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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 901 records · Page 50
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Research subjects for analytical estimation of core degradation at Fukushima-Daiichi nuclear power plant

Estimation of the accident progress and status inside the pressure vessels (RPV) and primary containment vessels (PCV) is required for appropriate conductance of decommissioning in the Fukushima-Daiichi NPP. For that, it is necessary to obtain additional experimental data and revised models for the estimation using computer codes with increased accuracies. The Japan Atomic Energy Agency (JAEA) has selected phenomena to be reviewed and developed, considering previously obtained information, conditions specific to the Fukushima-Daiichi NPP accident, and recent progress of experimental and analytical technologies. As a result, research and development items have been picked up in terms of thermal-hydraulic behavior in the RPV and PCV, progression of fuel bundle degradation, failure of the lower head of RPV, and analysis of the accident. This paper introduces the selected phenomena to be reviewed and developed, research plans and recent results from the JAEA's corresponding research programs. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Theoretical oscillator strengths, transition probabilities, and radiative lifetimes of levels in Pb V

Theoretical values of oscillator strengths and transition probabilities for 306 spectral lines arising from the 5d{sup 9}ns(n=7,8,9),5d{sup 9}np(n=6,7),5d{sup 9}6d, and 5d{sup 9} 5f configurations, and radiative lifetimes of 9 levels, of Pb V have been obtained. These values were obtained in intermediate coupling (IC) and using ab initio relativistic Hartree–Fock calculations including core-polarization effects. We use for the IC calculations the standard method of least squares fitting of experimental energy levels by means of computer codes from Cowan. We included in these calculations the 5d{sup 8}6s6p and 5d{sup 8}6s{sup 2} configurations. These calculations have facilitated the identification of the 214.25, 216.79, and 227.66 nm spectral lines of Pb V. In the absence of experimental results of oscillator strengths and transition probabilities, we could not make a direct comparison with our results. However, the Stark broadening parameters calculated from these values are in excellent agreement with experimental widening found in the literature. -- Highlights: •Theoretical values of transition probabilities of Pb V have been obtained. •We use for the IC calculations the standard method of least square. •The parameters calculated from these values are in agreement with the experimental values.

74 ATOMIC AND MOLECULAR PHYSICS↗
DOE OSTI2015

SU-E-P-45: An Analytical Formula for Deriving Mechanical Iso-Center of Rotational Gantry Treatment Unit Rotational Gantry Treatment Unit

Purpose: To present an analytical formula for deriving mechanical isocenter (MIC) of a rotational gantry treatment unit. The input data to the formula is obtained by a custom-made device. The formula has been implemented and used in an operational proton therapy facility since 2005. Methods: The custom made device consisted of 3 mutually perpendicular dial indicators and 5 clinometers, to obtain displacement data and gantry angle data simultaneously. During measurement, a steel sphere was affixed to the patient couch, and the device was attached to the snout rotating with the gantry. The displacement data and angle data were obtained simultaneously at angular increments of less than 1 degree. The analytical formula took the displacement and angle as input and derived the positions of dial indicator tips (DIT) position in room-fixed coordinate system. The formula derivation presupposes trigonometry and 3-dimentional coordinate transformations. Due to the symmetry properties of the defining equations, the DIT position can be solved for analytically without using mathematical approximations. We define the mean of all points in the DIT trajectory as the MIC. The formula was implemented in computer code, which has been employed during acceptance test, commissioning, as well as routine QA practice in an operational proton facility since 2005. Results: It took one minute for the custom-made device to acquire the measurement data for a full gantry rotation. The DIT trajectory and MIS are instantaneously available after the measurement. The MIC Result agrees well with vendor’s Result, which came from a different measurement setup, as well as different data analysis algorithm. Conclusion: An analytical formula for deriving mechanical isocenter was developed and validated. The formula is considered to be absolutely accurate mathematically. Be analyzing measured data of radial displacements as function of gantry angle, the formula calculates the MI position in room coordinate.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2019

Nonlinear Optimization in "Synchrotron Radiation Workshop" Code

"Synchrotron Radiation Workshop" (SRW) is an open source physical optics computer code for calculation of detailed characteristics of Synchrotron Radiation (SR) generated by relativistic electrons in magnetic fields of arbitrary configuration. SRW also supports high-accuracy simulations of fully- and partially-coherent radiation propagation through X-ray optical beamlines, facilitated by so-called “Virtual Beamline” module. In this present tech note, we discuss the nonlinear optimization features of SRW with emphasis on the optimization library in SRW. This optimization library not only can be used to optimize the simulation beamline in SRW, it also can be integrated with real beamline operations and optimize the parameters based on real measurements. We also show some demos of the optimization library recently developed in SRW.

36 MATERIALS SCIENCE↗
NASA NTRS1988

On the validation of a code and a turbulence model appropriate to circulation control airfoils

A computer code for calculating flow about a circulation control airfoil within a wind tunnel test section has been developed. This code is being validated for eventual use as an aid to design such airfoils. The concept of code validation being used is explained. The initial stages of the process have been accomplished. The present code has been applied to a low-subsonic, 2-D flow about a circulation control airfoil for which extensive data exist. Two basic turbulence models and variants thereof have been successfully introduced into the algorithm, the Baldwin-Lomax algebraic and the Jones-Launder two-equation models of turbulence. The variants include adding a history of the jet development for the algebraic model and adding streamwise curvature effects for both models. Numerical difficulties and difficulties in the validation process are discussed. Turbulence model and code improvements to proceed with the validation process are also discussed.

Viegas, J. R.↗
NASA NTRS1988

On the validation of a code and a turbulence model appropriate to circular control airfoils

A computer code for calculating flow about a circulation control airfoil within a wind tunnel test section was developed. This code is being validated for eventual use as an aid to design such airfoils. The concept of code validation being used is explained. The initial stages of the process were accomplished. The present code was applied to a low subsonic, 2-D flow about a circulation control airfoil for which extensive data exist. Two basic turbulence models and variants thereof were successfully introduced into the algorithm, the Baldwin-Lomax algebraic and the Jones-Launder two equation models of turbulence. The variants include adding a history of the jet development for the algebraic model and adding streamwise curvature effects for both models. Numerical difficulties and difficulties in the validation process are discussed. Turbulence model and code improvements to proceed with the validation process are also discussed.

Viegas, J. R.↗
DOE OSTI2016

An Overview on the role of nuclear data: analysis and evaluation - invited Eugene P. Wigner Reactor Physicist Award Winner

The design and calculation of a nuclear system requires the generation of cross-section libraries in the multi-group or continuous-energy form that is suitable for use in reactor physics computer codes. These working libraries are derived from general purpose libraries, named evaluated nuclear data libraries that are constructed based on data evaluations using physical models and theories. Presently, the most common used evaluated nuclear data libraries are the ENDF/B, JEFF, JENDL, CENDL, and BROND. These general purpose nuclear data libraries are the backbone of any design involving neutron-target interactions. The data for neutron interaction in these libraries spans from fractions of electron-volts (eV), that is 10-5 eV, up to millions of electron-volts (MeV), that is 20 MeV or even higher energies. The purpose of this paper is to provide an overview on the approach used to generate nuclear data for practical applications especially in the resonance energy region. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS1986

Analytical determination of propeller performance degradation due to ice accretion

A computer code has been developed which is capable of computing propeller performance for clean, glaze, or rime iced propeller configurations, thereby providing a mechanism for determining the degree of performance degradation which results from a given icing encounter. The inviscid, incompressible flow field at each specified propeller radial location is first computed using the Theodorsen transformation method of conformal mapping. A droplet trajectory computation then calculates droplet impingement points and airfoil collection efficiency for each radial location, at which point several user-selectable empirical correlations are available for determining the aerodynamic penalities which arise due to the ice accretion. Propeller performance is finally computed using strip analysis for either the clean or iced propeller. In the iced mode, the differential thrust and torque coefficient equations are modified by the drag and lift coefficient increments due to ice to obtain the appropriate iced values. Comparison with available experimental propeller icing data shows good agreement in several cases. The code's capability to properly predict iced thrust coefficient, power coefficient, and propeller efficiency is shown to be dependent on the choice of empirical correlation employed as well as proper specification of radial icing extent.

Miller, T. L.↗
DOE OSTI2016

Neutron Activation Foil and Thermoluminescent Dosimeter Responses to a Lead Reflected Pulse of the CEA Valduc SILENE Critical Assembly

This benchmark experiment was conducted as a joint venture between the US Department of Energy (DOE) and the French Commissariat à l'Energie Atomique (CEA). Staff at the Oak Ridge National Laboratory (ORNL) in the US and the Centre de Valduc in France planned this experiment. The experiment was conducted on October 13, 2010 in the SILENE critical assembly facility at Valduc. Several other organizations contributed to this experiment and the subsequent evaluation, including CEA Saclay, Lawrence Livermore National Laboratory (LLNL), the Y-12 National Security Complex (NSC), Babcock International Group in the United Kingdom, and Los Alamos National Laboratory (LANL). The goal of this experiment was to measure neutron activation and thermoluminescent dosimeter (TLD) doses from a source similar to a fissile solution critical excursion. The resulting benchmark can be used for validation of computer codes and nuclear data libraries as required when performing analysis of criticality accident alarm systems (CAASs). A secondary goal of this experiment was to qualitatively test performance of two CAAS detectors similar to those currently and formerly in use in some US DOE facilities. The detectors tested were the CIDAS MkX and the Rocky Flats NCD-91. The CIDAS detects gammas with a Geiger-Muller tube, and the Rocky Flats detects neutrons via charged particles produced in a thin 6 LiF disc, depositing energy in a Si solid-state detector. These detectors were being evaluated to determine whether they would alarm, so they were not expected to generate benchmark quality data.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Neutron Activation Foil and Thermoluminescent Dosimeter Responses to a Polyethylene Reflected Pulse of the CEA Valduc SILENE Critical Assembly

This benchmark experiment was conducted as a joint venture between the US Department of Energy (DOE) and the French Commissariat à l'Energie Atomique (CEA). Staff at the Oak Ridge National Laboratory (ORNL) in the US and the Centre de Valduc in France planned this experiment. The experiment was conducted on October 19, 2010 in the SILENE critical assembly facility at Valduc. Several other organizations contributed to this experiment and the subsequent evaluation, including CEA Saclay, Lawrence Livermore National Laboratory (LLNL), the Y-12 National Security Complex (NSC), Babcock International Group in the United Kingdom, and Los Alamos National Laboratory (LANL). The goal of this experiment was to measure neutron activation and thermoluminescent dosimeter (TLD) doses from a source similar to a fissile solution critical excursion. The resulting benchmark can be used for validation of computer codes and nuclear data libraries as required when performing analysis of criticality accident alarm systems (CAASs). A secondary goal of this experiment was to qualitatively test performance of two CAAS detectors similar to those currently and formerly in use in some US DOE facilities. The detectors tested were the CIDAS MkX and the Rocky Flats NCD-91. The CIDAS detects gammas with a Geiger-Muller tube and the Rocky Flats detects neutrons via charged particles produced in a thin 6 LiF disc depositing energy in a Si solid state detector. These detectors were being evaluated to determine whether they would alarm, so they were not expected to generate benchmark quality data.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2013

Safety Aspects of Dry Spent Fuel Storage and Spent Fuel Management - 13559

Dry storage systems are characterized by passive and inherent safety systems ensuring safety even in case of severe incidents or accidents. After the events of Fukushima, the advantages of such passively and inherently safe dry storage systems have become more and more obvious. As with the storage of all radioactive materials, the storage of spent nuclear fuel (SF) and high-level radioactive waste (HLW) must conform to safety requirements. Following safety aspects must be achieved throughout the storage period: - safe enclosure of radioactive materials, - safe removal of decay heat, - securing nuclear criticality safety, - avoidance of unnecessary radiation exposure. The implementation of these safety requirements can be achieved by dry storage of SF and HLW in casks as well as in other systems such as dry vault storage systems or spent fuel pools, where the latter is neither a dry nor a passive system. Furthermore, transport capability must be guaranteed during and after storage as well as limitation and control of radiation exposure. The safe enclosure of radioactive materials in dry storage casks can be achieved by a double-lid sealing system with surveillance of the sealing system. The safe removal of decay heat must be ensured by the design of the storage containers and the storage facility. The safe confinement of radioactive inventory has to be ensured by mechanical integrity of fuel assembly structures. This is guaranteed, e.g. by maintaining the mechanical integrity of the fuel rods or by additional safety measures for defective fuel rods. In order to ensure nuclear critically safety, possible effects of accidents have also to be taken into consideration. In case of dry storage it might be necessary to exclude the re-positioning of fissile material inside the container and/or neutron moderator exclusion might be taken into account. Unnecessary radiation exposure can be avoided by the cask or canister vault system itself. In Germany dry storage of SF in casks fulfills both transport and storage requirements. Mostly, storage facilities are designed as concrete buildings above the ground, but due to regional constraints, one storage facility has also been built as a rock tunnel. The decay heat is always removed by natural air flow; further technical equipment is not needed. The removal of decay heat and shielding had been modeled and calculated by state-of-the-art computer codes before such a facility has been built. TueV and BAM present their long experience in the licensing process for sites and casks and inform about spent nuclear fuel management and issues concerning dry storage of spent nuclear fuel. Different storage systems and facilities in Germany, Europe and world-wide are compared with respect to the safety aspects mentioned above. Initial points are the safety issues of wet storage of SF, and it is shown how dry storage systems can ensure the compliance with the mentioned safety criteria over a long storage period. The German storage concept for dry storage of SF and HLW is presented and discussed. Exemplarily, the process of licensing, erection and operation of selected German dry storage facilities is presented. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

An Evidence-Based Critique of U.S. Nuclear Waste Policy Using Data from Reactor and Cyclotron Sites - 18568

Concrete at nuclear facilities becomes radioactive when it is exposed to a neutron flux, and when such facilities are vacated, decommissioning activities are enacted to make the site safe for reuse. These activities generate wastes that the public expects to be managed in their best interest. This paper provides side-by-side comparisons of concrete bio-shielding encountered at typical reactor and cyclotron decommissioning projects. Three sites are considered, and results from characterization samples are used to demonstrate that the wastes generated by such sites are similar in terms of content and risk. First, the dose potential represented by each waste is determined from mass-based radioactivity concentration with the RESRAD-BUILD computer code. The resulting doses are translated into population-risk estimates by applying dose-response predictions from the National Academy of Sciences' 2006 BEIR VII report. In terms of possible adverse health outcomes, the risk of solid cancer or leukemia associated with the wastes this paper examines is 8 to 50 extra cases per 100,000 persons. The regulatory framework applicable to each site is also reviewed, and the waste management strategies used at each site are contrasted. This paper demonstrates that the wastes it examines are virtually indistinguishable. Contrary to this finding, the wastes are regulated, managed, and disposed in very different manners according to current U.S. nuclear waste policy. The value of this paper is that the side-by-side comparisons it presents provides an opportunity to examine the risk-benefit of existing policies in terms of the benefits or disparities they represent. (author)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Performance assessment for the geological disposal of Deep Burn spent fuel using TTBX

The behavior of Deep Burn Modular High Temperature Reactor Spent Fuel (DBSF) is investigated in the Yucca Mountain geological repository (YMR) with respect to the annual dose (Sv/yr) delivered to the Reasonably Maximally Exposed Individual (RMEI) from the transport of radionuclides released from the graphite waste matrix. Transport calculations are performed with a novel computer code, TTBX which is capable of modeling transport pathways that pass through heterogeneous geological formations. TTBX is a multi-region extension of the existing single region TTB transport code. Overall the peak annual dose received by the RMEI is seen to be four orders of magnitude lower than the regulatory threshold for exposure, even under pessimistic scenarios. A number of factors contribute to the favorable performance of DBSF. A reduction of one order of magnitude in the peak annual dose received by the RMEI is observed for every order of magnitude increase in the waste matrix lifetime, highlighting the importance of the waste matrix durability and suggesting graphite's utility as a potential waste matrix for the disposal of high-level waste. Furthermore, we see that by incorporating a higher fidelity far-field model the peak annual dose calculated to be received by the RMEI is reduced by two orders of magnitude. By accounting for the heterogeneities of the far field we have simultaneously removed unnecessary conservatisms and improved the fidelity of the transport model. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Tersoff Benchmarking of Be-C-H Interatomic Potential

To demonstrate prolonged fusion power in ITERs (International Thermonuclear Experimental Reactors), the selection of appropriate plasma facing materials (PFMs) is a vital issue that requires the use of interatomic potentials to simulate the plasma interactions with the PFMs. So it is no surprise that the use of computer codes such as LAMMPS to perform studies in order to determine proper PFMs is necessary based on the complexity of such studies. In recent Tokamak advancements, beryllium was chosen as a first wall material due to its low Z (lower energy loss due to plasma contamination), its low tritium inventory and its oxygen gettering quality. With its high melting point and ability to withstand high heat fluxes, beryllium makes a fine choice for a PFM. In this work, Be-C-H (beryllium-carbon-hydrogen) bonds were investigated in a LAMMPS Tersoff simulation, where the interatomic potentials for Be-C-H were converted to LAMMPS Tersoff format, and were validated by calculating the cohesive energy (E{sub coh}), lattice constant (a), and atomic volume (V{sub 0}) for comparison to Density Functional Theory (DFT), Analytical Bond-Order Potentials (ABOP) and experiment potentials. This benchmark serves as the validation of performing Be-C-H bombardment simulations using LAMMPS.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

obtain3D

This computer code uses the concept of the parallax to compute the x, y and z coordinates of points found using transmission electron microscopy (TEM), or any transmission imaging technique, using two images, each taken at a different perspective of the region containing the points. Points correspond, but are not limited, to the center of cavities or precipitates, positions of irradiation black dot damage, positions along a dislocation line, or positions along where an interface meets a free surface. The code allows the user to visualize the features containing the points in three dimensions. Features can include dislocations, interfaces, cavities, precipitates, inclusions etc. The x, y and z coordinates of the points are output in a text file as well. The program can also combine the x, y and z coordinates of the points with crystallographic directional information from diffraction pattern(s) to calculate dislocation line directions and interface plane normals.

Eftink, Benjamin Paul↗
DOE OSTI2017

A Gibbs Energy Minimization Approach for Modeling of Chemical Reactions in a Basic Oxygen Furnace

In modern steelmaking, the decarburization of hot metal is converted into steel primarily in converter processes, such as the basic oxygen furnace. The objective of this work was to develop a new mathematical model for top blown steel converter, which accounts for the complex reaction equilibria in the impact zone, also known as the hot spot, as well as the associated mass and heat transport. An in-house computer code of the model has been developed in Matlab. The main assumption of the model is that all reactions take place in a specified reaction zone. The mass transfer between the reaction volume, bulk slag, and metal determine the reaction rates for the species. The thermodynamic equilibrium is calculated using the partitioning of Gibbs energy (PGE) method. The activity model for the liquid metal is the unified interaction parameter model and for the liquid slag the modified quasichemical model (MQM). The MQM was validated by calculating iso-activity lines for the liquid slag components. The PGE method together with the MQM was validated by calculating liquidus lines for solid components. The results were compared with measurements from literature. The full chemical reaction model was validated by comparing the metal and slag compositions to measurements from industrial scale converter. The predictions were found to be in good agreement with the measured values. Furthermore, the accuracy of the model was found to compare favorably with the models proposed in the literature. The real-time capability of the proposed model was confirmed in test calculations.

36 MATERIALS SCIENCE↗