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

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At least 1,531 records · Page 85

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NASA NTRS1978

Planetary gamma-ray spectroscopy

The measured intensities of certain gamma rays of specific energies escaping from a planetary surface can be used to determine the abundances of a number of elements. The fluxes of the more intense gamma-ray lines emitted from 32 elements were calculated using current nuclear data and existing models for the source processes. The source strengths for neutron-capture reactions were modified from those previously used. The fluxes emitted form a surface of average lunar composition are reported for 292 gamma-ray lines. These theoretical fluxes were used elsewhere to convert the data from the Apollo gamma-ray spectrometers to elemental abundances and can be used with measurements from future missions to map the concentrations of a number of elements over a planet's surface. Detection sensitivities for these elements are examined and applications of gamma-ray spectroscopy for future orbiters to Mars and other solar-system objects are discussed.

Reedy, R. C.↗

NASA NTRS1978

Nuclear gamma rays from energetic particle interactions

Gamma ray line emission from nuclear deexcitation following energetic particle reactions is evaluated. The compiled nuclear data and the calculated gamma ray spectra and intensities can be used for the study of astrophysical sites which contain large fluxes of energetic protons and nuclei. A detailed evaluation of gamma ray line production in the interstellar medium is made.

Ramaty, R.↗

NASA NTRS1978

Planetary gamma-ray spectroscopy

Reedy, R. C.↗

NASA NTRS1982

Studies of neutron and proton nuclear activation in low-Earth orbit

The expected induced radioactivity of experimental material in low Earth orbit was studied for characteristics of activating particles such as cosmic rays, high energy Earth albedo neutrons, trapped protons, and secondary protons and neutrons. The activation cross sections for the production of long lived radioisotopes and other existing nuclear data appropriate to the study of these reactions were compiled. Computer codes which are required to calculate the expected activation of orbited materials were developed. The decreased computer code used to predict the activation of trapped protons of materials placed in the expected orbits of LDEF and Spacelab II. Techniques for unfolding the fluxes of activating particles from the measured activation of orbited materials are examined.

Laird, C. E.↗

NASA NTRS1990

Risk assessment methodologies for target fragments produced in high-energy nucleon reactions

Collisions of energetic protons with target nuclei yield nuclear products of large linear energy transfer, and it is usually assumed that the energy of the heavy products is absorbed locally with a quality factor of 20. Past methods of risk assessment have relied on the Bertini (1969) model to evaluate the nuclear reaction products. If the Q(F) of 20 is correct, then a significant underestimate of biological risk from the nuclear products could result from the use of Bertini's model. On the other hand, a Q(F) of 20 is probably too large for some of the lighter fragments, leaving prior risk estimates as possibly correct or even conservative. This issue is discussed here in the context of the recently proposed nuclear data set of Wilson et al. (1989).

Shinn, J. L.↗

NASA NTRS1992

The cosmological constant

The cosmological constant problem is examined in the context of both astronomy and physics. Effects of a nonzero cosmological constant are discussed with reference to expansion dynamics, the age of the universe, distance measures, comoving density of objects, growth of linear perturbations, and gravitational lens probabilities. The observational status of the cosmological constant is reviewed, with attention given to the existence of high-redshift objects, age derivation from globular clusters and cosmic nuclear data, dynamical tests of Omega sub Lambda, quasar absorption line statistics, gravitational lensing, and astrophysics of distant objects. Finally, possible solutions to the physicist's cosmological constant problem are examined.

Carroll, Sean M.↗

NASA NTRS2000

The Concentration of Oxygen (and Silicon) in Lunar Materials

As an aid for interpretation of nuclear data obtained remotely, we provide data on concentrations of oxygen and silicon in a variety of lunar materials.

Korotev, Randy L.↗

DOE OSTI2013

A Complex-Geometry Validation Experiment for Advanced Neutron Transport Codes

The Idaho National Laboratory (INL) has initiated a focused effort to upgrade legacy computational reactor physics software tools and protocols used for support of core fuel management and experiment management in the Advanced Test Reactor (ATR) and its companion critical facility (ATRC) at the INL.. This will be accomplished through the introduction of modern high-fidelity computational software and protocols, with appropriate new Verification and Validation (V&V) protocols, over the next 12-18 months. Stochastic and deterministic transport theory based reactor physics codes and nuclear data packages that support this effort include MCNP5[1], SCALE/KENO6[2], HELIOS[3], SCALE/NEWT[2], and ATTILA[4]. Furthermore, a capability for sensitivity analysis and uncertainty quantification based on the TSUNAMI[5] system has also been implemented. Finally, we are also evaluating the Serpent[6] and MC21[7] codes, as additional verification tools in the near term as well as for possible applications to full three-dimensional Monte Carlo based fuel management modeling in the longer term. On the experimental side, several new benchmark-quality code validation measurements based on neutron activation spectrometry have been conducted using the ATRC. Results for the first four experiments, focused on neutron spectrum measurements within the Northwest Large In-Pile Tube (NW LIPT) and in the core fuel elements surrounding the NW LIPT and the diametrically opposite Southeast IPT have been reported [8,9]. A fifth, very recent, experiment focused on detailed measurements of the element-to-element core power distribution is summarized here and examples of the use of the measured data for validation of corresponding MCNP5, HELIOS, NEWT, and Serpent computational models using modern least-square adjustment methods are provided.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2014

Role of Experiment Covariance in Cross Section Adjustments

This paper is dedicated to the memory of R. D. McKnight, which gave a seminal contribution in establishing methodology and rigorous approach in the evaluation of the covariance of reactor physics integral experiments. His original assessment of the ZPPR experiment uncertainties and correlations has made nuclear data adjustments, based on these experiments, much more robust and reliable. In the present paper it has been shown with some numerical examples the actual impact on an adjustment of accounting for or neglecting such correlations.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2015

A collision history-based approach to Sensitivity/Perturbation calculations in the continuous energy Monte Carlo code SERPENT

In this work, the implementation of a collision history-based approach to sensitivity/perturbation calculations in the Monte Carlo code SERPENT is discussed. The proposed methods allow the calculation of the eects of nuclear data perturbation on several response functions: the eective multiplication factor, reaction rate ratios and bilinear ratios (e.g., eective kinetics parameters). SERPENT results are compared to ERANOS and TSUNAMI Generalized Perturbation Theory calculations for two fast metallic systems and for a PWR pin-cell benchmark. New methods for the calculation of sensitivities to angular scattering distributions are also presented, which adopts fully continuous (in energy and angle) Monte Carlo estimators.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2016

Development of an Efficient Approach to Perform Neutronics Simulations for Plutonium-238 Production

Conversion of 238Pu decay heat into usable electricity is imperative to power National Aeronautics and Space Administration (NASA) deep space exploration missions; however, the current stockpile of 238Pu is diminishing and the quality is less than ideal. In response, the US Department of Energy and NASA have undertaken a program to reestablish a domestic 238Pu production program and a technology demonstration sub-project has been initiated. Neutronics simulations for 238Pu production play a vital role in this project because the results guide reactor safety-basis, target design and optimization, and post-irradiation examination activities. A new, efficient neutronics simulation tool written in Python was developed to evaluate, with the highest fidelity possible with approved tools, the time-dependent nuclide evolution and heat deposition rates in 238Pu production targets irradiated in the High Flux Isotope Reactor (HFIR). The Python Activation and Heat Deposition Script (PAHDS) was developed specifically for experiment analysis in HFIR and couples the MCNP5 and SCALE 6.1.3 software quality assured tools to take advantage of an existing high-fidelity MCNP HFIR model, the most up-to-date ORIGEN code, and the most up-to-date nuclear data. Three cycle simulations were performed with PAHDS implementing ENDF/B-VII.0, ENDF/B-VII.1, and the Hybrid Library GPD-Rev0 cross-section libraries. The 238Pu production results were benchmarked against VESTA-obtained results and the impact of various cross-section libraries on the calculated metrics were assessed.

depletion↗

DOE OSTI2015

Target Design Optimization of KIPT Neutron Source Facility

Argonne National Laboratory (ANL) of the United States developed and designed a neutron source facility for Kharkov Institute of Physics and Technology (KIPT) of Ukraine. The facility was constructed at Kharkov, Ukraine and its commissioning process has been started. The facility has an electron accelerator driving a subcritical assembly. The electron beam power is 100 kW using 100 MeV electrons. The subcritical assembly has WWR-M2 fuel assemblies with U-235 enrichment of 19.7 wt%. The facility will be utilized to perform basic and applied nuclear research, to produce medical isotopes, and to train young nuclear specialists. Solid target design with stacked disks is selected and each target disk is cooled by water from both side. Tungsten or natural uranium is the target material. This paper presents the target design optimization to maximize the neutron yield and the neutron flux level of the facility while satisfying the thermal-hydraulic design criteria. Monte Carlo computer code MCNPX is utilized for the neutron yield analyses with ENDF/B-VII.0 nuclear data libraries, as a function of target thickness using a simplified target model without cladding or coolant channels. It’s found that the neutron yield saturates at target thickness 60 ~ 70 mm for both tungsten and uranium materials. For the final target design with cladding and coolant channels, the total uranium thickness is 56.5 mm, and the total tungsten thickness is only 33.0 mm. The tungsten material has a large absorption cross section for thermal neutrons. If the tungsten thickness is increased, the neutron yield gain is offset by the neutron absorption reaction. The paper presents the design optimization analyses for both target materials.

electron accelerator↗

DOE OSTI2017

Impact of Reactor Operating Parameters on Cask Reactivity in BWR Burnup Credit

This paper discusses the effect of reactor operating parameters used in fuel depletion calculations onspent fuel cask reactivity, with relevance for boiling-water reactor (BWR) burnup credit (BUC)applications. Assessments that used generic BWR fuel assembly and spent fuel cask configurationsare presented. The considered operating parameters, which were independently varied in thedepletion simulations for the assembly, included fuel temperature, bypass water density, specificpower, and operating history. Different operating history scenarios were considered for the assemblydepletion to determine the effect of relative power distribution during the irradiation cycles, as wellas the downtime between cycles. Depletion, decay, and criticality simulations were performed usingcomputer codes and associated nuclear data within the SCALE code system. Results quantifying thedependence of cask reactivity on the assembly depletion parameters are presented herein.

Ilas, Germina↗

DOE OSTI2015

MONTE CARLO ANALYSES OF THE YALINA THERMAL FACILITY WITH SERPENT STEREOLITHOGRAPHY GEOMETRY MODEL

This paper analyzes the YALINA Thermal subcritical assembly of Belarus using two different Monte Carlo transport programs, SERPENT and MCNP. The MCNP model is based on combinatorial geometry and universes hierarchy, while the SERPENT model is based on Stereolithography geometry. The latter consists of unstructured triangulated surfaces defined by the normal and vertices. This geometry format is used by 3D printers and it has been created by: the CUBIT software, MATLAB scripts, and C coding. All the Monte Carlo simulations have been performed using the ENDF/B-VII.0 nuclear data library. Both MCNP and SERPENT share the same geometry specifications, which describe the facility details without using any material homogenization. Three different configurations have been studied with different number of fuel rods. The three fuel configurations use 216, 245, or 280 fuel rods, respectively. The numerical simulations show that the agreement between SERPENT and MCNP results is within few tens of pcms.

CUBIT↗

DOE OSTI2019

SCALE/Sampler Sensitivity Indices

The Sampler sequence of the SCALE code system for random sampling–based uncertainty analyses has recently been extended by including a sensitivity analysis. In uncertainty analyses with respect to uncertainties in the nuclear data, the determination of two multiple correlation coefficients—R2 and SPC2—allows for the identification of nuclide reactions that are significantly contributing to the investigated output uncertainty. Individual nuclide reactions are ranked by importance, allowing suggestions for additional measurements and evaluations to significantly reduce the observed output uncertainty. A brief description of the theory is provided herein, followed by application of this analysis to a sodium-cooled fast reactor pin cell.

Bostelmann, Rike↗

DOE OSTI2019

DETAILED DESIGN OF AN EPITHERMAL/INTERMEDIATE CRITICAL EXPERIMENT USING THE SANDIA NATIONAL LABORATORIES CRITICAL FACILITY

Nuclear criticality safety evaluations commonly take credit for neutron-absorbing materials. The normal practice for validating the calculational process to develop safety limits and margins of safety is to compare the safety case models with calculations of various critical experiments of a similar composition. Because there is a lack of critical experiments with various relevant neutron-absorbing materials mixed with fissile material, a new approach is being developed. This new approach uses small amounts of these credited materials in a critical assembly and assesses their worth and neutron energy spectral characteristics using absorption reaction rates. This approach is proposed as a new method for nuclear data validation. The assessment will involve measured central reactivity worth in a critical or near-critical assembly and subsequent comparison with calculated values from Monte Carlo computer analyses. Additionally, four configurations are proposed which replace fuel rods in the driver region with absorber rods. These configurations will allow for examination of the effects of the material in a more thermalized spectrum experiment selection in these scenarios. This paper presents some case studies examining the impact of these changes on some hypothetical safety analysis systems.This paper documents the detailed design portion of the critical experiment design process, focusing on development of a capability to test epithermal/ intermediate energy cross sections for materials using the 7uPCX critical experiment facility. Tantalum is specified as the test material to be examined.

Clarity, Justin↗

DOE OSTI2019

A Study of Fission Modes to Improve Nuclear Forensics

Obtain and evaluate charge and mass distributions from various databases and the literature Establish the mathematical formulism to decompose the selected mass distributions into their respective components of the multimodal fission models Compare and evaluate the results with prior work available from the literature Obtain, evaluate/validate experimentally-measured, photofission mass distributions using complementary INL-led research Apply decomposition methods to the photofission mass distributions Evaluate peak-to-valley ratios and specific isotopic ratios as a function of excitation energies Report and publish appropriate findings

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2013

Calculating infinite-medium {alpha}-eigenvalue spectra with a transition rate matrix method

The time-dependent behavior of the energy spectrum in neutron transport was investigated with a formulation, based on continuous-time Markov processes, for computing {alpha}-eigenvalues and eigenvectors in an infinite medium. For this, a research Monte Carlo code called TORTE was created and used to estimate elements of a transition rate matrix. TORTE is capable of using both multigroup and continuous-energy nuclear data, and verification was performed. Eigenvalue spectra for infinite homogeneous mixtures were obtained and an eigenfunction expansion was used to investigate transient behavior of the neutron energy spectrum. (authors)

97 MATHEMATICS AND COMPUTING↗