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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 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

RELAP5-3D Assessment Using High Temperature Test Facility Test PG-22

The RELAP5-3D computer code has been assessed using data from Test PG 22 in Oregon State University’s High Temperature Test Facility. Test PG 22 investigated the gas exchange between the primary coolant system helium and the reactor cavity simulation tank nitrogen following a heatup from ambient temperature. The heatup was terminated when the coolant temperature rise across the core reached 125°C. The code generally under predicted coolant and structure temperatures, likely the result of under predicting the core temperature and stored energy during the initial heatup, although there were no measured values to compare with. A full-loop natural circulation flow that occurred in the test was not predicted in any of the simulations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

NJOY21: Making NJOY Suitable for the 21. Century - 14409

NJOY has a long history of nuclear data processing. NJOY began as a successor to MINX (Multigroup Interpretation of Nuclear X-Sections), a Los Alamos code and was first publicly released in the summer of 1977. The initial version included modules such as RECONR for resonance reconstruction on a union energy grid, BROADR for Doppler broadening plus a new Multigroup averaging program, GROUPR and an associated output module, DTR. Soon thereafter the HEATR, ERRORR, and THERMR modules were added to NJOY. By the mid 1980's, NJOY had a worldwide user community. The current version of NJOY2012 was released in December 2012. NJOY2012 builds on the success of more than 45 years of nuclear data processing at Los Alamos National Laboratory. NJOY2012 was rewritten using Fortran 90/95, utilizing among other things the concepts of public and private variables, eliminating common blocks in lieu of Fortran modules and taking advantage of allocatable arrays. A new version of NJOY is being developed at Los Alamos National Laboratory; NJOY21: NJOY for the 21. Century. NJOY21 uses modern programming standards of quality, reliability, and maintainability. It is being designed from the ground up to utilize modern methods of interacting with computer codes by providing an interactive and scripting (e.g., Python) interface. This will allow scientists the ability to more easily work with the results of the various steps in the processing of nuclear data. NJOY21 will continue the legacy of NJOY by providing a trusted and verified processing code that is fast, flexible, easy to use and customize and capable of handling both modern and established nuclear data formats. This paper introduces NJOY21 by first relating the history of NJOY; the majority of the paper will discuss NJOY21. NJOY21 will take advantage of the many decades of nuclear data processing experience that is contained in NJOY2012 and previous versions of NJOY and extend the capabilities and convenience of processing nuclear data. We will discuss how NJOY21 will remain backwards compatible with NJOY2012 - i.e., how current NJOY2012 input decks will continue to work with NJOY21. We will also discuss the verification and validation of the NJOY21 code base. Finally we will discuss new features in NJOY21 that do not exist in NJOY2012. These features include changes in how a scientist can interact with NJOY to process and verify and validate nuclear data; and the ability to read and write new nuclear data formats. (authors)

42 ENGINEERING↗
DOE OSTI2019

Multiple Phase Screen Scintillation Modeling Code

This document describes the underlying theory and use of a computer code written to simulate scintillation effects on propagating electromagnetic (EM) signals. The Multiple Phase Screen (MPS) technique [1, 2, 3, 4] is employed to model the effects of scintillation on EM signals as they travel through the earth’s ionosphere.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2013

Calculation of the dose load during dismantling of large components in the process of decommissioning of nuclear installations

The paper is focused on the decommissioning process from the point of view of steam generator dismantling. A brief description of the steam generator (used in nuclear power plants with VVER 440 type reactor) and used computer code VISIPLAN 3D ALARA are given. The main part deals with the created model and dismantling strategy together with variable input parameters - decay time and decontamination. The obtained results - external exposure of workers and the influence of time and pre-dismantling decontamination - are studied. Also detailed analyses of every dismantling step considered are presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Multi-objective optimization of a PCHE-type intermediate heat exchanger using genetic algorithms

For the advanced nuclear reactor systems and power conversion units, it is important to optimize the design of intermediate heat exchangers (IHXs) which transfer heat from the primary loop to the secondary or intermediate loop. For a multi-objective optimization problem, a printed circuit heat exchanger-type IHX with zigzag and S-shaped fin channels is selected to be optimized in terms of both the total cost and the heat exchanger thermal effectiveness using genetic algorithms. A computer code is developed to search a set of optimized designs of the zigzag channel geometry. It is found that among the Pareto-optimal solutions, the total cost is approximately linearly related to the thermal effectiveness between 88 and 95% while it rises rapidly after the effectiveness exceeds 95%. An optimal flow channel diameter is also determined through the optimization. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Theoretical Prediction of Structural, Electronic, and Magnetic Properties of Rh{sub 2}YIn (Y = Nd, Sm) Full-Heusler Alloys

We have studied the structural, electronic, and magnetic properties of Rh{sub 2}YIn (Y = Nd, Sm) alloys based on rare earth elements, using the full-potential linearized augmented plane-wave method (FP-LAPW) implemented in the Wien2K computer code with the parameterization of the generalized gradient approximation (GGA) and the local density approximation (LDA). Our results reveal a metallic behavior for rare earth-based full-Heusler Rh{sub 2}YIn (Y = Nd, Sm) alloys. Indeed, the structure of the state density illustrates a polarization for the up orientation (metal character) and a pseudo gap at the down orientation.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2015

RADIOACTIVE POSITRON EMITTER PRODUCTION BY ENERGETIC ALPHA PARTICLES IN SOLAR FLARES

Measurements of the 0.511 MeV positron-annihilation line from solar flares are used to explore the flare process in general and ion acceleration in particular. In flares, positrons are produced primarily by the decay of radioactive positron-emitting isotopes resulting from nuclear interactions of flare-accelerated ions with ambient solar material. Kozlovsky et al. provided ion-energy-dependent production cross sections for 67 positron emitters evaluated from their threshold energies (some <1 MeV nucleon{sup –1}) to a GeV nucleon{sup –1}, incorporating them into a computer code for calculating positron-emitter production. Adequate cross-section measurements were available for proton reactions, but not for α-particle reactions where only crude estimates were possible. Here we re-evaluate the α-particle cross sections using new measurements and nuclear reaction codes. In typical large gamma-ray line flares, proton reactions dominate positron production, but α-particle reactions will dominate for steeper accelerated-ion spectra because of their relatively low threshold energies. With the accelerated-{sup 3}He reactions added previously, the code is now reliable for calculating positron production from any distribution of accelerated-ion energies, not just those of typical flares. We have made the code available in the online version of the Journal. We investigate which reactions, projectiles, and ion energies contribute to positron production. We calculate ratios of the annihilation-line fluence to fluences of other gamma-ray lines. Such ratios can be used in interpreting flare data and in determining which nuclear radiation is most sensitive for revealing acceleration of low-energy ions at the Sun.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Impact of Dutch nuclear fuel cycle scenarios on radionuclide inventory intended for disposal - 15201

One of the main subjects of Dutch five-year research program into geological disposal of radioactive waste, OPERA, is the post-closure radiological safety assessment. Reliable estimates of the radionuclide inventory and matrix composition are an important input for the safety assessment. The amounts and categories of waste foreseen to be disposed in 2130 the Netherlands have been compiled by the Dutch waste management organisation COVRA. Additionally, NRG has made a characterisation of the waste (radionuclides and matrix). This reference inventory is based on the Dutch base nuclear energy scenario of operating the Dutch single reactor, the Borssele NPP, until its presently scheduled shut-down in 2033. However, changes in the presently adopted nuclear fuel cycle strategy in the Netherlands may impact both the quantities of generated radioactive waste and its composition, and therefore the source term of the radiological safety assessment. To get a grip on this uncertainty a set of alternative future fuel cycle scenarios in the Netherlands, which are in compliance with scenarios formulated in the Dutch vision documents, has been considered and evaluated. The scenarios have been analysed with the computer code DANESS, 'Dynamic Analysis of Nuclear Energy System Strategies'. The result of the DANESS analyses concerns the amount and type of the radioactive waste to be finally disposed in a deep geological repository. This also includes the inventories of the individual radionuclides that are relevant for the post-closure safety assessment to be performed in a subsequent stage of the OPERA program. (authors)

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

An assessment of total depletion uncertainties for a GBC-32 dry storage cask with PLUS7 fuels using Monte Carlo sampling and two-way ANOVA F-test

It is important to accurately evaluate depletion uncertainties that exist due to the limitations of a computer code on accuracy in a nuclear depletion calculation for criticality safety analysis. In this work, the bias and bias uncertainty in the effective neutron multiplication factor (k{sub eff}) for the generic 32 PWR-assembly burnup credit (GBC-32) cask system were assessed using the Monte Carlo uncertainty sampling method for three different burnups and the isotopic biases and uncertainties obtained with the calculated and measured isotopic concentrations. The PLUS7 used nuclear fuel assemblies discharged from the Hanbit Nuclear Power Plant Unit 3 were assumed to be loaded into the GBC-32 cask. In addition, the statistical F-test in the two-way analysis of variance (ANOVA) was suggested to efficiently estimate the statistically converged total depletion uncertainty in k{sub eff}, with a significant reduction of computing efforts in the Monte Carlo criticality calculations, and it was applied to the GBC-32 cask system after a sensitivity analysis was performed to select the important nuclides having significant contributions to the total depletion uncertainty. With the statistical F-tests in the two-way ANOVA, 100 criticality calculations were needed to get a statistically converged total depletion uncertainty for 10 and 30 GWD/MTU, while 120 criticality calculations were needed for 50 GWD/MTU. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Investigations into validation of plutonium solutions for criticality safety analysis

Plutonium solution chemistry is among the most complex of all the elements in the periodic table. Under acidic conditions up to four different oxidation states can be present in significant amounts in a single solution. From the perspective of nuclear criticality safety, accurate information regarding the concentration of plutonium and other constituents in solution must be well-known in order to correctly model plutonium solution systems for analysis. This requires knowledge of the solution system along with the ability to make accurate assumptions and predictions of the chemical and physical properties of the solution system. If neutron transport methods are used to assess subcriticality in a nuclear criticality safety analysis, the models must be compared with benchmark experiments which are similar and the suitability of the computer code system must be quantified. This paper describes some investigations undertaken to support the validation of plutonium solution systems for nuclear criticality safety analysis and how the Whisper sensitivity/uncertainty (S/U) tool can provide information to support validation of such systems. Case study results for various solution systems such as aqueous plutonium nitrate, aqueous plutonium chloride and hypothetical metal-water mixture are presented. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Flow Field and Pressure Measurements in a 61-Pin Wire-Wrapped Bundle

Texas A and M University is conducting isothermal flow tests in a wire-wrapped, 61-pin hexagonal fuel bundle to support the research on advanced nuclear fuel development sponsored by the Department of Energy (DOE). The purpose of these tests is to produce high spatial- and temporal-resolution measurements of the flow and pressure measurements at different locations of the test bundle. The data resulting from these tests will be used to benchmark numerical simulations of the experiments using both industry-standard Reynolds-averaged Navier-Stokes (RANS) based Computation Fluid Dynamics (CFD) methods and a highly resolved Large Eddy Simulation (LES) computer code. This article described the main features of the experimental facility, its capabilities in producing experimental data of pressure drop and velocity fields at different axial locations and within different exterior and interior sub-channels, using advanced flow visualization techniques. Examples of experimental data outputs and typical locations of the measurements that will included in the research scope are also shown in this article. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Evaluation of the Inclusive Electron Scattering Observables in the Resonance Region from the Experimental Data

The method for evaluation of inclusive electron scattering observables in the resonance region from the experimental data is presented. The computer codes in the web page allow user to evaluate inclusive electron scattering, virtual photon-proton cross sections and the F{sub 1} and F{sub 2} structure functions at the invariant masses of the final hadrons 1.0 < W < 4.0 GeV and at photon virtualities 0.5 < Q {sup 2} < 7.0 GeV{sup 2}. The developed approach is of interest for analyses of the experimental data on exploration of the hadron structure in the experiments with electromagnetic probes.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2016

RISMC Industry Application no. 1 (ECCS/LOCA) - Core Characterization Automation: Lattice Codes Interface for PHISICS/RELAP5-3D

The U.S. NRC is currently proposing rulemaking '10 CFR 50.46c' to revise the loss-of-coolant-accident (LOCA) and emergency core cooling system (ECCS) acceptance criteria to include the effects of higher burn-up on cladding performance. The key implications of this proposition are that the cladding performance cannot be evaluated in isolation anymore. Both cladding and ECCS performance need to be considered in a coupled manner. This also implies that models for cladding performance as well as LOCA methodologies will need to be updated for regulatory purposes. Furthermore, efforts need to be made to find the limiting case within the regulatory input space. The Risk-Informed Safety Margin Characterization (RISMC) Pathway, as part of the DOE Light Water Reactor Sustainability (LWRS) research and development program initialed several programs at the Idaho National Laboratory (INL) to support the industry in the transition to the new rule and to offer potential solutions to this LOCA/ECCS analysis. One of these programs is the Industry Application no. 1 (IA1). Its purpose is to provide to the plant owner/operator a vehicle to inform decisions to manage margins related to compliance with the proposed 10 CFR 50.46c rule. IA1's goal is to develop an Integrated Evaluation Model (IEM) to revisit how uncertainties are propagated across the physical disciplines and data involved, as well as well as how risks are evaluated in a LOCA under 10 CFR 50.46c. This IEM is called LOCA Tool-kit for the US (LOTUS) and it connects five major disciplines involved in LOCA analysis. The focus of LOTUS is to establish the automation interfaces among the five disciplines. LOTUS will utilize in a first step existing computer codes. The risk-informed margins management approach for IA1 will provide a means of quantifying the impact on the key LOCA analysis figures-of-merit like peak cladding temperature (PCT), equivalent cladding reacted (ECR), etc. of a change in LOCA analysis inputs. The information that the risk analysis and associated tools provide can then be used for decision-making and margin management. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Simplex-stochastic collocation method with improved scalability

The Simplex-Stochastic Collocation (SSC) method is a robust tool used to propagate uncertain input distributions through a computer code. However, it becomes prohibitively expensive for problems with dimensions higher than 5. The main purpose of this paper is to identify bottlenecks, and to improve upon this bad scalability. In order to do so, we propose an alternative interpolation stencil technique based upon the Set-Covering problem, and we integrate the SSC method in the High-Dimensional Model-Reduction framework. In addition, we address the issue of ill-conditioned sample matrices, and we present an analytical map to facilitate uniformly-distributed simplex sampling.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2016

The Neutrino: A Better Understanding Through Astrophysics: Final Report

The final report for the award "The Neutrino: A Better Understanding Through Astrophysics" is given. The goals of the work were the following: to construct new theoretical approaches to the problem of neutrino propagation in media including where neutrino-neutrino interactions are important; to pioneer the use of new approaches, including super-scattering operators, for the evolution of neutrino thermal and statistical ensembles; to implement these new approaches in computer codes to study neutrino evolution in supernovae and other hot, dense environments; to increase the realism of simulated signals of a Galactic supernovae neutrino burst in current and future neutrino detectors; to study the simulated signals to determine the ability to extract information on the missing neutrino mixing parameters and the dynamics of the supernova explosion; and to study sterile neutrinos and non-standard interactions of neutrinos in supernovae and their effect upon the signal. Accomplishments made in these areas are described.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2018

RELAP5-3D Pre-test Prediction for High Temperature Test Facility Test PG-26

The RELAP5-3D computer code has been used to perform a pre-test prediction for Test PG-26 in the High Temperature Test Facility at Oregon State University. Test PG-26 will simulate a pressurized conduction cooldown transient following a trip of the primary coolant system gas circulator. The facility is initialized with a steady-state power of 700 kW and prototypic core inlet and outlet helium coolant temperatures of 259 and 687°C, respectively. Sensitivity calculations addressed the flow coastdown time, the reactor cavity cooling system flow rate and temperature, and decay power.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Particle Resuspension Simulation Capability to Substantiate DOE-HDBK-3010 Data

The data provided in the U.S. Department of Energy (DOE) Handbook, DOE-HDBK-3010 provides data such as airborne release fractions and respirable fractions which are widely used in safety analysis and facility design. These fractions are often used as conservative bounds, despite the fact that the original data may have been obtained at a different scale or in a different configuration than the end application. The use of computational techniques to augment these data can be used to extend laboratory-scale data applicability to facility-scale decisions. Our prior report described an approach using Sandia National Laboratories (SNL) computer codes to simulate particle releases from various accident scenarios. The relevant physics controlling the particle resuspension phenomena can be highly problem-dependent. They can include surface tension for particles or surfaces with liquid present, van der Waals forces for small particles, electrostatic forces for charged particles, and mechanical interactions with surface features larger than the particles sizes. The complexity of this phenomenon and recent developments have been described in a recent review by Henry and Minier. (authors)

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2013

Converged accelerated finite difference scheme for the multigroup neutron diffusion equation

Computer codes involving neutron transport theory for nuclear engineering applications always require verification to assess improvement. Generally, analytical and semi-analytical benchmarks are desirable, since they are capable of high precision solutions to provide accurate standards of comparison. However, these benchmarks often involve relatively simple problems, usually assuming a certain degree of abstract modeling. In the present work, we show how semi-analytical equivalent benchmarks can be numerically generated using convergence acceleration. Specifically, we investigate the error behavior of a 1D spatial finite difference scheme for the multigroup (MG) steady-state neutron diffusion equation in plane geometry. Since solutions depending on subsequent discretization can be envisioned as terms of an infinite sequence converging to the true solution, extrapolation methods can accelerate an iterative process to obtain the limit before numerical instability sets in. The obtained results have been compared to the analytical solution to the 1D multigroup diffusion equation when available, using FORTRAN as the computational language. Finally, a slowing down problem has been solved using a cascading source update, showing how a finite difference scheme performs for ultra-fine groups (104 groups) in a reasonable computational time using convergence acceleration. (authors)

97 MATHEMATICS AND COMPUTING↗