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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Effective Lanthanide Diffusivity through U-Zr Metallic Fuel

Fuel-cladding chemical interaction (FCCI) is one of the main concerns for the performance of U-Zr metallic fuels. Although FCCI results in complex phase transformations and morphological changes in both the fuel and cladding sides of the interface, the formation of a brittle layer in the cladding side of the interface (sometimes referred to as wastage) has one of the most significant impacts on the cladding mechanical integrity. This brittle layer is associated with the formation of intermetallic compounds between the cladding constituents Fe, Cr and the lanthanide fission product species. The formation of this cladding wastage layer has been the focus of recent efforts to develop a mechanistic modeling framework, including production of lanthanides in the fuel, transport through the fuel to the fuel-cladding interface, and phase transformation to form intermetallic phases. To calculate the rate of transport of lanthanides through the fuel, a multi-scale computational approach has been used. Atomistic calculations have been performed to determine the diffusivities of Nd, the most prevalent lanthanide species, through the solid fuel matrix and along pore surfaces. These diffusivities have been used as input parameters for a mesoscale model to determine an effective diffusion coefficient that accounts for the formation of porosity, its interconnection, and infiltration with bond sodium from the fuel-cladding gap. The effective diffusivity has been used in engineering-scale simulations using the BISON fuel performance code, and BISON simulations have been compared to representative validation cases from the EBR-II reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Full core LOCA analysis for BWR/6 - Methodology and first results

As part of collaboration between the Paul Scherrer Institute (PSI) and the Swiss regulator ENSI, a study aimed at core-wide estimates of the fuel rod behaviour during large-break loss-of-coolant-accidents (LB-LOCA) in BWR/6 plants was recently conducted on the basis of a Swiss reactor. To perform such full core LOCA analysis, a methodology consisting of a multi-physics sequence of interfaced best-estimate models and codes was established: CASMO-4E/SIMULATE-3 (C4/S3) for the core modelling, TRACE for the plant system analysis and FALCON for the fuel thermomechanical calculations during base irradiation (BI) as well as for the transient analyses (TA). As starting point, reference C4/S3 models validated for all operated cycles of the given plant were used to select as situation target, a recent cycle representative of modern BWR core loadings and fuel assembly designs. Based on the selected C4/S3 cycle model, PSI in-house computational modules were developed or updated in order to set up the TRACE full core model with regards to assembly geometries, fuel rods thermophysical properties and initial 3-D power distributions. Also, an interface to FALCON BI models was established for the transfer of the irradiation history, including primarily linear heat generation rates (LHGR), for an equivalent fuel pin. For each assembly, the power of the equivalent pin is defined with the maximum power of all pins at each axial position. However, in order to reduce the computational time of the fuel performance analyses, the whole core was subdivided into groups of assemblies, each represented by the most powerful assembly within the group. On this basis, the FALCON BI analyses were conducted with coupling to the PSI advanced model for gas release and swelling (GRWS-A) and an interface was developed to transfer the calculated fuel rod characteristics at end of irradiation to initialize the core fuel rods heat structures in the TRACE model. Subsequently, double-ended recirculation line rupture scenarios at hot-full-power (HFP) conditions were investigated with TRACE in order to determine a conservative transient case as function of cycle burnup, single failure criterion and availability of safety systems. For the most conservative case of this study, an interface was set-up to transfer the TRACE results as boundary conditions to the FALCON TA models, including thus primarily time and axial distributions of the clad outer temperatures. As final step, the FALCON TA were conducted for the representative assemblies in order to infer core-wide estimates of the fuel rod behaviour with specific focus on equivalent cladding reacted levels, hoop strains as well as margins to failure criteria. The objective of this paper is to present this methodology along with the obtained first results. Regarding the latter, it is found that principally, even for the most conservative LOCA scenario of this study, the clad temperature excursion remains mild enough at all core locations such that no significant rod ballooning and no cladding failure is predicted, something in line with corresponding analyses conducted earlier for BWR/4 plants. (authors)

42 ENGINEERING↗
DOE OSTI2024

A Bulk versus Nanoscale Hydrogen Storage Paradox Revealed by Material-System Co-Design

Metal hydrides are serious contenders for materials-based hydrogen storage to overcome constraints associated with compressed or liquefied H 2 . Their ultimate performance is usually evaluated using intrinsic material properties without considering a systems design perspective. An illustrative case with startling implications is (LiNH 2 +2LiH). Using models that simulate the storage system and associated fuel cell of a light-duty vehicle (LDV), the performance of the bulk hydrides is compared with a nanoscaled version in porous carbon (PC), (LiNH 2 +2LiH)@(6-nm PC). Using experimental material properties, the simulations show that (LiNH 2 +2LiH)@(6-nm PC) counterintuitively has higher usable gravimetric and volumetric capacities than the bulk counterpart on a system basis despite having lower capacities on a materials-only basis. Nanoscaling increases the thermal conductivity and lowers the desorption enthalpy, which consequently increases heat management efficiency. In a simulated drive cycle for fuel cell-powered LDV, the fuel cell is inoperable using bulk (LiNH 2 +2LiH) as the storage material but completes the drive cycle using the nanoscale material. Further, these results challenge the notion that nanoscaling incurs mass and volume penalties. Instead, the synergistic nanoporous host-hydride interaction can favorably modulate chemical and heat transfer properties. Moreover, a co-design approach considering application-specific tradeoffs is essential to accurately assess a material's potential for real-world hydrogen storage.

08 HYDROGEN↗
DOE OSTI2015

Ford/BASF/UM Activities in Support of the Hydrogen Storage Engineering Center of Excellence

Widespread adoption of hydrogen as a vehicular fuel depends critically on the development of low-cost, on-board hydrogen storage technologies capable of achieving high energy densities and fast kinetics for hydrogen uptake and release. As present-day technologies -- which rely on physical storage methods such as compressed hydrogen -- are incapable of attaining established Department of Energy (DOE) targets, development of materials-based approaches for storing hydrogen have garnered increasing attention. Material-based storage technologies have potential to store hydrogen beyond twice the density of liquid hydrogen. To hasten development of these ‘hydride’ materials, the DOE previously established three centers of excellence for materials storage R&D associated with the key classes of materials: metal hydrides, chemical hydrogen, and adsorbents. While these centers made progress in identifying new storage materials, the challenges associated with the engineering of the system around a candidate storage material are in need of further advancement. In 2009 the DOE established the Hydrogen Storage Engineering Center of Excellence with the objective of developing innovative engineering concepts for materials-based hydrogen storage systems. As a partner in the Hydrogen Storage Engineering Center of Excellence, the Ford-UM-BASF team conducted a multi-faceted research program that addresses key engineering challenges associated with the development of materials-based hydrogen storage systems. First, we developed a novel framework that allowed for a material-based hydrogen storage system to be modeled and operated within a virtual fuel cell vehicle. This effort resulted in the ability to assess dynamic operating parameters and interactions between the storage system and fuel cell power plant, including the evaluation of performance throughout various drive cycles. Second, we engaged in cost modeling of various incarnations of the storage systems. This analysis revealed cost gaps and opportunities that identified a storage system that was lower cost than a 700 bar compressed system. Finally, we led the HSECoE efforts devoted to characterizing and enhancing metal organic framework (MOF) storage materials. This report serves as a final documentation of the Ford-UM-BASF project contributions to the HSECoE during the 6-year timeframe of the Center. The activities of the HSECoE have impacted the broader goals of the DOE-EERE and USDRIVE, leading to improved understanding in the engineering of materials-based hydrogen storage systems. This knowledge is a prerequisite to the development of a commercially-viable hydrogen storage system.

08 HYDROGEN↗
DOE OSTI2012

Eigenvalue sensitivity studies for the Fort St. Vrain high temperature gas-cooled reactor to account for fabrication and modeling uncertainties

Uncertainties in the composition and fabrication of fuel compacts for the Fort St. Vrain (FSV) high temperature gas reactor have been studied by performing eigenvalue sensitivity studies that represent the key uncertainties for the FSV neutronic analysis. The uncertainties for the TRISO fuel kernels were addressed by developing a suite of models for an 'average' FSV fuel compact that models the fuel as (1) a mixture of two different TRISO fuel particles representing fissile and fertile kernels, (2) a mixture of four different TRISO fuel particles representing small and large fissile kernels and small and large fertile kernels and (3) a stochastic mixture of the four types of fuel particles where every kernel has its diameter sampled from a continuous probability density function. All of the discrete diameter and continuous diameter fuel models were constrained to have the same fuel loadings and packing fractions. For the non-stochastic discrete diameter cases, the MCNP compact model arranged the TRISO fuel particles on a hexagonal honeycomb lattice. This lattice-based fuel compact was compared to a stochastic compact where the locations (and kernel diameters for the continuous diameter cases) of the fuel particles were randomly sampled. Partial core configurations were modeled by stacking compacts into fuel columns containing graphite. The differences in eigenvalues between the lattice-based and stochastic models were small but the runtime of the lattice-based fuel model was roughly 20 times shorter than with the stochastic-based fuel model. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

OVERVIEW OF ALUMINUM CLADDING OXIDE PREDICTION MODELS FOR EUROPEAN HIGH POWER RESEARCH REACTORS

UMo/Al dispersion fuel clad with aluminum alloy is a primary candidate fuel form that is being developed for the European high power research reactors (EUHPRR). In BR2 tests, water oxidation of cladding grew thick due to high powers. Excessive cladding oxidation is a cause for serious fuel performance degradation in a test while fuel performance itself may be sound. It is necessary for a fuel designer to have a reliable model to predict cladding oxidation for fuel licensing activities. Using oxide data from BR2 tests, this study reviews the prominent oxide prediction models and cladding-to-coolant heat transfer correlations. Because the oxide growth models were based upon different coolant flow conditions from those of the EUHPRR, the effect of the heat transfer correlations at the cladding-to-coolant interface on the oxide thickness prediction were also examined. The best combination of the oxide prediction model and the heat transfer correlation applicable for the EUHPRR is recommended. Since the existing oxide models were all developed for aluminum alloy types different from those currently considered for the EUHPRR, the effect of cladding alloy type is also discussed.

Research reactor fuel, aluminum oxidation, in-pile↗
DOE OSTI2020

A detailed chemical kinetic modeling and experimental investigation of the low- and high-temperature chemistry of n-butylcyclohexane

Chemical kinetic models of gasoline, jet, and diesel fuels and their mixtures with bioderived fuels are needed to assess fuel property effects on efficiency, emissions, and other performance metrics in internal combustion and gas turbine engines. As these real fuels have too many fuel components to be included in a chemical kinetic model, surrogate fuels containing fewer components are used to represent them. These surrogate fuels mimic the chemical classes or molecular structures contained in the real fuel. One of the important chemical classes in gasoline, jet, and diesel fuels comprises cyclohexanes. Cyclohexanes comprise about 30% or more by weight in diesel fuel. Also, Mueller et al proposed n-butylcyclohexane (nBCH) as a component in a nine-component surrogate palette to represent the ignition properties, distillation curve, density, and molecular structures of a diesel certification fuel. Herein, experimental measurements of the ignition delay times (IDTs) of nBCH in a shock tube and in a rapid compression machine are reported over a wide range of temperature, pressure, and equivalence ratio important for enabling the validation of a chemical kinetic model for nBCH for combustion in diesel engines. The range of conditions are temperatures of 630–1420 K, pressures of 10, 30, and 50 bar, and equivalence ratios of 0.3, 0.5, 1.0, and 2.0 in ‘air’. A detailed chemical model is developed for nBCH to simulate its ignition at both low and high-temperature conditions and at relevant elevated pressures. The experimentally measured IDTs are used to improve and validate the chemical kinetic model.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2019

Separate-Effects Validation Experiments for Models of Fracture in Ceramic Nuclear Fuel

A number of advances have recently been made in modeling fracture initiation and propagation in ceramic light water reactor (LWR) fuel. These include the use of techniques such as peridynamics, the extended finite element method (XFEM), the discrete element method (DEM), and phase field fracture. Improved predictions of fracture in LWR fuel can benefit fuel performance simulations in a number of ways, with important implications for radial relocation, pellet-cladding mechanical interactions, and fragmentation during accident conditions. Although these modeling approaches appear promising, to-date most of their applications have been to fresh fuel. A major limitation in applying these approaches with confidence to a wider variety of conditions, such as fragmentation of high-burnup fuel, is that there is very limited data that can be directly used to validate the results of these fracture models. While fracture patterns are available from cross-sections of numerous irradiated fuel rods, the fuel rods have typically been subjected to complex irradiation histories. To validate models of fracture propagation, focused experiments that give snapshots of the initiation and propagation of fracture under well characterized prototypical reactor conditions are needed. Under a multi-institutional U.S. Department of Energy Integrated Research Project (IRP), three separate-effects validation experiments are underway to provide information on the propagation of cracks in LWR fuel under controlled conditions. The primary challenge in developing such an experiment is in providing conditions that result in a radial temperature distribution that is prototypical of LWR in-reactor conditions. Two out-of-reactor experiments are being planned that employ very different approaches to achieve such a temperature distribution. The first of these achieves this by employing a combination of inductive and resistive electrical heating to generate volumetric heating. The second experiment involves slowly heating a pellet in a metal rod to a high temperature, and then quenching that rod in a cold bath. The quenching process generates a temperature profile approximating that in an LWR fuel rod for a moment, but has a very different temperature history from the LWR. Finally, an experiment is being planned for the Idaho National Laboratory TREAT reactor that will use nuclear fission as a heat source, and which will employ a metal heat sink to replicate the thermal conditions provided by coolant in an LWR. These experiments all have their advantages and drawbacks, with varying levels of ability to reproduce actual LWR conditions and varying ability to use instruments to monitor quantities of interest over the course of the experiment that can be used to validate fracture models. For that reason, these different experimental approaches are complimentary to each other, and the combined results are expected to significantly improve confidence in the ability of computational models to simulate fracture in ceramic nuclear fuel.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Autoignition response of n -butanol and its blend with primary reference fuel constituents of gasoline.

We study the influence of blending n-butanol on the ignition delay times of n -heptane and iso-octane, the primary reference fuels for gasoline. The ignition delay times are measured using a rapid compression machine, with an emphasis on the low-to-intermediate temperature conditions. The experiments are conducted at equivalence ratios of 0.4 and 1.0, for a compressed pressure of 20 bar, with the temperatures at the end of compression ranging from 613 K to 979 K. The effect of n -butanol addition on the development of the two-stage ignition characteristics for the two primary reference fuels is also examined. The experimental results are compared to predictions obtained using a detailed chemical kinetic mechanism, which has been obtained by a systematic merger of previously reported base models for the combustion of the individual fuel constituents. In conclusion, a sensitivity analysis on the base, and the merged models, is also performed to understand the dependence of autoignition delay times on the model parameters.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2025

Fluoride-Cooled High-Temperature Pebble-Bed Reactor Reference Plant Model Updates

This work presents the latest improvements to, and investigations performed with, the Fluoride-Cooled High-Temperature Pebble-Bed Reactor reference plant models for the United States Nuclear Regulatory Commission. These models, developed with the Comprehensive Reactor Analysis Bundle, or BlueCRAB, serve as the foundation for the future development of detailed design evaluation models based on license applications. BlueCRAB is the code suite proposed for non-light-water reactor systems safety analyses, and it incorporates various simulation tools developed by the Nuclear Energy Advance Modeling and Simulation program, including the Griffin code for reactor physics, the Pronghorn and SAM codes for core thermal fluids, the BISON code for solid conduction and fuel performance, and the SAM code for system analysis. The primary objective of this work is to assess the level of readiness of BlueCRAB for modeling fluoride-cooled high-temperature pebble-bed reactors. We first developed numerical models in BlueCRAB that include the key physics for this technology, ensuring an adequate level of fidelity for modeling the core performance during accident scenarios. This was followed by simulation of transient scenarios, two loss-of-forced-cooling events (one protected and one unprotected), and two control rod withdrawal events (one delayed and one prompt supercritical reactivity insertion). The analysis includes comparisons between the 2-D thermal fluid porous media models in Pronghorn and SAM, comparisons between coupled Pronghorn-Griffin and coupled SAM-Griffin models for two loss-of-forced cooling events and one control rod withdrawal event, and comparisons between SAM single-solve and domain-overlapping approaches for multi-scale thermal fluid coupling. In addition, we performed comparisons between 3-D, 2-D, and 0-D neutronic models for the two control rod withdrawal scenarios with Pronghorn-Griffin. The results show that the BlueCRAB models led to physically intuitive solutions for the scenarios examined. The changes in the various scalar and vector fields such as the neutron flux, power, temperatures, densities, pressures, and velocities are all within the expected ranges, and their distributions can be explained from the system response of the transients and the geometric and material variations. Several comparisons suggest that the porous media models in Pronghorn and SAM can lead to similar solutions, even though they are based on different methodologies. The simulations demonstrate that there are differences between the various levels of fidelity, and it is advisable to have flexible tools that can cover the breadth and depth of needs that may arise in future technical evaluations. We believe that BlueCRAB’s capabilities represent a significant asset for confirmatory analyses aimed at resolving important safety questions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1979

Effect of time dependent flight loads on JT9D-7 performance deterioration

The results of a modal transient analysis of the engine/aircraft system are presented. The response of the JT9D to analytically simulated vertical gusts and landings was predicted using a NASTRAN finite element mathematical model of the JT9D/747 propulsion system. The NASTRAN finite element model of the propulsion system included engine structural models of the fan, low/high pressure compressors, diffuser/turbine cases, and high/low pressure rotors, as well as nacelle models of the inlet cowl, tailcone, and wing pylon. The analysis conducted predicts that an insignificant level of JT9D-7 performance deterioration would occur due to a typical vertical gust encounter or a typical revenue service landing. Analysis of a high sink rate landing with a heavy fuel load indicates the possibility of local wear, however, the lack of an accurate dynamic rotor/seal interference model precludes an accurate quantitative evaluation of performance change for this once-per-airframe-life event.

Jay, A.↗
DOE OSTI2022

Thermal Modelling of Advanced Test Reactor Fuel in a Generalized Dry Storage System

Star-CCM+, a computational fluid dynamics (CFD) software was used to conduct modeling and simulation of the thermal performance of a dry storage configuration consisting of Department of Energy Standardized Canisters (DOESCs) loaded with aluminum-clad spent nuclear fuel (ASNF.) The configuration includes nine DOESCs loaded with Advanced Test Reactor ASNF contained within a stainless steel overcanister centered in a ventilated, concrete overpack. The simulations were used to estimate the maximum temperatures reached by backfill gases inside the overcanister and DOESCs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Performing k eff Validation of As-Loaded Criticality Safety Calculations Using UNF-ST&DARDS: Applicable Experiment Selection [Slides]

This presentation discusses the UNF-ST&DARDS which performs many analyses for as-loaded SNF canisters, including criticality safety, shielding, thermal-hydraulic, and containment. It also discusses the experiment selection based on c k assessment of similarity and that the c k value of 0.8 or greater considered applicable for validation. 11 PWR SNF canisters (MPC-32) are used in this work. 1 model represents a failed fuel assembly as fresh, per the design basis and the remaining 10 models represent all 32 assemblies with depleted fuel. In conclusion, critical experiment selection can be performed with S/U techniques for as-loaded canisters in UNF-ST&DARDS. Sufficient benchmark experiments exist to support validation and additional MOX experiments that are a good match for commercial SNF would be a benefit to provide independent data. The presentations states that S/U techniques identify different pools of experiments for different systems and that the process is amenable to automation within UNF-ST&DARDS.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

Improving Durability of Turbine Components Through Trenched Film Cooling and Contoured Endwalls

The experimental and computational studies of the turbine endwall and vane models completed in this research program have provided a comprehensive understanding of turbine cooling with combined film cooling and TBC. To correctly simulate the cooling effects of TBC requires the use of matched Biot number models, a technique developed in our laboratories. This technique allows for the measurement of the overall cooling effectiveness which is a measure of the combined internal and external cooling for a turbine component. The overall cooling effectiveness provides an indication of the actual metal temperature that would occur at engine conditions, and is hence a more powerful performance indicator than the film effectiveness parameter that is commonly used for film cooling studies. Furthermore these studies include the effects of contaminant depositions which are expected to occur when gas turbines are operated with syngas fuels. Results from the endwall studies performed at Penn State University and the vane model studies performed at the University of Texas are the first direct measurements of the combined effects of film cooling and TBC. These results show that TBC has a dominating effect on the overall cooling effectiveness, which enhances the importance of the internal cooling mechanisms, and downplays the importance of the film cooling of the external surface. The TBC was found to increase overall cooling effectiveness by a factor of two to four. When combined with TBC, the primary cooling from film cooling holes was found to be due to the convective cooling within the holes, not from the film effectiveness on the surface of the TBC. Simulations of the deposition of contaminants on the endwall and vane surfaces showed that these depositions caused a large increase in surface roughness and significant degradation of film effectiveness. However, despite these negative factors, the depositions caused only a slight decrease in the overall cooling effectiveness on the endwall, and in some cases a slight increase in overall cooling effectiveness on the vane model. This was attributed to the insulating effects of the depositions which compensated for the negative factors.

Bogard, David G.↗
DOE OSTI2019

Summary of Grizzly Development for Advanced Reactor Structural Materials

This summary report contains an overview of work performed during Fiscal Year 2019 under the Nuclear Engineering Advanced Modeling and Simulation (NEAMS) Fuels Product Line work package entitled "MS-19IN020106 - GRIZZLY - INL". Grizzly is a code based on the MOOSE framework that has been under development for several years funded by the US Department of Energy’s Light Water Reactor Sustainability (LWRS) program. Prior to 2019, Grizzly development was focused on development of capabilities needed to simulate the effects of aging mechanisms in light water reactor (LWR) components. The goal of this LWR-focused work has been to develop tools that can be used to assess the ability of existing LWRs to perform safety in long term operation (LTO) scenarios. This LWR-focused work has primarily focused on the effects of reactor pressure vessel (RPV) embrittlement and aging of reinforced concrete structures. Grizzly development to support LWR degradation needs is still ongoing, but starting in 2019, Grizzly’s capabilities are being expanded to also meet growing needs for a tool to address structural component integrity and degradation in advanced reactors. A simulation capability for advanced reactor components has many of the same basic requirements that one for LWRs does, such as the ability to solve coupled systems of partial differential equations that arise from the fact that many of the problems of interest involve multiple coupled physics. The materials issues also inherently involve multiple length scales, with the response at the engineering scale being strongly affected by mechanisms occurring at atomistic and mesoscopic scales.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

A combined theoretical-experimental investigation of thermal transport in low-dose irradiated thorium dioxide

During reactor operation, nuclear fuels are subject to extreme temperature and irradiation conditions which can significantly degrade the fuel's thermal transport properties. The reduction in thermal conductivity of the fuel as a result of irradiation-induced lattice defects is arguably the most important fuel performance metric in regard to reactor efficiency and safety. Because thorium dioxide (ThO2) is suitable as a model system for more complex materials such as UO2 and its mixed oxides, we present a theoretical investigation of thermal conductivity reduction seen in defect-bearing thorium dioxide and compare directly to experimental measurements. Phonon-mediated thermal transport of the fuel is modeled by a solution to the Boltzmann transport equation (BTE) for phonons. A cluster dynamics (CD) model for lattice defect evolution during irradiation predicts defect densities which are used as input to the BTE for modeling phonon-defect scatterings. Phonon scatterings by lattice defects include those from point defects and vacancy clusters and interstitial clusters of various sizes. The CD model is benchmarked against structural defect characterization of irradiated thorium dioxide using electron microscopy. Thermal conductivity predicted by the BTE model is compared to measured values for irradiated thorium dioxide specimens below room temperature to isolate effects of phonon-defect scattering from intrinsic 3-phonon processes, which dominate at higher temperatures. The computed conductivity values are in partial agreement at temperatures close to room temperature while slight deviations are observed at the lowest measured temperatures, suggesting that implemented phonon-defect scattering cross-section expressions may not be adequate for low temperatures. The presented work provides a necessary investigation of the influence of irradiation induced defects on fuel performance and represents a first step toward a full characterization of phonon mediated thermal transport in irradiated materials with complex defect microstructure.

36 MATERIALS SCIENCE↗
DOE OSTI2014

BEAMS3D Neutral Beam Injection Model

With the advent of applied 3D fields in Tokamaks and modern high performance stellarators, a need has arisen to address non-axisymmetric effects on neutral beam heating and fueling. We report on the development of a fully 3D neutral beam injection (NBI) model, BEAMS3D, which addresses this need by coupling 3D equilibria to a guiding center code capable of modeling neutral and charged particle trajectories across the separatrix and into the plasma core. Ionization, neutralization, charge-exchange, viscous velocity reduction, and pitch angle scattering are modeled with the ADAS atomic physics database. Benchmark calculations are presented to validate the collisionless particle orbits, neutral beam injection model, frictional drag, and pitch angle scattering effects. A calculation of neutral beam heating in the NCSX device is performed, highlighting the capability of the code to handle 3D magnetic fields.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗