Fuel Cycle Chemistry

1
Fuel Cycle Chemistry

• Chemistry in the fuel cycle
• Uranium
• Separation
• Fluorination
• Fission products
• Advanced Fuel Cycle
• Fuel development
• Separations
• Environmental behavior
• Waste forms
• Focus on chemistry and radiochemistry in the fuel
  cycle

2
Reactor basics

• Utilization of fission process to create heat
• Heat used to turn turbine and produce electricity
• Requires fissile isotopes
• 233U, 235U, 239Pu
• Need in sufficient concentration and geometry
• 233U and 239Pu can be created in neutron flux
• 235U in nature
• Need isotope enrichment

induced fission cross section for 235U and 238U
as function of the neutron energy.
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Uranium chemistry

• Separation and enrichment of U
• Uranium separation from ore
• Solvent extraction
• Ion exchange
• Separation of uranium isotopes
• Gas centrifuge
• Laser

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Mining

• Uranium is reduced (tetravalent)
• Introduction of oxygen produces hexavalent
  uranium
• 222Rn daughter
• Ore mining or solution mining
• solution mining uses injection of sulfuric acid
  to dissolve U and solution is removed
• not all solution is removed
• minerals are solubilized
• seepage into aquifer (Dresden, Sachsen)

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Acid-Leach Process for U Milling
U ore
Water
Crushing Grinding
Slurry
H2SO4 Steam NaClO3
40-60C
Acid Leaching
Separation
Tailings
Organic Solvent
Solvent Extraction
NH4
Recovery, Precipitation
Drying (U3O8)
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Uranium Ore
http//www.cogema.fr/photos_gb/
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Yellowcake production
http//www.cogema.fr/photos_gb/
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U Fluorination
U ore concentrates
HNO3
Solvent extraction purification
Conversion to UO3
H2 Reduction
UO2
HF
UF4
F2
U metal
Mg
UF6
MgF2
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Fuel Fabrication
Enriched UF6
Calcination, Reduction
UO2
Pellet Control
40-60C
Tubes
Fuel Fabrication
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Aside Fluorination of UO2 by NH4HF2

• Degradation of TRISO Fuel and Kernel Matrices
  using Ammonium Bifluoride
• Chemical treatment of TRISO
• Concept Fluorination of graphite, SiC and
  actinide kernel by NH4HF2
• Solid-solid reactions have been observed between
  ammonium bifluoride and oxides of vanadium,
  zirconium, thorium, uranium, and plutonium
• Reaction with uranium dioxide at 25 C
• UO2(s) 4 NH4HF2(s) ? (NH4)4UF82H2O(s)

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Fluorination of UO2 Ball Mill at 25 C

• UO2 4 NH4HF2 ? (NH4)4UF82H2O
• 50 g finely-divided (30 mm) UO2 and 10 excess
  NH4HF2
• 20 minutes in a ball mill

?
13
U enrichment

• Utilizes gas phase UF6
• Gaseous diffusion
• lighter molecules have a higher velocity at same
  energy
• Ek1/2 mv2
• For 235UF6 and 238UF6
• 235UF6 impacts barrier more often

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Gas centrifuge

• Centrifuge pushed heavier 238UF6 against wall
  with center having more 235UF6
• Heavier gas collected near top
• Enriched UF6 converted into UO2
• UF6(g) 2H2O?UO2F2 4HF
• Ammonium hydroxide is added to the uranyl
  fluoride solution to precipitate ammonium
  diuranate
• 2UO2F2 6NH4OH ? (NH4)2U2O7 NH4F 3 H2O
• Calcined in air to produce U3O8 and heated with
  hydrogen to make UO2

15
Fission Product Chemistry

• Chemistry dictated by oxidation state
• Importance of isotopes related to half life
• Shorter half-lives important in reactor
  maintenance
• Longer lived isotopes important waste treatment
  and disposal
• Dose and heat

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Radionuclide Inventories

• Fission Products
• generally short lived (except 135Cs, 129I)
• ß,??emitters
• geochemical behavior varies
• Activation Products
• Formed by neutron capture (60Co)
• ß,??emitters
• Lighter than fission products
• can include some environmentally important
  elements (C,N)
• Actinides
• alpha emitters, long lived

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Fission products

• Kr, Xe
• Inert gases
• Xe has high neutron capture cross section
• Lanthanides
• Similar to Am and Cm chemistry
• High neutron capture cross sections
• Tc
• Redox state (Tc4, TcO4-)
• I
• Anionic
• 129I long lived isotope

20
Cesium and Strontium

• High yield from fission
• Both beta
• Some half-lives similar
• Similar chemistry
• Limited oxidation states
• Complexation
• Reactions
• Can be separated or treated together

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Alkali Elements

• 1st group of elements
• Li, Na, K, Rb, Cs
• Single s electron outside noble gas core
• Chemistry dictated by 1 cation
• no other cations known or expected
• Most bonding is ionic in nature
• Charge, not sharing of electron
• For elemental series the following decrease
• melting of metals
• salt lattice energy
• hydrated radii and hydration energy
• ease of carbonate decomposition

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Solubility

• Group 1 metal ions soluble in some non-aqueous
  phases
• Liquid ammonia
• Aqueous electron
• very high mobility
• Amines
• Tetrahydrofuran
• Ethylene glycol dimethyl ether
• Diethyl ether with cyclic polyethers

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Complexes

• Group 1 metal ions form oxides
• M2O, MOH
• Cs forms higher ordered chloride complexes
• Cs perchlorate insoluble in water
• Tetraphenylborate complexes of Cs are insoluble
• Degradation of ligand occurs
• Forms complexes with ß-diketones
• Crown ethers complex Cs
• Cobalthexamine can be used to extract Cs
• Zeolites complex group 1 metals
• In environment, clay minerals complex group 1
  metal ions

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Group 2 Elements

• 2nd group of elements
• Be, Mg, Ca, Sr, Ba, Ra
• Two s electron outside noble gas core
• Chemistry dictated by 2 cation
• no other cations known or expected
• Most bonding is ionic in nature
• Charge, not sharing of electron
• For elemental series the following decrease
• melting of metals
• Mg is the lowest
• ease of carbonate decomposition
• Charge/ionic radius ratio

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Complexes

• Group 2 metal ions form oxides
• MO, M(OH)2
• Less polarizable than group 1 elements
• Fluorides are hydroscopic
• Ionic complexes with all halides
• Carbonates somewhat insoluble in water
• CaSO4 is also insoluble (Gypsum)
• Nitrates can form from fuming nitric acid
• Mg and Ca can form complexes in solution
• Zeolites complex group 2 metals
• In environment, clay minerals complex group 2
  metal ions

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Trivalent Actinides

• Am, Cm
• Am does have oxidation states 3-6
• most prevalent state is 3
• Similar chemical behavior
• Trivalent lanthanides can be used as homologs
• Thermodynamic data can be interchanged
• Alpha emitters
• Produced through neutron capture

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A241 Isotopes
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Solution chemistry

• Oxidation states
• Am (3-6)
• Am3, AmO2, AmO22 can be made
• Am4 rapidly disproportionates in solution except
  concentrated fluoride or phosphate
• Cm (3 and 4)
• Cm4 is a strong oxidizing agent
• Cm4 can be stabilized in high fluoride
  concentrations forming CmF5- or CmF62-

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Trivalent State

• In solution forms
• Carbonates
• Hydroxides
• Organic complexes
• Easily separated from other actinides by redox
  properties
• Behaves similar to trivalent lanthanides

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Absorption and fluorescence process of Cm3
Optical Spectra
Fluorescence Process
H G F
Emissionless Relaxation
7/2
A
Excitation
Fluorescence Emission
Z
7/2
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Waste from Reactor

• For a typical 1000 MWe reactor
• 30 tons of spent fuel are produced each year
• 4-11 m3 of HLW
• up to 400 m3 of non-HLW
• Medium Level Waste or Low Level Waste
• Generally waste not from spent fuel
• Only LLW in USA (no MLW)
• Radionuclide Inventory
• Concerned about
• amount
• half-live
• decay mode

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Power Plants

• Spent Fuel
• Actinides, Fission, Activation Products
• Radionuclides from Fuel (in Kg)
• Isotope Starting Ending ?
• 235U 33 7.9 -25.1
• 236U 0 4.0 4
• 238U 967 942.9 -24.1
• 237Np 0 0.75 0.75
• Am, Cm 0 0.2 0.2
• Pu 0 9.05 9.05
• FP 0 35.1 35.1

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Solvent Extraction PUREX

• Based on separating aqueous phase from organic
  phase
• Used in many separations
• U, Zr, Hf, Th, Lanthanides, Ta, Nb, Co, Ni
• Can be a multistage separation
• Can vary aqueous phase, organic phase, ligands
• Uncomplexed metal ions are not soluble in organic
  phase
• Metals complexed by organics can be extracted
  into organic phase
• Considered as liquid ion exchangers

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Extraction Reaction

• Phases are mixed
• Ligand in organic phase complexes metal ion in
  aqueous phase
• Conditions can select specific metal ions
• oxidation state
• ionic radius
• stability with extracting ligands
• Phase are separated
• Metal ion removed from organic phase
• Evaporation
• Back Extraction

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Effect of nitric acid concentration on extraction
of uranyl nitrate with TBP
48
Reactions

• Tributyl Phosphate (TBP)
• (C4H9O)3PO
• Resonance of double bond between P and O
• UO22(aq) 2NO3-(aq) 2TBP(org)
  lt--gtUO2(NO3)2.2TBP(org)
• Consider Pu4

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Extraction Systems

• Automatic systems are available
• Separation of solutions based on density
• Organic usually lower density than water
• Chlorinated hydrocarbons tend to be denser than
  water
• Need to achieve phase separation before solution
  extraction

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Single Solvent Extraction Stage
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Aside Third phase formation

• Brief review of third phase formation
• Related prior research
• Laboratory methods
• Np third phase behavior
• Comparison with U and Pu
• Spectroscopic observations

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Third Phase Formation

• In liquid-liquid solvent extraction certain
  conditions cause the organic phase to split
• ?PUREX separations using tributyl
• phosphate (TBP)
• ?Possible with future advanced separations
• Limiting Organic Concentration (LOC) highest
  metal content in phase prior to split
• Light phase mostly diluent
• Heavy phase extractant and metal rich
• ?Problematic to safety!

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Actinide Third Phases
Light Phase
Pu(VI)
Heavy Phase
Pu(IV)
Aqueous Phase
U(VI)
Np(IV)
Np(VI)
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Importance to Safety

• Increased risk of criticality
• Phase inversion
• Difficulty in process fluid separations
• Carry-over of high concentration TBP to heated
  process units
• ?Possible contribution to Red Oil event at
  Tomsk, Russia

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Phase Inverted Plutonium
Light Organic
Heavy Organic
Aqueous
Inverted Organic
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Prior Research

• Majority of work focused on defining LOC
  boundary (reviewed by Rao and Kolarik)
• -Effects of temp., concs., acid, diluent,
  etc.
• Recent work on possible mechanisms
• -Reverse micelle evidence from neutron
  scattering (Osseo-Asare Chiarizia)
• Spectroscopic studies -UV, IR, EXAFS (Jensen)

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Reverse Micelle Theory
Possible Reverse Micelle

• Classical Stoichiometry

• UO22 2NO3- 2TBP ? UO2(NO3)2 ? 2TBP

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Role of the Metal

• LOC behavior well known for U(VI), U(IV), Pu(IV),
  and Th(IV)
• Little data available on Pu(VI)
• No data on any Np systems
• Mixed valence systems not understood

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Mixed Systems

• Observed effect of Pu(VI) in HPT vs. C12
• Large impact of presence of Pu(VI) in HPT
• -Indications heavy phase enriched in Pu(VI)
• Opposite found with U(VI) inhibiting phase
  separation in U(IV) system (Zilberman 2001)
• ?Suggests possible role of trinitrato species
• AnO2(NO3)3-

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Neptunium Study

• Unique opportunity to examine trends in the
  actinides LOC curve for U(IV) vs Pu(IV)
• - Effective ionic charge
• - Ionic radii
• - Stability constants for trinitrato species
• Never been investigated
• Ease of preparing both tetravalent and hexavalent
  nitrate solutions

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Neptunium Methods

• Worked performed at Argonne National Laboratory,
  Argonne, IL
• Stock prepared from nitric acid dissolution of
  237Np oxide stock
• Anion exchange purification
• -Reillex HPQ resin, hydroquinone (Pu reductant),
  hydrazine (nitrous scavenger)
• Np(IV) reduction with hydrogen peroxide reduction
• Np(VI) oxidation with concentrated HNO3 under
  reflux

Np(IV) nitrate
Np(VI) nitrate salt
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LOC Behavior

• Np(VI) near linear
• Np(IV) slight parabolic
• ? Appears between linear U(IV) and parabolic
  Pu(IV)
• Both curves similar resemblance to distribution
  values Purex systems
• ? Suggests possible link with metal-nitrate
  speciation

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Np Third Phase Boundaries
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Comparison with Other Actinides
LOC in 7M HNO3 / 1.1M TBP/dodecane 20-25 C, M
U Np Pu
An (IV) 0.08 (Wilson 1987) 0.15 0.27 (Kolarik 1979)
An (VI) No 3F (Chiarizia 2003) 0.17 0.10
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Organic Nitric Acid

• Balance available TBP and organic H
• Np(IV) mixture of the monosolvate (TBP HNO3)
  and hemisolvate (TBP 2HNO3) species
• Np(VI) hemisolvate
• ?Agrees with literature data on U(VI) and Th(IV)
  acid speciations

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Valence Trends An (IV)

• General trend decreasing LOC as ionic radii
  increases
• ?Lowest charge density lowest LOC
• Np intermediate between U and Pu
• Literature Th(IV) data consistent with trend

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Valence Trends An(VI)

• An(VI) LOC increases as ionic radii decreases
• ?Opposite trend for An(IV), including Th(IV)
• Charge density using effective cationic charge
  and 6-coordinate radii
• ?No evidence of correlation with charge density
  within error of effective charge data
• Oxo group interactions not fully considered
• ? Future work required

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Spectral Study Methods

• Look for spectral trends in Np(VI) system
• Examined trends for
• -LOC
• -Metal loading
• -Nitrate loading (using NaNO3)
• 5 mm quartz cuvette with Cary 5 Spectrometer

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LOC Spectra
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Metal Loading
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Nitrate Effects
Aqueous
Organic
H 4M, Np(VI) 0.03M
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Valence Scoping Experiments

• Examined various mixes of Pu(IV)/Pu(VI)
• Solutions prepared by method of slow addition of
  concentrated HNO3 to heated syrupy Pu nitrate
  solution
• Use UV-Vis peak analysis for determination of
  initial aqueous composition
• Perform mole balance on aqueous phase before and
  after contact for organic content of each valence
  species (some samples)

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Spectrum Mixed Valence Phases
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Third phase conclusions

• Third phase behavior measured in Np
• LOC trends consistent with U and Pu
• Np(IV) LOC trends with charge density
• No clear correlation for Np(VI)
• Spectroscopic evidence suggests possible role of
  trinitrato species in third phase

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Current and Future Fuel Cycles US Approach and
RD Programs Next Generation Nuclear Plant

• The Department is engaged in the review and
  approval process for the NGNP Acquisition
  Strategy
• Critical Decision from the Deputy Secretary is
  expected to be issued in a matter of weeks
• We expect to be able to make awards for NGNP in
  2005
• Expected NGNP to be gas cooled reactor
• TRISO fuel
• Prismatic
• Based on discussion amongst current researchers
• Not official
• Research coupled with Gen IV, AFCI, nuclear
  hydrogen and NERI
• NERI program covers all areas
• Increase university participation

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US DOE Advanced Fuel Cycle Initiative

• Advanced Fuel Cycle Initiative
• Administered by the Office of Nuclear Energy,
  Science and Technology
• Stems from National Energy Policy Development
  Group, May 2001
• Support expansion of nuclear energy in the United
  States
• Develop advanced nuclear fuel cycles and next
  generation technologies
• Develop advanced reprocessing and fuel treatment
  technologies

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AFCI Mission and Goals

• Mission
• Develop technologies for the transition to a
  stable, long-term, and politically acceptable
  advanced fuel cycle
• Waste
• Proliferation resistance
• Economics
• Safety
• Transition from once-through fuel cycle to an
  advanced closed fuel cycle
• Current focus on aqueous reprocessing additional
  research on pyroprocessing
• Goals
• Develop advanced, proliferation-resistant fuel
  cycle technologies for current and
  next-generation reactors
• Develop a recommendation on the need for a second
  repository in the 2007-2010 timeframe
• Repository and proliferation needs related to
  separation
• Near term focus on utilization of existing
  reactors for transmutation
• Reduce cost of geologic disposal by enhancing
  performance of Yucca Mountain
• Heat loading

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AFCI Research

• National Program on Development of New Nuclear
  Fuel Cycle
• Combines universities, national laboratories, and
  industry
• Long term view
• To be deployed in the future
• Training next generation of researchers
• Development of new facilities
• Super Atalante for separations and fuel
• Development of fuel cycle that combines
  separations and fuel design
• Utilization of separated material for fuel in new
  or existing reactors
• Address pressing nuclear issues facing the United
  States
• nuclear energy and waste management concerns
• declining US nuclear infrastructure
• Facilities and researchers
• global nuclear leadership
• Cooperation with international partners
• France
• Russia

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AFCI Research

• Separations
• Aqueous-organic
• Electrochemical separations in molten salt
• TRISO fuel
• Reprocessing and repository behavior of Pu, Np,
  Am fuel
• Fuel
• Inert matrix for existing light water reactors
• Advanced fuels with transuranic elements
• Oxides, carbides
• TRISO fuel
• Silica carbide coated fuel for gas reactors
• Production of coated Pu, Np, Am oxides
• Reactors
• Advanced light water reactors
• Gas reactors
• High temperature for H2 production
• TRISO fuel deep burn reactors
• Pu, Np, Am kernel
• Reactor physics and system analysis

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Overview of AFC reactors
86
AFCI separations

• Bulk U separations
• Precipitation
• Electrochemistry
• Disposal or re-enrichment
• Separation of actinides and fission products by
  group
• Transuranics (Np, Pu, Am, Cm)
• Solvent extraction
• For incorporation into fuel
• Discussion of Am and Cm separation
• Fuel fabrication
• Fission products
• Cs, Sr
• Separate disposal
• Lanthanides
• Separation from actinides

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Separation

• Primarily solvent extraction based on PUREX
• Organic phase tributylphosphate in dodecane
• Some inclusion of other organic ligand
• Acetohydroxamic acid
• Aqueous phase nitric acid at varying
  concentrations
• Other processes also examined
• Pyroprocessing
• Fluoride volatility

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Current Extraction Scheme

• UREX
• PUREX modification
• addition of the acetohydroxamic acid (AHA)
  reduces Pu
• Tetravalent Np and Pu forms aqueous phase AHA
  complexes
• U and Tc extracted
• CCD-PEG
• Cs and Sr extracted with chlorinated cobalt
  dicarbollide/polyethylene glycol (CCD/PEG )
• NPEX
• Np, Pu
• Nitric acid, acetohydroxamic acid, CH3COOH
• TRUEX
• Remaining fission products except lanthanides
• TBP with Diphenyl-N,N-dibutylcarbamoyl phosphine
  oxide (CMPO), oxalic acid
• CYANEX-301
• Am and Cm
• TBP, CYANEX-301

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Separation flowsheet
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High Level Waste and AFC

• Separation coupled to extension of repository
  utility
• Separation of heat generating isotopes
• Separation of long lived actinides
• Need 6000 MT reprocessing capacity

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Expect US Repository Need in 2100
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Current AFCI Direction

• Stress research for recommendation on second
  repository in 2007-2010
• explore new alternative approaches to provide
  confidence in selection
• advanced recycle research facility
• Investigate other advanced aqueous processes
• Defer building and construction
• Increase systems analysis and modeling
• Align separations with current US
  non-proliferation policy
• May need to emphasize group actinide separations
• Remove U, keep rest of actinides as group
• FY 2005 Budget at 68 M

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Evaluation of Fuel Cycles

• 6 cases were studied for 3 growth rates (0,
  1.8, 3.8)
• Once through LWR
• LWR Inert Matrix Fuel (IMF)(TRU) recycle in LWR
• start separations in 2025
• LWR MOX (Np, Pu, Am) one pass in LWR
• start separations in 2025
• LWR IMF (Pu, Np) Fast Reactor (FR) (TRU)
• LWR MOX (Pu, Np) one pass FR (TRU)
• LWR FR (TRU)

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Pu Inventory
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Repository Heat Loading
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Implementation of Fuel Cycle

• Separation facility identified as limiting factor
• Early large scale separation enterprise
• reactor capacity for recycled materials is not
  issue
• Delay in separation causes large inventory delay
  in 2nd tier reactor
• Exact separation scheme open to debate
• Ideally only fission products go to repository
• Separation and storage of Cm
• Decay of 244Cm will leave 245Cm
• Time before separation
• Analysis supports both 5 year and 30 year waiting
  period
• Different results based on heat loading and
  proliferation
• Issue is decay of 241Pu
• Need to prevent placement of 241Am, 238,240Pu and
  237Np in repository

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Separation Needs

• Both repository and proliferation resistance
  needs to be addressed
• Repository
• Reduction of heat loading
• Separation of Cs, Sr
• Removal of 241Pu and 241Am
• Environmental behavior of 237Np
• Proliferation
• Build up of fissile isotopes in fuel cycle
• Separation of Pu during reprocessing
• Procedure should not produce separated Pu stream

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Separations

• Current extraction research (Scheme 1)
• U and Tc with UREX
• U precipitation as nitrate
• Cs and Sr with CCD-PEG
• Np, Pu with NPEX
• Remaining fission products except lanthanides
  with TRUEX
• Am and Cm with CYANEX-301 and TPB
• Am and Cm separation
• New concept separation discussed (Scheme 2)
• U
• Cs, Sr, and long lived fission products
• No Pu or Np/Pu separation
• Actinide group separation
• Future separation research and development
  ongoing
• Am and Cm treatment
• Separation both from recycle or just Cm
• Mass separation can be applied to either scheme
• Initial U separation needed for both schemes

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Waste Forms and Packaging

• Components of Waste
• Radioactive Isotopes
• Other Materials
• Waste Forms
• Materials
• Characteristics
• Properties and analysis
• Spent Fuel

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Components of Waste

• Radioactive elements common in radioactive waste
• monovalent Cs
• divalent Sr, Co
• trivalent Am, Cm
• tetravalent Zr, Tc, Th, U, Pu
• heptavalent Np, Pu
• hexavalent U, Pu
• septvalent Tc
• Non radioactive elements need to be considered
• B, lanthanides, Si, Cu, Fe, Ni, Ti, Zr, C, H, O
• Each elements behave differently in the
  environment and needs to be considered separately

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Waste Placement and Packaging

• A disposal site will contain packaged waste
• Waste will have different sections and components
• Waste Form
• Form of the material containing the radioactive
  waste
• Canister
• Primary container of the waste form
• Consider robust canister
• Overpack
• Barrier surrounding canisters for up to 1 meter
• May not be used

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Waste Placement and Packaging

• Backfill
• Material placed into gallery
• Different possible backfill materials
• Bentonite, crushed geologic material
• High exchange capacity or low permeability
• Sleeve for removal may be included
• Drip shield
• Divert water from package

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Waste Package Requirements

• From 10CFR60
• Substantially complete (assuming anticipated
  processes and events) containment for 300 to 1000
  years after repository closure
• Release rate after 1000 years lt 10 ppm/year for
  inventory at 1000 years
• Retrievability at any time up to 50 years after
  emplacement starts

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Waste Forms

• In US, two existing high level waste forms
• Spent fuel
• Zircaloy clad
• 5 UO2
• Borosilicate glass
• SiO2-B2O3-Na2O
• 1-30 waste in the glass
• A number of other waste forms are being
  considered

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Waste Forms

• Ceramics
• For disposal of weapons grade Pu
• Very durable material
• Based on TiO2, ZrO2
• Up to 20 incorporation of waste
• For Weapons grade Pu, up to 10 Pu
• Zeolite ceramics examined for disposal of liquid
  metal reactor waste
• High Na contain precluded normal ceramics and
  glass

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Zeolite
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Waste Forms

• Other Glass
• Developed as potential candidates
• Pb-Fe phosphate
• Lanthanide borosilicate
• For weapons grade Pu
• Monazite
• Sulfur glass
• For Hanford waste
• Waste loading determines volume, radiation dose,
  and thermal property of glass

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Glass

• Inorganic product of fusion
• Cooled to a rigid condition
• No crystallization
• Amorphous material
• Any substance with rapid cooling

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Glass Structure

• Compound forming structural network
• Oxides of Si, B, and P
• Modifiers
• Decrease melting temperature
• Add favorable processing properties
• Can degrade stability, increase solubility
• Oxides of Na, K, Ca, Ba
• Intermediates
• Can be present in waste
• May act as network former, increase durability
• Oxides of Al, Fe

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Thermal Stability of Glass

• Devitrification
• Formation of crystals in glass
• Lower chemical stability
• Increased leaching
• Reduced waste loading
• Phase separation
• Liquid-liquid phase separation during formation

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Glass Radiation Stability

• Atomic displacement by heavy particle radiation
• Volume change
• Density changes by 1
• Depends upon glass chemical composition
• Crystallization
• Concern over stored energy in glass leading to
  cracking or crystallization
• Ionizing effect from ß and g
• Chemical effects
• Disordering
• Breaking of bonds causes increased corrosion
• Radiolytic processes in aqueous medium in contact
  with glass

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Glass Corrosion

• Formation of altered phase on glass surface
• Can inhibit diffusion of radionuclide out of
  glass
• Two possible method of radionuclide release
• Diffusion of radionuclide out of glass
• Depends upon chemical behavior of radionuclide
• Corrosion of glass with release of radionuclide
• Release depends upon glass
• Secondary phase formation varies for radionuclide

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Basic Dissolution Rate Equation RateSk(1-(Q/K)s S
surface area, krate coefficient, Q activity,
KKsp, sstoichiometric number for rate-limiting
reaction

• Need to consider colloids
• Chemical changes in near field can also effect
  glass dissolution

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Spent Fuel

• Barrier
• Zr cladding
• Zr corrosion
• Zr 2H2O -gt ZrO2 2H2
• Weakening of cladding
• Drop in thermal conductivity
• Radiation
• Atomic Displacement
• Neutron activation of Zr and Ni
• 3 phases of release
• Gap release, grain release, UO2 dissolution

116
Geochemistry

• Principles which control the behavior of
  dissolved groundwater constituents
• Behavior based on equilibrium concepts
• Provide insight into behavior
• Groundwater Constituents
• Concentration Units
• Molality (m) (mol/kg solute)
• Molarity (M) (mol/L)
• for dilute concentrations (lt0.2 M), mM
• Mass concentration (g/L, mg/L)
• Equivalent (valence per unit)
• Used for resins or humic substances (moles H/g)
• Often written in milliequivalents/g (meq/g)

117
Speciation and Transport
118
Temperature Effects

• Temperature effects can be described by enthalpy
  (?H) and entropy (?S)
• Gibbs Free Energy (?G) relates ?H and ?S
• ?G?H-T?S
• ?G-RTlnß
• T in K, R8.314 J/molK
• Temperature effect on ß can be described as

119
Oxidation-Reduction

• Charge of ions in solution
• Fe, Mn, Co, As, Cr, U, Np, Pu are some redox
  sensitive metal ions
• Eh-pH diagrams can show the oxidation states
• based on oxygen and hydrogen
• Eh is also written as pE
• O2(g) 4H 4e lt--gt 2 H2O
• 2H 2e lt--gt H2(g)
• At 25 C
• pE 20.8 - pH 0.25 logPO2
• pE - pH - 0.5 logPH2

120
Eh-pH diagram for Gohy Groundwater
121
Am and Cm at ORNL WAG-5 site

• pH near 7
• Carbonate system
• Use modified Gohy data
• FA stability less than HA
• FA(III) 2 µmol/L
• An3t Cm3 Am3 20 pmol/L
• aqueous carbonate concentration evaluated from
  the measured alkalinity
• ionic strength at 0.02 M

122
Data of WAG-5 Site

• Species logß
• AnFA(III) 6.090.12
• An(OH)FA(II) 13.040.20
• An(OH)2FA(I) 17.240.30
• An(CO3)FA(I) 12.740.30
• Also carbonate and hydrolysis data
• LC 0.279pH - 1.01
• maximum 1, minimum 0

123
Speciation Calculation for WAG-5 Site
Speciation Calculation for WAG-5 Site