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Radiation Shielding for Storage and Transportation Cask Using Depleted Uranium Oxide in Cementitious Matrices Leslie R

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Radiation Shielding for Storage and Transportation Cask Using Depleted Uranium Oxide in Cementitious Matrices Leslie R Radiation Shielding for Storage and Transportation Cask Using Depleted Uranium Oxide in Cementitious Matrices Leslie R. Dole Juan J. Ferrada Catherine H. Mattus U.S./Russian Depleted Uranium Workshop: Review of ISTC Projects Oak Ridge National Laboratory Oak Ridge, Tennessee May 17–21, 2004 OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY BACKGROUND •Combinations of gamma adsorbing DU and neutron absorbing hydrated binders are shown to effectively reduce the size and weight of storage, transport, and disposal casks •DUCRETE will become one of the materials of choice for advanced spent nuclear fuel (SNF) casks 9Requires the demonstration of low-cost fabrication processes 9Requires the demonstration of long-term durability under expected service conditions OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY Comparison of Equivalent Wall Thickness (in.) Shield Thickness 30 Neutrons Gamma 25 20 15 10 5 0 DUCRETE Uranium Metal Lead Metal Stainless Steel Concrete Dr. J. Sterbentz, INEEL J. W. Sterbentz, Shielding Evaluation of a Depleted Uranium Heavy Aggregate Concrete for Spent Fuel Storage Casks, INEL-94/0092 (Idaho Falls, Idaho: Idaho National Engineering and Environmental Laboratory, October 1994). OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY DUCRETE Casks are Considerably Smaller and lighter than Casks Constructed of Ordinary Concrete The DUCRETE cask is 35 tons lighter and 100 cm smaller in diameter than casks made from Cask Lid Air Outlet ordinary concrete. Multipurpose Sealed Basket (Shell Wall) SNF Fuel Assemblies Concrete Cask Liner Concrete Comparison of conventional and DUCRETE Air Inlet Duct spent-fuel dry storage casks/silos Air Entrance OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY For the USA, Estimated Cumulative Number of SNF Assemblies for PWR and BWR Light-Water Reactors Through 2020, Low-Case 300,000 250,000 200,000 150,000 100,000 Assemblies Cummulative Cummulative Number of Fuel Fuel of Number 50,000 PWR 0 PWR 2002 2004 2006 2008 BWR 2010 2012 2014 2016 2018 Total Year 2020 U.S. Department of Energy, Integrated Data Base Report —1994: U.S. Spent Nuclear Fuel and Radioactive Waste Inventories, Projections and Characteristics, DOE/RW-0006, Rev 11 (Washington, D.C.: September 1995), p. 36. OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY DUO2 Use in Storage/Transport Casks for USA Commercial SNF at 55.7 tonnes per Cask Potential Use of DUO2 in SNF Storage for 24 PWR and 61 BWR Assemblies per Storage Cask 500,000 400,000 300,000 , tonne 2 200,000 DUO 100,000 0 BWR BWR 2002 2004 2006 2008 2010 PWR 2012 2014 2016 2018 Total tonnes Year 2020 OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY Russian Dry Storage of SNF • No site selected, but has a potential candidate site near Chelyabinsk • Russia has proposed development of an international repository and to Import 30000 tons of spent fuel OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY Dry Storage and Transport Casks have been Developed for Russian SNF A Metal Concrete Cask for Storing and Transporting the Spent Fuel of RBMK Reactors A. Zubkov, V. N. Fromzel’, V. Yu. Vasil’ev, B. K. Danilin, V. A. Nikitin, and L. V. Fromzel’ A cask was developed for storing and transporting the spent fuel of a nuclear power station with RBMK reactors. Nuclear and radiation safety has been appraised and an analysis of the thermal state and the strength of the cask under normal and emergency conditions has been made. It is shown that the cask complies with the requirements of international and Russian regulations. OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY Russian RBMK Spent Fuel Cask with Heavy Concrete Heavy concrete with steel shot and barium sulfate GNB CONSTOR test cask for RBMK SNF OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY RMBK SNF Shipments in Russia A train carrying a load of SNF from a Ukrainian nuclear power plant arrived to Zheleznogorsk, Krasnoyarsk. OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY DUAGG Briquettes are Stabilized DU Aggregates with Basalt Sintering Agent To get high strengths Briquettes are pressed, solidified by liquid-phase sintering, crushed, and gap- graded for use in high- strength DUCRETE at 5000 to 6000 psi, (35–42 MPa) OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY Current Accelerated DUAGG Exposure Studies Using ASTM C289-94 Standard Test Method At a consistent surface-to-liquid ratio of 1:10, the sintered DUAGG samples are exposed to (1) distilled water (2) 1N sodium hydroxide standard solution (3) saturated water extract of high-alkali cement Simulating expected service conditions, there are three exposure temperatures and six times: 25, 66, and 150ºC at intervals of 30, 60, 90,180, 360, and 730 days OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY View of DUAGG Before Testing C B C A Detail of the surface (secondary electrons) Particle A contains Al Particle B contains Ti and some Mg Area C contains DUO particles surrounded by dark basalt OAK RIDGE NATIONAL LABORATORY 2 U. S. DEPARTMENT OF ENERGY DUAGG 6 Months in DI Water 50 µm Backscattered electrons A A Secondary electrons A 10 µm 50 µm 150°C — secondary electrons Particles A contain Ti and some Mg OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY DUAGG 6 months in Cement Pore Water 50 µm 67°C—Backscattered electrons 67°C — secondary electrons 20 µm 50 µm 67°C — secondary electrons Covered by CaCO3 and needle-like crystals containing Ca, Si, and some Al OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY Uranium Leached from DUAGG 0.160 0.140 0.120 0.100 0.080 % leached 0.060 0.040 0.020 0.000 O DI water- 1 m U. S. AKD R IDGE DI water- 2 m EPARTMENT OF DI water- 3 m N ATIONAL DI water- 6 m DI water- 13 m E L NERGYABORATORY NaOH - 1 m NaOH - 2 m NaOH - 3 m NaOH - 6 m NaOH - 13 m cement sol.- 1 m 20°C 66°C cement sol.- 2 m 150°C cement sol.- 3 m cement sol.- 6 m cement sol.- 13 m Comparison of Release Rates for Uranium with Data from the Literature •DUAGG tests: 0.25 mg/(m2•day) after 1 month at 66°C in DI water ~0.40 mg/(m2•day) after 13 months in cement pore solution 2 •UO2 (Thomas & Till) : 5 mg/(m •day) after 8 days at 70°C in DI water G. F. THOMAS and G. Till, “The Dissolution of Unirradiated UO2 Fuel Pellets under Simulated Disposal Conditions,” Nucl. Chem. Waste Manage. 5, 141–147 (1984) OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY DUCRETE Aggregate is Leach Resistant Leach Rate Test Results for DU 6 Subjected to USEPA TCLP Test 5 DU Material Concentration in Conc rete Leachate (ppm) 4 Magnetite DUCRETE 3 DUCRETE 0.42 Steel DUAGG 4 2 Lead Uranium UO2 170 Half Value Thickness (cm) 1 U O 420 0 3 8 0 5 10 15 20 UO3 6900 Material Density (g/cm3) UF4 7370 DUCRETE half value thickness comparison USEPA= U.S. Environmental Protection Agency OAK RIDGE NATIONAL LABORATORY TCLP= Toxicity Characteristic Leach Test U. S. DEPARTMENT OF ENERGY Conclusions • After 13 months of exposure, the release rate of uranium in a cement pore solution is low and shows that DUAGG is superior to pure UO2 • A protective layer of recrystallization products from the basalt phase of DUAGG cover the surface, slowing the release of uranium • In the cement pore solution, after 6 months of exposure, no deleterious products from the alkali- aggregate reaction were seen OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY Summary • Results show that DUAGG can be expected to be stable under the casks’ service conditions • We are continuing laboratory experiments to characterize DUAGG/DUCRETE materials and their behavior in SNF cask applications • We are pursuing a collaboration with Holtec International, Inc., and Russians to design and demonstrate the next generation of SNF storage and transport casks OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY Additional Technical features of DUCRETE • Thin cask walls and higher thermal conductivity increase thermal transport resulting in greater capacities for fuel bundles • Increased fracture toughness contributes significant physical protection against missiles and explosive blasts OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY DUCRETE Enables Increased Thermal Loading Capacity of Casks • Reduce wall thickness • Increase thermal conductivity • Both • Increase internal diameter and wall area by r2 OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY Theoretical Thermal Conductivities of Composites Concrete Volume percent of component W/(m•K) Cement Steel Lime- Barium UO2 Paste stone Sulfate powder Normal .25 .10 .65 2.9 Concrete Heavy .25 .10 .65 3.4 Concrete DUCRETE .25 .12 .63 7.0 OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY Fracture Toughness with Microreinforcement OAK RIDGE NATIONAL LABORATORY U. S. DEPARTMENT OF ENERGY ISTC Micro-Reinforcement Project Micro-reinforcement of Concrete in SNF Cask Shielding Performers: Lead Institution: Russian Federal Nuclear Center – All Russia Science and Research Institute of Experimental Physics (RFNC-VNIIEF), Project Manager: S. Ermichev, V. Shapovalov. Other participating institutes: VNIIKhT–Seredenko; RAS ICP–Gromov; VNIINM–Orlov, Sergeev; the RAS IHT–Mineev Description: The addition of micro-reinforcement to concrete structures increases their fracture toughness and therefore the ability of the concrete to absorb energy (e.g., rocket from a terrorist attack or airplane crash). This enhances the durability and reliability of an SNF cask under high-impulse stresses and provides for more effective physical protection of the cask contents. It is proposed that the tensile strength of concrete be increased by adding metal and/or polymeric fibers Benefits: The use of micro-reinforcement may eliminate, the need for rebar in construction of concrete structures specifically concrete SNF storage casks.
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