SEARCH · Engineering Papers

Results for “Monte Carlo Code”

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.

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,621 records · Page 90

All sources NASA NTRS DOE OSTI

DOE OSTI2011

Modeling of nanocluster formation by ion beam implantation

A theoretical model was developed to investigate the mechanism of the formation of nanoclusters via ion beam implantation. The evolution of nanoclusters, including the nucleation and growth process known as Ostwald ripening, was rebuilt using numerical simulations. The effects of implantation parameters such as the ion energy, ion fluence, and temperature on the morphology of implanted microstructures were also studied through integration with the Monte Carlo Transport of Ions in Matter code calculation for the distribution profiles of implanted ions. With an appropriate ion fluence, a labyrinth-like nanostructure with broad size distributions of nanoclusters formed along the ion implantation range. In a latter stage, a buried layer of implanted impurity developed. With decreasing ion energy, the model predicted the formation of precipitates on the surface. These simulation results were fully consistent with many experimental observations. With increased temperature, the characteristic length and size of nanostructures would increase due to the high mobility. This theoretical model provides an efficient numerical approach for fully understanding the mechanism of the formation of nanoclusters, allowing for the design of ion beam experiments to form specific nanostructures through ion-implantation technology.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

DOE OSTI2015

Radiometric characterization process for locating radioactivity hold-up and measuring non-uniformly distributed sources - 15342

A novel approach to facility radiometric characterization for non-uniformly distributed radiation sources has been developed by ANTECH and is currently being employed at the Argonne National Laboratory. It is being applied to the measurement of a wide range of objects and the interiors and contents of laboratories and buildings. The integrated non destructive assay measurement process involves the use of a scanning spectroscopic gamma ray camera as well as high purity Germanium detectors with different collimators and hence differing detector fields of view to measure and quantify non-uniformly distributed sources of radioactivity in laboratory equipment, glove boxes, storage tanks, pipework and general laboratory objects and process components. The purpose of the radiometric characterization is to quantify radiation sources and radioactive contamination levels in facilities containing nuclear material and to ensure compliance with facility characterization requirements and for decontamination and decommissioning planning and execution. The integrated measurement process also involves automated data capture and processing software and a database with which both measurement instrument operators and subject matter expert analysts can interact. An appropriate measuring instrument (gamma camera, scintillation detector, high purity Germanium detector system, neutron detector or alpha-beta counter) records the measurement data and the measurement site is recorded in a photograph. All of the data relevant to the measurement (or set of measurements for a particular object) are recorded in the database including measured gamma ray spectra, equipment description and relevant calibration, measurement site photograph and measurement descriptive data. Once data analysis protocols are established for specific geometries and measurement situations they can be applied to the measured data in a partially automated manner, reducing analysis and transcription errors. In addition to measurements and data analysis the process also utilises radiometric modelling computer codes including MicroShield and Monte Carlo Neutrons Photons in order to perform model calculations where appropriate. In a typical measurement, the ANTECH RadSearch spectroscopic gamma camera is utilized to find areas of concentrated activity so that these areas can be measured and calculated separately from areas with diffuse contamination. The use of the gamma camera in scanning mode helps to ensure that the assay does not overestimate the detected radiation source, which is a frequent consequence of using standard far field geometry and attenuation analysis codes that assume uniformity of both density and radioactive source distribution. Once the hot spots are identified, high purity Germanium detector measurements are made using various collimators with wide or narrow fields of view as required. These measurements are then processed through a semi-automatic calculation process assuming an appropriate source configuration as identified by RadSearch. These source configurations include point sources, line sources or surface sources. The Generalized Geometry Hold-up algorithms are employed to determine an estimate of the activity of the various sources that are identified in a measurement. The ANTECH integrated measurement process provides an automated calculation procedure where the analyst can select the source type, geometry and appropriate input data, and then the software calculates the activity by radionuclide for all sources identified in the measurement. The measurement and analysis process is controlled under the ANTECH ISO9001: 2008 Quality Assurance System and managed by a site specific Quality Assurance Plan. Both Argonne National Laboratory and Department of Energy technical reviewers have vetted the Integrated Measurement and Analysis Process. The application of the measurement and analysis process is described and measurement examples are provided. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Improved Cross Section Generation Capability of Griffin in FY22

The Griffin code is a Multiphysics Object-Oriented Simulation Environment (MOOSE) based reactor multiphysics analysis application jointly developed by Idaho National Laboratory and Argonne National Laboratory. The code includes a variety of deterministic steady-state transport solvers for fixed source, k-eigenvalue, adjoint, and subcritical multiplication as well as transient solvers for spatial dynamics with the improved quasi-static method. Griffin uses cross section data in the ISOXML format generated from external deterministic or Monte Carlo cross section generation codes. In recent years, the MC 2 -3 modules have been added to Griffin for fast reactor cross section generation, and the self-shielding application programming interface (SSAPI) was implemented in the ISOXML module for thermal reactor cross section generation. The on-the-fly slowing down method and double-heterogeneity treatment have been implemented to SSAPI and verified against particulate fuel-bearing graphite-moderated thermal reactor problems with high accuracy. This year, work has been focused on improving the cross section generation capability of ISOXML and streamlining the cross section generation procedures. In addition, the form function data were added to ISOXML in order to support the pin power reconstruction capability that was newly implemented in Griffin in this fiscal year. To facilitate the cross section generation using MC 2 -3 and SSAPI in Griffin, the cross section generation workflows have been set up for both fast and thermal spectrum reactors. The MOOSE action system tool was devised for fast spectrum problems, and the MOOSE stochastic tool was adopted to the branch calculation procedure for thermal spectrum problems. Meanwhile, to ensure the accuracy of group-constants, the thermal up-scattering kernel calculator accounting for resonance scattering was implemented in ISOXML, demonstrating the accurate computation of a Doppler-broadened scattering kernel of any Legendre order within a reasonable timescale. Other aspects of ISOXML, such as deletion solver and data, documentation, ISOXML file management, and the interface for Mixture, were improved as well.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Investigation of Delayed Neutron Sensitivities for Several ICSBEP Benchmarks using MCNP

The effective delayed neutron (β$_{eff}$) is a very important parameter for reactor and criticality applications. This parameter is equal to the difference in reactivity between delayed critical ($k_{eff}$ = 1, which requires both prompt and delayed neutrons to achieve criticality) and prompt critical ($k_p$ = 1, which requires only prompt neutrons to achieve criticality). This is often referred to as the delayed critical "window" (the region of criticality between delayed and prompt critical). β$_{eff}$ is a reactor kinetics parameters and depends on the nuclides in the system that undergo fission as well as the spectral characteristics of the system. Measurements of β$_{eff}$ have been performed for many criticality experiments. The EUCLID (Experiments Underpinned by Computational Learning for Improvements in nuclear Data) project at Los Alamos National Laboratory (LANL) aims to constrain nuclear data by using a suite of measurement types beyond $k_{eff}$. Our team has recently investigated the use of pulsed spheres for nuclear data validation. Several other measurement methods are also of interest, including β$_{eff}$ (investigated here) and reactivity coefficients (investigated in a separate work at this same meeting). One focus of our work is to determine if other methods are complimentary to the critical experiments already utilized for nuclear data validation. This is important because if a method has similar sensitivities then it will not be particularly useful for nuclear data validation as it will provide the same information as the critical experiments already being used. Here "similar" could refer to several characteristics, one being the shape of a sensitivity profile over energy. In the future, these methods will be utilized (with both existing and new experiments) in machine learning algorithms for nuclear validation, similar to what is currently done for criticality experiments. In order to use a measurement type for nuclear validation, it is necessary to obtain cross-section sensitivities for that parameter. This work looks at one approach to estimate β$_{eff}$ sensitivities by utilizing $k_{eff}$ sensitivities within Monte Carlo N-Particle ® Code Version 6.2. This is applied to several criticality benchmarks. Results are compared and the benefits and limitations of this approach are discussed.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2016

Deadtime Determination of Helium Detector Using MCNP Code

In detector systems, the minimum time separation between the events is called detector's deadtime. The The intrinsic properties of the detector and the pulse processing circuitry's characteristics are the sources of deadtime. Researchers have been working on improving a detector deadtime model that can implicitly characterize a detection system's behavior while reducing counting errors. There are two commonly known dead-time models: the 'Paralyzing' and the 'Non-paralyzing' models. In reality, detection systems fit neither of these idealized models perfectly, instead falling somewhere between the two models. The paralyzing model is mathematically expressed by eq. 1, where m is the measured count rate, n is the true count rate and t is dead-time. This model assumes that each event during the dead-time will reset it to a fixed duration, thus extending the deadtime. The dead-time extension depends on the count rate. According to the non-paralyzing model, deadtime is fixed after each detected event, and all events occurring during deadtime are lost. The fraction of time during which an apparatus is sensitive is 1-mt. Therefore, the fraction of the true number of events can be recorded. In this study, a Canberra {sup 3}He neutron detector (0.5NH1/1K) deadtime value was estimated using a combination of experimental subcritical assembly measurements and theoretical MCNP simulations. The deadtime of the Canberra 10 mm diameter {sup 3}He tube detector was calculated by comparing measured counts from different locations in the subcritical assembly at Missouri University of Science and Technology Reactor (MSTR) with MCNP (Monte Carlo N-Particle Code) simulations. The deadtime calculations are provided for both the paralyzing and non-paralyzing models for the Canberra 10 mm diameter {sup 3}He tube detector. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2012

Time-resolved ion energy distribution measurements using an advanced neutral particle analyzer on the MST reversed-field pinch

An advanced neutral particle analyzer (ANPA) capable of simultaneously measuring hydrogen and deuterium ions of energies up to 45 keV has recently been developed for use on the Madison Symmetric Torus. The charge-to-mass separation allows for separate analysis of bulk deuterium ions and hydrogen ions injected with a 1 MW, 25 keV neutral beam. Orientation of the ANPA allows sampling of different regions of ion velocity space; a radial viewport favors collection of ions with high v{sub Up-Tack }/ Double-Vertical-Line v Double-Vertical-Line while a recently installed tangential viewport favors ions with high v{sub Double-Vertical-Line Double-Vertical-Line }/ Double-Vertical-Line v Double-Vertical-Line , such as those from the core-localized fast ion population created by the neutral beam. Signals are observed in the ANPA's highest energy channels during periodic magnetic reconnection events, which are drivers of anisotropic, non-Maxwellian ion energization in the reversed-field pinch. ANPA signal strength is dependent on the background neutral density, which also increases during magnetic reconnection events, so careful analysis must be performed to identify the true change in the ion distribution. A Monte Carlo neutral particle tracing code (NENE) is used to reconstruct neutral density profiles based on D{sub {alpha}} line emission, which is measured using a 16-chord filtered photodiode array.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2015

MCNP Progress & Performance Improvements

Twenty-eight slides give information about the work of the US DOE/NNSA Nuclear Criticality Safety Program on MCNP6 under the following headings: MCNP6.1.1 Release, with ENDF/B-VII.1; Verification/Validation; User Support & Training; Performance Improvements; and Work in Progress. Whisper methodology will be incorporated into the code, and run speed should be increased.

42 ENGINEERING↗

DOE OSTI2014

SU-E-T-380: Particle Microdosimetry Study Based On 3D-Cylindrical Silicon Radiaton Detectors

Purpose: A new design of a solid-state-microdetector based on silicon 3D microfabrication and its performance to characterise Lineal energy, Specific Energy, dose, LET and other microdosimetric variables required for modelling particle relative biological effectiveness (RBE) is presented. Methods: A microdosimeter formed by a matrix of independent sensors with well-defined micrometric cylindrical shape and with a volume similar to those of cellular dimensions is used to measure microdosimetric variables. Each sensor measures the radiation deposited energy which, divided by the mean cord length of the sensors, provides us with the Linear Energy (y) of the radiation as well as its energy distribution, and frequencymean. Starting from the these distributions in different points of a proton beam, we generate biophysical data (e.g. Linear Energy Transfer (LET), Specific Energy (z), etc…) needed for relative biological effectiveness (RBE) calculations radiation effect models used in particle radiotherapy treatment planning. In addition, a Tissue Equivalent Proportional Counter (TEPC) will be used as baseline to calibrate the “y” magnitude of the microdosimeter unit-cells. Results: The experimental measurements will soon be carried out at the Perelman Center for Advanced Medicine (University of Pennsylvania), which provides proton beam for clinical research proposals. The results of distributions measured of the microdosimetric variables from the first tests developed in the proton facility will be presented and compared with Monte Carlo simulations using the Geant4 code. Conclusion: The use of 3D microdosimeters such as the one presented here will enhance the accuracy of RBE calculations normally affected by the inherent uncertainty of monte carlo simulations due to the approximation of material composition and energy dependent physical laws involved in such calculations. The effect of such approximations will be quatified by comparison with absolute measurement of radiation quality parameters.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2013

Uncertainty Underprediction in Monte Carlo Eigenvalue Calculations

It is well-known that statistical estimates obtained from Monte Carlo criticality simulations can be adversely affected by cycle-to-cycle correlations in the fission source, which can lead to estimates of statistical uncertainties that are lower than the true uncertainty by a factor of 5 or more. However, several other more fundamental issues such as adequate source sampling over the fissionable regions and source convergence can have a significant impact on the uncertainties for the calculated eigenvalue and localized tally means, and these issues may be mistaken for effects resulting from cycle-to-cycle correlations. In worst-case scenarios, the uncertainty may be underpredicted by a factor of 40 or more. Since Monte Carlo methods are widely used in criticality safety applications and are increasingly being used for benchmarking reactor analyses, an in-depth understanding of the effects of these issues must be developed in order to support the practical use of Monte Carlo software packages. A rigorous statistical analysis of eigenvalue and localized tally results in Monte Carlo criticality calculations is presented using the SCALE/KENO-VI (continuous-energy version) and MCNP codes. The purpose of this analysis is to investigate the underprediction of uncertainty and its sensitivity to problem characteristics and calculational parameters using two of the most widely used Monte Carlo criticality codes. For the problems considered here, which are fuel rod and fuel assembly problems with reflecting boundary conditions on all four horizontal sides, we show that adequate source convergence along with proper specification of Monte Carlo parameters can reduce the magnitude of uncertainty underprediction to reasonable levels, below a factor of 2 in most cases.

Mervin, Brenden T↗

DOE OSTI2012

Hybrid parallel code acceleration methods in full-core reactor physics calculations

When dealing with nuclear reactor calculation schemes, the need for three dimensional (3D) transport-based reference solutions is essential for both validation and optimization purposes. Considering a benchmark problem, this work investigates the potential of discrete ordinates (Sn) transport methods applied to 3D pressurized water reactor (PWR) full-core calculations. First, the benchmark problem is described. It involves a pin-by-pin description of a 3D PWR first core, and uses a 8-group cross-section library prepared with the DRAGON cell code. Then, a convergence analysis is performed using the PENTRAN parallel Sn Cartesian code. It discusses the spatial refinement and the associated angular quadrature required to properly describe the problem physics. It also shows that initializing the Sn solution with the EDF SPN solver COCAGNE reduces the number of iterations required to converge by nearly a factor of 6. Using a best estimate model, PENTRAN results are then compared to multigroup Monte Carlo results obtained with the MCNP5 code. Good consistency is observed between the two methods (Sn and Monte Carlo), with discrepancies that are less than 25 pcm for the k{sub eff}, and less than 2.1% and 1.6% for the flux at the pin-cell level and for the pin-power distribution, respectively. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

A recoverable gas-cell diagnostic for the National Ignition Facility

The high-fluence neutron spectrum produced by the National Ignition Facility (NIF) provides an opportunity to measure the activation of materials by fast-spectrum neutrons. A new large-volume gas-cell diagnostic has been designed and qualified to measure the activation of gaseous substances at the NIF. This in-chamber diagnostic is recoverable, reusable and has been successfully fielded. Data from the qualification of the diagnostic have been used to benchmark an Monte Carlo N-Particle Transport Code simulation describing the downscattered neutron spectrum seen by the gas cell. We present early results from the use of this diagnostic to measure the activation of {sup nat}Xe and discuss future work to study the strength of interactions between plasma and nuclei.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2015

Complik: A versatile tool for automatizing the representation of material data in MCNP models - 14412

At the Paul Scherrer Institut, the CASMO/SIMULATE code system is used as reference scheme for core analyses of the Swiss reactors. Besides, Monte Carlo (MC) based neutron transport codes are in parallel being established both for out-of-core applications and full core analyses, in order to provide higher fidelity solutions to problems beyond the modelling capabilities of the standard core modelling methods. In this perspective, a necessity is to be able to conduct the MC calculations at any operating conditions, requiring thus proper initialisation of the thermal-hydraulic feedback as well as at any burnup, prompting the need for a detailed modelling of all material zones in the irradiated fuel assemblies. To tackle the later challenge, the development of the COMPLINK tool, aimed at initialising large-scale MC models of burned core configurations, using as basis upstream 2-D assembly/ 3-D core analyses results, was initiated and is presented in this paper. A main feature is the capability to extend in an automatic manner, a user given MC model in order to represent material data with a higher resolution, keeping track of the location of such materials also within nested repeated structures. This capability prevents the user from the burden of developing specific computational framework for a target application. Since the same functionalities are required to analyse out-of-core spent fuel managements systems, a first verification of COMPLINK is illustrated on the basis of a criticality safety evaluation (CSE) with MCNP6.1 of a Swiss PWR wet storage pool loaded with fresh and burned fuel assemblies. Starting from CASMO-5 depletion calculations at each axial layer of the depleted fuel assemblies and reconstructed based on realistic irradiation histories, COMPLINK is then applied to load the isotopic compositions into the material zones of the corresponding MCNP6.1 assembly models. On this basis, the CSE analyses are conducted with focus on investigating axial and radial modelling effects. All in all, the obtained results are found to be well in line with the impact on CSE expected at the considered burnup levels, providing thus confidence in the correct implementation of the COMPLINK functionalities and confirming its effectiveness as well as efficiency for the initialization of all material zones in large-scale MC models of complex and strongly heterogeneous burned configurations. (authors)

42 ENGINEERING↗

DOE OSTI2013

Direct Modeling of Scintillator Thickness for Optimal Light Output and Spatial Resolution

It is common in x-ray radiography to use scintillators (e.g., BGO or LSO) to convert x-rays to visible light, which is then recorded by an imaging system. The response of the scintillator depends fundamentally on its thickness, with respect to both its visible light emittance and its spatial resolution. This is important for optimizing light output, signal to noise ratio, or optical response time. Given that it is often cost-prohibitive to procure a variety of scintillator samples and empirically test the performance, it is essential to be able to model and accurately simulate the performance of a scintillator with respect to thickness and other properties, and a direct way of doing this is using Monte Carlo-based radiation transport codes. Such simulations can be expensive in terms of computational time, and the codes are not easily obtained. In this work we first show such simulations, and demonstrate that there is a natural trade-off between light output of a scintillator and its spatial resolution. We then derive a first-principles model that accurately approximates the light output, using straightforward calculations that can be performed quickly with any basic computing software. We compare the results to those obtained from Monte Carlo simulations and show that our simplified model can be used to analyze the tradeoff between emittance and resolution nearly as well as using a full-scale radiation transport code.

Monte Carlo methods, X-ray imaging, scintillation ↗

NASA NTRS2014

A Monte Carlo Analysis of the Thrust Imbalance for the RSRMV Booster During Both the Ignition Transient and Steady State Operation

This paper presents the results of statistical analyses performed to predict the thrust imbalance between two solid rocket motor boosters to be used on the Space Launch System (SLS) vehicle. Two legacy internal ballistics codes developed for the Space Shuttle program were coupled with a Monte Carlo analysis code to determine a thrust imbalance envelope for the SLS vehicle based on the performance of 1000 motor pairs. Thirty three variables which could impact the performance of the motors during the ignition transient and thirty eight variables which could impact the performance of the motors during steady state operation of the motor were identified and treated as statistical variables for the analyses. The effects of motor to motor variation as well as variations between motors of a single pair were included in the analyses. The statistical variations of the variables were defined based on data provided by NASA's Marshall Space Flight Center for the upgraded five segment booster and from the Space Shuttle booster when appropriate. The results obtained for the statistical envelope are compared with the design specification thrust imbalance limits for the SLS launch vehicle

Foster, Winfred A., Jr.↗

NASA NTRS1993

Transport methods and interactions for space radiations

This report presents a brief history leading to the involvement of the Langley Research Center of the National Aeronautics and Space Administration (NASA) in space-radiation physics and protection. Indeed, a relatively complete summary of technical capability as of the summer of 1990 is given. The Boltzmann equations for coupled ionic and neutronic fields are presented and inversion techniques for the Boltzmann operator are discussed. Errors generated by the straight ahead approximation are derived and are shown to be negligible for most problems of space-radiation protection. A decoupling of projectile propagation from the target fields greatly simplifies the Boltzmann equations and allows an analytic solution of the target fragment transport. Analytic and numerical methods of solving the projectile transport equations are discussed. The nuclear physics underlying the coefficients in the Boltzmann equation is discussed. A coupled-channel optical model is found as a consequence of the loose binding of nuclear matter and closure of the nuclear states in high-energy reactions. Transport solutions with the developed data base are used with laboratory experiments to validate both the transport code and the data base. Numerical benchmarks and comparison with Monte Carlo calculations are also used for code validation.

Wilson, John W.↗

NASA NTRS2014

A Monte Carlo Analysis of the Thrust Imbalance for the Space Launch System Booster During Both the Ignition Transient and Steady State Operation

Foster, Winfred A., Jr.↗

DOE OSTI2016

Algorithmic Improvements for Portable Event-Based Monte Carlo Transport Using the Nvidia Thrust Library

High performance computing environments are progressively moving towards many-core computing architectures. The Los Alamos National Laboratory Trinity machine, available in late 2016, will use both Intel Xeon Haswell processors and Intel Xeon Phi Knights Landing many integrated core (MIC) coprocessors. The Lawrence Livermore National Laboratory Sierra machine, available in 2018, will use an IBM PowerPC architecture along with Nvidia graphics processing units (GPUs). Applications that must work in this supercomputing environment must continue to adapt in order to take advantage of the diverse hardware architectures that are coming. A significant consideration is not only the performance of the application on a given platform but also the portability of the application to other platforms. The algorithmic improvements presented in this paper build upon recently-reported work on event-based Monte Carlo transport in the ALPSMC code that models particle transport in one-dimensional binary stochastic media. That paper discussed the lack of available vectorization in the traditional history-based algorithm used for Monte Carlo transport and presented a data parallel event-based algorithm implemented using the Nvidia Thrust library for portability. The performance of the data parallel event-based algorithm implemented using Thrust was compared to a native CUDA implementation. The conclusions from that work were that the Thrust library abstraction technique caused too significant a loss in performance but that the event-based method was a viable option that should be further investigated. In this paper, we describe algorithmic improvements to the data parallel event-based algorithm previously presented. We made further algorithmic optimizations to the event-based CUDA implementation, most notably: data structure changes, a new conditional particle removal scheme in the event-based process, and the use of multiple GPUs. In addition to improvements to the algorithm, we re-implemented the Thrust version from the now further optimized CUDA version, giving a greater chance for success at a performant abstraction. Finally, we revisited our previous assumptions about the inability of the history-based method to achieve performance on vector style architectures such as the MICs and GPUs, with surprising and promising results. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2016

Integration of OpenMC methods into MAMMOTH and Serpent

OpenMC, a Monte Carlo particle transport simulation code focused on neutron criticality calculations, contains several methods we wish to emulate in MAMMOTH and Serpent. First, research coupling OpenMC and the Multiphysics Object-Oriented Simulation Environment (MOOSE) has shown promising results. Second, the utilization of Functional Expansion Tallies (FETs) allows for a more efficient passing of multiphysics data between OpenMC and MOOSE. Both of these capabilities have been preliminarily implemented into Serpent. Results are discussed and future work recommended.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗