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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 757 records · Page 42
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Metallographic examinations and hydrogen measurements of high-burnup spent nuclear fuel cladding

In the US, commercial spent nuclear fuel (SNF) is transferred to interim dry storage casks where it will be stored for decades awaiting transport to a consolidated interim storage facility or a geologic repository. Because the fuel rod cladding is the first barrier against any radioactive material release, understanding the behavior of SNF cladding, particularly at high burnup (HBU), in dry storage conditions is crucial to safely store and transport the spent fuel. In this study, a series of metallographic examinations and cladding hydrogen measurements were conducted on HBU SNF cladding at Oak Ridge National Laboratory as a part of the High Burnup Spent Fuel Data Project, which is sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE). Here, t o investigate the effect of simulated drying conditions on the cladding, three as-received fuel rods with different cladding materials—M5, ZIRLO, and Zircaloy-4—were heated to 400°C and then slow-cooled to room temperature. The pellet and cladding were then qualitatively and quantitatively analyzed and compared in terms of pellet crack morphology, HBU rim, waterside oxide, cladding H, and cladding hydride morphologies. This paper presents and discusses the results of these analyses in detail.

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

Evaluation of Hydrogen Gas Generation and Permeation in the 9979 Type AF Shipping Package

The 9979 Type AF Shipping Packaging is a cost-effective radioactive material package designed by Savannah River National Laboratory (SRNL) that consists of two primary components: a foamed outer drum for structural protection and an inner containment drum. The packaging was designed to transport Highly Enriched Uranium (HEU), Low Enriched Uranium (LEU), and other isotopes not exceeding a Type A quantity. These contents have the potential to generate flammable hydrogen gas during transport due to the degradation of hydrogenous materials (e.g. water vapor, plastics, etc) by high-energy alpha radiation. Since 10 CFR 71.43(h) prohibits the incorporation of packaging features explicitly designed for continuous venting, alternative justification was required to demonstrate that the requirements of 10 CFR 71.43(d) for negligible reactions between packaging components and contents were satisfied. An analysis was performed to demonstrate that the potential for hydrogen gas generation over a one-year period was limited by the effects of permeation through the packaging materials. The rate of hydrogen collection was evaluated for both the inner containment drum and the outer structural drum under 10 CFR 71.71 Normal Conditions of Transport (NCT). The analysis concludes that the Lower Flammability Limit (LFL) for hydrogen gas will not be reached in a one-year shipping period assuming a minimum void volume is maintained within the drum.

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

Radionuclide identification algorithm for organic scintillator-based radiation portal monitor

We have developed an algorithm for on-the-fly radionuclide identification for radiation portal monitors using organic scintillation detectors. The algorithm was demonstrated on experimental data acquired with our pedestrian portal monitor on moving special nuclear material and industrial sources at a purpose-built radiation portal monitor testing facility. The experimental data also included common medical isotopes. The algorithm takes the power spectral density of the cumulative distribution function of the measured pulse height distributions and matches these to reference spectra using a spectral angle mapper. F-score analysis showed that the new algorithm exhibited significant performance improvements over previously implemented radionuclide identification algorithms for organic scintillators. Reliable on-the-fly radionuclide identification would help portal monitor operators more effectively screen out the hundreds of thousands of nuisance alarms they encounter annually due to recent nuclear-medicine patients and cargo containing naturally occurring radioactive material. Lastly, portal monitor operators could instead focus on the rare but potentially high impact incidents of nuclear and radiological material smuggling detection for which portal monitors are intended.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

Generic HLW Repository In Bedded Salt In Germany: Emplacement Technologies and Repository Designs - 18288

In Germany, a new site selection act for a repository for HLW and SNF entered into force in July 2013 and was amended in spring 2017 [1]. It stipulates that the safety of repository systems for heat-generating radioactive waste and spent nuclear fuel in different host rock formations be compared. This requires that generic repository concepts and adequate safety- and safety demonstration concepts for all potential host rock formations in Germany (salt, clay and crystalline rock) exist. For several decades, development of disposal concepts and safety analyses for a repository in a salt dome was prioritized while repository concepts for clay and crystalline rock have been taken into account for two decades only. So far, bedded salt formations have not been the subject of comparative investigations. Thus, in summer 2015, the Federal Ministry for Economic Affairs and Energy (BMWi) launched the R and D project KOSINA to develop repository designs and carry out safety analyses for different emplacement alternatives in a generic bedded salt formation and a generic salt pillow. KOSINA is an acronym for 'Concept development for a generic repository for heat-generating waste in bedded salt formations in Germany as well as development and testing of a safety and demonstration concept'. Together with its partners BGR, GRS, and IfG, DBE Technology GmbH has successfully carried out an almost three years lasting scientific programme, which will be concluded in April 2018. The core areas of the programme are the development of generic geologic models, including derivation of model parameters, the development of a safety and safety demonstration concept, the development of technical repository designs for four different emplacement alternatives, the analysis of geomechanical integrity, the evaluation of operational safety as well as the analysis of radiological consequences. An interim report, which compiles the basic data of the waste inventory, the repository design requirements, the data basis for geologic model developments, and the outline of a safety and safety demonstration concept, was already published in February 2016. The main achievements of the R and D project KOSINA are the development of four different repository designs, two per each generic geologic model (flat-bedded salt, salt pillow) based on thermomechanical calculations. This includes the implementation of a safety concept that provides containment of the radioactive materials with regard to the technical designs of waste packages, transport and emplacement technologies, and backfilling and sealing concepts. Furthermore, the integrity of the geologic barriers could be demonstrated for all four repository designs, and the long-term predictions showed no radiological releases during the demonstration period of one million years. (authors)

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

Burnt Mountain Alaska RTG Extraction, Transport and Disposal Project - 17359

The Air Force Technical Applications Center (AFTAC), headquartered at Patrick Air Force Base (AFB), Florida, removed excess nuclear material from one of the center's seismic arrays on Burnt Mountain, Alaska, after nearly 15 years of un-operability, and transferred it to a the low-level waste disposal facility at the Nevada National Security Site (NNSS). Ten radioisotope thermoelectric generators (RTGs), were used as a power source by AFTAC at its array 97 kilometers (60 miles) north of the Arctic Circle. Originally placed on Burnt Mountain in 1973, the first RTG was installed to power a single seismometer. Twelve years later, due to the increase in communications equipment and demand for additional power, nine more RTGs were placed at the site. For the next 17 years, the RTGs were operated and maintained by AFTAC's Detachment 460 based at Eielson AFB near Fairbanks, Alaska, and were used because of their high reliability and low maintenance requirements to determine if regional seismic activity was caused by nuclear explosions or naturally-occurring events such as earthquakes, volcanic activity, etc. AFTAC's primary mission is to verify compliance with nuclear test ban treaties. In 1992, a tundra fire encroached on the Burnt Mountain site and damaged some of the data cables. Despite the fact the fire did not impact the generators themselves, it raised public concern among nearby inhabitants about the safety of using radioactive material as a power source at the station. The Air Force, in coordination with multiple agencies and mission partners, made the decision to remove the RTGs and relocate them to NNSS for permanent disposal. (authors)

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

2013 Annual Summary Report for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada National Security Site, Nye County, Nevada: Review of the Performance Assessments and Composite Analyses

The Maintenance Plan for the Performance Assessments and Composite Analyses for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada Test Site (National Security Technologies, LLC 2007a) requires an annual review to assess the adequacy of the performance assessments (PAs) and composite analyses (CAs), with the results submitted to the U.S. Department of Energy (DOE) Office of Environmental Management. The Disposal Authorization Statements for the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) also require that such reviews be made and that secondary or minor unresolved issues be tracked and addressed as part of the maintenance plan (DOE 1999a, 2000). The U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office performed an annual review of the Area 3 and Area 5 RWMS PAs and CAs for fiscal year (FY) 2013. This annual summary report presents data and conclusions from the FY 2013 review, and determines the adequacy of the PAs and CAs. Operational factors (e.g., waste forms and containers, facility design, and waste receipts), closure plans, monitoring results, and research and development (R&D) activities were reviewed to determine the adequacy of the PAs. Likewise, the environmental restoration activities at the Nevada National Security Site (NNSS) relevant to the sources of residual radioactive material that are considered in the CAs, the land-use planning, and the results of the environmental monitoring and R&D activities were reviewed to determine the adequacy of the CAs. Important developments in FY 2013 include the following: • Development of a new Area 5 RWMS closure inventory estimate based on disposals through. FY 2013 • Evaluation of new or revised waste streams by special analysis. • Development of version 4.115 of the Area 5 RWMS GoldSim PA/CA model.

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

Maintenance Plan for the Composite Analysis of the Hanford Site, Southeast Washington

The U.S. Department of Energy (DOE) manuals for radioactive waste management, DOE M 435.1-1 Chg 2 and DOE-STD-5002-2017, require that the Hanford Site maintain site performance assessments and composite analyses (CAs). This document describes the plan for maintaining the CA that supports waste disposal and remedial actions for the Hanford Site. An initial CA of the site was issued in 1998, conditionally approved in 1999, received further analysis to satisfy conditions in an addendum in 2001, and was approved in 2002. This document meets the maintenance plan requirements described in DOE M 435.1-1 Chg 2 and DOE-STD-5002-2017 and implements the requirements of the disposal authorization related to the CA for the U.S. Department of Energy, Richland Operations Office, the responsible field Office, and its contractors. In addition to providing current plans for revisions to the CA, this maintenance plan describes the work expected to be completed for ongoing reviews of new information; data related to residual sources of radioactive material, land use, and monitoring; and research and development related activities being conducted at the Hanford Site and within other U.S. Department of Energy programs. The main emphasis of this maintenance plan is to identify additional data and information to support the next revision of the Hanford Site CA. Improvements are expected regarding the knowledge of waste characterization and inventory, site-specific geotechnical data, and final disposition of major disposal facilities and waste sites.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Seismic Performance of Deeply Embedded Conceptual Advanced Reactors: Three-Dimensional Nonlinear Soil-Structure Interaction Analyses

The U.S. Nuclear Regulatory Commission’s functional containment concept provides advanced nuclear power plant designers with more flexibility in terms of the civil/structural design if the appropriate set of barriers for prevention of radioactive material release exist. Some of the conceptual advanced reactor structures, without the traditional pressure boundaries of large containment structures, are proposed to be deeply embedded or buried into soil. This approach is expected to provide (1) lesser seismic demands on the structures and safety-critical structures, (2) eased regulatory efforts and overall design against other external hazards such as aircraft impact, and (3) overall cost savings. Further, one of the important aspects of assessing the technical and economic viability of deeply embedding advanced reactor buildings is to assess the seismic performance with the understanding of effects with material and geometric nonlinearities. This study investigates the seismic response of deeply embedded or buried advanced reactors by conducting three-dimensional nonlinear soil-structure interaction analyses. Although the results indicate that there is a general trend of decreased seismic response with increased embedment depths, the change in the dynamic environment with different embedment depths and the nonlinear environment under high-intensity seismic inputs may result in increased peak response at increased embedment depths.

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

Quality quandaries: Statistical detective work to understand the isotopic ratios in Drum 68660 and the radioactive release at WIPP

An incident at the Department of Energy's Waste Isolation Pilot Plant (WIPP) in 2014 resulted in the release of radioactive material into the environment. Initially, it was known that at least one drum in WIPP, identified as drum 68660, was involved. However, questions remained. Could the air-monitor isotopic ratios measured in WIPP at the time of the release be explained by materials in drum 68660 or were other drums involved? Could internal conditions in drum 68660 have caused the breach? What were the implications for 68660's sister drum? These questions needed to be answered as quickly as possible. Here, this analysis, which was completed in three weeks, combined combinatorics and uncertainty analysis to provide scientists with the timely evidence they needed to either answer these important questions or to design experiments to answer them.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Waste Characterization Using Gamma Ray Spectrometry with Automated Efficiency Optimization - 13404

Gamma ray spectrometry using High Purity Germanium (HPGe) detectors is commonly employed in assaying radioactive waste streams from a variety of sources: nuclear power plants, Department of Energy (DOE) laboratories, medical facilities, decontamination and decommissioning activities etc. The radioactive material is typically packaged in boxes or drums (for e.g. B-25 boxes or 208 liter drums) and assayed to identify and quantify radionuclides. Depending on the origin of the waste stream, the radionuclides could be special nuclear materials (SNM), fission products, or activation products. Efficiency calibration of the measurement geometry is a critical step in the achieving accurate quantification of radionuclide content. Due to the large size of the waste items, it is impractical and expensive to manufacture gamma ray standard sources for performing a measurement based calibration. For well over a decade, mathematical efficiency methods such as those in Canberra's In Situ Object Counting System (ISOCS) have been successfully employed in the efficiency calibration of gamma based waste assay systems. In the traditional ISOCS based calibrations, the user provides input data such as the dimensions of the waste item, the average density and fill height of the matrix, and matrix composition. As in measurement based calibrations, the user typically defines a homogeneous matrix with a uniform distribution of radioactivity. Actual waste containers can be quite nonuniform, however. Such simplifying assumptions in the efficiency calibration could lead to a large Total Measurement Uncertainty (TMU), thus limiting the amount of waste that can be disposed of as intermediate or low activity level waste. To improve the accuracy of radionuclide quantification, and reduce the TMU, Canberra has developed the capability to optimize the efficiency calibration using the ISOCS method. The optimization is based on benchmarking the efficiency shape and magnitude to the data available in the analyzed gamma ray spectra. Data from measurements of a given item in multiple counting geometries are among the powerful benchmarks that could be used in the optimization. Also, while assaying a waste stream with fission products and activation products emitting gamma lines of multiple energies, optimizing the efficiency on the basis of line activity consistency is very effective. In the present paper, the ISOCS- based optimization methodology is applied to measurement scenarios involving multiple counting geometries, and multi-gamma-line radionuclides. Results will be presented along with accuracy and precision estimates for each measurement. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Atmospheric plume progression as a function of time and distance from the release point for radioactive isotopes

The International Monitoring System contains up to 80 stations around the world that have aerosol and xenon monitoring systems designed to detect releases of radioactive materials to the atmosphere from nuclear tests. A rule of thumb description of plume concentration and duration versus time and distance from the release point is useful when designing and deploying new sample collection systems. This paper uses plume development from atmospheric transport modeling to provide a power-law rule describing atmospheric dilution factors as a function of distance from the release point.

atmospheric plume progression↗
DOE OSTI2015

Nevada National Security Site Environmental Report 2014--Summary

This Nevada National Security Site Environmental Report (NNSSER) was prepared to satisfy DOE Order DOE O 231.1B, “Environment, Safety and Health reporting.” Its purpose is to (1) report compliance status with environmental standards and requirements, (2) present results of environmental monitoring of radiological and nonradiological effluents, (3) report estimated radiological doses to the public from releases of radioactive material, (4) summarize environmental incidents of noncompliance and actions taken in response to them, (5) describe the National Nuclear Security Administration Nevada Field Office Environmental (NSA/NFO) Management System and characterize its performance, and (6) highlight significant environmental programs and efforts. The three major NNSS missions include National Security/Defense, Environmental Management, and Nondefense. The major programs that support these missions are Stockpile Stewardship and Management, Nonproliferation and Counterterrorism, Nuclear Emergency Response, Strategic Partnership Project, Environmental Restoration, Waste Management, Conservation and Renewable Energy, Other Research and Development, and Infrastructure. The major facilities that support the programs include the U1a Facility, Big Explosives Experimental Facility (BEEF), Device Assembly Facility, Dense Plasma Focus Facility, Joint Actinide Shock Physics Experimental Research Facility, Radiological/Nuclear Countermeasures Test and Evaluation Complex, Nonproliferation Test and Evaluation Complex (NPTEC), Radiological/Nuclear Weapons of Mass Destruction Incident Exercise Site, Area 5 Radioactive Waste Management Complex (RWMC), and the Area 3 Radioactive Waste Management Site (RWMS).

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Hemicellulose and starch citrate chitosan foam adsorbents for removal of arsenic and other heavy metals from contaminated water

Arsenic and other heavy metal contaminants in water are a significant global health threat. In this study, low-cost, sulfur-free, sustainable, water-insoluble materials with heavy metal remediation properties were produced from renewable resources such as starch, xylan, citric acid, and chitosan. Synthesized starch citrate-chitosan (SCC) foam and xylan citrate-chitosan (XCC) foam were flexible, porous, and elastic. The foams’ arsenic uptake in water was significantly greater than five different commercial metal remediating agents. The mercury and lead uptakes with the synthesized foams were similar to the performance of a commercial sulfur-based product, SorbaTech 450 (ST450). However, the cadmium and selenium uptakes were comparatively lower. The complexation of arsenic with oxygen and nitrogen of the SCC foam was shown with time-of-flight secondary ion mass spectrometry (TOF-SIMS). The XCC foam was also shown to adsorb potassium iodide (KI) at a similar rate to sodium chloride. This may be used to remediate water contaminated with radioactive materials, such as iodine 131.

Materials Science↗
DOE OSTI2018

Degradation of ILW Drums Due to Internal Metallic Corrosion - 18036

Within the British nuclear inventory, metallic Intermediate Level Waste (ILW) is encapsulated in grout and stored in stainless steel drums. Based on recent inspections held in the relevant storage facilities (Sellafield, Cumbria), a proportion of these drums have exhibited considerable distortion, raising concerns about the long-term safety of the packages. Distortion may be caused by the generation of voluminous and potentially flammable substances resulting from metallic corrosion occurring within the drums' core. Potential oxygen influx in the event of steel fracture and consequently drum's failure could potentially cause release or even ignition of the encapsulated radioactive material. The mechanical degradation of the encapsulants (stainless steel and grout) induced from the progressive corrosion of the encapsulated metallic components (and the associated volume expansion) was investigated using both Finite Element Modelling (FEM) techniques as well as experimental trials. A parametric study using simplified 3D FE models was performed to determine their structural integrity under different scenarios. Synchrotron X-Ray tomography (XRT) was used to monitor the behaviour of miniaturized ILW drum-simulants, whilst the latter were subjected to controlled accelerated corrosion in situ. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

A DETAILED KINEMATIC MAP OF CASSIOPEIA A'S OPTICAL MAIN SHELL AND OUTER HIGH-VELOCITY EJECTA

We present three-dimensional (3D) kinematic reconstructions of optically emitting material in the young Galactic supernova remnant Cassiopeia A (Cas A). These Doppler maps have the highest spectral and spatial resolutions of any previous survey of Cas A and represent the most complete catalog of its optically emitting material to date. We confirm that the bulk of Cas A's optically bright ejecta populate a torus-like geometry tilted approximately 30 Degree-Sign with respect to the plane of the sky with a -4000 to +6000 km s{sup -1} radial velocity asymmetry. Near-tangent viewing angle effects and an inhomogeneous surrounding circumstellar material/interstellar medium environment suggest that this geometry and velocity asymmetry may not be faithfully representative of the remnant's true 3D structure or the kinematic properties of the original explosion. The majority of the optical ejecta are arranged in several well-defined and nearly circular ring-like structures with diameters between approximately 30'' (0.5 pc) and 2' (2 pc). These ejecta rings appear to be a common phenomenon of young core-collapse remnants and may be associated with post-explosion input of energy from plumes of radioactive {sup 56}Ni-rich ejecta that rise, expand, and compress non-radioactive material. Our optical survey encompasses Cas A's faint outlying ejecta knots and exceptionally high-velocity NE and SW streams of S-rich debris often referred to as ''jets''. These outer knots, which exhibit a chemical make-up suggestive of an origin deep within the progenitor star, appear to be arranged in opposing and wide-angle outflows with opening half-angles of Almost-Equal-To 40 Degree-Sign.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2018

Evaluation of Repurposing Archetypal Preventive Radiological/Nuclear Detectors to the Consequence Management Mission

In the first hours or days after an unplanned release of radioactive material to the environment, the radiation detection instruments most widely available to local first responders may be those currently fielded for interdiction missions. This study investigated how such preventative radiological / nuclear detection instruments could perform if repurposed to consequence management missions. A representative sample of three archetypes (body-worn, human carried, and other/large detection volume) encompassed six categories: personal radiation detector, extended range personal radiation detector, personal emergency radiation detector, radio-isotope identification device, human portable detector/backpack, and vehicle mounted large detection volume. Overall 19 models of equipment were included in the study. Laboratory evaluations were designed to assess the capabilities of the instruments in four consequence management missions: exposure rate, integrated exposure, radiation survey, and contamination screening. As applicable, the evaluations included measurement of exposure rate, integrated exposure, overrange response, and angular response. The results were compared to benchmarks from the American National Standards Institute ANSI N42.49A. The performance of the instruments for initial screening for contamination was assessed by an automated radioactive source moving past the detectors at various speeds and distances. The results demonstrate that if the equipment is used in accordance with the mission analysis and categories, and within the original equipment manufacturer specifications, it is possible to achieve sufficient accuracy to estimate and document doses to responders, plan entries into contaminated areas, detect contamination, and protect the public, until such time as outside resources arrive with sufficient numbers of standard health physics instruments and personnel dosimetry to replace the preventative radiological / nuclear detection instruments. Finally, this evaluation campaign was conducted to complement the National Council on Radiation Protection and Measurements Report-179 on Guidance for Emergency Response Dosimetry.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Concept Feasibility Report for Electroplating Zirconium onto Uranium Foil - Year 2

The Fuel Fabrication Capability within the U.S. High Performance Research Reactor Conversion Program is funded through the National Nuclear Security Administration (NNSA) NA-26 (Office of Material Management and Minimization). An investigation was commissioned to determine the feasibility of using electroplating techniques to apply a coating of zirconium onto depleted uranium/molybdenum alloy (U-10Mo). Electroplating would provide an alternative method to the existing process of hot roll-bonding zirconium foil onto the U-10Mo fuel foil during the fabrication of fuel elements for high-performance research reactors. The objective of this research was to develop a reproducible and scalable plating process that will produce a uniform, 25 μm thick zirconium metal coating on U-10Mo foil. In previous work, Pacific Northwest National Laboratory (PNNL) established a molten salt electroplating apparatus and protocol to plate zirconium metal onto molybdenum foil (Coffey 2015). During this second year of the research, PNNL furthered this work by moving to the U-10Mo alloy system (90 percent uranium:10 percent molybdenum). The original plating apparatus was disassembled and re-assembled in a laboratory capable of handling low-level radioactive materials. Initially, the work followed the previous year’s approach, and the salt bath composition was targeted at the eutectic composition (LiF:NaF:ZrF4 = 26:37:37 mol%). Early results indicated that the formation of uranium fluoride compounds would be problematic. Other salt bath compositions were investigated in order to eliminate the uranium fluoride production (LiF:NaF = 61:39 mol% and LiF:NaF:KF = 46.5:11.5:42 mol% ). Zirconium metal was used as the crucible for the molten salt. Three plating methods were used—isopotential, galvano static, and pulsed plating. The molten salt method for zirconium metal application provided high-quality plating on molybdenum in PNNL’s previous work. A key advantage of this approach is that plating can be performed under conditions that would greatly reduce the quantity of intermetallics that form at the interface between the zirconium and U-10Mo; unlike roll bonding, the molten salt plating approach would allow for complete coverage of the U-10Mo foil with zirconium. When utilizing the experimental parameters developed for zirconium plating onto molybdenum, a uranium fluoride reaction product was formed at the Zr/U-10Mo interface. By controlling the initial plating potential, the uranium fluoride could be prevented; however, the targeted zirconium thickness (25 ±12.5 μm) could not be achieved while maintaining 100% coverage.

36 MATERIALS SCIENCE↗