Immobilisation and Disposal Options for the Management of Separated Uranium

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Immobilisation and Disposal Options for the
Management of Separated Uranium
Joe Small, NNL Risley
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Background

• The UK has significant quantities of uranic
  materials, comprising
• uranium hexafluoride (UF6) resulting from
  isotopic enrichment of oxide fuels
• uranium trioxide (UO3) product from the
  reprocessing of used nuclear fuels and
• smaller quantities of separated uranium in other
  chemical forms
• The Nuclear Decommissioning Authority (NDA) has
  identified a range of options for the management
  of these materials
• interim storage prior to conditioning for direct
  disposal
• indefinite storage and
• reuse.
• Under the NDA Direct Research Portfolio the
  National Nuclear Laboratory (NNL) are developing
  a research programme to underpin the management
  strategy for uranium.

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The disposal option - talk outline

• Current chemical forms of UF6 UO3
• Radiological considerations
• nature of the hazard
• what type of facility/concept may be required?
• Chemical forms suitable for disposal
• conditioning and conversion processes
• Encapsulation/immobilisation
• cement
• polymers
• ceramic technologies
• Summary and future research

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Current chemical forms of uranium

• UF6 25,000 t U
• UO3
• Magnox reprocessing 30,000 t U
• THORP 5,000 t U
• 2,000 tU of other materials
• UF4, U3O8, UO2, U-metal, UC
• not considered in this presentation

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Uranium hexafluoride (hex tails)

• UF6 is the volatile chemical form used in the
  235U enrichment process
• tails are the residue depleted in 235U and 234U
• currently stored in hex cylinders
• UF6 reacts strongly with water yielding HF
• conversion to a more stable form is required for
  continued safe storage
• conversion of enriched 235UF6 to UO2 is integral
  to oxide fuel manufacture in the UK
• conversion processes of depleted tails to other
  chemical forms are established in the US and
  France.
• UO2, U3O8, UF4

http//www.nda.gov.uk/sites/capenhurst/
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Uranium trioxide from reprocessing

• In contrast to UF6, the UO3 products are specific
  to the UK nuclear fuel cycle.
• Magnox product (MDU) is depleted in 235U
• THORP product (TPU) is enriched in 235U and 234U
• Contain trace levels of fission/activation
  products and actinides from the original fuels.
• Dry UO3 powder is stored in 200l mild steel or
  stainless steel drums.
• UO3 powder has a tendency to hydrate, leading to
  expansion.
• Further conditioning of the UO3 powder will be
  required prior to disposal.

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Radiotoxicity of separated uranium(initially
1/10 that of natural uranium isotopic composition)
Natural Uranium
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Radiotoxicity of spent fuel and HLW(Initially
103 times that of natural uranium)
Geological Disposal Options for High-Level Waste
and Spent Fuel, Baldwin et al, 2008, for NDA-RWMD
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Radiological impact on the cementitious ILW
concept - groundwater pathway (based on GPA03)

• 238U inventory is 36 times that of GPA03
• But estimated increase in peak risk by 8e-7 y-1
• Reason
• separated uranium is depleted in 234U
• GPA03 inventory enriched in 234U
• Compared to the GPA03 the increase in risk from
  additional uranium is,
• comparable to the effect of removing the
  near-field solubility and sorption constraints,
• significantly lower than the effect of sorption
  in the geosphere, and
• less than that which results from a 10 reduction
  in the geosphere path length.
• Thus the additional inventory of uranium would
  likely be accommodated within the range of
  geosphere properties at suitable sites in the UK.
  

Calculated by scaling the inventory and reference
case results of the GPA03, (UK Nirex Report
N/080).
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Disposal concepts and chemical forms of uranium

• A geological disposal concept is envisaged as
  being appropriate for disposal of separated
  uranium
• The average activity is 26 GBq/te and would be
  classified as ILW (gt 4 GBq/te)
• Given the long half lives of uranium and the
  increase in radiotoxicity over time geological
  containment will be necessary to isolate the
  uranium from human intrusion.
• Over such long periods of time (gt105 y) less
  reliance can be placed on engineered barriers
  (e.g. containers, backfill) in minimising uranium
  and daughters mobility in groundwater.
• The geological barrier is thus of primary
  importance.
• The chemical form of the uranium should be
  compatible with the host rock geology/geochemistry
  as well as with any backfill that is used.

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Short-listed chemical forms suitable for disposal

• Generally, naturally occurring mineral phases
• UO2 - reducing conditions
• U3O8 - reducing conditions
• hydrated UO3 (schoepite) oxidising conditions,
  high solubility ?
• Calcium uranate (CaUO4) stable phase under
  cementitious conditions
• Uranium silicates reducing and oxidising
  conditions, neutral pH
• Uranium phosphates reducing and oxidising
  conditions, neutral pH
• Ceramic (Synroc) phases low solubility and low
  dissolution rates
• Conversions routes from UF6/UO3
• Dry routes oxides, CaUO4, phosphates?
• Wet routes phosphates, silicates

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Encapsulation in cement and polymers

• Technologies to manufacture Depleted Uranium (DU)
  shielding materials are established mainly from
  US programmes
• Consider UF6 deconversion products UO2, UF4
• Potential for reuse in spent fuel storage and
  disposal
• Cement technologies better established and
  probably more compatible with geological disposal
• Experimental research required to examine
  processing of UK specific uranium e.g.
• Settling of U - requirement for superplasticisers
• Reactions during curing UO3 gt CaUO4.
• Chemical pre-treatments

DUAGG product Sintered UO2 with basalt binder
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Use of ceramic technologies

• Research has been established from the NDA Pu
  disposition programme on the use of ceramics
• U used as a chemical surrogate for Pu in trials
• A range of fully crystalline and mixed
  glass-ceramics developed e.g. to immobilise Pu
  and fluoride containing residues
• Ceramics may have uses for enriched stocks
  including HEU.
• Hot Isostatic Pressing (HIP) technologies have
  wider application to producing uranium wasteforms
• consolidate powders at relatively low
  temperatures
• combined chemical processing
• contain fission product contamination
• large volume feasible

Glass-ceramic wasteforms
HIP products
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Volume considerations

• The packaged volume of separated uranium will
  depend on the density of the de-converted/immobili
  sed product
• Criticality issue for enriched batches might
  also limit package loading, unless these
  materials are blended
• For the existing ILW concept the weight limits of
  standard packages constrains the volume
• In this case packaged volume would increase by
  12 (from 241,000 m3 )
• For higher densities (7 g/cm3) and increased
  weight limits volume increase is 4
• Compared to the HLW/Spent fuel concept
• For the highest likely density (e.g. UO2) the
  volume of uranium would be equivalent to the
  total volume of HLW/SF canisters.

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Consideration of disposal concepts in defining
research on uranium immobilisation

• Given the volume of separated uranium and the
  long-term geological containment required some
  alternatives to the current ILW and HLW/SF
  concepts are apparent.
• At this stage of research it is appropriate to
  bear these in mind.
• Co-location of uranium specific concepts
• shallower disposal based on the low solubility of
  some uranium minerals at neutral pH under a range
  of redox conditions,
• or deeper concepts more reliant on geosphere
  retardation.
• Variants of the current ILW GDF design and the
  developing HLW/SF concept may also be considered
• increased weight limits of ILW packages
• alternative backfills (rock?)
• closer spacing of HLW/SF packages
• less durable canisters/containers.

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Summary

• To support NDA strategy a research programme has
  been initiated considering the disposal of
  separated uranium in the case that these
  materials are declared as wastes.
• Research has been identified to further examine
  the chemical conversion processes and
  immobilisation technologies required to produce
  uranic wasteforms suitable for geological
  disposal.
• Due to the very long half lives of the uranium,
  disposal concepts will be reliant on geological
  containment more so than containers and
  engineered barriers.
• An initial wide range of candidate chemical forms
  have been identified, which may be stable under a
  range of geological conditions.
• In developing this research there is scope to
  optimise the design of the wasteform and the
  disposal concept with a specific geological site.

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Acknowledgements

• NDA
• Paul Gilchrist
• Colin Rhodes
• NNL
• Conversion Processes - Duncan Coppersthwaite
• Cement Polymer encapsulation - Ed Butcher
• Ceramics - Charlie Scales
• Disposal - Alan Wareing, Andras Paksy, Candy
  Lean, Helen Steele
• The views expressed here are those of the NNL
  team and do not necessarily represent those of
  the NDA