scholarly journals Conceptual Design of a MEDE Treatment System for Sodium Bonded Fuel

2008 ◽  
Author(s):  
Carl E. Baily ◽  
Karen A. Moore ◽  
Collin J. Knight ◽  
Peter B. Wells ◽  
Paul J. Petersen ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Department Of Energy ◽  
Test Facility ◽  
Evaporation Chamber ◽  
Oak Ridge ◽  
Uranium Fuel ◽  
Fuel Assemblies ◽  
Metal Fuel ◽  
Enriched Uranium

Unirradiated sodium bonded metal fuel and casting scrap material containing highly enriched uranium (HEU) is stored at the Materials and Fuels Complex (MFC) on the Idaho National Laboratory (INL). This material, which includes intact fuel assemblies and elements from the Fast Flux Test Facility (FFTF) and Experimental Breeder Reactor-II (EBR-II) reactors, as well as scrap material from the casting of these fuels, has no current use under the terminated reactor programs for both facilities. The Department of Energy (DOE), under the Sodium-Bonded Spent Nuclear Fuel Treatment Record of Decision (ROD), has determined that this material could be prepared and transferred to an off-site facility for processing and eventual fabrication of fuel for commercial nuclear reactors. A plan is being developed to prepare, package, and transfer this material to the DOE HEU Disposition Program Office (HDPO), located at the Y-12 National Security Complex in Oak Ridge, Tennessee. Disposition of the sodium bonded material will require separating the elemental sodium from the metallic uranium fuel. A sodium distillation process known as MEDE (Melt-Drain-Evaporate), will be used for the separation process. The casting scrap material needs to be sorted to remove any foreign material or fines that are not acceptable to the HDPO program. Once all elements have been cut and loaded into baskets, they are then loaded into an evaporation chamber as the first step in the MEDE process. The chamber will be sealed and the pressure reduced to approximately 200 mtorr. The chamber will then be heated as high as 650 °C, causing the sodium to melt and then vaporize. The vapor phase sodium will be driven into an outlet line where it is condensed and drained into a receiver vessel. Once the evaporation operation is complete, the system is de-energized and returned to atmospheric pressure. This paper describes the MEDE process as well as a general overview of the furnace systems, as necessary, to complete the MEDE process.

scholarly journals Deployment and Operation of the ES-3100 Type B Shipping Container

2006 ◽  
Author(s):  
Jeffrey G. Arbital ◽  
Dean R. Tousley ◽  
Dennis B. Miller
Keyword(s):  
National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Department Of Energy ◽  
Radioactive Material ◽  
Oak Ridge ◽  
Shipping Container ◽  
Enriched Uranium ◽  
Transportation Research ◽  
Regulatory Commission ◽  
The U.S

The U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) is shipping, for disposition purposes, bulk quantities of fissile materials, primarily highly enriched uranium (HEU). The U.S. Department of Transportation (DOT) specification 6M container has been the workhorse for NNSA and many other shippers of radioactive material since the 1980s. However, the 6M does not conform to the packaging requirements in the Code of Federal Regulations (10 CFR 71) and, for that reason, is being phased out for use in the DOE secure transportation system by the end of 2006. BWXT Y-12 developed and licensed the ES-3100 container to replace the DOT 6M. The ES-3100 was certified by the Nuclear Regulatory Commission (NRC) in April 2006. The process of deploying the new package began in June 2005 and is planned to be completed in July 2006. The package will be fully operational and completely replace the DOT 6M at the Y-12 National Security Complex (Y-12) by October 2006. This paper reviews the deployment process and the mock loading station that was installed at National Transportation Research Center (NTRC) of Oak Ridge National Laboratory. Specialized equipment, tools, and instrumentation that support the handling and loading operations of the ES-3100 are described in detail. Loading options for other user sites are explored in preparation for deployment of this new state-of-the-art shipping container throughout the DOE complex and the private sector.

Advanced Research and Technology Development Fossil Energy Materials Program

1988 ◽  
Vol 110 (4) ◽  
pp. 670-676
Author(s):  
R. R. Judkins ◽  
R. A. Bradley
Keyword(s):  
Technology Development ◽  
National Laboratory ◽  
Development Program ◽  
Ceramic Composites ◽  
Department Of Energy ◽  
Energy Materials ◽  
Alloy Development ◽  
Fossil Energy ◽  
Oak Ridge ◽  
Research Activities

The Advanced Research and Technology Development (AR&TD) Fossil Energy Materials Program is a multifaceted materials research and development program sponsored by the Office of Fossil Energy of the U.S. Department of Energy. The program is administered by the Office of Technical Coordination. In 1979, the Office of Fossil Energy assigned responsibilities for this program to the DOE Oak Ridge Operations Office (ORO) as the lead field office and Oak Ridge National Laboratory (ORNL) as the lead national laboratory. Technical activities on the program are divided into three research thrust areas: structural ceramic composites, alloy development and mechanical properties, and corrosion and erosion of alloys. In addition, assessments and technology transfer are included in a fourth thrust area. This paper provides information on the structure of the program and summarizes some of the major research activities.

Development of Testing Methodologies for Onsite Radioactive Material Storage Containers

2008 ◽  
Author(s):  
Matthew R. Feldman
Keyword(s):  
Lawrence Livermore National Laboratory ◽  
National Laboratory ◽  
Nuclear Material ◽  
Department Of Energy ◽  
Radioactive Material ◽  
Development Effort ◽  
Nuclear Facilities ◽  
Los Alamos National Laboratory ◽  
Oak Ridge ◽  
Transportation Technologies

Based on a recommendation from the Defense Nuclear Facilities Safety Board, the Department of Energy (DOE) Office of Nuclear Safety Policy and Assistance (HS-21) has recently issued DOE Manual 441.1-1 entitled Nuclear Material Packaging Manual. This manual provides guidance regarding the use of non-engineered storage media for all special nuclear material throughout the DOE complex. As part of this development effort, HS-21 has funded the Oak Ridge National Laboratory (ORNL) Transportation Technologies Group (TTG) to develop and demonstrate testing protocols for such onsite containers. ORNL TTG to date has performed preliminary tests of representative onsite containers from Lawrence Livermore National Laboratory and Los Alamos National Laboratory. This paper will describe the testing processes that have been developed.

scholarly journals Completion of Experimental Breeder Reactor-II Sodium Processing at Argonne National Laboratory

2002 ◽  
Author(s):  
Mary D. McDermott ◽  
Charles D. Griffin ◽  
Daniel K. Baird ◽  
Carl E. Baily ◽  
John A. Michelbacher ◽  
...  
Keyword(s):  
Sodium Hydroxide ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
The United States ◽  
Department Of Energy ◽  
Test Facility ◽  
United States Department ◽  
Safe Condition ◽  
And Control ◽  
Low Level Radioactive Waste

The Experimental Breeder Reactor - II (EBR-II) at Argonne National Laboratory - West (ANL-W) was shutdown in September 1994 as mandated by the United States Department of Energy. Located in eastern Idaho, this sodium-cooled reactor had been in service since 1964, and was a test facility for fuels development, materials irradiation, system and control theory tests, and hardware development. The EBR-II termination activities began in October 1994, with the reactor being maintained in an industrially and radiologically safe condition for decommissioning. With the shutdown of EBR-II, its sodium coolant became a waste necessitating its reaction to a disposal form. A Sodium Process Facility (SPF), designed to convert sodium to 50 wt% sodium hydroxide, existed at the ANL-W site, but had never been operated. The SPF was upgraded to current standards and codes, and then modified in 1998 to convert the sodium to 70 wt% sodium hydroxide, a substance that solidifies at 65°C (150°F) and is acceptable for burial as low level radioactive waste in Idaho. In December 1998, the SPF began operations. Working with sodium and highly concentrated sodium hydroxide presented some unique operating and maintenance conditions. Several lessons were learned throughout the operating period. Processing of the 330 m3 (87,000 gallons) of EBR-II primary sodium, 50 m3 (13,000 gallons) of EBR-II secondary sodium, and 290 m3 (77,000 gallons) of Fermi-1 primary sodium was successfully completed in March 2001, ahead of schedule and within budget.

scholarly journals Probabilistic dose assessment for personnel during decommissioning of spent nuclear fuel storage facility of transportable NPP “Pamir-630D”

2019 ◽  
Vol 64 (3) ◽  
pp. 378-384
Author(s):  
A. V. Kuzmin ◽  
A. V. Radkevich ◽  
V. P. Petrushkevich ◽  
N. D. Kuzmina
Keyword(s):  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Accurate Determination ◽  
Probabilistic Approach ◽  
Argonne National Laboratory ◽  
Department Of Energy ◽  
Dose Assessment ◽  
Storage Facility ◽  
Nuclear Facilities ◽  
Uncertainty Estimates

The aim of the current work is to perform a probabilistic dose assessment to quantify the relative importance of the source data uncertainties contribution towards the uncertainty estimates of collective and maximum individual doses of personnel during decommissioning of a storage facility. A probabilistic approach to the analysis of dose loads, including the analysis of sensitivity and uncertainty with respect to the input parameters of the used calculation models of dose assessment, allows to determine the most sensitive parameters, inaccuracies in the task of which lead to significant uncertainties in the estimates of dose loads on personnel and, therefore, require more accurate determination of conservative boundary values in deterministic analysis and safety justification. The calculations were performed by applying the code RESRAD-BUILD 3.50, developed by the Argonne National Laboratory of the US Department of Energy. The obtained results allow us to rank the parameters of the computational model according to the degree of their influence on the uncertainty of the final estimates of the dose loads on personnel, to develop recommendations for optimizing dose loads when performing radiation-hazardous work during nuclear facilities decommissioning.

scholarly journals Calculation of Metallocene Ionization Potentials via Auxiliary Field Quantum Monte Carlo: Towards Benchmark Quantum Chemistry for Transition Metals

2021 ◽  
Author(s):  
Benjamin Rudshteyn ◽  
John Weber ◽  
Dilek Coskun ◽  
Pierre A. Devlaminck ◽  
Shiwei Zhang ◽  
...  
Keyword(s):  
Quantum Monte Carlo ◽  
National Science Foundation ◽  
National Laboratory ◽  
Computer Time ◽  
Department Of Energy ◽  
National Science Foundation Grant ◽  
Science And Engineering ◽  
Oak Ridge ◽  
National Science ◽  
The U.S

Main Document<div>Supporting Information</div><div>XYZ Coordinates of Structures</div><div><br></div><div><div> An award of computer time was provided by the Innovative and Novel Computational Impact on Theory and Experiment (INCITE) program. This research used resources of the Oak Ridge Leadership Computing Facility at the Oak Ridge National Laboratory, which is supported by the Office of Science of the U.S. Department of Energy under Contract No. DE-AC05-00OR22725.</div><div>This work used the Extreme Science and Engineering Discovery Environment (XSEDE), which is supported by National Science Foundation grant number ACI-1548562. In particular, we used San Diego Computing Center's Comet resources under grant number TG-CHE190007 and allocation ID COL151.</div><div>The Flatiron Institute is a division of the Simons Foundation.</div></div>

Creep Strength and Microstructure of AL20-25+Nb Alloy Sheets and Foils for Advanced Microturbine Recuperators

2006 ◽  
Author(s):  
Philip J. Maziasz ◽  
John P. Shingledecker ◽  
Neal D. Evans ◽  
Yukinori Yamamoto ◽  
Karren L. More ◽  
...  
Keyword(s):  
Phase I ◽  
Phase Ii ◽  
Creep Resistance ◽  
National Laboratory ◽  
Rupture Life ◽  
Cost Effective ◽  
Test Facility ◽  
Oak Ridge ◽  
Collaborative Program ◽  
Wide Range

The Oak Ridge National Laboratory (ORNL) and ATI Allegheny-Ludlum began a collaborative program in 2004 to produce a wide range of commercial sheets and foils of the new AL20-25+Nb stainless alloy, specifically designed for advanced microturbine recuperator applications. There is a need for cost-effective sheets/foils with more performance and reliability at 650–750°C than 347 stainless steel, particularly for larger 200–250 kW microturbines. Phase I of this collaborative program produced the sheets and foils needed for manufacturing brazed plated-fin (BPF) aircells, while Phase II provided foils for primary surface (PS) aircells, and modified processing to change the microstructure of sheets and foils for improved creep-resistance. Phase I sheets and foils of AL20-25+Nb have much more creep-resistance than 347 steel at 700–750°C, and foils are slightly stronger than HR120 and HR230. Preliminary results for Phase II show nearly double the creep-rupture life of sheets at 750°C/100 MPa, with the first foils tested approaching the creep resistance of alloy 625 foils. AL20-25+Nb alloy foils are also now being tested in the ORNL Recuperator Test Facility.

Postirradiation Ductility Demonstration Tests of Weapons-Derived MOX Fuel Cladding

2004 ◽  
Author(s):  
W. J. McAfee ◽  
W. R. Hendrich ◽  
T. E. McGreevy ◽  
C. A. Baldwin ◽  
N. H. Packan
Keyword(s):  
Synergistic Effects ◽  
National Laboratory ◽  
Department Of Energy ◽  
Test Method ◽  
Test Results ◽  
Fuel Cladding ◽  
Oak Ridge ◽  
Mox Fuel ◽  
Uranium Plutonium ◽  
Cladding Material

The U.S. Department of Energy (DOE) Fissile Materials Disposition Program (FMDP) is pursuing reactor irradiation of mixed uranium-plutonium oxide (MOX) fuel for disposal of surplus weapons-usable plutonium. Since most of the MOX fuel utilization experience has been with reactor-grade plutonium, it is desired to demonstrate that the unique properties of the surplus weapons-derived or weapons-grade (WG) plutonium do not compromise the applicability of this MOX experience base. A related question to be addressed for weapons-derived MOX fuel is that of ductility loss of the cladding. While irradiation induced loss of ductility has long been known and quantified for many cladding materials, the potential synergistic effects of irradiation and the unique constituents (i.e., gallium) of weapons-derived MOX fuel are not known. As part of an extensive fuel qualification research program conducted by Oak Ridge National Laboratory (ORNL), a new test method was developed and validated to measure the room temperature ductility and hoop tensile properties of MOX fuel cladding. The cladding material is a zirconium alloy designated as Zr-4 manufactured by Sandvick Corporation. This paper is a summary of the test method developed and of demonstration test results obtained for MOX cladding irradiated to 21 GWd/MT [7 × 1020 n/cm2 (E &gt; 1 MeV)].

Seismic Evaluation of the Oak Ridge National Laboratory High Flux Isotope Reactor Support Assembly

2006 ◽  
Author(s):  
Gustavo A. Aramayo
Keyword(s):  
Structural Component ◽  
National Laboratory ◽  
Department Of Energy ◽  
Seismic Excitation ◽  
High Flux ◽  
Seismic Evaluation ◽  
Oak Ridge ◽  
Oak Ridge National Laboratory

The support assembly of the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) was modeled to determine the assembly’s response to a seismic excitation. The compliance of this structural component to established U. S. Department of Energy (USDOE) standards [1, 2] is evaluated.

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