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At least 613 records · Page 34
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Removal of High Specific Activity Fission Products from Uranyl Sulfate Waste Solutions

The Savannah River National Laboratory (SRNL) is currently providing support to SHINE Medical Technologies (SHINE) which plans to deploy a low energy, accelerator-based neutron source to fission low enriched U in a uranyl sulfate target solution for 99 Mo production. The 99 Mo is initially separated from the fission products and target solution by an extraction column. Subsequent washing of the column will generate waste solutions containing residual U and fission product activity. A small number of high specific activity fission products (e.g., 90 Sr, 137 Cs, and 144 Ce) in these streams will likely control the classification of the low level waste (LLW). If a sufficient amount of the high specific activity isotopes are separated from the SHINE waste streams and concentrated into a waste form, it would be possible to treat a majority of the wash solutions from the column operations as a lower class of LLW (Class A versus Class B or C or Class B versus Class C). The high specific activity fission product elements could then be dispositioned as a much smaller volume of waste rather than requiring the disposal of the entire waste stream at the higher waste classification. The Savannah River Site (SRS) has experience with using monosodium titanate (MST) and crystalline silicotitanate (CST) to remove Cs and Sr from high salt content waste solutions generated during the reprocessing of nuclear fuels and targets. Both of these materials have worked very well for their intended purposes at the SRS where the fission product elements are removed from highly alkaline waste. On the other hand, SHINE waste streams from the extraction column contain H 2 SO 4 which makes the solution acidic. Additionally, the SRS waste streams do not contain other fission product elements such as transition metals or lanthanides because they precipitate upon neutralization of the SRS waste and are not present in the supernate which is dispositioned as LLW following treatment. As such, there are inherent differences between SHINE and SRS waste treatment strategies. Savannah River National Laboratory was tasked with performing scoping studies to see if MST and CST would remove Sr, Cs, and Ce from an acidic mixed metal simulant solution. Batch contact experiments were performed using MST and two CST type materials. The MST material is a 15 wt % powder in 0.15 M NaOH slurry. The MST showed low adsorption for elements of interest from acidic solution. Furthermore, the powder size makes MST non-ideal for column operations. A CST IE-911 ion exchange material had high Cs adsorption, moderate Sr, and marginal Ce adsorption. Based on adsorption of all species, the ion exchange capacity was found to be 0.032 meq/mL. A bench-top column experiment to measure elemental breakthrough curves was performed using CST IE-911 where chromatographic separations of the mixed simulant were expected to occur. While most elements behaved as expected, the lanthanide series, containing Ce, broke though the column earlier than expected. The second CST material, CST R9120, displayed high adsorption for all elements in the acidic mixed simulant solutions in a batch contact study, and had a calculated loading capacity of 0.091 meq/mL. Future studies to develop a waste treatment flowsheet should focus on CST R9120 to treat SHINE waste solutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Status Report on Development of a Cask to Enable Pulsed Neutron Characterization of Irradiated Fuel

We present the design of a shielding cask that allows pulsed neutron characterization of irradiated fuel rodlets prior to their destructive examination in hot cells. The goal is to provide complementary and informative measurements that will inform subsequent hot cell PIE examinations by identifying typical and atypical regions with respect to microstructure, tomographic data, or isotope densities. Time-of-flight neutron diffraction has the potential to offer efficient, non-destructive and non-contact microstructural characterization of irradiated fuel pins with spatial resolution of 1 mm3 to 1 cm3 while for energy-resolved neutron imaging (and by extension tomography) a resolution of 100 μm was demonstrated. The potential results include crystallinity vs. amorphous volumes and microstructural information such as phase compositions and textures from the diffraction data as well as distances (e.g. pellet to cladding), cracks, and isotopic distributions of minor actinides, fission products as well as fission gas partial pressures e.g. in the plenum from energy resolved neutron imaging. The capability will identify regions of atypical behavior and provide cost effective bulk characterization of entire samples prior to destructive evaluation in hot cells. This effort is a collaboration between LANL, INL, ORNL, and UC Berkeley. A key facet of the initiative is the ability to cost effectively ship irradiated fuels from INL to the pulsed spallation neutron source at LANSCE. The irradiated fuel would be loaded in the custom designed cask (designated RaMHaM) at INL hot cells which then could be shipped in a BRR Type B shipment container between INL and LANL. No hot cell is required after the loading of the sample at INL in the pathway laid out in this report, greatly simplifying handling at LANL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Oak Ridge Response to Versatile Test Reactor Environmental Impact Statement Data Request

The Versatile Test Reactor (VTR) is a fast-spectrum test reactor being developed in the United States under the direction of the US Department of Energy Office of Nuclear Energy (DOE-NE). The VTR mission is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. The conceptual design of the 300 MWth sodium-cooled metallic-fueled pool-type fast reactor has been led by the US National Laboratories in collaboration with General Electric–Hitachi and Bechtel National, Inc. In support of the VTR project, DOE issued a Notice of Intent (NOI) in the Federal Register on August 5, 2019, announcing the intent to prepare an Environmental Impact Statement (EIS) in accordance with the National Environmental Policy Act (NEPA) and its implementing regulations. The EIS will evaluate alternatives for a versatile reactor–based fast-neutron source facility and associated facilities for the preparation, irradiation, and post-irradiation examination (PIE) of test/experimental fuels and materials. Specifically, the NOI identified two siting alternatives for the VTR reactor facility: Idaho National Laboratory (INL) or Oak Ridge National Laboratory (ORNL). In addition, the NOI also specified two siting alternatives for VTR fuel fabrication: INL and the Savannah River Site (SRS). This report provides information in response to data requests made to ORNL to fill in site-specific knowledge gaps to develop a high-quality EIS. The responses provided are not required to provide full details in every aspect; instead, they adequately bound possible environmental impacts or provide sufficient information to adequately assess likely environmental impacts. This work is being performed under a subcontract from INL to ORNL using DOE-NE funds and is directed by DOE-NE and DOE-ID. Leidos has been contracted by DOE-NE to write the VTR EIS, so most data requests have come from Leidos but were often routed through INL or DOE-ID. DOE-ID is overseeing the NEPA and EIS processes for the VTR project. Leidos will use the information provided in this report to inform the VTR EIS and will also cite this document to establish a clear, publicly available source of the information. Section 2 of this report briefly describes the proposed ORNL VTR Alternative and illustrates the location of the proposed site for the ORNL VTR Alternative. Sections 3 through 7 provide direct responses to data requests received by ORNL. These sections use a tabular format in which data requests are divided into separate items to be addressed; the items are numbered, the data requests are restated with more topical information included, and then the responses are provided. Initial data requests and follow-on requests for additional information (RAIs) are combined under the original data request fields. Finally, Section 8 presents summarized conclusions and describes future work.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Tomographic Image Reconstruction for the Parallel-Slit Ring Collimator Fast Neutron Emission Tomography System

For the past three years, Oak Ridge National Laboratory (ORNL) has been developing a passive fast-neutron emission tomography capability. The goal of this development is the ability to quantify the neutron source strength of individual fuel pins (rods) in spent nuclear fuel assemblies. Such a system could be used to measure the burnup of each fuel pin in a spent fuel assembly in order to take burnup credit when loading dry storage casks or to count individual fuel pins in spent fuel assemblies for safeguards purposes. At present, a laboratory prototype imager is under construction. The purpose of this prototype is to demonstrate imaging capability sufficient to resolve individual fuel pins in spent fuel assemblies. This report documents the development of the iterative reconstruction code used to perform tomographic image reconstruction, the imager response calculation used by the reconstruction code, and the results of reconstructions of simulated tomographic imaging measurements for the prototype imager design.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Fundamental Neutron Physics: Theory and Analysis

The research topic of the project focus on the theoretical study of the possibility to search for manifestations of new physics beyond the Standard model in fundamental neutron physics, with emphasis on fundamental symmetries and "exotic'' interactions. The mail goals concentrate on the search for Time Reversal Invariance Violation (TRIV) in neutron nuclei scattering, and on possible manifestations of new physics in neutron-antineutron oscillations and their interactions with nuclei. Theoretical methods for analysis and interpretations of TRIV parameters in neutron scattering with the possible background interactions, and for unambiguous separation of the effects related to manifestations of new physics have been developed. The obtained results will be applied for the development of experimental programs in fundamental neutron physics at the Spallation Neutron Sources (SNS) in USA, Japan, and Europe.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Involute Working Group – Validation of CFD Turbulence Models for Steady-State Safety Analysis of Corner Geometry

Worldwide, three research reactors have fuel plates curved as a circle-involute (a spiral generated around a circle): Oak Ridge National Laboratory High Flux Isotope Reactor located in Tennessee, U.S.A. [1], Institut Laue-Langevin High Flux Reactor located in Grenoble, France [2], and Technical University of Munich Research Neutron Source Heinz Maier-Leibnitz located in Garching, Germany [3]. All three reactors are currently using highly enriched uranium ( 235 U/U ≥ 20 wt%) as fuel, and all three are actively engaged in activities to convert to low-enriched uranium ( 235 U/U < 20 wt%) fuel.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Hydrogen Diffusivity Measurements of YH 1.87 Moderator Material with Incoherent Quasielastic Neutron Scattering

Yttrium hydride is an optimal choice of moderator material for thermal neutron spectrum reactors requiring small core volumes. The Transformation Challenge Reactor (TCR) program at Oak Ridge National Laboratory (ORNL) aims to develop an additively manufactured microreactor core by leveraging recent advances in materials, manufacturing, data analytics, and high-fidelity modeling and simulation. Yttrium hydride was selected as the moderator material for TCR due to its high moderating power, which enables the spatially efficient reactor design and thermal stability that is superior when compared with zirconium hydride. Hydrogen desorption from yttrium hydride is expected at elevated temperatures, posing safety and performance concerns. As a lifetime component in advanced reactors, appropriate strategies to mitigate hydrogen release from yttrium hydride over long-term reactor operation are required. Understanding hydrogen mobility in yttrium hydride provides the scientific basis for developing a hydrogen barrier for hydride moderator and predicting the hydrogen redistribution within the material during various operational conditions. This study investigated the hydrogen motion in YH 1.87 as a function of temperature ranging from 200 to 1,173 K using incoherent quasielastic neutron scattering (IQNS) at the Spallation Neutron Source (SNS). The results are presented and discussed in terms of hydrogen self-diffusion coefficients, activation energy for translational hydrogen motion, and hydrogen jump distances. Based on the IQNS data, YH 1.87 is found to be stable at high temperatures with no translational H motion below 1,023 K within the IQNS detection limits. In the temperature range of 1,073–1,173 K, hydrogen diffusivity in YH 1.87 is $D = 4.57 X 10^{⁻3} [m^2/s] exp (⁻\frac{1.73[eV]}{RT}$). The hydrogen concentration’s dependence on hydrogen diffusivity is also discussed in this report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Time-of-Flight Calculations with Multiple Beam Phase Monitors: Calibration, Jitter Analysis & Energy Measurement

This report describes time-of-flight (ToF) beam energy measurements with multiple phase monitors in a beam line with no acceleration. Measurement uncertainty is quantified via rigorous treatment of beam-based phase monitor calibration and jitter analysis of the accelerator system. Experimental results in the Superconducting Linac (SCL) of the Spallation Neutron Source (SNS) are presented.

43 PARTICLE ACCELERATORS↗
DOE OSTI2020

Conceptual design for a blanket tritium extraction test stand

Tritium breeding is fundamentally required for a sustainable fusion fuel cycle, yet the technological readiness of blanket technology lags far behind other fusion systems. Breeder concepts are divided into solid and liquid media, where solid breeders typically rely on a sweep gas, such as helium, to carry away tritium from lithium containing ceramic materials, and liquid breeders produce tritium from lithium containing eutectics (e.g., PbLi) or molten salts (FLiBe). In each case, tritium must be harvested from the breeding medium. A promising method for tritium extraction is through a vacuum permeator, in which a concentration gradient from the tritium- containing fluid promotes diffusion through a membrane with high hydrogen permeability to the vacuum. This technology has been demonstrated for hydrogen gas systems using Pd and PdAg permeators, but relatively little work has been done to test tritium extraction from PbLi. A Tritium Extraction eXperimental (TEX) loop is being designed to test tritium extraction in a vacuum permeator configuration. The system design is such that it will allow the testing of tritium extraction from both helium and PbLi. A phased approach is being taken that will allow testing of small specimens for fundamental permeation measurements, to multi-meter component testing at near-prototypic conditions. A molten PbLi loop is challenging due to the toxic and explosive nature of Pb and Li, respectively, radiological concerns by introducing tritium, and high temperatures involved in such a system. In addition, PbLi corrosion is a significant issue at high temperatures (>400C). The TEX system will not employ a neutron source for volumetric production of tritium. Herein we present the design and methods used for 1) pumping PbLi, 2) introducing deuterium or tritium into the PbLi, 3) quantifying the amount of deuterium in the loop, 4) extracting deuterium and tritium, and 5) quantifying the total amount of extracted deuterium or tritium from the permeator. In addition, the safety design for operating such a system will be discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Second Target Station Computer Science and Math Workshop Report

Discovery science drives innovation and underpins the technological advances that will solve some of society’s most challenging issues, including clean energy technologies, better medicines, safe potable water, and addressing aging infrastructures, including transportation. Many of these advances will result from basic research into new materials and new ways to optimize our use of existing materials. Oak Ridge National Laboratory’s neutron sources provide cutting-edge scientific tools to probe the structure and dynamics of matter in unique ways. This insight is fundamental to advancing our ability to discover, design, control, and use new materials to address society’s most pressing needs.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2023

Volume 4: Detection Systems and Ultra-Cold Neutrons

As part of the Sustaining and Enhancing Nuclear Science (SENSe) initiative at Oak Ridge National Laboratory (ORNL), the prospect of adding new detection systems to support High Flux Isotope Reactor (HFIR) operations and to advance scientific research has prompted many ideas and discussions regarding potential features, configurations, locations, and applications. A working group of ORNL staff members was organized to further develop the concepts and recommending one or more configurations to best support future HFIR operations and scientific capacities and to provide order-of-magnitude cost estimates and timing. The areas of investigation included fast access detection systems, non-scattering beamline instruments, shielded detection instruments, and an ultracold neutron source. In each case, the focus was to develop world leading capabilities that would be unmatched by any other facility.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Optical Simulations for the VENUS Neutron Imaging Instrument

The VENUS Neutron Imaging instrument at the Spallation Neutron Source has gone through a multitude of optical design changes since its original conception. The goal of this report is to quantify the performance of the most recent optical design and provide insight into potential improvements to this design as needed to ensure optimal performance of the instrument. The results will come in the form of a refined and well developed simulation that provides as best a representation of the performance of the instrument as reasonably achievable. This simulation will be available in a repository alongside relevant simulation output needed to inform any future decisions on the design of the instrument.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Uncertainty Aware Deep Learning for Particle Accelerators

Standard deep learning models for classification and regression applications are ideal for capturing complex system dynamics. However, their predictions can be arbitrarily inaccurate when the input samples are not similar to the training data. Implementation of distance aware uncertainty estimation can be used to detect these scenarios and provide a level of confidence associated with their predictions. In this paper, we present results from using Deep Gaussian Process Approximation (DGPA) methods for errant beam prediction at Spallation Neutron Source (SNS) accelerator (classification) and we provide an uncertainty aware surrogate model for the Fermi National Accelerator Lab (FNAL) Booster Accelerator Complex (regression).

Rajput, Kishansingh↗
DOE OSTI2023

Evaluated Nuclear Data Requirements in Support of Fission and Fusion Applications [Slides]

This presentation touches on thermal scattering cross section and scattering lengths including criticality benchmarks (fission). Secondly, development and completeness of charged ­particle libraries: criticality benchmarks (fission), neutron source (fission/fusion), simulation benchmarks and light elements generation (fusion). Additionally, this presentation details on the covariances (methodology and completeness) with adjustment and optimization of nuclear data libraries (fission/fusion). Finally nuclear data reproducibility and conclusions are discussed.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

CINDER90 and CINDER2008 Comparison for Second Target Station Analysis

A preliminary target activation analysis for the Second Target Station (STS) at the Spallation Neutron Source (SNS) located at Oak Ridge National Laboratory (ORNL) was preformed using Monte Carlo N-Particle X (MCNPX), CINDER90, and the corresponding multi-group cross section libraries distributed with CINDER90. This preliminary analysis has been used to guide some design decisions of the STS target design. Using the MCNPX output files from this preliminary analysis, a comparison of the activation analysis using CINDER90 and CINDER2008 was performed. While the CINDER90 code has been validated for use in spallation source activation and transmutation calculations, and compares well with measurements, the transition to the modernized version of the code (CINDER2008) and the updated cross section libraries is desirable. The goal of this comparison is to evaluate the differences and implications for the STS activation analysis. Overall, the total activity differs between CINDER2008 and CINDER90 by no more than 20% for the materials evaluated in this report. The largest discrepancies were observed in the decay gamma intensity densities with CINDER90 calculating a factor of 2 higher than CINDER2008.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2024

A High-level Design for Bidirectional Data Streaming to High-Performance Computing Systems from External Science Facilities

Cutting-edge science is increasingly data-driven due to the emergence of scientific machine learning models that can guide scientists toward fruitful areas of exploration. Experimental science facilities such as light and neutron sources, particle colliders, and radio astronomy telescopes are also producing raw measurement data at rates that exceed available data storage and computing capacity at those facilities. As a result, scientific workflows are being developed that concurrently couple experiments at science facilities with high-performance computing (HPC) facilities to enable analysis of experimental data while the experiment is ongoing, and where analysis results are potentially fed back to the experiment for use in experimental control and/or steering in a time-sensitive manner. Our goal is to design, prototype, and deploy a new capability for the Oak Ridge Leadership Computing Facility (OLCF) that enables such workflows through support for bidirectional, memory-based streaming of data from external experiments into and out of OLCF HPC systems. This high-level design document describes the related work and motivating use cases that inform our understanding of the technical requirements for this capability, and describes a proposed architectural solution that meets these requirements and our plans for demonstrating the capability.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2024

Analysis of Second Target Station Target Segment Removal Dose Rates

This report documents the analysis of the dose rate fields in two configurations of the target system of the Second Target Station (STS) at the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL) in support of the remote handling target segment removal operation. The two configurations detailed in this report are both assuming 8 hours since beam was incident on the target. The first configuration assumes all shielding is in place just as it was during operation and allows for confirmation of hands-on maintenance capability of the target drive components. The second configuration assumes that a hatch in the target drive room ceiling has been opened to the high bay, the target segment removable core vessel shield block has been removed, and there is a direct line-of-sight to the target segment some 5 m below the target drive room elevation. Dose rate fields in both configurations are documented in this report along with details of the methods of computation. The dose rate fields included in this report are intended to support the engineering of remote handling special equipment and the planning of remote handing operations. The models and analysis detailed in this report are in support of the preliminary design of the STS and future analysis is needed to support the final design of the STS.

61 RADIATION PROTECTION AND DOSIMETRY↗