R&D for Fast Reactor Fuel Cycle at IGCAR

P.R.Vasudeva RaoDirector

Chemistry, Metallurgy and Materials Groups

Indira Gandhi Centre for Atomic ResearchKalpakkam, India

Asian nuclear prospect 2008

Strategies for Fuel Cycle Strategies for Fuel Cycle

Fuel cycle facilities to cater to two or more reactors: economy of scale

Enhanced burn-up (200,000 MWd/t) : lower quantities of fuel to be processed per MW of electricity

Minimum cooling time, matching fuel handling intervals of reactor: reduced out-of-pile inventory, better utilisation of valuable fissile material

Reduced waste production through optimisation of flow sheets for fabrication and reprocessing

Recovery of minor actinides and “valuable” fission products: reduced waste volume, with added societal benefits

Uranium Plutonium Carbide Fuel for FBTRMark I Fuel: (U0.3, Pu0.7)CMark II Fuel: U0.45Pu0.55C

The Mark I fuel has reached 155 GWd/t burn-up without any fuel pin failure

Carbide Fuel Cycle of FBTR

Fuel fabricated at BARC

Comprehensive Post-Irradiation Examination was carried out at various burn-up values to understand the fuel behaviour and obtain safety clearances for enhancing burn-up

Fuel discharged from FBTR with burn-up up to 150 GWd/t has been reprocessed, for closing the fuel cycle

Thermal conductivity

Laser VaporisationMass spectrometry

Capacity to develop, ab-initio, new fuel concepts and closing the fuel cycle with benchmark performance indices

R & D on Carbide Fuel CycleThermophysical properties of high Pu content

fuel (first measurements)• Thermal conductivity, Carbon potential

Thermochemical Modeling: • Oxygen limit to reduce Pu volatility during

fabrication • Carbon potential, Pu segregation, CO pressure

Post-irradiation examination in inert atmosphere hot cells: destructive and non-destructive

Reprocessing flow sheet: dissolution behaviour, third phase formation

PIE Facilities for Performance Assessment of FBTR Fuel

REMOTE METALLOGRAPHY

U 233 FUELLED REACTOR FOR N-RADIOGRAPHY

LASER DISMANTLING OF IRRADIATED FSA

METROLOGY & NDT EQUIPMENTS

SMALL SPECIMEN TEST EQUIPMENT (BI)

HOT CELL FACILITY (INERT ATMOSPHERE)

REMOTE CNC MACHINE

FISSION GAS EXTRACTION & ANALYSIS

HIGH TEMPERATURE REMOTE TENSILE TEST MACHINE

Comprehensive R & D in Fast reactor Fuel Reprocessing Comprehensive R & D in Fast reactor Fuel Reprocessing

Extractants for fuel reprocessing

Extractants for MA recovery

Extractants / resins for recovery of valuable fission products

Development of novel contactors, pumps and metering equipment

Extensive use of Modeling and Simulation for extraction processes and equipment performance

Materials development for longer for Purex process plant life

Fast reactor Fuel Reprocessing : Extractants

Extractants for Purex processHigher homologues of Tributyl Phosphate (like Tri-iso-Amyl Phosphate)Long chain dialkyl amides (DOHA etc.)

Extractants for MA recoveryCMPOAmides

Extractants / resins for recovery of selected fission products

16-stage ejector mixer-settler for flow sheet

development

Alternate Trialkyl phosphates for fast reactor fuel reprocessing

• A variety of trialkylphosphates synthesised and comprehensively characterised ( extraction behaviour, radiation and chemical degradation)

• Bulk synthesis of Tri-n-amyl phosphate

• Mixer settler runs under progress

O

O

O

O

O

RO- Group

O

O

O

O

RO- Group

Triamyl phosphate: a candidate for fast reactor fuel reprocessing

Room temperature ionic liquids for actinide and fission product recovery

• Exploiting unique properties of RTILs to reduce waste generation (Ex-El process)

• Use of RTILs in place of high temperature molten salts for electrorefining of metal fuels

• Use of RTILs for treatment of waste- eg. Recovery from waste

SIMPSEX (SIMulation Program for Solvent EXtraction) : nuclear SX of U, Pu and Nitric acid in FBR fuel reprocessing flowsheets.

PUThEX : Uranium separation from Th

SIMPACTR: Code for simulation of actinide recovery.

SIMPUREXAE: Code for PUREX process with Alternate Extractant

Extensive validation of Computer Codes by from experimental data in the literature

Solvent Extraction Modeling

Strip 1 (0.01 N Acid),

Base Flow

Strip 2 (4N Acid),Base Flow

Max

imum

PuC

onc.

g/L

Variation of Max. Pu Concentration

in the HC contactor

Waste Management

• Simplification of process routes for reducing waste volumes

• Development of processes for recovery of minor actinides and fission products (eg. Pd, Cs) from HLW

• Use of “green” processes such as SFE for waste treatment

• Development of glass and ceramic matrices for immobilisation

Partitioning of actinides from HLWSynthesis and characterisation of CMPO & DMDOHEMA:

• Purification method developed for CMPO• Bulk synthesis achieved• Extraction studies & Third phase formation studies

carried out• Mixer settler experiments with CMPO and TODGA in

process• Experiments with HLW from FBTR fuel reprocessing

by Jan 2009

Separation of An and Ln:

• 1,6-Bis triazynyl pyridine syntheised and characterised• Chromatographic studies on Am/Ln separations in

process

Recovery of Pu

SFE Facility in Glove Box

Supercritical Fluid Extraction (SFE)

.

0 15 30 45 60 75 900

20

40

60

80

100

PuR

ecov

ery

(%)

Extraction Period (min)

Pu(III)

SC-CO2+CMPO for extrn.

Lab scale SFE facility established in glove boxLab scale SFE facility established in glove box

Quantitative extraction of Quantitative extraction of U,PuU,Pu and Am from and Am from

tissue waste demonstratedtissue waste demonstrated

Extraction of silicone oil from fuel Extraction of silicone oil from fuel microspheresmicrospheres

(produced in sol(produced in sol--gel process) establishedgel process) established

SolventSolvent--free route developed for SFE using solid free route developed for SFE using solid

extractantsextractants

Recovery of residual actinides and fission productsRecovery of residual actinides and fission products

from salt waste of from salt waste of pyrochemicalpyrochemical processprocess

Iron Phosphate Glass for Fast Reactor Waste

• High loading (20 wt %) of waste demonstrated using simulated HLW

• Low volatility of Cesium • Low leaching rate• “Difficult” elements such as Pd don’t

segregate• Conditions for glass formation &

physical properties of glass : better than BSG

1 0 2 0 3 0 4 0 5 0 6 0 7 0 8 0

I P F W

F ig . 1 X R D p a t t e r n o f t h e I P G a n d I P F W

I P G

Arb.

uni

ts

2 Θ

4 0 0 6 0 0 8 0 0 1 0 0 0 1 2 0 0 1 4 0 0

- 2 5

- 2 0

- 1 5

- 1 0

- 5

0

5

1 0

I P F W

∆ T

(µV

)

T / K8 5

9 0

9 5

1 0 0

D T A T G A

m e l t in gt e m p e r a t u r e

C r y s t a l l i s a t i o n t e m p e r a t u r e

T g

F ig . 2 T G A / D T A c u r v e o f I P G a n d I P F W

0 50 100 150 200 25099.2

99.4

99.6

99.8

100.0

Fig. 8 Weight (%) of the IP5C5 glass as a function of time (min.)

Wei

ght (

%)

time / min

Cans with simulated synroc after hot isostaticpressing

High level waste immobilisation

Ceramic waste matrices development- 100g size monoliths fabricated and characterised

Studies show that Synroc C is efficient for immobilisinghigh level waste

Bulk synthesis (kg) & fabrication of Synroc monoliths containing simulated HLW expected from FBTR under progress

Palladium recovery from HLW

• Polyvinyl pyridine based resin developed indigenously

• Palladium quantitatively extracted from nitric acid medium

• RTIL based extraction-electrodeposition process

Studies also under progress for recovery of Cs and Sr from waste

# Facility for fabrication of test fuel pins through sol-gel / sphere-pac route

# First test pins of MOX fuel to be introduced in FBTR in Dec. 2008

# Sphere-pac pin with two fractions, (770 µm MOX microspheres and 115 µm UO2 microspheres)

# Coarse fraction fabricated using silicone oil column; Fine fraction fabricated by jet entrainment method

# Test pins with minor actinides will be fabricated for irradiation in FBTR

Sol-gel Vibrocompaction

Vibrocompaction and welding set up of the test fuel pin fabrication facility

Microsphere filling & vibrocompaction station

Hardware loading stationWelding station

Metallic FuelMetallic Fuel

1.Fuel Fabrication including sodium bonding

2. Studies on fuel properties

3. Fuel Reprocessing by pyrochemical routes

4. Irradiation programme including recycled fuels

5. Development of technologyfor waste arising out ofpyrochemical processes

Areas of R & D

# Development of sodium bonded , injection cast, U Pu-6 %Zr ternary alloy

# Development of Mechanically bonded, U-Pu binary alloy fuel, clad with and without coating

# Measurement of thermophysicalproperties, (Cp, K)

# Test fuel pin irradiations from FBTR –2009

# Facility for fabrication of metal fuel pins on a regular basis for FBTR

# Design of fuel initiated at IGCAR

Metallic fuel Development

R & D for R & D for PyrochemicalPyrochemical ReprocessingReprocessingMolten salt electrorefining: modeling as well as experimentation being pursued

Development of corrosion resistant materials and coatings

Equipment development (electrorefiner, consolidation system..)

Molten salt loops for long term corrosion studies on materials

Studies on room temperature ionic liquids as alternates for molten salts

-1.2 -0.8 -0.4 0.0 0.4

-50.0

-25.0

0.0

25.0

50.0

Zr2+/Zr

Ia2

Zr/Zr2+

798 K

Ic2

Curr

ent (

mA)

Potential (V) vs Ag(I)/Ag

Cyclic Voltammograms of Zr ions in LiCl-KCl at 798 K, in presence of Zrmetal, show only Zr2+ ions ; Zr4+

ions are absent

Unexposed 1000 hrs

Partially stabilised Zirconiacoatings on steel after 1000 h exposure to salt

10 20 30 40 50 60 70

XRD of glass-bonded sodalitewasteform

THANK YOU

If I have a thousand ideas and only one turns out to be good, I am satisfied.

Alfred Nobel: born Oct. 21, 1833