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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
DOE OSTI2024

Electricity and natural gas tariffs at United States wastewater treatment plants

Abstract Wastewater treatment plants (WWTPs) are large electricity and natural gas consumers with untapped potential to recover carbon-neutral biogas and provide energy services for the grid. Techno-economic analysis of emerging energy recovery and management technologies is critical to understanding their commercial viability, but quantifying their energy cost savings potential is stymied by a lack of well curated, nationally representative electricity and natural gas tariff data. We present a dataset of electricity tariffs for the 100 largest WWTPs in the Clean Watershed Needs Survey (CWNS) and natural gas tariffs for the 54 of 100 WWTPs with on-site cogeneration. We manually collected tariffs from each utility’s website and implemented data checks to ensure their validity. The dataset includes facility metadata, electricity tariffs, and natural gas tariffs (where cogeneration is present). Tariffs are current as of November 2021. We provide code for technical validation along with a sample simulation.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2022

Extension and Validation of NEMA-Style Dual-Ring Controller in SUMO

Until recently, SUMO users could not model the behavior of a ring-and-barrier traffic signal via existing signal types, which left North American SUMO users without the direct ability to capture traffic dynamics of their local networks. This work presents the methods, implementation overview, and validation of a 'dual-ring' NEMA style traffic controller which has recently been added to the main SUMO code base. A brief explanation of the 'dual-ring' implementation is also provided as context for those new to this type of traffic controller. The foundation for this work was presented at the SUMO User Conference 2021 by researchers at the US Department of Energy's National Renewable Energy Laboratory, but was not integrated into SUMO code base at the time. Following the initial inclusion of the controller to SUMO, the authors began validation of the SUMO controller against an Econolite software-in-the-loop (SIL) traffic signal controller configured with actual setup parameters from controllers in a real-world three-intersection corridor in Tuscaloosa, Alabama, USA. This paper documents the process of adding new features to the controller code as well as validating their implementation through simulation-based and automated grey-box testing is presented in this paper. Key features such as fully-actuated operation, various timing offset plans, proper next-phase fit algorithms and more, have been added and validated against this SIL system. Though not an exhaustive demonstration of features, this work is intended make more users aware of this extension of SUMO capabilities.

MATHEMATICS AND COMPUTING↗
DOE OSTI2019

Continuous Integration Service at Fermilab

This paper describes the current architecture of Continuous Integration (CI) service developed at Fermilab, encountered successes and difficulties, as well as future development plans. The current experiment code has hundreds of contributors that provide new features, bug fixes, and other improvements. Version control systems help developers to collaborate in contributing software for their experiments, while the CI system helps developers keep their code healthy. The Fermilab CI service allows experiments and projects to test and validate their offline production and analysis code on the supported platforms. It is built on top of Jenkins, designed to be set up from a configuration file that provides implementation for each phase of the CI workflow, and able to validate experiment code through grid jobs. This CI service provides a dashboard for easy access to logs and statistical graphs. Since the CI service has been adopted by Fermilab experiments/projects, it proved to be very useful to intercept issues in their code early on and get them fixed before running it in production. Currently the CI service is in use by the ArgoNeuT, DUNE, g-2, LArIAT, MINERvA, mu2e, NOvA, SBND and uBooNE experiments and by the following projects: ART and LArSoft software suites, GENIE, and Glidein-WMS. The CI service is under active development and planning to support code profiling.

Di benedetto, Vito↗
DOE OSTI2021

Evaluation of gamma-ray transmission through rectangular collimator slits for application in nuclear fuel spectrometry

Gamma-ray spectrometry is widely applied in several science fields, and in particular in non-destructive gamma scanning and gamma emission tomography of irradiated nuclear fuel. Usually, a collimator is used in the experimental setup, to selectively interrogate a region of interest in the fuel. For the optimization of instrument design, as well as for planning measurement campaigns, predictive models for the transmitted gamma-ray intensity through the collimator are needed. Commonly, Monte Carlo Radiation Transport tools are used for accurate prediction of gamma-ray transport, however, the long computation time requirements when used in low-efficiency experimental setups present challenges. Here, the full-energy peak intensity transmitted through a rectangular collimator slit was examined. A uniform planar surface source emitting isotropically was considered, and the rate of photons reaching an ideal counter plane on the opposite side of the collimator was evaluated by analytical integration. To find a closed-form primitive function, some idealizations were required, and thereby parametric models were obtained for the optical field of view, dependent on slit dimensions (length, height and width) and source-to-collimator distance. For contributions from outside the optical field of view, where a closed-form expression cannot be found, fast numerical integral methods were instead used. The results were validated using the Monte Carlo code MCNP6 and show an agreement within three percent for the numerical method. For the analytical method, deviations up to tens of percent were obtained, which is deemed to still be sufficient for instrument design and measurement planning, where often the order of magnitude of the count rate is not a priori known. The method is planned for use in iterative optimization routines in the design of Gamma Emission Tomography devices, as well as for the prediction of gamma spectra obtained in the planning of fuel inspections. An application of the proposed method was demonstrated in spectrum prediction for a short cooling-time fuel rod test from the Halden reactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

System Level Analysis Software Validation with Argonne's THETA Experimental Facility (Final CRADA Report)

Under this CRADA, the Contractor will be generating and collecting sodium-based experimental data with modern instrumentation. The Contractor will work symbiotically with the Participant to generate SAS4A/SASSYS-1 and SAM computational models of the THETA experimental facility for validation of the system-level analysis codes utilized by Oklo, Inc. This effort will be specifically focused on the design basis event space scoped by Chapter 15 of the NRC’s Standard Review Plan (NUREG-0800) with an emphasis on natural circulation and thermal stratification phenomena in a prototypic sodium-cooled fast reactor (SFR).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Modeling and Validation of Irradiation Damage in Ni-based Alloys for Long-Term LWR Applications (Final Report)

As light water reactor (LWR) plant lives are extended, the need for predictive modeling tools for materials degradation increase in order to ensure safe operation and plan for component replacements. The Grizzly code, which is built on the MOOSE multi-physics simulation environment, is being developed for exactly this purpose. To extend Grizzly to include capabilities for modeling Ni-based alloys, we propose a US/UK integrated program to address thermal and irradiation-induced transformations mechanisms of Alloys 690 and 625. Alloy 690 is widely used in existing LWR plants due to its superior stress corrosion cracking (SCC) resistance compared to Alloy 600. Alloy 625 is used in more limited applications but offers the benefits of both high strength (in the aged condition) and corrosion/SCC resistance. Research has shown that both alloys can undergo phase changes due to thermal or irradiation exposure. In the precipitation-hardened condition, Alloy 625 “softens” during neutron irradiation as the strengthening precipitates decompose and metastable precipitates form. However, the nature and rates of these transformations as a function of exposure conditions are not well understood. Similarly, the effects of these thermal and irradiation-induced microstructural changes on mechanical properties require evaluation. The proposed program combines thermal and irradiation experiments, mechanical testing, microstructural characterization using state-of-the art analytical techniques, atomistic modeling, micro-and macro-scale modeling via Grizzly. This multi-pronged approach yields several benefits: 1) atomistic scale models capture the physics of the mechanisms of phase transformations, which can be generalized to other systems, 2) the modeling results (from atomistic to continuum) will be validated experimentally to ensure the predictive aspect, 3) the microstructure/property relationships captured with mechanical testing provides information at the engineering scale that is needed to validate the integration of the proposed new capability into Grizzly

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

APOLLO: a facility-scale differentiable virtual accelerator at Fermilab FAST/IOTA

As the design complexity of modern accelerators grows, there is more interest in using advanced simulations that have fast execution time or yield additional insights like gradients. The FAST/IOTA facility has been working on implementing and experimentally validating an end-to-end digital twin that is both fast and gradient-aware, allowing for rapid prototyping of new software and experiments with minimal beam time costs. Our framework integrates physics and ML codes for linac and ring simulation through a set of generic interfaces between surrogate and physics-based sections. To reproduce device inputs and outputs, system state is exposed as a deterministic event loop in a specialized discrete event simulator architecture. Because Fermilab is undergoing control system transition, several APIs were implemented as final user interfaces - a fully asynchronous EPICS soft IOC, a gRPC-based Data Pool Manager (DPM), and legacy ACNET protocols. We discuss implementation details as well as challenges handling live data assimilation and future plans to extend modelling to main complex proton accelerators like PIPII and Booster.

Kuklev, Nikita [Fermilab]↗
DOE OSTI2025

Qualification of Digitized Legacy Fast Reactor Data

The Integral Fast Reactor (IFR) fuel compatibility test program (1984-1994) included a variety of fuel pin examinations conducted at the Hot Fuel Examination Facility (HFEF) and the Alpha-Gamma Hot Cell Facility (AGHCF). Hard copy data records of these examinations have been recovered, scanned, and preserved in PDF format. Many hard copy records are now qualified in accordance with an NRC-approved Quality Assurance Program Plan (QAPP), and there is an ongoing effort to qualify additional legacy records. This legacy fuel performance data is vital to support design and licensing of fast reactors with validation of state-of-the-art codes and advanced methods for design and analysis. Stakeholders can most easily utilize this data when the PDF scans have been converted into digital data tables. However, qualification of the scanned hard copy data does not qualify the digital data file resulting from the digitization of the data contained in the record; the subject matter expert (SME) must make a review of the digitized data table as well before it can be designated as qualified. This report outlines a peer review process to qualify the digital data file(s), typically in CSV format, corresponding to hard copy records in accordance with the existing QAPP.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Comparison of 3-dimensional dose reconstruction system between fluence-based system and dose measurement-guided system

COMPASS system (IBA Dosimetry, Schwarzenbruck, Germany) and ArcCHECK with 3DVH software (Sun Nuclear Corp., Melbourne, FL) are commercial quasi-3-dimensional (3D) dosimetry arrays. Cross-validation to compare them under the same conditions, such as a treatment plan, allows for clear evaluation of such measurement devices. In this study, we evaluated the accuracy of reconstructed dose distributions from the COMPASS system and ArcCHECK with 3DVH software using Monte Carlo simulation (MC) for multi-leaf collimator (MLC) test patterns and clinical VMAT plans. In a phantom study, ArcCHECK 3DVH showed clear differences from COMPASS, measurement and MC due to the detector resolution and the dose reconstruction method. Especially, ArcCHECK 3DVH showed 7% difference from MC for the heterogeneous phantom. ArcCHECK 3DVH only corrects the 3D dose distribution of treatment planning system (TPS) using ArcCHECK measurement, and therefore the accuracy of ArcCHECK 3DVH depends on TPS. In contrast, COMPASS showed good agreement with MC for all cases. However, the COMPASS system requires many complicated installation procedures such as beam modeling, and appropriate commissioning is needed. In terms of clinical cases, there were no large differences for each QA device. The accuracy of the compass and ArcCHECK 3DVH systems for phantoms and clinical cases was compared. Both systems have advantages and disadvantages for clinical use, and consideration of the operating environment is important. The QA system selection is depending on the purpose and workflow in each hospital.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2022

Improved Verification and Validation Testing and Tools including Nuclear Criticality Safety Applications with the MCNP6.3® Code [Abstract]

A new Python-based framework has been developed to enable a more consistent layout with automatable setup, execution, and documentation of all verification and validation (V&V) test suites previously established for use with the MCNP code. In this paper, the new general framework for the V&V test suites is discussed, including information on all of the current capabilities and plans for future capabilities. For nuclear criticality safety applications, the existing V&V benchmark problems within the criticality, extended criticality, and analytic k-effective test suites have been ported into this new framework. In addition, the status and updates to the Rossi-alpha and subcritical multiplicationtest suites will be discussed. Some V&V results exercising new MCNP6.3 capabilities will be demonstrated.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2022

Extension of the energetic particle transport kick model in TRANSP to multiple fast ion species

Abstract Alfvénic instabilities (AEs) are well known to cause enhanced transport of energetic particles (EPs) in fusion devices. Most studies until now have focused on characterizing and understanding AE stability in single-species plasmas heated by neutral beams (NB), where deuterium is typically used as both main plasma species and NB fuel. As the fusion community moves toward fusion reactors that target burning plasma conditions, such as ITER, the single-species picture breaks down. Burning plasmas, which will use a mix of deuterium and tritium (DT) as main fuel, also feature the presence of several supra-thermal fusion products such as alpha particles, protons, helium isotopes and high-energy tritium ions. This work presents the extension of the EP transport kick model implemented in the TRANSP time-dependent tokamak transport code to study the combined effect of multiple EP species on AE stability and, in turn, the response of different EP species to plasma instabilities in terms of their redistribution and losses. Further validation of the enhanced model is planned based on experimental results expected from the JET DT campaign scheduled for 2021, in preparation for ITER plasmas and beyond.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Test Plan for the Horizontal Dry Cask Simulator

The thermal performance of commercial spent nuclear fuel dry storage casks is evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission(NRC). Canistered dry storage cask systems rely on ventilation between the inner canister and the overpack to convect heat away from the canister to the surrounding environment for both horizontal and vertical configurations. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. While several testing programs have been previously conducted, these earlier validation studies did not integrate all the physics or components important in a modern, horizontal dry cask system. The purpose of the investigation described in this test plan is to produce data sets that can be used to benchmark the codes and best practices presently used to determine cladding temperatures and induced cooling air flows in modern horizontal dry storage systems. The horizontal dry cask simulator(HDCS) has been designed to generate this benchmark data and add to the existing knowledgebase. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure(MAWP)rating of 2,400 kPa at400 °C. An existing electrically heated but otherwise prototypic boiling water reactor(BWR), Incoloy-clad test assembly will be deployed inside of a representative storage basket and canister. An insulated sheet metal enclosure will be used to mimic the thermal properties of the concrete vault enclosure used in a modern horizontal storage system. Radial and axial temperature profiles along with induced cooling air flow will be measured for a wide range of decay powers and representative(and higher)cask pressures using various backfills of helium, argon, or air. The single assembly geometry with well-controlled boundary conditions simplifies computational requirements while preserving relevant physics. The proposed test apparatus integrates all the underlying thermal-hydraulics important to defining the performance of a modern horizontal storage system. These include combined-mode heat transfer from the electrically-heated assembly to the canister walls and the primarily natural-convective heat transfer from the canister to the cooling air flow passing through the horizontal vault enclosure. The objective of the HDCS is not to reproduce the performance of a commercial dry storage system for any given set of operational parameters. Rather ,the objective is to capture the dominant physics in a well-characterized test apparatus. The close coupling between the thermal response of the canister system and the resulting induced cooling air flow rate is of particular importance. While incorporating the best available information based on thermal-hydraulic scaling arguments as well as previous vertical testing, this test plan is subject to changes due to improved understanding or from as built deviations to designs. As-built conditions and actual procedures will be documented in the final test report.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Validation of the Computational Fluid Dynamics Method using the Aboveground Configuration of the Dry Cask Simulator

Dry cask applicants submit their spent nuclear fuel dry storage and transportation cask designs to the U.S. Nuclear Regulatory Commission (NRC) for certification under Title 10 of the Code of Federal Regulations (10 CFR) Part 72, 'Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste'. The NRC staff performs its technical review of these designs in accordance with 10 CFR Part 72 and NUREG-1536, 'Standard Review Plan for Spent Fuel Dry Storage Systems at a General License Facility,' Revision 1, issued July 2010. To ensure that the cask and fuel material temperatures of the dry cask storage system remain within the allowable limits or criteria for normal, off-normal, and accident conditions, the NRC staff performs a thermal review as part of the technical review. Recent applications increasingly have used thermalhydraulic analyses using computational fluid dynamics (CFD) codes (e.g., ANSYS FLUENT) to demonstrate the adequacy of the thermal design. The NRC performs validation studies of the FLUENT CFD code to assist it in making regulatory decisions to ensure adequate protection for storage and transportation casks. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. In efforts to further validate these designs, the NRC in conjunction with the Department of Energy and Sandia National Laboratories (SNL) are conducting an experiment to capture the effects of elevated helium pressures and accurately portray the external convection of aboveground and belowground canister dry cask systems.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Feasibility of a multigroup Boltzmann–Fokker–Planck solution for electron beam dose calculations

Legacy nuclear-reactor Boltzmann solvers start clinical deployment as an alternative to Monte Carlo (MC) codes and Fermi–Eyges semiemprical models in radiation oncology treatment planning. Today’s certified clinical solvers are limited to photon beams. In this paper, ELECTR, a state-of-the-art multigroup electron cross sections generation module in NJOY is presented and validated against Lockwood’s calorimetric measurements, EGS-nrc and GEANT-4 for 1–20 MeV unidirectional electron beams. The nuclear-reactor DRAGON-5 solver is upgraded to access the library and solve the Boltzmann–Fokker–Planck (BFP) equation. A variety of heterogeneous radiotherapy and radiosurgery phantom configurations were used for validation purpose. Case studies include a thorax benchmark, that of a typical breast Intra-Operative Radiotherapy and a high-heterogeneity patient-like benchmark. For all beams, 100% of the water voxels satisfied the American Association of Physicists in Medicine accuracy criterion for a BFP-MC dose error below 2%. At least, 97.0% of adipose, muscle, bone, lung, tumor and breast voxels satisfied the 2% criterion. The average BFP-MC relative error was about 0.56% for all voxels, beams and materials combined. By irradiating homogeneous slabs from Z = 1 (hydrogen) to Z = 99 (einsteinium), we reported performance and defects of the CEPXS mode [US. Sandia National Lab., SAND-89-1685] in ELECTR for the entire periodic table. For all Lockwood’s benchmarks, NJOY-DRAGON dose predictions are within the experimental data precision for 98% of voxels.

42 ENGINEERING↗
DOE OSTI2016

Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). In conclusion, this paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

IER 489: Critical Unresolved Region Integral Experiment (CURIE) CEDT Phase-2 Final Design

The final design for the CURIE integral benchmark experiment is presented in this report. Measurements will be performed on the Comet vertical assembly machine and include various critical configurations consisting of High Enriched Uranium (HEU) metal plates and Teflon moderator plates surrounded by a thick copper reflector. The various configurations with different Teflon plate thickness planned allow for complimentary measurements with differing nuclear data sensitivity in the Unresolved Resonance Region (URR). Similar to previously performed Zeus experiments, the proposed work will serve to validate vital nuclear data quantities and uncertainties and identify deficiencies in current radiation transport codes. Additionally, it will assess the applicability of optimization approaches to nuclear critical experiment design. The final experiment design is detailed in this report and includes design, optimization, uncertainty and continuous-energy nuclear data sensitivity analysis for the final experimental configurations that will provide the basis for a future benchmark evaluation intended for submission to the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook [1]. This document discusses the method, approach, considerations and final design chosen with conclusions. It also includes cost and schedule. The basis and principles of this document come from the preliminary design report [2]. Section 2 discusses the content and potential impact; section 3 gives an overview of the nuclear data and unresolved resonance region; section 4 gives a summary of the work done during the preliminary design; section 5 discusses the machine learning method and principles used for final design; section 6 summaries the results from final design; section 7 provides conclusions and the range of configurations chosen to carry forward to experiment execution; and section 8 includes a summary of costs.

42 ENGINEERING↗
DOE OSTI2016

Plasma Interactions with Mixed Materials and Impurity Transport

The project brings together three discipline areas at LLNL to develop advanced capability to predict the impact of plasma/material interactions (PMI) on metallic surfaces in magnetic fusion energy (MFE) devices. These areas are (1) modeling transport of wall impurity ions through the edge plasma to the core plasma, (2) construction of a laser blow-off (LBO) system for injecting precise amounts of metallic atoms into a tokamak plasma, and (3) material science analysis of fundamental processes that modify metallic surfaces during plasma bombardment. The focus is on tungsten (W), which is being used for the ITER divertor and in designs of future MFE devices. In area (1), we have worked with the University of California, San Diego (UCSD) on applications of the UEDGE/DUSTT coupled codes to predict the influx of impurity ions from W dust through the edge plasma, including periodic edge-plasma oscillations, and revived a parallel version of UEDGE to speed up these simulations. In addition, the impurity transport model in the 2D UEDGE code has been implemented into the 3D BOUT++ turbulence/transport code to allow fundamental analysis of the impact of strong plasma turbulence on the impurity transport. In area (2), construction and testing of the LBO injection system has been completed. The original plan to install the LBO on the National Spherical Torus Experiment Upgrade (NSTX-U) at Princeton and its use to validate the impurity transport simulations is delayed owing to NSTX-U being offline for substantial magnetic coil repair period. In area (3), an analytic model has been developed to explain the growth of W tendrils (or fuzz) observed for helium-containing plasmas. Molecular dynamics calculations of W sputtering by W and deuterium (D) ions shows that a spatial blending of interatomic potentials is needed to describe the near-surface and deeper regions of the material.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Verification of the Two-Fluid Model in the Spectral Element Code Nek-2P for Adiabatic Flow Conditions

The simulation of multiphase flow is a difficult physical, computational, and numerical problem. Sophisticated tools are necessary to provide results encompassing a wide range of physical phenomena. The development of these tools must be supported by rigorous analysis and ongoing efforts of verification and validation to ensure the accuracy of simulations. The Nek-2P multiphase code is being developed at Argonne with the general goal of simulating a wide range of multiphase flows and the primary goal of simulating boiling flows in LWR fuel assemblies. As part of this, there is an ongoing effort to verify the code and validate the underlying physical models. The objective of this work is to present recent results in the verification of Nek-2P. As a starting point, this work is focused on adiabatic flows for the sake of simplicity with the immediate next planned step to expand the effort to include heat and mass transfer. Details of the verification of the implemented mass fraction transport, momentum conservation, and pressure equations in Nek-2P are presented. Results of simulations are compared to analytic solutions when possible, however due to the complexity of multiphase flows, these are seldom available. To overcome this difficulty, an independent model is used to provide solutions for comparison.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗