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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 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

ITER Port Interspace Pressure Calculations

The ITER ('the way' in Latin) project in Cadarache, France, will operate the largest tokamak ever built to demonstrate the physical and technical feasibility of nuclear fusion as an energy source. The ITER vacuum vessel is equipped with 50 access ports. Each port has an opening in the bio-shield that communicates with a dedicated port cell. During tokamak operation, the bio-shield opening must be closed with a concrete plug to shield the radiation coming from the plasma. This port plug separates the port cell into a port interspace (between vacuum vessel closure lid and port plug) on the inner side and a port cell on the outer side. This paper presents calculations of pressures and temperatures in the ITER port interspace after a double-ended guillotine break (DEGB) of a pipe in the tokamak cooling water system (TCWS). It is assumed that the DEGB occurs during the worst possible conditions-during water baking operation, with water at 523 K (250 deg. C) and 4.4 MPa. These conditions are more severe than during normal tokamak operation with the water at 398 K (125 deg. C) and 2 MPa. Two computer codes are employed in these calculations: RELAP5-3D Version 4.2.1 to calculate blowdown releases from the pipe break, and MELCOR, Version 1.8.6 to calculate pressures and temperatures in the port interspace. A sensitivity study has been performed to optimize the maximum peak pressure in the port interspace volume of ITER. The optimum case was for a labyrinth area (FL131) of 0.5 m{sup 2} and a relief panel (FL135) kept open after the first cycle. This case keeps the pressure in the port interspace under 150 kPa and the temperatures under 385 K. (authors)

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Synthesis of polycaprolactone-titanium oxide multilayer films by nanosecond laser pulses and electrospinning technique for better implant fabrication

In this study, a combination of electrospinning and laser texturing is introduced as a novel method for increasing the biocompatibility of metal implants. Besides having a rough laser treated surface, the implant benefits from the high porosity and better wettability of an electrospun fibrous structure, which is a more favorable environment for cell proliferation. Titanium samples were patterned using a nanosecond laser beam and were placed as collectors in an electrospinning machine. They were then soaked in simulated body fluid for four weeks. Energy Dispersive X-ray and X-Ray Diffraction results indicate significantly more hydroxyapatite formation on laser treated samples with nanoscale fibers deposited on their surface. This shows that having a laser treated surface underneath the fibrous layer can improve short-term biocompatibility even before degradation of fibers. The thermal conductivity of the electrospun layer, measured using a Hot Disk Transient Plane Source instrument and computer code, was shown to be considerably lower than that of titanium and very close to bone. The presence of this layer can therefore be beneficial in making the implant more compatible to a biological medium. In case of dental implants, it was shown that this layer can act as a thermal barrier while a hot beverage is consumed and it can decrease the temperature rise by about 60%, which avoids any possible damage to newly formed cells during the healing period.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2020

A Study on Radionuclides Selection for DCGLs Development for Site Release of KRR Units 1 and 2 site in Korea - 20256

The selection of Derived Concentration Guideline Levels (DCGLs) in nuclear facilities decommissioning is important because they serve as criteria for site characterization and final status surveys. DCGLs should be derived for KAERI TRIGA Mark II and III research reactor site clearance and prior to the work, a radionuclides list for DCGLs consideration was created. Radionuclides selection for DCGLs was performed according to the procedure proposed by the US NRC based on the characterization survey related to the decommission site. By evaluating the dose contribution of each radionuclide using D and D computer code, radionuclides with a dose contribution fraction of 0.1% or more for the entire radionuclide were selected. As a result, the selected radionuclides were beta radionuclides H-3, C-14 and Sr-90 and gamma radionuclides Co-60, Cs-137, Eu-152 and Eu-154. The selected radionuclides can be presented for consultation with domestic regulatory agencies in the future and will be used as the technical base of DCGLs for the final status survey. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2018

Coupling Exascale Multiphysics Applications: Methods and Lessons Learned

With the growing computational complexity of science and the complexity of new and emerging hardware, it is time to re-evaluate the traditional monolithic design of computational codes. One new paradigm is constructing larger scientific computational experiments from the coupling of multiple individual scientific applications, each targeting their own physics, characteristic lengths, and/or scales. We present a framework constructed by leveraging capabilities such as in-memory communications, workflow scheduling on HPC resources, and continuous performance monitoring. This code coupling capability is demonstrated by a fusion science scenario, where differences between the plasma at the edges and at the core of a device have different physical descriptions. This infrastructure not only enables the coupling of the physics components, but it also connects in situ or online analysis, compression, and visualization that accelerate the time between a run and the analysis of the science content. Results from runs on Titan and Cori are presented as a demonstration.

Choi, Jong Youl↗
DOE OSTI2016

NSRD-11: Computational Capability to Substantiate DOE-HDBK-3010 Data.

Safety basis analysts throughout the U.S. Department of Energy (DOE) complex rely heavily on the information provided in the DOE Handbook, DOE - HDBK - 3010, Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities, to determine radionuclide source terms. In calculating source terms, analysts tend to use the DOE Handbook's bounding values on airborne release fractions (ARFs) and respirable fractions (RFs) for various categories of insults (representing potential accident release categories). This is typically due to both time constraints and the avoidance of regulatory critique. Unfortunately, these bounding ARFs/RFs represent extremely conservative values. Moreover, they were derived from very limited small-scale bench/laboratory experiments and/or from engineered judgment. Thus, the basis for the data may not be representative of the actual unique accident conditions and configurations being evaluated. The goal of this research is to develop a more accurate and defensible method to determine bounding values for the DOE Handbook using state-of-art multi-physics-based computer codes. This enables us to better understand the fundamental physics and phenomena associated with the types of accidents in the handbook. In this year, this research included improvements of the high-fidelity codes to model particle resuspension and multi-component evaporation for fire scenarios. We also began to model ceramic fragmentation experiments, and to reanalyze the liquid fire and powder release experiments that were done last year. The results show that the added physics better describes the fragmentation phenomena. Thus, this work provides a low-cost method to establish physics-justified safety bounds by taking into account specific geometries and conditions that may not have been previously measured and/or are too costly to perform.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2013

Multi-cycle boiling water reactor fuel cycle optimization

In this work a new computer code, BWROPT (Boiling Water Reactor Optimization), is presented. BWROPT uses the Parallel Simulated Annealing (PSA) algorithm to solve the out-of-core optimization problem coupled with an in-core optimization that determines the optimum fuel loading pattern. However it uses a Haling power profile for the depletion instead of optimizing the operating strategy. The result of this optimization is the optimum new fuel inventory and the core loading pattern for the first cycle considered in the optimization. Several changes were made to the optimization algorithm with respect to other nuclear fuel cycle optimization codes that use PSA. Instead of using constant sampling probabilities for the solution perturbation types throughout the optimization as is usually done in PSA optimizations the sampling probabilities are varied to get a better solution and/or decrease runtime. The new fuel types available for use can be sorted into an array based on any number of parameters so that each parameter can be incremented or decremented, which allows for more precise fuel type selection compared to random sampling. Also, the results are sorted by the new fuel inventory of the first cycle for ease of comparing alternative solutions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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↗
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↗
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↗