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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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All sourcesNASA NTRSDOE OSTI
NASA NTRS2007

High Fidelity, Fuel-Like Thermal Simulators for Non-Nuclear Testing: Analysis and Initial Test Results

Non-nuclear testing can be a valuable tool in the development of a space nuclear power system, providing system characterization data and allowing one to work through various fabrication, assembly and integration issues without the cost and time associated with a full ground nuclear test. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Testing with non-optimized heater elements allows one to assess thermal, heat transfer. and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. High fidelity thermal simulators that match both the static and the dynamic fuel pin performance that would be observed in an operating, fueled nuclear reactor can vastly increase the value of non-nuclear test results. With optimized simulators, the integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics and assess potential design improvements at relatively small fiscal investment. Initial conceptual thermal simulator designs are determined by simple one-dimensional analysis at a single axial location and at steady state conditions; feasible concepts are then input into a detailed three-dimensional model for comparison to expected fuel pin performance. Static and dynamic fuel pin performance for a proposed reactor design is determined using SINDA/FLUINT thermal analysis software, and comparison is made between the expected nuclear performance and the performance of conceptual thermal simulator designs. Through a series of iterative analyses, a conceptual high fidelity design is developed: this is followed by engineering design, fabrication, and testing to validate the overall design process. Test results presented in this paper correspond to a "first cut" simulator design for a potential liquid metal (NaK) cooled reactor design that could be applied for Lunar surface power. Proposed refinements to this simulator design are also presented.

Bragg-Sitton, Shannon M.↗
DOE OSTI2016

Statistical and sensitivity analysis of failing rods in EPR LBLOCA

A statistical fuel failure analysis procedure has been developed, and is now applied in full scale to a large break loss-of-coolant accident (LBLOCA) scenario in EPR-type nuclear power plant. The goal is to statistically determine the number of failing fuel rods. It is also important to bring out the underlying causes for rod failures, and therefore a sensitivity analysis procedure for the preceding complex calculation chain is outlined and adopted. In order to produce a statistically reliable estimation on the number of failing rods, a large number of simulations are required. In the analysis, single rods are simulated with the coupled fuel performance - thermal hydraulics code FRAPTRAN-GENFLO. The statistically varied factors are divided into global and local by their range of influence. Transient thermal hydraulic and power history boundary conditions and model parameters are global and affect all the rods. Fuel rod manufacturing parameters and other rod position related factors are local and are related to a certain rod. The number of global scenarios is 59 as dictated by the Wilks' formula, and 1000 randomly sampled rods are simulated in each scenario. The thermal hydraulic and transient power boundary conditions for the fuel performance code are calculated with a statistical version of the system code APROS. The steady-state irradiation histories are simulated with FRAPCON, and the steady-state power histories used as boundary conditions for FRAPCON are obtained from SIMULATE calculations. Thus, a multistage calculation chain is required to consolidate the procedure. As an outcome, in the worst global scenario, 1.4% of the simulated rods failed. It can be concluded that the Finnish safety regulations, i.e. max. 10% of the rods allowed to fail, are met. In order to find out which input parameters have significance to the outcome of the analysis, a sensitivity analysis is done. At the moment, only the worst global scenario is considered but later on it is possible to include also the sensitivities of the global factors. Due to complexity of the existing data, first the relevant input parameters for the sensitivity analysis have to be specified. Data visualization with a cobweb graph is used for the screening. Then, selected sensitivity measures are calculated between the chosen input and output parameters. The sensitivity indices calculated are the Borgonovo's delta measure, the first order Sobol sensitivity index, and squared Pearson correlation coefficients. The first mentioned is a novelty in this context. As an outcome, the most relevant parameters with respect to the cladding integrity were determined to be the decay heat power during the transient, the thermal hydraulic conditions in the rod's location in the reactor, and the steady-state irradiation history of the rod as represented in this analysis by the rod burnup. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Transitional cores and fuel cycle analyses in support of MIT reactor low enriched uranium fuel conversion

The Massachusetts Institute of Technology Reactor (MITR) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. At 6 MW, it delivers neutron flux and energy spectrum comparable to power light water reactors (LWRs) in a compact core using highly enriched uranium (HEU) fuel. In the framework of non-proliferation policy, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel based on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow conversion of U.S. high performance research reactors (USHPRRs) like the MITR. The Preliminary Safety Analysis Report (PSAR) for the MITR LEU conversion has been submitted and currently under review by the U.S. Nuclear Regulatory Commission (NRC). A transition core plan, from 22 fresh LEU fuel elements (i.e., beginning-of-life) gradually to 24 of them arranged in an equilibrium configuration, is expected to serve as an appendix chapter in the final SAR. The current study reported here presents the fuel cycle development, which eventually leads to the transition core plan. The results confirm the equilibrium state, where both the shim bank movement (i.e., core reactivity) as well as the content of fissile materials stabilize, can be achieved by fixed-pattern fuel management. Fission density has been evaluated for a number of fully discharged LEU fuel elements, using a best-estimate approach. The fuel cycle calculations also generate power profiles at each core state. A steady-state thermal-hydraulic safety analysis has thus been performed, where onset of nucleate boiling (ONB) is considered as the safety criterion. The results confirm significant margin to ONB at all analyzed transition and equilibrium fuel cycle states.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Sensitivity and Uncertainty Quantification of Transition Scenario Simulations

This report documents the first collective attempt at developing and applying capabilities to quantify uncertainties, assess parametric sensitivities, and optimize multiple parameters and metrics in fuel cycle simulations generated by the SA&I Campaign. To do this, external codes that were designed to perform sensitivity analysis and uncertainty quantification (SA&UQ) needed to be coupled to the SA&I Campaign’s nuclear fuel cycle simulators (NFCS). In FY20, two approaches were pursued: 1) coupling Cyclus to an ORNL-internal code called MOT (Metaheuristic Optimization Tool) and 2) coupling DYMOND to the opensource SA&UQ tool kit Dakota. The primary objective of having these NFCS/SA&UQ coupled capabilities is to better inform DOE-NE and other stakeholders on the results generated from the NFCS. For a given set of fuel cycle strategies, policies, and technology assumptions that make up a fuel cycle scenario, these NFCS have traditionally been used by the SA&I Campaign to provide quantitative answers in terms of year-by-year mass flows, infrastructure requirements, costs, etc. With these newly developed coupled capabilities, the SA&I Campaign can now efficiently simulate hundreds or thousands of these scenarios, sample large ranges of parameters and assumptions, and use the unique features of the SA&UQ tools to process the data. This enables providing answers with known and propagated uncertainties, determining the sensitivity of important metrics to different parameters and assumptions, quantifying how much fuel cycle and technology parameters impact each other, and producing optimized fuel cycle strategies for single and multiple variables. To demonstrate these new capabilities, the Cyclus/MOT was used to model several scenarios ranging from simple fleet retirements to transitions to advanced reactors. Specifically, for a transition scenario from LWRs to SFRs and advanced LWRs, uncertainty quantification, sensitivity analysis, and optimization studies were applied to cases involving single and multiple parameter (input) and single and multiple metric (output) variations. In addition, a similar transition scenario was modeled to demonstrate how to optimize the reprocessing capacity parameter to minimize two performance metrics while taking into account uncertainties from two other parameters. Lastly, a depletion module based on SCALE/ORIGEN was added in Cyclus to simulate the third scenario that was designed to quantify the impact of the modeling assumption that all LWR used nuclear fuel have the same burnup. The newly developed DYMOND/Dakota capability was also applied to a transition scenario from the existing fleet to small modular reactors and fast reactors. This particular scenario involves not only explicit isotopic depletion via ORIGEN-2, but also includes multirecycling and utilizing the criticality search feature to determine the fresh fuel composition of recycled fuel, a feature unique to the DYMOND NFCS. A large database of simulations were run with 4 main parameters that were sampled: start date of reprocessing, reprocessing capacity, energy demand growth rate, and advanced reactor share of the fleet. The 4 main metrics were uranium consumption, enrichment requirements, waste generation, and levelized cost of electricity using data from the Cost Basis Report. The demonstrated SA&UQ results include those that inform on how to choose parameters to avoid “failed” scenarios, Sobol’ indices that inform on the importance of various parameters individually and synergistically, and Analysis of Variance (ANOVA) studies that decompose parameter ranges into groups and informs on whether variations are statistically significant.

Feng, B.↗
DOE OSTI2016

Gamma emission tomography measurements of fuel assemblies at the Halden reactor

Gamma emission tomography measurements have been performed at the Halden Reactor Project using a dedicated instrument, which utilizes a high-resolution gamma-ray detector allowing for spectroscopic analysis of the measurement data. Measurements have been performed on two fuel assemblies consisting of 9 and 13 rods, respectively, in order to characterize the rod-wise radioactive source distribution (i.e. fission and activation products) within the fuel. The 9-rod fuel assembly consisted of five fuel rods at 50 MWd/kgUO{sub 2} and four fuel rods at 26 MWd/kgUO{sub 2}, where all rods had cooled for approximately 22 years at the time of measurement. The rods in the 13-rod assembly all had a burnup of ∼4.5 MWd/kgUO{sub 2}, and cooled for approximately 1.5 years at the time of measurement. In this paper, the tomographic data was reconstructed using the Filtered Backprojection technique, where no consideration to gamma-ray attenuation in the fuel was given. Due to the varying burnup and cooling times between the assemblies, the spectroscopic data also varied between the respective sets of measurements. The high-resolution detector used in the measurements allowed for tomographic reconstruction of many gamma-ray peaks corresponding to various fission products and activation products present in the fuel and structural materials. For instance the gamma tomography technique has shown rod-wise variation in the contents of fission products between the high-burnup and low-burnup rods. The rod-internal distribution of fission products was also observed where the volatile fission product {sup 137}Cs is observed to have migrated to the outer pellet surface in fuel rods in the 9-rod assembly, and where the gaseous fission product {sup 85}Kr is observed to have a uniform distribution in each fuel rod's gas plenum.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2006

Sodium Borohydride/Hydrogen Peroxide Fuel Cells For Space Application

This viewgraph presentation examines Sodium Borohydride and Hydrogen Peroxide Fuel Cells as they are applied to space applications. The topics include: 1) Motivation; 2) The Sodium Borohydride Fuel Cell; 3) Sodium Borohydride Fuel Cell Test Stands; 4) Fuel Cell Comparisons; 5) MEA Performance; 6) Anode Polarization; and 7) Electrode Analysis. The benefits of hydrogen peroxide as an oxidant and benefits of sodium borohydride as a fuel are also addressed.

fuel cells↗
DOE OSTI2016

Sensitivity Analysis of an Experimental Breeder Reactor II Fuel Assembly

The Experimental Breeder Reactor II (EBR-II) was selected as a candidate for a reactor physics benchmark evaluation due to its unique features and capabilities. In connection with the benchmark evaluation, the goal of this work was to perform a sensitivity analysis. The EBR-II was a sodium cooled, pool type, metal fueled, fast spectrum reactor. EBR-II operated from 1964 through 1994. It had a maximum heat output of 62.5 MW which could produce 20 MW of electricity. The design of EBR-II consisted of 637 hexagonal shaped, removable assemblies. The assemblies were divided into three regions: the core, an inner blanket, and an outer blanket. The core region contained assemblies with 91 fuel rods in each. The fuel was enriched uranium metal and was clad within stainless steel. The core region also included two safety assemblies and eight control assemblies with the difference being that these assemblies only contained 61 fuel rods. Additional assemblies in the core region included stainless steel dummies, half worth drivers, and experimental/instrumentation assemblies. The inner blanket region initially consisted of depleted uranium for fuel breeding, but was replaced, after proving the concept of breeding, with stainless steel reflectors for the purpose of irradiation. The outer blanket region consisted of depleted uranium. Figure 1 shows a cross section view of the EBR-II core layout. Each assembly was approximately 234 cm long with the uranium fuel section being 33 cm long. Above and below the fuel were stainless steel neutron reflectors. Additionally, orifices at the bottom of the assembly allowed for sodium coolant to flow upward, with larger holes in the subassemblies at the center where the greatest heat was produced. The core of EBR-II was enclosed by a stainless steel reactor vessel which was comprised of a radial shell, a lower grid plenum structure, and an upper reactor vessel cover that served as a neutron shield. Surrounding the reactor vessel radially were layers of graphite and borated graphite shielding. The entire assembly is submerged 3 meters in a pool of sodium. The pool contained up to 337,000 liters of sodium which was forced through the core and blanket regions using two centrifugal pumps that were also submerged within the sodium pool. The design of EBR-II included many safety features. The compatibility between the sodium and stainless steel prevented corrosion and therefore the risk of a radioactive release due to corrosion. The sodium coolant was kept at near atmospheric pressure which made any leak more controllable. The metal fuel, cladding, and coolant allowed for heat to be removed easily. The large volume of sodium allowed for heat to be absorbed without problems occurring. Lastly, the reactor would expand as temperature increased which would result in greater neutron leakage. This extra leakage caused a large negative temperature reactivity feedback which automatically shut the reactor down in case of an accident. EBR-II had unique capabilities which were proven during the Integral Fast Reactor initiative. This program lasted from 1984 to 1994. The goal of this program was to solve many of the perceived problems with nuclear energy through scientific means. The safety features of EBR-II were put to the test in April 1986. The tests simulated loss of reactor coolant flow and loss of a heat sink. For both of these tests, the reactor was at 100% power and was not scrammed. During both tests, the reactor shut itself down and achieved a safe steady state temperature all due to the design of the reactor. It was because of these tests that EBR-II was chosen for a reactor physics benchmark evaluation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Approach to proliferation risk assessment based on multiple objective analysis framework

The approach to the assessment of proliferation risk using the methods of multi-criteria decision making and multi-objective optimization is presented. The approach allows the taking into account of the specifics features of the national nuclear infrastructure, and possible proliferation strategies (motivations, intentions, and capabilities). 3 examples of applying the approach are shown. First, the approach has been used to evaluate the attractiveness of HEU (high enriched uranium)production scenarios at a clandestine enrichment facility using centrifuge enrichment technology. Secondly, the approach has been applied to assess the attractiveness of scenarios for undeclared production of plutonium or HEU by theft of materials circulating in nuclear fuel cycle facilities and thermal reactors. Thirdly, the approach has been used to perform a comparative analysis of the structures of developing nuclear power systems based on different types of nuclear fuel cycles, the analysis being based on indicators of proliferation risk.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2026

Cost Analysis of Heavy-Duty Vehicle Proton Exchange Membrane Fuel Cell Stationary Power Plants

Heavy-duty PEM fuel cells could be a low-cost, low emission alternative to combustion turbines for re-electrifying hydrogen if used as part of a long duration grid energy storage systems. Many studies expect heavy duty PEM fuel cell production costs to reduce as manufacturing volumes ramp up and their expected durability of 25,000-30,000 hours aligns well with a 30-year life for hydrogen seasonal energy storage plants that would likely operate less than 10% of the year. The labor, material, and equipment costs associated with installing PEM fuel cells and their required balance of plant for stationary applications have not been thoroughly explored, however. This study performs a detailed design and cost analysis of a 100 MW stationary PEM fuel cell power plant, capturing costs such as cooling, power electronics, pipes, valves, fittings, cabling, conduit, concrete foundations, buildings, and land. It employs methods consistent with NLR's solar PV benchmarking cost analysis and annual technology baseline to derive the total installed costs of stationary PEM fuel cell plants that utilize heavy duty PEM fuel cells.

08 HYDROGEN↗
DOE OSTI2016

Complete Sensitivity/Uncertainty Analysis of LR-0 Reactor Experiments with MSRE FLiBe Salt and Perform Comparison with Molten Salt Cooled and Molten Salt Fueled Reactor Models

In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7 LiF + BeF 2 ) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in 7 Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2016

NDARC NASA Design and Analysis of Rotorcraft - Input, Appendix 4

The NDARC code performs design and analysis tasks. The design task involves sizing the rotorcraft to satisfy specified design conditions and missions. The analysis tasks can include off-design mission performance analysis, flight performance calculation for point operating conditions, and generation of subsystem or component performance maps. The principal tasks (sizing, mission analysis, flight performance analysis) are shown in the figure as boxes with heavy borders. Heavy arrows show control of subordinate tasks. The aircraft description consists of all the information, input and derived, that denes the aircraft. The aircraft consists of a set of components, including fuselage, rotors, wings, tails, and propulsion. This information can be the result of the sizing task; can come entirely from input, for a fixed model; or can come from the sizing task in a previous case or previous job. The aircraft description information is available to all tasks and all solutions. The sizing task determines the dimensions, power, and weight of a rotorcraft that can perform a specified set of design conditions and missions. The aircraft size is characterized by parameters such as design gross weight, weight empty, rotor radius, and engine power available. The relations between dimensions, power, and weight generally require an iterative solution. From the design flight conditions and missions, the task can determine the total engine power or the rotor radius (or both power and radius can be fixed), as well as the design gross weight, maximum takeoff weight, drive system torque limit, and fuel tank capacity. For each propulsion group, the engine power or the rotor radius can be sized. Missions are defined for the sizing task, and for the mission performance analysis. A mission consists of a number of mission segments, for which time, distance, and fuel burn are evaluated. For the sizing task, certain missions are designated to be used for design gross weight calculations; for transmission sizing; and for fuel tank sizing. The mission parameters include mission takeoff gross weight and useful load. For specified takeoff fuel weight with adjustable segments, the mission time or distance is adjusted so the fuel required for the mission equals the takeoff fuel weight. The mission iteration is on fuel weight or energy. Flight conditions are specified for the sizing task, and for the flight performance analysis. For the sizing task, certain flight conditions are designated to be used for design gross weight calculations; for transmission sizing; for maximum takeoff weight calculations; and for anti-torque or auxiliary thrust rotor sizing. The flight condition parameters include gross weight and useful load. For flight conditions and mission takeoff, the gross weight can be maximized, such that the power required equals the power available. A flight state is defined for each mission segment and each flight condition. The aircraft performance can be analyzed for the specified state, or a maximum effort performance can be identified. The maximum effort is specified in terms of a quantity such as best endurance or best range, and a variable such as speed, rate of climb, or altitude.

Rotorcraft Input↗
NASA NTRS2022

NDARC - NASA Design and Analysis of Rotorcraft: Input and Data Structures - Vol 2

The NDARC code performs design and analysis tasks. The design task involves sizing the rotorcraft to satisfy specified design conditions and missions. The analysis tasks can include off-design mission performance analysis, flight performance calculation for point operating conditions, and generation of subsystem or component performance maps. Figure 1-1 illustrates the tasks. The principal tasks (sizing, mission analysis, flight performance analysis) are shown in the figure as boxes with heavy borders. Heavy arrows show control of subordinate tasks. The aircraft description (figure 1-1) consists of all the information, input and derived, that defines the aircraft. The aircraft consists of a set of components, including fuselage, rotors, wings, tails, and propulsion. This information can be the result of the sizing task; can come entirely from input, for a fixed model; or can come from the sizing task in a previous case or previous job. The aircraft description information is available to all tasks and all solutions (indicated by light arrows). The sizing task determines the dimensions, power, and weight of a rotorcraft that can perform a specified set of design conditions and missions. The aircraft size is characterized by parameters such as design gross weight, weight empty, rotor radius, and engine power available. The relations between dimensions, power, and weight generally require an iterative solution. From the design flight conditions and missions, the task can determine the total engine power or the rotor radius (or both power and radius can be fixed), as well as the design gross weight, maximum takeoff weight, drive system torque limit, and fuel tank capacity. For each propulsion group, the engine power or the rotor radius can be sized. Missions are defined for the sizing task, and for the mission performance analysis. A mission consists of a number of mission segments, for which time, distance, and fuel burn are evaluated. For the sizing task, certain missions are designated to be used for design gross weight calculations; for transmission sizing; and for fuel tank sizing. The mission parameters include mission takeoff gross weight and useful load. For specified takeoff fuel weight with adjustable segments, the mission time or distance is adjusted so the fuel required for the mission (burned plus reserve) equals the takeoff fuel weight. The mission iteration is on fuel weight or energy. Flight conditions are specified for the sizing task, and for the flight performance analysis. For the sizing task, certain flight conditions are designated to be used for design gross weight calculations; for transmission sizing; for maximum takeoff weight calculations; and for antitorque or auxiliary thrust rotor sizing. The flight condition parameters include gross weight and useful load. For flight conditions and mission takeoff, the gross weight can be maximized, such that the power required equals the power available. A flight state is defined for each mission segment and each flight condition. The aircraft performance can be analyzed for the specified state, or a maximum effort performance can be identified. The maximum effort is specified in terms of a quantity such as best endurance or best range, and a variable such as speed, rate of climb, or altitude. The aircraft must be trimmed, by solving for the controls and motion that produce equilibrium in the specified flight state. Different trim solution definitions are required for various flight states. Evaluating the rotor hub forces may require solution of the blade flap equations of motion.

NDARC↗
NASA NTRS2022

NDARC - NASA Design and Analysis of Rotorcraft: Input - Vol 3

The NDARC code performs design and analysis tasks. The design task involves sizing the rotorcraft to satisfy specified design conditions and missions. The analysis tasks can include off-design mission performance analysis, flight performance calculation for point operating conditions, and generation of subsystem or component performance maps. Figure 1-1 illustrates the tasks. The principal tasks (sizing, mission analysis, flight performance analysis) are shown in the figure as boxes with heavy borders. Heavy arrows show control of subordinate tasks. The aircraft description (figure 1-1) consists of all the information, input and derived, that defines the aircraft. The aircraft consists of a set of components, including fuselage, rotors, wings, tails, and propulsion. This information can be the result of the sizing task; can come entirely from input, for a fixed model; or can come from the sizing task in a previous case or previous job. The aircraft description information is available to all tasks and all solutions (indicated by light arrows). The sizing task determines the dimensions, power, and weight of a rotorcraft that can perform a specified set of design conditions and missions. The aircraft size is characterized by parameters such as design gross weight, weight empty, rotor radius, and engine power available. The relations between dimensions, power, and weight generally require an iterative solution. From the design flight conditions and missions, the task can determine the total engine power or the rotor radius (or both power and radius can be fixed), as well as the design gross weight, maximum takeoff weight, drive system torque limit, and fuel tank capacity. For each propulsion group, the engine power or the rotor radius can be sized. Missions are defined for the sizing task, and for the mission performance analysis. A mission consists of a number of mission segments, for which time, distance, and fuel burn are evaluated. For the sizing task, certain missions are designated to be used for design gross weight calculations; for transmission sizing; and for fuel tank sizing. The mission parameters include mission takeoff gross weight and useful load. For specified takeoff fuel weight with adjustable segments, the mission time or distance is adjusted so the fuel required for the mission (burned plus reserve) equals the takeoff fuel weight. The mission iteration is on fuel weight or energy. Flight conditions are specified for the sizing task, and for the flight performance analysis. For the sizing task, certain flight conditions are designated to be used for design gross weight calculations; for transmission sizing; for maximum takeoff weight calculations; and for antitorque or auxiliary thrust rotor sizing. The flight condition parameters include gross weight and useful load. For flight conditions and mission takeoff, the gross weight can be maximized, such that the power required equals the power available. A flight state is defined for each mission segment and each flight condition. The aircraft performance can be analyzed for the specified state, or a maximum effort performance can be identified. The maximum effort is specified in terms of a quantity such as best endurance or best range, and a variable such as speed, rate of climb, or altitude. The aircraft must be trimmed, by solving for the controls and motion that produce equilibrium in the specified flight state. Different trim solution definitions are required for various flight states. Evaluating the rotor hub forces may require solution of the blade flap equations of motion.

NDARC↗
NASA NTRS2023

NDARC - NASA Design and Analysis of Rotorcraft: Input - Appendix 3

The NDARC code performs design and analysis tasks. The design task involves sizing the rotorcraft to satisfy specified design conditions and missions. The analysis tasks can include off-design mission performance analysis, flight performance calculation for point operating conditions, and generation of subsystem or component performance maps. Figure 1-1 illustrates the tasks. The principal tasks (sizing, mission analysis, flight performance analysis) are shown in the figure as boxes with heavy borders. Heavy arrows show control of subordinate tasks. The aircraft description (figure 1-1) consists of all the information, input and derived, that defines the aircraft. The aircraft consists of a set of components, including fuselage, rotors, wings, tails, and propulsion. This information can be the result of the sizing task; can come entirely from input, for a fixed model; or can come from the sizing task in a previous case or previous job. The aircraft description information is available to all tasks and all solutions (indicated by light arrows). The sizing task determines the dimensions, power, and weight of a rotorcraft that can perform a specified set of design conditions and missions. The aircraft size is characterized by parameters such as design gross weight, weight empty, rotor radius, and engine power available. The relations between dimensions, power, and weight generally require an iterative solution. From the design flight conditions and missions, the task can determine the total engine power or the rotor radius (or both power and radius can be fixed), as well as the design gross weight, maximum takeoff weight, drive system torque limit, and fuel tank capacity. For each propulsion group, the engine power or the rotor radius can be sized. Missions are defined for the sizing task, and for the mission performance analysis. A mission consists of a number of mission segments, for which time, distance, and fuel burn are evaluated. For the sizing task, certain missions are designated to be used for design gross weight calculations; for transmission sizing; and for fuel tank sizing. The mission parameters include mission takeoff gross weight and useful load. For specified takeoff fuel weight with adjustable segments, the mission time or distance is adjusted so the fuel required for the mission (burned plus reserve) equals the takeoff fuel weight. The mission iteration is on fuel weight or energy. Flight conditions are specified for the sizing task, and for the flight performance analysis. For the sizing task, certain flight conditions are designated to be used for design gross weight calculations; for transmission sizing; for maximum takeoff weight calculations; and for antitorque or auxiliary thrust rotor sizing. The flight condition parameters include gross weight and useful load. For flight conditions and mission takeoff, the gross weight can be maximized, such that the power required equals the power available. A flight state is defined for each mission segment and each flight condition. The aircraft performance can be analyzed for the specified state, or a maximum effort performance can be identified. The maximum effort is specified in terms of a quantity such as best endurance or best range, and a variable such as speed, rate of climb, or altitude. The aircraft must be trimmed, by solving for the controls and motion that produce equilibrium in the specified flight state. Different trim solution definitions are required for various flight states. Evaluating the rotor hub forces may require solution of the blade flap equations of motion.

NDARC↗
DOE OSTI2015

Modeling and Simulation of Used Nuclear Fuel During Transportation with Consideration of Hydride Effects and Cyclic Fatigue

The objective of this work is to understand the integrity of Used Nuclear Fuel (UNF) during transportation. Previous analysis work has been performed to look at the integrity of UNF during transportation but these analyses have neglected to analyze the effect of hydrides and flaws (fracture mechanics models to capture radial cracking in the cladding). In this study, the clad regions of interest are near the pellet-pellet interfaces. These regions can experience more complex stress-states than the rest of the clad during cooling and have a greater possibility to develop radially reoriented hydrides during vacuum drying.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Diffusion behaviors between Fe and Pd-containing metallic fuel

Historical interest in metallic fuel, including U-Zr and U-Pu-Zr, is due to the high thermal conductivity, high burnup capability, high fissile atom density, and good compatibility with the coolant in sodium-cooled fast reactor conditions. The notable fuel application issue is the fuel-cladding chemical interaction (FCCI). The major cause of FCCI is the lanthanide fission products that migrate to the fuel periphery and deposit on the fuel cladding. Chemical interactions between the lanthanides and cladding will result in a cladding breach. Thus additives were sought that would immobilize the lanthanides by forming stable additive-lanthanide compounds. Palladium (Pd) is being explored as a potential additive. The microstructure of metallic fuels, including U-Zr and U-Pu-Zr, with Pd addition have been characterized. The post-irradiation examination on the U-13Zr-4Pd wt% alloy has found a pronounced effect of Pd on the fuel-cladding interface. The current study performed preliminary tests of diffusion couple and microstructure analysis, in order to investigate the chemical interactions between cladding and fuel with Pd addition.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2010

Seal Analysis for the Ares-I Upper Stage Fuel Tank Manhole Cover

Techniques for studying the performance of Naflex pressure-assisted seals in the Ares-I Upper Stage liquid hydrogen tank manhole cover seal joint are explored. To assess the feasibility of using the identical seal design for the Upper Stage as was used for the Space Shuttle External Tank manhole covers, a preliminary seal deflection analysis using the ABAQUS commercial finite element software is employed. The ABAQUS analyses are performed using three-dimensional symmetric wedge finite element models. This analysis technique is validated by first modeling a heritage External Tank liquid hydrogen tank manhole cover joint and correlating the results to heritage test data. Once the technique is validated, the Upper Stage configuration is modeled. The Upper Stage analyses are performed at 1.4 times the expected pressure to comply with the Constellation Program factor of safety requirement on joint separation. Results from the analyses performed with the External Tank and Upper Stage models demonstrate the effects of several modeling assumptions on the seal deflection. The analyses for Upper Stage show that the integrity of the seal is successfully maintained.

Phillips, Dawn R.↗
DOE OSTI2017

Shielding Analysis Capability of UNF-ST&DARDS

The Used Nuclear Fuel Storage, Transportation and Disposal Analysis Resource and Data System (UNF-ST&DARDS) is used to perform dose rate calculations for spent nuclear fuel (SNF) transportation packages based on the actual physical and nuclear characteristics (i.e., assembly design, burnup, initial enrichment, and cooling time) of the as-loaded SNF. Nuclear fuel data, transportation package model templates, and SNF canister loading map information residing within the tool facilitate automated generation of SCALE input files for radiation source term and dose rate calculations. Transportation package specific models developed for UNF-ST&DARDS dose rate analyses are described in detail. UNF-ST&DARDS dose rate analyses were performed in this paper for over 400 SNF canisters from 16 sites in their designated transportation casks. For simplicity, representative dose rate calculation results are presented as a function of time (i.e., selected calendar years between 2020 and 2100) for 73 SNF canisters in dry storage at four sites. For these canisters, the projected maximum dose rate values at 2 m from the lateral surfaces of the vehicle under normal conditions of transport (NCT) would vary between 1.9 and 11.5 mrem/h in 2020. Five SNF canisters will exceed the regulatory dose rate limit of 10 mrem/h at 2 m in 2020, and the analyzed SNF canisters will comply with regulatory dose rate limits by 2030. An analysis of the impact on the dose rate of fuel failure and reconfiguration during transportation indicated that the maximum dose rate for hypothetical accident conditions will be unaffected, and the NCT maximum dose rate at 2 m would have a maximum increase by a factor of 1.7 for a representative pressurized water reactor package and by a factor of 2.6 for a representative boiling water reactor package relative to intact fuel. Finally, analysis of the actual heat loading and the dose rate at 2 m from the lateral surface of the vehicle for the five SNF canisters exceeding the NCT regulatory dose rate limit of 10 mrem/h in 2020 showed that the dose rate could be more limiting with respect to regulatory requirements than the heat loading; i.e., the canister transportability date may be evaluated based on the transportation package’s external dose rate.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗