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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 1,477 records · Page 82
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Metal nanoparticles as lubricant additives

A lubricant composition comprising: (i) silver or gold nanoparticles, each of which is encapsulated by a layer of alkylthiol or alkylamine molecules; (ii) palladium or platinum nanoparticles, each of which is encapsulated by a layer of alkylthiol or alkylamine molecules; and (iii) a fluid in which components (i) and (ii) are present. Further described are methods for applying the lubricant composition onto a mechanical device for which lubrication is beneficial, with resulting improvement in friction and wear reduction and/or corrosion inhibition.

Ihala Gamaralalage, Chanaka Kapila Kumara↗
DOE OSTI2022

Evaluation of Bifacial Modules and PV Technologies with Combined-Accelerated Stress Testing

In view of the increasing interest and market share of bifacial cells and modules, suitable substrates such as glass and transparent backsheets along with ethylene vinyl acetate (EVA) and polyolefin elastomer (POE) encapsulants were examined in combined-accelerated stress testing (C-AST) to evaluate and compare degradation modes. Testing with both monofacial and bifacial cells, we found glass-glass modules with monofacial cells led to greater grid finger breakage than those with polymeric backsheets. This is attributed to previous x-ray topography work showing higher stress in cells in glass-glass modules than glass backsheet modules. Consistent with the objectives of C-AST, which stresses modules at levels corresponding to the limits seen in the natural environment, we observed the UV-fluorescence signatures of modules tested in C-AST (considering the degradation associated with developing chromophores, moisture penetration and photobleaching effects) to be like those in fielded modules, more so than other chamber stress testing implemented for comparison. Finally, among multiple cell degradation modes that we observed such as light-induced degradation (LID), we differentiated susceptibility to potential-induced degradation (PID) on the back of the bifacial Passivated Emitter Back Cells (PERC) in C-AST. Confirming with ex-situ tests, we found polarization-type PID most prevalent in glass/glass modules with EVA as would be anticipated considering the greater leakage current through such module encapsulation. Unlike PID tests performed in the dark, which can lead to false positive PID test results, field-representative illumination is experienced by the modules on the front and back side while -1200 V system voltage is applied in C-AST, supporting the conclusion that this module type would be susceptible to PID in the field.

bifacial modules↗
DOE OSTI2022

Carbon-based volatile and non-volatile memristors

An ultrathin, carbon-based memristor with a moiré superlattice potential shows prominent ferroelectric resistance switching. The memristor includes a bilayer material, such as Bernal-stacked bilayer graphene, encapsulated between two layers of a layered material, such as hexagonal boron nitride. At least one of the encapsulating layers is rotationally aligned with the bilayer to create the moiré superlattice potential. The memristor exhibits ultrafast and robust resistance switching between multiple resistance states at high temperatures. The memristor, which may be volatile or nonvolatile, may be suitable for neuromorphic computing.

Jarillo-Herrero, Pablo↗
DOE OSTI2022

Solar mirrors and methods of making solar mirrors having improved properties

An article for reflecting solar energy includes a coating stack having solar reflecting films and metal oxide films, the coating stack applied on a major surface of a glass substrate, and a protective overcoat comprising a first and a second surface, wherein the first surface of the protective overcoat is disposed toward the solar reflective films and metal oxide films; and a polymer encapsulant over outer wall surfaces of the coating stack, the second surface of the protective overcoat and over peripheral edges of the coated article, the encapsulant having a base layer, a top layer and metallic corrosion-inhibitive material in the base layer.

14 SOLAR ENERGY↗
DOE OSTI2022

Wavelength-selective specularly reflecting photovoltaic module and manufacture thereof

A photovoltaic module comprises a plurality of photovoltaic cells, and a polymeric film positioned on an incident light side of the plurality of photovoltaic cells, wherein the polymeric film transmits a range of wavelengths of the incident light spectrum and specularly reflects wavelengths outside of the range. An encapsulant layer is in contact with the polymeric film. The polymeric film may have a first surface area larger than a second surface area of the layer of photovoltaic cells. The polymeric film may have one or more through-holes so that encapsulant can penetrate through the through-holes at elevated temperature during lamination thereby bonding to a front glass of the photovoltaic module.

14 SOLAR ENERGY↗
DOE OSTI2023

Electronic and Optical Properties of NV Center in Diamonds for Sensing Applications: Experimental and First Principles Density Functional Theory Approach

The nitrogen-vacancy (NV) center in a nanodiamond (ND) crystal is one of the best candidate materials for quantum sensing and metrology applications at elevated environmental conditions. The NV center can be used to achieve an unprecedented level of sensitivity for sensing applications at high temperatures and pressures. The NV center is also proven to be useful for magnetic field sensing. Furthermore, it can be used for applications in the sensing of valuable minerals such as rare earth elements (REEs) from the crude and waste subsurface materials. This work presents computational and experimental studies on electronic and optical properties of bulk diamond with N and NV defects. ND in MOF/polymer materials will be experimentally encapsulated. The transverse spin relaxation time of the NV center will be measured in bare and encapsulated ND. Different doping elements will be introduced on the surfaces and present surface electronic properties to study the surfaces with shallow NV center for quantum sensing applications.

Paudel, Hari P.↗
DOE OSTI2023

Isolated data processing modules

This disclosure relates to computer systems that isolate data processing modules and methods of operating the same. In one embodiment of a method, a computer system receives a first data processing request for a first data file. The computer system may determine that a first data type of the first data file is supported by an encapsulated data processing module and determine a first communication protocol from a set of communication protocols that can be used to receive the first data file. The computer system can then receive the first data file from the first storage location in accordance with the first communication protocol and the encapsulated data processing module may be executed to convert the first data file in the first data format into second data in a second data format.

Penton, Joshua Michael↗
DOE OSTI2023

Crosslinked polymers with tunable coefficients of thermal expansion

Curatives and their resulting thermosets and other crosslinked polymers can reduce thermal expansion mismatch between an encapsulant and objects that are encapsulated. This can be accomplished by incorporating a negative CTE moiety into the thermoset resin or polymer backbone. The negative CTE moiety can be a thermal contractile unit that shrinks as a result of thermally induced conversion from a twist-boat to chair or cis/trans isomerization upon heating. Beyond CTE matching, other potential uses for these crosslinked polymers and thermosets include passive energy generation, energy absorption at high strain rates, mechanophores, actuators, and piezoelectric applications.

Redline, Erica Marie↗
DOE OSTI2013

Radiation Stability of Benzyl Tributyl Ammonium Chloride towards Technetium-99 Extraction - 13016

A closed nuclear fuel cycle combining new separation technologies along with generation III and generation IV reactors is a promising way to achieve a sustainable energy supply. But it is important to keep in mind that future recycling processes of used nuclear fuel (UNF) must minimize wastes, improve partitioning processes, and integrate waste considerations into processes. New separation processes are being developed worldwide to complement the actual industrialized PUREX process which selectively separates U(VI) and Pu(IV) from the raffinate. As an example, the UREX process has been developed in the United States to co-extract hexavalent uranium (U) and hepta-valent technetium (Tc) by tri-n-butyl phosphate (TBP). Tc-99 is recognized to be one of the most abundant, long-lived radio-toxic isotopes in UNF (half-life, t{sub 1/2} = 2.13 x 10{sup 5} years), and as such, is targeted in UNF separation strategies for isolation and encapsulation in solid waste-forms for final disposal in a nuclear waste repository. Immobilization of Tc-99 by a durable solid waste-form is a challenge, and its fate in new advanced technology processes is of importance. It is essential to be able to quantify and locate 1) its occurrence in any new developed flowsheets, 2) its chemical form in the individual phases of a process, 3) its potential quantitative transfer in any waste streams, and consequently, 4) its quantitative separation for either potential transmutation to Ru-100 or isolation and encapsulation in solid waste-forms for ultimate disposal. In addition, as a result of an U(VI)-Tc(VII) co-extraction in a UREX-based process, Tc(VII) could be found in low level waste (LLW) streams. There is a need for the development of new extraction systems that would selectively extract Tc-99 from LLW streams and concentrate it for feed into high level waste (HLW) for either Tc-99 immobilization in metallic waste-forms (Tc-Zr alloys), and/or borosilicate-based waste glass. Studies have been launched to investigate the suitability of new macro-compounds such as crown-ethers, aza-crown ethers, quaternary ammonium salts, and resorcin-arenes for the selective extraction of Tc-99 from nitric acid solutions. The selectivity of the ligand is important in evaluating potential separation processes and also the radiation stability of the molecule is essential for minimization of waste and radiolysis products. In this paper, we are reporting the extraction of TcO{sub 4}{sup -} by benzyl tributyl ammonium chloride (BTBA). Experimental efforts were focused on determining the best extraction conditions by varying the ligand's matrix conditions and concentration, as well as varying the organic phase composition (i.e. diluent variation). Furthermore, the ligand has been investigated for radiation stability. The ?-irradiation was performed on the neat organic phases containing the ligand at different absorbed doses to a maximum of 200 kGy using an external Co-60 source. Post-irradiation solvent extraction measurements will be discussed. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Disposition of Radioisotope Thermoelectric Generators Currently Located at the Oak Ridge National Laboratory - 12232

Under the American Recovery and Reinvestment Act (ARRA), the U.S. Department of Energy (DOE) awarded SEC Federal Services Corporation (SEC) a 34-building demolition and disposal (D and D) project at the Oak Ridge National Laboratory (ORNL) that included the disposition of six Strontium (Sr-90) powered Radioisotope Thermoelectric Generators (RTGs) stored outside of ORNL Building 3517. Disposition of the RTGs is very complex both in terms of complying with disposal facility waste acceptance criteria (WAC) and U.S. Department of Transportation (DOT) requirements for packaging and transportation in commerce. Two of the RTGs contain elemental mercury which requires them to be Land Disposal Restrictions (LDR) compliant prior to disposal. In addition, all of the RTGs exceed the Class C waste concentration limits under Nuclear Regulatory Commission (NRC) Waste Classification Guidelines. In order to meet the LDR requirements and Nevada National Security Site (NNSS) WAC, a site specific treatability variance for mercury was submitted to the U.S. Environmental Protection Agency (EPA) to allow macro-encapsulation to be an acceptable treatment standard for elemental mercury. By identifying and confirming the design configuration of the mercury containing RTGs, the SEC team proved that the current configuration met the macro-encapsulation standard of 40 Code of Federal Regulations (CFR) 268.45. The SEC Team also worked with NNSS to demonstrate that all radioisotope considerations are compliant with the NNSS low-level waste (LLW) disposal facility performance assessment and WAC. Lastly, the SEC team determined that the GE2000 Type B cask met the necessary size, weight, and thermal loading requirements for five of the six RTGs. The sixth RTG (BUP-500) required a one-time DOT shipment exemption request due to the RTG's large size. The DOT exemption justification for the BUP-500 relies on the inherent robust construction and material make-up of the BUP- 500 RTG. DOE-ORO, SEC, and the entire SEC RTG team are nearing the conclusion of the Sr-90 RTG disposition challenge - a legacy now 50 years in the making. Over 600,000 Ci of Sr-90 waste await disposal and its removal from ORNL will mark an historical moment in the clean-up of the cold-war legacy in the ORNL central industrial area. Elimination (i.e., removal) of the RTGs will reduce security risks at ORNL and disposal will permanently eliminate security risks. The RTGs will eventually decay to benign levels within a reasonable timeframe relative to radiological risks posed by long-lived isotopes. The safety authorization basis at ORNL Building 3517 will be reduced enabling greater operational flexibility in future clean-out and D and D campaigns. Upon disposition the Department of Energy will realize reduced direct and indirect surveillance and maintenance costs that can be reapplied to accelerated and enhanced clean-up of the Oak Ridge Reservation. At present, waste profiles for the RTGs are developed and under review by NNSS RWAP staff and approval authorities. Disposition schedule is driven by the availability of compliant shipping casks necessary to safely transport the RTGs from ORNL to NNSS. The first disposal of the RCA RTG is expected in April 2012 and the remaining RTGs disposed in 2012 and 2013. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

GRAFEC: A New Spanish Program to Investigate Waste Management Options for Radioactive Graphite - 12399

Spain has to manage about 3700 tons of irradiated graphite from the reactor Vandellos I as radioactive waste. 2700 tons are the stack of the reactor and are still in the reactor core waiting for retrieval. The rest of the quantities, 1000 tons, are the graphite sleeves which have been already retrieved from the reactor. During operation the graphite sleeves were stored in a silo and during the dismantling stage a retrieval process was carried out separating the wires from the graphite, which were crushed and introduced into 220 cubic containers of 6 m{sup 3} each and placed in interim storage. The graphite is an intermediate level radioactive waste but it contains long lived radionuclides like {sup 14}C which disqualifies disposal at the low level waste repository of El Cabril. Therefore, a new project has been started in order to investigate two new options for the management of this waste type. The first one is based on a selective decontamination of {sup 14}C by thermal methods. This method is based on results obtained at the Research Centre Juelich (FZJ) in the Frame of the EC programs 'Raphael' and 'Carbowaste'. The process developed at FZJ is based on a preferential oxidation of {sup 14}C in comparison to the bulk {sup 12}C. Explanations for this effect are the inhomogeneous distribution and a weaker bounding of {sup 14}C which is not incorporated in the graphite lattice. However these investigations have only been performed with graphite from the high temperature reactor Arbeitsgemeinschaft Versuchsreaktor Juelich AVR which has been operated in a non-oxidising condition or research reactor graphite operated at room temperature. The reactor Vandellos I has been operated with CO{sub 2} as coolant and significant amounts of graphite have been already oxidised. The aim of the project is to validate whether a {sup 14}C decontamination can also been achieved with graphite from Vandellos I. A second possibility under investigation is the encapsulation of the graphite in a long term stable glass matrix. The principal applicability has been already proved by FNAG. Crushed graphite mixed with a suitable glass powder has been pressed at elevated temperature under vacuum. The vacuum is required to avoid gas enclosures in the obtained product. The obtained products, named IGM for 'Impermeable Graphite Matrix', have densities above 99% of theoretical density. The amount of glass has been chosen with respect to the pore volume of the former graphite parts. The method allows the production of encapsulated graphite without increasing the disposal volume. This paper will give a short overview of characterisation results of different irradiated graphite materials obtained at CIEMAT and in the Carbowaste project as well as the proposed methods and the actual status of the program including first results about leaching of non-radioactive IGM samples and hopefully first tendencies concerning the C-14 separation from graphite of Vandellos I by thermal treatment. Both processes, the thermal treatment as well as the IGM, have the potential to solve problems related to the management of irradiated graphite in Spain. However the methods have only been tested with different types of i-graphite and virgin graphite, respectively. Only investigations with real i-graphite from Spain will reveal whether the described methods are applicable to graphite from Vandellos I. However all partners are convinced that one of these new methods or a combination of them will lead to a feasible option to manage i-graphite in Spain on an industrial scale. (authors)

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

Developing High-Energy Laser-Driven Bremsstrahlung Radiation Sources for Imaging and Analysis of Nuclear Waste Packages - 17300

Radiographic imaging of small scale simulated nuclear waste packages (a 28 mm uranium penny encapsulated in grout) was achieved using a single pulse exposure from an x-ray source driven by a high-power laser. The high-energy laser pulse was generated at the Vulcan laser facility, which focused the photon burst onto a tantalum target foil in order to produce a bright burst of highly penetrating x-rays (with energy >500 keV) from a source size less than 0.5 mm. The sample was imaged using BAS-TR and BAS-SR image plates alongside a Thallium doped Caesium Iodide scintillator-based detector coupled to CCD chips for rapid image acquisition. The penny was resolved to sub-mm accuracy over a 30 cm{sup 2} scan area from a single shot. In addition, on a parallel beamline, neutron generation (of 10{sup 7}-10{sup 9} neutrons per steradian per pulse) was achieved, thereby demonstrating the potential application for criticality testing of waste packages. The principle concern regarding intermediate level waste (ILW) containers is the internal volumetric expansion due to the formation of corrosion products. Therefore, simulated waste samples with hydride corrosion have been formed and characterised in order to emulate this phenomenon. In addition, samples containing controlled isotopic ratios have been formed. This next generation of simulated waste samples have been produced in preparation for a second phase of analysis at the Vulcan laser facility to quantify the identification ability of radiographic imaging and characterisation accuracy of neutronics measurements. This feasibility study successfully demonstrated non-destructive radiography of encapsulated, high density, nuclear material and this paper details the next phase of technological development. (authors)

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

Commissioning of a GeoMelt Vitrification Plant at the Central Laboratory on the Sellafield Site - 17561

The National Nuclear Laboratory (NNL) and Kurion Veolia (Kurion) have installed a GeoMelt{sup R} In-Container Vitrification (ICV){sup R} radioactive waste treatment system in the Central Laboratory Active Rig Hall at Sellafield. This installation was completed in 2016. GeoMelt incorporates a range of patented and proprietary vitrification processing technologies that are configured in a variety of ways depending on the project and the waste. GeoMelt transforms hazardous chemical and radioactive wastes into an inert vitreous monolith. Hazardous organics are destroyed by pyrolysis and heavy metals and radionuclides are immobilized within the glass matrix where they are not preferentially leached into the environment. GeoMelt IVC produces a durable wasteform for a large variety of input wastes. The batch nature of the process which involves single-use refractory offers a wide range of process benefits over traditional vitrification technologies including: - a wide range of waste chemistries allowing high waste loadings; - compatibility with heterogeneous wastes including wastes with high metal content; - a technological simple solution without stirrers, bubblers, etc.; - No refractory corrosion concerns or melter drain failure scenarios; - Flexibility in glass chemistry and melter/disposal container configuration allows the wasteform to be specifically tailored to the storage or disposal site requirements. As part of the 2016 system commissioning at NNL's Central Laboratory, three demonstration melts were performed, one funded by Sellafield Ltd. and two by the Nuclear Decommissioning Authority (NDA). The first melt processed inactive soil with an inactive caesium tracer. The second melt processed inactive soil with a Cs-137 tracer. The third melt processed active waste soil from a variety of NDA nuclear licensed sites. The work conducted during commissioning was multi-purpose; it successfully demonstrated the benefits that the GeoMelt technology could bring to the UK NDA estate while performing equipment testing/troubleshooting and operator training. A significant benefit of the GeoMelt ICV technology is volume reduction for wastes which would otherwise experience a volume increase if encapsulated. Another benefit is the ability to treat difficult radioactive waste streams not amenable to encapsulation (e.g., reactive and/or organic wastes) with a very low energy consumption. The purpose of the first commissioning melt, Plant Performance Demonstration (PPD)-1, was to operate the GeoMelt system under typical conditions and to perform a post-melt mass balance of the caesium as a surrogate for Cs-137 in order to access retention of caesium in the final glass product. The melt processed a mixture of inactive soil obtained near the Sellafield site (as analogue for contaminated Sellafield soil), additives, and glass frit. An inactive caesium tracer was added to this melt as Cs{sub 2}CO{sub 3}. The purpose of the second commissioning melt, PPD-2, was to operate the GeoMelt system under typical conditions and to perform a post-melt activity balance using a Cs-137 tracer in order to assess retention of Cs-137 in the final glass product. Materials processed were inactive soil, glass frit, additives, and a radioactive caesium tracer (CsCl). The purpose of the third melt, Experiment-1 (EX-1) was to treat active waste and inactive waste simulant, as a demonstration of simultaneous processing of multiple waste streams. EX-1 consisted of processing active soil, Magnox sludge simulant, inactive soil, additives, and frit. The active soil used for this melt was obtained from the Hunterston A, Bradwell, Capenhurst, and Sellafield nuclear licensed sites. Since most soils are high in silica, the concept for this experiment was to use the active soils to provide glass formers while co-processing Magnox sludge simulant which does not contain glass formers. Beyond demonstration of capability areas of focus of this series of demonstrations included processing efficiency (kWh/kg treated) and caesium retention (%). This paper summarizes operations and results of the three melts. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Design and Construction of the Paddys Run Streambank Stabilization Project at the Fernald Preserve, Harrison, Ohio - 16165

The Fernald Preserve is a former uranium-processing plant that underwent extensive remediation pursuant to CERCLA and is now managed by the US DOE Office of Legacy Management. While remediation of buildings and soil contamination was completed in 2006, aquifer remediation is ongoing. Paddys Run is a second-order stream that runs to the south along the western side of the Fernald Preserve. The Paddys Run watershed encompasses nearly 41 km{sup 2} (16 mi{sup 2}), including most of the Fernald site. Field personnel conducting routine site inspections in March 2014 observed that Paddys Run was migrating east via bank erosion into the 'Pit 3 Swale', an area of known surface-water contamination. The soil there was certified pursuant to site regulatory agreements and meets all final remediation levels. However, weekly surface-water monitoring is conducted from two puddles within the swale area, when water that exceeds the final remediation levels is present. Paddys Run had migrated east approximately 4 m (13 ft) in 2 years and was approximately 29 m (95 ft) from the sample location. This rapid migration threatened existing conditions that allowed for continued monitoring of the swale area and also threatened Paddys Run water quality. Therefore, DOE and regulators determined that the east bank of Paddys Run required stabilization. This was accomplished with a design that included the following components: relocation of approximately 145 m (475 ft) of streambed 9 m (30 ft) west, installation of a rock toe along the east bank, installation of two cross-vane in-stream grade-control structures, stabilization of a portion of the east bank using soil encapsulated lifts, and regrading, seeding, and planting within remaining disturbed areas. In an effort to take advantage of low-flow conditions in Paddys Run, construction was initiated in September 2014. Weather delays and subsurface flow within the Paddys Run streambed resulted in an interim shutdown of the project area in December 2014. Construction activities resumed in April 2015, with completion in November 2015. To date, this stabilization project has been successful. The regraded bank and streambed have remained stable, and no compromise to installed cross-vanes, the rock toe, or the soil encapsulated lifts has been observed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

In-Situ Gamma-Ray Assay of the 235-F Plutonium Fuel Form Facility at the Savannah River Site - 16293

The Plutonium Fuel Form (PuFF) Facility is located in Building 235-F near the geographic center of the Savannah River Site. Between 1978 and 1984, the facility was used to produce approximately 165 kilograms of iridium-encapsulated Pu-238 spheres and pellets for use as radioisotope thermal generators, primarily for the space program. The facility is divided between two cell lines, the east cell line used to process powdered Pu-238 oxide raw material into fuel forms and the west cell line used to encapsulate the fuel forms in iridium. Scoping in-situ gamma-ray assays were performed in the PuFF Facility in 2006 and estimated that hundreds of grams of Pu-238 holdup were within the facility. Using this holdup estimate as a source term, SRS has performed a risk analysis that indicated a seismic event that induces a full-facility fire in 235-F could lead to a 28,800 rem dose to a co-located worker. Based on this risk assessment, SRS is taking steps to decontaminate the facility. One of the first steps taken has been to improve upon the quality of the in-situ gamma-ray assay data. Carts and collimators were specially designed to survey the equipment in the PuFF Facility. The cell interiors were assayed along with the furnaces and storage coolers that protrude beneath the cells. While the previous scoping work consisted of 32 measurements, the current series of measurements included nearly 400 measurements using both high-purity germanium and lanthanum bromide detectors. Data analysis for the current set of measurements was conducted with greater rigor as well. MCNP5 was used to evaluate a variety of possible physical distributions for the Pu-238 source term and estimate cross-talk between neighboring cells. Data analysis was performed using three gamma-rays emitted by Pu-238 (99.85 keV, 152.7 keV, and 766.4 keV) providing three independent estimates of the mass of Pu-238 holdup in each of the cells. This also allowed the application of matrix correction factors to adjust gamma-ray attenuation. The weighted mean of these three results was used as the best estimate of Pu-238 holdup in the PuFF facility. This work lowered the estimate of holdup in the PuFF facility by approximately 100 grams compared to the previous estimate, a 0.55's difference. The uncertainty in the Pu-238 holdup was also reduced substantially relative to the 2006 scoping assay. Further radiological measurements are planned for the near future including in-situ gamma-ray assays of wing cabinets in the east cell line maintenance area, the transfer tube between the two cell lines, and some of the furnaces and storage coolers. A germanium gamma imager, the GeGI, will also be used to map out the distribution of the Pu-238 source term in the PuFF Facility. (authors)

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

Radionuclide Behaviour During Thermal Treatment of LLW and ILW Using In-Container Vitrification - 19239

Thermal treatment is being considered in the UK for the treatment of a wide range of low and intermediate level wastes (LLW and ILW). To support this objective, a range of in-container vitrification trials were conducted to test a wide range of simulated waste streams. The vitrified products have been visually inspected and samples removed for chemical analysis. This paper presents preliminary results of this analysis with some comment on the spatial and chemical heterogeneity of the products. Treatment of a sludge simulant in a steel skip failed to fully melt the metal and produced a glass encapsulated waste-form with clearly visible porosity. Clinoptilolite produced a dense uniform glass but with some chemical heterogeneity. Some operational issues were encountered when treating water-rich sludge, but these were successfully resolved and resulted in a broadly homogenous vitrified product. Melting simulated mixed decommissioning waste produced a chemically homogenous product, despite the heterogenous nature of the waste stream. Uranium and concrete contained in a steel 'top hat' presented significant treatment issues. The top hat failed to fully melt, and this appears to have prevented the waste contained within the top hat from mixing with the glass external to the top hat. As with the steel skip containing sludge, this again left a glass encapsulated, rather than fully vitrified, product. Attempts to melt bulk PVC (an analogue of soft PCM) initially led to significant operational issues with the melt being terminated early and the waste apparently combusting rather than melting. This was resolved by pre-treating the PVC by shredding it and mixing this with the frit to give a larger contact area for melting. This approach produced a homogenous well mixed glass. Overall it appears the properties of the melt product are highly dependent on the physical and chemical properties of the waste stream being treated. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Multi-Physics Coupling to Model a RbCl-RbCl-Ga Target Stack

A computational model has been developed to simulate the performance of the Los Alamos Isotope Production Facility's (IPF) routine production target configuration, which consists of two Inconel-encapsulated RbCl salt targets and one niobium-encapsulated gallium target separated by water cooling channels. The methodology involves iteratively coupling particle transport and thermal hydraulic codes to determine the time dependent behavior of the target. Phase and density changes seen in targets are known to impact beam penetration and distribution. In addition, evidence of energy distributions seen in experimental radionuclide yield measurements at IPF do not match those seen in previous MCNP predictions. Density reduction as a result of heat deposition affects proton energy loss along the particle's track, impacting both radionuclide yields and target design. Accurate modeling of the thermal behavior of these multi-phase targets and the corresponding impact on beam behavior will significantly improve our understanding of incident and exit proton energy distributions and resulting radionuclide yields in downstream targets. Thermal hydraulic modeling, including computational fluid dynamics (CFD) and heat transfer, was performed with ANSYS CFX while particle transport was performed with the Monte Carlo N-Particle code (MCNP). (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

The criticality safety program of the Swedish nuclear fuel and waste management Co (SKB)

SKB is co-owned by the owners of the Swedish nuclear power plants (NPP) and is responsible for the back-end of all Swedish spent nuclear fuel and radioactive waste from the Swedish nuclear program. The responsibility starts from when the fuel leaves the NPP and ends at the planned final repository. That means that SKB has the responsible for the safety at the intermediate pool storage (Clab), the encapsulation plant (Clink), the Spent Fuel Repository and the transportation between these sites. In March 2011 SKB applied to the Swedish Radiation Safety Authority (SSM) for a permit to build a final repository for spent fuel and an encapsulation plant (Clink). The application contained a criticality safety analysis for the canisters to be used in Clink and the copper canister to be used in the final repository. In 2012 SSM requested several complements to the criticality safety analysis in the application and about the same time they also required Clab to modernise the Safety Analysis Report. Our owners (the NPPs) had also requested SKB to take care of higher enrichments and higher burnup fuel. A prerequisite to meet this was extended Burnable Absorber credit (peak reactivity) and Burn-up credit. Up to this stage SKB had relied heavily on consultants to do criticality safety analysis and had no specific criticality safety program or methodology. However it was now evident to SKB that we needed to take an overall view on the criticality safety area and not only lift the quality of the actual criticality analysis but also to handle organisational and competence issues. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗