8-10 February 2010 Recycling strategy : systems evolution…
(Fuel cycle)
• The new requirements :
– Save resource – Minimize environment impact (waste nocivity)
– Increase resistance vs proliferation risks
– New purposes, other than electricity (heat, H2, dessalinization)…
– And obviously : safety, costs…
(Reactors)
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 16 Gen IV Nuclear Reactor, 8-10 February 2010 The Evolution of Nuclear Power to fast systems
Future Advanced Advanced Systems Current Reactors Reactors First Reactors
1950 1970 1990 2010 2030 2050 2070 2090
Generation I
UNGG Generation II CHOOZ REP 900 Generation III REP 1300 N4 EPR Generation IV ? CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 17 Gen IV Nuclear Reactor, 8-10 February 2010 Transition scenarios between generations in France
¾ Major role of LWRs in the 21st century Current PWRs (Gen II ): life time extension (> 40 years) Gen III/III+ PWRs : current PWRs replacement around 2015 (EPR) – Operation during 21st century ¾ A transition scenario between LWRs and Fast Neutrons Systems
1975 2000 2025 2050 2075 Reactors Life time extension Gen IV Current Fleet EPR Source : EDF, ENC 2002
Cycle
U (Udep, URT) Recycling Cycle Gen IV Pu (MOX recycling) Gen IV M.A. ---> glasses or recycling Disposal M.A. + F.P. ---> glasses F.P. ---> glasses
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 18 Gen IV Nuclear Reactor, 8-10 February 2010 Future fuel cycle options, Reactor and Treatment
U U U R T FP&MA R T FP R T FP Pu MA U U Pu U Pu AM U and Pu recycled, MA heterogeneous MA homogeneous PUREX recycling recycling
U « double strata » U R T FP R T FP&MA MA U Pu U Pu U and Pu recycled, ADS T COEX
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 19 Gen IV Nuclear Reactor, 8-10 February 2010 THE FUEL
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 20 Gen IV Nuclear Reactor, 8-10 February 2010 Different fuels
UOX homogeneous UO2 UO2 UO2 PuO2 UO2 UO2 UO2 PuO2 UO2 UO2 UO2 UO2 MAO2 UO2 UO2 UO2 UO2 UO2 UO2 UO2 UO2 UO2 PuO2 UO2 UO2 UO2 UO2 UO2 heterogeneous MOX UO2 UO2 UO2 UO2 UO2 UO2 PuO2 UO2 UO2 PuO2 MAO2 UO2 PuO2 UO2 UO2 PuO2 UO2 UO2 UO2 UO2 UO2 UO2 UO2 UO2 UO2 UO2 MAO2 PuO2 UO2 UO2 UO2 PuO2 UO2 UO2 UO2 UO2 UO2 MAO2
Winter Seminar Demänovà, Slovakia 21 Gen IV Nuclear Reactor, 8-10 February 2010 What characterize the MA bearing fuels Transmutation fuel development is considerably more challenging than conventional fuels :
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 22 Gen IV Nuclear Reactor, 8-10 February 2010 Oxide versus metal, carbide or nitride ?
- Metal - a good option for SFR fuel, in the US - a large experimental program in the US National Labs - MA experiments in the Idaho ATR (UPuAmZr) - Futurix experiment in Phénix (UPuAmZr and PuAmZr)
- Carbide - a good option for SFR fuel, “almost” reference option for GFR fuel - France (limited) and India (large) experience on UPuC - no MA-UPuC experiment so far ? - pyrophoricity of divided material, fabrication and reprocessing in safe industrial conditions ?
-Nitride - a good experience in Japan, France and US - a complicated fabrication process, N15 enrichment - stability of Am nitride compounds as function of temperature ? - Futurix experiment in Phénix (PuAmZrN and PuAmN)
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 23 Gen IV Nuclear Reactor, 8-10 February 2010 THE PARTITIONING
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 24 Gen IV Nuclear Reactor, 8-10 February 2010 The French Act of December 1991, for 15 years
Waste management in France: which goals? Goal 1 research : to reduce the overall radwaste quantity by solutions that would allow partitioning and transmutation of long lived radionuclides present in HL LL nuclear waste Goal 2 research : to study the feasibility of reversible or irreversible repository in deep geological formations Positive results of Underground Laboratory: Dossier Argile 2005 : clay in Meuse/Haute Marne site
Goal 3 research : to study conditioning processes and feasibility of long term storage, above-grade or below-grade Sub surface long term storage demonstration gallery Galerie constructed at CEA-Marcoule ; long term storage demonstration containers (scale 1) BDeadline set in law : 2006
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 25 Gen IV Nuclear Reactor, 8-10 February 2010 The french Act of June 2006, up to 2015 and 2020
Ö Key-points for waste management, and future R and D :
• A national plan on radioactive materials and radioactive waste management (up-grading by Parliament every three years) • A program on R & D with a time schedule to implement this plan • A secured financing of radioactive waste management and R & D (dedicated fund) • A step by step program of HLLL waste management, including the 3 complementary solutions :
– Geological disposal for the final HL waste (dossier and Parliament decision in 2015, operation in 2025) ; existing wastes will be disposed – Intermediate storage for industrial flexibility
– MA Recycling : - R & D in the framework of Gen IV Systems; international framework - interest (as consequences on the geological repository) to be proved with a cost/benefit detailed analysis by 2012 - 2012 decision : demo by 2020 ; industrialization by 2040…
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 26 Gen IV Nuclear Reactor, 8-10 February 2010 What strategy in process development?
Ligand design and assessment Development of separation processes
Organic synthesis Optimisation of of novel ligands solvent formulation Assessment of complexing and Selection extracting criteria properties Assessment of Molecular ligand radiolytic structural & hydrolytic studies stability Actinide Science Selection criteria Flowsheet modelling & design
Implementation of Process cold, spiked and HA counter- current tests
Winter Seminar Demänovà, Slovakia 27 Gen IV Nuclear Reactor, 8-10 February 2010 The principle of solvent extraction
Step1 Step 2 Step 3 Preparation Emulsion Separation
Organic solution
[M ]org D()or K D = Aqueous solution []M aq
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 28 Gen IV Nuclear Reactor, 8-10 February 2010 Extraction of a specie
Pulsed Column (PC) (M)aq ' (M)org
Very diverse mechanisms: - solvatation (complexes) - ion exchange (anions, cations)
- chelation Mixer-Settler (MS) (all organic species) [M ]+ [complex1]+ [complex 2]+ ... D = M [][M + complex1 + complex ][2 + ... ] (all aqueous species)
For two elements, M1 and M2 to be separated : D M 1 FS M /M = 1 2 D M 2
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 29 Gen IV Nuclear Reactor, 8-10 February 2010 Solvent extraction
• Counter-current separation process
Extractant(s) Solvent treatment Diluent
EXTRACTION BACK-EXTRACTION
Raffinate Stripping Solution M2 Feed M1 Stripping M1 + M2 solution
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 30 Gen IV Nuclear Reactor, 8-10 February 2010 Requirements for process development
• Extraction efficiency – Distribution ratios of separated elements • High distribution ratios at the extraction step O D × M A >> 1 Ö E(%) > 50% • Small distribution ratios at the stripping step O D × M A<< 1 Ö E(%) < 50% DM • High selectivity: FS = 1 >> 1 M1/M2 D M2 • Fast kinetics of mass transfer – Both at extraction and stripping steps • Implementation in short time contactors • Ligand stability – Both vs acidic hydrolysis and radiolysis
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 31 Gen IV Nuclear Reactor, 8-10 February 2010 Actinides properties
SYMBOL
OXIDATION STATES
CHEMICAL BEHAVIOR: Light actinides (Th to Pu): specific redox (PUREX basis) Ö Np recovery is achievable by PUREX Heavier actinides (from Am) behave like lanthanides Ö An(III) recovery requires new separation processes
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 32 Gen IV Nuclear Reactor, 8-10 February 2010 Fuel cycle, the MA heterogeneous recycling option
URANIUM
REACTOR TREATMENT FPs
MAs
Pu,U,(Np) U
- U, Pu, Np by COEXTM - Am (and Cm) separation : (DIAMEX-SANEX), (SANEX- TODGA)… - Am (and Cm) recycled on dedicated « targets-blankets »
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 33 Gen IV Nuclear Reactor, 8-10 February 2010 Actinide recovery by solvent extraction
A possible route…
Minor Actinides SPENT FUEL
Uranium and Np Plutonium PUREXPUREX
Fission, Activation, COMPLEMENTARY COMPLEMENTARY Am and Corrosion Products STEPSSTEPS Cm
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 34 Gen IV Nuclear Reactor, 8-10 February 2010 The PUREX Process
TriButylPhospahte TBP U TBP
SPENT FUEL HNO 3 Pu
U, Pu, DISSOLUTION FPs,MAs EXTRACTION solution
HULLS FPs, MAs
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 35 Gen IV Nuclear Reactor, 8-10 February 2010 An evolution: from PUREX to COEX™: co-extraction (U-Pu-Np)
Feed Increase the non proliferation resistance Extraction/scrubbing « no pure plutonium alone » FP + MA Pu,U,Np U
Np scrubbing Pu stripping U stripping
100% of initial Pu, Np nd U/Pu ~ 1 HAN To 2 U cycle ?
Oxidizing
U(VI),Pu(IV),Np(V,VI)
Extraction/scrubbing Pu stripping HAN FP U(IV) U(VI),Pu(III),Np(IV)
Concentration and Oxidizing Reduction buffer storage U(VI), Pu(IV), Np(V,VI) U(IV),U(VI),Pu(III),Np(IV)
To Adjustment UVI scrubbing co-conversion U(IV),Pu(III),Np(IV) b
a Atalante/L15 U(IV) CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 36 36 Gen IV Nuclear Reactor, 8-10 February 2010 Partitioning research steps
Ö A true challenge : a sophisticated partitioning chemistry under highly radioactive conditions
Ö Fundamental research : a wide co-operative framework - exploration : new extracting systems - fundamentals : in-depth study of mechanisms at work A few hundreds Ö Applied research : of new molecules - process design - lab experiments on actual spent fuel material Scale : 1/10000 - « demonstration » experiments : integration, representativeness, long-lasting performance Scale : 1/100 à 1/1000
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 37 Gen IV Nuclear Reactor, 8-10 February 2010 Actinides Partitioning : the research path used
1 2 H He 3 4 5 6 7 8 9 10 Li Be B C N O F Ne 11 12 13 14 15 16 17 18 Na Mg Al Si P S Cl A 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 K Ca Sc Ti V Cr Mn Fe Co Ni Cu Zn Ga Ge As Se Br Kr 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 Rb Sr Y Zr Nb Mo Tc Ru Rh Pd Ag Cd In Sn Sb Te I Xe 55 56 72 73 74 75 76 77 78 79 80 81 82 83 84 85 86 Cs Ba Ln Hf Ta W Re Os Ir Pt Au Hg Tl Pb Bi Po At Rn 87 88 104 105 106 107 108 109 110 Spent fuel Fr Ra An Rf Db Sg Bh Hs Mt Uun 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 LANTHANIDES La Ce Pr Nd Pm Sm Eu Gd Tb Dy Ho Er Tm Yb Lu 89 90 91 92 93 94 95 96 97 98 99 100 101 102 103 ACTINIDES Ac Th Pa U Np Pu Am Cm Bk Cf Es Fm Md No Lr
NOYAUX LOURDS PRODUITS D’ACTIVATION U, PRODUITS DE FISSION PRODUITS DE FISSION ET D’ACTIVATION Np Pu PUREXPUREX
O O Fission CO-EXTRACTIONCO-EXTRACTION H3C CH3 DIAMEXDIAMEX N N products ofof AnAn etet LnLn C8H17 C2H4 C8H17 O C6H13 Am DMDOHEMA PARTITIONINGPARTITIONING Ln SANEXSANEX Cm betweenbetween AnAn && LnLn O O P HDEHPO O H
Winter Seminar Demänovà, Slovakia 38 Gen IV Nuclear Reactor, 8-10 February 2010 Np Extraction demonstration hot run
Feed HA U, Pu, Np, P.F.
HNO3 4.5 M ~1 L/h F.P. scrub U, Pu, Np Acidification ( > 99%) 2 extractions (PC) F.P. scrub (PC) Tc scrubbing (MS)
4 m high Pulsed columns
Scrub TBP 30%
Np < 0.1 % (d.l.), F.P. - ~ 15 kg of UOX fuel, Complementary 52 GWd/t extraction (MS) TBP 30% - CEA Marcoule - April 2005 Np < 0.5 % (d.l.), F.P.
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 39 Gen IV Nuclear Reactor, 8-10 February 2010 Am3+ and Cm3+ recovery : a 2-step strategy
Spent Fuel O P O O O
U, Pu PUREXPUREX Np TBPTBP
Fission CO-EXTRACTIONCO-EXTRACTION 11 Products ofof MAsMAs andandLnLn
(e.g., CMPO, TRUEX processes using hard-donor ligands)
SEPARATIONSEPARATION Ln 22 Am, Cm ofof MAsMAs fromfromLnLn
(e.g. SANEX, TALSPEAK processes Two step strategy Two step strategy using soft-donor ligands)
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 40 Gen IV Nuclear Reactor, 8-10 February 2010 DIAMEX demonstrative hot run, November 2005
PUREX raffinate 15 kg genuine fuel
NaOH HEDTA Solvent Treatment HNO3 Am ~ 150 mg/L (ECRAN) HNO 3 Cm ~ 15 mg/L H2C2O4 Ln ~ 2.5 g/L 4 m high HEDTA V ~ 1 L/h HNO Pulsed columns 3 H2C2O4 HEDTA
Back-extraction- An-Ln Scrubbing extraction extraction (MS) An+ Ln An+ Ln FP (CP) HNO (CP) (CP) 3 0.65 M DMDOHEMA/TPH DIAMEX Production Am, Cm, Ln
Am ~ 0.015 % FP Am, Cm, Ln Cm < 0.002 % Raffinate > 99.9%
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 41 Gen IV Nuclear Reactor, 8-10 February 2010 SANEX demonstrative hot run, December 2005
HNO3 NaOH HEDTA
extraction-scrubbing O O H C CH Solvent DIAMEX 3 N N 3 DMDOHEMA 0.5 M C H C H C H Production 8 17 2 4 8 17 + HDEHP 0.3 M O C H TPH Am, Cm, Ln 6 13 HNO DMDOHEMA ~ 45 mL/h 3
O O P HDEHP O O Raffinate Am, Cm < 0.01% H
O O O O
N N O
Ln OH O An stripping Ln stripping scrubbing HEDTA
Ln HEDTA HNO3 An Am, Cm > 99.9% Citric ac. F.D.Ln ~ 80 Am < 0.06 % Cm < 0.03 %
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 42 Gen IV Nuclear Reactor, 8-10 February 2010 Simplified TODGA-SANEX process, March 2009
• Co-extraction An (III) and Ln (III) with TODGA, using HNO3 4N 104
U(VI) Np (IV ) 3 Np (V) 10 Pu(III) Pu(IV) C H Am(III ) C8H17 8 17 Cm (I II) 102 TODGA concentration: 0.1M/ n-do deca ne N N 1 O 10 C H C8H17 8 17 D(An) O O 100
10-1
10-2
10-3 10-2 10-1 100 101
• Selective back-extraction of An (III) HNO3 concentration (M) • With polyamino-carboxylic hydrophile complexing agent
COEX Back Extr. An Extr. TODGA Back Extr. Ln Raffinate An + Ln
PF An DTPA Ln pH 2 + NO3 • Advantages : simple scheme, TODGA synthesis low cost • Drawbacks : high sensitivity of the Am-Cm back extraction step to pH and temperature CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 43 Gen IV Nuclear Reactor, 8-10 February 2010 Fuel cycle, the MA heterogeneous recycling option
U
R T FP
U Pu AM
MA homogeneous recycling
- U recovery: GANEX 1 -Pu, Np, Am (and Cm) separation : GANEX 2 - Am (and Cm) recycled in “homogeneous fuel” CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 44 Gen IV Nuclear Reactor, 8-10 February 2010 Fuel cycle, the MA homogeneous recycling option
The grouped actinide GANEX concept RECYCLE Spent fuel
DISSOLUTION UPuNpAmCm Conversion
U 1st step U Extraction Pu+MA
2nd step An Stripping Ln Stripping
F.P. Ln
WASTE
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 45 Gen IV Nuclear Reactor, 8-10 February 2010 The GANEX process
HA spent fuel solution ∼ 4M HNO3
Excellent compromise DEHiBA Monoamide U between complete Uranium
O extraction, and good
N U(VI)/Pu(IV) selectivity
TRU + PF
O O H C 3 H CH3 N C N
C H C H 8 17 C H 8 17 O O 2 4 P O O OH C6H13
HDEHP- Adaptation of the process DMDOHEMA DIAMEX-SANEX TRU HEDTA/citric acid already demonstrated in 2005
PF (dont Ln)
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 46 Gen IV Nuclear Reactor, 8-10 February 2010 GANEX demonstrative hot runs, 2008
1st step : U selective extraction
Mono Amide Solvent clean-up (performed successfully in June 2008) Extraction Scrubbing U stripping
Np, Pu, Am, Cm + Feed HNO3 U Diluted FPs HNO3 U > 99.9%
2nd step : Pu-Np-Am-Cm co-recovery (diamide-based process) (performed successfully in November 2008)
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 47 Gen IV Nuclear Reactor, 8-10 February 2010 Recycling process : consolidation towards industrialization
Long term effects (solvent treatment)
O O
H3C CH N N 3 Modeling C H C H Modeling 8 17 C H 8 17 2 4 (flowsheet, sensitivity, operation) O C6H13
Extractant synthesis Core of the process demonstrated at small scale DMDOHEMA + HDEHP An(III) + Ln(III) Ln(III) Extr-scrub. An stripping Ln stripping
DTPA PF Raffinat An pH 3-4 Ln PUREX Interface FP solution adapted Co-conversion to vitrification Equipments (Concent. Calcination)
3000 Essai Nd 17/11/2005 2900 2800 suivi ANL , resultats analyses et calcul (Implementation, 2700 (Implementation, 2600 2500 2400 2300 2200 2100 2000 1900 ANL AP ANL ASPA extrapolation) 1800 extrapolation) In situ ANL CP 1700 Analyse CBA AP 1600 Analyse CBA ASPA
) Analyse CBA CP -1 1500 Calcul AP 1400 Calcul ASPA 1300 Calcul CP 1200 [Nd] (mg.L [Nd] 1100 1000 900 800 700 analysis 600 500 400 300 200 100 0 17/11 17/11 17/11 17/11 17/11 17/11 17/11 18/11 18/11 18/11 18/11 18/11 18/11 18/11 18/11 18/11 18/11 18/11 (cations, 17:00 18:00 19:00 20:00 21:00 22:00 23:00 00:00 01:00 02:00 03:00 04:00 05:00 06:00 07:00 08:00 09:00 10:00 Temps pH...)
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 48 Gen IV Nuclear Reactor, 8-10 February 2010 Modelling
Speciation in organic phase, thermodynamics
Radiolysis • For what? – Reliable Calculation Codes • Draft flowsheets in industrial scale Interfacial reactions kinetics, thermodynamics • Minimise number of experimental tests, steps and measurements Speciation in • Carry out operating analysis and transitory calculations aqueous phase, Thermodynamics •How? – Phenomenological representation
• Main equilibria during MA selective stripping
O
DIAMEX-SANEX: N
N
O
O
O O O 3+ + O O
• Extraction: M + n (HDEHP) Ù [M, 3DEHP,HDEHP] + 3 H A N
O
m
O O O
N N N N
O O
O
O O O O O
N N N O O O O
O 3+ + O
O O O N
• Complexation: M + HEDTA Ù (M,HEDTA) + 3 H N
N N O SANEX-TODGA: O 3+ - • Extraction: M + 3 NO3 + 3 TODGA Ù [M(NO3)3,3 TODGA] • Complexation: M3+ + HEDTA Ù (M,HEDTA) + 3 H+
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 49 Gen IV Nuclear Reactor, 8-10 February 2010 Solvent clean up
• Objective : assessment of long term solvent behaviour – Clean up efficiency destined to remove acidic degradation products issued from hydrolysis/radiolysis
Continuous long term runs Simulating 1-2 years of industrial operation
Ö No significant impact on Hydraulic and Recovery performances
Solvent studied: • DMDOHEMA/TPH • DMDOHEMA/HDEHP/TPH Yet to be tested: Ö TODGA Ö DIAMEX-SANEX MARCOULE γ-Irradiation loop
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 50 Gen IV Nuclear Reactor, 8-10 February 2010 Solvent clean up
• Need of solvent purification step • Recover pure diluent • Extractant concentration • Decontamination/some degradation products Thermosensitive molecules
• Assessment • Evaporation under vacuum • Flash distillation • Supercritical carbon dioxide extraction
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 51 Gen IV Nuclear Reactor, 8-10 February 2010 The main results of P and T research programs
• New possibilities, continuous progress : – Extraction molecules have been developed at lab scale, and pre industrial demonstrations have proved successful : industrial partitioning is feasible
– Transmutation of Fission Products (I, Cs, Tc) is either not feasible or unrealistic – MAs transmutation is unrealistic in PWR – Transmutation experiments, at pin scale, have been carried out for americium and neptunium in a power reactor, such as Phénix, which demonstrates the feasibility of their transmutation in SFR – Feasibility of transmutation in ADS is not obtained today, and research must going on
– Realistic scenario : Transmutation can be carried out only with the new generation of fast neutron reactors, as a contribution to sustainable nuclear energy by homogeneous or heterogeneous mode
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 52 Gen IV Nuclear Reactor, 8-10 February 2010 Pyrochemical reprocessing
9EBRII spent fuel reprocessing in US-INL,
For TRU recovery, were proposed : • either the use of a liquid cathode • or the use of a liquid-liquid reductive extraction step Pyrochemical reprocessing
9mixed oxide recovery in Russia-RIAR for Mox.
Limited to U [or (U,Pu)O2] recovery on solid inert cathode
Dry Dimitrovgrad Process Production of MOX grains convenient for Vibropac technology (will be used to prepared BN800 fuel The PyroReprocessing in European Research Projects
Electrorefining on Liquid-Liquid Solid Aluminium Reductive Cathode in molten Extraction in chloride molten fluoride
T = 460°C
Bi Solid Al Electrorefining in molten chloride
• Based on the IFR concept, the process is centered on the selective electrorefining of An on solid aluminium cathode in molten chloride Conf. of Cl waste
Exhaustive Precipitation Electrolysis FIltration
Winter Seminar Demänovà, Slovakia 56 Gen IV Nuclear Reactor, 8-10 February 2010 Liquid-liquid reductive extraction
• This process is centered on the selective extraction of An in molten fluoride / liquid aluminium.
Fuel Dissolution
Conf. of metal. waste
Winter Seminar Demänovà, Slovakia 57 Gen IV Nuclear Reactor, 8-10 February 2010 FUEL REFABRICATION
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 58 Gen IV Nuclear Reactor, 8-10 February 2010 General consideration on existing industrial conversion for U-Pu • The PUREX process – AREVA’s UP2 & UP3 (La Hague) – Hydrometallurgy, separated purification cycles for U and Pu Fuel processing Manufacturing of MOX Dissolution Partitioning Pu Conversion Fuels (oxalic) + UO2
Used fuel U Pu Pu U + Pu U Conversion Actinides FP + MAs (peroxide, ADU,…) + FP
UOX U
– Conversion • For uranium and plutonium : conversion of each actinide into oxide (following a specific purification cycle), via continuous oxalic precipitation for Pu and continuous peroxide precipitation (in France) for U.
•PuO2 = one of the starting material for MOX fabrication (MELOX process). CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 59 Gen IV Nuclear Reactor, 8-10 February 2010 General considerations on « existing » industrial conversion for (U,Pu) • Advanced processes for Gen III+/IV fuel cycle (e.g. COEXTM) – Ex: hydrometallurgy, co-processing of uranium and plutonium Fuel processing Manufacturing of MOX Dissolution Partitioning U+Pu Fuels oxalic co-conversion TM Used fuel COEX
Pu + U Actinides U+Pu FP + MAs + FP M2+x(U2-x,Pux)(C2O4)5O2 hyd.
(U,Pu)O2
– Co-conversion • Co-conversion of U and Pu into oxide following the partitioning steps and more integrated to the fuel refabrication (MOX).
• U+Pu co-conversion product (U,Pu)O2 = fuel precursor (LWR then FR).
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 60 Gen IV Nuclear Reactor, 8-10 February 2010 General considerations on future conversion, for MAs • Innovative processes: e.g. COEXTM + MAs Transmutation (het.) – U+Pu comanagement and separated MAs recycling (het. mode) Fuel Treatment Recycling Co-conversion Dissolution Partitioning Spent U fuel
COEXTM
Actinides U+Pu Actinides FP + FP DX-SX
U+MA
– Mixed compounds – a major UPu stream + a specific MAs stream • e.g. oxalic process:
• M2+x(U2-x,Pux)(C2O4)5O2 hyd. + M2+x((U,Np)2-x,(Am,Cm)x)(C2O4)5O2 hyd.
• (U,Pu)O2 + (U,MA)O2-x
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 61 Gen IV Nuclear Reactor, 8-10 February 2010 General considerations on future conversion, for MAs
• Advanced processes: e.g. GANEX with MAs transmut. (homogeneous) – Homogeneous recycling of all the actinides together Fuel treatment
Dissolution Partitioning Co-conversion Recycling U Spent Fuel
GANEX
Actinides or U+Pu Actinides +Np + FP PUREX +Am FP +DX-SX +Cm
Σ Actinides
– One mixed actinides end-product – One fuel precursor • e.g. oxalic process:
• M2+x((U,Np)2-x,(Pu, Am,Cm)x)(C2O4)5O2 hyd.
• (U,Np,Pu,Am,Cm)O2-x
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 62 Gen IV Nuclear Reactor, 8-10 February 2010 Methods of coconverting actinides into oxide
• Conversion and coconversion existing/possible methods – Coprecipitation + Thermal treatment • Oxalic Pu(IV), Pu(III), An(VI) • Hydroxide U(VI), Pu(IV), An(III) • Carbonate U et Pu(VI), An(V) • Peroxide… U(VI), Pu(IV) – Sol-Gel + Thermal Treatment U(VI), Pu(IV), An(III), An(V) • Colloidal Sol-Gel • Internal or external gelation (gelating additives) – Thermal Denitration U(VI), Pu(IV) • Direct Denitration with/without additives • Plasma Denitration • MW Denitration • Denitration within a template matrix (e.g. ceramic beads) B New requirements for GenIII/GenIV systems • Economy of resources (recycling of U and Pu) • Increased proliferation resistance (co-processing of actinides) • MAs management (P long-term radiotoxicity and volume of waste) High activity and remote operation, thermal and radiolysis effect… Innovative concepts of fuels and reactors… B Innovative processes CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 63 Gen IV Nuclear Reactor, 8-10 February 2010 CONCLUSION
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 64 Gen IV Nuclear Reactor, 8-10 February 2010 General Objective
Spent Fuel
Process Hydro Pyro Development
FUEL FUEL Dissolution Dissolution
GANEX ELECTRO REDUCTIVE PUREX PUREX PUREX 1st Cycle REFINING EXTRACTION
Alternative Innovative 1 Cycle GANEX Waste DIAMEX Electrochemical SANEX SANEX 2nd Cycle Processes management
Advanced Partitioning Grouped Actinide Management SANEX Strategies Strategies Fuel Refabrication Studies
Other Process European Integration Projects …
Assessment of reprocessing flowsheets for different fuel cycle scenarios with a view to their future demonstration at the pilot level
Winter Seminar Demänovà, Slovakia 65 Gen IV Nuclear Reactor, 8-10 February 2010 The next generation of reprocessing facilities
Gen IV LWR & FR fuels > 2000 t/year New technologies
> 2040
Gen II UOx – LWR fuel Gen III 1000 t/year ∼ 2020 PUREX process MOX & UOx LWR fuels Higher throughput Integrated fuel processing and recycling : COEX Advanced separation technologies
1990
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 66 Gen IV Nuclear Reactor, 8-10 February 2010 Looking for a long term management strategy?
Ö Results already in use : - improved industrial facilities and processes - since 1991, significant waste volume reduction (by a factor of 6) - feasibility of disposal
Ö A continuous improvement process to be continued: - for opening the scope of possible solutions - and defining future electricity generating systems
Ufrom mine
Comb.Spent. Traitement et Ultimate - Minimising ultimate waste uséfuel waste Re-fabrication- FP - Optimal use of energy-bearing natural resources GEN IV NR
Actinides - Proliferation resistance
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 67 Gen IV Nuclear Reactor, 8-10 February 2010 General conclusion: in France, towards the 2012 milestone
¾ Recycling options, for sustainable FR systems
¾ Some options still open (what, and how), assess benefits/cost ratio by 2012 : a progressive step by step approach (from U and Pu first, Am to MAs recycling?)
¾ A need for flexible processes?
¾ - A specific new and important program on reprocessing modelling
¾ - A consolidation program for industrial potentiality by 2012
¾ - From separated MA solutions to Am and MA-bearing experimental fuels: to be tested at pin scale in the ASTRID SFR after 2020 …
And in ALLEGRO in Europe…
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 68 Gen IV Nuclear Reactor, 8-10 February 2010 The two mainstays of the future nuclear energy
3rd generation reactors RENAISSANCE avoid spent fuel with advanced recycling accumulation! proven technologies
Nuclear energy for the 21st century
SUSTAINABILITY 4th generation reactors with appropriate fuel cycle options R T FP
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 69 Gen IV Nuclear Reactor, 8-10 February 2010 Thank you for your attention
CEA / Nuclear Energy Division Winter Seminar Demänovà, Slovakia 70 Gen IV Nuclear Reactor, 8-10 February 2010