Depleted Uranium Dioxide–Steel Cermet For Spent-Nuclear-Fuel Waste Packages
Charles W. Forsberg Vinod K. Sikka
Oak Ridge National Laboratory P.O. Box 2008; Oak Ridge, TN 37831-6180 Tel: (865) 574-6783; Email: [email protected]
Session: Fuel Cycle and Waste Management: General American Nuclear Society Winter Meeting Washington D.C. November 14, 2000
The submitted manuscript has been authored by a contractor of the U.S. Government under contract DE-AC05-00OR22725. Accordingly, the U.S. Government retains a nonexclusive, royalty-free license to publish or reproduce the published form of this contribution, or allow others to do so, for U.S. Government purposes. Outline
• Characteristics of cermets • Near-term waste waste package (WP) applications • Long-term repository benefits • Disposal of excess depleted uranium (DU) • Conclusions Characteristics of Cermets A CERMET IS A CERAMIC-METAL COMPOSITE MATERIAL • Ceramic and metal are separate phases • Properties between ceramics and metals • Large quantities of some cermets are produced (>105 t/y) • A cermet for this application includes: − Ceramic: DU dioxide (DUO2) − Metal: steel (the continuous phase) • Steel-UO2 cermets have been made with 60 to 90 vol % UO2 (nuclear fuels) ORNL DWG 2000-211 Steel
Cermet Steel (Continuous Phase)
Depleted Uranium Dioxide
Stel ORNL WG 20-21D1 Cermet (ContinDeplteSuos PhaedUrnim
CERMET is a CER amic MET allic Composite ORNL DWG 2000-212 Molten Steel
Water Jet Depleted Uranium Water Dioxide Particulates
Cermet Powder Metal Powder Production
Oxide-Metal Top Steel Powder Mix
“Picture Frame” Assembly
Bottom Steel
Powder Vacuum Degas Mix Degas And Weld Weld Edges Steel Heat And Fuse Cermet Heat Roll Cermet
Example Method for Cermet Production Near-Term WP Applications There Are Three Near-Term Cermet Applications In A Waste Package (Before Geological Degradation of the WP)
• Radiation shielding in a self-shielded WP • Structural components in the WP • WP basket ORNL DWG 2000-210
Edge Void Optional Shielding SNF Assembly Slot Layer: Cermet (22.64 cm x 22.64 cm Space x 458.5 cm) Structural Support: Stainless Steel or Cermet
Corrosion Inside Resistance: length C-22( 2 cm) 458.5 cm Steel( 1 cm)
Neutron Absorber: Cermet with Added Neutron Absorber Inside diameter Thermal Shunt: 142.4 cm Aluminum (0.5 cm)
Basket Structure
Uses of Cermet in PWR Fuel Assembly Waste Package A DUO2-Steel Cermet is Compatible With WP Requirements • Chemical compatibility is required − Cermet steel is compatible (or identical) with WP steel − DUO2 is in the same chemical form as is spent nuclear fuel (SNF) UO2 • A cermet changes DUO2 properties − Adds mechanical strength and ductility − Provides contamination control (clean surfaces) • A cermet meets multiple requirements Cermets Are Effective Radiation- Shielding Materials
• Cermets provide gamma-radiation shielding and limited neutron shielding • Unlike ceramics, cermets are ductile (no cracks) • DUO2 cermets can provide full gamma- radiation shielding − Density: 25% higher than steel − WP weight: ~100 t for a self-shielded WP Cermets Have Several Desirable Characteristics As Basket Materials
• Cermets can incorporate other neutron absorbers in addition to DUO2 • For long-term performance benefits (see following), cermet basket mixes DUO2 with the SNF • Cermet properties can be adjusted to meet varying requirements − Variable DUO2 to metal ratios − Alternative choices of metals Cermets May Also Be Attractive In SNF or HLW Storage or Transport Casks
• Same cermet applications (structural, radiation shielding, and basket) as in WPs • Allows beneficial use of excess materials − Use excess depleted uranium as DUO2 (lower- cost uranium form than metal) − Use slightly contaminated recycle steel (except clean steel for external surfaces) • Higher-density shield material than steel • Excellent end-of-life waste form − Iron in cermet minimizes solubility of DUO2 − High-integrity non-dispersible waste form Long-Term Repository Benefits There Are Potential Long-Term Benefits Of Using A Cermet In A Repository WP
• Reduce potential for long-term nuclear criticality in the repository • Reduce radionuclide release rates from the SNF to the accessible environment Cermet Degradation Mechanisms Are Defined; Thus, Predictions Of Long-term Behavior Are Possible • With a steel cermet, the steel oxidizes first − Steel oxidizes first if there is a mixture of steel
and UO2 (experimental observation) − Steel, the outer layer, oxidizes first • Oxidation process will initially result in a
iron oxide–DUO2 particulate mixture, which is intermixed with the SNF • DUO2 will ultimately oxidize to hydrated, higher oxidation states Long-Term Repository Benefits
Potential for long-term nuclear criticality in the repository Cermets May Reduce The Long- Term Potential For Nuclear Criticality In The Repository • Natural reactors (Oklo) have occurred with 1.3 wt % 235U in 238U • Average fissile content of light-water reactor (LWR) SNF is ~1.5 wt % equivalent 235U in 238U; potential for criticality exists • Other SNF has higher enrichment levels • DUO2 in the cermet lowers WP enrichment so that long-term criticality is less likely Natural Uranium Enrichment Levels Over Geological Time 6
Global ratio 5
3.8% Maximum 4 LWR SNF assay
U-235 depletion in 3 Oklo reactor zones
1.5% LWR SNF 2 fissile assay Isotopic fissile assay (%) assay fissile Isotopic
1.3% 1
1.0% Homogeneous criticality system limit
0 Past -2 -1 Present Time into the past (billion years)
ORNL DWG 97C-59 Methods Exist To Demonstrate That Repository Criticality Will Not Occur
• Isotopically dilute fissile materials with DU • Surround fissile materials with DU and show isotopic dilution over time (concept described herein) • Dilute 235U during uranium migration by isotopic exchange with 238U in rock − Model SNF and WP degradation − Follow 235U migration over time − Show that criticality does not occur ORNL DWG 2000-267
Waste form Groundwater
DUO 2 SNF cermet assembly
235U in 238U Rock
235U
238 Mix DU with 235 U Pack 235U in 238U Show no criticality U and show isotopic will occur until mixing over time geological isotopic dilution occurs
Methods to Assure Nuclear Criticality Will Not Occur in a Repository The Potential For Long-Term Near-Field Criticality In A Degraded Waste Package Is Difficult To Estimate Uncertain Geometry
Degraded SNF with enriched uranium Uncertain Composition
Density Particle size Solubility (leaching) separation separation separation
ORNL DWG 2000-252 Isotopic Dilution Of Spent Nuclear Fuel Uranium With DU Degraded Clad
Fuel Rod
Depleted Uranium Dioxide Between Rods or Fuel Assemblies Uranium Dioxide Fuel Pellet Uranium Dissolution Groundwater (Unsaturated Front with Uranium)
Isotopic Exchange of Uranium in Solution with Solid Uranium Uranium Saturated Groundwater
ORNL DWG 98C-65 Long-Term Repository Benefits
Reduce radionuclide release rates from SNF WPs DUO2 Cermets May Reduce SNF WP Radionuclide Releases By Providing More Time For Radioactive Decay • Delay by preservation of SNF UO2 − Maintain chemically reducing conditions in WP − Saturate groundwater with DUO2 to slow SNF UO2 dissolution • Delay by retarding radionuclide transport via groundwater from the SNF − Ion-exchange and absorption of radionuclides on
hydrated DUOx and hydrated iron oxides − Filter radioactive colloids from groundwater UO2 Ore Deposits Have Existed For Long Times In Environments Similar To Yucca Mountain (YM) • UO2 is chemically unstable in oxidizing environments such as YM • SNF UO2 oxidation will release radionuclides • UO2 ore deposits exist in such environments by sacrificial chemical reactions on the outer edges and elsewhere in the ore deposits • The same mechanisms can be used to
preserve SNF UO2 with DUO2 in the WP The Interiors Of Natural Uranium Ore Deposits Of Uranium Dioxide Are Preserved For Extended Times By Degradation Of Uranium Oxides On The Outside Surfaces Of The Ore Deposit Oxidizing Groundwater
Reduced Groundwater Flow with Uranium Oxidation and Insoluble Crust Subsequent Expansion from Formation of Uranium Silicates
Slow Uranium Dissolution with Maintain Chemically Uranium-Saturated Reducing Conditions Groundwater by Uranium Oxidation Natural UO2+x Ore Body
ORNL DWG 99C-362 SNF WP With DU02 Fill To Improve Repository Performance
Groundwater
Reduced Groundwater Flow with Fill Expansion Insoluble Crust from Formation of DU Silicates
Maintain Chemically Slow SNF Dissolution Reducing Conditions with DU-Saturated (Intact SNF) by DUO2 Groundwater Oxidation
Waste Package
Long-term Spent Nuclear Criticality Control Fuel Assembly by Isotopic Dilution Radionuclide Retention by Uranium Oxide Adsorption
ORNL DWG 99C-351 After WP Failure, Cermets May Reduce Radionuclide Releases By Reacting With Oxygen • SNF and SNF UO2 are stable under chemically reducing conditions • Oxygen removal creates chemically reducing conditions that preserve SNF − Cermet iron removes oxygen by iron oxidation to iron oxides − Cermet DUO2 removes oxygen by DUO2 oxidation to (DU)3O8 and DUO3*xH2O; DU Oxides In The 4+ Valence State Help Maintain Chemically Reducing Groundwater (With Low SNF Uranium Solubility) In The WP
FUEL PINS
REPOSITORY WASTE PACKAGE URANIUM OXIDE FUEL ELEMENTS PARTICULATES (bundle of pins)
URANIUM OXIDE PARTICULATES
REDUCING DEGRADED CLAD GROUNDWATER OXIDIZING GROUNDWATER DEPLETED URANIUM OXIDATION U+4 U +6 URANIUM DIOXIDE FUEL PELLET (fission products and actinides in UO2 crystals)
ORNL DWG 96C-333 Cermet DUO2 May Reduce SNF UO2 Dissolution Rates By Saturating The Groundwater With Uranium
• DUO2 saturates groundwater coming into the WP with uranium • Uranium-saturated groundwater can not dissolve SNF uranium with release of encapsulated radionuclides • SNF radionuclide releases are slowed. DU Fill Saturates WP Groundwater With
Uranium And Thus Minimizes SNF UO2 Dissolution
FUEL PINS
REPOSITORY WASTE PACKAGE URANIUM OXIDE FUEL ELEMENTS PARTICULATES (bundle of pins)
URANIUM OXIDE PARTICULATES
URANIUM-SATURATED DEGRADED CLAD WATER LOW-URANIUM GROUNDWATER URANIUM DISSOLUTION URANIUM DIOXIDE FUEL PELLET (fission products and actinides in UO2 crystals)
ORNL DWG 96C-332 Cermets May Reduce Radionuclide Releases By Slowing Radionuclide Transport In WP Groundwater • Adsorption and ion exchange − Iron and UO2 oxidation products absorb and exchange ions with many radionuclides − Delays radionuclide transport in groundwater • Filtration − Many radionuclides (particularly actinides) form colloids (small particulates transported by groundwater) − Iron and DUO2 oxidation products act as filters Disposal of Excess DU Use Of A DUO2 Cermet Would Provide A Disposal Method For Excess DU • Growing DU inventory − 4 to 6 t of DU produced per t of LWR fuel − Worldwide inventory ~106 t (40% in U.S.) − Low consumption (~103 t/year) • Geological disposal is a preferred option − Meets all disposal requirements − Oxide is the preferred form (no compatibility issues and same chemical form as that of SNF) • Option could use most or all DU Paducah Gaseous Diffusion Plant, CTD WS 1150 (ORO 88-899) Portsmouth Gaseous Diffusion Plant, CTD WS 1149 Conclusions Conclusions
• DUO2-steel cermets can potentially replace structural, shielding, and basket components of WPs • There are multiple potential long-term advantages − Long-term criticality control − Reduced radionuclide releases from WPs − Disposal of excess DU • Further examination of cermets in WPs is warranted Backup Information DUO2 Is The Preferred Form Of Uranium In The Cermet • The repository is primarily designed to
accept SNF uranium in the form of UO2 − No chemical compatibility issues with DUO2 − Massive knowledge base on behavior of UO2 in the repository environment • Cermet manufacturing may require DUO2 − Other uranium oxides decompose to UO2 and release gases upon heating − Cermet manufacturing methods require high processing temperatures and no gas releases A Cermet WP Could Consume Large Quantities of Contaminated Recycle Steel
• The YM repository will ultimately require ~11,000 WPs • WP gross weights may vary from 42 (current baseline) to 100+ tons − Alternative designs have different weights − Different programmatic viewpoints about the preferred WP (self-shielded?) • Potential demand for recycled steel is between 50,000 and 500,000 tons • If slightly contaminated steel is used, the external layer would be clean steel so as to avoid handling complications Needed Cermet Activities To Create A Cermet WP Option
• Define WP requirements, evaluate production routes, and estimate economics • Prepare and test cermets using alternative production techniques (cost vs performance) • Refine best production techniques • Effect technology transfer Cermet Use Will Require Close Cooperation Between And Among Potential Suppliers and Users • Suppliers − DU [U.S. Department of Energy (DOE)–NE) − Slightly contaminated steel (DOE–EM) • Potential Users − DOE SNF program (DOE–EM) − YM Project (DOE–RW) − HLW and SNF storage programs (DOE–EM) ORNL DWG 2000-267
Waste form Groundwater
DUO 2 SNF cermet assembly
235U in 238U Rock
235U
238 Mix DU with 235 U Pack 235U in 238U Show no criticality U and show isotopic will occur until mixing over time geological isotopic dilution occurs
Methods to Assure Nuclear Criticality Will Not Occur in a Repository