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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 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

MeV per Nucleon Ion Irradiation of Nuclear Materials with High Energy Synchrotron X-ray Characterization

The combination of MeV/Nucleon ion irradiation (e.g. 133 MeV Xe) and high energy synchrotron x-ray characterization (e.g. at the Argonne Advanced Photon Source, APS) provides a powerful characterization method to understand radiation effects and to rapidly screen materials for the nuclear reactor environment. Ions in this energy range penetrate ~10 μm into materials. Over this range, the physical interactions vary (electronic stopping, nuclear stopping and added interstitials). Spatially specific x-ray (and TEM and nanoindentation) analysis allow individual quantification of these various effects. Hard x-rays provide the penetration depth needed to analyze even nuclear fuels. Here, this combination of synchrotron x-ray and MeV/Nucleon ion irradiation is demonstrated on U-Mo fuels. A preliminary look at HT-9 steels is also presented. We suggest that a hard x-ray facility with in situ MeV/nucleon irradiation capability would substantially accelerate the rate of discovery for extreme materials.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

A boundary value “reservoir problem” and boundary conditions for multi-moment multifluid simulations of sheaths

Multifluid simulations of plasma sheaths are increasingly used to model a wide variety of problems in plasma physics ranging from global magnetospheric flows around celestial bodies to plasma–wall interactions in thrusters and fusion devices. For multifluid problems, accurate boundary conditions to model an absorbing wall that resolves a classical sheath remain an open research area. This work justifies the use of vacuum boundary conditions for absorbing walls to show comparable accuracy between a multifluid sheath and lower moments of a continuum-kinetic sheath.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Partial flocculation for spray drying of spherical mixed metal oxide particles

Abstract Particles are an important design consideration for many proposed solar thermochemical water splitting reactor concepts. These particles must maintain physical integrity and chemical performance at very high temperatures (>900°C) while flowing within a system. Spray drying is a commonly used, scalable method to produce solid spherical particles that have the best opportunity to meet these requirements. Spray drying is used to produce hercynite (FeAl 2 O 4 ) particles using pH ‐modification of a charge‐stabilized sol. This process involves inducing partial flocculation in nanoparticle suspensions to yield larger particles that are spherical and structurally robust. A commercial boehmite suspension is characterized along with a mixture of the boehmite suspension with an iron oxide suspension. Zeta potential near zero and peak yield stress measurements at the same pH indicate flocculation, but associated sedimentation behavior is not observed. The behavior of both the boehmite‐only and mixed suspensions both seem to be driven by the behavior of the boehmite suspension alone. Spray dried particles from the mixed suspension at the critical pH gave significantly larger and more spherical particles, which also indicates flocculation. This relatively simple technique shows promise for new mixed‐metal material formulations and scalable applications.

Ehrhart, Brian D.↗
DOE OSTI2016

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↗
DOE OSTI2016

MCNP Progress for the Nuclear Criticality Safety Program

The DOE/NNSA Nuclear Criticality Safety Program (NCSP) supports research, development, maintenance, and verification and validation of the MCNP{sup R} Monte Carlo radiation transport software package for nuclear criticality safety (NCS) customers within DOE/NNSA. The MCNP Monte Carlo code has been used for high-fidelity analyses of criticality safety problems since the 1970's. This paper summarizes MCNP progress during FY 2015 and early FY 2016. Activities and accomplishments are summarized in five major areas: - MCNP6 status, - Whisper status, - Verification and validation testing, - User support and training, - Work in progress. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

A data-driven framework for error estimation and mesh-model optimization in system-level thermal-hydraulic simulation

We report over the past decades, several computer codes have been developed for simulation and analysis of thermal-hydraulics and system response in nuclear reactors under operating, abnormal transient, and accident conditions. However, simulation errors and uncertainties still inevitably exist even while these codes have been extensively assessed and used. In this work, a data-driven framework (Optimal Mesh/Model Information System, OMIS) is formulated and demonstrated to estimate simulation error and suggest optimal selection of computational mesh size (i.e., nodalization) and constitutive correlations (e.g., wall functions and turbulence models) for low-fidelity, coarse-mesh thermal-hydraulic simulation, in order to achieve accuracy comparable to that of high-fidelity simulation. Using results from high-fidelity simulations and experimental data with many fast-running low-fidelity simulations, an error database is built and used to train a machine learning model that can determine the relationship between local simulation error and local physical features. This machine learning model is then used to generate insight and help correct low-fidelity simulations for similar physical conditions. The OMIS framework is designed as a modularized six-step procedure and accomplished with state-of-the-art methods and algorithms. A mixed-convection case study was performed to illustrate the entire framework.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

SELF-LIMITING TRANSIENT PULSE SIMULATION METHOD EXHIBITING TIME LAG PHENOMENON USING MAMMOTH

Coupling of multi-physics and multi-scale allows the exploration of phenomenon not previously studied with high fidelity. The heat diffusion time lag in TREAT-like fuel requires both multi-physics and multi-scale coupling to model the full transient behavior. The reactor physics code MAMMOTH, which has the ability to couple the heat conduction equation and the neutron diffusion equation, simulated a self-limiting transient pulse meeting these requirements. The MOOSE MultiApp system allowed the coupling of many micro-scale simulations with a single macro-scale simulation. The micro-scale geometry was a 0.0044 cm diameter UO2 fuel grain surrounded by a graphite moderator. The macro-scale geometry was a homogeneous fuel with the same isotope ratio of the micro-scale material surrounded by a graphite reflector. Two temperature-dependent two energy group cross-section libraries were computed using Serpent 2 for use in MAMMOTH. Both the homogeneous unit cell and heterogeneous unit cell cross-section library had infinite reactor boundary conditions. The macro-scale simulation had finite boundary conditions. By varying the feedback temperature, macro or micro, given to the cross-sections, the effect of the time lag was explored using a Picard coupling method. The time lag increases the peak power density by 9 % to 10 % and the energy deposited by 8 % to 9 %. The effect of differing grain and moderator temperature at the micro-scale was also explored showing differences in time lag behavior. Future research will improve the fidelity of the method through representative macro-scale geometries, temperature-dependent cross-sections, and greater energy group resolution.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Boron-Coated Straw Collar for Uranium Neutron Coincidence Collar Replacement

The objective of this project was to design and optimize, in simulation space, an active neutron coincidence counter (or collar) using boron-coated straws (BCSs) as a non- 3 He replacement to the Uranium Neutron Coincidence Collar (UNCL). UNCL has been used by the International Atomic Energy Agency (IAEA) and European Atomic Energy Community (Euratom) since the 1980s to verify the 235 U content in fresh light water reactor fuel assemblies for safeguards purposes. This report documents the design and optimization of the BCS collar.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Temperature compensated transfer function for probability of detection (POD) in high temperature ultrasonic NDE using low temperature signals - 249

Ultrasonic piezoelectric based transducers offer the potential for in-coolant NDT measurements and imaging required at high temperature for the safe operation of advanced small modular reactors (SMRs). Current work estimates a significant 97% reduction in the probability of detection (POD) for an artificial flaw size of 2{lambda} due to degradation of the transducer performance at elevated temperature. The source of variability for POD in the current work is the temperature dependence of material matrix of piezoelectric material PZT-5A used in the sensor. The physics based model developed using finite element method quantifies reduction in the scattering amplitude for the side drilled hole (SDH) from 15 C to 195 C. Using model assisted probability of detection (MAPOD) approach, sensor temperature effect on the POD is demonstrated for SDHs in the low carbon steel specimen. Scattering amplitudes for the room temperature simulated data is validated with the experimental data. A temperature correction factor and transfer factors are reported to transform the simulated temperature effect from physics based model to the experimental data for estimation of high temperature POD using low temperature ultrasonic measurements. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Prediction on Underestimation of Statistical Uncertainty in Monte Carlo Eigenvalue Calculation for Two-dimensional Heterogeneous Color Set Geometry

In a Monte Carlo calculation, statistical uncertainty is inevitably included in the result. The statistical uncertainty is commonly estimated by the central limit theorem without considering correlation among generations. The statistical uncertainty is usually expected to decrease at the inverse square root of the number of neutron histories. However, in a Monte Carlo eigenvalue calculation in a large geometry whose dominance ratio is close to unity, the central limit theorem would not be valid and thus the statistical uncertainty of fission source distribution can be significantly underestimated when inter-cycle correlation is not appropriately taken into account. The true statistical uncertainty can be calculated by statistical processing of the results obtained by multiple Monte Carlo runs using different initial random seeds, but this approach has a drawback of large calculation cost. Therefore, prediction of true statistical uncertainty by alternate approach is desirable. One of the solutions for this issue is an evaluation method of statistical uncertainty based on the eigenfunction expansion and the Autoregressive model (AR model). In this method, the true statistical uncertainty is theoretically derived by applying the AR model to behavior of expansion coefficients of eigenfunction in fission source distribution during cycles. Underestimation ratio of statistical uncertainty of fission source distribution can be predicted by this method. Note that the underestimation ratio of statistical uncertainty is defined by the ratio of estimated error by a Monte Carlo calculation to the true statistical error. In the previous study, the proposed method was verified in a simple condition such as one energy group, one dimensional homogeneous slab geometry. Contributions of this summary are as follows. Firstly, expression of the noise term that is appeared in the AR model is theoretically derived. Secondly, prediction on underestimation ratio of statistical uncertainty is carried out in a more general case, i.e., flux of each fuel cell in 2 x 2 heterogeneous assembly geometry taken from the C5G7 benchmark problem. Through the present study, applicability of the theoretical prediction for underestimation of statistical uncertainty in practical problems will be verified. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Richard Freeman Post (14 November 1918–7 April 2015)

Richard Freeman Post, a member of the Nuclear Fusion Board of Editors in 1976 and again in the 1990s, passed away April 7, 2015 [1]. He was 96. Dick Post initiated magnetic mirror research in the US and served as Deputy Associate Director for Magnetic Fusion Energy at the Lawrence Livermore National Laboratory. Dick contributed three seminal ideas— yin-yang mirrors, high current neutral beams and a way to stabilize mirror plasmas—that culminated in the first demonstration of a plasma at thermonuclear temperatures, in the 2XIIB experiment in 1976. He also invented a method for direct conversion of plasma energy to electricity as a way to improve the efficiency of mirror fusion reactors. His review papers published in 1956 in Reviews of Modern Physics and in 1959 in Annual Review of Nuclear Science were among the first consistent reviews of hot plasmas for controlled fusion. These influential papers were immediately translated into Russian and published in Russia; the 1959 paper as a separate book. In 1987, Dick published a comprehensive review of mirror research that filled the whole October issue of Nuclear Fusion [2]. Dick also pioneered research on high-speed flywheel energy storage employing novel magnetic bearings that presaged a new approach to magnetically levitated trains. He remained actively engaged in mirror research and in early 2000 came up with an idea of a 'kinetic stabilizer' for axisymmetric tandem mirrors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Initial assessment of clad/base metal systems for MSR

In support of the development of new design and construction rules and analysis methods for cladded structure for use under the anticipated service lifetime and cyclic service conditions in Molten Salt Reactors (MSR), research work has been initiated at ORNL for assessment of physical, metallurgical and mechanical properties of clad/base metal systems. This report summarizes the results for the nickel cladded SS316H system in FY18.

36 MATERIALS SCIENCE↗
DOE OSTI2016

The Need for Precise and Well-documented Experimental Data on Prompt Fission Neutron Spectra from Neutron-induced Fission of {sup 239}Pu

The spectrum of neutrons emitted promptly after {sup 239}Pu(n,f)—a so-called prompt fission neutron spectrum (PFNS)—is a quantity of high interest, for instance, for reactor physics and global security. However, there are only few experimental data sets available that are suitable for evaluations. In addition, some of those data sets differ by more than their 1-σ uncertainty boundaries. We present the results of MCNP studies indicating that these differences are partly caused by underestimated multiple scattering contributions, over-corrected background, and inconsistent deconvolution methods. A detailed uncertainty quantification for suitable experimental data was undertaken including these effects, and test-evaluations were performed with the improved uncertainty information. The test-evaluations illustrate that the inadequately estimated effects and detailed uncertainty quantification have an impact on the evaluated PFNS and associated uncertainties as well as the neutron multiplicity of selected critical assemblies. A summary of data and documentation needs to improve the quality of the experimental database is provided based on the results of simulations and test-evaluations. Given the possibly substantial distortion of the PFNS by multiple scattering and background effects, special care should be taken to reduce these effects in future measurements, e.g., by measuring the {sup 239}Pu PFNS as a ratio to either the {sup 235}U or {sup 252}Cf PFNS.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

SCALE Code System

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.SCALE 6.2 provides many new capabilities and significant improvements of existing features.New capabilities include:• ENDF/B-VII.1 nuclear data libraries CE and MG with enhanced group structures,• Neutron covariance data based on ENDF/B-VII.1 and supplemented with ORNL data,• Covariance data for fission product yields and decay constants,• Stochastic uncertainty and correlation quantification for any SCALE sequence with Sampler,• Parallel calculations with KENO,• Problem-dependent temperature corrections for CE calculations,• CE shielding and criticality accident alarm system analysis with MAVRIC,• CE depletion with TRITON (T5-DEPL/T6-DEPL),• CE sensitivity/uncertainty analysis with TSUNAMI-3D,• Simplified and efficient LWR lattice physics with Polaris,• Large scale detailed spent fuel characterization with ORIGAMI and ORIGAMI Automator,• Advanced fission source convergence acceleration capabilities with Sourcerer,• Nuclear data library generation with AMPX, and• Integrated user interface with Fulcrum.Enhanced capabilities include:• Accurate and efficient CE Monte Carlo methods for eigenvalue and fixed source calculations,• Improved MG resonance self-shielding methodologies and data,• Resonance self-shielding with modernized and efficient XSProc integrated into most sequences,• Accelerated calculations with TRITON/NEWT (generally 4x faster than SCALE 6.1),• Spent fuel characterization with 1470 new reactor-specific libraries for ORIGEN,• Modernization of ORIGEN (Chebyshev Rational Approximation Method [CRAM] solver, API for high-performance depletion, new keyword input format)• Extension of the maximum mixture number to values well beyond the previous limit of 2147 to ~2 billion,• Nuclear data formats enabling the use of more than 999 energy groups,• Updated standard composition library to provide more accurate use of natural abundances, andvi• Numerous other enhancements for improved usability and stability.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2016

Updates to the Generation of Physics Data Inputs for MAMMOTH Simulations of the Transient Reactor Test Facility - FY2016

The INL is currently evolving the modeling and simulation (M&S) capability that will enable improved core operation as well as design and analysis of TREAT experiments. This M&S capability primarily uses MAMMOTH, a reactor physics application being developed under Multi-physics Object Oriented Simulation Environment (MOOSE) framework. MAMMOTH allows the coupling of a number of other MOOSE-based applications. This second year of work has been devoted to the generation of a deterministic reference solution for the full core, the preparation of anisotropic diffusion coefficients, the testing of the SPH equivalence method, and the improvement of the control rod modeling. In addition, this report includes the progress made in the modeling of the M8 core configuration and experiment vehicle since January of this year.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Risk Informed, Performance-Based, Technology-Inclusive Regulatory Infrastructure: Technology-Inclusive Determination of Mechanistic Source Terms for Offsite Dose-Related Assessments for Advanced Nuclear Reactor Facilities

This report summarizes a risk informed, performance-based and technology-inclusive approach to determine source terms for dose related assessments at advanced nuclear facilities. This approach uses a graded process which allows both the non-mechanistic source terms calculation methods, which adopt conservative approaches and assumptions based on known physical and chemical principles, and more importantly the mechanistic source term calculation methods, which consider design-specific scenarios and use best estimate models with uncertainty quantification for a range of licensing basis events (LBEs), to be used for the design and licensing of advanced nuclear technologies. The source terms developed with this graded approach, and radionuclide inventories elsewhere in the facility that are determined during source term analysis, can be used to address licensing issues to support the application processes of 10 CFR Part 50 for a construction permit and operating license or 10 CFR Part 52 for a Combined Operating License (COL), Standard Design Certification, Early Site Permit, Standard Design Approval or Manufacturing License. They can also be used for other purposes including equipment environmental qualification, control room habitability analyses, and assessments of severe accident risks in environmental impact statements. There are many advanced reactor concepts being developed including high temperature gas-cooled reactor (HGTR), sodium-cooled fast reactor (SFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), micro-reactor, etc. The graded approach presented in this report for source terms determination is, to the extent possible, generic to any of these reactor designs and to future reactor designs. This report provides information on the review of the regulatory foundation for use of conservative bounding source terms as well as event-specific mechanistic source terms for advanced nuclear reactor designs.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

High-fidelity simulation-driven model development for coarse-grained computational fluid dynamics

Nuclear reactor safety analysis requires analysis of a broad range of accident scenarios and determining their consequences. For a nuclear power plant behavior, it is impossible to obtain experimental data at sufficiently large scales to support calibration and validation of various system models. In single-phase flow convective problems, high-resolution methods of Computational Fluid Dynamics (CFD) such as Direct Numerical Simulation (DNS) and Large Eddy Simulation (LES) can provide high fidelity results under conditions where physical measurements are unavailable. However, such high-resolution simulations are computationally expensive and not suitable for simulation of long transient scenarios in nuclear reactor accidents. In this work, we investigate the use of a high fidelity simulation-driven approach to model sub-grid scale (SGS) effect in Coarse Grained Computational Fluid Dynamics CG-CFD. Notably, fine-mesh simulations are used to construct physics-informed statistical surrogate of coarse-mesh SGS model. For an initial analysis, we consider a case of turbulent natural convection in a volumetrically heated fluid layer with a thermally insulated lower boundary and isothermal upper boundary. This scenario of unstable stratification is relevant to turbulent natural convection in a severe nuclear reactor accident, as well as in containment mixing and passive cooling. For this type of flow, the SGS effect is modeled by an added turbulent diffusivity in the energy equation. It is shown that a global correction for the energy equation is sufficient to achieve a significant improvement to the CG-CFD prediction of thermal mixing in the fluid layer. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

University Research in Support of TREAT Modeling and Simulation, FY 2016

Idaho National Laboratory is currently evolving the modeling and simulation (M&S) capability that will enable improved core operation as well as design and analysis of TREAT experiments. This M&S capability primarily uses MAMMOTH, a reactor physics application being developed under the Multi-physics Object Oriented Simulation Environment (MOOSE) framework. MAMMOTH allows the coupling of a number of other MOOSE-based applications. In support of this research, INL is working with four universities to explore advanced solution methods that will complement or augment capabilities in MAMMOTH. This report consists of a collection of year end summaries of research from the universities performed in support of TREAT modeling and simulation. This research was led by Prof. Sedat Goluoglu at the University of Florida, Profs. Jim Morel and Jean Ragusa at Texas A&M University, Profs. Benoit Forget and Kord Smith at Massachusetts Institute of Technology, Prof. Leslie Kerby of Idaho State University and Prof. Barry Ganapol of University of Arizona. A significant number of students were supported at various levels though the projects and, for some, also as interns at INL.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗