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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

Introduction of plutonium systems to the nuclear criticality slide rule

AWE (UK), IRSN (France), LLNL (USA) and ORNL (USA) began a long term collaboration effort in 2015 to update the nuclear criticality Slide Rule for the emergency response to a nuclear criticality accident. This document, published almost 20 years ago, gives order of magnitude estimates of key parameters, such as number of fissions and doses (neutron/gamma), useful for emergency response teams and public authorities. This paper presents, firstly, the motivation and the long term objectives for this update, then it introduces the initial critical configuration, and finally it presents the preliminary results obtained with modern 3D radiation transport codes and nuclear data for new configurations including plutonium systems. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Adsorption Modeling at the Front and Back Ends of the Nuclear Fuel Cycle

Adsorption is a diverse and complex physicochemical process that has many different applications in chemical, environmental, and nuclear sciences. For example, adsorption can be found at both the front and back ends of the nuclear fuel cycle for (i) capturing uranium from seawater to use as a new source of nuclear fuel and (ii) treatment of the off-gases produced when reprocessing used nuclear fuel. Accurate modeling of these processes is important for the design of efficient capture, recovery, and removal systems for the particular aqueous or gaseous species of interest. A set of algorithms and methods designed for flexibility and portability will greatly improve the ability to rigorously model such a complex set of systems. This approach allows more specific models to be built upon the more generic methods, which gives the overall modeling effort larger breadth and greater usability. Due to the complexity of adsorption processes, it is necessary to develop relatively complex mechanistic models to accurately predict the behavior of engineered adsorption systems. Capture and recovery systems designed for nuclear energy need to have a high level of accuracy to reduce uncertainties and aid in the implementation of separations technologies for efficient nuclear energy systems. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Quantification of Residual Water in Spent Fuel Dry Storage Canisters Using Mass Spectrometry

The purpose of this report is to document updates on the apparatus to simulate commercial vacuum drying procedures at the Nuclear Energy Work Complex at Sandia National Laboratories. Validation of the extent of water removal in a dry spent nuclear fuel storage system based on drying procedures used at nuclear power plants is needed to close existing technical gaps. Operational conditions leading to incomplete drying may have potential impacts on the fuel, cladding, and other components in the system during subsequent storage and disposal, such as fuel degradation; cladding corrosion, embrittlement, or breaching; and the creation of a flammable environment via radiolysis of water. A general lack of data suitable for model validation of commercial nuclear canister drying processes necessitates well-designed investigations of drying process efficacy and water retention. Scaled tests that incorporate relevant physics and well-controlled boundary conditions are essential to provide insight and guidance to the simulation of prototypic systems undergoing drying processes. This report documents details on the quantification of residual water in the Advanced Drying Cycle Simulator (ADCS), an apparatus built to simulate commercial drying procedures and quantify the amount of residual water remaining in a pressurized water reactor (PWR) fuel assembly after drying. The ADCS was constructed with a prototypic 17×17 PWR fuel skeleton and waterproof heater rods to simulate decay heat. The ADCS is outfitted with thermocouples to measure the thermal response of the ADCS to simulated decay heats and internal helium fill pressures relevant to commercial drying procedures. The ADCS is also instrumented with pressure transducers to measure the pressures and vacuum levels observed during simulated commercial drying. The most unique instrumentation used for quantifying residual water in the ADCS is a Hiden Analytical HPR-30 mass spectrometer (MS), which measures gas compositions of the ADCS internal free volume, based on partial pressures calculated from relative proportions of gas molecules detected by the MS. This report details the methodology used to implement MS measurements in quantifying residual water in the ADCS. This methodology includes the calibration of the HPR-30 MS to a Buck Research Instruments CR-4 chilled mirror hygrometer, which itself is calibrated to a NIST-traceable standard. Data collected by both the MS and the chilled mirror hygrometer from water/helium mixtures ranging from 150 to 500,000 ppmv water in helium were used to generate calibration curves, establishing a source of verification of MS measured water contents. Details regarding water content measurement uncertainties are included in this report, defining the accuracy and verifiability of the HPR-30 MS in measuring residual water content in simulated dry storage canister environments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Modeling thermochemistry of fuel and coupling to fuel performance codes

A thermochemical library, Thermochimica, utilizes new algorithms and solvers designed for calculating equilibria of multicomponent and multiphase systems and has been coupled to the nuclear fuel performance code BISON. Thermochimica utilizes a generalized thermochemical database for urania-based fuels that has been under development for a number of years and continues to be expanded. One of the key motivations for use of thermodynamic models is that they can better represent the physics of oxygen migration. The coupling of the BISON code with Thermochimica and examples of its use in representing fuel phenomena are described. In the current implementation in BISON, fluxes are driven by the elemental concentration gradient. However, equilibrium is characterized by the absence of spatial variations of chemical potentials of the system species/components. The driving force to equilibrium (species flux) is proportional to the deviation from equilibrium, i.e., the gradient in chemical potential. It is therefore necessary to integrate chemical potential driven diffusion in the BISON representation of oxygen transport. The diffusion kernels based on thermodynamic models have been developed to simulate multicomponent transport in light water reactor (LWR) materials. The Thermochimica library was used to calculate various thermodynamics properties needed for transport calculations. The library also can calculate properties that are specific to various transport mechanisms for different fuel materials. For example, it can calculate defect site fractions among the sublattices in oxide fuels as a function of temperature, burnup, and stoichiometry. The transport and thermodynamic models are being integrated with other simulation efforts such as calculation of mobility functions that couple fluxes and chemical gradients. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Integrated Reflection Seismic Monitoring and Reservoir Modeling for Geologic CO 2 Sequestration

The US DOE/NETL CCS MVA program funded a project with Fusion Petroleum Technologies Inc. (now SIGMA) to model the proof of concept of using sparse seismic data in the monitoring of CO 2 injected into saline aquifers. The goal of the project was to develop and demonstrate an active source reflection seismic imaging strategy based on deployment of spatially sparse surface seismic arrays. The primary objective was to test the feasibility of sparse seismic array systems to monitor the CO 2 plume migration injected into deep saline aquifers. The USDOE/RMOTC Teapot Dome (Wyoming) 3D seismic and reservoir data targeting the Crow Mountain formation was used as a realistic proxy to evaluate the feasibility of the proposed methodology. Though the RMOTC field has been well studied, the Crow Mountain as a saline aquifer has not been studied previously as a CO 2 sequestration (storage) candidate reservoir. A full reprocessing of the seismic data from field tapes that included prestack time migration (PSTM) followed by prestack depth migration (PSDM) was performed. A baseline reservoir model was generated from the new imaging results that characterized the faults and horizon surfaces of the Crow Mountain reservoir. The 3D interpretation was integrated with the petrophysical data from available wells and incorporated into a geocellular model. The reservoir structure used in the geocellular model was developed using advanced inversion technologies including Fusion's ThinMAN™ broadband spectral inversion. Seal failure risk was assessed using Fusion's proprietary GEOPRESS™ pore pressure and fracture pressure prediction technology. CO 2 injection was simulated into the Crow Mountain with a commercial reservoir simulator. Approximately 1.2MM tons of CO 2 was simulated to be injected into the Crow Mountain reservoir over 30 years and subsequently let 'soak' in the reservoir for 970 years. The relatively small plume developed from this injection was observed migrating due to gravity to the apexes of the double anticline in the Crow Mountain reservoir of the Teapot dome. Four models were generated from the reservoir simulation task of the project which included three saturation models representing snapshots at different times during and after simulated CO 2 injection and a fully saturated CO 2 fluid substitution model. The saturation models were used along with a Gassmann fluid substitution model for CO 2 to perform fluid volumetric substitution in the Crow Mountain formation. The fluid substitution resulted in a velocity and density model for the 3D volume at each saturation condition that was used to generate a synthetic seismic survey. FPTI's (Fusion Petroleum Technologies Inc.) proprietary SeisModelPRO™ full acoustic wave equation software was used to simulate acquisition of a 3D seismic survey on the four models over a subset of the field area. The simulated acquisition area included the injection wells and the majority of the simulated plume area.

02 PETROLEUM↗
DOE OSTI2025

Investigating the Role of Accident Tolerant Cladding on Source Term Reduction for High-Burnup PWRs Using MELCOR

The use of accident tolerant fuel (ATF) cladding can increase coping times during and beyond design basis accidents. While such gains may be incremental, they provide a margin that can potentially be recovered to enable high-burnup (HBU) operation. Realizing such a margin requires demonstrating that the combination of HBU and ATF has not led to an overall increase in source term. This study investigates the influence of cladding technology (Zr-based, Cr-coated Zr, and FeCrAl) and fuel cycle length (18 and 24 months) on radiological dose at the boundary of the exclusion zone for a four-loop pressurized water reactor to investigate whether ATF claddings can provide such benefits. We analyze a recovered large break loss-of-coolant accident scenario to investigate the impact of transient timescale on the benefits of such coping time increases. The simulations have been performed using the MELCOR and MELCOR Accident Consequence Code System codes. For the cases analyzed, increased fuel cycle length did not necessarily increase radionuclide release and hydrogen generation, as these were found to be sensitive to the core power distribution. Similarly, off-site dose consequence is dominated by short-lived radionuclides that tend to saturate earlier in the burnup, so higher burnup operation did not necessarily increase the source term for the phenomena and transients analyzed here. Delays in recovery of the lowpressure safety injection system increase hydrogen production and radionuclide release, especially between 780 s and 1620 s, due to the nonlinear oxidation and core degradation behavior. Results show that Cr-coated Zr enhances safety by delaying heatup and gap release. Here, when uncertainty propagation on oxidation properties is considered, FeCrAl exhibits the lowest overall radionuclide release and off-site dose throughout the spectrum. However, while the considered “base model” performance is superior under delayed injection scenarios, upper-bound cases display hydrogen generation risk comparable to the Zr-based cladding.

Accident Tolerant Fuel↗
DOE OSTI2024

Fusion Energy Research at Idaho National Laboratory: Experimentation and Simulation to Support Safety and Rapid Technology Development

Research into fusion energy is growing rapidly, responding to a call for sustainable sources of energy to replace fossil fuels and mitigate climate change. Within the United States, at least, researchers are also responding to the “Bold Decadal Vision” proposed by the White House, seeking to have a commercially relevant fusion pilot plant deployed within a decade. Before this can become a reality, many Fusion Science & Technology (FS&T) gaps remain. For over 45 years, Idaho National Laboratory has been at the forefront of addressing these FS&T gaps in the context of fusion safety and technology via the operation of world-leading experimental facilities within the Safety and Tritium Applied Research (STAR) Facility. Here, INL focuses on the tritium fuel cycle, conceptual system design studies, risk assessment, waste management, and materials safety. Modeling and Simulation (M&S) has also been a component of this portfolio of research, but, early on, focused on individual systems. Since 2019, active development and research on integrated whole device modeling tools based on the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework has been undertaken. This has culminated in a MOOSE-based version of the Tritium Migration and Analysis Program (TMAP), an INL code historically focused on tritium permeation and trapping within fusion systems. More recently, INL Laboratory Directed Research and Development funds have been used to create the Fusion ENergy Integrated multiphys-X (FENIX) code focused on scrape-off layer plasma physics and the first wall of a magnetically confined fusion device. This talk will focus on an overview of INL activities in the FS&T research area, with a particular focus on recent M&S activities and results.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

Corrosion property of 9Cr-ODS steel in nitric acid solution for spent nuclear fuel reprocessing

Corrosion tests of oxide dispersion strengthened with 9% Cr (9Cr-ODS) steel, which is one of the desirable materials for cladding tube of sodium-cooled fast reactors, in pure nitric acid solution, spent FBR fuel solution, and its simulated solution were performed to understand the corrosion behavior in a spent nuclear fuel reprocessing. In this study, the 9Cr-ODS steel with lower effective chromium content was evaluated to understand the corrosion behavior conservatively. As results, the tube-type specimens of the 9Cr-ODS steels suffered severe weight loss owing to active dissolution at the beginning of the immersion test in pure nitric acid solution in the range from 1 to 3.5 M. In contrast, the weight loss was decreased and they showed a stable corrosion in the higher nitric acid concentration, the dissolved FBR fuel solution, and its simulated solution by passivation. The corrosion rates of the 9Cr-ODS steel in the dissolved FBR fuel solution and its simulated solution were 1-2 mm/y and showed good agreement with each other. The passivation was caused by the shift of corrosion potential to noble side owing to increase in nitric acid concentration or oxidative ions in the dissolved FBR fuel solution and the simulated spent fuel solution. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Considerations for Thermal Annealing of Zr-Alloy Cladding During Dry Storage

In January 2025, the U.S. Department of Energy (DOE) sponsored a technical workshop with fuel vendors that provided a forum for increased collaboration on backend fuel cycle research. The workshop reviewed observations of thermal annealing of irradiated zirconium alloys in simulated dry storage conditions and discussed implications to the industry’s interest in reducing wet storage to optimize operational flexibility and the expected trend of increasing discharge burnup (and decay heats), which could lead to an increase in dry storage temperatures. One approach to accommodate increased decay heats in dry storage is to increase the regulatory temperature limit during dry storage, which is now 400°C. However, this could lead to thermal annealing of the cladding, which will tend to increase creep rates, creep strains, and ductility while decreasing yield strength. These considerations are compounded by bonding between the pellet and cladding at high burnup, which results in the pellet carrying more load during fuel rod deformations, especially in bending, which is the dominant deformation phenomenon in canister drop analyses. Because the U.S. Nuclear Regulatory Commission (NRC) recommends yield strength as a primary failure criterion for accident analyses in storage and transportation, particularly for canister drop scenarios, the purpose of the workshop was to develop options for alternative failure criterion that could replace yield strength in canister drop analyses. Topics discussed were (1) the regulatory framework, (2) the effect of thermal annealing on microhardness and tensile properties of cladding materials, (3) the effect of thermal annealing on fuel rod failure during bending, bending fatigue, and pinch loading, and (4) an alternative failure criterion for canister drop analyses. This paper provides a summary of the key discussions and outcomes of the workshop

Cantonwine, Paul [ORNL] (ORCID:0009000522247033)↗
DOE OSTI2015

Synchrotron characterization of nanograined UO 2 grain growth

This activity is supported by the US Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Product Line (FPL) and aims at providing experimental data for the validation of the mesoscale simulation code MARMOT. MARMOT is a mesoscale multiphysics code that predicts the coevolution of microstructure and properties within reactor fuel during its lifetime in the reactor. It is an important component of the Moose-Bison-Marmot (MBM) code suite that has been developed by Idaho National Laboratory (INL) to enable next generation fuel performance modeling capability as part of the NEAMS Program FPL. In order to ensure the accuracy of the microstructure based materials models being developed within the MARMOT code, extensive validation efforts must be carried out. In this report, we summarize our preliminary synchrotron radiation experiments at APS to determine the grain size of nanograin UO 2 . The methodology and experimental setup developed in this experiment can directly apply to the proposed in-situ grain growth measurements. The investigation of the grain growth kinetics was conducted based on isothermal annealing and grain growth characterization as functions of duration and temperature. The kinetic parameters such as activation energy for grain growth for UO 2 with different stoichiometry are obtained and compared with molecular dynamics (MD) simulations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Assessment of Pellet-Clad Interaction Indicators in Watts Bar Unit 1 using the VERA Framework

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity, multiphysics simulations of light water reactors (LWRs) by coupling a variety of codes within the Virtual Environment for Reactor Analysis (VERA). One of the primary goals of CASL is to predict local cladding failure through pellet-clad interaction (PCI). This capability is currently being pursued through several different approaches, such as with Tiamat, which is a simulation tool within VERA that more tightly couples the MPACT neutron transport solver, the CTF thermal hydraulics solver, and the MOOSE-based Bison-CASL fuel performance code. However, the process in this paper focuses on running fuel performance calculations with Bison-CASL to predict PCI using the multicycle output data from coupled neutron transport/thermal hydraulics simulations. In recent work within CASL, Watts Bar Unit 1 has been simulated over 12 cycles using the VERA core simulator capability based on MPACT and CTF. Using the output from these simulations, Bison-CASL results can be obtained without rerunning all 12 cycles, while providing some insight into PCI indicators. Multi-cycle Bison-CASL results are presented and compared against results from the FRAPCON fuel performance code. There are several quantities of interest in considering PCI and subsequent fuel rod failures, such as the clad hoop stress and maximum centerline fuel temperature, particularly as a function of time. Bison-CASL performs single-rod simulations using representative power and temperature distributions, providing high-resolution results for these and a number of other quantities. This will assist in identifying fuels rods as potential failure locations for use in further analyses.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Use of Monte Carlo simulation to quantify CIPS risk

Crud Induced Power Shift (CIPS), caused by the accumulation of boron compounds in corrosion product deposits on high power fuel assemblies, is an important operational issue for Pressurized Water Reactors (PWRs). The Boron-induced Offset Analysis (BOA) computer code has been developed by EPRI in collaboration with Westinghouse to help utility engineers avoid onset of CIPS. The code, used with an established procedure, predicts the maximum boron mass that will be accumulated in the upcoming cycle for a given core design and plant chemistry and operating program. The calculated boron mass above a given threshold predicts the risk of CIPS to the upcoming cycle. BOA has been successful in correctly predicting the risk for onset or lack of CIPS in many cycles. However, like most codes modeling complex physical phenomena, it is subject to uncertainties in several of the inputs used to describe these phenomena. The current BOA procedure results in a single 'best estimate' prediction, which has been judged to provide engineers with a practical way to assess CIPS risk. As such, however, there is a probability that the actual plant CIPS performance may be underestimated. This paper describes a method to assess the probability (with an associated confidence level) of no CIPS, mild to moderate CIPS, or severe CIPS using BOA for a given core design. This is accomplished through the use of a Monte Carlo simulation. The simulation is applied to several plant examples, and the results are evaluated in terms of the likelihood of mild or severe CIPS. A suggested set of criteria for decision making regarding CIPS risk is presented in the conclusion. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Equilibrium cycle pin by pin transport depletion calculations with DeCART

As the Advanced Fuel Cycle Initiative (AFCI) program has matured it has become more important to utilize more advanced simulation methods. The work reported here was performed as part of the AFCI fellowship program to develop and demonstrate the capability of performing high fidelity equilibrium cycle calculations. As part of the work here, a new multi-cycle analysis capability was implemented in the DeCART code which included modifying the depletion modules to perform nuclide decay calculations, implementing an assembly shuffling pattern description, and modifying iteration schemes. During the work, stability issues were uncovered with respect to converging simultaneously the neutron flux, isotopics, and fluid density and temperature distributions in 3-D. Relaxation factors were implemented which considerably improved the stability of the convergence. To demonstrate the capability two core designs were utilized, a reference UOX core and a CORAIL core. Full core equilibrium cycle calculations were performed on both cores and the discharge isotopics were compared. From this comparison it was noted that the improved modeling capability was not drastically different in its prediction of the discharge isotopics when compared to 2-D single assembly or 2-D core models. For fissile isotopes such as U-235, Pu-239, and Pu-241 the relative differences were 1.91%, 1.88%, and 0.59%), respectively. While this difference may not seem large it translates to mass differences on the order of tens of grams per assembly, which may be significant for the purposes of accounting of special nuclear material. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Fabrication of New NFA-SiC Composites for Nuclear Applications

Cladding materials play the critical and fundamental function of fuel confinement during the safe operation of a nuclear power system. With the development of a new generation of nuclear reactors, investigation of the corresponding advanced cladding materials is attracting intensive scrutiny. Nanostructured ferritic alloy (NFA), which has excellent irradiation resistance and mechanical performance, is one primary candidate for future fission and fusion reactors. The NFA alloys are composed of nano-grained Fe-Cr alloy domains and Y-Ti-O-Cr-Fe enriched nano-clusters. The nano-clusters can not only prevent dislocation glide, grain growth, and grain boundary slip, but also act as sinks to trap helium atoms and radiation-generated point defects. In addition, NFAs can provide high temperature strength and prominent thermal stability, which are beneficial to nuclear cladding applications. Silicon carbide (SiC) is another nuclear cladding candidate. Due to its physical and chemical stability, it shows very good refractory properties at high temperatures up to 2000 deg. C, which has enabled its wide applications in combustion chamber and aerospace engines. In addition, SiC has the remarkable characteristic of radiation stability. SiCf/SiC composite has been developed as an alternative cladding material to metallic alloys. Theoretical simulation is also being developed to understand irradiation induced defects in SiC. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Atomistic Simulations in Support of UO{sub 2} Fuel Performance Modeling

Large fission gas atoms such as Xe have very low solubility in the UO{sub 2} fuel matrix, which drives precipitation and growth of intra-granular fission gas bubbles, segregation of gas atoms to grain boundaries and evolution of intergranular bubbles. The inter-granular bubbles develop percolation networks leading to release of the fission gas. Fission gas bubbles induce fuel swelling and the fuel thermal conductivity is reduced by both fission gas bubbles and dispersed gas atoms. When fission gas is released from the fuel to the plenum, the pressure on the clad walls increases and the gap thermal conductivity decreases. All of these detrimentally affect fuel performance. The thermal conductivity of UO{sub 2} is low across a wide temperature range compared to iso-structural materials such as ThO{sub 2}, which is caused by the coupling between phonons carrying heat and uranium spins or magnetism providing a scattering mechanism with very low mean free path. For in-pile conditions the thermal conductivity is further reduced due to, for example, radiation damage, generation of fission products and fission gases and fuel microstructure evolution. This results in a complex burn-up dependent thermal conductivity. Thermal conductivity and fission gas evolution are two of the most important properties determining UO{sub 2} nuclear fuel performance. Both properties have significant uncertainty. In order to reduce this uncertainty, new models are being developed based on multi-scale simulations that attempt to improve the description of material properties and also couple them to the microstructure state of the material. In this paper we highlight atomistic calculations and simulations used to develop meso-scale models of the evolution of fission gases and thermal conductivity in UO{sub 2}. Literature examples of the application of these results and the corresponding models to fuel performance simulations will also be given. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Using Machine Learning to Analyze Factors Determining Cycle-to-Cycle Variation in a Spark-Ignited Gasoline Engine

In this work, we have applied a machine learning (ML) technique to provide insights into the causes of cycle-to-cycle variation (CCV) in a gasoline spark-ignited (SI) engine. The analysis was performed on a set of large eddy simulation (LES) calculations of a single cylinder of a four-cylinder port-fueled SI engine. The operating condition was stoichiometric, without significant knock, at a load of 16 bar brake mean effective pressure (BMEP), at an engine speed of 2500 revolutions per minute. A total of 123 cycles was simulated. Of these, 49 were run in sequence, while 74 were run in parallel. For the parallel approach, each cycle is initialized with its own synthetic turbulent field to generate CCV, as part of another work performed by us. In the current work, we used 3D information from all 123 cycles to compute flame topology and pre-ignition flow-field metrics. We then evaluated correlations between these metrics, and peak cylinder pressure (PCP) employing an ML technique called random forest. The computed metrics form the inputs to the random forest model, and PCP is the output. This model captures the effect of all inputs, as well as interactions between them owing to its decision-tree structure. The goal of this work is to demonstrate (as a first step) that ML models can implicitly learn complex relationships between pre-ignition flow-fields, flame shapes, and the eventual outcome of the cycle (whether a cycle will be a high or a low cycle).

42 ENGINEERING↗
DOE OSTI2016

Multicomponent adsorption of radioactive iodine and krypton using ETS-10 supported carbon nano-polyhedrons

Over 13% of the world's energy demand is supplied by nuclear power plants in 31 countries. This demand is growing; by 2030 the IAEA predicts a 100% increase in the demand for nuclear energy. The increase in nuclear energy production imposes a growing strain on the uranium fuel supply. At current consumption rates the world's supply of uranium is projected to last ∼100 years. Following the IAEA's prediction for increased nuclear energy, uranium supplies will dwindle even more quickly. However, this consequence can be limited as ∼95% of uranium is left in used nuclear fuel (UNF). Reprocessing of UNF will utilize this remaining uranium and extend the lifetime of the world's uranium supply. One of the concerns with nuclear reprocessing is that it frees radionuclides created by the fission of uranium to be released into the atmosphere through off-gas. The major constituents of off-gas include: {sup 85}Kr, {sup 129}I, {sup 14}C, {sup 3}H, and various Xe isotopes. {sup 85}Kr and {sup 129}I present the largest threat due to their harmful environmental effects and abundance. They are the most practical contaminants to be removed to meet requirements for radionuclide release. Removal of these contaminants has been a challenge since the 1950's, and several methods including cryogenic distillation, fluorocarbon absorption, wet scrubbing, and adsorption have been employed. Currently, adsorption on porous solid sorbents shows the most promising results. Until now, several sorbents such as silver impregnated mordenite, silver Engelhard titanosilicate 10, chalcogen based aerogels, and metal organic framework, MOFs have been investigated for this purpose. Despite these advances, development of economical and highly active sorbents for complete removal and immobilization of volatile radioactive gases remains a key challenge in realizing advanced fuel cycles. This work is focused on the multicomponent of {sup 85}Kr and {sup 129}I from a simulated off-gas with an Engelhard titanosilicate 10 (ETS-10) supported carbon nano-polyhedron sorbent. Adsorption studies were performed to determine iodine sorption capacity as a function of temperature and iodine inlet concentration. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗