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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 847 records · Page 47
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

CASMO5/TSUNAMI-3D spent nuclear fuel reactivity uncertainty analysis

The CASMO5 lattice physics code is used in conjunction with the TSUNAMI-3D sequence in ORNL's SCALE 6 code system to estimate the uncertainties in hot-to-cold reactivity changes due to cross-section uncertainty for PWR assemblies at various burnup points. The goal of the analysis is to establish the multiplication factor uncertainty similarity between various fuel assemblies at different conditions in a quantifiable manner and to obtain a bound on the hot-to-cold reactivity uncertainty over the various assembly types and burnup attributed to fundamental cross-section data uncertainty. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Management of Legacy Spent Nuclear Fuel Wastes at the Chalk River Laboratories: The Challenges and Innovative Solutions Implemented - 13301

AECL's Fuel Packaging and Storage (FPS) Project was initiated in 2004 to retrieve, transfer, and stabilize an identified inventory of degraded research reactor fuel that had been emplaced within in-ground 'Tile Hole' structures in Chalk River Laboratories' Waste Management Area in the 1950's and 60's. Ongoing monitoring of the legacy fuel storage conditions had identified that moisture present in the storage structures had contributed to corrosion of both the fuel and the storage containers. This prompted the initiation of the FPS Project which has as its objective to design, construct, and commission equipment and systems that would allow for the ongoing safe storage of this fuel until a final long-term management, or disposition, pathway was available. The FPS Project provides systems and technologies to retrieve and transfer the fuel from the Waste Management Area to a new facility that will repackage, dry, safely store and monitor the fuel for a period of 50 years. All equipment and the new storage facility are designed and constructed to meet the requirements for Class 1 Nuclear Facilities in Canada. (authors)

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DOE OSTI2013

Discussing spent nuclear fuel in high school classrooms: addressing public fears through early education

The Inreach program combines the Deep River Science Academy (DRSA) 'learning through research' approach with state of the art communication technology to bring scientific research to high school classrooms. The Inreach program follows the DRSA teaching model where a university student tutor works on a research project with scientific staff at AECL's Chalk River Laboratories. Participating high school classes are located across Canada. The high school students learn about the ongoing research activities via weekly web conferences. In order to engage the students and encourage participation in the conferences, themed exercises linked to the research project are provided to the students. The DRSA's Inreach program uses a cost-effective internet technology to reach a wide audience, in an interactive setting, without anyone leaving their desks or offices. An example Inreach research project is presented here: an investigation of the potential of the Canadian supercritical water cooled reactor (SCWR) concept to burn transuranic elements (Np, Pu, Am, Cm) to reduce the impact of used nuclear fuel. During this project a university student worked with AECL (Atomic Energy of Canada Limited) researchers on technical aspects of the project, and high school students followed their progress and learned about the composition, hazards, and disposition options for used nuclear fuel. Previous projects included the effects of tritium on cellular viability and neutron diffraction measurement of residual stresses in automobile engines.

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DOE OSTI2012

New Design for an HLW Repository (for Spent Fuel and Waste from Reprocessing) in a Salt Formation in Germany - 12213

In autumn 2010, after a 10-year moratorium, exploration was resumed in Gorleben, the potential site for a German HLW repository. At the same time, the Federal Government launched a two-year preliminary safety analysis to assess whether the salt dome at Gorleben is suitable to host all heat-generating radioactive waste generated by German NPPs based on the waste amounts expected at that time. The revised Atomic Energy Act of June 2011 now stipulates a gradual phase-out of nuclear energy production by 2022, which is 13 years earlier than expected in 2010. A repository design was developed which took into account an updated set of data on the amounts and types of expected heat-generating waste, the documented results of the exploration of the Gorleben salt dome, and the new 'Safety Requirements Governing the Final Disposal of Heat-Generating Radioactive Waste' of 30 September, 2010. The latter has a strong influence on the conceptual designs as it requires that retrievability of all waste containers is possible within the repository lifetime. One design considered that all waste containers will be disposed of in horizontal drifts of a geologic repository, while the other design considered that all waste containers will be disposed of in deep vertical boreholes. For both options (emplacement in drifts/emplacement in vertical boreholes), the respective design includes a selection of waste containers, the layout of drifts, respectively lined boreholes, a description of emplacement fields, and backfilling and sealing measures. The design results were described and displayed and the differences between the two main concepts were elaborated and discussed. For the first time in both repository designs the requirement was implemented to retrieve waste canisters during the operational phase. The measures to fulfill this requirement and eventually the consequences were highlighted. It was pointed out that there arises the need to keep transport- and storage casks in adequate numbers and interim storage facilities available until the repository is closed. (authors)

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DOE OSTI2012

Operational Challenges of Extended Dry Storage of Spent Nuclear Fuel - 12550

As a result of the termination of the Yucca Mountain used fuel repository program and a continuing climate of uncertainty in the national policy for nuclear fuel disposition, the likelihood has increased that extended storage, defined as more than 60 years, and subsequent transportation of used nuclear fuel after periods of extended storage may become necessary. Whether at the nation's 104 nuclear energy facilities, or at one or more consolidated interim storage facilities, the operational challenges of extended storage and transportation will depend upon the future US policy for Used Fuel Management and the future Regulatory Framework for EST, both of which should be developed with consideration of their operational impacts. Risk insights into the regulatory framework may conclude that dry storage and transportation operations should focus primarily on ensuring canister integrity. Assurance of cladding integrity may not be beneficial from an overall risk perspective. If assurance of canister integrity becomes more important, then mitigation techniques for potential canister degradation mechanisms will be the primary source of operational focus. If cladding integrity remains as an important focus, then operational challenges to assure it would require much more effort. Fundamental shifts in the approach to design a repository and optimize the back-end of the fuel cycle will need to occur in order to address the realities of the changes that have taken place over the last 30 years. Direct disposal of existing dual purpose storage and transportation casks will be essential to optimizing the back end of the fuel cycle. The federal used fuel management should focus on siting and designing a repository that meets this objective along with the development of CIS, and possibly recycling. An integrated approach to developing US policy and the regulatory framework must consider the potential operational challenges that they would create. Therefore, it should be integral to these efforts to redefine retrievability to apply to the dual purpose cask, and not to apply to individual assemblies. (authors)

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DOE OSTI2015

Expected count rate for the Self- Interrogation Neutron Resonance Densitometry measurements of spent nuclear fuel

The Self-Interrogation Neutron Resonance Densitometry (SINRD) is a passive neutron technique that aims at a direct quantification of {sup 239}Pu in the fuel assemblies by measuring the attenuation of the neutron flux in the energy region close to the 0.3 eV resonance of {sup 239}Pu. The {sup 239}Pu mass is estimated by calculating the SINRD signature, that is the ratio between the neutron flux integrated over the fast energy region and around the 0.3 eV resonance region. The SINRD measurement approach considered in this study consists in introducing a small neutron detector in the central guide tube of a PWR 17x17 fuel assembly. In order to measure the neutron flux in the energy regions defined in the SINRD signature, different detector types are used. The response of a bare {sup 238}U fission chamber is considered for the determination of the fast neutron flux, while other thermal-epithermal detectors wrapped in neutron absorbers are envisaged to measure the neutron flux around the resonance region. This paper provides an estimation of the count rate that can be achieved with the detector types proposed for the SINRD measurement. In the first section a set of detectors are evaluated in terms of count rate and sensitivity to the {sup 239}Pu content, in order to identify the optimal measurement configuration for each detector type. Then a study is performed to increase the count rate by increasing the detector size. The study shows that the highest count rate is achieved by using either {sup 3}He or {sup 10}B proportional counters because of the high neutron efficiency of these detectors. However, the calculations indicate that the biggest contribution to the measurement uncertainty is due to the measurement of the fast neutron flux. Finally, similar sensitivity to the {sup 239}Pu content is obtained by using the different detector types for the measurement of the neutron flux close to the resonance region. Therefore, the count rate associated to each detector type will play a major role in the selection of the detector types used for the SINRD measurement. (authors)

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DOE OSTI2022

A Framework to Assess Advanced Reactor Spent Fuel Management Facility Deployment

Previous planning and prioritization for LWR SNF management investigated the risks and uncertainties of deploying facilities such as consolidated interim storage [1, 2, 3, 4]. As part of that work, activities and milestones were collected into success precedence diagrams that charted a path to achieving facility deployment [1]. In that framework, activities are any research, development, design, or decision required to achieve an intermediate goal; milestones are activity endpoints and mark the completion of deliverables. Milestones can be thought of as achievements required to reach the final goal of facility deployment; activities are the means by which milestones are accomplished. In planning, activities and milestones are compiled into comprehensive flow charts that visualize the steps necessary for deployment. This framework has been used to quantify risks, timelines, and costs of deploying SNF management facilities.

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DOE OSTI2017

Spent Nuclear Fuel Disposition at the Savannah River Site - 17153

For over forty years, the H-Canyon facility at the Savannah River Site (SRS) performed remotely operated radiochemical separations of irradiated targets and fuels to produce materials for national defense. Although the materials production mission has ended, the facility continues to play an important role in the stabilization and safe disposition of proliferable nuclear materials. (authors)

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DOE OSTI2017

The Gd based Neutron Absorber for Spent Fuel Storage and Transport Cask - 17636

The criticality safety calculations were performed for integral model which has single rack wall of stainless steel added neutron absorber material. The integral model was verified by reference model composed of 3 layer; basket cell, neutron absorber and sheath. For the neutron absorber material, Gd was selected due to its high neutron absorption cross section in thermal region compared to B. The mathematical model based on validated code, KENO-Va in SCALE was used and the data library is ENDF/B-VII 238-Group. The k{sub eff} was calculated for changes in Gd and B contents in flooded condition. The Gd added stainless steel shows higher performance than B at same atomic percent and reactivity reduction decreases as absorber content increases. Therefore, the Gd 0.47 at% was selected for 'optimum Gd content' for effective use of neutron absorber. With Gd 0.47 at%, B, Sm, Er and Eu was considered for additional neutron absorber and B shows the best performance among them. The Gd 0.47 at% with B 2.05 at% was selected for (optimum neutron absorber content'. (authors)

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DOE OSTI2015

Issue focused stakeholder collaboration: reaching consensus on spent nuclear fuel transportation operational practices - 15465

This paper examines stakeholder and issue-focused approaches used by Department of Energy (DOE) programs to collaborate with key stakeholders on transportation practices to ship radioactive material, particularly used nuclear fuel (UNF) and transuranic waste (TRU). In accordance with the Administration's Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste, the current effort by DOE to plan for the development of a future transportation system to ship UNF from commercial reactors, with an initial focus on shutdown reactors, is using approaches similar to those used in other successful shipping campaigns to collaborate with stakeholders about sustainable transportation practices. As part of a coordinated approach to stakeholder outreach and collaboration, DOE is currently developing a draft Transportation Planning Framework to provide a template for future shipments of UNF from commercial reactors and is engaged with states and tribes and other stakeholders in discussions on the institutional and operational components of that Framework. (authors)

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DOE OSTI2016

A Parametric Analysis of Factors Affecting Calculations of Estimated Dose Rates from Spent Nuclear Fuel Shipments - 16458

Planning for transportation of SNF requires estimation of the radiation dose to the members of the public associated with the transportation. This dose is expected to be very low compared to natural background radiation because dose rates and exposure durations are minimal. A preliminary investigation was performed to estimate the dose rate from a SNF transportation package as a function of distance and evaluate factors affecting the estimated dose rate, such as local environment modeling. Effects on the dose rate are determined for changes in temperature and relative humidity of the air surrounding a transportation package and for different compositions of the ground underneath it. The impact of different vapor content in air was evaluated to be small, within the statistical uncertainty of the dose rate estimates. The neutron dose rate contributions from scattering interactions within different soil/pavement materials were from approximately 20% to 70% at 25 cm from the ground and from approximately 10% to 40% at 175 cm from the ground. The effect of photon scattering interactions within different soil/pavement materials was an increase in photon dose rate of less than 20% at 25 cm from the ground and less than 10% at 175 cm from the ground. The impact of different ground materials on the total dose rate would depend on the relative contributions of photon and neutron dose rates to the total dose rate. (authors)

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DOE OSTI2016

Separation of Zirconium from Zircaloy-2 for Spent Nuclear Fuel Cladding Applications using a Chloride Volatility Process

The United States produces a large amount of cladding material from used nuclear fuel, expected to approach 1,000 MT/year for the next 50 years. The recovery of zirconium from the cladding material would represent a ∼50% savings on the geologic waste disposal requirement, and savings of >$40 M/year due to the cost of hafnium-free zirconium. Previous simulation studies have shown that in the irradiated cladding, the major activation products found are {sup 125}Sb, {sup 60}Co, and {sup 93m}Nb, and the major fission product found is {sup 137}Cs. These radioisotopes make it impossible to treat the cladding as low level waste. In a chlorination method, zirconium cladding reacts with Cl{sub 2} at 350-380 deg. C to produce mainly ZrCl{sub 4}. The zirconium tetrachloride has a relatively low boiling point (sublimes at 330 deg. C) which allows its separation from other products, and so it can be treated as low level waste for disposal or recycling. This chlorination approach has been proposed for the reprocessing of zirconium. However, previous work has observed element impurities (Fe, Sn, Cr, Cs, Sb, and Nb) in the recovered ZrCl{sub 4} meaning that additional steps are necessary in a chlorination process previous to the reduction of ZrCl{sub 4} to the metallic form. The chlorination process for zirconium reprocessing has been highly encouraged as it has proven more efficient than previous mechanical processes, redox processes, and other chemical methods such as iodination. The overall goal of this project is to develop the road map for a chlorination process that will yield high purity zirconium. The current study focuses on the recovery of ZrCl{sub 4} from non-irradiated alloys used as nuclear fuel cladding. (authors)

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DOE OSTI2019

Use of In Situ Acoustic Backscatter Systems to Characterize Spent Nuclear Fuel and its Separation in a Thickener - 19288

To aid in the transportation, dewatering and storage of radioactive UK legacy waste sludges at Sellafield Ltd., the application of an Acoustic Backscatter System (ABS) was investigated in this study, as a method for monitoring suspended solids concentration and consolidation of cohesive and aggregated sludges (that form the basis of many wastes at Sellafield). An ABS represents a particularly promising technology, as they can be used with minimal intrusion, to measure both particle size and concentration depending on prior system knowledge, through inversion of the return echo voltage response of pulse-echo signals in the MHz range. To assess the application for continuous, high concentration (∼30 % v/v) thickened wastes, an ABS was utilised to characterise a pseudo-steady state laboratory thickener over hours, using flocculated calcite as a representative test material. Measurements were taken both horizontally through a consolidated bed using a 1 MHz transducer and vertically downwards with a 2 MHz transducer to measure sediment concentration above the settled bed and track bed height via the interface reflection. Characterisation of flocculated aggregate diameters was initially determined to be ∼510 μm, using an in-line Focussed Beam Reflectance Measurement (FBRM) device. The submerged 2 MHz probe indicated that there were only limited changes to the dispersion concentration or size over time. Changes in settled bed density, due to increasing bed height and changes in the underflow flow rate, were qualitatively measured via differences to the acoustic signal attenuation over time. An initial increase in attenuation with bed height was observed as the bed densified due to local compression. Once an equilibrium bed height was reached, the acoustic attenuation remained fairly constant with time, although the bulk underflow density was reduced with the increase in underflow rate. The solids residence time in the bed, determined from a transitive volume-balance model, was found to be ∼3200 s for pseudo-steady state operation at a maintained bed height of ∼0.175 m. Results of the study highlight the potential the ABS as a remote process monitoring tool for both relatively dilute suspensions and concentrated thickened mineral sludges, with potential applications across waste processing sites at Sellafield. (authors)

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