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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 865 records · Page 48
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Progress Towards Treating the Sludges Produced During Spent Fuel Processing - 19546

As a consequence of over 50 years of Magnox reprocessing, Sellafield site is in possession of significant quantities of Magnox sludge. This sludge material is the by-product of corrosion of the magnesium based Magnox metal that clads the Magnox fuel elements. Retrievals and processing of this material poses a significant challenge as corrosion of Magnox metal has occurred under a variety of conditions and over a significant period of time, in a variety of locations. Devising processes to treat heterogeneous materials offers a significant challenge to identify technologies with robust envelopes of acceptance, to effectively operate together. This is further compounded by difficulties defining test programmes and materials for testing that will provide assurance that nonradioactive testing is suitably representative or bounding. An overview of the significant body of work undertaken to process Sellafield's Magnox sludge is provided, from the thousands of commercially sensitive reports generated over the past five decades. This covers the sampling, characterisation, strategy and process evolution, technology development and test materials. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Modification of an Agitated Tube Reactor for use in the Reprocessing of Spent Nuclear Fuel

Aims: Convert an Agitated Tube Reactor (ATR) into a counter-current solvent extraction device. Assess effect of liquid fill level on liquid dispersion. The agitated tube reactor (ATR) is an array of connected tubes. The ATR laterally shakes in a sinusoidal path, moving the agitator bar to create mixing. Originally designed by AM Technology as a cocurrent reactor, the ATR has many features that are potentially useful in the nuclear industry: Tubes are connected by non-moving seals; Reduction in seal damage and reduced risk of expensive repairs in radioactive environments; Minimal axial dispersion within a tube; Plug flow; Narrower residence time distributions that traditional technology; Multiple tubes allow for multiple reactions in series; Converting to a counter-current design allows the ATR to be used for solvent extraction; Reprocessing or decommissioning. A 180 ml PERSPEX ATR tube with no liquid flow through was constructed and filled with a 1:1 solvent:aqueous ratio of kerosene and water. The tube was filled such that the total liquid filled 25 %, 50 %, 75 % or 100 % of the tube. A Photron Fastcam SA5, positioned at the side and the end of the tube was set to a frame rate of 5400 fps with a resolution of 1024 x 1024 and 6000 fps with a resolution of 1024 x 896 respectively. Droplets were recorded after 2 minutes of shaking to ensure good mixing. Droplet sizes are used to determine optimum fill level in the ATR. The shaking frequency follows a sinusoidal path and was set to 5 Hz. The images shown in the results section were taken at the peak of the sinusoidal motion. To convert to a counter-current design the two liquids need to be separated at the end of each tube. Liquids have been allowed to separate under gravity separation. Inclusion of baffles to reduce remixing by sloshing. Wall with height corresponding to 25 % fill volume allows kerosene to overflow into separate outlet. Increase in liquid fill level decreases the dispersion of the two liquid phases. Higher dispersion of droplets decreases surface area and therefore increases mass transfer in solvent extraction. 25 % should provide the greatest mass transfer though more work needs to completed to confirm this.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

OPTIMUS{sup TM} Packaging for Intermediate Level Wastes and Spent Fuel - 20129

NAC's OPTIMUS{sup TM} packaging is a family of Type B(U)F transportation packages designed for maximum flexibility and cost-efficiency to support shipment of a wide range of challenging wastes and materials. They are small, modular packaging designs meeting DOT weight limits for road transport. OPTIMUS{sup TM} contents include low level waste, mixed low-level waste, irradiated fuel waste and intermediate level waste. Contents can also include aerosol cans with compressed gas and liquefied gas propellants and standard DOT 3E lecture bottles for shipment to treatment facilities. OPTIMUS{sup TM} packaging provides a cost-effective solution for challenging and unique waste contents combined with design simplicity and operational flexibility meeting the requirements of both SSR-6 and 10 CFR 71 for Type B(U)F packaging. The modular design of the packaging components combined with the flexibility to add modified internals for enhanced packaging performance makes OPTIMUS{sup TM} the preferred solution for many radioactive material transportation needs. The OPTIMUS{sup TM} product line currently includes two packaging designs; OPTIMUS{sup TM}-L and OPTIMUS{sup TM}-H. OPTIMUS{sup TM}-L is a lightweight transportation packaging intended for low-activity wastes, whereas OPTIMUS{sup TM}-H is a larger, heavier package intended for high-activities wastes. Due to its small size and weight, six (6) to ten (10) OPTIMUS{sup TM}-L packages can be accommodated on a single legal-weight truck shipment, depending on the weight of the contents. For typical payloads, two (2) OPTIMUS{sup TM}-H packages can be accommodated on a single legal-weight truck shipment. The OPTIMUS{sup TM} packaging share the same cask containment vessel (CCV) in the OPTIMUS{sup TM}-H and OPTIMUS{sup TM}-L. The large cavity size of the CCV, which can accommodate a 416-liter (110-gallon) drum, combined with the small size, low weight and modularity of the OPTIMUS{sup TM} packaging provides unmatched flexibility. In this paper, NAC provides a technical overview of the OPTIMUS{sup TM} packaging and identifies the design features and technology advancements making the OPTIMUS{sup TM} a readily adaptable and flexible solution for packaging processing facilities, reactor and decommissioning wastes. Optimized shielding performance through the addition of internal shielding components and alternative shipping configurations are discussed. In addition, an overview of the first major deployment of the OPTIMUS{sup TM} packaging for the Whiteshell Laboratories Closure Project (WLCP) is discussed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Exploring the Use of Muon Momentum for Detection of Nuclear Material Within Shielded Spent Nuclear Fuel Dry Casks

The inspection of the storage containers that house used nuclear fuel is an important issue facing the nuclear industry; currently, there are limited options available to provide for even minimal inspections. During the long storage life of the fuel, intermediate handling and transportation activities may take place that can create concerns regarding fuel integrity (safety) and continuity of knowledge (safeguards and security). Accidents and natural disasters also contribute to this concern. The reliable and safe management of used nuclear fuel is among the critical tasks to be addressed for the advancement of fission-based nuclear energy in the United States and elsewhere. Since the early 1950's, when the first nuclear power plant began to produce electricity, approximately 65,000 metric tons of used nuclear fuel have been generated in the United States. The postponement of the opening of a permanent repository until 2048 will lead to increased used fuel accumulation at reactor sites. A rapid increase in storage installations with more than 9,000 casks storing {approx}420,000 assemblies is anticipated by 2050 even if no new nuclear power plants are built. After used nuclear fuel has been placed inside a cask, the cask is sealed, not allowing for visual inspection. Conventional methods, e.g., x-rays, are limited in that they cannot penetrate dense, well-shielded objects. More sophisticated techniques, such as penetrating neutrons or proton radiography, require the use of an expensive accelerator. Efforts by Ziock et al. using neutron and gamma-ray imagers concluded that '...the signatures are deemed insufficient to uniquely identify individual casks of the same type...'. It was found that the gamma-ray and neutron images were dominated by diffusion and thermalization effects resulting from the large amount of shielding. One promising mode of inspection would use cosmic-ray muons. These muons are naturally generated, are able to penetrate dense materials, and are incident from all directions above the horizon. Tomography performed using these muons could potentially provide efficient, inexpensive, remote assessment of used nuclear fuel. However, current muon applications are limited in that they have not included momentum measurement. If momentum measurement could be incorporated, that would represent a big step forward because it would allow more accurate estimation of the muon observables, e.g., scattering variance. This approach has the potential to reconstruct the contents of a cask with improved resolution and identify partial defects, fuel assemblies, and individual fuel rods.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Dose Rate Analysis of As-Loaded Spent Nuclear Fuel Casks - Paper 97

Used Nuclear Fuel Storage, Transportation and Disposal Analysis Resource and Data System (UNF-ST[and]DARDS) is a comprehensive data and calculation capability for UNF safety assessments currently being developed at Oak Ridge National Laboratory through a collaborative effort of multiple national laboratories and industry participants. The purpose of UNF-ST[and]DARDS is to provide a unified domestic UNF system of data and analysis capabilities in support of U.S. Department of Energy waste management and fuel cycle-related objectives. UNF-ST[and]DARDS is designed to provide safety assessments for UNF storage, transportation, and disposal systems based on actual characteristics of the as-loaded UNF (e.g., fuel assembly burnup, enrichment, decay time, cladding integrity, etc.). The cask-specific safety evaluations allow quantification of realistic safety margins and conditions and provide the technical basis for addressing concerns related to potential fuel failure (e.g., for high-burnup fuel). This paper describes the UNF-ST[and]DARDS methodology for calculating radiation dose rates for U.S. Nuclear Regulatory Commission (NRC) licensed casks based on actual UNF canister contents. Realistic dose rate values based on actual canister contents may be used in place of bounding dose rate values to support development of repackaging operations procedures, evaluation of radiation-related transportation risks, and communication with stakeholders. UNF-ST[and]DARDS capabilities have been recently expanded to include automated radiation source term calculations and dose rate analyses of as-loaded transportation casks. This unique capability determines realistic dose rate values based on actual canister UNF contents that may be used in radiation dose assessments to demonstrate safe transportation of commercial UNF. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Criticality calculation package and its use for a spent fuel storage pool with steam layer - 14324

Criticality studies need several calculations with small variations in a number of parameters. Therefore, a calculation package written at VTT in order to facilitate sensitivity studies in criticality calculations was used. This package is based on MCNP5 [1], and consists of a data transfer and analysis program written in Perl, and a set of pseudo inputs for the transport and storage systems including the fuel assembly. The pseudo inputs contain key words that are replaced by a given value for the specific parameter. In this paper, the package is described in more detail including the new modifications needed for this study. The usability will be demonstrated with a study of the fuel pool in the case of steam gathered under the support plates of the pool. In the new pool model the steam layer was added as a water layer with a very low density. The density of this layer, i.e. the steam density, could be varied as well as the thickness of the layer. This package was then used to study the reactivity effect of the moderator density, the steam density and the thickness of the steam layer. As can be expected the strongest effect was obtained with the moderator density. However, the steam density seemed not to affect the reactivity while the thickness of the layer had a small effect. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Impact of modeling choices on inventory and in-cask criticality calculations for FORSMARK 3 BWR spent fuel - 14441

Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational data available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10x10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Redox‐Mediated Electrochemical Regeneration of Spent LiFePO 4 Battery Cathodes

Direct recycling of lithium-ion battery cathodes offers considerable appeal over metallurgical approaches. Here, we demonstrate a mediated electrochemical method for direct regeneration of degraded LiFePO 4 (LFP). The approach uses a redox mediator, iron propylenediamine tetraacetate, that undergoes electrochemical reduction and is circulated through an external reservoir, where it supplies the electrons needed to regenerate LFP in the presence of Li + ions derived from LiOH oxidation. Rapid outer-sphere electron transfer is observed from the mediator to the degraded LFP material. This feature, together with good aqueous solubility of the mediator (0.3 M), supports current densities up to 100 mA/cm 2 , and this electrochemical recycling process is demonstrated on 100 g scale. 57 Fe Mössbauer spectroscopy is used to monitor the correction of structural defects in the degraded LFP, providing the basis for regeneration of LFP that matches the electrochemical performance of pristine LFP.

Electrochemical Relithiation↗
DOE OSTI2023

Noble gas constraints on spent fuel irradiation histories

Fission gas isotopic compositions are sensitive to a variety of reactor operating parameters that include the neutron flux, neutron fluence at different neutron energies, and operating temperature. Measurements of fission gas isotopic compositions thus have potential to constrain reactor simulations for nuclear forensics and safeguards applications. In this paper, we present Kr and Xe isotope measurements from a suite of samples obtained from locations that span the axial length a fuel pin with a well-characterized irradiation history and compare these data to spatially resolved reactor simulations. Here, we observed positive correlations between fluence sensitive isotopic ratios and burnup and between a flux sensitive ratio and power, although some discrepancies are observed between the measured data and model predictions. These differences may be due to simplifications in the model and/or inaccuracies in the cross sections. A much broader measurement to model comparison is required to better understand the discrepancies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗