ZirconiaZirconia inert matrix fuels for inert matrix fuels for plutonium and minor actinides plutonium and minor actinides burning in reactorsburning in reactors

Claude Degueldre

Paul Scherrer Institute, Villigen, Switzerland& Uni Geneva, Switzerland

IAEA TECHNICAL MEETING“ADVANCED FUEL PELLET MATERIALS AND FUEL ROD

DESIGNS FOR WATER COOLED REACTORS“

23-26 Nov. 2009, PSI, Villigen, Switzerland

IMF Goal and Objective• Desired goal: eliminate

minor actinides and Pu excesses• Desired objective: use them to

produce energy in reactors• Opt for an economical

ecological safe andsustainable solution

economicalecological safesustainable

R&D work required• Material selection for fast or thermal reactors• Neutronics optimisation• System studies

homogeneity vs heterogeneity for: fuel material and fissile at the FUEL ASSEMBLY CORE level

• Inventory after high burn-up• Dispositions: multirecycling leaching solubility or

once-through with low solubility for geodisposal

Degueldre & Paratte, Concepts for an inert matrix fuel, an overview, J. Nucl. Mater. 274 (1999) 1-6

Examples of Inert Matrices to replace 238UO2

Inert Matrix type Inert Matrix formulaElement C, Mg, Al, Si, Cr,V, Zr, Mo, W

Inter-metallics AlSi, AlZr, ZrSiAlloy Stainless steel, zirconium alloys

Carbide 11B4C, SiC, TiC, ZrCNitrides AlN, TiN, ZrN, CeN,

Binary oxide MgO, Y2O3, ZrO2, CeO2

Ternary oxide MgAl2O4, Y3Al5O12, ZrSiO4

Oxide solid solution YyZr1-yO2-y/2, Mg(1-x)Al(2+x)O(4-x)

Examples of Inert Matrix design and additives

Design CompositionSolid solution AnzYyPuxZr1-yO2-ψ*

Cercer MgAl2O4 - YyPuxZr1-yO2-y/2*

Cermet Zr - YyPuxZr1-y-xO2-y/2*

Metmet PuAl4*-Al

Additive FormulaBurnable poison B, Gd, Dy, Ho, Er, Eu, Np, Am

Resonance additive W, Th, U

Stabiliser Y2O3, CaO in ZrO2

Concept suggested at PSI for Pu-IMF utilization in LWR

C. Degueldre, Zirconia inert matrix for plutonium utilisation and minor actinides disposition in reactors, J. Alloys Compounds, 444-445 (2007), 36-41

. The three levels for IMF utilization in light-water reactors considering homogeneous vs. heterogeneous systems at the fuel, assembly, and core levels.

The fuel is either a solid solution ceramic homogeneously doped with plutonium (red) or heterogeneously doped with some uranium (green), or is a composite material with particulates or microspheres (again plutonium-doped red, uranium-doped green) imbedded in inert matrix material.

The fuel assemblies themselves may be homogeneous (all fuel rods in a given assembly contain IMF, red) or heterogeneous (red IMF rods distributed among green UO2 fuel—e.g. the French Advanced Plutonium Assembly (APA) concept).

The reactor core may also be loaded homogeneously (with red IMF assemblies), or the UO2 core may be partially loaded with some IMF assemblies forming a heterogeneous core loading.

ZirconiaZirconia IMF concept @PSIIMF concept @PSI• With IMF no new Pu is produced (contrary to MOX)

• (Er,Y,Pu,Zr)O2-x solid solution with fissile & burnable poison.

• The plutonium in spent IMF foreseen for disposal is devaluated.

• For the material qualification, the relevant fuel properties are: - fissile and component densities, - chemical stability and inertness ,- porosity, micro/nano structural studies, - thermal conductivity, - stability under irradiation, - efficient retention of fission products and - solubility: a key property for the disposal of the spent fuel.

Cooperative work fissile and component densities

• Basic neutronicproperties calculations

• Comparison code and library

• Neutron physics benchmark: progress & <2% consistency forEr2O3-PuO2-ZrO2

• To be made with An• Ref. proceedings of

IMF3, ENEA 1997

Zirconia IM conceptchemical stability

• ZrO2-MgO• ZrO2-CaO• ZrO2-SrO

• ZrO2-ScO1.5

• ZrO2-YO1.5

• ZrO2-LaO1.5

• ZrO2-TiO2

• ZrO2-HfO2

• ZrO2-CeO2

ZrO2 – LnO1.5 phase diagrams

Effect of additives on the SS

porosity, micro/nano structural studies(Er,Y,Am,Pu,Zr)O(Er,Y,Am,Pu,Zr)O22--xx IMF IMF density/porositydensity/porosity

• Ceramography• X-ray Tomo• ND• XRD

• IMF data:- Density- Porosity- Fissile densityAlso to modelthermal conductivity

Small Angle Neutron Scattering

Use of SANS for investigating nano-porosity in simulated inert matrix fuel.

Samples of (Er0.05Y0.10Ce0.10Zr0.75)O2-xwith porosity 8%

Investigations for 2 detector positions

Investigation with 2 sample thicknesses to track multiscattering.

Detection of pores around 300 nm ↑

Data to be complemented by ceramography or tomographicinvestigations.

Simulated IMF

Simulated Inert Matrix FuelSimulated Inert Matrix Fuel

• Picture of a zirconia 1 x 1 x 10 mm prism of Er0.05Y0.10Ce0.10Zr0.75O2-x, tomogramme, recorded with 1024x1024 CCD camera, image obtained after filtering, scale in μm per pixel, work performed at the ESRF on the High Energy Beamline ID15.

• Visualisation of pores and cracks in this dense matrixdown to 1 μm

• Porosity studies.

Degueldre, Pouchon, Streit, Zaharko, Di Michiel, Analysis of porous features in zirconia based inert matrix, impact on the material qualification, Prog. Nucl. Energy, 38 (2001) 241-246

Analoge and real solid solutionBurghartz, Ledergerber, Ingold, Heimgartner, Degueldre X-ray diffraction analysis and data interpretation of stabilized zirconia inert matrix doped with plutonium, Prog. Nucl.Energy, 38(2001) 241-246

Analoge solid solution

Degueldre, Pouchon, Streit, Zaharko, Di Michiel, Analysis of porous features in zirconia based inert matrix, impact on the material qualification, Prog. Nucl. Energy, 38(2001) 241-246

Bulk density studies

Simulated and U doped Inert matrix fuelSimulated and U doped Inert matrix fuel• Rdfs in Er0.05Y0.10U/Ce0.10Zr0.75O2-x,

showing element specific changes in local environments resulting from substituting U for Ce. The larger size of U(IV) relative to Ce(IV) results in an expansion of the second and third nnfor Zr and Er, whose rdfs are otherwise very similar in both the Ceand U compounds. The substantial differences in the Y environment, including splitting of the O shell at 2.3 Å and increased separation of the more distant nn O set, and in Ce/U environment including the splitting of the second shell cations, imply strong interaction between the Y and U sites that results in preferential Y-U second nn aggregation.

• P. Villella, S. Conradson, F. Espinosa, S. Foltyn, K. Sickafus, J. Valdez, C. Degueldre. Phys. Rev. B 4101 (2001) 58-68.

Y-10 Er-5 Ce-10Y-10 Er-5 U-10

Longer secondshell distance in

U compound

Longerthird

shell Odistance in

U compound

2.0 4.0

Zr rd

f

Zr

R (Å)

Substantial differencesin all shells, especiallyO shells, between Ce-

and U-compounds

2.0 4.0

Y rd

f

Y

R (Å)

Shifts to longerdistances in U compound,

otherwise very similar

2.0 4.0

Er rd

fEr

R (Å)

U-nn O distances longer than Ce;Second shell totally different

2.0 4.0

Ce/

U rd

f

Ce/U

R (Å)

Simulated Inert matrix fuelSimulated Inert matrix fuel

• Comparing XAFS & XRD

• for the 1st shell average∑ξ(rMi+rO) compared to a√3/4 within ± 1.6%,

• for the 2nd shell average∑ξ(rMi+rMj) compared to a√2/2 within ± 2.5%

• Degueldre & Conradson. Appl. Phys. A 73 (2001) 489-494.

Conradson, Degueldre, Espinosa, Foltyn, Sickafus, Valdez, Villella. Prog. Nucl. Energy. 38 (2001) 221-230.

Thermal conductivity (model) of inert matrix fuelThermal conductivity (model) of inert matrix fuel

Er0.05Y0.10Pu/Ce0.10Zr0.75O2-x,thermal conductivity was evaluated using Klemenstheory accounting the effect of photonic scattering centers. The hyperbolic thermal conductivity trend with temperature for pure zirconiais reduced by - isotopes, impurities, - dopants, oxygen vacancies,- defects and porous features.

C. Degueldre, T. Arima, Y.W. Lee, Thermal conductivity of zirconiabased inert matrix fuel: use and abuse of the formal models for testing new experimental data, J. Nucl. Mater. 319 (2003) 6-14.

Thermal conductivity (exp.) of inert matrix fuelThermal conductivity (exp.) of inert matrix fuel

Er0.05Y0.10Pu/Ce0.10Zr0.75O2-x,thermal conductivity was measured The hyperbolic thermal conductivity trend with temperature for pure zirconiais reduced by - isotopes, impurities, - dopants, oxygen vacancies,- defects and porous features.

C. Degueldre, T. Arima, Y.W. Lee, Thermal conductivity of zirconiabased inert matrix fuel: use and abuse of the formal models for testing new experimental data, J. Nucl. Mater. 319 (2003) 6-14.

Degueldre, Pouchon, Döbeli,. Sickafus, Hojou, Ledergerber, Abolhassani-Dadras, Behaviour of implanted xenon in yttria-stabilised zirconia as inert matrix of a nuclear fuel, J. Nucl. Mater. 289 (2001) 115-121

Pouchon, Degueldre, Döbeli, Enhanced retention of cesium in yttria stabilised zirconia by co-implantation of iodine, Progr. Nucl. Energy 38 (2001) 275-276

Efficient

retention of

fission products

IMF R&D work in Switzerland

• At PSI: IMF since 1995• ZrO2 & ZrN material

preparation and testing• Neutronics calculation

for LWR• Disposition• Organising

IMF workshops

COOPERATIONPSI, EPFL, ETHZUni Geneva Fabrication of Inert-Matrix-Fuel (IMF)

for plutonium incineration; irradiation tests in the Halden Research Reactor (N)

ZrO2 IMF R&D work in Switzerland

• Integral measurements with a plutonium inert matrix fuel rod in a heterogeneous light water reactor lattice

• Sectional view of a SVEA-96 BWR fuel assembly indicating the voided moderator conditions simulated in LWR-PROTEUS Core 3A

• No major uncertainties in pin power distribution predictions, neither for the MOX nor for the Pu-Er-Zroxide IMF investigated

• ○UO2 ●Mesur. • ⊗ γ scan BPd pin

Neutron radiography of two IMF segments during the OTTO experiment

IMF composition: (Er,Y,Pu,Zr)O2-x Capsule 1 and (Y,Pu,U,Zr)O2-x Capsule 2.

Pellet diameter 8.00 mm, stack length 67.0 and 67.7 mm, density of plutonium fissile at beginning of life: 0.37 and 0.34 g cm-3, density 5.80 and 6.02 g cm-3 respectively. This image was obtained after one cycle.

C. Degueldre, Ch. Hellwig, Study of a zirconia based inert matrix fuel under irradiation,J. Nucl. Mater. 320, (2003) 96-105

Central fuel temperature of IMF&MOX @ Halden RP

Central fuel temperature and FGR of IMF&MOX @ Halden RP

Ledergerber, Degueldre, Heimgartner, Pouchon, Kasemeyer, Inert matrix fuel for the utilisation of plutonium, Progr. Nucl. Energy, 38 (2001) 301-308

Plutonium inventory as a function of IMF irradiation time in PWR

0 200 400 600 800 1000 1200 14000

10

20

30

40

50

60

70

80

90

100

0

10

20

30

40

50

60

70

80

90

100

Pu-240 + 242

Pu-239 + 241

Standard MOX

Uranium free matrix

Plut

oniu

m in

vent

ory

[%, o

f ini

tial l

oadi

ng]

Full power days

Berthou, Degueldre, Magill, Transmutation characteristics in thermal and fast neutron spectra: application to americium J. Nucl. Mater., 320 (2003) 156-162

SolubilitySolubility

Glass 10-3 M

Sapphire10-7 M

Zirconia10-10 M

SolubilitySolubility

Zirconiasolubility: <10-10 M for pH 3.0 - 8.5

Data from:■ Kovalenko et al., Russ. J. Inorg. Chem. 6 (1961) 272.● Adair et al., Ceram. Trans 1 (1987) 135.▼ Pouchon et al., Progr. Nucl. Energy 38 (2001) 443.▲ Egberg et al., J. Sol. Chem., 133 (2004) 47.♦ Michel, 2005 These University of Nantes, (2005).

R&D R&D forfor spentspent zirconiazirconia IMF IMF dispositiondisposition

• Zirconia IMF as a Nonproliferationbarrier

• Material doped with fissile and low solubility or leaching rate

-11

-10

-9

-8

-7

-6

-5

-4

-3

-2

-1

-6 -5 -4 -3 -2 -1 0log [CO3] tot

log

[Zr]

tot

model, log beta0(1,5) = 43

model, log beta0(1,4) = 42

oxy-hydroxide, maxima

m-ZrO2, 250 days reaction time

Fig. 6. Solubility of monoclinic zirconia and oxy-hydroxide precipitate as a function of total carbonateconcentration: comparison of experimental data with computed solubility curves. []: concentration inmole per litre (M); beta0(1,I) stand for the stability constant of the complex Zr(CO3)i

(4-2i). Note the porewater around the waste package is expected to reach a log [CO3]tot between -3 and –2 corresponding toa solubility ranging from 10-9 to 10-8 M.

M. A. Pouchon, E. Curti, C. Degueldre, L. Tobler, The influence of carbonate complexes on the solubility of zirconia: new experimental data, Prog. Nucl. Energy, 38,, (2001), 443-446

SEM comparison of secondary (a,c) and back scattered (b,d) SEM comparison of secondary (a,c) and back scattered (b,d) electron images electron images [Lumpkin, 1999][Lumpkin, 1999]..

Conditions: intergrowth between uranpyrochlore and baddeleyite from Jacupiranga carbonatitecomplex, Brazil. General features and microfracturing, alteration in the uranpyrochlore (lighter gray, (U,Ca,Ce)2(Nb.Ta)2O6(OH,F) and the baddeleyite (darker gray, ZrO2) 500 & 200 μm width.

Very high resistance to proliferation risksVery high resistance to proliferation risks

• Low leaching potential in acids

(HNO3 or HF) even under pressure and elevated temperature or with mixtures of acids.

• High mechanical resistance

crushing larger amounts of (Y,Zr)O2-xfor dissolution is very difficult.

10th IMF workshop, 1-2 Jun. 2005 EMRS-Strasbourg

• Comparing material in research reactors

• Testing new datalibraries e.g. for MA

• Characterisingmaterials (IMF PIE)

• Discussing fuel designs• J.Nucl.Mater. 352, 3 (2006)

11th IMF workshop, 10-12 Oct. 2006 INL, Park City

• Promoting MgO-ZrO2 material in LWV• Benchmark for MA as BP for IMF in LWR• Characterising materials (IMF PIE)• Discussing fuel utilisation

o What is role of IMF in future advanced nuclear scenarios with potential mix of nuclear systems which may include LWRs, FRs, GCR, ADS?

o Fuel vs. targetso Must take into consideration uranium resources (conversion ratio, re-

cycle, etc.) o Address remaining TRUo New vs. existing reactorso Impact on performance and safety depends on reactor characteristics

• Production of a CD report (2006)

Status of the IMF initiative in 2009

1995 - 2009 11 Meetings : CH3, It, Fr, EC, Nl, Jp, UK, EC.450 Participants, 15 countries (Aus, Ar, B, Cd, CH, Cz, D, Fr, I, In, Is, Jp, Kr, Nl, Ru,

S, UK, US), and 3 international organisations (EC, IAEA, OECD)Universities: Michigan, POLIMI, Aachen, Ben Gourion, Delft,

Geneva, Lausanne, New Mexico, Ontario, Osaka, Paris, Purdue, Darmstadt, Florida, Kyushu.

Nat. Lab. : AECL, ANSTO, CEA, CNRS, ENEA, ITU, JAERI, KAERI, IPPE, FZJ, LANL, NRG, ORNL, PNL, PSI, VNIINM, ANL, INL…..

Industry: BNFL, COGEMA, FRAMATOME-ANP, NRG, SKODA,137 Papers published in: J. Nucl. Mater. (1999, 2003 & 2006)

Prog. Nucl. Energy (2001).120 Communications : 7 internal reports

(ENEA,CEA,NRG,BNFL,INL)

Todays IMF programmes and perspectives

Results gained from the IMF programmes :

• Concepts & optimised conditions for IMF utilisation• Possible IMF solutions, together with their scope

(cycle length, safety parameters, etc.) will be defined • Zirconia IMF may be fabricated and used in reactor • Spent fuel dispositions are evaluated and selected

Once through then geodisposal• A basis of knowledge on inert matrix fuels will be

established to allow deployment in reactor

IMF12-workshop in 2012?