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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 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Elemental source attribution signatures for calcium ammonium nitrate (CAN) fertilizers used in homemade explosives

Calcium ammonium nitrate (CAN) is a widely available fertilizer composed of ammonium nitrate mixed with some form of calcium carbonate such as limestone or dolomite. CAN is also frequently used to make homemade explosives. The potential of using elemental profiling and chemometrics to match both pristine and reprocessed CAN fertilizers to their factories for use in future forensic investigations was examined. Inductively coupled plasma-mass spectrometry (ICP-MS) analysis was performed on 64 elements in 125 samples from 11 CAN stocks from 6 different CAN factories. Fisher ratio, degree-of-class-separation, and partial least squares discriminant analysis (PLSDA) were used to develop a model using the concentrations of Na, V, Mn, Cu, Ga, Sr, Ba and U to classify a validation set of CAN samples into 5 factory groups; one group was two factories from the same fertilizer company. In terms of the pristine CAN samples, i.e., unadulterated prills, 64% of the test samples were matched to their correct factory group with zero false positives. The same PLSDA model was used to correctly match 100% of the CAN samples that were reprocessed by crushing and mixing the CAN with powdered sugar. In the case of crushed CAN samples mixed with aluminum powder, correct matches were made for zero to 100% of the samples depending on the factory the CAN originated. Remarkably, for one factory, 100% of the ammonium nitrate samples that were extracted from CAN using tap or bottled water were matched to the correct CAN factory group. Lastly, the water-insoluble (calcium carbonate) portions of CAN provided a greater degree of discrimination between factories than the water-soluble portions of CAN. In summary, this work illustrates that sourcing unadulterated CAN fertilizer can potentially be done with high frequency and high confidence using elemental profiling and chemometrics while the sourcing of reprocessed CAN is dependent on how much an adulterant alters the recovered elemental profile of CAN.

chemical forensics↗
DOE OSTI2016

Metal–organic framework with optimally selective xenon adsorption and separation

Nuclear energy is considered among the most viable alternatives to our current fossil fuel based energy economy.1 The mass-deployment of nuclear energy as an emissions-free source requires the reprocessing of used nuclear fuel to mitigate the waste.2 One of the major concerns with reprocessing used nuclear fuel is the release of volatile radionuclides such as Xe and Kr. The most mature process for removing these radionuclides is energy- and capital-intensive cryogenic distillation. Alternatively, porous materials such as metal-organic frameworks (MOFs) have demonstrated the ability to selectively adsorb Xe and Kr at ambient conditions.3-8 High-throughput computational screening of large databases of porous materials has identified a calcium-based nanoporous MOF, SBMOF-1, as the most selective for Xe over Kr.9,10 Here, we affirm this prediction and report that SBMOF-1 exhibits by far the highest Xe adsorption capacity and a remarkable Xe/Kr selectivity under relevant nuclear reprocessing conditions. The exceptional selectivity of SBMOF-1 is attributed to its pore size tailored to Xe and its dense wall of atoms that constructs a binding site with a high affinity for Xe, as evident by single crystal X-ray diffraction and molecular simulation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Application of sensitivity analysis in DYMOND/Dakota to fuel cycle transition scenarios

The ability to perform sensitivity analysis has been enabled for the nuclear fuel cycle simulator DYMOND through its coupling with the design and analysis toolkit Dakota. To test and demonstrate these new capabilities, a transition scenario and multi-parameter study were devised. The transition scenario represents a partial transition from the US nuclear fleet to a closed fuel cycle with small modular LWRs and fast reactors fueled by reprocessed used nuclear fuel. Four uncertain parameters in this transition were studied – start date of reprocessing, total reprocessing capacity, the nuclear energy demand growth, and the rate at which the fast reactors are deployed – with respect to their impact on four response metrics. The responses – total natural uranium consumed, maximum annual enrichment capacity required, total disposed mass, and total cost of the nuclear fuel cycle – were chosen based on measures known to be of interest in transition scenarios and to be significantly impacted by the varying parameters. Furthermore, analysis of this study was performed both from the direct sampling and through surrogate models developed in Dakota to calculate the global sensitivity measures Sobol’ indices. This example application of this new capability showed that the most consequential parameter to most metrics was the share of new build capacity that is fast reactors. However, for the cost metric, the scaling factor of the energy demand growth was significant and had synergistic behavior with the fast reactor new build share.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Extraction of Neptunium, Plutonium, Americium, Zirconium, and Technetium by Di-(2-Ethylhexyl)- Iso -Butyramide (DEH i BA) at High Metal Loadings

Increased focus on carbon neutral energy has generated a resurgence of interest in nuclear power, and in particular advanced reactors which are likely to utilize high assay low enriched uranium (HALEU). This in turn could increase the economic attractiveness of recovering still partially enriched uranium from used nuclear fuel. Concomitant to development of advanced reactors, advanced reprocessing schemes should be developed which address the disadvantages to well established reprocessing schemes. The present study focuses on using di-(2-ethylhexyl)-iso-butyramide (DEHiBA) under high metal loading conditions for the reprocessing of used nuclear fuel. The elements examined in the study include the dominant transuranic actinides (Np, Pu, Am) as well as the often-problematic Tc and Zr. Further, by increasing the concentration of the extractant from the more commonly reported 1.0 M – 1.5 M, the extraction of hexavalent actinides is substantially increased, while maintaining effective rejection of tri, tetra, and pentavalent actinides, particularly in the presence of high loadings of uranium. In conclusion, the extraction of Zr by 1.5 M DEHiBA is noted to be negligible by comparison to tributyl phosphate (TBP), however the coextraction of Tc with U is observed to be nominally twice the quantity that is extracted by TBP indicating a need for effective Tc management.

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

Isotope anomalies in the FE-group elements in meteorites and connections to nucleosynthesis in AGB stars

We study the effects of neutron captures in AGB stars on “Fe-group” elements, with an emphasis on Cr, Fe, and Ni. These elements show anomalies in {sup 54}Cr, {sup 58}Fe, and {sup 64}Ni in solar system materials, which are commonly attributed to supernovae (SNe). However, as large fractions of the interstellar medium (ISM) were reprocessed in AGB stars, these elements were reprocessed, too. We calculate the effects of such reprocessing on Cr, Fe, and Ni through 1.5 M{sub ⊙} and 3 M{sub ⊙} AGB models, adopting solar and 1/3 solar metallicities. All cases produce excesses of {sup 54}Cr, {sup 58}Fe, and {sup 64}Ni, while the other isotopes are little altered; hence, the observations may be explained by AGB processing. The results are robust and not dependent on the detailed initial isotopic composition. Consequences for other “Fe group” elements are then explored. They include {sup 50}Ti excesses and some production of {sup 46,47,49}Ti. In many circumstellar condensates, Ti quantitatively reflects these effects of AGB neutron captures. Scatter in the data results from small variations (granularity) in the isotopic composition of the local ISM. For Si, the main effects are instead due to variations in the local ISM from different SN sources. The problem of Ca is discussed, particularly with regard to {sup 48}Ca. The measured data are usually represented assuming terrestrial values for {sup 42}Ca/{sup 44}Ca. Materials processed in AGB stars or sources with variable initial {sup 42}Ca/{sup 44}Ca ratios can give apparent {sup 48}Ca excesses/deficiencies, attributed to SNe. The broader issue of galactic chemical evolution is also discussed in view of the isotopic granularity in the ISM.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2024

A recyclable self-healing composite with advanced sensing property

Polymer-based composites frequently encounter damage, often lurking beneath the surface and proving challenges to their early detection and repair. While material-based sensors show promise for encoding self-sensing properties within these composites, their in situ healing and reprocessability remain significant challenges. Therefore, the overarching goal of this study is the creation of a reprocessable polymeric composite encoded with self-healing attributes and the ability to autonomously sense damage. At the core of this innovation are vitrimers, a polymeric material characterized by a covalently adaptive dynamic network responsive to external factors such as heat. They combine thermoset-like resilience with thermoplastic-like flowability on demand under external stimuli. We nanoengineer a polyester-based vitrimeric polymer by incorporating piezoresistive carbon nanotubes (CNTs) as reinforcing elements that not only enhance its mechanical strength but also create a percolation network within the composite, thereby enabling piezoresistive self-sensing properties, all the while preserving the intrinsic self-healing capabilities offered by the vitrimeric matrix. The fabrication process of the composite involves a solvent-free in situ polymerization method that combines epoxy and anhydride-containing monomers with ~ 0.1 wt.% of CNTs. Once it was established that the introduction of CNTs into the polymeric matrix did not compromise the mechanical properties of the composite, their strain-sensing properties were characterized by applying cyclic loading while measuring their electrical resistance. Strikingly, CNT-enhanced vitrimer composite consistently retains its mechanical and sensing properties through repeated cycles of reshaping and reprocessing, underscoring its potential as a robust distributed strain sensor. This polyester-based vitrimeric composite is also easily recyclable without harsh chemical treatments. Preliminary findings from this study conclusively demonstrate that the bulk composite boasts both self-sensing capabilities and in situ detect healing properties, charting a promising course towards the development of a mechanically resilient multifunctional composite that seamlessly integrates selfsensing and healing capabilities.

Rohewal, Sargun Singh↗
DOE OSTI2019

Non-Aqueous Electrochemical Fluorination of Used Nuclear Fuel as an Advanced Separation Process

Development of efficient and environmentally benign methods to reprocess used nuclear fuel (UNF) will enable technologies for a nuclear renaissance. Electrochemical and fluorination methods for reprocessing UNF have been proposed, but combinations of electrochemical and fluorination methods have not been investigated. Electrochemical fluorination may reduce waste volumes compared to the main large-scale aqueous methods for processing used nuclear fuel. A non-aqueous electrochemical fluorination reprocessing method has been developed and demonstrated that enables gas phase uranium recovery while allowing for potential control of the reaction using a single process. Thermodynamic modeling showed a galvanic reaction between UNF and a fluorinating agent, such as NF 3 , in a molten fluoride electrolyte was possible and could selectively fluorinate U to UF 6 . To verify the model results, a reactor system for electrochemical fluorination was constructed along with vessels for product collection and analysis. Initial trials of the electrochemical reaction were performed and characterized by several electrochemical methods. The electrochemical oxidation of U using NF 3 as an oxidizing agent was demonstrated and UF 6 as a reaction product was detected.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Utah FORGE: Seismic Reflection Data

This is 2D and 3D seismic reflection data from Utah FORGE reprocessed during Phase 2c. The readme file containing an explanation of the data including data formats, software that can be used, processing, and projection and datum used. The Reprocessing document gives the rationale for reprocessing and shows examples of the improvements that were obtained. For all 3D and 2D data the following datasets were created and output in SEG-Y format: - Unmigrated Time - Prestack Time Migration (PSTM), Unenhanced (UnEnh) and Enhanced (Enh) - Prestack Depth Migration (PSDM), Unenhanced (UnEnh) and Enhanced (Enh) - Velocity Model used for Migration

15 GEOTHERMAL ENERGY↗
DOE OSTI2024

Hestia SW-IFL Onroad Fossil Fuel Carbon Dioxide (FFCO2) product: Road segment-level annual FFCO2 emissions across Arizona (2017-2022), version 1.1

The SW-IFL onroad fossil fuel carbon dioxide (FFCO2) emissions data product represents CO2 emissions from the combustion of fossil fuels by motor vehicles (e.g., passenger cars, trucks, buses, motorcycles) traveling on designated roadways. The emissions are represented geographically on each road segment within the state of Arizona spanning the 2017 to 2022 time period. This data product was developed as part of the Southwest Urban Corridor Integrated Field Laboratory (SW-IFL) project, which aims to provide new knowledge and tools that address urban environmental issues by integrating high-resolution observations, modeling, and civic engagement. The emissions data are provided in CSV (input data, ONR_FFCO2_AZ_county.csv) and GeoPackage form (output polyline objects - about 786,000 road segments, XXXX_AZ_v1.1.gpkg) designated by road class (interstates, arterials, collectors, local). The metadata file (Metadata_SW-IFL_Onroad_annualFFCO2_v1.1.docx) provides details about attributes and data formats. The GeoPackage emissions data are provided separately for local roads and nonlocal roads (interstates, arterials, collectors). The method file (Methods_SW-IFL_Onroad_annualFFCO2_v1.1.docx) describes the data processing flow and data sources. Update on 2024-04-17: Updates were made to both the input emission data file (.csv) and output segment-level emission file (.gpkg). There was an update in county-level emission input data (ONR_FFCO2_AZ_county.csv) and the entire road segments were reprocessed to reflect this update.Update on 2024-04-29: Update was made to one output segment-level emission file (Nonlocal_AZ_v1.0.gpkg). There was an error in the AADT values and the data were reprocessed to reflect this update.Update on 2024-10-22: Temporal coverage was extended to include 2022. VMT values were recalculated using new AADT data and the entire road segments were reprocessed to reflect these updates.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2017

Synthesis and Characterization of 5- and 6- Coordinated Alkali Pertechnetates

The local chemistry of technetium-99 ( 99 Tc) in oxide glasses is important for understanding the incorporation and long-term release of Tc from nuclear waste glasses, both those for legacy defense wastes and fuel reprocessing wastes. Tc preferably forms Tc(VII), Tc(IV), or Tc(0) in glass, depending on the level of reduction of the melt. Tc(VII) in oxide glasses is normally assumed to be isolated pertechnetate TcO 4 - anions surrounded by alkali, but can occasionally precipitate as alkali pertechnetate salts such as KTcO 4 and NaTcO 4 when Tc concentration is high. In these cases, Tc(VII) is 4-coordinated by oxygen. A reinvestigation of the chemistry of alkali-technetium-oxides formed under oxidizing conditions and at temperatures used to prepare nuclear waste glasses showed that higher coordinated alkali Tc(VII) oxide species had been reported, including those with the TcO 5 - and TcO 6 - anions. The chemistry of alkali Tc(VII) and other alkali-Tc-oxides is reviewed, along with relevant synthesis conditions. Additionally, we report attempts to make 5- and 6-coordinate pertechnetate compounds of K, Na, and Li, i.e. TcO 5 - and TcO 6 -. It was found that higher coordinated species are very sensitive to water, and easily decompose into their respective pertechnetates. It was difficult to obtain pure compounds, but mixtures of the pertechnetate and other phase(s) were frequently found, as evidenced by x-ray absorption spectroscopy (XAS), neutron diffraction (ND), and Raman spectroscopy. Low temperature electron paramagnetic resonance (EPR) measurements showed the possibility of Tc(IV) and Tc(VI) in Na 3 TcO 5 and Na 5 TcO 6 compounds. Here, it was hypothesized that the smaller counter cation would result in more stable pertechnetates. To confirm the synthesis method, LiReO 4 and Li 5 ReO 6 were prepared, and their Raman spectra match those in the literature. Subsequently, the Tc versions LiTcO 4 and Li 5 TcO 6 were synthesized and characterized by ND, Raman spectroscopy, XANES, and EXAFS. The Li 5 TcO 6 was a marginally stable compound that appears to have the same structure as that known for Li 5 ReO 6 . Implications of the experimental work on stability of alkali technetate compounds and possible role in the volatilization of Tc are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2012

The Multi-Isotope Process (MIP) Monitor Project: FY12 Progress and Accomplishments

The Multi-Isotope Process (MIP) Monitor, being developed at Pacific Northwest National Laboratory (PNNL), provides an efficient approach to monitoring the process conditions in reprocessing facilities in support of the goal of "...(minimization of) the risks of nuclear proliferation and terrorism." The MIP Monitor measures distributions of a suite of indicator (radioactive) isotopes present within product and waste streams of a nuclear reprocessing facility. These indicator isotopes are monitored on-line by gamma spectrometry and compared, in near-real-time, to spectral patterns representing "normal" process conditions using multivariate pattern recognition software. The monitor utilizes this multivariate analysis and gamma spectroscopy of reprocessing streams to detect small changes in the gamma spectrum, which may indicate changes in process conditions. Multivariate analysis methods common in chemometrics, such as principal component analysis (PCA) and partial least squares regression (PLS), act as pattern recognition techniques, which can detect small deviations from the expected, nominal condition. By targeting multiple gamma-emitting indicator isotopes, the MIP Monitor approach is compatible with the use of small, portable, relatively high-resolution gamma detectors that may be easily deployed throughout an existing facility. The automated multivariate analysis can provide a level of data obscurity, giving a built-in information barrier to protect sensitive or proprietary operational data. Proof-of-concept simulations and experiments have been performed in previous years to demonstrate the validity of this tool in a laboratory setting. Development of the MIP Monitor approach continues to evaluate the efficacy of the monitor for automated, real-time or near-real-time application. This report details follow-on research and development efforts sponsored by the U.S. Department of Energy Fuel Cycle Research and Development related to the MIP Monitor for fiscal year 2012 (FY12).

fuel characterization↗
DOE OSTI2012

Iodine Sorbent Performance in FY 2012 Deep Bed Tests

Nuclear fission results in the production of fission products and activation products, some of which tend to be volatile during used fuel reprocessing and evolve in gaseous species into the reprocessing facility off-gas systems. Analyses have shown that I-129, due to its radioactivity, high potential mobility in the environment, and high longevity (half life of 15.7 million years), can require control efficiencies of up to 1,000x or higher to meet regulatory emission limits. Iodine capture is an important aspect of the Separations and Waste Forms Campaign Off-gas Sigma Team (Jubin 2011, Pantano 2011). Deep-bed iodine sorption tests for both silver-functionalized Aerogel and silver zeolite sorbents were performed during Fiscal Year 2012. These tests showed that: • Decontamination factors were achieved that exceed reasonably conservative estimates for DFs needed for used fuel reprocessing facilities in the U.S. to meet regulatory requirements for I-129 capture. • Silver utilizations approached or exceeded 100% for high inlet gas iodine concentrations, but test durations were not long enough to approach 100% silver utilization for lower iodine concentrations. • The depth of the mass transfer zone was determined for both low iodine concentrations (under 10 ppmv) and for higher iodine concentrations (between 10-50 ppmv); the depth increases over time as iodine is sorbed. • These sorbents capture iodine by chemisorption, where the sorbed iodine reacts with the silver to form very non-volatile AgI. Any sorbed iodine that is physisorbed but not chemically reacted with silver to form AgI might not be tightly held by the sorbent. The portion of sorbed iodine that tends to desorb because it is not chemisorbed (reacted to form AgI) is small, under 1%, for the AgZ tests, and even smaller, under 0.01%, for the silver-functionalized Aerogel.

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

Real-Time Detection Methods to Monitor TRU Compositions in UREX+Process Streams

The U.S. Department of Energy has developed advanced methods for reprocessing spent nuclear fuel. The majority of this development was accomplished under the Advanced Fuel Cycle Initiative (AFCI), building on the strong legacy of process development R&D over the past 50 years. The most prominent processing method under development is named UREX+. The name refers to a family of processing methods that begin with the Uranium Extraction (UREX) process and incorporate a variety of other methods to separate uranium, selected fission products, and the transuranic (TRU) isotopes from dissolved spent nuclear fuel. It is important to consider issues such as safeguards strategies and materials control and accountability methods. Monitoring of higher actinides during aqueous separations is a critical research area. By providing on-line materials accountability for the processes, covert diversion of the materials streams becomes much more difficult. The importance of the nuclear fuel cycle continues to rise on national and international agendas. The U.S. Department of Energy is evaluating and developing advanced methods for safeguarding nuclear materials along with instrumentation in various stages of the fuel cycle, especially in material balance areas (MBAs) and during reprocessing of used nuclear fuel. One of the challenges related to the implementation of any type of MBA and/or reprocessing technology (e.g., PUREX or UREX) is the real-time quantification and control of the transuranic (TRU) isotopes as they move through the process. Monitoring of higher actinides from their neutron emission (including multiplicity) and alpha signatures during transit in MBAs and in aqueous separations is a critical research area. By providing on-line real-time materials accountability, diversion of the materials becomes much more difficult. The objective of this consortium was to develop real time detection methods to monitor the efficacy of the UREX+ process and to safeguard the separated TRUs against unlawful diversion from within a processing facility. To achieve this, a comprehensive strategy was implemented to incorporate traditional detectors and advanced Tensioned Metastable Fluid (TMFD) metastable fluid detectors (developed, in part, under this project) into a novel detector assembly coupled to the UREX+ centrifugal contactor array. The sections below provide a brief summary of the technical achievements completed during this project. The principal outcomes are documented in more complete details contained the doctoral dissertations and masters theses, journal papers, conference proceedings and additional items for more than the 35 publications that are listed in the program bibliography in Section 3.

McDeavitt, Sean M.↗
DOE OSTI2013

Annual Report on Improved Accuracy of Pu Assay of Spent Fuel Project

The objective of this work is to improve the accuracy of plutonium assay for nuclear fuel reprocessing. This work addresses such questions as: 1) How much Pu has been recovered in a reprocessing facility?; 2) Is a reprocessing operation consistent with safeguards?; and 3) How much Pu is unaccounted for in a safeguarded facility? This improved Pu accuracy will be achieved by direct measurement of Pu using monochromatic excitation and detection. These elemental measurements will not only provide direct concentration measurements of plutonium in spent fuel dissolver solution. The increased accuracy of the Pu assay addresses the heart of safeguards by providing greater accuracy and less error in the Pu measurements of these facilities. This is further achieved by using the high energy Ka line of Pu at 103.65 keV. This allows for the direct measurement of Pu through container walls such as Zircaloy tubing which is used for the cladding of nuclear fuel rods. While ultrahigh energy X-ray fluorescence (UHEXRF) has been done at synchrotron facilities, X-ray optic technology has advanced to make possible, laboratory based instrumentation to enable such high energy measurements to be done on the bench top. Such instrumentation can change the plutonium assay measurement paradigm by offering safer, more rapid and more accurate Pu measurements through container walls nondestructively.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2014

SRNL CRP progress report [Development of Melt Processed Ceramics for Nuclear Waste Immobilization]

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multiphase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing.

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

Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208 (Final Report)

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

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

Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208 (Final Report)

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

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

Advanced Burner Reactor with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization

This study assesses the feasibility of designing Seed and Blanket (S&B) Sodium-cooled Fast Reactor (SFR) to generate a significant fraction of the core power from radial thorium fueled blankets that operate on the Breed-and-Burn (B&B) mode without exceeding the radiation damage constraint of presently verified cladding materials. The S&B core is designed to maximize the fraction of neutrons that radially leak from the seed (or “driver”) into the subcritical blanket and reduce neutron loss via axial leakage. The blanket in the S&B core makes beneficial use of the leaking neutrons for improved economics and resource utilization. A specific objective of this study is to maximize the fraction of core power that can be generated by the blanket without violating the thermal hydraulic and material constraints. Since the blanket fuel requires no reprocessing along with remote fuel fabrication, a larger fraction of power from the blanket will result in a smaller fuel recycling capacity and lower fuel cycle cost per unit of electricity generated. A unique synergism is found between a low conversion ratio (CR) seed and a B&B blanket fueled by thorium. Among several benefits, this synergism enables the very low leakage S&B cores to have small positive coolant voiding reactivity coefficient and large enough negative Doppler coefficient even when using inert matrix fuel for the seed. The benefits of this synergism are maximized when using an annular seed surrounded by an inner and outer thorium blankets. Among the high-performance S&B cores designed to benefit from this unique synergism are: (1) the ultra-long cycle core that features a cycle length of ~7 years; (2) the high-transmutation rate core where the seed fuel features a TRU CR of 0.0. Its TRU transmutation rate is comparable to that of the reference Advanced Burner Reactor (ABR) with CR of 0.5 and the thorium blanket can generate close to 60% of the core power; but requires only one sixth of the reprocessing and fabrication capacity per unit of core power. Nevertheless, these high-performance cores were designed to set upper bounds on the S&B core performance by using larger height and pressure drop than those of typical SFR design. A study was subsequently undertaken to quantify the tradeoff between S&B core design variables and the core performance. This study concludes that a viable S&B core can be designed without significant deviation from SFR core design practices. For example, the S&B core with 120cm active height will be comparable in volume, HM mass and specific power with the S-PRISM core and could fit within the S-PRISM reactor vessel. 43% of this core power will be generated by the once-through thorium blanket; the required capacity for reprocessing and remote fuel fabrication per unit of electricity generated will be approximately one fifth of that for a comparable ABR. The sodium void worth of this 120cm tall S&B core is significantly less positive than that of the reference ABR and the Doppler coefficient is only slightly smaller even though the seed uses a fertile-free fuel. The seed in the high transmutation core requires inert matrix fuel (TRU-40Zr) that has been successfully irradiated by the Fuel Cycle Research & Development program. An additional sensitivity analysis was later conducted to remove the bias introduced by the discrepancy between radiation damage constraints -- 200 DPA applied for S&B cores and fast fluence of 4x1023 n(>0.1MeV)/cm2 applied for ABR core design. Although the performance characteristics of the S&B cores are sensitive to the radiation damage constraint applied, the S&B cores offer very significant performance improvements relative to the conventional ABR core design when using identical constraint.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗