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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 1,063 records · Page 59
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Experimental study of the energy dependence of the total cross section for the {sup 6}He + {sup nat}Si and {sup 9}Li + {sup nat}Si reactions

New experimental measurements of the total reaction cross sections for the {sup 6}He + {sup nat}Si and {sup 9}Li + {sup nat}Si processes in the energy range of 5 to 40 A MeV are presented. A modified transmission method based on high-efficiency detection of prompt n–γ radiation has been used in the experiment. A bump is observed for the first time in the energy dependence σ{sub R}(E) at E ∼ 10–30 A MeV for the {sup 9}Li + {sup nat}Si reaction, and existence of the bump in σ{sub R}(E) at E ∼ 10–20 A MeV first observed in the standard transmission experiments is experimentally confirmed for the {sup 6}He + {sup nat}Si reaction. Theoretical analysis of the measured {sup 6}He + {sup nat}Si and {sup 9}Li + {sup nat}Si reaction cross sections is performed within the microscopic double folding model. Disagreement is observed between the experimental and theoretical cross sections in the region of the bump at the energies of 10 to 20 A MeV, which requires further study.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2015

On the fine structure of spectra of the inelastic-electron-scattering cross section and the Si surface parameter

Reflection electron-energy loss spectra are obtained for a series of Si samples with different crystallographic orientations, prepared under different technological conditions. Using the experimental spectra, the electron energy loss dependences of the product of the mean inelastic free path and differential inelastic electron scattering cross section are calculated. A new technique is suggested for analyzing the spectra of inelastic electron scattering cross section by simulating experimental spectra with the use of the three-parameter Tougaard universal cross section functions. The results of the simulation are used to determine the nature of loss peaks and to calculate the surface parameter.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Analytic error analysis of cross section interpolation methods in nodal diffusion codes - II: Numerical results

This paper is the second part of a two-part paper that documents the numerical results for the partial derivatives model presented in part I. In this paper, we derive the error bounds for the analytical point-wise error expression and verify our bounds with numerical experiments. The point-wise error expressions make available, and bound, the sources that contribute to the total error of the interpolated cross section in terms of the Lagrange interpolation errors and the model form error. MPACT is used to generate two-group homogenized cross sections for Westinghouse's AP1000 Region 4 lattice to evaluate the accuracy of the bounds. Error bounds calculated over a grid are compared to numerical data for uni-variate and multi-variate interpolation. The point-wise error bounds of a typical case matrix - two branches in each state variable - are displayed for bi-variate interpolation in the state variables: moderator density, fuel temperature, and boron concentration. The error bounds are shown to be highly accurate compared to numerical results, and in accordance with the underlying physics. We then discuss and show how the sources of error contribute to the total error, and consider the improvement of each error source. Finally, we mention future work such as propagating our cross section error bounds through a reactivity calculation. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Pu239 Cross-Section Variations Based on Experimental Uncertainties and Covariances

Algorithms and software have been developed for producing variations in plutonium-239 neutron cross sections based on experimental uncertainties and covariances. The varied cross-section sets may be produced as random samples from the multi-variate normal distribution defined by an experimental mean vector and covariance matrix, or they may be produced as Latin-Hypercube/Orthogonal-Array samples (based on the same means and covariances) for use in parametrized studies. The variations obey two classes of constraints that are obligatory for cross-section sets and which put related constraints on the mean vector and covariance matrix that detemine the sampling. Because the experimental means and covariances do not obey some of these constraints to sufficient precision, imposing the constraints requires modifying the experimental mean vector and covariance matrix. Modification is done with an algorithm based on linear algebra that minimizes changes to the means and covariances while insuring that the operations that impose the different constraints do not conflict with each other.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2013

Modeled Neutron Induced Nuclear Reaction Cross Sections for Radiochemsitry in the region of Thulium, Lutetium, and Tantalum I. Results of Built in Spherical Symmetry in a Deformed Region

We have developed a set of modeled nuclear reaction cross sections for use in radiochemical diagnostics. Systematics for the input parameters required by the Hauser-Feshbach statistical model were developed and used to calculate neutron induced nuclear reaction cross sections for targets ranging from Terbium (Z = 65) to Rhenium (Z = 75). Of particular interest are the cross sections on Tm, Lu, and Ta including reactions on isomeric targets.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

The ENDF/B-VIII.0 Compendium Volume II: Neutron Cross Section Rank-Order Analyses

Version 8 of the Evaluated Nuclear Data File (ENDF/B-VIII.0) contains a wealth of nuclear data representing the most recent collection of evaluated neutron cross section measurements, isotopic radioactive decay, and much more. This Compendium has been prepared as a easily-navigable reference of the contents of the ENDF/B-VIII.0 library, while making no statement regarding the accuracy of those contents. The Compendium is split into four volumes. Volume II presents a series of tables of reaction cross section values for thermal (0.025 eV), fission (1 MeV), and fusion (14 MeV) energies, each sorted by cross section value, isotope, and element-then-isotope to facilitate quick searches.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Absolute integral cross sections for the state-selected ion–molecule reaction N{sub 2}{sup +}(X{sup 2}Σ{sub g}{sup +}; v{sup +} = 0–2) + C{sub 2}H{sub 2} in the collision energy range of 0.03–10.00 eV

Using the vacuum ultraviolet laser pulsed field ionization-photoion source, together with the double-quadrupole–double-octopole mass spectrometer developed in our laboratory, we have investigated the state-selected ion–molecule reaction N{sub 2}{sup +}(X{sup 2}Σ{sub g}{sup +}; v {sup +} = 0–2, N{sup +} = 0–9) + C{sub 2}H{sub 2}, achieving high internal-state selectivity and high kinetic energy resolution for reactant N{sub 2}{sup +} ions. The charge transfer (CT) and hydrogen-atom transfer (HT) channels, which lead to the respective formation of product C{sub 2}H{sub 2}{sup +} and N{sub 2}H{sup +} ions, are observed. The vibrationally selected absolute integral cross sections for the CT [σ {sub CT}(v {sup +})] and HT [[σ {sub HT}(v {sup +})] channels obtained in the center-of-mass collision energy (E {sub cm}) range of 0.03–10.00 eV reveal opposite E {sub cm} dependences. The σ {sub CT}(v {sup +}) is found to increase as E {sub cm} is decreased, and is consistent with the long-range exothermic CT mechanism, whereas the E {sub cm} enhancement observed for the σ {sub HT}(v {sup +}) suggests effective coupling of kinetic energy to internal energy, enhancing the formation of N{sub 2}H{sup +}. The σ {sub HT}(v {sup +}) curve exhibits a step at E {sub cm} = 0.70–1.00 eV, suggesting the involvement of the excited C{sub 2}H{sub 2}{sup +}(A{sup 2}Σ{sub g}{sup +}) state in the HT reaction. Contrary to the strong E {sub cm} dependences for σ {sub CT}(v {sup +}) and σ {sub HT}(v {sup +}), the effect of vibrational excitation of N{sub 2}{sup +} on both the CT and HT channels is marginal. The branching ratios and cross sections for the CT and HT channels determined in the present study are useful for modeling the atmospheric compositions of Saturn's largest moon, Titan. These cross sections and branching ratios are also valuable for benchmarking theoretical calculations on chemical dynamics of the titled reaction.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

The {sup 95}Zr( n , γ ){sup 96}Zr Cross Section from the Surrogate Ratio Method and Its Effect on s -process Nucleosynthesis

The {sup 95}Zr( n , γ ){sup 96}Zr reaction cross section is crucial in the modeling of s -process nucleosynthesis in asymptotic giant branch stars because it controls the operation of the branching point at the unstable {sup 95}Zr and the subsequent production of {sup 96}Zr. We have carried out the measurement of the {sup 94}Zr({sup 18}O, {sup 16}O) and {sup 90}Zr({sup 18}O, {sup 16}O) reactions and obtained the γ -decay probability ratio of {sup 96}Zr* and {sup 92}Zr* to determine the {sup 95}Zr( n , γ ){sup 96}Zr reaction cross sections with the surrogate ratio method. Our deduced Maxwellian-averaged cross section of 66 ± 16 mb at 30 keV is close to the value recommended by Bao et al., but 30% and more than a factor of two larger than the values proposed by Toukan and Käppeler and Lugaro et al., respectively, and routinely used in s -process models. We tested the new rate in stellar models with masses between 2 and 6 M {sub ⊙} and metallicities of 0.014 and 0.03. The largest changes—up to 80% variations in {sup 96}Zr—are seen in models of mass 3–4 M {sub ⊙}, where the {sup 22}Ne neutron source is mildly activated. The new rate can still provide a match to data from meteoritic stardust silicon carbide grains, provided that the maximum mass of the parent stars is below 4 M {sub ⊙}, for a metallicity of 0.03.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2012

Measurement of the inclusive W-+/- and Z/gamma* cross sections in the e and mu decay channels in pp collisions at root s=7 TeV with the ATLAS detector

The production cross sections of the inclusive Drell-Yan processes W{sup {+-}} {yields} {ell}{nu} and Z/{gamma}* {yields} {ell}{ell} ({ell} = e, {mu}) are measured in proton-proton collisions at {radical}s = 7 TeV with the ATLAS detector. The cross sections are reported integrated over a fiducial kinematic range, extrapolated to the full range, and also evaluated differentially as a function of the W decay lepton pseudorapidity and the Z boson rapidity, respectively. Based on an integrated luminosity of about 35 pb{sup -1} collected in 2010, the precision of these measurements reaches a few percent. The integrated and the differential W{sup {+-}} and Z/{gamma}* cross sections in the e and {mu} channels are combined, and compared with perturbative QCD calculations, based on a number of different parton distribution sets available at next-to-next-to-leading order.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2016

A CUMULATIVE MIGRATION METHOD FOR COMPUTING RIGOROUS TRANSPORT CROSS SECTIONS AND DIFFUSION COEFFICIENTS FOR LWR LATTICES WITH MONTE CARLO

A new method for computing homogenized assembly neutron transport cross sections and dif- fusion coefficients that is both rigorous and computationally efficient is proposed in this paper. In the limit of a homogeneous hydrogen slab, the new method is equivalent to the long-used, and only-recently-published CASMO transport method. The rigorous method is used to demonstrate the sources of inaccuracy in the commonly applied “out-scatter” transport correction. It is also demonstrated that the newly developed method is directly applicable to lattice calculations per- formed by Monte Carlo and is capable of computing rigorous homogenized transport cross sections for arbitrarily heterogeneous lattices. Comparisons of several common transport cross section ap- proximations are presented for a simple problem of infinite medium hydrogen. The new method has also been applied in computing 2-group diffusion data for an actual PWR lattice from BEAVRS benchmark.

Cross sections↗
DOE OSTI2012

Proposal for the utilization of the total cross section covariances and its correlations with channel reactions for sensitivity and uncertainty analysis

An alternate method for the estimation of the global uncertainty on criticality, using the total cross section and its covariances, is proposed. Application of the method with currently available covariance data leads to an unrealistically large prediction of the global uncertainty on criticality. New covariances for total cross section and individual reactions are proposed. Analysis with the proposed covariance matrices is found to result in a global uncertainty for criticality consistent with the traditional method. Recommendations are made to evaluators for providing total cross section covariances. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Comparison of Serpent and HELIOS-2 as applied for the PWR few-group cross section generation

This paper discusses recent modifications to the Serpent Monte Carlo code methodology and related to the calculation of few-group diffusion coefficients and reflector discontinuity factors The new methods were assessed in the following manner. First, few-group homogenized cross sections calculated by Serpent for a reference PWR core were compared with those generated 1 commercial deterministic lattice transport code HELIOS-2. Second, Serpent and HELIOS-2 fe group cross section sets were later employed by nodal diffusion code DYN3D for the modeling the reference PWR core. Finally, the nodal diffusion results obtained using the both cross section sets were compared with the full core Serpent Monte Carlo solution. The test calculations show that Serpent can calculate the parameters required for nodal analyses similar to conventional deterministic lattice codes. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Study on generating of thermal neutron scattering cross sections for LiH

LiH is designated as a promising moderator and shielding material because of its low density, high melting point and large fraction of H atoms. However, lack of the thermal neutron cross sections of LiH makes numerical calculation deviate from experimental data to some extent. As a result, it is necessary to study LiH thermal kernel effect. The phonon property of LiH has been investigated by first-principles calculations using the plane-wave pseudo potential method with CASTEP code. The scattering law and the thermal neutron scattering cross sections for Li and H have been generated using this distribution. The results have been compared with zirconium hydride data. The GASKET and NJOY/LEAPR codes have been used in the calculation of scattering law, whose results have been compared with the reference; the discrepancy mainly comes from phonon spectrums and its expansion. LEAPR had the capability to compute scattering through larger energy and momentum transfers than GASKET did. By studying LiH phonon spectrum and constructing the model of LiH thermal kernel and scattering matrix, the ACE format LiH thermal neutron cross sections for MCNP software could be made and used for reactor Neutronics calculation. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

McStas Analysis of Monoenergetic Thermal Neutron Beam Generation for Total Cross Section Measurements

The thermal neutron scattering law (TSL) of a given material represents data that is essential for assessing the criticality safety of a neutronic system in which this material is utilized. Traditionally, the TSL is developed through elaborate calculations that combine input and models based on condensed matter physics and scattering theory. Nonetheless, measurements of key parameters (e.g., the total cross section at low energies) can help in benchmarking and validating TSL libraries such as those used in the ENDF/B database. In this work, the setup of experiments, at the PULSTAR reactor of North Carolina State University (NCSU), for creating monoenergetic thermal neutron beams in support of total scattering cross section measurements in the thermal energy range is explored using McStas simulations and experimental verification. A McStas model has been established of the NPD facility at the NCSU PULSTAR reactor to explore the production of monoenergetic thermal neutron beams in support of various applications including total cross section measurements. The McStas results clearly demonstrate the viability of six monochromatic thermal neutron beams with substantial intensities that allow future applications at the NPD facility. The experimental results reasonably support the computational model predictions. Based on these results, transmission thermal neutron experiments can be conducted for data validation studies. Furthermore, the model also allows the evaluation of design changes to expand the portfolio of available monoenergetic thermal neutron beams. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Resonance Self-Shielding Using Deterministic Transport Methods with Pointwise Cross Sections (Invited) - Eugene P. Wigner Reactor Physics Award

The difficulty of predicting resonance absorption has been recognized since the birth of reactor physics. A paper published by Wigner, et al. in 1955 describes pioneering theoretical work done in 1941 to provide an estimate for the resonance absorption in heterogeneous uranium lattices. This work introduced many fundamental concepts, including the narrow resonance (NR) approximation and the Wigner rational expression for the escape probability from a fuel lump, and it established an equivalence relation for resonance absorption in homogeneous and heterogeneous systems. During the 1960's, Bondarenko proposed an approach in which resonance shielding factors are precomputed and parametrized as a function of the background cross section by using the NR approximation and equivalence theory. This provided an efficient technique for treating resonance absorption in reactor lattices. The Bondarenko method and its modern variations (e.g., certain subgroup techniques) have become the predominate method for resonance shielding calculations in reactor physics. However, this approach has several limitations, most of which are well known, such as the treatment of resonance interference effects and its reliance on homogeneous-heterogeneous equivalence theory relationships. Other limitations are more subtle and less generally recognized, such as neglecting the impact of self-shielding the multigroup (MG) elastic removal cross section and the effects of directionally dependent shielding on the total cross section. This paper and the accompanying presentation apply deterministic solutions of the continuous energy (CE) neutron transport equation to examine common approximations made in resonance methods for a uniform lattice. The calculated fine-structure spectra are used to calibrate the impact of the approximations and describe how results from CE flux solution can be incorporated in the Bondarenko self-shielding method to improve the accuracy. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Uncertainty Quantification of Activation Due to Cross Section Data in Neutron Shielding Calculation

Quantification of activation in structure materials around the nuclear reactor through calculation is necessary before start of decommissioning. Then a neutron shielding calculation is carried out to estimate activation of structural materials. Neutron shielding calculation is performed out for spatially large regions around a reactor through deep penetration calculation, calculation result would have significant uncertainty. In order to minimize total amount of structure materials treated as radioactive waste, quantitative estimation on uncertainty of structural activation and its reduction is important. The uncertainties due to calculation arises from that of input data and that of calculation method. Examples of uncertainty of input data is those for nuclear reaction cross section and atom density of the nuclide in the structure. In this study, we focus on the uncertainty of cross section, which would be one of the dominant uncertainties, and try to estimate uncertainty of activation due to cross section data in a neutron shielding calculation. The random sampling method (RS) is used as the uncertainty quantification method. With the RS method, uncertainty estimation can be carried out without modification of computer codes for a neutron shielding calculation. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Improved Gamma Yield and Interaction Cross-Section Libraries of MC2-3

The MC{sup 2}-3 code developed by Argonne National Laboratory (ANL) has been widely used for multigroup cross section generation for fast reactor design and analysis. The neutron cross sections generated with MC{sup 2}-3 have been validated with various fast critical experiments. The gamma production data and interaction cross sections were also verified with the EBR-II benchmark problem. The current gamma calculation process is based on the 21-group gamma libraries generated with the NJOY code, which does not include the delayed gamma production and does not account for the self-shielding effect on the KERMA (Kinetic Energy Released in Material) factor. In order to improve gamma prediction accuracy, the gamma libraries of MC{sup 2}-3 have recently been extended from 21 to 94 groups, including delayed gamma data and kinetic energy release per reaction data. The latter is used to take into account the self-shielding effect on KERMA factors. Gamma transport calculation capabilities have also been added to MC{sup 2}-3. In this paper, we present the updated data and gamma transport capabilities along with the verification tests with ABTR and EBR-II benchmark problems. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Spectral-average cross section validation in LR-0 reactor spectrum of {sup 55}Mn(n,2n), {sup 90}Zr(n,2n) and {sup 127}I(n,2n) reactions - 25362

This article deals with the validation of spectral average cross-sections (SACS) of {sup 55}Mn(n,2n){sup 54}Mn, {sup 90}Zr(n,2n){sup 89}Zr and {sup 127}I(n,2n){sup 126}I reactions in well-defined LR-0 reactor spectrum. The measurement of SACS for selected reactions is of high interest because they are used for the practical reactor dosimetry and also for refinement of prompt fission neutron spectra of {sup 235}U in the region of higher energies. Various experiments for description of neutron field in reactor and detector characterization, which are necessary for validation, are described. The spectral average cross-sections are derived from Net Peak Areas of irradiated samples measured using a semiconductor high purity germanium detector. The spectral average cross section in {sup 235}U prompt fission neutron spectrum was experimentally determined to be 0.239 ± 0.015 mb for {sup 55}Mn, 1.209 ± 0.052 mb for {sup 127}I and 0.107 ± 0.005 mb for {sup 90}Zr. The obtained results were compared with MCNP6 calculations using ENDF/B-VII.0, ENDF/B-VII.1, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND-2010, CENDL-3.1 and IRDFF-1.05 nuclear data libraries. Notable discrepancies were found in {sup 90}Zr(n,2n){sup 89}Zr and {sup 127}I(n,2n){sup 126}I reactions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗