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At least 685 records · Page 38
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Development of a Method for Measurement of Photonuclear-Reaction Cross Sections with the Aid of Simulated Quasimonoenergetic-Photon Spectra

A procedure for measuring photonuclear-reaction cross sections by employing the method of quasimonochromatization of the bremsstrahlung photon spectrum is developed. This procedure permits determining reaction cross sections via measuring the reaction yield at three fixed values of the electron energy. The procedure is equivalent to measuring the cross section in question for quasimonochromatic photons. Bremsstrahlung-photon spectra are obtained using state-of-the-art simulation means involving the GEANT4 code as applied to a specific experiment with allowance for its geometry, target thicknesses, electron-beam parameters, etc. The results obtained by simulating quasimonochromatic distributions of bremsstrahlung photons are presented for experiments at the LUE-8-5 linear electron accelerator of Institute for Nuclear Research (Moscow, Russian Academy of Sciences). These results are found to be in good agreement with the experimental energy distribution of the cross section for the reaction {sup 111}Cd(γ, γ') in the energy range between 5 and 8 MeV.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2013

Measurement of inelastic, single- and double-diffraction cross sections in proton–proton collisions at the LHC with ALICE

Measurements of cross sections of inelastic and diffractive processes in proton–proton collisions at LHC energies were carried out with the ALICE detector. The fractions of diffractive processes in inelastic collisions were determined from a study of gaps in charged particle pseudorapidity distributions: for single diffraction (diffractive mass M X <200 GeV/ c 2 ) \(\sigma_{\mathrm{SD}}/\sigma_{\mathrm {INEL}} = 0.21 \pm 0.03, 0.20^{+0.07}_{-0.08}\) , and \(0.20^{+0.04}_{-0.07}\) , respectively at centre-of-mass energies \(\sqrt{s} = 0.9, 2.76\mbox{, and }7~\mbox{TeV}\) ; for double diffraction (for a pseudorapidity gap Δ η >3) σ DD / σ INEL =0.11±0.03,0.12±0.05, and \(0.12^{+0.05}_{-0.04}\) , respectively at \(\sqrt{s} = 0.9, 2.76\mbox{, and }7~\mbox{TeV}\) . To measure the inelastic cross section, beam properties were determined with van der Meer scans, and, using a simulation of diffraction adjusted to data, the following values were obtained: \(\sigma_{\mathrm{INEL}} = 62.8^{+2.4}_{-4.0} (\mathrm{model}) \pm1.2 (\mathrm{lumi})\) mb at \(\sqrt{s} = 2.76\ \mbox{TeV}\) and \(73.2^{+2.0}_{-4.6} (\mathrm{model}) \pm2.6 (\mathrm{lumi})~\mbox{mb}\) at \(\sqrt{s} = 7\ \mbox{TeV}\) . The single- and double-diffractive cross sections were calculated combining relative rates of diffraction with inelastic cross sections. The results are compared to previous measurements at proton–antiproton and proton–proton colliders at lower energies, to measurements by other experiments at the LHC, and to theoretical models.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2022

Analytic error analysis of cross section interpolation methods in nodal diffusion codes - I : Theory

This paper discusses two cross section interpolation methods commonly found in popular nodal codes; the partial derivatives and multiple tables models. The motivation for choosing a model, and thus a case matrix structure, is a trade off between accuracy and computational cost. Due to decades of experience, there are default structures that are sufficient for current light water reactor analysis. However, this is not necessarily the case for advanced reactor designs. Therefore, it is advantageous to understand the sources of error in cross section interpolation models so that the quality of a case matrix may be improved. A mathematical framework for these models is presented in this work that provides a more rigorous connection between the nuclear engineering field's cross section interpolation methods and the broader mathematical field of function approximation. The two cross section models examined in this paper were found to utilize Lagrange interpolation and are a subset of Lagrange tensor products. Classical results of Lagrange polynomial error analysis were then applied to the partial derivative and multiple tables models to derive expressions for the total point-wise error. The analytical results classify the total error into two parts: the model form error and interpolation error. Finally, based on our observations, a better foundation for improving the quality of a case matrix is proposed. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

New Tools to Prepare ACE Cross-section Files for MCNP Analytic Test Problems

Monte Carlo calculations using one-group cross sections, multigroup cross sections, or simple continuous energy cross sections are often used to: (1) verify production codes against known analytical solutions, (2) verify new methods and algorithms that do not involve detailed collision physics, (3) compare Monte Carlo calculation methods with deterministic methods, and (4) teach fundamentals to students. In this work we describe 2 new tools for preparing the ACE cross-section files to be used by MCNP ® for these analytic test problems, simple_ace.pl and simple_ace_mg.pl.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Regional Geologic Cross Sections for Potential Storage and Containment Zones in the MRCSP Region

The Midwest Regional Carbon Sequestration Partnership (MRCSP) has incorporated the work of geologic research teams (Geoteams) in its regional characterization, project planning and carbon dioxide (CO 2 ) injection implementation work since the partnership was established by the U.S. Department of Energy (DOE) in 2003. Over this 16-year period, the cohort of Geoteams has grown from five to ten states and has contributed to the characterization of geologic sequestration opportunities, refinement of reservoir and seal data, and supported injection efforts through both predictive and post-injection assessments. Five geologic cross sections (labeled A through E) were constructed onshore across the parts of the MRCSP Region west of the Valley and Ridge Province, and one section (labeled F) was constructed beginning onshore in New Jersey and running eastward to offshore areas examined for carbon storage. Regional and local stratigraphic units were used in these cross sections to show current nomenclature as well as highlighting specific units/features pertinent to carbon capture, utilization and storage (CCUS) opportunities in the region. Each cross section uses 15 to 25 deep wells and other geologic information to illustrate the relative depth, thickness and location of rock units in the subsurface.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2025

Alloy 625 Qualification Pathway for ASME Section III Division 5 Class A construction

The American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section III Division 5 provides construction and inspection rules to ensure the safety of nuclear components operating at elevated temperatures, defined as those operating above 700°F (370°C) or 800°F (425°C), depending on the material type. At present, there are six materials approved for Class A (high safety significance) component construction under ASME BPVC Section III Division 5. Out of these six materials, five materials are steels or iron-based alloys and one material, Alloy 617, is a nickel alloy. Nickel alloys are stronger than steels; however, Alloy 617 consists of 10-15% cobalt, and activation is a concern under irradiation. Therefore, there is a need to qualify new nickel alloy with lower cobalt content. Alloy 625 is one candidate which has similar mechanical properties at elevated temperature compared to Alloy 617 and has a maximum of 1% cobalt. Although Alloy 625 material properties have been developed in the past to support allowable stress development under ASME Section II, new test results would be needed to qualify this material for elevated temperature component construction under ASME Section III Division 5. The purpose of this report is to describe a path towards developing a Nuclear Code Case to qualify Alloy 625 (UNS N06625; Grade 1 and Grade 2) for elevated temperature nuclear use in accordance with the ASME BPVC rules. This report reviews the existing database on Alloy 625, presents proposed test campaign, and discusses potential paths for accelerating the accelerated material qualification process.

36 - MATERIALS SCIENCE↗
NASA NTRS1984

The radar cross section of dielectric disks

A solution is presented for the backscatter (monostatic) radar cross section of dielectric disks of arbitrary shape, thickness, and dielectric constant. The result is obtained by employing a Kirchhoff-type approximation to obtain the fields inside the disk. The internal fields induce polarization and conduction currents from which the scattered fields and the radar cross section can be computed. The solution for the radar cross section obtained in this manner will be shown to agree with known results in the special cases of normal incidence, thin disks, and perfect conductivity. It will also be shown that the solution can be written as a product of the reflection coefficient of an identically oriented slab times the physical optics solution for the backscatter cross section of a perfectly conducting disk of the same shape. This result follows directly from the Kirchhoff-type approximation without additional assumptions.

Radar cross section↗
NASA NTRS2002

Cross-sectional structural parameters from densitometry

Bone densitometry has previously been used to obtain cross-sectional properties of bone from a single X-ray projection across the bone width. Using three unique projections, we have extended the method to obtain the principal area moments of inertia and orientations of the principal axes at each scan cross-section along the length of the scan. Various aluminum phantoms were used to examine scanner characteristics to develop the highest accuracy possible for in vitro non-invasive analysis of cross-sectional properties. Factors considered included X-ray photon energy, initial scan orientation, the angle spanned by the three scans (included angle), and I(min)/I(max) ratios. Principal moments of inertia were accurate to within +/-3.1% and principal angles were within +/-1 degrees of the expected value for phantoms scanned with included angles of 60 degrees and 90 degrees at the higher X-ray photon energy (140 kVp). Low standard deviations in the error (0.68-1.84%) also indicate high precision of calculated measurements with these included angles. Accuracy and precision decreased slightly when the included angle was reduced to 30 degrees. The method was then successfully applied to a pair of excised cadaveric tibiae. The accuracy and insensitivity of the algorithms to cross-sectional shape and changing isotropy (I(min)/I(max)) values when various included angles are used make this technique viable for future in vivo studies.

NASA Center ARC↗
NASA NTRS2019

Light Ion Double-Differential Cross Section Parameterization and Results from the SHIELD Transport Code

Light ions and neutrons have been shown to make large contributions to space radiation dose equivalent for realistic shielding scenarios. Efficient and accurate calculations of light ion double-differential cross sections are required for input into space radiation transport codes. A thermal proton cross section model is developed which includes proton production from the three sources of projectile, central fireball, and target. It is shown that this three-source model is able to explain the low momentum shoulder seen in proton spectra. Using the coalescence model, the thermal proton model is used to calculate light ion double-differential cross sections employed in space radiation transport codes. The three-source model is also seen to be essential to explain the light ion shoulders, which are even more pronounced than the proton shoulders. Comparisons are also made to the cross section models used in the SHIELD transport code.

Space radiation↗
NASA NTRS2020

Effect of Atmospheric Wind and Upstream Buildings on the Flow Quality in the Test Section of the 80-by 120-Foot Wind Tunnel

The effect of upwind buildings on flow quality in the test section of the 80- by-120-Foot Wind Tunnel housed at the National Full-Scale Aerodynamics Complex at NASA Ames Research Center was modeled at reduced scale. A 1/50th-scale model of the 80x120 leg was placed on the turntable in the test section of the full-scale NFAC 80x120 wind tunnel. Ambient wind was simulated by means of a fixed linear array of 6 blowers mounted upstream of the NFAC turntable. NFAC turntable rotation was used to change ambient (atmospheric) wind direction. Models of existing upwind buildings and the planned Google Campus buildings (Googleplex) were constructed at 1/50th-scale. In addition, a variety of upwind blockage conditions were studied, using rectangular bodies, to simulate either a 2D or 3D blockage directly ahead of the inlet. Computer-controlled surveys of test section flow quality were measured using a Cobra probe to provide all three orthogonal velocity components. Surprisingly, measurements indicate that test section flow quality will be unaffected or slightly improved by the upwind presence of the Google Campus. A possible explanation is discussed.

open-circuit wind tunnel↗
NASA NTRS2020

Characterization of the NASA Glenn Research Center 8- by 6-Foot Supersonic Wind Tunnel Supersonic Test Section (2020 Test)

Recent customer requirements for improved Mach number stability for sonic boom signature measurements has piqued interest in the use of the solid-wall supersonic test section in the NASA Glenn Research Center 8- by 6-Foot Supersonic Wind Tunnel. To supply customers with understanding of the operating conditions and flow quality in this region of the tunnel, characterization tests were performed in the solid-wall supersonic test section in January 2020. The test entry covered the operating range of the facility for three-drive-motor operation. Calibration models were generated along with an improved understanding of operational considerations for this region of the tunnel. The data set acquired included centerline axial and planar flow-field distributions to quantify test section flow quality. The collected data provided an improved understanding of operational considerations for this region of the tunnel as well as allowed for the development of detailed calibration models for the solid-wall supersonic test section.

Supersonic↗
DOE OSTI2012

One-group fission cross sections for plutonium and minor actinides inserted in calculated neutron spectra of fast reactor cooled with lead-208 or lead-bismuth eutectic

The paper is dedicated to one-group fission cross sections of Pu and MA in LFRs spectra with the aim to increase these values by choosing a coolant which hardens neutron spectra. It is shown that replacement of coolant from Pb-Bi with Pb-208 in the fast reactor RBEC-M, designed in Russia, leads to increasing the core mean neutron energy. As concerns fuel Pu isotopes, their one-group fission cross sections become slightly changed, while more dramatically Am-241 one-group fission cross section is changed. Another situation occurs in the lateral blanket containing small quantities of minor actinides. It is shown that as a result of lateral blanket mean neutron energy hardening the one-group fission cross sections of Np-237, Am-241 and Am-243 increases up to 8-11%. This result allows reducing the time of minor actinides burning in FRs. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Thermal neutron capture cross section of gadolinium by pile-oscillation measurements in MINERVE

Natural gadolinium is used as a burnable poison in most LWR to account for the excess of reactivity of fresh fuels. For an accurate prediction of the cycle length, its nuclear data and especially its neutron capture cross section needs to be known with a high precision. Recent microscopic measurements at Rensselaer Polytechnic Inst. (RPI) suggest a 11% smaller value for the thermal capture cross section of {sup 157}Gd, compared with most of evaluated nuclear data libraries. To solve this inconsistency, we have analyzed several pile-oscillation experiments, performed in the MINERVE reactor. They consist in the measurement of the reactivity variation involved by the introduction in the reactor of small-samples, containing different mass amounts of natural gadolinium. The analysis of these experiments is done through the exact perturbation theory, using the PIMS calculation tool, in order to link the reactivity effect to the thermal capture cross section. The measurement of reactivity effects is used to deduce the 2200 m.s-1 capture cross section of {sup nat}Gd which is (49360 {+-} 790) b. This result is in good agreement with the JEFF3.1.1 value (48630 b), within 1.6% uncertainty at 1{sigma}, but is strongly inconsistent with the microscopic measurements at RPI which give (44200 {+-} 500) b. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Validating the ENDF-B/VII{sup 235}U(n{sub th},f) prompt fission neutron spectrum using updated dosimetry cross sections (IRDFF)

The International Reactor Dosimetry File IRDF-2002 released in 2004 by the IAEA contains cross-section data and corresponding uncertainties for 66 dosimetry reactions. New cross-section evaluations have become available recently that re-define some of these dosimetry reactions for reactor applications including: 1) high fidelity evaluation work undertaken by one of the authors (KIZ); 2) evaluations from the ENDF/B-VII libraries that cover reactions within the International Evaluation of Neutron Cross-Section Standards; and 3) evaluations from JENDL-3.1 and JENDL-4 libraries. Overall, 37 new evaluations of dosimetry reactions have been assessed to determine whether they should be adopted to update and improve IRDF-2002. A new dosimetry library (International Reactor Dosimetry File for Fission and Fusion - IRDFF) was assembled based on new evaluations combined with selected IRDF-2002 evaluations. A grand-total of 74 dosimetry reactions are included into the IRDFF dosimetry library available at www-nds.iaea.org/IRDFFI. The assembled library was used to validate the {sup 235}U(n{sub th},f) ENDF-B/VII.0 prompt fission neutron spectrum. An excellent average C/E value of 1.002 +/- 0.02 is achieved for reactions with mean neutron energy of the integrated response (E50%) lower than 11 MeV. C/E data for reactions with E50%-response higher than 11 MeV decreases up to 0.8. We conclude that the ENDF-B/VII.0 {sup 235}U(n{sub th},f) prompt fission neutron spectrum from 1-11 MeV is validated within quoted uncertainties by available integral measurements in {sup 235}U(n{sub th},f) neutron field. Further investigations for high-threshold reactions are needed and new measurements of spectrum average cross sections for those reactions in the {sup 235}U(n{sub th},f) neutron field are recommended. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Benchmark for Free-Gas Elastic Scattering with the Heavy-Gas Model and Piecewise-Constant Cross Sections

Accurate Monte Carlo calculations involving thermal neutrons require that codes implement the thermal scattering physics correctly. The heavy-gas model, which is valid for large atomic mass, with constant cross sections in energy has an analytic solution for the thermal neutron spectrum that can be readily used to verify the implementation of the free-gas thermal scattering routines. Previous work has shown that MCNP6.1 correctly computes the free-gas thermal flux with constant cross sections. This summary extends the verification to piecewise-constant cross sections in energy that can better approximate physically realistic cross sections. The obtained solutions for the piecewise-constant cases agree with the respective thermal spectra obtained by MCNP6.1, further verifying that it is accurately simulating the free-gas model. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Production of the thermal scattering law and neutron scattering cross sections for ice from first principles - 14503

Water ice can exist in many polymorphic phases as a result of the wide range of geometric ordering possible for the hydrogen bonds between H{sub 2}O molecules. Ice Ih, which can form under normal atmospheric pressure at 273 K, is the most common type. Ab initio density functional theory, incorporating the generalized gradient approximation (GGA) for the exchange-correlation energy, is used to simulate the hexagonal structure of ice Ih and calculate phonon dispersion relations with harmonic lattice dynamics. While the unit cell for ice Ih contains four molecules with particular alignments of the H{sub 2}O molecules possible, the structure has globally disordered proton arrangement. To improve capturing this random distribution, a 12-molecule unit cell is incorporated along with a 2x2x2 supercell. The total phonon density of states is determined by randomly sampling dispersion relations in the first Brillouin zone. The translational, librational, bending, and stretching modes are produced with high-quality resolution and are consistent with the optical peaks and ranges observed in Raman, infrared, and inelastic neutron scattering experiments [1]. The phonon spectrum is used to calculate the S(α,β) thermal scattering law for ENDF [2] publication in the incoherent approximation, which is appropriate due to the relatively small coherent nuclear scattering cross section for H{sub 2}O. These theoretical results are compared to experimental S(α,β) at 268 K [3] and trends are found to be in good agreement. Finally, total scattering cross sections are calculated by integrating over S(α,β) to determine the incoherent inelastic cross sections and then separately adding the absorption and incoherent elastic components. These results are compared to experimental total cross sections for ice Ih at 115 K over a range of incident energies [4] and found to provide excellent agreement. (authors)

42 ENGINEERING↗
DOE OSTI2015

Calculation and use of pointwise effective cross-sections in the unresolved resonance region - 14605

The pointwise neutron elastic scattering, absorption, fission and total cross-sections in the Unresolved Resonance Region (URR) can be calculated with a code C-007 [1]. The volume of these data is large and estimated to make about 106 numbers for each of the reactions. These data volumes may cause time and memory problems in solving radiation transport by the Monte Carlo method. There are methods to thin the data. These are, for example, the widely used probability tables [2] and subgroups [3]. The paper offers another data thinning approach which was successfully used for the Resolved Resonance Region (RRR) [4,5]. It transforms the pointwise resonance cross-sections into the pointwise monotone ones within some energy intervals with the use of cross-section probability distribution functions P(σ). A similar approach was used to calculate pointwise cross-sections in the URR [6]. (authors)

42 ENGINEERING↗
DOE OSTI2024

Fine Temperature Grid Continuous Energy Cross Section Generation for Monte Carlo Analysis of Xe-100 Design

The standard “A Compact ENDF (ACE)” data libraries used by Monte Carlo based reactor physics codes calculations are provided by Los Alamos National Laboratory (LANL) with a temperature interval mostly of 300 K (e.g. 300 K, 600 K, 900 K) for the cross sections and between 100 K and 200 K for the thermal scattering libraries (TSL). However, some codes such as MCNP lack capability to perform on-the-fly temperature interpolation during simulation both for neutron and TSL cross-sections. To evaluate the impact related to Doppler broadening and spectrum shift associated with TSL changes, this paper explores the potential of adopting a temperature grid finer than the ones contained in the standard data libraries. A 50 K temperature grid was employed to quantify the error in neutronics calculations due to temperature grid resolution. This was achieved by comparing the results of this study (50 K temperature interval) against the results obtained with standard data libraries (>100 K temperature interval). While the adopted grid primarily relies on the ENDF/B-VII.1 library, for neutron cross-sections, it utilizes ENDF/B-VIII.0 library for TSL. The analyses confirmed that the accuracy of neutronics calculations is satisfactory when using a 50 K temperature grid. Notably, adopting a 50 K temperature grid, as opposed to standard libraries or coarser temperature grids, could lead to a difference of no more than a few hundred pcm in dk for both fresh fuel and burnt fuel. The most sensitive reaction type to the temperature grid was as expected identified as the capture cross-section of U-238.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗