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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 1,585 records · Page 88

Near Real-Time Nondestructive Active Inspection Technologies Utilizing Delayed γ-Rays and Neutrons for Advanced Safeguards

In this two year project, the research team investigated how delayed γ-rays from short-lived fission fragments detected in the short interval between irradiating pulses can be exploited for advanced safeguards technologies. This program contained experimental and modeling efforts. The experimental effort measured the emitted spectra, time histories and correlations of the delayed γ-rays from aqueous solutions and solid targets containing fissionable isotopes. The modeling effort first developed and benchmarked a hybrid Monte Carlo simulation technique based on these experiments. The benchmarked simulations were then extended to other safeguards scenarios, allowing comparisons to other advanced safeguards technologies and to investigate combined techniques. Ultimately, the experiments demonstrated the possible utility of actively induced delayed γ-ray spectroscopy for fissionable material assay.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Enhancements to the SHARP Build System and NEK5000 Coupling

The SHARP project for the Department of Energy's Nuclear Energy Advanced Modeling and Simulation (NEAMS) program provides a multiphysics framework for coupled simulations of advanced nuclear reactor designs. It provides an overall coupling environment that utilizes custom interfaces to couple existing physics codes through a common spatial decomposition and unique solution transfer component. As of this writing, SHARP couples neutronics, thermal hydraulics, and structural mechanics using PROTEUS, Nek5000, and Diablo respectively. This report details two primary SHARP improvements regarding the Nek5000 and Diablo individual physics codes: (1) an improved Nek5000 coupling interface that lets SHARP achieve a vast increase in overall solution accuracy by manipulating the structure of the internal Nek5000 spatial mesh, and (2) the capability to seamlessly couple structural mechanics calculations into the framework through improvements to the SHARP build system. The Nek5000 coupling interface now uses a barycentric Lagrange interpolation method that takes the vertex-based power and density computed from the PROTEUS neutronics solver and maps it to the user-specified, general-order Nek5000 spectral element mesh. Before this work, SHARP handled this vertex-based solution transfer in an averaging-based manner. SHARP users can now achieve higher levels of accuracy by specifying any arbitrary Nek5000 spectral mesh order. This improvement takes the average percentage error between the PROTEUS power solution and the Nek5000 interpolated result down drastically from over 23 % to just above 2 %, and maintains the correct power profile. We have integrated Diablo into the SHARP build system to facilitate the future coupling of structural mechanics calculations into SHARP. Previously, simulations involving Diablo were done in an iterative manner, requiring a large amount manual work, and left only as a task for advanced users. This report will detail a new Diablo build system that was implemented using GNU Autotools, mirroring much of the current SHARP build system, and easing the use of structural mechanics calculations for end-users of the SHARP multiphysics framework. It lets users easily build and use Diablo as a stand-alone simulation, as well as fully couple with the other SHARP physics modules. The top-level SHARP build system was modified to allow Diablo to hook in directly. New dependency handlers were implemented to let SHARP users easily build the framework with these new simulation capabilities. The remainder of this report will describe this work in full, with a detailed discussion of the overall design philosophy of SHARP, the new solution interpolation method introduced, and the Diablo integration work. We will conclude with a discussion of possible future SHARP improvements that will serve to increase solution accuracy and framework capability.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE F 241.3_DE-EE0006097_FINAL

In this project, Optony deployed the Sustainable Energy for Business Districts (SEBIZ) model to address China’s surging urban energy needs and greenhouse gas emissions via a district-level Energy Efficiency and Renewable Energy (EERE) and smart grid technology deployment strategy. The SEBIZ approach built local capacity and demonstrated the opportunities and benefits for ongoing improvements to and investments in energy infrastructure through market-based project development and an innovative, highly-effective applied public policy development model. Furthermore, the SEBIZ program used real-world examples and resources to formulate appropriate policy, program and educational guidance for similar areas and cities around China and internationally.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

Status Report on NEAMS PROTEUS/ORIGEN Integration

The US Department of Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program has contributed significantly to the development of the PROTEUS neutron transport code at Argonne National Laboratory and to the Oak Ridge Isotope Generation and Depletion Code (ORIGEN) depletion/decay code at Oak Ridge National Laboratory. PROTEUS’s key capability is the efficient and scalable (up to hundreds of thousands of cores) neutron transport solver on general, unstructured, three-dimensional finite-element-type meshes. The scalability and mesh generality enable the transfer of neutron and power distributions to other codes in the NEAMS toolkit for advanced multiphysics analysis. Recently, ORIGEN has received considerable modernization to provide the high-performance depletion/decay capability within the NEAMS toolkit. This work presents a description of the initial integration of ORIGEN in PROTEUS, mainly performed during FY 2015, with minor updates in FY 2016.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Destructive examination of shipping package 9975-02101

Destructive and non-destructive examinations have been performed on the components of shipping package 9975-02101 as part of the comprehensive Model 9975 package surveillance program. This package is one of ten high-wattage packages that were selected for field surveillance in FY15, and was identified to contain several non-conforming conditions. Most of these conditions (mold, stains, drum corrosion, calculated fiberboard dimensions and fiberboard damage) relate to the accumulation of water in the outer and lower portions of the cane fiberboard assembly. In the short term, this causes local but reversible changes in the fiberboard properties. Long-term effects can include the permanent loss of fiberboard properties (thus far observed only in the bottom fiberboard layers) and reduced drum integrity due to corrosion. The observed conditions must be fully evaluated by KAC to ensure the safety function of the package is being maintained. Three of the other nine FY15 high-wattage packages examined in the K-Area Complex showed similar behavior. Corrosion of the overpack drum has been seen primarily in those packages with relatively severe fiberboard degradation. Visual examination of the drums in storage for external corrosion should be considered as a screening tool to identify additional packages with potential fiberboard degradation. Where overpack drum corrosion has been observed, it is typically heaviest adjacent to the stitch welds along the bottom edge. It is possible that changes to the stitch weld design would reduce the degree of corrosion in this area, but would not eliminate it. Several factors can contribute to the concentration of moisture in the fiberboard, including higher than average initial moisture content, higher internal temperature (due to internal heat load and placement with the array of packages), and the creation of additional moisture as the fiberboard begins to degrade.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Multiscale Modeling and Uncertainty Quantification for Nuclear Fuel Performance

In this project, we will address the challenges associated with constructing high fidelity multiscale models of nuclear fuel performance. We (*) propose a novel approach for coupling mesoscale and macroscale models, (*) devise efficient numerical methods for simulating the coupled system, and (*) devise and analyze effective numerical approaches for error and uncertainty quantification for the coupled multiscale system. As an integral part of the project, we will carry out analysis of the effects of upscaling and downscaling, investigate efficient methods for stochastic sensitivity analysis of the individual macroscale and mesoscale models, and carry out a posteriori error analysis for computed results. We will pursue development and implementation of solutions in software used at Idaho National Laboratories on models of interest to the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Distribution Development for STORM Ingestion Input Parameters

The Sandia-developed Transport of Radioactive Materials (STORM) code suite is used as part of the Radioisotope Power System Launch Safety (RPSLS) program to perform statistical modeling of the consequences due to release of radioactive material given a launch accident. As part of this modeling, STORM samples input parameters from probability distributions with some parameters treated as constants. This report described the work done to convert four of these constant inputs (Consumption Rate, Average Crop Yield, Cropland to Landuse Database Ratio, and Crop Uptake Factor) to sampled values. Consumption rate changed from a constant value of 557.68 kg / yr to a normal distribution with a mean of 102.96 kg / yr and a standard deviation of 2.65 kg / yr. Meanwhile, Average Crop Yield changed from a constant value of 3.783 kg edible / m 2 to a normal distribution with a mean of 3.23 kg edible / m 2 and a standard deviation of 0.442 kg edible / m 2 . The Cropland to Landuse Database ratio changed from a constant value of 0.0996 (9.96%) to a normal distribution with a mean value of 0.0312 (3.12%) and a standard deviation of 0.00292 (0.29%). Finally the crop uptake factor changed from a constant value of 6.37e -4 (Bq crop /kg)/(Bq soil /kg) to a lognormal distribution with a geometric mean value of 3.38e -4 (Bq crop /kg)/(Bq soil /kg) and a standard deviation value of 3.33 (Bq crop /kg)/(Bq soil /kg)

60 APPLIED LIFE SCIENCES↗

UFD Proof-of-Concept Demonstration

The BISON Spent Nuclear Fuel modeling capability has been demonstrated through simulations of the complete nuclear fuel lifecycle including in-reactor irradiation, spent fuel pool cooling, vacuum drying and finally dry cask storage. The importance of hydrogen in the cladding and the formation of hydrides at various stages of the fuel lifecycle have been a focus of the initial development work. In addition, the integration in BISON of a hydride damage model has improved the prediction of the cladding mechanical behavior. The work described in this report was performed under funding from the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. This report is issued in satisfaction of the level 2 milestone M2MS-17IN0201012 on UFD Proof-of-Concept Demonstration for the INL Engineering-Scale Fuel Performance (BISON) Workpackage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Graphite Microstructural Characterization Using Time-Domain and Correlation-Based Ultrasonics

Among techniques that have been used to determine elastic modulus in nuclear graphites, ultrasonic methods have enjoyed wide use and standards using contacting piezoelectric tranducers have been developed to ensure repeatability of these types of measurements. However, the use of couplants and the pressures used to effectively couple transducers to samples can bias measurements and produce results that are not wholly related to the properties of the graphite itself. In this work, we have investigated the use of laser ultrasonic methods for making elastic modulus measurements in nuclear graphites. These methods use laser-based transmitters and receivers to gather data and do not require use of ultrasonic couplants or mechanical contact with the sample. As a result, information directly related to the elastic responses of graphite can be gathered even if the graphite is porous, brittle and compliant. In particular, we have demonstrated the use of laser ultrasonics for the determination of both Young’s modulus and shear modulus in a range of nuclear graphites including those that are being considered for use in future nuclear reactors. These results have been analyzed to assess the contributions of porosity and microcracking to the elastic responses of these graphites. Laser-based methods have also been used to assess the moduli of NBG-18 and IG-110 where samples of each grade were oxidized to produce specific changes in porosity. These data were used to develop new models for the elastic responses of nuclear graphites and these models have been used to infer specific changes in graphite microstructure that occur during oxidation that affect elastic modulus. Specifically, we show how ultrasonic measurements in oxidized graphites are consistent with nano/microscale oxidation processes where basal plane edges react more readily than basal plane surfaces. We have also shown the use of laser-based methods to perform shear-wave birefringence measurements and have shown how these measurements can be used to assess elastic anisotropy in nuclear graphites. Using models developed in this program, ultrasonic data were interpreted to extract orientation distribution coefficients that could be used to represent anisotropy in these materials. This demonstration showed the use of ultrasonic methods to quantify anisotropy and how these methods provide more detailed information than do measurements of thermal expansion – a technique commonly used for assessing anisotropy in nuclear graphites. Finally, we have employed laser-based, ultrasonic-correlation techniques in attempts to quantify aspects of graphite microstructure such as pore size and distribution. Results of these measurements indicate that additional work must be performed to make this ultrasonic approach viable for quantitative microstructural characterization.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Warthog Framework Update: Depletion Capabilities

The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program the US Department of Energy (DOE) Office of Nuclear Energy supports two simulation frameworks to provide advanced simulation techniques for analyzing nuclear reactors. Primarily at Idaho National Laboratory (INL), the fuels product line (FPL) is developing the fuel performance code BISON from Hales et al. [2016], which depends on the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework developed by Gaston et al. [2009]. Likewise, at Argonne National Laboratory (ANL), in the reactor product line (RPL), PROTEUS is being developed by Lee et al. [2013] and Nek5000 by Fischer et al. [2015] based on the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) framework from Tautges et al. [2013]. MOOSE employs a bottom-up approach to multiphysics coupling, where component codes are built upon a common framework. SHARP employs a top-down approach in which legacy component codes are connected after being separately developed. Each approach has benefits and disadvantages, but this work focuses on having the two approaches work in harmony.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Atmospheric testbed workshop, August 5-6, 2013

A primary goal of the Climate and Environmental Sciences Division (CESD) is to improve the predictive ability of regional and global climate models (GCMs). One set of tools used for evaluating and improving model predictions and projections are testbeds that combine data derived from observations and model simulations. In these testbeds, outputs of model simulations can be compared to observations of the climate system in order to identify errors and/or biases in the model simulations and determine the specific model processes that need improvements. Testbeds can also be used to provide scientific insights into dominant processes and process interactions as well as increase our understanding of the role of various physical processes involved in a particular case study or meteorological event. CESD currently funds multiple activities, which are developing testbeds to evaluate and improve atmospheric model parameterizations across a range of models, scales, and processes. For each of these testbeds, the Atmospheric Radiation Measurement (ARM) Climate Research Facility represents a primary source of observational data for evaluating the models. The Atmospheric System Research (ASR), Earth System Modeling (ESM), Regional and Global Climate Modeling (RGCM), and ARM program managers co-sponsored a workshop on atmospheric testbed activities across CESD. The workshop was held August 5-6, 2013, at the U. S. Department of Energy (DOE) Germantown facility. The principal objectives of the workshop were to survey the CESD testbed portfolios; identify the distinctive capabilities of each testbed as well as commonalities between the testbeds; identify areas for coordination and collaboration among testbeds; improve linkages between testbed activities and the ARM facility; identify gaps in testbed activities; and identify ways to improve coordination and communication among the testbeds and between the testbeds and the broader scientific community.

54 ENVIRONMENTAL SCIENCES↗

Warthog: Coupling Nek5000 Thermal Hydraulics to BISON Fuel Performance through the Giraffe Interface

The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program under the US Department of Energy (DOE) consists of three product lines. The Fuels Product Line (FPL), hosted at Idaho National Laboratory (INL), has produced a multiphysics framework with the name Multiphysics Object Oriented Simulation Environment (MOOSE) (Gaston et al. 2015). The FPL has also produced several high-fidelity physics codes including the fuel performance code Bison (Hales et al. 2016). The Reactors Product Line (RPL), hosted at Argonne National Laboratory (Argonne), has developed the high-order spectral-element Computational Fluid Dynamics (CFD) code Nek5000 (Fischer et al. 2015). The Integration Product Line (IPL), hosted at Oak Ridge National Laboratory (ORNL), is tasked with ensuring tools from the other product lines can be coupled together so that users may perform high-fidelity multiphysics simulations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Fuel Cycle Technologies 2016 Achievements Report

The U.S. Department of Energy (DOE), Office of Nuclear Energy (DOE- NE) established the Fuel Cycle Technologies (FCT) program to address challenges such as growing stockpiles of used nuclear fuel (UNF) and high-level waste (HLW), and an urgency to enhance accident tolerance of the existing reactor fleet. Resolution of these challenges involves developing systems that reduce waste while improving resource utilization and safety. Towards this goal, the FCT program implemented a research and development (R&D) science-based approach, integratesing theory and experiment with high performance computing modeling and simulation. The FCT program establishes and strengthens collaborations with experts at international agencies, universities, and industry to develop the nuclear fuel cycle of the future. DOE-NE national laboratories and their international counterparts provide extensive expertise and experience and access to critical R&D facilities. Universities provide a wellspring of innovative ideas while industry provides the venue to integrate the expertise, experience, and ideas into commercial practice.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Preparation for a PROTEUS Software Quality Assurance Plan

PROTEUS is the neutron transport solver package developed at Argonne National Laboratory under the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. For the purposes of this report, we consider the PROTEUS package to include the three solvers PROTEUS-NODAL, PROTEUS-MOC, and PROTEUS-SN, as well as the PROTEUS mesh tools and utilities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Performance and Energy Implications for Heterogeneous Computing Systems: A MiniFE Case Study

Heterogeneous computing systems, which employ a mix of general-purpose (GP) processors and accelerators such as graphics processing units (GPUs) or Field Programmable Gate Arrays (FPGAs), have the potential to offer much higher performance and lower energy usage than homogeneous systems. However, designing heterogeneous computing systems to achieve high performance and low energy usage is a challenging task. Designs that offer higher performance do not necessarily lead to lower energy consumption. Furthermore, mapping of applications to different computing devices can play a key role in performance and energy tradeoff. In this report, we present a detailed performance and energy study of executing a specific mini-application on different heterogeneous systems. The results show that hardware choices, application implementations, and mapping of applications to hardware can all significantly impact system performance and energy consumption and that the impact on performance and energy can be quite different. This study forms a basis for modeling the interdependencies of program structures and hardware execution units, which could be used to guide design space exploration.

97 MATHEMATICS AND COMPUTING↗

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↗

Collaborative study on Earthquake Ground Motion Simulation Using Rupture Dynamics

The objective of this multi-year project between the LLNL and Kyoto University, is the investigation of effects of rupture dynamics on near-fault ground motion amplitude for strike slip earthquakes through rupture dynamics simulations. The work presented in this report was in part performed by Jikai Sun, a graduate student from Kyoto university, during his visit to the LLNL as a summer student. He was trained by Dr. Pitarka in using computer programs for generating rupture models and performing rupture dynamics simulations.

58 GEOSCIENCES↗

SAM Code Validation using the Compact Integral Effects Test (CIET) Experimental Data

The primary objective of this study is to validate the system analysis code, SAM, using experimental data from the Compact Integral Effects Test (CIET) facility. The System Analysis Module (SAM) is an advanced and modern system analysis tool being developed at Argonne National Laboratory under the U.S. DOE Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. Originally aiming at supporting sodium fast reactor (SFR) design, it is now extended to cover other non-light water reactor (non-LWR) concepts, such as high-temperature gas-cooled reactor (HTGR) designs, molten salt-fueled, and molten salt-cooled reactor designs. To support SAM code development for the wide range of non-LWR applications, it is of paramount importance to validate the code against experiments relevant to these reactor concepts. The Compact Integral Effects Test (CIET) facility, which was designed to reproduce the thermal-hydraulics response of fluoride salt-cooled high-temperature reactors (FHR) under both forced- and natural-circulation conditions, has been identified and selected as one of benchmark test facilities for SAM code validation. Among available CIET experiments, three sets of them with distinctive characteristics were selected for SAM code validation purpose. The three sets experiments are: 1) Power step change transient tests; 2) DHX-DRACS natural circulation tests; and 3) heater frequency response tests. For all tests, steady-state or transient, SAM-predicted results show very good agreement with experimental data. The successful validation of SAM against these selected CIET experiments demonstrates that the computer code is well suited for thermal-hydraulics analysis of FHR designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗