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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 973 records · Page 54
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

High Flux Isotope Reactor Low-Enriched Uranium Low Density Silicide Fuel Design Parameters

High Flux Isotope Reactor (HFIR) highly enriched uranium (HEU) to low-enriched uranium (LEU) conversion activities are ongoing as part of the Department of Energy (DOE) National Nuclear Security Administration (NNSA)’s nuclear nonproliferation mission. Design activities studying the conversion of HFIR from HEU to LEU fuel explored different fuel design features and shapes with a low density uranium-silicide dispersion (U 3 Si 2 -Al) fuel, which has a uranium density of 4.8 gU/cm 3 . The goal of these studies is to generate several HFIR LEU fuel designs of varying fuel fabrication complexity that meet the current HEU performance metrics and safety requirements. The documented designs will serve as references for fuel fabrication and qualification activities. Recent advancements in modeling and simulation tools enable quick prototyping of fuel designs. Shift, a Monte Carlo neutron transport and depletion tool optimized for high-performance computing (HPC) architectures, is used for efficient fuel cycle and performance metrics calculations. The HFIR Steady State Heat Transfer Code (HSSHTC) is used to vet the thermal safety margin. Also, a new automation tool that connects all fuel design analysis steps, named Python HFIR Analysis and Measurement Engine (PHAME), has been developed to expedite the design study in an efficient and reproducible manner. Leveraging these tools, several candidate fuel designs were selected for varying fabrication complexity. This report provides design feature details for four selected HFIR LEU low density U 3 Si 2 -Al fuel designs and their corresponding performance and safety metrics. Nominal, best-estimate design parameters and irradiation conditions, including fission rate densities, power densities, heat fluxes, and cumulative fission densities are provided for candidate fuel designs relevant to framing irradiation experiments to support fuel qualification efforts. Simulations show that the low density U 3 Si 2 -Al, with design features to enhance safety, can meet HEU core performance metrics and safety requirements if the reactor power is increased from 85 MW (HEU) to 95 MW (LEU) and if the active fuel length is increased from 50.80 cm (HEU) to 55.88 cm (LEU).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Computational modeling and validation of a modified Marple cascade impactor

Oak Ridge National Laboratory (ORNL) is investigating respirable aerosol fractions from failed high-burnup (HBU) commercial spent nuclear fuel (CSNF) rods (Montgomery et al., 2022). The work aims to characterize the source term of radioactive aerosols that could result from mechanical failure and handling of failed fuel rods during storage and/or transportation. The experiment involves capturing and characterizing aerosols from a four-point bending experiment of fueled rod segments at ORNL’s Irradiated Fuels Examination Laboratory (IFEL). For this, a Marple cascade impactor (Series 290) with the ability to collect particles up to aerodynamic equivalent diameters of ~15 μm was tested for feasibility. The commercially available Marple cascade impactor design was modified for the experiment, where the cascade inlet was replaced with a 3D printed inlet nozzle. Collection efficiency curves were then generated with the modified impactor, and the cascade’s performance was tested using ISO dust capture experiments followed by SEM analysis of the collection and computational modeling. The current work highlights the performance testing and experimental validation of this modified Marple Impactor for spent nuclear fuel aerosol collection.

Kumar, Vineet↗
DOE OSTI2014

Performance Assessment Modeling and Sensitivity Analyses of Generic Disposal System Concepts

The Used Fuel Disposition Campaign (UFDC) of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) is conducting research and development (R&D) on generic deep geologic disposal systems (i.e., repositories) for high-activity nuclear wastes (i.e., used nuclear fuel (UNF) and high-level radioactive waste (HLW)) that exist today or that could be generated in future fuel cycles. This report describes specific activities in FY2014 toward the development of an enhanced generic disposal system modeling and analysis capability that utilizes high performance computing (HPC) environments to simulate important multi-physics phenomena and couplings associated with the potential behavior of a geologic repository for UNF and HLW.

Sevougian, S. David↗
DOE OSTI2021

Sensitivity and Uncertainty of the IFR-1 BISON Benchmark

The fuel performance code BISON is being used to evaluate metallic fuel for a new fast-spectrum test reactor called the Versatile Test Reactor, which is being considered by the US Department of Energy. To quantify the accuracy of BISON predictions, researchers at Oak Ridge National Laboratory have been developing a series of benchmarks based on legacy metallic fuel experiments. As part of this effort, the sensitivity of BISON predictions to variations in model inputs and the uncertainties associated with BISON predictions must be established. This report summarizes efforts to perform a comprehensive sensitivity analysis (SA) and uncertainty quantification (UQ) on a benchmark based on the IFR-1 experiment. For the SA, at least one input was chosen from every BISON model and physics module used in the benchmark. The inputs were varied individually in a series of BISON simulations. The resulting variations in benchmark predictions were normalized to calculate sensitivities. The strongest sensitivities were identified and used to inform input selections for the UQ. The UQ was performed using the Monte Carlo UQ method. A literature review was conducted to estimate uncertainty distributions for the selected inputs, and values were sampled randomly from each distribution in a series of BISON simulations. Variations in the benchmark predictions were used to estimate uncertainty distributions and confidence intervals. It was found that nearly 100% of benchmark predictions matched the corresponding legacy values within the confidence intervals. However, this is at least partially because of the wide confidence intervals associated with the benchmark predictions. The uncertainty contributions of assumptions in the benchmark, experimental uncertainties, and BISON models were quantified. Some analysis was performed to identify inputs that contributed to the uncertainties. Finally, recommendations are made for future benchmark development and future BISON development.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Core design and performance of the Westinghouse lead fast reactor with UO 2 and MOX configurations

For this work, Westinghouse partnered with Argonne National Laboratory to design, model and optimize UO 2 - and MOX-fueled core designs for a medium size (950 MWt) Lead Fast Reactor that was pursued by Westinghouse. Using Argonne’s suite of reactor analysis codes together with Westinghouse fuel cost economic models, thousands of candidate cores were considered to achieve the economics-optimized cores presented in this paper. This optimization process considered detailed reactor physics, fuel performance, transient performance, and fuel economics models. The reactor performance of the resulting optimized UO 2 - and MOX-fueled core designs are described and compared in this paper. Both cores show fuel performance and transient behavior that is considered acceptable for the optimization presented herein, while further testing campaigns on material performance in high-temperature liquid lead will be required to confirm acceptability at the operating conditions chosen. A multi-batch strategy was selected for the UO 2 core for best fuel utilization with minimum fuel inventory costs. A single-batch fuel management was instead selected for the MOX core to maximize cycle length and minimize the impact of the longer refueling outage resulting from the higher decay heat of the discharged MOX fuel relative to the discharged UO 2 fuel, requiring a longer cooling time before dry-lift of discharged fuel could take place.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Diagnosing magnetized liner inertial fusion experiments on Z

Magnetized Liner Inertial Fusion experiments performed at Sandia's Z facility have demonstrated significant thermonuclear fusion neutron yields (∼10{sup 12} DD neutrons) from multi-keV deuterium plasmas inertially confined by slow (∼10 cm/μs), stable, cylindrical implosions. Effective magnetic confinement of charged fusion reactants and products is signaled by high secondary DT neutron yields above 10{sup 10}. Analysis of extensive power, imaging, and spectroscopic x-ray measurements provides a detailed picture of ∼3 keV temperatures, 0.3 g/cm{sup 3} densities, gradients, and mix in the fuel and liner over the 1–2 ns stagnation duration.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS1972

Technology for design of transport aircraft. Lecture notes for MIT courses: Seminar 1.61 freshman seminar in air transportation and graduate course 1.201, transportation systems analysis

The design parameters which determine cruise performance for a conventional subsonic jet transport are discussed. It is assumed that the aircraft burns climb fuel to reach cruising altitude and that aeronautical technology determines the ability to carry a given payload at cruising altitude. It is shown that different sizes of transport aircraft are needed to provide the cost optimal vehicle for different given payload-range objectives.

Simpson, R. W.↗
DOE OSTI2018

Status Report on the Component Models Developed in the Modelica Framework: Reverse Osmosis Desalination Plant & Thermal Energy Storage

This report has been prepared as part of an effort to design and build a Modeling and Simulation (M&S) framework to assess the economic viability of a Nuclear-Renewable Hybrid Energy System (N-R HES). In order to facilitate dynamic M&S of such an integrated system, research groups in multiple national laboratories and universities have been developing various subsystems as dynamic physics-based components using the Modelica programming language. In Fiscal Years (FYs) 2015, Idaho National Laboratory (INL) performed a dynamic analysis of two region-specific N-R HES configurations, including the gas-to-liquid (natural gas to Fischer-Tropsch synthetic fuel) and brackish water Reverse Osmosis (RO) desalination plants as industrial processes. In FYs 2016–2017, INL developed two additional subsystems in the Modelica framework: (1) a high-temperature steam electrolysis plant as a high priority industrial plant to be integrated with a light water reactor within an N-R HES and (2) a gas turbine power plant as a secondary energy supply. In FY 2018, the RO desalination system model developed in FY 2015 has been updated such that the model is compatible with the most recent version of the ThermoPower library. Special attention has been given to the controller settings based on process models, aiming to improve process dynamics and controllability. A dynamic performance analysis of the updated RO desalination plant was carried out to evaluate the technical feasibility (load-following capability) of such a system operating under highly variable conditions requiring flexible output. Simulation results involving several case studies show that the suggested control scheme could maintain the controlled variables (including the variable electrical load and RO feed pressure) within desired limits under various plant operating conditions. The results also indicate that the proposed RO plant could provide operational flexibility to participate in energy management at the utility scale by dynamically optimizing the use of excess plant capacity within an N-R HES. For a small-scale energy storage system, a sensible Thermal Energy Storage (TES) model has been developed in the Modelica Framework in FY 2018.

25 ENERGY STORAGE↗
DOE OSTI2015

MONK10: Burnup credit capability - 14596

MONK{sup R} is a Monte Carlo code for nuclear criticality and reactor physics analyses. It has a proven track record of application to the whole of the nuclear fuel cycle and is well established in the UK criticality community. Furthermore it is increasingly being used for reactor physics analysis (as described at ICNC 2011), which makes it an ideal tool for burn-up credit (BUC) calculations. Throughout the paper, example calculations based on a PWR are presented to illustrate the capabilities of the MONK10 code. In order to account for the spatial dependence of material burn-up it has in the past been necessary to design models with multiple regions and materials specifically to allow material burn-up to vary spatially. This is very labour intensive and difficult to change at a later stage. A new code version, MONK10, was released last year which includes the facility to allow a burn-up (BU) mesh to be superimposed on an existing model in order to account for the spatial dependence of the burn-up. This facility is used to consider the effect of radial position of a fuel element in a PWR core on BUC. Additionally, a thermal hydraulics (TH) mesh can be used to specify region dependent temperature. This, coupled with the fact that MONK10 also incorporates an on-the-fly Doppler broadening methodology facilitates the modelling of spatially dependent temperatures for the different components. A TH mesh is used to superimpose a temperature profile on a PWR based model and the effect of this on BUC is considered. The burn-up modelling in MONK has been benchmarked against the ANSWERS WIMS deterministic reactor physics code. Once the burn-up calculation has been completed and the depleted fuel compositions determined the spent fuel compositions can be transferred into a model of a storage facility or transport flask in order to perform burn-up credit analysis. The initial model is usually described as the donor model and the latter model as the receiver model. This transfer is carried out using the COWL option which allows the specification of a material in the receiver model based on the material's composition in a given BU mesh cell from the donor model. This allows compositions and densities to be transferred and also allows user specified adjustments to be made. For example, this could include omitting the fission products in order to estimate their contribution to burn-up credit and provide an actinide-only analysis. The effect of excluding appropriate nuclides is presented. An example of how the ANSWERS SPRUCE code can be used to quantify uncertainty in a BUC calculation is also presented. (authors)

42 ENGINEERING↗
NASA NTRS1999

A Performance Assessment of Eight Low-Boom High-Speed Civil Transport Concepts

A performance assessment of eight low-boom high speed civil transport (HSCT) configurations and a reference HSCT configuration has been performed. Although each of the configurations was designed with different engine concepts, for consistency, a year 2005 technology, 0.4 bypass ratio mixed-flow turbofan (MFTF) engine was used for all of the performance assessments. Therefore, all original configuration nacelles were replaced by a year 2005 MFRF nacelle design which corresponds to the engine deck utilized. The engine thrust level was optimized to minimize vehicle takeoff gross weight. To preserve the configuration's sonic-boom shaping, wing area was not optimized or altered from its original design value. Performance sizings were completed when possible for takeoff balanced field lengths of 11,000 ft and 12,000 ft, not considering FAR Part 36 Stage III noise compliance. Additionally, an arbitrary sizing with thrust-to-weight ratio equal to 0.25 was performed, enabling performance levels to be compared independent of takeoff characteristics. The low-boom configurations analyzed included designs from the Boeing Commercial Airplane Group, Douglas Aircraft Company, Ames Research Center, and Langley Research Center. This paper discusses the technology level assumptions, mission profile, analysis methodologies, and the results of the assessment. The results include maximum lift-to-drag ratios, total fuel consumption, number of passengers, optimum engine sizing plots, takeoff performance, mission block time, and takeoff gross weight for all configurations. Results from the low-boom configurations are also compared with a non-low-boom reference configuration. Configuration dependent advantages or deficiencies are discussed as warranted.

Baize, Daniel G.↗
DOE OSTI2015

Reaction-space analysis of homogeneous charge compression ignition combustion with varying levels of fuel stratification under positive and negative valve overlap conditions

Full-cycle computational fluid dynamics simulations with gasoline chemical kinetics were performed to determine the impact of breathing and fuel injection strategies on thermal and compositional stratification, combustion and emissions during homogeneous charge compression ignition combustion. The simulations examined positive valve overlap and negative valve overlap strategies, along with fueling by port fuel injection and direct injection. The resulting charge mass distributions were analyzed prior to ignition using ignition delay as a reactivity metric. The reactivity stratification arising from differences in the distributions of fuel–oxygen equivalence ratio (Φ FO ), oxygen molar fraction (χ O2 ) and temperature (T) was determined for three parametric studies. In the first study, the reactivity stratification and burn duration for positive valve overlap valve events with port fuel injection and early direct injection were nearly identical and were dominated by wall-driven thermal stratification. nitrogen oxide (NO) and carbon monoxide (CO) emissions were negligible for both injection strategies. In the second study, which examined negative valve overlap valve events with direct injection and port fuel injection, reactivity stratification increased for direct injection as the Φ FO and T distributions associated with direct fuel injection into the hot residual gas were positively correlated; however, the latent heat absorbed from the hot residual gas by the evaporating direct injection fuel jet reduced the overall thermal and reactivity stratification. These stratification effects were offsetting, resulting in similar reactivity stratification and burn durations for the two injection strategies. The higher local burned gas temperatures with direct injection resulted in an order of magnitude increase in NO, while incomplete combustion of locally over-lean regions led to a sevenfold increase in CO emissions compared to port fuel injection. The final study evaluated positive valve overlap and negative valve overlap valve events with direct injection. Furthermore, relative to positive valve overlap, the negative valve overlap condition had a wider reactivity stratification, a longer burn duration and higher NO and CO emissions associated with reduced fuel–air mixing.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2016

Effects of calcium in ash on the corrosion performance of Ni-based alloys in simulated oxy-fuel environment

Increasing the efficiency of coal power plants requires raising the operating temperature above 650°C. However, coal ash can severely attack alloy materials at high temperature. For example, the corrosion rates of commercial Fe- and Ni-based alloys are generally greater than 2 mm/year at 750°C in the gas environment of oxy-fuel combustion. Thus, a critical study is needed to determine the effect of the constituents in the ash on corrosion and find an approach to reduce the corrosion rates in an ash-laden environment at high temperature. The role of CaO in the ash (typical of U.S. Western coal ash) has been investigated in laboratory exposure environments with various structural alloys. Detailed results are presented on weight change, scale thickness, internal penetration, microstructural characteristics of corrosion products, and the cracking of scales for the alloys after exposure at 750°C. The thermal stability of K3Al(SO4)3 under the environment of oxy-fuel combustion was determined by thermogravimetric analysis and differential thermal analysis. The reaction of this low melting temperature salt with the CaO-containing ash is discussed. In addition, we performed synchrotron nanobeam X-ray analysis to study the phase and chemical composition of the oxide layers on the alloy surface. Results from these studies are used to address the role of CaO in ash in the long-term corrosion performance of alloys.

36 MATERIALS SCIENCE↗
NASA NTRS2015

Subsonic Ultra Green Aircraft Research: Truss Braced Wing Design Exploration - Phase II - Volume I

This report summarizes the Truss Braced Wing (TBW) work accomplished by the Boeing Subsonic Ultra Green Aircraft Research (SUGAR) team, consisting of Boeing Research and Technology, Boeing Commercial Airplanes, General Electric, Georgia Tech, Virginia Tech, NextGen Aeronautics, and Microcraft. A multi-disciplinary optimization (MDO) environment defined the geometry that was further refined for the updated SUGAR High TBW configuration. Airfoil shapes were tested in the NASA TCT facility, and an aeroelastic model was tested in the NASA TDT facility. Flutter suppression was successfully demonstrated using control laws derived from test system ID data and analysis models. Aeroelastic impacts for the TBW design are manageable and smaller than assumed in Phase I. Flutter analysis of TBW designs need to include pre-load and large displacement non-linear effects to obtain a reasonable match to test data. With the updated performance and sizing, fuel burn and energy use is reduced by 54% compared to the SUGAR Free current technology Baseline (Goal 60%). Use of the unducted fan version of the engine reduces fuel burn and energy by 56% compared to the Baseline. Technology development roadmaps were updated, and an airport compatibility analysis established feasibility of a folding wing aircraft at existing airports.

Bradley, Marty K.↗
NASA NTRS1983

Develop and test fuel cell powered on-site integrated total energy system

Test results are presented for a 24 cell, two sq ft (4kW) stack. This stack is a precursor to a 25kW stack that is a key milestone. Results are discussed in terms of cell performance, electrolyte management, thermal management, and reactant gas manifolding. The results obtained in preliminary testing of a 50kW methanol processing subsystem are discussed. Subcontracting activities involving application analysis for fuel cell on site integrated energy systems are updated.

Kaufman, A.↗
NASA NTRS2020

CFD Predictions of Soot & CO Emissions Generated by a Partially-Fueled 9-Element Lean-Direct Injection Combustor

A study was undertaken to investigate the CO & soot emissions generated by a partially-fueled 9- element LDI (Lean-Direct Injection) combustor configuration operating in the idle range of jet engine conditions. In order to perform the CFD analysis, several existing soot/chemistry models were implemented into the OpenNCC (Open National Combustion Code). The calculations were based on a Reynolds-Averaged Navier Stokes (RANS) simulation with standard k-epsilon turbulence model, a 62- species jet-a/air chemistry, a 2-equation soot model, & a Lagrangian spray solver. A separate transport equation was solved for all individual species involved in jet-a/air combustion. In the test LDI configuration we examined, only five of the nine injectors were fueled with the major pilot injector operating at an equivalence ratio of near one and the other four main injectors operating at an equivalence ratio near 0.55. The calculations helped to identify several reasons behind the soot & CO formation in different regions of the combustor. The predicted results were compared with the reported experimental data on soot mass concentration (SMC) & emissions index of CO (EICO). The experimental results showed that an increase in either T3 and/or F/A ratio lead to a reduction in both EICO & SMC. The predicted results were found to be in reasonable agreement. However, the predicted EICO differed substantially in one test condition associated with higher F/A ratio.

Raju, M. S.↗
NASA NTRS2020

Analysis of a 100-W Regenerative Fuel Cell Demonstration

A 100-W regenerative fuel cell (RFC) demonstration was successfully completed at the NASA Glenn Research Center on December 11, 2014. This effort was funded by Advanced Space Power Systems milestone and represented the first demonstration of passive laboratory-scale stacks arranged as an RFC, consisting of a non-flow-through fuel cell and a static-feed electrolyzer. Once developed, these stacks present the potential to minimize system mass, volume, and parasitic power in comparison to existing terrestrial electrochemical systems and nonnuclear power generation options. Testing involved the fuel cell powering a simulated load then recycling fuel cell product water by subsequently electrolyzing it to regenerate hydrogen and oxygen gases for fuel cell operations. A total of five cycles were completed over a 2-week period, yielding cell voltage data for stack performance and quality measures for regenerated reactants. Overall, round-trip efficiency was calculated to be 47.4 percent. Gas crossover was observed in electrolyzer gas products, as residual gas analysis showed 1.1 percent hydrogen in oxygen and 0.4 percent oxygen in hydrogen. Fuel cell product water was determined to be more acidic and conductive and have a higher fluoride content when compared to deionized water.

fuel cell↗
DOE OSTI2024

Scale effects on core design, fuel costs, and spent fuel volume of pressurized water reactors

The desire to improve the economic competitiveness and deployment pace of nuclear energy through modularization, manufacturing, and series production had led to the development of smaller size reactors. As the standard 17x17 fuel technology is mainly maintained in the pressurized water reactors (PWRs) category, this translates into a lower number of fuel assemblies in the core and sometimes a reduced fuel height. To assess the impact of such scale change in core design on fuel cycle cost and spent fuel volume, a scoping analysis tool is developed based on infinite lattice calculations, leakage, fuel management reduced models, and levelized unit cost of electricity (LCOE) estimate. As such, cost dynamics driven by fuel specific power, burnup, core leakage, feed, cycle length, fuel assembly height as well as uranium market data are captured with consistent set of assumptions and analysis methods. A selection of 5 reactor designs representative of leading PWR developers is assessed and compared. Pursuing higher specific powers and optimal burnups are highlighted as the main fuel cost reduction drivers, nevertheless, practical limitations and opportunities must be evaluated to establish the feasibility of such enhanced fuel operation. In consequence, a detailed core design is performed using SIMULATE3 code for 5 PWR variations including natural and forced coolant circulation modes, two reactor scales, power densities of 73, 112, and 123 kW/l and higher discharge burnups. Design and optimization are performed at the lattice level, for the reflector, and at the core loading level. Satisfactory steady-state operation including power distribution, coolant operating limits, and reactivity requirements are analyzed and reported in this paper. The fuel economics of the detailed core designs confirm the scoping analysis findings. Despite the unlocked power uprates in small PWRs, the achievable burnup for a given fuel specific power requires more enrichment and shorter fuel height results in higher fabrication costs per mass of fuel, which makes scaling down core size a more expensive endeavor on the fuel cycle front. Spent fuel volumes are reported for the PWRs designed in this paper. Furthermore, these volumes are driven by the core average discharge burnup regardless of the scale in consideration. Additional cost and core performance aspects related to heavy reflector gains, fuel-reflector substitution, and disposal cost policy in the U.S. are examined.

42 ENGINEERING↗
DOE OSTI2019

MELCOR Modeling of Accident Scenarios at a Facility for Aqueous Reprocessing of Spent Nuclear Fuel

The work presented in this report applies the MELCOR code to evaluate potential accidents in non-reactor nuclear facilities, focusing on Design Basis Accidents. Ten accident scenarios were modeled using NRC's best-estimate severe accident analysis code, MELCOR 2.2. The accident scenarios simulated a range of explosions and/or fires related to a nuclear fuel reprocessing facility. The objective was to evaluate the radionuclide source term to the environment following initiating explosion and/or fire events. The simulations were performed using a MELCOR model of the Barnwell Nuclear Fuel Plant, which was decommissioned before beginning reprocessing operations. Five of the accident scenarios were based on the Class 5 Design Basis Accidents from the Final Safety Analysis Report. Three of the remaining accident scenarios include sensitivity studies on smaller solvent fires. The final two accidents included an induced fire from an initial explosion. The radionuclide inventory was developed from ORIGEN calculations of spent PWR fuel with an initial enrichment of 4.5% U-235 by weight. The fuel aged for five years after a final 500-day irradiation cycle. The burn-up was conservatively increased to 60 GWd/MTU to bound current US operations. The results are characterized in terms of activity release to the environment and the building decontamination factor, which is related the leak path factor used in Department of Energy safety analyses. The MELCOR 2.2 results consider adverse consequences to the filters, ventilation system, and structures as a result of the explosions and fires. The calculations also include best-estimate models for aerosol transport, agglomeration, and deposition. The new calculations illustrate best-estimate approaches for predicting the source term from a reprocessing facility accident.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗